Containment Code Validation Matrix

Nuclear SafetyNEA/CSNI/R(2014)3May 2014www.oecd-nea.org

containment Code Validation Matrix

Unclassified NEA/CSNI/R(2014)3 Organisation de Coopération et de Développement Économiques Organisation for Economic Co-operation and Development 23-May-2014 ___________________________________________________________________________________________

English text only NUCLEAR ENERGY AGENCY COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

Containment Code Validation Matrix

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Complete document available on OLIS in its original format This document and any map included herein are without prejudice to the status of or sovereignty over any territory, to the delimitation of international frontiers and boundaries and to the name of any territory, city or area.

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/CSN

I/R(2014)3

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English text only

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ORGANISATION FOR ECONOMIC CO-OPERATION AND DEVELOPMENT

The OECD is a unique forum where the governments of 34 democracies work together to address the economic, social and

environmental challenges of globalisation. The OECD is also at the forefront of efforts to understand and to help governments respond to new developments and concerns, such as corporate governance, the information economy and the challenges of an ageing population. The

Organisation provides a setting where governments can compare policy experiences, seek answers to common problems, identify good

practice and work to co-ordinate domestic and international policies.

The OECD member countries are: Australia, Austria, Belgium, Canada, Chile, the Czech Republic, Denmark, Estonia, Finland,

France, Germany, Greece, Hungary, Iceland, Ireland, Israel, Italy, Japan, Luxembourg, Mexico, the Netherlands, New Zealand, Norway,

Poland, Portugal, the Republic of Korea, the Slovak Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The European Commission takes part in the work of the OECD.

OECD Publishing disseminates widely the results of the Organisation’s statistics gathering and research on economic, social and

environmental issues, as well as the conventions, guidelines and standards agreed by its members.

This work is published on the responsibility of the OECD Secretary-General. The opinions expressed and arguments employed herein do not necessarily reflect the official

views of the Organisation or of the governments of its member countries.

NUCLEAR ENERGY AGENCY

The OECD Nuclear Energy Agency (NEA) was established on 1 February 1958. Current NEA membership consists of 31 countries: Australia, Austria, Belgium, Canada, the Czech Republic, Denmark, Finland, France, Germany, Greece, Hungary, Iceland, Ireland, Italy,

Japan, Luxembourg, Mexico, the Netherlands, Norway, Poland, Portugal, the Republic of Korea, the Russian Federation, the Slovak

Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, the United Kingdom and the United States. The European Commission also takes part in the work of the Agency.

The mission of the NEA is:

– to assist its member countries in maintaining and further developing, through international co-operation, the scientific, technological and legal bases required for a safe, environmentally friendly and economical use of nuclear energy for peaceful

purposes, as well as

– to provide authoritative assessments and to forge common understandings on key issues, as input to government decisions on nuclear energy policy and to broader OECD policy analyses in areas such as energy and sustainable development.

Specific areas of competence of the NEA include the safety and regulation of nuclear activities, radioactive waste management,

radiological protection, nuclear science, economic and technical analyses of the nuclear fuel cycle, nuclear law and liability, and public

information.

The NEA Data Bank provides nuclear data and computer program services for participating countries. In these and related tasks, the NEA works in close collaboration with the International Atomic Energy Agency in Vienna, with which it has a Co-operation Agreement, as

well as with other international organisations in the nuclear field.

This document and any map included herein are without prejudice to the status of or sovereignty over any territory, to the delimitation of

international frontiers and boundaries and to the name of any territory, city or area.

Corrigenda to OECD publications may be found online at: www.oecd.org/publishing/corrigenda.

© OECD 2014

You can copy, download or print OECD content for your own use, and you can include excerpts from OECD publications, databases and multimedia products in your own documents, presentations, blogs, websites and teaching materials, provided that suitable acknowledgment of

the OECD as source and copyright owner is given. All requests for public or commercial use and translation rights should be submitted to

[email protected]. Requests for permission to photocopy portions of this material for public or commercial use shall be addressed directly to the

Copyright Clearance Center (CCC) at [email protected] or the Centre français d'exploitation du droit de copie (CFC) [email protected].

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THE COMMITTEE ON THE SAFETY OF NUCLEAR INSTALLATIONS

“The Committee on the Safety of Nuclear Installations (CSNI) shall be responsible for the

activities of the Agency that support maintaining and advancing the scientific and technical knowledge

base of the safety of nuclear installations, with the aim of implementing the NEA Strategic Plan for 2011-

2016 and the Joint CSNI/CNRA Strategic Plan and Mandates for 2011-2016 in its field of competence.

The Committee shall constitute a forum for the exchange of technical information and for

collaboration between organisations, which can contribute, from their respective backgrounds in research,

development and engineering, to its activities. It shall have regard to the exchange of information between

member countries and safety R&D programmes of various sizes in order to keep all member countries

involved in and abreast of developments in technical safety matters.

The Committee shall review the state of knowledge on important topics of nuclear safety science

and techniques and of safety assessments, and ensure that operating experience is appropriately accounted

for in its activities. It shall initiate and conduct programmes identified by these reviews and assessments in

order to overcome discrepancies, develop improvements and reach consensus on technical issues of

common interest. It shall promote the co-ordination of work in different member countries that serve to

maintain and enhance competence in nuclear safety matters, including the establishment of joint

undertakings, and shall assist in the feedback of the results to participating organisations. The Committee

shall ensure that valuable end-products of the technical reviews and analyses are produced and available to

members in a timely manner.

The Committee shall focus primarily on the safety aspects of existing power reactors, other

nuclear installations and the construction of new power reactors; it shall also consider the safety

implications of scientific and technical developments of future reactor designs.

The Committee shall organise its own activities. Furthermore, it shall examine any other matters

referred to it by the Steering Committee. It may sponsor specialist meetings and technical working groups

to further its objectives. In implementing its programme the Committee shall establish co-operative

mechanisms with the Committee on Nuclear Regulatory Activities in order to work with that Committee

on matters of common interest, avoiding unnecessary duplications.

The Committee shall also co-operate with the Committee on Radiation Protection and Public

Health, the Radioactive Waste Management Committee, the Committee for Technical and Economic

Studies on Nuclear Energy Development and the Fuel Cycle and the Nuclear Science Committee on

matters of common interest.”

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TABLE OF CONTENTS

TABLE OF CONTENTS ................................................................................................................................ 5

EXECUTIVE SUMMARY ........................................................................................................................... 15

LIST OF THE CCVM WRITING GROUP MEMBERS (2010 – 2013) ...................................................... 16

ACRONYMS ................................................................................................................................................ 21

1 INTRODUCTION .................................................................................................................................. 25

1.1 Background ..................................................................................................................................... 25 1.2 Objectives and Scope ...................................................................................................................... 25 1.3 Structure of the Report .................................................................................................................... 25 1.4 References ....................................................................................................................................... 26

2 GENERAL OVERVIEW OF CONTAINMENT AND ACCIDENT PROGRESSION ........................ 27

2.1 Plant Types ...................................................................................................................................... 27 2.1.1 Light Water Reactors ................................................................................................................. 27 2.1.2 Pressurized Heavy Water Reactors (CANDU) .......................................................................... 34

2.2 Accident Progression ...................................................................................................................... 37 2.2.1 LWR Accident Progression ....................................................................................................... 37 2.2.2 CANDU Accident Progression .................................................................................................. 39

2.3 References ....................................................................................................................................... 41

3 PHENOMENA ....................................................................................................................................... 42

3.1 Containment Thermalhydraulics Phenomena ................................................................................. 67 3.1.1 P1-1 - Stratification ................................................................................................................... 67 3.1.2 P1-2 - Flashing (Flashing Discharge) ........................................................................................ 69 3.1.3 P1-3 - Boiling Heat and Mass Transfer ..................................................................................... 70 3.1.4 P1-4 - Critical Heat Flux (CHF) ................................................................................................ 71 3.1.5 P1-5 - Heat Conduction in Solids .............................................................................................. 72 3.1.6 P1-6 - Convection Heat Transfer (Natural and Forced) ............................................................ 73 3.1.7 P1-7 - Thermal Diffusion in Fluids (No Experiments) .............................................................. 74 3.1.8 P1-8 - Radiation Heat Transfer (No Experiments) .................................................................... 75 3.1.9 P1-9 - Condensation on Surfaces ............................................................................................... 77 3.1.10 P1-10 - Pool Surface Evaporation and Condensation ............................................................ 78 3.1.11 P1-11 - Heat Removal by Dousing ........................................................................................ 79 3.1.12 P1-12 - Liquid Re-Entrainment (Resuspension) .................................................................... 80 3.1.13 P1-13 - Direct Contact Condensation .................................................................................... 81 3.1.14 P1-14 - Momentum Induced Mixing in Gases ...................................................................... 82 3.1.15 P1-15 - Buoyancy Induced Mixing in Gases ......................................................................... 84 3.1.16 P1-16 - Pressure Wave Propagation in Water ....................................................................... 85 3.1.17 P1-17 - Mixing in Water Pools .............................................................................................. 86 3.1.18 P1-18 - Mass Diffusion in Vapour ........................................................................................ 88 3.1.19 P1-19 - Laminar Flow (No Experiments) .............................................................................. 89 3.1.20 P1-20 - Turbulent Flow ......................................................................................................... 90

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3.1.21 P1-21 - Critical Flow (Choked Flow) .................................................................................... 91 3.1.22 P1-22 - Laminar/Turbulent Leakage Flow ............................................................................ 93 3.1.23 P1-23 - Vent Clearing ............................................................................................................ 94 3.1.24 P1-24 - Pool Swell / Air Injection ......................................................................................... 96 3.1.25 P1-25 - Interfacial Drag (No Experiments) ........................................................................... 98 3.1.26 P1-26 - Liquid Film Flow .................................................................................................... 100 3.1.27 P1-27 - Gas Dissolved in Water (No Experiments) ............................................................. 101 3.1.28 P1-28 - Gas Entrainment by Spray Droplets (Dousing) ...................................................... 102 3.1.29 P1-29 - Heat and Mass Transfer of Spray Droplets (Dousing) ............................................ 103 3.1.30 P1-30 - Droplet Interaction (Dousing) ................................................................................. 104 3.1.31 P1-31 - Mixing by Sprays .................................................................................................... 105 3.1.32 P1-32 - Turbulence Induced by Sprays ............................................................................... 106

3.2 Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena .............................. 107 3.2.1 P2-1 - Deflagration .................................................................................................................. 107 3.2.2 P2-2 - Hydrogen Flame Acceleration (FA) ............................................................................. 109 3.2.3 P2-3 - Deflagration-to-Detonation Transition (DDT) ............................................................. 110 3.2.4 P2-4 - Hydrogen Detonation.................................................................................................... 111 3.2.5 P2-5 - Quenching of Detonations by Geometrical Constrains ................................................ 112 3.2.6 P2-6 - Quenching ..................................................................................................................... 113 3.2.7 P2-7 - Hydrogen Diffusion Flame (Standing Flame) .............................................................. 114 3.2.8 P2-8 - Hydrogen Mitigation - Passive Autocatalytic Recombiners ......................................... 115 3.2.9 P2-9 - Hydrogen Ignition by PARs (Weak Ignition) ............................................................... 116 3.2.10 P2-10 - Hydrogen Mitigation by Hydrogen Ignitors (Mild Ignition) .................................. 117 3.2.11 P2-11 - Strong Ignition of Hydrogen ................................................................................... 118 3.2.12 P2-12 - Jet Ignition of Hydrogen ......................................................................................... 119 3.2.13 P2-13 - Radiolysis (Hydrogen Production by Water Radiolysis) ........................................ 120 3.2.14 P2-14 - Effect of Droplets on Hydrogen Combustion ......................................................... 121

3.3 Aerosol and Fission Product Behaviour Phenomena .................................................................... 122 3.3.1 P3-1 - Aerosol Formation in a Flashing Jet ............................................................................. 122 3.3.2 P3-2 - Aerosol Formation in a Steam Jet ................................................................................. 123 3.3.3 P3-3 - Aerosol Impaction (Jet Impingement) .......................................................................... 124 3.3.4 P3-4 - Thermophoresis ............................................................................................................ 125 3.3.5 P3-5 - Diffusiophoresis ............................................................................................................ 126 3.3.6 P3-6 - Liquid Aerosol Evaporation ......................................................................................... 127 3.3.7 P3-7 - Condensation on Aerosols ............................................................................................ 128 3.3.8 P3-8 - Gravitational Agglomeration ........................................................................................ 129 3.3.9 P3-9 - Diffusional Agglomeration ........................................................................................... 130 3.3.10 P3-10 - Turbulent Agglomeration of Aerosols .................................................................... 131 3.3.11 P3-11 - Drop Breakup.......................................................................................................... 132 3.3.12 P3-12 - Gravitational Settling (Drop Settling)..................................................................... 133 3.3.13 P3-13 - Diffusional Deposition ............................................................................................ 134 3.3.14 P3-14 - Inertial Deposition of Aerosols (Also called Impaction) ........................................ 135 3.3.15 P3-15 - Turbulent Deposition of Aerosols........................................................................... 136 3.3.16 P3-16 - Re-volatilisation ...................................................................................................... 137 3.3.17 P3-17 - Aerosol Removal in Leakage Paths ........................................................................ 138 3.3.18 P3-18 - Pool Scrubbing of Aerosols .................................................................................... 139 3.3.19 P3-19 - Radionuclide Transport .......................................................................................... 140 3.3.20 P3-20 - Radionuclide Decay Heat (No Experiments) .......................................................... 141 3.3.21 P3-21 - Release Rate Change Due to Oxidizing Environment ............................................ 142 3.3.22 P3-22 - Containment Chemistry Impact on Source Term ................................................... 143 3.3.23 P3-23 - Ruthenium Volatility and Behaviour in Containment ............................................ 144

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3.3.24 P3-24 - Aerosol Removal by Sprays (Dousing) .................................................................. 145 3.3.25 P3-25 - Re-suspension (Dry) ............................................................................................... 146 3.3.26 P3-26 - Re-entrainment (Wet) ............................................................................................. 147 3.3.27 P3-27 - Aerosol De-agglomeration ...................................................................................... 148

3.4 Iodine Chemistry Phenomena ....................................................................................................... 149 3.4.1 P4-1 - Aqueous Phase Oxidation and Reduction of Iodine Species ........................................ 149 3.4.2 P4-2 - Inorganic Iodine Hydrolysis ......................................................................................... 150 3.4.3 P4-3 - Inorganic Iodine Radiolysis in Water Phase ................................................................. 151 3.4.4 P4-4 - Homogeneous Organic Reactions in Water Phase........................................................ 152 3.4.5 P4-5 - Iodine Reactions with Surfaces in the Water Phase...................................................... 153 3.4.6 P4-6 - Iodine reactions with surfaces in the gas phase ............................................................ 154 3.4.7 P4-7 - Silver Iodine Reactions in the Water Phase .................................................................. 155 3.4.8 P4-8 - Gas Phase Radiolytic Oxidation of Molecular Iodine (I2) (Iodine/Ozone Reaction) ... 156 3.4.9 P4-9 - Homogeneous Organic Iodine Reactions in Gas Phase ................................................ 157 3.4.10 P4-10 - RI (Organic Iodine) Radiolytic Destruction ........................................................... 158 3.4.11 P4-11 - Interfacial Mass Transfer ........................................................................................ 159 3.4.12 P4-12 - Decomposition of Iodides (CsI) by Heat-up in PARs ............................................ 160 3.4.13 P4-13 - Iodine Filtration ...................................................................................................... 161 3.4.14 P4-14 - Volatile Iodine Trapping by Airborne Droplets ..................................................... 162 3.4.15 P4-15 - Iodine Retention in Leakage Paths ......................................................................... 163 3.4.16 P4-16 - I2 Interaction with Aerosols .................................................................................... 164 3.4.17 P4-17 - Iodine Wash-down .................................................................................................. 165 3.4.18 P4-18 - Pool Scrubbing of Iodine ........................................................................................ 166 3.4.19 P4-19 - Iodine Release from Flashing Pool or Flashing Jet ................................................ 167

3.5 Core Melt Distribution and Behaviour in Containment Phenomena ............................................ 168 3.5.1 P5-1 - Corium Release from Failed Dry Reactor Pressure Vessel .......................................... 168 3.5.2 P5-2 - Corium Entrainment Out of the Reactor Primary Vessel with Lateral Breaches ......... 170 3.5.3 P5-3 - Corium Particles Generation from the Corium Pool..................................................... 171 3.5.4 P5-4 - Corium Particles Generation from the Two Phase Jet .................................................. 172 3.5.5 P5-5 - Corium Particles Entrainment ....................................................................................... 173 3.5.6 P5-6 - Corium Particles Trapping ............................................................................................ 174 3.5.7 P5-7 - Direct Containment Heating ......................................................................................... 175 3.5.8 P5-8 - Corium Jet Break-up in Water Pool .............................................................................. 176 3.5.9 P5-9 - FCI and Steam Explosion - Melt into Water Ex-Vessel (Melt Quenching) ................. 177 3.5.10 P5-10 - Pressure Load on Corium Retention Devices ......................................................... 178 3.5.11 P5-11 - Particulate Debris Bed Formation .......................................................................... 179 3.5.12 P5-12 - Corium Debris (Solid) Heat Transfer ..................................................................... 180 3.5.13 P5-13 - Molten Core Concrete Interaction .......................................................................... 181 3.5.14 P5-14 - Corium Melt Stratification ...................................................................................... 182 3.5.15 P5-15 - Corium Spreading ................................................................................................... 183 3.5.16 P5-16 - Molten Corium Heat Transfer ................................................................................ 184 3.5.17 P5-17 - Corium Evaporation/Vaporization .......................................................................... 185 3.5.18 P5-18 - Corium Solidification/Crust Formation .................................................................. 186 3.5.19 P5-19 - Cracking (Crust) ..................................................................................................... 187 3.5.20 P5-20 - Ex-Vessel Corium Coolability, Top Flooding ........................................................ 188 3.5.21 P5-21 - Ex-Vessel Corium Catcher - Coolability and Water Bottom Injection .................. 189 3.5.22 P5-22 - Ex-Vessel Corium Catcher - Corium-Ceramics Interaction and Properties ........... 190 3.5.23 P5-23 - Effect of Non Homogeneous Ablation on Gate Ablation ....................................... 191 3.5.24 P5-24 - Crust Anchorage ..................................................................................................... 192 3.5.25 P5-25 - Radionuclide Release from MCCI and Core Catchers ........................................... 193 3.5.26 P5-26 - Core Catchers with External Cooling ..................................................................... 194

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3.5.27 P5-27 - Oxidation of Corium ............................................................................................... 195 3.5.28 P5-28 - Corium Attack of Metallic Liner ............................................................................ 196 3.5.29 P5-29 - Corium Release from Failed Flooded Reactor Pressure Vessel ............................. 197

3.6 Systems Phenomena ...................................................................................................................... 198 3.6.1 P6-1 - Ventilation Systems ...................................................................................................... 198 3.6.2 P6-2 - Behaviour of Doors, Burst Membranes, Rupture Discs etc. ......................................... 199 3.6.3 P6-3 - Air Cooler (Fan Cooler) Heat Transfer......................................................................... 200 3.6.4 P6-4 - Pump Performance including Sump Clogging (No Experiments) ................................ 201 3.6.5 P6-5 - Passive Cooling by Internal and External Condensers ................................................. 202 3.6.6 P6-6 - Aerosol Removal in EFADS ........................................................................................ 204

4 EXPERIMENTS .................................................................................................................................. 205

4.1 Containment Thermalhydraulics Experiments .............................................................................. 256 4.1.1 E1-1 - Flow through Interconnected Vessels........................................................................... 256 4.1.2 E1-2 - Bruce LAC Test in Air, Test No. 50............................................................................. 259 4.1.3 E1-3 - LSGMF GMBT001 ...................................................................................................... 260 4.1.4 E1-4 - LSGMF GMUS001 ...................................................................................................... 262 4.1.5 E1-5 - AECL-SP Dousing Test No. 1 ...................................................................................... 263 4.1.6 E1-6 - FIPLOC F2 ................................................................................................................... 265 4.1.7 E1-7 - VANAM M3 (ISP-37) .................................................................................................. 267 4.1.8 E1-8 - EREC LB LOCA Test 1 ............................................................................................... 268 4.1.9 E1-9 - EREC LB LOCA Test 5 ............................................................................................... 271 4.1.10 E1-10 – EREC MSLB Test 7 .............................................................................................. 273 4.1.11 E1-11 - EREC MSLB Test 9 ............................................................................................... 275 4.1.12 E1-12 - EREC SLB G02 ...................................................................................................... 277 4.1.13 E1-13 - HDR V44 (ISP-16) ................................................................................................. 279 4.1.14 E1-14 - HDR T31.5 (ISP-23) .............................................................................................. 280 4.1.15 E1-15 - HDR E11.2 (ISP-29) .............................................................................................. 281 4.1.16 E1-16 - HDR E11.4 ............................................................................................................. 283 4.1.17 E1-17 - GKSS M1 ............................................................................................................... 284 4.1.18 E1-18 - MISTRA ISP-47 ..................................................................................................... 287 4.1.19 E1-19 - MISTRA M7 .......................................................................................................... 290 4.1.20 E1-20 - MISTRA-M8 .......................................................................................................... 293 4.1.21 E1-21 - MISTRA-MASP ..................................................................................................... 295 4.1.22 E1-22 - NUPEC M-7-1 (ISP-35) ......................................................................................... 297 4.1.23 E1-23 - NUPEC M-8-2 ........................................................................................................ 298 4.1.24 E1-24 - PANDA ISP-42, Phase A ....................................................................................... 300 4.1.25 E1-25 - PANDA ISP-42, Phase C ....................................................................................... 302 4.1.26 E1-26 - PANDA ISP-42, Phase E ....................................................................................... 303 4.1.27 E1-27 - PANDA ISP-42, Phase F ........................................................................................ 304 4.1.28 E1-28 - PANDA BC4 .......................................................................................................... 305 4.1.29 E1-29 - SVUSS G02 ............................................................................................................ 307 4.1.30 E1-30 - THAI TH1 .............................................................................................................. 309 4.1.31 E1-31 - THAI TH2 .............................................................................................................. 311 4.1.32 E1-32 - THAI TH7 .............................................................................................................. 312 4.1.33 E1-33 - THAI TH10 ............................................................................................................ 313 4.1.34 E1-34 - THAI TH13 (ISP-47) ............................................................................................. 314 4.1.35 E1-35 - THAI HM2 ............................................................................................................. 315 4.1.36 E1-36 - TOSQAN ISP-47 .................................................................................................... 316 4.1.37 E1-37 - TOSQAN Condensation Tests ............................................................................... 318 4.1.38 E1-38 - TOSQAN Test 113 ................................................................................................. 320

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4.1.39 E1-39 - TOSQAN Spray Tests ............................................................................................ 322 4.1.40 E1-40 - University of Wisconsin Flat Plate Condensation Tests......................................... 324 4.1.41 E1-41 - CONAN SARNET Benchmark No. 1 .................................................................... 325 4.1.42 E1-42 - CONAN SARNET2 Benchmark No. 2 .................................................................. 327 4.1.43 E1-43 - CSTF Tests ............................................................................................................. 328 4.1.44 E1-44 - Marviken Test 18 .................................................................................................... 329 4.1.45 E1-45 - CARAIDAS EVAP and COND tests ..................................................................... 331 4.1.46 E1-46 - TOSQAN sump tests .............................................................................................. 333 4.1.47 E1-47 - CALIST PWR spray test ........................................................................................ 335 4.1.48 E1-48 - MISTRA LOWMA ................................................................................................ 337 4.1.49 E1-49 - PANDA OECD/SETH tests ................................................................................... 339 4.1.50 E1-50 - PANDA OECD/SETH-2 ........................................................................................ 342 4.1.51 E1-51 - CYBL Boiling Tests ............................................................................................... 345 4.1.52 E1-52 - ULPU CHF Tests ................................................................................................... 347 4.1.53 E1-53 - SULTAN CHF Tests .............................................................................................. 349 4.1.54 E1-54 - SBLB Boiling Tests ................................................................................................ 351 4.1.55 E1-55 – Small Scale Burst Test Experiments ...................................................................... 353

4.2 Hydrogen Behaviour (Combustion, Mitigation and Generation) Experiments ............................ 355 4.2.1 E2-1 - LSVCTF S01 ................................................................................................................ 355 4.2.2 E2-2 - LSVCTF S03 ................................................................................................................ 357 4.2.3 E2-3 - BMC Hx series ............................................................................................................. 358 4.2.4 E2-4 - BMC Ix series ............................................................................................................... 363 4.2.5 E2-5 - BMC Gx Series............................................................................................................. 366 4.2.6 E2-6 - BMC Kx Series............................................................................................................. 369 4.2.7 E2-7 - BMC Ex Series ............................................................................................................. 374 4.2.8 E2-8 - ENACEFF SARNET2 Tests ........................................................................................ 376 4.2.9 E2-9 - ENACEFF SARNET Test (Run 703) ........................................................................... 378 4.2.10 E2-10 - ENACEFF SARNET Test (Run 717) ..................................................................... 379 4.2.11 E2-11 - ENACEFF Run 765 (ISP-49) ................................................................................. 380 4.2.12 E2-12 - ENACEFF Run 736 (ISP-49) ................................................................................. 381 4.2.13 E2-13 - ENACEFF Run 733 (ISP-49) ................................................................................. 382 4.2.14 E2-14 - DRIVER HYCOM MC 003 ................................................................................... 383 4.2.15 E2-15 - DRIVER HYCOM MC 012 ................................................................................... 384 4.2.16 E2-16 - FZK R 0498_09 ...................................................................................................... 385 4.2.17 E2-17 - DRIVER HYCOM MC 043 ................................................................................... 386 4.2.18 E2-18 - DRIVER HYCOM HC 020 .................................................................................... 387 4.2.19 E2-19 - DRIVER HYCOM-HC027 .................................................................................... 388 4.2.20 E2-20 - RUT HYC01 ........................................................................................................... 389 4.2.21 E2-21 - RUT HYC12 ........................................................................................................... 390 4.2.22 E2-22 - RUT HYC14 ........................................................................................................... 391 4.2.23 E2-23 - VGES Tests ............................................................................................................ 392 4.2.24 E2-24 - NTS Tests ............................................................................................................... 394 4.2.25 E2-25 - PET Tubes .............................................................................................................. 395 4.2.26 E2-26 - THAI HD Series (Combustion Tests) .................................................................... 399 4.2.27 E2-27 - THAI HR Series (PAR Tests) ................................................................................ 402 4.2.28 E2-28 - THAI Hydrogen Combustion During Spray Operation.......................................... 405 4.2.29 E2-29 - DFF SFSER01 ........................................................................................................ 408 4.2.30 E2-30 - LSVCTF S02 .......................................................................................................... 410 4.2.31 E2-31 - LSVCTF DC ........................................................................................................... 411 4.2.32 E2-32 - LSVCTF 3C ........................................................................................................... 412 4.2.33 E2-33 - LSVCTF CIC.......................................................................................................... 413

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4.2.34 E2-34 - Gammacell Radiolysis Tests .................................................................................. 414 4.2.35 E2-35 - LACOMECO UFPE2 ............................................................................................. 416 4.2.36 E2-36 - LACOMECO HYGRADE10 ................................................................................. 418 4.2.37 E2-37 - LACOMECO HYGRADE09 ................................................................................. 420 4.2.38 E2-38 - LACOMECO HYGRADE03 ................................................................................. 421 4.2.39 E2-39 - LACOMECO HYDET06 ....................................................................................... 422 4.2.40 E2-40 - LACOMECO HYDET07 ....................................................................................... 424 4.2.41 E2-41 - H2PAR E 12 ........................................................................................................... 425 4.2.42 E2-42 - H2PAR E 13 ........................................................................................................... 426 4.2.43 E2-43 - H2PAR E 3 ............................................................................................................. 427 4.2.44 E2-44 – KIT DDT Tests in CHANNEL Facility ................................................................. 428 4.2.45 E2-45 – KIT Jet Ignition Tests in HPHR Facility ............................................................... 430 4.2.46 E2-46 – KIT Geometric Quenching of Detonation Tests in the HYKA-A1 Facility .......... 432 4.2.47 E2-47 – Cheikhravat Experiments on Effect of Spray on Hydrogen Combustion .............. 434 4.2.48 E2-48 – Bjerketvedt Experiments on Effect of Spray on Hydrogen Combustion ............... 436

4.3 Aerosol and Fission Product Behaviour Experiments................................................................... 437 4.3.1 E3-1 - AHMED OECD benchmark ......................................................................................... 437 4.3.2 E3-2 - KAEVER CsI series ..................................................................................................... 439 4.3.3 E3-3 - KAEVER K187 (ISP-44) ............................................................................................. 440 4.3.4 E3-4 - KAEVER K148 (ISP-44) ............................................................................................. 441 4.3.5 E3-5 - KAEVER K188 (ISP-44) ............................................................................................. 442 4.3.6 E3-6 - LACE LA2 ................................................................................................................... 443 4.3.7 E3-7 - LACE LA4 ................................................................................................................... 445 4.3.8 E3-8 – LACE LA5 and LA6.................................................................................................... 446 4.3.9 E3-9 - Phebus FPT-1 (ISP-46) ................................................................................................. 448 4.3.10 E3-10 - POSEIDON PA10 .................................................................................................. 450 4.3.11 E3-11 - BMC VANAM M2................................................................................................. 451 4.3.12 E3-12 - VICTORIA test 58 ................................................................................................. 452 4.3.13 E3-13 - CSTF ABCOVE Tests ............................................................................................ 454 4.3.14 E3-14 - CSTF ACE ............................................................................................................. 456 4.3.15 E3-15 - CARAIDAS Aerosol washout by single droplet tests ............................................ 457 4.3.16 E3-16 - Whiteshell Flashing Jet Tests ................................................................................. 459 4.3.17 E3-17 - Clarkson College Brownian Agglomeration .......................................................... 461 4.3.18 E3-18 - JAERI Thermophoresis Tests ................................................................................. 462 4.3.19 E3-19 - PITEAS Diffusiophoresis Tests (PDI 08, PDI 09, PDI 11 and PDI 12)................. 463 4.3.20 E3-20 - PITEAS Aerosol Condensation Tests (PCON 01 to PCON 05) ............................. 464 4.3.21 E3-21 - Aerosol Deposition in Turbulent Vertical Conduits (Sehmel) ............................... 465 4.3.22 E3-22 - Aerosol Deposition in Turbulent Vertical Conduits (Forney) ................................ 466 4.3.23 E3-23 - Aerosol Deposition in Turbulent Vertical Conduits (Friedlander) ......................... 467 4.3.24 E3-24 - Aerosol Deposition in Turbulent Vertical Conduits (Liu) ...................................... 468 4.3.25 E3-25 - Aerosol Deposition in Turbulent Vertical Conduits (Wells) .................................. 469 4.3.26 E3-26 - CSE Fission Product Transport Tests ..................................................................... 470 4.3.27 E3-27 - CSE Aerosol Removal Tests .................................................................................. 472 4.3.28 E3-28 - LASS-SGTR ........................................................................................................... 474 4.3.29 E3-29 - MCE, UCE and HCE Tests .................................................................................... 476 4.3.30 E3-30 - GBI Tests ................................................................................................................ 479 4.3.31 E3-31 - Aerosol Trapping Effects in Containment Penetration (A. Watanabe) ................. 481 4.3.32 E3-32 - Aerosol transfer through cracked concrete walls.................................................... 483 4.3.33 E3-33 - Whiteshell Steam Jet Experiments ......................................................................... 485 4.3.34 E3-34 - WALE .................................................................................................................... 487 4.3.35 E3-35 – AEREST (Aerosol resuspension shock tube) ........................................................ 490

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4.3.36 E3-36 – VANAM-M4 ......................................................................................................... 492 4.3.37 E3-37 – THAI Aer-1, Aer-3 and Aer-4 tests ....................................................................... 493 4.3.38 E3-38 – Phebus FPT4 Revaporization ................................................................................. 494 4.3.39 E3-39 – Ruthenium Revolatilisation Studies at VTT .......................................................... 495 4.3.40 E3-40 – Ruthenium Transport and Revolatilisation Studies at KFKI ................................. 496 4.3.41 E3-41 – Ruthenium deposition studies at Chalmers University .......................................... 497 4.3.42 E3-42 – Ruthenium Revolatilisation Studies at IRSN ......................................................... 498

4.4 Iodine Chemistry Experiments ...................................................................................................... 499 4.4.1 E4-1 - CFTF Charcoal Filter Test ............................................................................................ 499 4.4.2 E4-2 - RTF P9T3 ..................................................................................................................... 501 4.4.3 E4-3 - RTF P9T1 ..................................................................................................................... 502 4.4.4 E4-4 - RTF P9T2 ..................................................................................................................... 503 4.4.5 E4-5 - RTF P10T2 ................................................................................................................... 504 4.4.6 E4-6 - RTF P10T3 ................................................................................................................... 505 4.4.7 E4-7 - RTF P11T1 ................................................................................................................... 506 4.4.8 E4-8 - RTF P0T2 ..................................................................................................................... 507 4.4.9 E4-9 - RTF P10T1 ................................................................................................................... 508 4.4.10 E4-10 - RTF PHEBUS RTF1 .............................................................................................. 509 4.4.11 E4-11 - EPICUR Test Series S1, S2 and S3 ........................................................................ 510 4.4.12 E4-12 - THAI Iod-09 ........................................................................................................... 511 4.4.13 E4-13 - THAI Iod-11 ........................................................................................................... 512 4.4.14 E4-14 - THAI Iod-12 ........................................................................................................... 513 4.4.15 E4-15 - THAI Iod-13 ........................................................................................................... 515 4.4.16 E4-16 - THAI Iod-14 ........................................................................................................... 516 4.4.17 E4-17 - THAI Iod-25 ........................................................................................................... 517 4.4.18 E4-18 - THAI Iod-26 ........................................................................................................... 518 4.4.19 E4-19 - THAI AW ............................................................................................................... 519 4.4.20 E4-20 - THAI HR31 ............................................................................................................ 521 4.4.21 E4-21 - THAI HR32 ............................................................................................................ 522 4.4.22 E4-22 - LASS-GIRS DABASCO ........................................................................................ 523 4.4.23 E4-23 - OECD-THAI2 Gaseous Iodine Release from Flashing Jet Test ............................ 524 4.4.24 E4-24 - CAIMAN 97/02 test ............................................................................................... 526 4.4.25 E4-25 - CAIMAN 2001/01 Test .......................................................................................... 528 4.4.26 E4-26 – Iodine Clean-Up in a Steam Suppression System .................................................. 529

4.5 Core Melt Distribution and Behaviour in Containment Experiments ........................................... 531 4.5.1 E5-1 - IET Experiments - Zion Geometry ............................................................................... 531 4.5.2 E5-2 - IET Experiments - Surry Geometry.............................................................................. 534 4.5.3 E5-3 - FARO Tests .................................................................................................................. 541 4.5.4 E5-4 - DISCO-C Tests ............................................................................................................. 544 4.5.5 E5-5 - DISCO-H Tests ............................................................................................................ 546 4.5.6 E5-6 - DISCO-A2 .................................................................................................................... 548 4.5.7 E5-7 - KROTOS JRC Tests ..................................................................................................... 550 4.5.8 E5-8 - SERENA-2 KROTOS and TROI Commissioning Tests.............................................. 557 4.5.9 E5-9: SERENA-2 KROTOS and TROI Tests ......................................................................... 562 4.5.10 E5-10 - MCCI-1 Tests CCI Tests 1-3; SSWICS tests 1-7 ................................................... 564 4.5.11 E5-11 - MCCI-2 Tests CCI Tests 4-6; SSWICS tests 8-13; WCB-1 .................................. 566 4.5.12 E5-12 - ECO Tests ............................................................................................................... 569 4.5.13 E5-13 - BALI Ex-Vessel Tests ............................................................................................ 570 4.5.14 E5-14 - BALISE Tests ......................................................................................................... 571 4.5.15 E5-15 - VULCANO VB-U7 (EPR concrete) ...................................................................... 572 4.5.16 E5-16 - VULCANO VW-U1 (COMET bottom flooding) .................................................. 573

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4.5.17 E5-17 - VULCANO VE-U7 ................................................................................................ 575 4.5.18 E5-18 – SURC-1 and SURC-2 ............................................................................................ 576 4.5.19 E5-19 - SURC-3 .................................................................................................................. 579 4.5.20 E5-20 - SURC-3A ............................................................................................................... 580 4.5.21 E5-21 - SURC-4 .................................................................................................................. 581 4.5.22 E5-22 - BETA V5.1 ............................................................................................................. 583 4.5.23 E5-23 - ACE Phase C Tests L1, L2, L4, L5, L6, and L7 .................................................... 584 4.5.24 E5-24 - MACE Tests M0, M1b, M3b, M4, and MSET-1 ................................................... 585 4.5.25 E5-25 - COLIMA CA-U4.................................................................................................... 588 4.5.26 E5-26 - BURN-1 .................................................................................................................. 589 4.5.27 E5-27 – SWISS-1 and SWISS-2 ......................................................................................... 590 4.5.28 E5-28 – HSS-1 and HSS-3 .................................................................................................. 591 4.5.29 E5-29 - TURC1T and TURC1SS ........................................................................................ 593 4.5.30 E5-30 – TURC2 and TURC3 .............................................................................................. 594 4.5.31 E5-31 - LSL-1,2,3 ................................................................................................................ 595 4.5.32 E5-32 - LBL-1,2,3 ............................................................................................................... 596 4.5.33 E5-33 - LSCRBR-1,2,3 ....................................................................................................... 597 4.5.34 E5-34 - COIL-1 ................................................................................................................... 598 4.5.35 E5-35 - WETCOR-1 ............................................................................................................ 599 4.5.36 E5-36 - FRAG ..................................................................................................................... 600 4.5.37 E5-37 - 1DHtFlx .................................................................................................................. 602 4.5.38 E5-38 – MC Tests ................................................................................................................ 603 4.5.39 E5-39 – Plate Tests .............................................................................................................. 604

4.6 Systems Experiments .................................................................................................................... 605 4.6.1 E6-1 - CSE EFADS Tests ........................................................................................................ 605 4.6.2 E6-2 - ACE-CSTF EFADS Tests ............................................................................................ 607 4.6.3 E6-3 - ACE-LSFF EFADS Tests............................................................................................. 609

5 PHENOMENA VS. EXPERIMENTS CROSS MATRIX ................................................................... 611

6 SUMMARY ......................................................................................................................................... 614

Tables

Table 3-1 Containment Thermalhydraulics Phenomena ............................................................................ 43 Table 3-2 Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena .......................... 49 Table 3-3 Aerosol and Fission Product Behaviour Phenomena ................................................................. 52 Table 3-4 Iodine Chemistry Phenomena .................................................................................................... 57 Table 3-5 Core Melt Distribution and Behaviour in Containment Phenomena ......................................... 60 Table 3-6 Systems Phenomena .................................................................................................................. 66 Table 4-1 List of Information Provided for Each Experiment ................................................................. 205 Table 4-2 Containment Thermalhydraulics Experiments ........................................................................ 207 Table 4-3 Hydrogen Behaviour (Combustion, Mitigation and Generation) Experiments ....................... 221 Table 4-4 Aerosol and Fission Product Behaviour Experiments ............................................................. 227 Table 4-5 Iodine Chemistry Experiments ................................................................................................ 235 Table 4-6 Core Melt Distribution and Behaviour in Containment Experiments ..................................... 243 Table 4-7 Systems Experiments ............................................................................................................... 255 Table 4.1.1-1 Test Matrix for AECL Interconnected Vessels Tests ........................................................ 257 Table 4.1.51-1 CYBL Boiling Test Matrix .............................................................................................. 346 Table 4.2.3-1 H2-Deflagration Tests Performed (“Hx Tests”) ................................................................ 361 Table 4.2.4-1 H2 Igniter Tests Performed (Utilities’ Program, “Ix Tests”)............................................. 364

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Table 4.2.5-1 H2-Mitigation Tests Performed (“Gx Tests” and VGB Test Ix11) ................................... 367 Table 4.2.6-1 Test Matrix for BMC Kx Series Tests ............................................................................... 370 Table 4.2.6-2 Initial Conditions for BMC Kx Series Tests ..................................................................... 371 Table 4.2.6-3 Initial Conditions for BMC Kx Series Tests (continued) .................................................. 372 Table 4.2.7-1 Test Matrix for BMC Ex Series Tests ............................................................................... 375 Table 4.5.1-1 IET Experiments in Zion Geometry .................................................................................. 531 Table 4.5.2-1 IET Experiments in Surry Geometry ................................................................................. 534 Table 4.5.9-1 Test Matrix for SERENA-2 KROTOS and TROI Experiments ........................................ 562 Table 5-1 List of Phenomenon without Identified Experiments for CFD Validation .............................. 611

Figures

Figure 2.1.1-1 German Four Loop PWR Type Konvoi NPP .................................................................... 29 Figure 2.1.1-2 Typically VVER1000/V320 NPP [2.3] ............................................................................. 30 Figure 2.1.1-3 German BWR Type 72 NPP ............................................................................................. 32 Figure 2.1.2-1 CANDU 6 Containment .................................................................................................... 34 Figure 2.1.2-2 CANDU 6 Reactor Assembly ........................................................................................... 36 Figure 2.1.2-3 CANDU 6 Fuel Channel ................................................................................................... 37 Figure 4.1.1-1 Layout of AECL Interconnected Vessels ........................................................................ 256 Figure 4.1.1-2 Elbows used in AECL Interconnected Vessels ............................................................... 257 Figure 4.1.3-1 AECL Large Scale Gas Mixing Facility ......................................................................... 260 Figure 4.1.8-1 EREC BC-V-213 Facility................................................................................................ 269 Figure 4.1.17-1 Schema of the GKSS Test Facility ................................................................................ 284 Figure 4.1.18-1 MISTRA Facility for ISP-47 Test ................................................................................. 287 Figure 4.1.19-1 MISTRA Facility for M7 Test ...................................................................................... 290 Figure 4.1.20-1 MISTRA Facility for M8 Test ...................................................................................... 293 Figure 4.1.23-1 NUPEC Facility ............................................................................................................ 298 Figure 4.1.24-1 Comparison Between ESBWR and PANDA Facility ................................................... 301 Figure 4.1.28-1 Layout of the PANDA Facility ..................................................................................... 306 Figure 4.1.29-1 SVUSS Vertical vessel (~11 m³) with Bubble Condenser ............................................ 308 Figure 4.1.30-1 THAI Facility - General Layout with Removable Internals. ......................................... 310 Figure 4.1.41-1 Layout of CONAN SARNET Benchmark No. 1 .......................................................... 325 Figure 4.1.48-1 MISTRA Facility for LOWMA Test ............................................................................ 337 Figure 4.1.49-1 PANDA Facility for OECD/SETH Tests ...................................................................... 339 Figure 4.1.50-1 PANDA Facility for OECD/SETH-2 Tests................................................................... 342 Figure 4.1.55-1 Schematic of the Small Scale Burst Test Facility ......................................................... 353 Figure 4.2.3-1 Hx-, Ix- and Gx-Test Geometries A to E ........................................................................ 359 Figure 4.2.3-2 Hx-, Ix- and Gx-Test Geometries G to K ........................................................................ 360 Figure 4.2.25-1 Schematic Illustration of the 3 Cases to be Investigated in the PET-tube ..................... 396 Figure 4.2.25-2 Schematic of PET Facility in Configuration 1 .............................................................. 396 Figure 4.2.25-3 Schematic of PET Facility in Configuration 2 .............................................................. 397 Figure 4.2.26-1 THAI HD-tests Instrumentation .................................................................................... 400 Figure 4.2.27-1 THAI HR-tests: General Experimental Set-up .............................................................. 403 Figure 4.2.28-1 THAI Hydrogen Combustion During Spray Operation ................................................ 406 Figure 4.2.35-1 HYKA-A2 Facility ........................................................................................................ 416 Figure 4.2.36-1 HYKA-A3 Facility ........................................................................................................ 419 Figure 4.2.39-1 Test Section for LACOMECO HYDET06 Test ............................................................ 422 Figure 4.2.44-1 Schematic of CHANNEL Test Facility ......................................................................... 428 Figure 4.2.45-1 HPHR Test Section ....................................................................................................... 430 Figure 4.2.46-1 Schematic of HYKA-A1 Facility .................................................................................. 432

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Figure 4.3.34-1 Schematic of the WALE Test Facility .......................................................................... 487 Figure 4.4.23-1 THAI Facility for Gaseous Iodine Release from Flashing Jet Test ............................... 524 Figure 4.4.24-1 Layout of the CAIMAN Test Facility ........................................................................... 526 Figure 4.5.1-1 Surtsey Vessel ................................................................................................................. 532 Figure 4.5.2-1 Models of Surry Structures in the Containment Technology Test Facility ..................... 535 Figure 4.5.2-2 Side-View of the Experiment Setup used in the IET/CTTF Tests (IET-9, 10 and 11) ... 536 Figure 4.5.2-3 Model of the Surry Bottom Head Used in the IED Experiments .................................... 537 Figure 4.5.2-4 Models of Surry Structures in the Surtsey Test Facility ................................................. 538 Figure 4.5.2-5 Side-View of the Experiment Setup used in the IET/Surtsey Test (IET-12) .................. 539 Figure 4.5.7-1 Schematic of the KROTOS JRC Facility ........................................................................ 551 Figure 4.5.8-1 Schematic of the KROTOS CEA Facility ....................................................................... 558 Figure 4.5.8-2 X-Ray Radioscopy for the KROTOS CEA Facility ........................................................ 559 Figure 4.5.8-3 Schematic of the TROI Facility ...................................................................................... 560 Figure 4.5.18-1 Schematic of SURC Test Facility ................................................................................. 577 Figure 4.5.24-1 Schematic of MACE Test ............................................................................................. 586

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EXECUTIVE SUMMARY

The Committee on the Safety of Nuclear Installations (CSNI) formed the CCVM (Containment Code

Validation Matrix) task group in 2002. The objective of this group was to define a basic set of available

experiments for code validation, covering the range of containment (ex-vessel) phenomena expected in the

course of light and heavy water reactor design basis accidents and beyond design basis accidents/severe

accidents. It was to consider phenomena relevant to pressurised heavy water reactor (PHWR), pressurised

water reactor (PWR) and boiling water reactor (BWR) designs of Western origin as well as of Eastern

European VVER types. This work would complement the two existing CSNI validation matrices for

thermal hydraulic code validation (NEA/CSNI/R(1993)14) and In-vessel core degradation

(NEA/CSNI/R(2001)21).

The report initially provides a brief overview of the main features of a PWR, BWR, CANDU and

VVER reactors. It also provides an overview of the ex-vessel corium retention (core catcher). It then

provides a general overview of the accident progression for light water and heavy water reactors. The

main focus is to capture most of the phenomena and safety systems employed in these reactor types and to

highlight the differences.

This CCVM contains a description of 127 phenomena, broken down into 6 categories:

Containment Thermalhydraulics Phenomena

Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena

Aerosol and Fission Product Behaviour Phenomena

Iodine Chemistry Phenomena

Core Melt Distribution and Behaviour in Containment Phenomena

Systems Phenomena

A synopsis is provided for each phenomenon, including a description, references for further

information, significance for DBA and SA/BDBA and a list of experiments that may be used for code

validation.

The report identified 213 experiments, broken down into the same six categories (as done for the

phenomena). An experiment synopsis is provided for each test. Along with a test description and

references, the synopsis also identifies the availability of the report and data, phenomena covered by the

test, type of test (separate effect, combined effect or integral test), covers DBA and/or SA/BDBA

conditions, range of key experimental parameters and past code validation/ benchmarks.

This CCVM has identified experiments for 93% of the phenomena requiring validation. However, if

only experiments suitable for CFD validation are considered, then only about half of the phenomena are

covered by this CCVM.

It is recommended that this work be reviewed in 5 years time to include new experiments and to

attempt to close the identified experiment gaps (phenomena lacking suitable experiments for validation).

NEA/CSNI/R(2014)3

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List of the CCVM Writing Group Members (2010 – 2013)

CANADA

CHIN, Yu-Shan (Sammy) Tel: +1 (613) 584-3311 ext. 46405

Fuel and Fuel Channel Safety Branch Email: [email protected]

AECL-Chalk River Laboratories

Station 88

Chalk River, Ontario, KOJ1J0

MATHEW, P.M. (Mani) (retired)

Fuel and Fuel Channel Safety Branch


Station 88


GLOWA, Glenn Tel: +1 (613) 584-3311 ext. 46052

Reactor Chemistry & Corrosion Branch Email: [email protected]


Station 86


DICKSON, Ray Tel: +1 (613) 584-3311 ext. 43381



Station 88


LIANG, Zhe (Rita) Tel: +1 (613) 584-3311 ext. 44484



Station 88


LEITCH, Brian Tel: +1 (613) 584-3311 ext. 43962



Station 88


BARBER, Duncan Tel: +1 (613) 584-3311 ext. 44851



Station Keys


VASIC, Aleks Tel: +1 (613) 584-3311 ext. 46111

Thermalhydraulics Branch Email: [email protected]


Station Keys


mailto:[email protected]







NEA/CSNI/R(2014)3

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FRANCE

BENTAIB, Ahmed Tel: +33 1 58 35 98 54

IRSN/PSN/SAG/BPhAG Email: [email protected]

31 av. de la Division LECLERC

B.P. 17, F92262 Fontenay-aux-Roses CEDEX

JOURNEAU, Christophe Tel: +33 (0)4 42 25 41 21

CEA Cadarache Email: [email protected]

DTN/STRI/LMA

F-13108 St Paul les Durance

MALET, Jeanne Tel: + (33-1) 69 08 87 40

Research Engineer Email: [email protected]

PSN-RES/SCA

DSU/SERAC

Institut de Radioprotection et de Sûreté

Centre d'Etudes de Saclay, Bâtiment 450

B.P. 63, F-91192 Gif-sur-Yvette CEDEX

STUDER, Etienne Tel: +33 1 69 08 83 92

MISTRA Project Manager Email: [email protected]

DEN/DM2S/STMF/LATF

Commissariat à l'Energie Atomique (CEA)

Centre d'Etudes Nucléaires de Saclay

F-91191 Gif-sur-Yvette Cedex

MEYNET, Nicolas Tel: +33 1 58 35 72 04

IRSN/PSN/SAG/BPhAG Email: [email protected]

77-83 Avenue du Général de Gaulle

B.P. 17, F92262 Fontenay-aux-Roses CEDEX

PILUSO, Pascal Tel: +33 (0)4 42 25 25 09

CEA Cadarache-DEN/DTN/dir Email: [email protected]

Bâtiment 710

13108 Saint Paul lez Durance Cédex-FRANCE

GÉLAIN, Thomas Tel: +33 1 69 08 50 61

CEA Saclay Email: [email protected]

B.P. 63

F-91192 Gif-sur-Yvette CEDEX

MICHIELSEN, Nathalie Tel: +33 1 69 08 62 37


B.P. 63










NEA/CSNI/R(2014)3

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PEILLON, Samuel Tel: +33 1 69 08 62 47


B.P. 63


PORCHERON, Emmanuel Tel: +33 1 69 08 62 47


B.P. 63


ALBIOL, Thierry Tel: 04 42 19 97 94

IRSN/PSN-RES/SEREX - Bât 327 Email: [email protected]

BP 3

13115 St Paul Lez Durance Cedex

CLÉMENT, Bernard Tel: 04 42 19 94 70

IRSN/PSN-RES/SAG Email: [email protected]

Bât. 702

BP 3

13115 St Paul Lez Durance Cedex

GERMANY

SONNENKALB, Martin Tel: +49 221 2068-770

Gesellschaft für Anlagen- und Reaktorsicherheit Email: [email protected]

(GRS) MbH

Leiter der Abteilung Barrierenwirksamkeit

Schwertnergasse 1

D-50667 Köln

WEBER, Gunter Tel: +49 89 32004 506


(GRS) mbH

Forschungszentrum

D-85748 Garching

KLEIN-HESSLING, Walter Tel: +49 221 2068 670


(GRS) mbH

Schwertnergasse 1

D-50667 Köln

ARNDT, Siegfried Tel: +49 30 88589 129


(GRS) mbH

Schwertnergasse 1

D-50667 Köln









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YANEZ, Jorge Tel: +49 (0) 721 6082 6502

Institute for Nuclear and Energy Technologies Email: [email protected]

Karlsruhe Institute of Technology

Kaiserstrasse 12, 76131

Karlsruhe, Germany

KOTCHOURKO, Alexei Tel: +49 (0) 721 6082 4028


Karlsruhe Institute of Technology (KIT)


Karlsruhe, Germany

KUZNETSOV, Mike Tel: +49 (0) 721 6082 4716


Karlsruhe Institute of Technology


Karlsruhe, Germany

ITALY

SANGIORGI, Marco Tel: +39 0516098901

Ente per le Nuove Tecnologie Email: [email protected]

l'Energia e l'Ambiente (ENEA)

Via Martiri di Monte Sole, 4

I-40129 Bologna

SPAIN

FONTANET, Joan Tel: +34 91 346 6577

Unit of Nuclear Safety Research Email: [email protected]

Divison of Nuclear Fission

CIEMAT (Edif. 12)

Avda. Complutense, 40

28040 Madrid

HERRANZ, Luis E. Tel: +34 91 346 6219

Unit of Nuclear Safety Research Email: [email protected]

CIEMAT (Edif. 12)

Avda. Complutense, 40

28040 Madrid

GARCIA DE LA RUA, Carmen Tel: +34 91 346 0226

Engineering Department Email: [email protected]

Consejo de Seguridad Nuclear (CSN)

Calle Justo Dorado Dellmans, 11

28040 Madrid








NEA/CSNI/R(2014)3

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SANTIAGO, Aleza Enciso Tel: +34 91 346 0218

Engineering Department Email: [email protected]

Consejo de Seguridad Nuclear (CSN)

Calle Justo Dorado Dellmans, 11

28040 Madrid

SWITZERLAND

ANDREANI , Michele Tel: +41 56 310 2687

Laboratory for Thermal-Hydraulics Email: [email protected]

Paul Scherrer Institut

CH-5232 Villigen PSI

DREIER, Jörg Tel: +41 56 310 2681

Scientific Programs Email: [email protected]

Nuclear Energy and Safety


OHSA/E04


PALADINO, Domenico Tel: +41 56 310 4373

Laboratory for Thermal-Hydraulics Email: [email protected]



UNITED STATES OF AMERICA

LEE, Richard Y. Tel: +1 301 251 7526

U.S. Nuclear Regulatory Commission (NRC) Email: [email protected]

Office of Nuclear Regulatory Research

M.S. CSB-A07M

Washington, DC 20555-0001

OECD Nuclear Energy Agency, Issy-les-Moulineaux, FRANCE

AMRI, Abdallah Tel: +33 1 45 24 10 54

OECD-NEA / Nuclear Safety Division Fax: +33 1 45 24 11 29

Le Seine St-Germain Email: [email protected]

12 bd des Iles

F-92130 Issy-les-Moulineaux







NEA/CSNI/R(2014)3

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Acronyms

ABWR Advanced Boiling Water Reactor

AEC Atomic Energy Commission

AECL Atomic Energy of Canada Limited

AICC Adiabatic Isochoric Complete Combustion

ANL Argonne National Laboratory

AP1000 Advanced Pressurized Reactor (design by Westinghouse)

ARVI Assessment of Reactor Vessel Integrity

BC Bubble Condenser

BCC Bubble Condenser Containment

BDBA Beyond Design Basis Accident

BMWi Bundesministerium für Wirtschaft und Technologie

BNWL Battelle NorthWest Laboratory

BR Blockage Ratio

BWR Boiling Water Reactor

CANDU CANada Deuterium Uranium

CCVM Containment Code Validation Matrix

CEA Commissariat à l'Energie Atomique

CFD Computational Fluid Dynamics

CHF Critical Heat Flux

CIEMAT Centro de Investigaciones Energéticas, Medioambientales y

Tecnológicas

C-J Chapman-Jouguet equilibrium detonation model

CNRS Centre National de la Recherche Scientifique

COM COMbined Effects test

CRDM Control Rod Drive Mechanisms

CSN Consejo de Seguridad Nuclear

CSNI Committee on the Safety of Nuclear Installations

DBA Design Basis Accident

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DCH Direct Containment Heating

DDT Deflagration-to-Detonation Transition

DF Decontamination Factor

DGRS Drywell Gas Recirculation System

DW Dry Well

EC European Commission

ECCS Emergency Core Cooling System

EDF Électricité de France

EFADS Emergency Filtered Air Discharge System

ENEA Italian National Agency for New Technologies, Energy and Sustainable

Economic Development

EPR European Pressurised Reactor

EPRI Electric Power Research Institute

ERMSAR European Review Meeting on Severe Accident Research

ESBWR Economic Simplified Boiling Water Reactor

EURATOM European Atomic Energy Community

FA Flame Acceleration

FCI Fuel Coolant Interaction

FP Fission Product

GDCS Gravity Driven Cooling System

GRS Gesellschaft für Anlagen- und Reaktorsicherheit

HEDL Hanford Engineering Development Laboratory

HPME High Pressure Melt Ejection

INT INTegral tests (entire system)

IRSN Institut de Radioprotection et de Sûreté

IRWST In-Containment Refueling Water Storage Tank

ISP International Standard Problem

IVR In-Vessel Retention

JRC-Ispra Joint Research Centre of the European Commission, Ispra, Italy

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KAERI Korea Atomic Energy Research Institute

KIT Karlsruhe Institute of Technology

LAC Local Air Cooler

LB-LOCA Large Break Loss Of Coolant Accident

LDA Laser Doppler Anemometry

LOCA Loss Of Coolant Accident

LP Lumped Parameter

LWR Light Water Reactor

MCCI Molten Core Concrete Interaction

MV Main Vent

NEA Nuclear Energy Agency

NPP Nuclear Power Plant

NRC Nuclear Regulatory Commission

OECD Organization for Economic Co-operation and Development

PAR Passive Autocatalytic Recombiner

PCC Passive Containment Cooling

PCCS Passive Containment Cooling System

PDI Phase-Doppler Interferometry

PHTS Primary Heat Transport System

PIRT Phenomena Identification and Ranking Table

PIV Particle Image Velocimetry

PSI Paul Scherrer Institut

PSS Pressure Suppression System

PWR Pressurized Water Reactor

RCS Reactor Coolant System

RI Radiolytic Iodine

RPV Reactor Pressure Vessel

SA Severe Accident

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SAMG Severe Accident Management Guideline

SARNET Severe Accident Research NETwork of Excellence

SARNET2 Severe Accident Research NETwork of Excellence 2

SBO Station Black Out

SE Separate Effects test

SG Steam Generator

SGTR Steam Generator Tube Rupture

SNL Sandia National Laboratories

SOAR State Of the Art Report

SRV Safety Relief Valve

TEDA Triethylene Di-Amine

US DOE United States Department Of Energy

VB Vacuum Breakers

VTT VTT Technical Research Centre of Finland

VVER Vodo-Vodyanoi Energetichesky Reactor

WW Wet Well

WWER Wasser-Wasser-Energie-Reaktor

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1 INTRODUCTION

1.1 Background

Safety analysis for nuclear power plants require validated codes, and these in turn require

suitable experimental datasets. In early 2000, there existed two NEA/CSNI documents that provided

a matrix of experiments that can be used for code validation of in-vessel phenomena [1.1] and for

thermal-hydraulics in the heat transport system [1.2]. However, it was recognized at the time that

there was a lack of a similar validation matrix document for containment (ex-vessel) phenomena, and

the CSNI CCVM (Containment Code Validation Matrix) task group was formed in 2002 as a result.

The CCVM task group was asked to document a basic set of experimental data that can be applied to

validate containment phenomena. This report would be a compliment to the work that has been done

in the State-of-the-Art Report on Containment Thermal Hydraulics and Hydrogen Distribution [1.3],

in which the most important phenomena for simulating containment thermal hydraulics and hydrogen

distribution during the core damaged phase of a PWR severe accident are cross-referenced to

internationally available experiments, mostly integrated ones. Thus, the CCVM work would extend

from thermal hydraulics and hydrogen distribution to include combustion processes, aerosol [1.4] and

fission product behaviour including iodine chemistry, and core melt behaviour.

1.2 Objectives and Scope

The objective and scope of the containment code validation matrix report is to define a basic set

of available experiments for comparison of measured and calculated parameters covering the full

range of containment phenomena expected in the course of light and heavy water reactor design basis

accidents and beyond design basis/severe accidents. It will consider phenomena and processes

relevant to pressurised heavy water reactor (PHWR), pressurised water reactor (PWR) and boiling

water reactor (BWR) designs of Western origin as well as of Eastern European VVER types.

1.3 Structure of the Report

The report is broken down into 6 sections. The present section provides an introduction. The

second section provides a general overview of the various reactor types and accident scenarios.

Section 3 provides a detailed description for the 127 containment phenomena. It also shows the

significance of each phenomenon for DBA and SA/BDBA accidents. More than 200 experiments

were identified by this project and they are described in Section 4. A phenomena vs. experiments

cross-matrix is discussed in Section 5 (Appendix A contains the actual cross-matrix). The cross-

matrix can be used to identify which experiment can be used to validate a particular phenomenon. As

well, this Section 5 identifies the gaps in this validation matrix (phenomena for which no experiments

are available in this validation matrix document). The final section provides a summary and a

recommendation to revisit this work in 5 years time to address the gaps.

To make navigating this document easier, all of the phenomena and experiments titles are

hyperlinked to their respective section of the report.

Some of the text is highlighted in red to denote that information was not available (or there was

some uncertainity) at the time this document was written.

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1.4 References

[1.1] OECD, “In-Vessel Core Degradation Code Validation Matrix”, NEA/CSNI/R(2001)21,

2001 February

[1.2] OECD, “Separate Effects Test Matrix for Thermal-Hydraulic Code Validation”,

NEA/CSNI/R(93)14, 1993 September

[1.3] OECD, “State-of-the-Art Report on Containment Thermal Hydraulics and Hydrogen

Distribution”, NEA/CSNI/R(99)16, 1999 June

[1.4] Hans-Josef Allelein, Ari Auvinen, Joanne Ball , Salih Güntay, Luis Enrique Herranz,

Akihide Hidaka , Alain V. Jones, Martin Kissane, Dana Powers , Gunter Weber , “State-of-

the-Art Report on Nuclear Aerosols”, NEA/CSNI/R(2009)5, 2009 December

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2 GENERAL OVERVIEW OF CONTAINMENT AND ACCIDENT PROGRESSION

2.1 Plant Types

The objective of this chapter is to provide a general overview of the various nuclear power plants

considered in the preparation of this containment (ex-vessel) code validation matrix. The plant types

are limited to:

Light Water moderated Reactors (LWRs)

o Pressurized Water Reactors (PWRs)

o Eastern European PWRs of type VVER

o Boiling Water Reactors (BWRs)

Heavy water moderated reactors

o Canadian Deuterium-Uranium (CANDU) pressurized water reactors

2.1.1 Light Water Reactors

This section provides an overview of the design (mainly focused on containment) of western

PWRs and BWRs and Eastern European VVER-1000 types. As well, there is a section on ex-vessel

corium retention (or core catchers) which is a new concept for Gen III+ reactors (and also VVER

1000 91/99).

The size of western PWRs range from single loop 510 MW(t) to large four loop 4270 MW(t)

units. Despite such differences, which are mainly reflected in the fission product and core material

inventories; there are no substantial differences in the basic nuclear and thermal-hydraulic parameters

of the reactor circuit [2.1]. The design of the containment however, can be significantly different, and

two types – steel containments within a concrete reactor building and prestressed concrete

containment with inner steel liner – exist. In Framatome and Westinghouse PWRs, typically spray

systems are installed to limit the containment pressure in design basis accidents (DBA) while the

German PWR containments are full pressure containments where no spray systems are needed nor

installed. The long term heat removal from the containment is provided by different systems,

sometimes including air ventilation systems. Typically measures to prevent hydrogen combustions

during DBA and severe accidents are installed, like thermal recombiners, ignitors or passive

autocatalytic recombiners (PAR). To prevent a long term containment over-pressure failure, many

plants are back fitted with containment venting systems (some of them connected with filter devices),

as part of their accident management concepts. The systems are connected to the peripheral part of

the containment and have a separate off-gas pipe towards the stack or the environment.

The size of BWRs and the construction of the containments vary significantly. The earlier

designs had less reactor power, in the order of several 100 MW(t), while the latest ones have power

output similar to that of the large PWRs. Likewise, the containment design and size varies

significantly, with both types – steel containments within a concrete reactor building and prestressed

concrete containment with inner steel liner – present. Well known BWR containments are of the type

Mark I, II or III. In Germany two different designs exists, the so called BWR type 69 (steel

containment) and the BWR type 72 (prestressed concrete containment with inner steel liner). In most

BWR containments, spray systems are installed in the drywell (though spray systems are not

necessarily safety relevant in all installations). Most of the BWR containments are inerted by

nitrogen during normal operation. If not, typically ignitors or passive autocatalytic recombiners

(PAR) are installed to prevent hydrogen combustions during DBA and severe accidents. To prevent a

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long term containment over-pressure failure in case of an accident, filtered or unfiltered containment

venting systems may be backfitted as part of accident management concepts. The systems are

connected typically to the gas phase of the condensation pool of the containment and have a separate

off-gas pipe towards the stack or the environment.

2.1.1.1 PWR

The description of a PWR is focused on the German four loop PWR Konvoi type1

(Figure 2.1.1-1) with additional text provided to cover additional features found in other PWR

containments.

The containment system of the Konvoi-units consists of the containment and the reactor building

surrounding it (see Figure 2.1.1-1). The containment provides a barrier against the release of

radioactive substances. It consists of a spherical steel vessel with a diameter of 56 m and a (nominal)

wall thickness of 38 mm that is designed to withstand the pressures and temperatures that could occur

during DBAs. The lower spherical part rests on a concrete foundation, and apart from that, the

containment is self-supported. The containment contains the entire reactor coolant system which is

under operating pressure, the spent fuel pool and parts of the directly connecting safety systems and

reactor auxiliary systems. The containment is the third barrier for compliance with the protection

objective “limitation of activity release”. During operation, the containment is continuously

ventilated and the rooms not containing the main components are accessible so that inspections,

preparatory work for inspections or fuel handling in the spent fuel pool may take place during plant

operation.

The reactor systems are located in the containment, while the emergency core cooling systems

(ECCS) are located in the reactor building annulus and reactor auxiliary building. The main reactor

systems, particularly those important to safety, are volume control system, extra borating system, and

coolant treatment. The containment related systems are hydrogen mixing system with thermal

recombiners, exhaust system, and nuclear ventilation system. Four independent trains of safety

systems exist (high pressure and low pressure injection system, accumulators). There is no

containment spray system in this design, though containment sprays exists for other PWRs, like the

Framatome and Westinghouse designs.

The reactor building, which consists of a hemispherical dome and a cylindrical base, surrounds

the containment. The reactor building has a wall thickness of approximately 1.8 m and rests on a

foundation. It is designed to protect the containment against external hazards, e.g., impacts by aircraft

or explosive detonations. The area between the lower cylindrical part of the reactor building and the

containment forms an annulus where parts of the safety systems are designed redundantly, and where

parts of the reactor systems are located. In case of an accident involving an increase in either pressure

or temperature in the containment, the containment isolation and with some time delay the annulus

isolation is triggered and the emergency sub-atmospheric pressure system in the annulus is started.

This system has the task to retain the sub-atmospheric pressure in the reactor building annulus and to

filter potential leakages from the containment vessel before discharge.

1 Provided by Martin Sonnekalb, GRS Cologne

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Figure 2.1.1-1 German Four Loop PWR Type Konvoi NPP

2.1.1.2 VVER-1000

During the design phases of the Russian VVER-1000 NPP2, great scientific efforts were made to

meet the optimal core configuration based on the small series of V-187, V-302 and V-338 models,

and large series of V-320 reactors [2.3], [2.4]. These reactor series show design differences in the

number of the control and protection system cluster rods (drives and bundles). The containment

design was similar (Figure 2.1.1-2). Although a limited number of the smaller reactors (V-187, V-

302, V-338) were built, it was primarily the type V-320 that was deployed commercially. The

VVER-1000/V-320 reactor types are light water moderated and light water-cooled PWR designs with

an electrical power output capacity of 1000 MW (thermal power output of 3000 MW(t)). The primary

system consists of four main coolant loops with one horizontal steam generator and one main coolant

pump each.


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Figure 2.1.1-2 Typically VVER1000/V320 NPP [2.3]

The containment encloses a confinement system consisting of 63 compartments in which the

pressurized components of the primary system are located. Its net volume is about 61,000 m3. It is a

pre-stressed reinforced concrete containment with an 8 mm internal steel liner. The containment

design overpressure is 0.41 MPa, covering the pressure peak after a double-ended rupture of the main

primary loop of 0.85 m diameter. A negative pressure of a maximum of 200 Pa is maintained in the

containment during normal operation.

The typical feature of the containment is that its shell is a part of the reactor building, with a

square ground plan and a side length of 66 m. It is separated from the non-leak-tight lower parts of

the building by a 3 m reinforced concrete plate, but connected to one rectangular leak-tight room in

the lower part of the building that houses the main emergency core cooling recirculation sump.

The emergency core cooling pumps, along with other support equipment, are located in the lower

part of the reactor building. The square building extends above the containment shell base plate, up to

about 40 m above ground, and protects a large part of the containment shell against external impact

and also improving the shielding of the primary system. There is a narrow gap between the

cylindrical shell and the cylindrical inner shaft in the square building.

The containment-spray system is needed to control the effects of leakage accidents associated

with the primary and secondary system within the containment, and is a three-train system resistant to

external impacts. The three trains are physically separated and emergency power supplied. They use

the common boric acid storage tank (sump) of the emergency cooling system as a water source.

During a potential accident, it functions to:

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Depressurize containment

Wash out airborne fission products

Discharge a part of the residual heat

Ensure emergency filling of the spent-fuel pool via a connecting line to the pool cooling system.

2.1.1.3 BWR

This section provides a description of a BWR, focused on the German BWR type 723

(Figure 2.1.1-3). Additional information regarding other types of BWR are provided as needed.

The containment-concept of BWR type 72 consists of the internally located separate containment

vessel (primary containment made of prestressed concrete with inner steel liner) and the outside

reactor building or secondary containment. Both buildings are based on a common foundation plate

with a diameter of 52 m and thickness of 3 m.

The containment vessel consists of pre-stressed concrete cylinder with an outer diameter of 29 m

and a height of about 40 m. The inner surface is covered with a gas proof steel shell of approximately

8 mm thickness. Inside the containment, there are the reactor pressure vessel and the pressure

suppression system, which consists of the drywell and wetwell (suppression pool). The wetwell has a

water pool with approx. 3,000 m3 deionised water, to condense the escaping steam during the loss-of-

coolant accident considered in the design (double-ended rupture of the main coolant line, the so called

2A break), thus limiting the pressure within the containment and the load to the building. During

events which lead to increased activity release into the containment, a direct sealing is ensured

because pipes penetrating the containment are equipped at least with two isolation valves, where one

of these is arranged inside and the other outside the containment, unless it is not conflicting with

safety related reasons (e.g., reactor scram). Thus the containment serves as an activity barrier for safe

enclosure of radioactive material, which is also efficient during events with leakages from the reactor

coolant pressure boundary.

The pressure suppression system has the task to condense the escaping steam in case of loss-of-

coolant accidents, thus suppressing the pressure. Furthermore, it is considered as a passive part of the

emergency cooling. The pressure suppression system consists of the wetwell, the condensate pipes

from the drywell into the wetwell, and the check valves between the wetwell and the drywell. The

water pool in the wetwell serves as the water supply for feeding the reactor pressure vessel for the

emergency cooling and residual heat removal systems and as a substitute heat sink in case of any loss-

of-coolant accidents where the main heat sink is not available.


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Figure 2.1.1-3 German BWR Type 72 NPP

The secondary containment (containment building) consists of ferroconcrete with an outside

diameter of 50 m and a thickness of 1.8 m. It encloses the containment and serves first of all as an

additional shielding of the surrounding area against ionising radiation, furthermore it protects against

external events caused by natural events like, e.g., earthquakes and flood, as well as aircraft crash,

fire, explosion blast wave and acts of sabotage. Additionally, the secondary containment serves for

retention of potential leakages from the containment so that these are controlled via the sub-

atmospheric pressure holding system and released through suspended solids filter and activated

carbon filter to the vent stack.

The spent fuel pool is located in the secondary containment above the containment. The

containment head has to be removed for fuel loading.

In case of an accident with pressure or temperature increase in the containment, the containment

isolation is triggered and the emergency sub-atmospheric pressure system is started. This system has

the task to retain the sub-atmospheric pressure in the reactor building and to filter potential leaking

from the containment vessel before discharge to the atmosphere.

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2.1.1.4 Corium Retention

There are two concepts for corium retention, in-vessel and ex-vessel, though only ex-vessel

corium retention is discussed below.

Some generation III+ reactors (and the VVER 1000-91/99) employ an ex-vessel core retention

concept, commonly called a core catcher. A brief summary of the core catchers described in [2.2] is

provided in this section.

In the EPR design, the corium is first accumulated on top of a sacrificial concrete layer located in

the lower pit. After the sacrificial layer is ablated and the melt plug (located under the sacrificial

concrete layer) has been thermally destroyed, a path opens that leads the melt down a lateral pathway

to be spread across a 170 m2 metallic structure, core catcher. As the melt spreads in the core catcher,

gravity-drive water fills the space below and around the core catcher, and eventually over the sides of

the core catcher to cover the top of the melt. The bottom and sides of the core catcher have

integrated, open cooling channels to transfer heat from the melt to the coolant.

The VVER 1000-91-99 (have been built in Tianwan nuclear power plant in China) core catcher

concept is based on a crucible that is located under the reactor pressure vessel. The crucible is a large

vessel (filled with a sacrificial material that is capable of fully oxidizing the metallic melt and prevent

any focussing effects) which would catch the core melt if the reactor pressure vessel were to fail. The

wall and floor of the core catcher is a ring heat exchanger, with re-circulating cooling water. The

sacrificial material would invert the oxidic and steel layers in the molten pool, and once inverted,

water could be poured on top of the melt (oxidic melt surface) without danger of steam explosions or

hydrogen production by steam-metal reactions.

Forschungszentrum Karlsruhe (FZK) developed another core catcher concept, called COMET,

based on the fragmentation of corium and porosity formation. The basemat would be protected by a

water cooled layer of sacrificial concrete. The bottom of the sacrificial concrete would be supplied by

water through channels or a porous concrete layer, and the water would then be forced upwards

through the melt, with the resulting water evaporation breaking up the melt and creating a porous

solidified mass from which heat could easily be removed. The corium would be expected to solidify

within 1 hour with this design, and any further release of fission products from the corium would

cease as a result.

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2.1.2 Pressurized Heavy Water Reactors (CANDU)

There are basically two types of CANDU reactors, single unit CANDU 6 (~600 MWe) and

multi-unit CANDU stations (each station has four to eight units, each station has a total generating

capacity ranging from 2,160 to 3,520 MWe). The biggest difference in the containment between the

single and multi-unit CANDU is the use of a vacuum building by the multi-unit stations. In a multi-

unit station, all units are connected by a fuelling machine duct to a common vacuum building. This

vacuum building is kept at a very low absolute pressure and is designed to contain any steam release

and to quickly return the containment internal volume to a negative gauge pressure. As well, the

vacuum buildings are equipped with a filtered discharge (Emergency Filtered Atmospheric Discharge

System) to keep the vacuum building sub-atmospheric during an accident.

The containments for a CANDU are generally larger than a PWR or BWR of a similar

generation capacity. Unless otherwise stated, the remaining discussion on CANDU will be for a

CANDU 6, see Figure 2.1.2-1.

Figure 2.1.2-1 CANDU 6 Containment

A CANDU 6’s containment has a free volume of 50,300 m3. The containment consists of a

cylindrical pre-stressed concrete shell, 0.9 m thick, with an internal height of 51.9 m and diameter of

41.5 m. The floor of the reactor compartment is 2.45 m thick concrete. This includes the floor of the

reactor vault, which is lined with 5/8” thick steel.

The common safety systems (related to containment) for CANDU are:

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Thick concrete containment building to withstand high pressures. The inner concrete surface is

either lined with steel (stainless steel or carbon steel coated with a coal-tar-epoxy layer for

corrosion protection), or a polyurethane liner

Dousing system (supplied by a large tank of water located in the dome of the containment) with

nozzles spraying water upwards.

Local Air Coolers

Passive Autocatalytic Recombiners (PARs) to recombine hydrogen

Ignitors to ignite hydrogen at concentrations near the lean flammability limit (Note: not all

CANDUs are equipped with ignitors)

The CANDU pressure tubes act like the reactor pressure vessel for a LWR. They are located

inside a calandria vessel (for a CANDU 6, radius of 3.8 m and length of 5.9 m with an internal

free volume of about 220 m3) that is filled with heavy water at about 70ºC and near atmospheric

pressure (~123 kPa(a)). The calandria vessel is then located inside a reactor vault (or calandria

vault), see Figure 2.1.2-2. The reactor vault has a total free volume of about 534 m3 and is filled

with light water at ~38ºC.

Emergency core cooling (emergency coolant injection system) – maintains a flow of coolant

through the primary heat transport system through all channels after the primary heat transport

system de-pressurizes.

Filtered air discharge system (CANDU 6, used during normal operation to ventilate containment

atmosphere) and emergency filtered air discharge system (multi-unit stations, maintain sub-

atmospheric pressure in vacuum building during an accident).

One of the key differences between the CANDU and PWR/BWR is the use of pressure tubes in

lieu of a large single reactor pressure vessel. For a PWR/BWR, ex-vessel terminology is meant

outside of the reactor pressure vessel. For a CANDU, ex-vessel is meant to be outside of the

calandria vessel.

Another unique feature of a CANDU is its online refuelling capability. This is possible by the

use of a fuelling machine that attaches to the end fitting of each fuel channel (Figure 2.1.2-3). The

end fitting can be involved in a special case of a small LOCA described in Section 2.2.2.2.2.

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Figure 2.1.2-2 CANDU 6 Reactor Assembly

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Figure 2.1.2-3 CANDU 6 Fuel Channel

2.2 Accident Progression

Nuclear power plants are designed with safety systems to deal with Design Basis Accidents

(DBA). As well, the bulk of containment phenomena occurring during a DBA are encompassed by a

Beyond Design Basis Accidents (BDBA) or Severe Accidents (SA). As such, a general overview of a

severe accident progression is provided in this section. In addition, for a CANDU, a few additional

DBA accident scenarios are discussed which involve limited fuel damage. These overviews will

mainly be focused on containment behaviour.

2.2.1 LWR Accident Progression

This section provides a brief description of the important aspects of a LWR accident progression.

A much more detailed description can be found in [2.2].

2.2.1.1 PWR Accident Progression

Severe accidents are characterised by several phases and initiated by the same events like the

ones considered in DBA analyses. In contrary, additional safety system failures lead to the loss of

core cooling and the transition into a severe accident. First the in-vessel phase until global RPV

failure at high or at low pressure (depending on the success of the accident management measure to

depressurize the reactor coolant system before its failure). Second, an ex-vessel phase, starting with

melt discharge from the RPV follows. The melt is discharged into the reactor cavity and dependent

on the plant design a dry or wet molten core concrete interaction (MCCI) follows. Phenomena like

direct containment heating or steam explosion may become relevant as well. The long term phase is

characterised by pressure increase due to non-condensable gas release from MCCI and water

evaporation being in contact with the melt in the cavity. Different accident management measure are

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implemented often to mitigate the consequences, like the use of PARs or igniters, the activation of

spray systems and of the filtered containment venting to prevent containment over pressure failure.

Analyses showed that the total amount of hydrogen gas generated during the in-vessel phase of a

severe accident is equivalent to an oxidation fraction of 35-60% of the total quantity of zirconium in

the reactor. The higher amounts of H2 have been calculated for scenarios with late start and longer

duration of the core melt process and for high pressure core melt scenarios.

As a result of the MCCI reaction under dry or wet conditions in the cavity, there is an ongoing

release of hydrogen, carbon monoxide, carbon dioxide, and steam. The rates are dependent on the

concrete composition. The released hydrogen (in- and ex-vessel) and carbon monoxide (ex-vessel)

can be recombined by a passive autocatalytic recombiner until the oxygen has been completely

consumed in the containment, so that large combustions in the containment challenging its integrity

are prevented.

The amount of energy released from the primary circuit into the containment before RPV failure,

as well as the size and location of the break, has a great influence on the pressure, convection flows,

local gas concentration, and long term behaviour of the containment and especially on the operation

of the PARs and the filtered containment venting. Light water reactors equipped with a core catcher

will sequester the corium (when it melts though the bottom of the reactor pressure vessel) and prevent

it from reaching the basemat and progressing the accident to a MCCI phase.

2.2.1.2 BWR Accident Progression

Ex-vessel severe accident scenarios that could possibly occur in a BWR would involve local

reactor pressure vessel failure, though these would likely be at lower pressure due to the automatic

depressurisation of the vessel. The initiating events to be considered and the assumed system failures

are similar to those that could occur with PWRs. The total amount of hydrogen that could be

generated prior to a breach in the reactor pressure vessel may be equivalent to that produced from the

oxidation of 15-70% of the total quantity of zirconium in the reactor. The lower values of hydrogen

production are typical for some low pressure core melt scenarios with steam starved conditions. Due

to the automatic depressurization of the RPV, the probability of high pressure cases with high

generation of hydrogen (melt release under high pressure and DCH phenomena) is much less

compared to other cases. Different accident management measures are generally employed to prevent

uncontrolled combustion of hydrogen, and include in most cases making the containment atmosphere

inert by filling it with nitrogen gas, but also by employing passive autocatalytic recombiners and

igniters.

In general, the ex-vessel phase of a BWR severe accident involves melt that is released from a

local failure in the reactor pressure vessel into the reactor cavity (typically called the control rod

driving room) , which is part of the lower containment. It is likely that there would already water in

the cavity before RPV failure, so steam explosions could be expected. A phase of the accident

involving either dry or wet MCCI could follow, and the melt could go on to penetrate into the

basemat.

Leakages to the surrounding reactor building cannot be excluded in all cases.

The filtered containment venting connected to the wetwell gas space could be used one or more

times to prevent an over pressure containment failure, and could be done before or after the failure of

the reactor pressure vessel. Venting would be initiated at the containment design pressure, and would

be the ultimate heat sink in the case where the wetwell heat removal fails.

A spray system located typically in the upper drywell may be used as well to limit the pressure

increase or to decrease the airborne content of aerosols inside the containment.

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Core-catcher concepts also exist for Gen III+ BWRs (ESBWR and ABWR) and will also prevent

MCCI.

2.2.2 CANDU Accident Progression

2.2.2.1 CANDU Station Blackout (Severe Accident)

This accident scenario description is for a single unit CANDU 6. The accident progression is

similar for a multi-unit station, but with the use of the vacuum building to maintain containment sub-

atmospheric as discussed for the large loss of coolant accident case in Section 2.2.2.2.1. The

postulated station blackout accident sequence starts with a loss of Class IV and Class III power,

causing the loss of pumps used in systems such as the PHTS, moderator cooling, shield cooling, steam

generator feedwater, and re-circulating cooling water. The accident progresses to core damage and

disassembly (i.e., a severe accident), because of the loss of:

i) the primary heat transport system inventory,

ii) long-term emergency core cooling,

iii) moderator cooling, and

iv) shield and calandria vault cooling.

In general terms, the basic sequence of events, postulated to occur during a station blackout severe

accident, are as follows:

a. Loss of Class IV and III power results in immediate shutdown of reactor and run down of pumps

(i.e., end of active heat sinks).

b. Steam generator (SG) secondary side water boils off, removing core decay heat in the short term.

c. Depletion of steam generator water, and subsequent primary coolant heat up and pressurization.

d. Primary heat transport system liquid relief valves open and close, venting primary coolant into the

degasser condenser. Degasser condenser pressurizes and vents primary coolant to containment.

Containment pressurization triggers dousing.

e. Fuel is uncovered and heats up as the PHTS inventory decreases.

f. Dry fuel channels overheat and rupture, depressurizing the PHTS as primary coolant discharges

into the calandria vessel.

g. Moderator is partially expelled out of the calandria vessel when the fuel channel rupture

pressurizes calandria vessel, bursting the rupture disks in the calandria vessel relief ducts. Fuel

channel bellows may also fail from channel rupture, in some scenarios, allowing moderator to

drain. Remaining moderator boils off as the fuel channels dry out on the inside, increasing heat to

moderator.

h. Fuel channel sections (that are uncovered on outside of calandria tube) overheat, disassemble and

drop onto lower intact channels still cooled by remaining moderator. Some debris relocates to

bottom of calandria vessel, where remaining moderator quenches it.

i. Majority of core drops to bottom of calandria vessel, perhaps in a sudden core collapse.

Moderator is boiled off by debris at the bottom of the calandria vessel.

j. Calandria vault water cools calandria vessel and thus core debris.

k. Calandria vault water boils off. Calandria vessel assumed to fail due to debris heat-up after the

vault water level decreases below the elevation of the top of the debris in the calandria vessel.

l. Core debris relocates onto calandria vault floor. Remaining vault water boiled off.

m. Molten core concrete interaction and melt through of calandria vault floor.

n. Molten corium relocates to reactor building basement, quenched by remaining water.

NEA/CSNI/R(2014)3

40

The containment (airlock seal blow out) may fail during several of the events listed above, in

particular during periods of rapid steam generation that produce pressure peaks.

During many of the accident stages, the core debris can release radionuclides (fission products

and actinides) to the containment. The release rates can be significant during the initial fuel heat up

(via sheath failures), the suspension and heat up of core debris in the calandria vessel, and the

core-concrete interaction.

Combustible gases can also form due to Zr-steam reactions in the early part of the accident.

Later, MCCI could result in hydrogen and carbon monoxide production. Likewise, radiolysis may

produce appreciable amounts of hydrogen over the long term of a severe accident.

2.2.2.2 CANDU DBA with Limited Fuel Damage

2.2.2.2.1 CANDU Large LOCA

A Large LOCA in a CANDU involves a break in the heat transport system pressure boundary of

sufficient magnitude (break in larger diameter piping than a fuel channel) that significant voids occur

in the core. Because of the positive coolant-void-reactivity, a reactor power excursion occurs and the

reactor regulating system is not capable of maintaining reactivity balance.

A more severe case of a LB-LOCA involves impairment of the emergency coolant injection

system and is described in this section.

The initial LOCA results in a short term pressurization of containment due to the coolant

discharge out of the rupture in the heat transport system boundary. The dominant behaviour during

this period is coolant flashing, leading to containment pressurization, aerosol formation and activation

of the dousing system.

For a CANDU 6 station (single unit), the pressure peaks at the end of the short term

pressurization period (break discharge begins to decrease and dousing acts to remove heat from the

containment). Continued decrease in the break discharge and continued heat removal by local air

coolers and containment structures leads to a slow decrease in containment pressure, eventually

reaching atmospheric pressure after several days.

For a multi-unit station with a negative pressure containment system, the containment is vented

to the vacuum building. This allows the containment to remain sub-atmospheric for several days. As

the vacuum building pressure rises, the emergency filtered air discharge systems (EFADS) are

employed by the operator to dry, filter and vent the vacuum building atmosphere, to maintain it sub-

atmospheric.

The impairment of the emergency coolant injection system results in degraded cooling in a large

number of fuel channels. The fuel overheats and large quantities of fission products and hydrogen are

generated and released into containment. Once in containment, the hydrogen may be removed by

operating ignitors and passive autocatalytic recombiners. The hydrogen concentration for this type of

accident is expected to be about 4% (uniform distribution). There is a possibility that higher hydrogen

concentration pockets may develop and combustion may occur. Because of the large water

inventories in a CANDU, the containment is expected to be wet. Any fission-products (with the

exception of noble gases) are expected to be in aerosol-form in containment (condense or be

nucleation sites for steam condensation).

The pressure tubes in the broken loops can strain, ballooning outwards into contact with the

calandria tube, thereby increasing the heat transfer to the moderator. For a loss of emergency core

NEA/CSNI/R(2014)3

41

coolant injection, long-term heat removal from the broken loop is provided by the moderator (which

is actively cooled). Cooling of the pressure/calandria tubes by the moderator maintains channel

integrity and accident progression is eventually halted.

2.2.2.2.2 End Fitting Failure (Small LOCA)

A unique accident relevant to CANDU is a small LOCA due to an end fitting failure. The

containment behaviour is similar to a LB-LOCA, but at much slower rate. The unique feature of this

accident is that it can result in ejection of fuel (from a single channel, CANDU 6 contains 12 fuel

bundles per channel) into the containment. The ejected fuel is cooled only by air-steam-water mixture

and the maximum fuel temperature is expected to be less than the melting temperature. However, the

temperature may be close to that at which the fuel oxidation rate is at a maximum. Fission products

will be released directly into containment, but only for a short period following the ejection, the fuel

will then cool down and releases become minimal.

2.3 References

[2.1] OECD, “In-Vessel Core Degradation Code Validation Matrix”, NEA/CSNI/R(2001)21,

2001 February

[2.2] B.R. Sehgal (editor), “Nuclear Safety in Light Water Reactors Severe Accident

Phenomenology”, Elsevier Inc., 2012

[2.3] GRS, “Quick Look Reports of the Russian Nuclear Power Plants”, GRS-V-2.2.4/1-98, 1998

January

[2.4] J. Dienstbier, “VVER-1000 Specific Design Features”, Nuclear Research Institute Řež plc,

Report prepared for Severe Accident Research Network (SARNET) Network of Excellence,

Contract FISO-CT-2004-509065, ÚJV Z-1368-T, 2005 January

NEA/CSNI/R(2014)3

42

3 PHENOMENA

The containment phenomena of relevance for DBA and SA/BDBA are indentified in this

chapter. The phenomena are broken down into 6 categories and the information is provided in a

combination of the following tables and a separate section for each phenomenon (Sections 3.1 to 3.6):

Table 3-1 - Containment Thermalhydraulics Phenomena

Table 3-2 - Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena

Table 3-3 - Aerosol and Fission Product Behaviour Phenomena

Table 3-4 - Iodine Chemistry Phenomena

Table 3-5 - Core Melt Distribution and Behaviour in Containment Phenomena

Table 3-6 - Systems Phenomena

The tables show the significance of the phenomena to DBA and SA/BDBA accidents and also

the experiments which exhibit this phenomenon. It should be noted that the significance level were

arrived by a general consensus of the members and did not undergo any structured ranking method

(i.e., PIRT). Details of each phenomenon are provided in Sections 3.1 to 3.6.

NEA/CSNI/R(2014)3

43

Table 3-1


Phenomena Number and Title Significance Experiments Exhibiting this

Phenomenon DBA SA/BDBA

P1-1 - Stratification Major Major E1-3 - LSGMF GMBT001

E1-4 - LSGMF GMUS001

E1-6 - FIPLOC F2

E1-7 - VANAM M3 (ISP-37)

E1-15 - HDR E11.2 (ISP-29)

E1-24 - PANDA ISP-42, Phase A

E1-27 - PANDA ISP-42, Phase F

E1-28 - PANDA BC4

E1-31 - THAI TH2

E1-32 - THAI TH7

E1-33 - THAI TH10

E1-34 - THAI TH13 (ISP-47)

E1-35 - THAI HM2

E1-50 - PANDA OECD/SETH-2

E3-11 - BMC VANAM M2

P1-2 - Flashing (Flashing

Discharge)

Major Minor E1-13 - HDR V44 (ISP-16)

E1-14 - HDR T31.5 (ISP-23)

E1-44 - Marviken Test 18

E1-49 - PANDA OECD/SETH tests

E1-51 - CYBL Boiling Tests

E1-52 - ULPU CHF Tests

P1-3 - Boiling Heat and Mass

Transfer

Minor Major E1-24 - PANDA ISP-42, Phase A

E1-25 - PANDA ISP-42, Phase C

E1-26 - PANDA ISP-42, Phase E


E1-46 - TOSQAN sump tests



E1-53 - SULTAN CHF Tests

E1-54 - SBLB Boiling Tests

P1-4 - Critical Heat Flux (CHF) Minor Major E1-51 - CYBL Boiling Tests


E1-53 - SULTAN CHF Tests


P1-5 - Heat Conduction in

Solids

Major Major E1-6 - FIPLOC F2

E1-8 - EREC LB LOCA Test 1


E1-10 – EREC MSLB Test 7

E1-11 - EREC MSLB Test 9

E1-12 - EREC SLB G02

E1-17 - GKSS M1

E1-22 - NUPEC M-7-1 (ISP-35)

E1-29 - SVUSS G02


NEA/CSNI/R(2014)3

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Table 3-1




P1-6 - Convection Heat

Transfer (Natural and Forced)








E1-13 - HDR V44 (ISP-16)

E1-14 - HDR T31.5 (ISP-23)

E1-15 - HDR E11.2 (ISP-29)

E1-16 - HDR E11.4

E1-17 - GKSS M1

E1-18 - MISTRA ISP-47

E1-19 - MISTRA M7

E1-20 - MISTRA-M8

E1-22 - NUPEC M-7-1 (ISP-35)

E1-29 - SVUSS G02

E1-30 - THAI TH1

E1-31 - THAI TH2

E1-32 - THAI TH7

E1-33 - THAI TH10

E1-34 - THAI TH13 (ISP-47)

E1-35 - THAI HM2

E1-36 - TOSQAN ISP-47

E1-37 - TOSQAN Condensation Tests






E3-2 - KAEVER CsI series

E3-3 - KAEVER K187 (ISP-44)




P1-7 - Thermal Diffusion in

Fluids (No Experiments)

Minor Minor (No Experiments)

P1-8 - Radiation Heat Transfer

(No Experiments)

Minor Major/Minor (No Experiments)

NEA/CSNI/R(2014)3

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Table 3-1




P1-9 - Condensation on

Surfaces








E1-13 - HDR V44 (ISP-16)

E1-15 - HDR E11.2 (ISP-29)

E1-16 - HDR E11.4

E1-17 - GKSS M1


E1-19 - MISTRA M7

E1-20 - MISTRA-M8

E1-22 - NUPEC M-7-1 (ISP-35)





E1-29 - SVUSS G02

E1-30 - THAI TH1

E1-31 - THAI TH2

E1-32 - THAI TH7

E1-33 - THAI TH10

E1-34 - THAI TH13 (ISP-47)

E1-35 - THAI HM2



E1-40 - University of Wisconsin Flat Plate

Condensation Tests

E1-41 - CONAN SARNET Benchmark

No. 1

E1-42 - CONAN SARNET2 Benchmark

No. 2








E3-9 - Phebus FPT-1 (ISP-46)

E5-1 - IET Experiments - Zion Geometry

E5-2 - IET Experiments - Surry Geometry

NEA/CSNI/R(2014)3

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Table 3-1




P1-10 - Pool Surface

Evaporation and Condensation

Major Major E1-8 - EREC LB LOCA Test 1





E1-17 - GKSS M1




E1-29 - SVUSS G02

E1-31 - THAI TH2



P1-11 - Heat Removal by

Dousing

Major Major E1-5 - AECL-SP Dousing Test No. 1

P1-12 - Liquid Re-Entrainment

(Resuspension)

Minor Minor E5-4 - DISCO-C Tests

P1-13 - Direct Contact

Condensation






E1-17 - GKSS M1




E1-28 - PANDA BC4

E1-29 - SVUSS G02




P1-14 - Momentum Induced

Mixing in Gases

Major Major E1-3 - LSGMF GMBT001



E1-19 - MISTRA M7

E1-20 - MISTRA-M8

E1-23 - NUPEC M-8-2


E1-28 - PANDA BC4

E1-48 - MISTRA LOWMA





NEA/CSNI/R(2014)3

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Table 3-1




P1-15 - Buoyancy Induced

Mixing in Gases

Major Major E1-3 - LSGMF GMBT001


E1-6 - FIPLOC F2


E1-15 - HDR E11.2 (ISP-29)

E1-16 - HDR E11.4

E1-22 - NUPEC M-7-1 (ISP-35)

E1-23 - NUPEC M-8-2

E1-34 - THAI TH13 (ISP-47)

E1-35 - THAI HM2




E1-48 - MISTRA LOWMA






P1-16 - Pressure Wave

Propagation

Minor Major E1-55 – Small Scale Burst Test

Experiments

E5-12 - ECO Tests

P1-17 - Mixing in Water Pools Major Major E1-8 - EREC LB LOCA Test 1





P1-18 - Mass Diffusion in

Vapour

Major Major E1-50 - PANDA OECD/SETH-2

P1-19 - Laminar Flow (No

Experiments)

Minor Minor (No Experiments)

P1-20 - Turbulent Flow Major Major E1-1 - Flow through Interconnected

Vessels


E1-19 - MISTRA M7

E1-20 - MISTRA-M8

E3-22 - Aerosol Deposition in Turbulent

Vertical Conduits (Forney)


Vertical Conduits (Friedlander)


Vertical Conduits (Liu)


Vertical Conduits (Wells)

P1-21 - Critical Flow (Choked

Flow)

Major Major E1-1 - Flow through Interconnected

Vessels


P1-22 - Laminar/Turbulent

Leakage Flow

Major Major

NEA/CSNI/R(2014)3

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Table 3-1




P1-23 - Vent Clearing Major Minor E1-8 - EREC LB LOCA Test 1





E1-17 - GKSS M1





E1-28 - PANDA BC4

E1-29 - SVUSS G02


P1-24 - Pool Swell / Air

Injection

Major Minor E1-8 - EREC LB LOCA Test 1





E1-17 - GKSS M1

E1-29 - SVUSS G02


E5-12 - ECO Tests

P1-25 - Interfacial Drag (No

Experiments)

Minor Minor

P1-26 - Liquid Film Flow Major Major E1-8 - EREC LB LOCA Test 1


P1-27 - Gas Dissolved in Water

(No Experiments)

Minor Minor

P1-28 - Gas Entrainment by

Spray Droplets (Dousing)

Major Major E1-38 - TOSQAN Test 113

E1-39 - TOSQAN Spray Tests

E1-47 - CALIST PWR spray test


P1-29 - Heat and Mass Transfer

of Spray Droplets (Dousing)



E1-21 - MISTRA-MASP

E1-45 - CARAIDAS EVAP and COND

tests


P1-30 - Droplet Interaction

(Dousing)

Minor Minor

P1-31 - Mixing by Sprays Major Major E1-38 - TOSQAN Test 113

E1-39 - TOSQAN Spray Tests


P1-32 - Turbulence Induced by

Sprays

Minor Minor

NEA/CSNI/R(2014)3

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Table 3-2


Phenomena Number and

Title

Significance Experiments Exhibiting this


P2-1 - Deflagration Major Major E2-1 - LSVCTF S01

E2-2 - LSVCTF S03

E2-3 - BMC Hx series

E2-4 - BMC Ix series

E2-5 - BMC Gx Series

E2-6 - BMC Kx Series

E2-7 - BMC Ex Series

E2-8 - ENACEFF SARNET2 Tests

E2-9 - ENACEFF SARNET Test (Run

703)


717)

E2-11 - ENACEFF Run 765 (ISP-49)



E2-14 - DRIVER HYCOM MC 003


E2-16 - FZK R 0498_09


E2-18 - DRIVER HYCOM HC 020

E2-19 - DRIVER HYCOM-HC027

E2-20 - RUT HYC01

E2-21 - RUT HYC12

E2-22 - RUT HYC14

E2-23 - VGES Tests

E2-24 - NTS Tests

E2-25 - PET Tubes

E2-26 - THAI HD Series (Combustion

Tests)

E2-27 - THAI HR Series (PAR Tests)

E2-30 - LSVCTF S02

E2-31 - LSVCTF DC

E2-32 - LSVCTF 3C

E2-35 - LACOMECO UFPE2

E2-36 - LACOMECO HYGRADE10



E2-39 - LACOMECO HYDET06




E5-6 - DISCO-A2

NEA/CSNI/R(2014)3

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Table 3-2



Title



P2-2 - Hydrogen Flame

Acceleration (FA)

N/A Major E2-3 - BMC Hx series

E2-6 - BMC Kx Series

E2-7 - BMC Ex Series

E2-8 - ENACEFF SARNET2 Tests


703)


717)






E2-16 - FZK R 0498_09


E2-18 - DRIVER HYCOM HC 020

E2-19 - DRIVER HYCOM-HC027

E2-20 - RUT HYC01

E2-21 - RUT HYC12

E2-22 - RUT HYC14

E2-23 - VGES Tests

E2-24 - NTS Tests

E2-25 - PET Tubes

E2-35 - LACOMECO UFPE2






E5-6 - DISCO-A2

P2-3 - Deflagration-to-

Detonation Transition (DDT)

N/A Major

P2-4 - Hydrogen Detonation N/A Major E2-16 - FZK R 0498_09



P2-5 - Quenching of

Detonations by Geometrical

Constrains

Minor Major/Minor

P2-6 - Quenching Major Major E2-12 - ENACEFF Run 736 (ISP-49)


P2-7 - Hydrogen Diffusion

Flame (Standing Flame)

Minor Major/Minor E2-29 - DFF SFSER01

E2-33 - LSVCTF CIC

E5-6 - DISCO-A2

P2-8 - Hydrogen Mitigation -

Passive Autocatalytic

Recombiners

Major Major E2-5 - BMC Gx Series

E2-27 - THAI HR Series (PAR Tests)

E2-41 - H2PAR E 12

E2-42 - H2PAR E 13

E2-43 - H2PAR E 3

E4-21 - THAI HR32

NEA/CSNI/R(2014)3

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Table 3-2



Title



P2-9 - Hydrogen Ignition by

PARs (Weak Ignition)

Major Major E2-27 - THAI HR Series (PAR Tests)

E2-41 - H2PAR E 12

P2-10 - Hydrogen Mitigation

by Hydrogen Ignitors (Mild

Ignition)

Major Major E2-3 - BMC Hx series

E2-4 - BMC Ix series

E2-5 - BMC Gx Series

P2-11 - Strong Ignition of

Hydrogen

Major Major

P2-12 - Jet Ignition of

Hydrogen

Major Major E2-31 - LSVCTF DC

E2-32 - LSVCTF 3C

E2-33 - LSVCTF CIC

P2-13 - Radiolysis (Hydrogen

Production by Water

Radiolysis)

Minor Major E2-34 - Gammacell Radiolysis Tests

P2-14 - Effect of Droplets on

Hydrogen Combustion

N/A Major E2-24 - NTS Tests

E2-28 - THAI Hydrogen Combustion

During Spray Operation

NEA/CSNI/R(2014)3

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Table 3-3




P3-1 - Aerosol Formation in a

Flashing Jet

Major Minor E3-16 - Whiteshell Flashing Jet Tests

E3-34 - WALE

P3-2 - Aerosol Formation in a

Steam Jet

Major Minor E3-33 - Whiteshell Steam Jet Experiments

P3-3 - Aerosol Impaction (Jet

Impingement)

Major Minor E3-34 - WALE

P3-4 - Thermophoresis Minor Major E1-43 - CSTF Tests

E3-14 - CSTF ACE

E3-18 - JAERI Thermophoresis Tests

E3-19 - PITEAS Diffusiophoresis Tests

(PDI 08, PDI 09, PDI 11 and PDI 12)


Vertical Conduits (Sehmel)

E3-26 - CSE Fission Product Transport

Tests

E3-34 - WALE

P3-5 - Diffusiophoresis Minor Major E1-43 - CSTF Tests

E3-6 - LACE LA2

E3-7 - LACE LA4


E3-13 - CSTF ABCOVE Tests

E3-14 - CSTF ACE



E3-20 - PITEAS Aerosol Condensation

Tests (PCON 01 to PCON 05)




Tests

P3-6 - Liquid Aerosol

Evaporation

Minor Major E3-14 - CSTF ACE


Tests

NEA/CSNI/R(2014)3

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Table 3-3




P3-7 - Condensation on

Aerosols

Minor Major E1-7 - VANAM M3 (ISP-37)

E1-43 - CSTF Tests

E3-1 - AHMED OECD benchmark





E3-6 - LACE LA2

E3-7 - LACE LA4


E3-12 - VICTORIA test 58


E3-14 - CSTF ACE

E3-16 - Whiteshell Flashing Jet Tests






Tests

P3-8 - Gravitational

Agglomeration

Major Major E1-43 - CSTF Tests

E3-1 - AHMED OECD benchmark

E3-6 - LACE LA2

E3-7 - LACE LA4





P3-9 - Diffusional

Agglomeration

Minor Major E1-43 - CSTF Tests

E3-6 - LACE LA2

E3-7 - LACE LA4



E3-14 - CSTF ACE

E3-17 - Clarkson College Brownian

Agglomeration


Tests

NEA/CSNI/R(2014)3

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Table 3-3




P3-10 - Turbulent

Agglomeration of Aerosols

Major Major E3-6 - LACE LA2

E3-7 - LACE LA4














P3-11 - Drop Breakup Minor Major/Minor

P3-12 - Gravitational Settling

(Drop Settling)

Major Major E1-7 - VANAM M3 (ISP-37)





E3-6 - LACE LA2

E3-7 - LACE LA4




E3-14 - CSTF ACE








Tests

E3-34 - WALE

E3-36 – VANAM-M4

P3-13 - Diffusional Deposition Minor Major E1-7 - VANAM M3 (ISP-37)




P3-14 - Inertial Deposition of

Aerosols (Also called

Impaction)

Minor Major E3-21 - Aerosol Deposition in Turbulent


E3-28 - LASS-SGTR

E3-34 - WALE

E5-25 - COLIMA CA-U4

NEA/CSNI/R(2014)3

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Table 3-3




P3-15 - Turbulent Deposition of

Aerosols

Major Major E3-14 - CSTF ACE












Tests

E3-28 - LASS-SGTR

E3-34 - WALE

P3-16 - Re-volatilisation Major Major E3-38 – Phebus FPT4 Revaporization

E3-39 – Ruthenium Revolatilisation

Studies at VTT

E3-40 – Ruthenium Transport and

Revolatilisation Studies at KFKI

E3-41 – Ruthenium deposition studies at

Chalmers University


Studies at IRSN

P3-17 - Aerosol Removal in

Leakage Paths

Major Major E3-27 - CSE Aerosol Removal Tests

E3-31 - Aerosol Trapping Effects in

Containment Penetration (A. Watanabe)

E3-32 - Aerosol transfer through cracked

concrete walls


P3-18 - Pool Scrubbing of

Aerosols

Major Major E3-10 - POSEIDON PA10


P3-19 - Radionuclide Transport Major Major E1-7 - VANAM M3 (ISP-37)


E4-13 - THAI Iod-11

E4-14 - THAI Iod-12

P3-20 - Radionuclide Decay

Heat (No Experiments)

Minor Major

P3-21 - Release Rate Change

Due to Oxidizing Environment

Major Major E3-29 - MCE, UCE and HCE Tests

E3-30 - GBI Tests


Studies at VTT



P3-22 - Containment Chemistry

Impact on Source Term

Major Major E3-9 - Phebus FPT-1 (ISP-46)

NEA/CSNI/R(2014)3

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Table 3-3




P3-23 - Ruthenium Volatility

and Behaviour in Containment

Major Major E3-39 – Ruthenium Revolatilisation

Studies at VTT



E3-41 – Ruthenium deposition studies at

Chalmers University


Studies at IRSN

P3-24 - Aerosol Removal by

Sprays (Dousing)

Major Major E3-15 - CARAIDAS Aerosol washout by

single droplet tests

P3-25 - Re-suspension (Dry) N/A Major E3-35 – AEREST (Aerosol resuspension

shock tube)

E3-36 – VANAM-M4

E3-37 – THAI Aer-1, Aer-3 and Aer-4

tests

P3-26 - Re-entrainment (Wet) Minor Major E3-8 – LACE LA5 and LA6

P3-27 - Aerosol De-

agglomeration

N/A Minor E3-28 - LASS-SGTR

NEA/CSNI/R(2014)3

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Table 3-4




P4-1 - Aqueous Phase

Oxidation and Reduction of

Iodine Species

Major Major E4-2 - RTF P9T3

E4-3 - RTF P9T1

E4-4 - RTF P9T2

E4-5 - RTF P10T2

E4-6 - RTF P10T3

E4-7 - RTF P11T1

E4-8 - RTF P0T2

E4-9 - RTF P10T1

E4-10 - RTF PHEBUS RTF1

E4-22 - LASS-GIRS DABASCO

E4-24 - CAIMAN 97/02 test

E4-25 - CAIMAN 2001/01 Test

P4-2 - Inorganic Iodine

Hydrolysis


E4-3 - RTF P9T1

E4-4 - RTF P9T2

E4-5 - RTF P10T2

E4-6 - RTF P10T3

E4-7 - RTF P11T1

E4-8 - RTF P0T2

E4-9 - RTF P10T1




P4-3 - Inorganic Iodine

Radiolysis in Water Phase


E4-3 - RTF P9T1

E4-4 - RTF P9T2

E4-5 - RTF P10T2

E4-6 - RTF P10T3

E4-7 - RTF P11T1

E4-8 - RTF P0T2

E4-9 - RTF P10T1


E4-11 - EPICUR Test Series S1, S2 and

S3



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Table 3-4




P4-4 - Homogeneous Organic

Reactions in Water Phase


E4-3 - RTF P9T1

E4-4 - RTF P9T2

E4-5 - RTF P10T2

E4-6 - RTF P10T3

E4-7 - RTF P11T1

E4-8 - RTF P0T2

E4-9 - RTF P10T1



S3



P4-5 - Iodine Reactions with

Surfaces in the Water Phase


E4-3 - RTF P9T1

E4-4 - RTF P9T2

E4-5 - RTF P10T2

E4-6 - RTF P10T3

E4-7 - RTF P11T1

E4-8 - RTF P0T2

E4-9 - RTF P10T1



S3



P4-6 - Iodine reactions with

surfaces in the gas phase


E4-3 - RTF P9T1

E4-4 - RTF P9T2

E4-5 - RTF P10T2

E4-6 - RTF P10T3

E4-7 - RTF P11T1

E4-8 - RTF P0T2

E4-9 - RTF P10T1



S3

E4-12 - THAI Iod-09

E4-13 - THAI Iod-11

E4-14 - THAI Iod-12



P4-7 - Silver Iodine Reactions

in the Water Phase

Minor Major E3-9 - Phebus FPT-1 (ISP-46)

P4-8 - Gas Phase Radiolytic

Oxidation of Molecular Iodine

(I2) (Iodine/Ozone Reaction)

Major Major E4-15 - THAI Iod-13

E4-16 - THAI Iod-14

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Table 3-4




P4-9 - Homogeneous Organic

Iodine Reactions in Gas Phase

Major Major E4-11 - EPICUR Test Series S1, S2 and

S3



P4-10 - RI (Organic Iodine)

Radiolytic Destruction


E4-3 - RTF P9T1

E4-4 - RTF P9T2

E4-5 - RTF P10T2

E4-6 - RTF P10T3

E4-7 - RTF P11T1

E4-8 - RTF P0T2

E4-9 - RTF P10T1




P4-11 - Interfacial Mass

Transfer


E4-3 - RTF P9T1

E4-4 - RTF P9T2

E4-5 - RTF P10T2

E4-6 - RTF P10T3

E4-7 - RTF P11T1

E4-8 - RTF P0T2

E4-9 - RTF P10T1



S3

E4-12 - THAI Iod-09

E4-22 - LASS-GIRS DABASCO



P4-12 - Decomposition of

Iodides (CsI) by Heat-up in

PARs

Minor Major E2-27 - THAI HR Series (PAR Tests)

E4-20 - THAI HR31

P4-13 - Iodine Filtration Major Major E4-1 - CFTF Charcoal Filter Test

P4-14 - Volatile Iodine

Trapping by Airborne Droplets

Major Major E4-22 - LASS-GIRS DABASCO

P4-15 - Iodine Retention in

Leakage Paths

Major Major E3-27 - CSE Aerosol Removal Tests

E3-31 - Aerosol Trapping Effects in

Containment Penetration (A. Watanabe)

P4-16 - I2 Interaction with

Aerosols

Minor Major E4-17 - THAI Iod-25

E4-18 - THAI Iod-26

P4-17 - Iodine Wash-down Major Major E4-19 - THAI AW

P4-18 - Pool Scrubbing of

Iodine

Major Major

P4-19 - Iodine Release from

Flashing Pool or Flashing Jet

Major Major E4-23 - OECD-THAI2 Gaseous Iodine

Release from Flashing Jet Test

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Table 3-5




P5-1 - Corium Release from

Failed Dry Reactor Pressure

Vessel

N/A Major E5-4 - DISCO-C Tests

E5-5 - DISCO-H Tests

P5-2 - Corium Entrainment Out

of the Reactor Primary Vessel

with Lateral Breaches



P5-3 - Corium Particles

Generation from the Corium

Pool




Generation from the Two Phase

Jet

N/A Major E5-1 - IET Experiments - Zion Geometry


E5-4 - DISCO-C Tests



Entrainment




Trapping

N/A Major E5-5 - DISCO-H Tests

P5-7 - Direct Containment

Heating



E5-4 - DISCO-C Tests


E5-6 - DISCO-A2

P5-8 - Corium Jet Break-up in

Water Pool

N/A Major E5-3 - FARO Tests

E5-7 - KROTOS JRC Tests

E5-8 - SERENA-2 KROTOS and TROI

Commissioning Tests

E5-9: SERENA-2 KROTOS and TROI

Tests

P5-9 - FCI and Steam

Explosion - Melt into Water Ex-

Vessel (Melt Quenching)




Commissioning Tests


Tests

E5-12 - ECO Tests

P5-10 - Pressure Load on

Corium Retention Devices




Commissioning Tests


Tests

P5-11 - Particulate Debris Bed

Formation




Commissioning Tests


Tests

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Table 3-5




P5-12 - Corium Debris (Solid)

Heat Transfer

N/A Major E5-8 - SERENA-2 KROTOS and TROI

Commissioning Tests


Tests

E5-28 – HSS-1 and HSS-3

E5-36 - FRAG

P5-13 - Molten Core Concrete

Interaction

N/A Major E5-10 - MCCI-1 Tests CCI Tests 1-3;

SSWICS tests 1-7

E5-11 - MCCI-2 Tests CCI Tests 4-6;

SSWICS tests 8-13; WCB-1

E5-13 - BALI Ex-Vessel Tests

E5-15 - VULCANO VB-U7 (EPR

concrete)

E5-16 - VULCANO VW-U1 (COMET

bottom flooding)

E5-17 - VULCANO VE-U7

E5-18 – SURC-1 and SURC-2

E5-19 - SURC-3

E5-20 - SURC-3A

E5-21 - SURC-4

E5-22 - BETA V5.1

E5-23 - ACE Phase C Tests L1, L2, L4,

L5, L6, and L7

E5-24 - MACE Tests M0, M1b, M3b, M4,

and MSET-1

E5-26 - BURN-1

E5-29 - TURC1T and TURC1SS

E5-30 – TURC2 and TURC3

E5-31 - LSL-1,2,3

E5-32 - LBL-1,2,3

E5-33 - LSCRBR-1,2,3

E5-34 - COIL-1

P5-14 - Corium Melt

Stratification

N/A Major E5-14 - BALISE Tests


E5-19 - SURC-3

E5-20 - SURC-3A

E5-21 - SURC-4




E5-31 - LSL-1,2,3

E5-32 - LBL-1,2,3

E5-33 - LSCRBR-1,2,3

E5-34 - COIL-1

E5-35 - WETCOR-1

P5-15 - Corium Spreading N/A Major E5-3 - FARO Tests

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Table 3-5




P5-16 - Molten Corium Heat

Transfer




SSWICS tests 1-7



E5-13 - BALI Ex-Vessel Tests


concrete)


bottom flooding)


E5-19 - SURC-3

E5-20 - SURC-3A

E5-21 - SURC-4


L5, L6, and L7


and MSET-1

E5-26 - BURN-1



E5-31 - LSL-1,2,3

E5-32 - LBL-1,2,3

E5-33 - LSCRBR-1,2,3

E5-34 - COIL-1

E5-37 - 1DHtFlx

P5-17 - Corium

Evaporation/Vaporization

N/A Minor E5-10 - MCCI-1 Tests CCI Tests 1-3;

SSWICS tests 1-7




E5-19 - SURC-3

E5-20 - SURC-3A

E5-21 - SURC-4


L5, L6, and L7


and MSET-1




E5-31 - LSL-1,2,3

E5-32 - LBL-1,2,3

E5-33 - LSCRBR-1,2,3

E5-34 - COIL-1

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Table 3-5




P5-18 - Corium

Solidification/Crust Formation


SSWICS tests 1-7




concrete)


bottom flooding)



E5-19 - SURC-3

E5-20 - SURC-3A

E5-21 - SURC-4


L5, L6, and L7


and MSET-1

E5-26 - BURN-1




E5-31 - LSL-1,2,3

E5-32 - LBL-1,2,3

E5-33 - LSCRBR-1,2,3

E5-34 - COIL-1

E5-36 - FRAG

P5-19 - Cracking (Crust) N/A Major E5-10 - MCCI-1 Tests CCI Tests 1-3;

SSWICS tests 1-7




L5, L6, and L7


and MSET-1

P5-20 - Ex-Vessel Corium

Coolability, Top Flooding


SSWICS tests 1-7




and MSET-1

E5-27 – SWISS-1 and SWISS-2


E5-35 - WETCOR-1


Catcher - Coolability and Water

Bottom Injection


SSWICS tests 1-7




bottom flooding)

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Table 3-5





Catcher - Corium-Ceramics

Interaction and Properties

N/A Major

P5-23 - Effect of Non

Homogeneous Ablation on Gate

Ablation

N/A Minor E5-15 - VULCANO VB-U7 (EPR

concrete)


bottom flooding)


P5-24 - Crust Anchorage N/A Major/

Minor


SSWICS tests 1-7




and MSET-1


E5-35 - WETCOR-1

P5-25 - Radionuclide Release

from MCCI and Core Catchers

N/A Minor E5-18 – SURC-1 and SURC-2

E5-19 - SURC-3

E5-20 - SURC-3A

E5-21 - SURC-4


L5, L6, and L7


and MSET-1





E5-31 - LSL-1,2,3

E5-32 - LBL-1,2,3

E5-33 - LSCRBR-1,2,3

E5-34 - COIL-1

E5-35 - WETCOR-1

E5-36 - FRAG

P5-26 - Core Catchers with

External Cooling



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Table 3-5




P5-27 - Oxidation of Corium N/A Major E5-1 - IET Experiments - Zion Geometry



SSWICS tests 1-7




E5-19 - SURC-3

E5-20 - SURC-3A

E5-21 - SURC-4


L5, L6, and L7


and MSET-1



E5-31 - LSL-1,2,3

E5-32 - LBL-1,2,3

E5-33 - LSCRBR-1,2,3

E5-34 - COIL-1

P5-28 - Corium Attack of

Metallic Liner

N/A Major

(for BWRs)

E5-38 – MC Tests

E5-39 – Plate Tests

P5-29 - Corium Release from

Failed Flooded Reactor

Pressure Vessel

N/A Major

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Table 3-6

Systems Phenomena



P6-1 - Ventilation Systems Major* Major*

P6-2 - Behaviour of Doors, Burst

Membranes, Rupture Discs etc.

Major* Major* E1-8 - EREC LB LOCA Test 1





P6-3 - Air Cooler (Fan Cooler)

Heat Transfer

Major Major E1-2 - Bruce LAC Test in Air, Test No.

50

P6-4 - Pump Performance

including Sump Clogging

Major Major

P6-5 - Passive Cooling by

Internal and External Condensers

Major Major E1-28 - PANDA BC4

P6-6 - Aerosol Removal in

EFADS

Major Major E6-1 - CSE EFADS Tests

E6-2 - ACE-CSTF EFADS Tests

E6-3 - ACE-LSFF EFADS Tests

* – especially for containments with multiple rooms

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3.1 Containment Thermalhydraulics Phenomena

3.1.1 P1-1 - Stratification

Description:

Fluid density differences can lead to stratification into layers. Density variation may be due to

temperature difference or composition. Therefore, vertical distributions of temperature and gas

species concentrations are in general non-similar.

Most accident scenarios involve the transport of fluids between vessels. Depending on the

density differences between the fluids in the vessels, as well as a variety of other parameters (flow

rate, geometry of the fluid release and of the receiving vessel, interaction between various

mass/momentum/heat sources, etc.), the fluid in the receiving vessel can either remain well mixed or

become vertically stratified. Moreover, stratification can be produced by differential heating within a

fluid body, differences in the temperature of the structures, and by condensation/evaporation. In

partially divided enclosures (such as containment buildings), fluid trapping in partitions can lead to

stable stratified conditions, which in turn can shut down large-loop natural circulation, and as such

hinder global mixing throughout the building.. Stratification can be stable or unstable, steady or

transient. In the transient case, the motion of the density interface (stratification front) is of

paramount importance. Although the phenomena are often 3-D, stratification is more often described

as a 1-D phenomenon governed by density differences in the vertical direction. This phenomenon is

strongly connected to forced mixing because stratification occurs when the driving forces to cause

forced mixing are weaker than natural buoyancy. In this chapter, the conditions leading to mixing and

stratification are discussed. Erosion/break-up of stratification is discussed in the chapters related to

phenomena:

P1-14 - Momentum Induced Mixing in Gases

P1-15 - Buoyancy Induced Mixing in Gases

P1-17 - Mixing in Water Pools

Stratification occurs in gas spaces (e.g., dry containments, Drywell of BWRs, etc.) as well as in

water bodies (e.g., ECCS water storage pools, BWR Wetwell, etc.). The first issue is covered here,

the second being addressed in phenomenon P1-17.

One of the most investigated issues in relation to gas stratification in large volumes is the

development of criteria for the establishment of stratified conditions and of computational methods

for predicting gas distribution. The development of stratification depends, among other things, on the

stability of the horizontal currents at the density interface and the overturning of the flow due to the

interaction with vertical wall (as would occur in enclosures with large aspect ratio). Stratification also

affects the inter-compartment transport, which is often controlled by small density differences

between adjacent compartments. In the case of hydrogen release, the main concern is the formation

of explosive mixtures in some regions of the containment. Accurate prediction of stratification is also

important for the evaluation of the effect and performance of accident mitigation measures and

performance of equipment for long-term cooling (see phenomenon P6-3 - Air Cooler (Fan Cooler)

Heat Transfer). In BWRs and other designs with small free volume, stratification affects the amount

of gas transported to the suppression chamber and therefore the system pressure.

References:

B. Gebhart, Y. Yaluria, R. Mahajan, B. Sammakia, “Buoyancy-Induced Flows and Transport”,

Hemisphere Publishing Corp., p. 334, 1988

A. Bejan, “Convective Heat Transfer”, John Wiley & Sons, 2nd

Edition, pp. 247-251, 1995

NEA/CSNI/R(2014)3

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P.F. Peterson, “Scaling and Analysis of Mixing in Large Stratified Volumes”, Int. J. Heat Mass

Transfer, Vol. 37, Supplement 1, pp. 97-106, 1994 March

H. Zhao and P.F. Peterson, “An Overview of Modeling Methods for Thermal Mixing and

Stratification in Large Enclosures for Reactor Safety Analysis”, The 8th Int. Topical Meeting on

Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8), N8P0079, Shanghai, China, 2010

October 10-14

N.B. Kaye and G.R. Hunt, “Overturning in a filling box”, J. Fluid Mech., Vol. 576, pp. 297-323, 2007

M. Andreani, F. Putz, T.V. Dury, C. Gjerloev and B.L. Smith, “On the Application of Field Codes to

the Analysis of Gas Mixing in Large Volumes: Case studies using CFX and GOTHIC”, Annals of

Nuclear Energy, Vol. 30, pp. 685-714, 2003

Prepared by: M. Andreani (PSI)

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3.1.2 P1-2 - Flashing (Flashing Discharge)

Description:

The discharge from a reservoir or vessel at a high pressure to either a vessel at a lower pressure,

or to atmosphere, is governed by factors like flow area, pressure difference, wall friction, exit losses,

state of the fluid (liquid, gas, or two phase mixture), and upstream flow regime. As the fluid flows to

the break, there may be a physical point in the system at which a critical pressure differential is

reached. Any subsequent changes in the downstream pressure no longer influence the fluid velocity

or mass flow. When this happens, the fluid velocity is at sonic velocity and the flow is said to be

critical. The actual point at which the flow becomes critical may be internally within the reservoir or

it may be at the break plane.

During the process of discharge or blowdown, the acceleration of the fluid and the subsequent

flashing of the discharge are important in determining the conditions of the discharge in the

downstream volume. This flashing phenomenon involves depressurization of the fluid from the break

plane to the asymptotic plane at which point the overall pressure is considered to be at the ambient

level. As the two-phase fluid is being depressurized, rapid vaporization and expansion of the fluid

occurs. Formation of very fine droplets may result. The end state of the fluid depends on the fluid

conditions at the break plane (fluid temperature, pressure and quality), fluid velocity and the pressure

differences between the break plane and the ambient.

In the case of primary or secondary pipe breaks, coolant is vented through a break opening from

an upstream reservoir to the containment volume. As the fluid accelerates through the break opening,

the local static pressure decreases while kinetic energy increases. Two-phase conditions occur at the

break plane if upstream conditions are that of saturated or slightly subcooled liquid. For the majority

of breaks considered in safety analyses, the pressure at the break plane is many times that of the

ambient pressure inside containment and choking occurs. Subsequently, the fluid expands rapidly

(within one to three equivalent break diameter) to the “asymptotic plane” where the local static

pressure is the same as the ambient containment pressure. Local jet velocity is also very high. Other

important phenomena include such factors as jet expansion and impingement on structures.

References:

J.C. Leung and M.A. Grolmes, “The discharge of two-phase flashing flow in a horizontal duct”,

AIChE Journal, Vol. 33, Issue 3, pp. 524-527, 1987 April

F.P. Incropera and D.P. Dewitt, “Fundamentals of Heat and Mass Transfer”, 5th Edition, John Wiley

& Sons, pp. 230-240, 2000

Prepared by: Y.S. Chin (AECL)

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3.1.3 P1-3 - Boiling Heat and Mass Transfer

Description:

In the literature, boiling is often defined as a phase change process in which vapour bubbles are

formed either on a heated surface or in a superheated liquid layer adjacent to the heated surface. It

differs from evaporation at predetermined vapour/gas-liquid interfaces because it also involves

creation of these interfaces at discrete sites on the heated surface. In general, however, it is the liquid-

to-vapour phase change that occurs when the temperature of the liquid is higher than the saturation

temperature at the corresponding to the liquid pressure. Boiling of a water body can occur by

deposition of heat, which can be supplied by a hot surface, by a hot fluid or by thermal radiation.

Boiling will also occur when the pressure is reduced below the saturation pressure corresponding to

the temperature of the bulk of the liquid. In this case, one refers to this phenomenon as to “flashing”.

Boiling heat transfer plays a role in steam explosions as well.

Different heat transfer regimes have been identified. Sub-cooled boiling involves formation of

bubbles at the surface when the bulk fluid temperature is below the boiling temperature. Saturated

boiling occurs when the bulk fluid is at the boiling temperature. In film boiling, the liquid is no

longer in direct contact with the heating surface, and the vapour blankets all or an appreciable portion

of the heating surface. For systems where the temperature is controlled (such as heat exchangers),

transition boiling occurs, which is characterised by rapid alternating of periods of film boiling and

periods when the surface is wet.

Boiling heat transfer is relevant for external cooling of the reactor vessel under severe accident

conditions. Boiling could also occur in the suppression pool of a BWR if the water is not adequately

cooled during a design base accident, or following depressurisation to prevent the pressure to increase

above design limits under severe accident scenario. In a PWR containment, evaporation of the sump

is considered in many realistic accident scenarios. Boiling can also occur, due to either volume

heating generated by deposited radionuclides, or to extreme decay heat release, or to controlled

depressurisation caused by venting or fast depressurisation caused by containment leakages. Boiling

of the sump prevents the use of pumps feeding safety equipment.

References:

V.K. Dhir, “Boiling Heat Transfer”, Ann. Rev. Fluid Mech., Vol. 30, pp. 1, 1998

N.I. Kolev, “Uniqueness of the Elementary Physics Driving Heterogeneous Nucleate Boiling and

Flashing”, Nuclear Engineering and Technology, Vol. 38, No. 2, Special Issue on ICAPP ’05, pp.

175-184, 2006

R.E. Henry and H.K. Fauske, “External Cooling of a Reactor Vessel under Severe Accident

Conditions”, Nucl. Eng. and Design, Vol. 139, pp. 31-43, 1993

J. Malet, M. Bessiron and C. Perrotin, “Modelling of Water Sump Evaporation in a CFD Code for

Nuclear Containment Studies”, Nucl. Eng. and Design, Vol. 214, pp. 1726-1735

J. Malet, O. Degrees du Lou and T. Gélain, Water evaporation over sump surface in nuclear

containment studies: CFD and LP codes validation on TOSQAN tests, submitted to Nucl. Eng. and

Design, 2012-2013

Hans-Josef Allelein, Ari Auvinen, Joanne Ball , Salih Güntay, Luis Enrique Herranz, Akihide Hidaka

, Alain V. Jones, Martin Kissane, Dana Powers , Gunter Weber , “State-of-the-Art Report on Nuclear

Aerosols”, NEA/CSNI/R(2009)5, p. 93, 2009 December


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3.1.4 P1-4 - Critical Heat Flux (CHF)

Description:

The maximum or critical heat flux represents the upper limit of nucleate boiling and marks the

termination of efficient cooling conditions on the surface. It is also known as dryout, departure from

nucleate boiling (DNB), burnout or boiling crisis. The mechanisms of dryout and DNB are different.

Dryout occurs under high quality (annular) flow conditions, when the thickness of the liquid film falls

below a critical limit and the surface of the heated structure remains in contact with steam. DNB

occurs for subcooled or low quality conditions, when the heat flux is so large that the resulting vapour

production hinders the liquid from wetting the surface. The mechanism of DNB is still a subject of

fundamental research, and various theories have been proposed. Burnout usually refers to the

negative effect of DNB on a surface. Boiling crisis is a general term for the DNB and dryout

mechanisms.

CHF is a phenomenon of interest for two aspects of containment related safety issues: in-vessel

retention under severe accident conditions and performance of condensers designed for the long term

decay heat removal.

In the case of a severe accident, a management strategy is to flood the reactor cavity, submerging

the reactor vessel. The concept is based on the idea that the external cooling will be able to prevent

the failure of the RPV and to arrest the downward relocation of the molten core. In general, it is

assumed that the primary system will be depressurized and the lower head will be fully submerged

before the core debris deposit on the inside of the lower head. With the lower head fully submerged,

the concept is based on the assumption that the outside surface will remain in nucleate boiling (at

100°C). This requires that the geometry of the vessel and reactor cavity provides a flow path for the

cooling water sufficiently wide to permit to the vapor to escape. Under these conditions, the lower

head failure can be avoided if CHF is not reached at any point on its surface. The mechanisms

leading to CHF are different at the various positions, and result in an angular dependence of the

limiting heat flux. Important parameters that affect CHF are surface wettability, wall thickness and

the strength of the convective flow (which depends on the geometry). Since wettability of the

material is important, coating of the lower head with porous material has been considered for

increasing the limiting heat flux.

References:

V.K. Dhir, “Boiling Heat Transfer”, Ann. Rev. Fluid Mech., Vol. 30, p. 380, 1998

T.G. Theofanous and T.N. Dinh, “High Heat Flux Boiling and Burnout as Microphysical Phenomena:

Mounting Evidence and Opportunities”, Multiphase Science and Technology, Vol. 18, No. 1, pp. 1-

26, 2006

A.E. Bergles, “What is the Real Mechanism of CHF in Pool Boiling?”, Proc. of Int. Conf. on Pool and

External Flow Boiling, 165-170, Santa-Barbara, USA, 2002

S.G. Kandlikar, “Insight into Mechanisms and Review of available Models for Critical Heat Flux

(CHF) in Pool Boiling”, 1st Int. Conf. on Heat Transfer, Fluid Dynamics and Thermodynamics

(HFFAT2002), Kruger National Park, South Africa, 2002 April 8-10

T.G. Theofanous, J.P. Tu, A.T. Dinh and T.N. Dinh, “The Boiling Crisis Phenomenon. Part I:

Nucleation and Nucleate Boiling Heat Transfer”, Experimental Thermal and Fluid Science, Vol. 26,

pp. 775–792, 2002

T.G. Theofanous, C. Liu, S. Additon, S. Angelini, O. Kymäläinen and T. Salmassi, “In-Vessel

Coolability and Retention of a Core Melt”, Nuclear Engineering and Design, Vol. 169, pp. 1-48, 1997


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3.1.5 P1-5 - Heat Conduction in Solids

Description:

Solid heat conduction is the process by which heat flows from a region of higher temperature to a

region of lower temperature within a solid medium or between different solid mediums in direct

physical contact. In conduction heat flow, the energy is transmitted by direct molecular

communication without appreciable displacement of the molecules. The conduction heat transfer

depends on the material geometry, conductivity and temperature gradient.

Conduction is the only mechanism by which heat can flow in opaque solids.

In a composite material, the temperature drop across an interface between materials may be

appreciable and this temperature drop is attributed to the thermal contact resistance at the interface.

This resistance is due principally to surface roughness effects.

Note: It includes both steady-state and transient conduction heat transfer.

References:

F.P. Incropera and D.P. DeWitt, “Fundamentals of Heat and Mass Transfer”, 2nd

Edition, John Wiley

& Sons, 1985


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3.1.6 P1-6 - Convection Heat Transfer (Natural and Forced)

Description:

Convection heat transfer is the energy transfer between a surface and a fluid moving over the

surface. The fluid motion can be due to “free” or “natural” convection whereby the fluid motion

results from density differences within the fluid arising from the temperature differences or fluid

components (i.e., lighter gases). The fluid motion can also be forced convection, whereby the fluid

motion is driven by an external force. Diffusion can also result in fluid motion, but is dominated by

natural or forced convection flows.

Convection heat transfer can also occur between a vapour and a liquid (film or droplets).

Note: Includes convection heat transfer from superheated steam to a surface.

References:

F.P. Incropera and D.P. DeWitt, “Fundamentals of Heat and Mass Transfer”, 2nd

Edition, John Wiley

& Sons, 1985


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3.1.7 P1-7 - Thermal Diffusion in Fluids (No Experiments)

Description:

The thermal diffusivity of a fluid is given by the thermal conductivity (k) divided by the

volumetric heat capacity (ρCp). It relates to the rate of heat transfer due to a temperature gradient in

the fluid. Heat transfer coefficients are usually computed using correlations which require thermal

conductivity data. The thermal conductivities of gases generally increase with increasing

temperature, whereas the thermal conductivities of most liquids decrease with increasing temperature.

For mixtures, thermal conductivity and thermal diffusivity can be estimated by the Wilke's

approximate method, which is analogous to methods applied for estimating the dynamic viscosity, µ,

of a fluid.

Experiments to measure the thermal diffusivity in gases or liquids are described in the referenced

MLM reports. These are fluid properties and additional experiments will not be requested. The

theory of thermal diffusion in fluids is well established and documented in the open literature. Thus,

no experimental results are needed.

Generally, CFD tools are validated on academic problems of heat transfer: natural convection in

a differentially heated square cavity with low Rayleigh number (De Vahl Davis, 1983) or laminar

fully developed heated channel flow in which self-similar solutions are available (Cebeci, 2002).

References:

R.B. Bird, W.E. Stewart and E.N. Lightfoot, “Transport phenomena”, Wiley, 1965

T. Cebeci, “Convective heat transfer”, 2nd

revised Edition, Springer, 2002

G. De Vahl Davis, “Natural convection of air in a square cavity: A bench mark numerical solution”,

Int. Journal for Numerical Methods in Fluids, pp. 249–264, 1983 May/June

J.I. Lin, “Thermodynamics of thermal diffusion - Thermal diffusion in Liquids”, Report MLM-36144,

1988

W.L. Taylor, “Thermodynamics of thermal diffusion - Thermal diffusion in Gases”, Report MLM-

3614, 1988

C.R. Wilke, “A viscosity equation for gas mixtures”, The Journal of Chemical Physics, Vol. 18, 4,

1950 April

Prepared by: E. Studer (CEA)

4 The publication, MLM-3614, contains two parts, one on diffusion in liquids and the other on diffusion in

gases, is publically available from the DOE/OSTI website (http://www.osti.gov/bridge)

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3.1.8 P1-8 - Radiation Heat Transfer (No Experiments)

Description:

Thermal radiation is a form of electromagnetic radiation, which is detected as heat or light, and is

generally composed of infrared and/or visible radiation. The intensity of thermal radiation heat

transfer between two bodies is proportional to the difference between the fourth powers of the

absolute temperatures. Therefore, the importance of radiation increases with the temperature levels.

In addition, no medium needs to be present between the two bodies for radiant exchange to

occur. If a medium exists between radiating surfaces, it can interact with radiative heat transfer, and

is called “participating” medium. In general a medium absorbs, scatters and emits energy. Its

capability to attenuate the radiation depends on the sum of the absorption and scattering coefficients,

which is called extinction coefficient. The optical thickness is defined as the integrated value of the

extinction coefficient over the thickness of the medium. If the optical thickness is much less than

unity, the medium is practically transparent to radiation, otherwise is considered opaque and must be

considered when calculating radiative heat transfer between two bodies. In general, scattering in

gaseous media is negligible, but can become important when dense clouds of particles (aerosols,

drops, etc.) are involved. Gases with symmetric diatomic molecules (such as hydrogen, oxygen and

nitrogen) are transparent to infrared radiation and do not need to be considered. Heat transfer is

mostly affected by water vapour, because of its strong emission bands. Additionally, CO and CO2

(and other gaseous products produced by core-concrete interaction during a severe accident) must be

considered in the analyses.

Finally, aerosols and dispersed core debris during a direct containment heating event may also

contribute to radiative heat transfer. Due to the large dimensions of the containment, the optical

thickness is always large, and the fluid is always a participating medium. Condensate liquid films

deposited on the walls have also to be considered, because a thin layer increases the emissivity of

concrete and steel structures. In principle, radiative heat transfer interacts with droplets (e.g., during

spray injection) as well, but this is usually not considered.

Overall, the heat transfer depends on the temperature of the surfaces and of the fluid, the

geometric arrangement of the surfaces, the surface emissivity and the properties of the intervening

media. Since radiative heat transfer occurs simultaneously with conduction and convection, it must

be considered whenever it cannot be proved to be negligible. For instance, radiant emission and

absorption can affect the heat transfer in a convection boundary layer.

In containment analysis, two forms of radiative heat transfer are considered:

Radiation between structures and media (vapour/gas/aerosols/debris). For most cases, radiation

represents a correction to convective heat transfer, and it can be added to it. In the case of direct

containment heating, radiation plays a major role where the core debris produced by the high

pressure melt ejection transfers heat to the air-steam atmosphere, and could cause a strong

pressurisation of the containment, and possibly even its failure. Also for evaluating the

performance of catalytic foils for hydrogen recombination, radiative heat transfer has to be

considered.

Radiation between the surfaces of the structures. The radiative heat exchange between the various

surfaces depends on the matrix of the view factors

Radiation between a liquid film and surrounding structures has also been considered for

evaluating the external cooling of the steel shell of the AP1000.

For the conditions relevant for containment thermal-hydraulics (range of pressures, gas mixtures

of steam, air and other gases, presence of liquid droplets and other dispersed phases), radiation heat

transfer is practically always associated with convective heat transfer from the structures and

interfacial heat and mass transfer between gas and particles/droplets. Moreover, the emissivity of the

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metallic surfaces is often difficult to characterise, as it varies with the age of the structure (oxidation,

wear, etc.). The experimental validation of the models implemented in the codes would therefore be

questionable, because the heat transfer modes are tightly coupled. On the other hand, radiation heat

transfer is governed by a nearly exact transport equation, where only the optical properties of the

steam, gas mixtures, and aerosols employ empirical or semi-empirical models. These considerations

suggest that the validation of the models could be better based on the comparison with benchmark

solutions, obtained by exact, although computationally very expensive methods (e.g., Montecarlo

simulations). Reference solutions could be obtained using hybrid methods, which combine the

performance of the two more practical (and popular) Discrete Ordinates (S-N) and Spherical

Armonics (P-N) methods, which cannot be used alone for the entire range of optical thicknesses and

properties of the medium (Maathangi (2011)).

References:

R. Siegel and J.R. Howell, “Thermal Radiation Heat Transfer”, 2nd

Edition, Hemisphere Publishing

Corporation, pp. 1-2, pp. 424-427, p. 619, p. 697, 1981

D.W. Condiff, D.H. Cho and S.H. Chan, “Heat Radiation Through Steam In Direct Containment

Heating”, American Nuclear Society and Atomic Industrial Forum joint meeting, Paper 32,

Washington, DC, USA, OSTI ID: 6889970, 1986 November

P. Royl, G. Necker and J.R. Travis, “GASFLOW Simulation of Hydrogen Recombination with

Radiation Transport from Catalytic Foils in the Recombiner Foil Test HDR E11.8.1”, Jahrestagung

Kerntechnik, Dresden, Germany, 2001 May 15-17

J. Woodcock et al., “WGOTHIC Application to AP600 and AP1000”, WCAP-15862, pp. 3-16 to 3-

18, 2004 March

Maathangi Sankar, “A Hybrid Discrete Ordinates - Spherical Harmonics Method for Solution of the

Radiative Transfer Equation in Multi-Dimensional Participating Media”, Master Thesis, Ohio State

University, 2011


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3.1.9 P1-9 - Condensation on Surfaces

Description:

Film and dropwise condensation occurs when a vapor contacts a cool surface. This type of

phenomenon is a very efficient means of heat transfer, and represents an important heat sink. The

vapor is cooled to its saturation temperature, releasing both sensible and latent heat to the surface.

The condensate forms on the surface as either a liquid “film” that covers the entire condensing surface

or as individual “droplets”.

Analytical work on condensation was pioneered by Nüsselt, and first reported in 1918, with the

formulation of the problem of pure vapor condensation. Experimentally, the effect of small amounts

of noncondensable species was identified during the 1920s. Sparrow in the 1960s analyzed the

phenomenon and demonstrated that the presence of noncondensable species is the dominating

contribution in the degradation of heat transfer coefficients. The analogy between the heat and mass

transfer phenomena concerning a chemical species emitted or absorbed at a surface was proposed by

Chilton and Colburn in 1934. This analogy is still widely used to model steam condensation on

surfaces in containment codes.

Empirical correlations like those of Uchida and Tagami have been extensively applied for large

scale containment analysis. Since the 1990s, various databases have been produced on condensation

phenomena in presence of noncondensable species like those of Anderson, Debhi, Huhtiniemi, and

recently, separate effect test facilities (CONAN and COPAIN) have provided valuable experimental

results to validate CFD codes.

References:

W. Nüsselt, “Die oberflachenkondensation des wasserdampfes”, Z. Ver. Deutsch. Ing., Vol. 60, pp.

541–569, 1916

T.H. Chilton and A.P. Colburn, “Evaporation of water into a laminar stream of air and superheated

steam”, Ind. Eng. Chem., Vol. 26, pp. 373–380, 1934

R. Leontiev, “Théorie des échanges de chaleur et de masse”, MIR editors, Moscow, 1979

H. Uchida, A. Oyama and Y. Togo, “Evaluation of post-incident cooling system of LWR’s”, In Proc.

Int. Conf. Peaceful Uses of Atomic Energy, Vol. 13, pp. 93–102, 1965

A.A. Dehbi, M. Golay and M.S. Kazimi, “Condensation experiments in steam-air and steam-air-

helium mixtures under turbulent natural convection”, In Proceeding of National Heat Transfer

Conference , pp. 19–28, Minneapolis, USA, 1991

I.K. Huhtiniemi and M.L. Corradini, “Condensation in the presence of noncondensable gases”,

Nuclear Engineering and Design, Vol. 141, pp. 429–446, 1993

M.H. Anderson, L.E. Herranz and M.L. Corradini, “Experimental analysis of heat transfer within the

AP600 containment under postulated accident conditions”, Nuclear Engineering and Design, Vol.

185, pp. 153–172, 1998

X. Cheng, “Experimental data base for containment thermalhydraulic analysis”, Nuclear Engineering

and Design, Vol. 204, 1-3, pp. 267-284, 2001 February

W. Ambrosini, N. Forgione and F. Oriolo, “Experimental and CFD analysis on condensation heat

transfer in a square cross section channel”, Int. Proceeding of the NURETH 11 Conference, Avignon,

France, 2005 October 2-6

M. Bucci, “Experimental and computational analysis of condensation phenomena for the Thermal-

hydraulic analysis of LWRs containments”, PhD thesis, University of Pisa, 2009.


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3.1.10 P1-10 - Pool Surface Evaporation and Condensation

Description:

Evaporation and condensation, and the associated transfers of latent heat at the liquid pool

surface, represent two opposite phase change phenomena. The free surface of a liquid pool is under a

total pressure, which is the sum of the partial pressures of the noncondensable air and water vapor in

the atmosphere.

The meaning of evaporation is limited to liquid evaporation only at the free surface of a pool that

is, without vapor bubble formation in and release from the liquid pool. Similarly, only condensation

at the free surface of a pool is considered here, without distributed precipitation or fog formation

within the bulk atmosphere, or condensation on other surfaces.

When evaporation occurs at a free surface of a liquid, it is usually called free surface

evaporation. The driving force is the density difference of water vapor between the gas mixture just

above the water surface and the ambient surroundings. The conditions, under which water

evaporation occurs, may be categorized according to the flow regime of the system (laminar/turbulent

conditions in free or forced convection), and correlations that describe evaporation can be either

empirical or based on heat and mass transfer analogy. The latter are more general and not restricted

by the experimental conditions. However, there are two possible source of errors related to the fact

that the mass transfer coefficient is directly proportional to Sherwood number and diffusion

coefficients. These correlations are also based on stagnant water pool with uniform temperature.

Convection motion due to temperature gradient will probably affect the surface temperature.

Condensation of steam/noncondensable gas mixture on a free water surface operates if the water

pool surface temperature is below the saturation temperature. The phenomenon is in a certain sense

equivalent to wall condensation as long as convection is not occurring in the water pool.

References:

M.T. Pauken, “An experimental investigation of combined turbulent free and forced evaporation”,

Experimental thermal and fluid science, Vol. 18, pp. 334-340, 1999

J. Malet et al., “Modeling of water sump evaporation in a CFD code for nuclear containment studies”,

Nuclear Engineering and Design, Vol. 241, 5, pp. 1736-1745, 2011

J. Malet, O. Degrees du Lou and T. Gélain, “Water evaporation over sump surface in nuclear

containment studies: CFD and LP codes validation on TOSQAN tests”, submitted to Nucl. Eng. and

Design, 2012-2013


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3.1.11 P1-11 - Heat Removal by Dousing

Description:

Heat can be removed from a hot air steam mixture, in a completely or partially closed room, by a

cooler dousing water spray. During the flight in the air steam atmosphere, sensible heat transfer and

condensation heats up the water droplets. Mass transfer by steam condensation on a droplet results in

some increase of the droplet size. If a droplet traverses a hot and dry atmosphere some water may

evaporate from the droplet surface, therby transferring heat to the droplet through its latent heat. The

net effect of dousing is to reduce the containment pressure and temperature.

References:

J. Malet, F. Dumay, E. Porcheron, P. Lemaitre and J. Vendel, TOSQAN spray benchmark –

TOSQAN test 101: Spray activation in air-steam mixture - Code-experiment comparison report,

Rapport IRSN/DSU/SERAC/LEMAC/05-07, 2005

J. Malet, E. Porcheron, J. Vendel, L. Blumenfeld and I. Tkatschenko, SARNET spray benchmark:

TOSQAN and MISTRA Specification report Rev. 1, Rapport IRSN/DSU/SERAC/LEMAC/06-11,

2006

J. Malet and P. Métier, SARNET spray benchmark: thermalhydraulic part, TOSQAN 101, Code-

experiment comparison report, IRSN Technical Report DSU/SERAC/LEMAC/07-03, 2007

E. Porcheron, P. Lemaitre, A. Nuboer, V. Rochas and J. Vendel, Experimental investigation in the

TOSQAN facility of heat and mass transfers in a spray for containment application, Nuclear

Engineering and Design, Vol. 237, pp. 1862-1871, 2007

P. Lemaitre and E. Porcheron, Analysis of heat and mass transfers in two-phase flow by coupling

optical diagnostic techniques, Exp. Fluids, Vol. 45, pp. 187–201, 2008

P. Lemaitre, E. Porcheron and A. Nuboer, “Study of Heat Transfer and Mass Transfer in a Spray for

Containment Application: Analysis of the Influence of Spray Temperature at the Injection Point”,

Nuclear Technology, Vol. 175, pp. 553-571, 2011

P. Lemaitre and E. Porcheron, “Study of Heat and Mass Transfers in a Spray for Containment

Application: Analysis of the Influence of the Spray Mass Flow Rate”, Nuclear Engineering and

Design, Vol. 239, pp. 541–550, 2009

J. Malet, L. Blumenfeld, S. Arndt, M. Babic, A. Bentaib, F. Dabbene, P. Kostka, S. Mimouni, M.

Movahed, S. Paci, Z. Parduba, J. Travis and E. Urbonavicius, “Sprays in Containment: Final Results

of the SARNET Spray Benchmark”, Nuclear Engineering and Design, Vol. 241, pp. 2162-2171, 2011

Prepared by: J. Malet (IRSN)

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3.1.12 P1-12 - Liquid Re-Entrainment (Resuspension)

Description:

Liquid re-entrainment (resuspension) refers to the mechanism whereby the tops of surface waves

in a liquid phase are stripped away by the flowing air/vapour mixture and then dispersed in a droplet

phase. This phenomenon is highly dependent on the velocity of the gas passing over the water

film/surface and the depth of the water pool.

Note: This phenomenon is primary of interest only for multi-unit CANDU stations. The reason is as

follows:

Re entrainment may occur in the early stages of a LOCA accident. A portion of the pooled liquid

discharge may be re entrained as it falls through the fuelling machine hatchway by the high

velocity flows present early in the accident. A multi unit station has the geometry for such a

situation. The fuelling machine hatch is the large opening in the floor of a Bruce/Darlington type

containment reactor vault and connects to the fuelling machine duct, which provides the long

flowpath to the vacuum building via the pressure relief duct.

It is this geometry in combination with the vacuum induced high velocities and the collection of

water on the floor of this pathway, which makes this phenomenon possible for multi unit stations.

The velocities in the fuelling machine vault could reach velocities much greater than 20 m/s if the

primary heat transport system break occurred in this room. At these velocities it is possible to lift

water droplets from a pool or strip a water film from a wall surface and entrain them in the flow.

To date, no suitable experimental data has been identified that can validate this phenomenon

under conditions of interest for CANDU LOCA analysis.

References:

M. Ishii and K. Mishima, “Droplet Entrainment Correlation in Annular Two Phase Flow”, Int. Journal

of Mass Heat Transfer, Vol. 32, No. 10, 1835 1846, 1989

S. Sugawara, “Droplet Deposition and Entrainment Modelling Based on the Three Fluid Model”,

Nuclear Engineering and Design, Vol. 122, 67 84, 1990

S. Sugawara and Y. Miyamoto, “FIDAS: Detailed Subchannel Analysis Code Based on the Three

Fluid and Three Field Model”, Nuclear Engineering and Design, Vol. 120, 147 161, 1990


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3.1.13 P1-13 - Direct Contact Condensation

Description:

In some nuclear reactor designs (BWR, EPR, etc.), safety systems can lead to the discharge of

steam from the reactor cooling systems to large reservoir of sub-cooled water. This steam makes

direct contact with the sub-cooled water and is condensed. Pressure oscillations may also be induced.

This phenomenon can be affected by the presence of noncondensable gas like hydrogen and is also

linked to pool scrubbing of fission products.

Youn et al. (2003) have reported a condensation regime map for low steam mass flux function of

steam mass flux and pool temperature. Six different regimes have been identified including chugging

(bubble forms and collapse), oscillating condensation and stable condensation. Norman et al. (2010)

have also conducted experiments on this phenomenon with a particular care of scaled-down strategy

and experiments show that air addition can lead in certain conditions to thermal stratification of the

water pool.

References:

D.H. Youn et al., “The direct contact condensation of steam in a pool at low mass flux”, Journal of

Nuclear Science and Technology, Vol. 40, 10, pp. 881-885, 2003 October

T.L. Norman et al., “Jet-plume condensation of steam-air mixtures in sub-cooled water, Part 1:

Experiments”, Nuclear Engineering and Design, Vol. 240, pp. 524-532, 2010


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3.1.14 P1-14 - Momentum Induced Mixing in Gases

Description:

Gas/vapour can be mixed by momentum induced motion caused by a vapour/gas jet. Two types

of mixing are be considered: mixing of the injected fluid into an ambient atmosphere, and mixing of a

stratified ambient atmosphere through some sort of forced convection.

Injection of fluid: A jet is formed by fluid injection into a quiescent or relatively quiescent

atmosphere. The relatively high velocity gradients present at the jet boundary, relative to a

plume, result in enhanced gas mixing by convective eddies and entrainment of the ambient

atmosphere into the expanding jet. Moreover, depending on the aspect ratio of the enclosure

and the strength of the jet, the wall jet resulting from the impact with the structures can

produce very large recirculation patterns, with cells that in an open geometry can extend over

the entire fluid domain. In the case of inclined or downwards oriented injections, intense flow

circulation zones can exist below the injection elevation as well.

In the case of a buoyant jet (which is nearly always the case in containment applications), the

momentum dominated zone will extend to a certain distance from the source. At large

distance, buoyancy effect will control the flow, and at intermediate distances both momentum

and buoyancy will be important. Various criteria have been proposed for defining the

boundaries between the three jet zones for vertical, upward jets. For large containments (such

as for PWR), various studies arrived at the conclusion that due to the large apertures, tall free

spaces, and interaction of the break flow with the structures, the flow originating from the

break becomes buoyancy dominated before reaching the top of the dome soon after the start

of the blow-down. Under these conditions, stratification is likely to build up (see Issue 1.01).

For more compact containments, the scenario is less clear. For other jet orientations, and

especially for negatively buoyant jets, the region of momentum controlled flow is more

difficult to identify, and the penetration depth (fountain region) depends on the competing

effects of inertia and buoyancy.

Another effect of large-motions produced by jets is the enhancement of heat and mass transfer

between the fluid and the wall.

Mixing in a stratified volume: an important aspect of containment thermal-hydraulics is the

stability of stratification under the effect of jets and plumes. For instance, the layering of

hydrogen produced in a phase of the accident could be destroyed by the injection of pure

steam from the break in a later phase of the accident. Depending on the strength of the jet and

of the density differences between the fluid layers, mixing will occur as a fast process

controlled by the penetration of the jet through the hydrogen-rich region and dilution, or as a

slow erosion process controlled by buoyancy-dominate mixing at the density interface.

The mixing can also be driven by the local acceleration of fluid, e.g., from a higher pressure

source or from a fan. High momentum jets can be produced by leakages and internal venting

between compartments (e.g., due to opening of vacuum breakers in BWRs or rupture foils in

the EPR). In certain advanced BWR designs (e.g., ESBWR), fans are provided to return the

non-condensable gases to the Drywell, with the aim to reduce the pressure increase in the

Wetwell, which controls the pressure of the system.

A special case of momentum induced mixing is that produced by the activation of the spray

(see Issue 1.28).

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References:

Gebhart, B., Yaluria, Y., Mahajan, R., Sammakia, B.,” Buoyancy-Induced Flows and Transport”,

Hemisphere Publishing Corp., p. 677, 1988.

Papanicolau, P.N. and List, E.J., “Investigations of round vertical turbulent buoyant jets”, J. Fluid

Mech., Vol. 195, pp. 341-391

Peterson, P.F., “Scaling and Analysis of Mixing in Large Stratified Volumes”, Int. J. Heat Mass

Transfer, Vol. 37, Supplement 1, pp. 97-106, 1994 March

Kuhn, S-Z., Kang, H.K., and Peterson, P.F., “Study of Mixing and Augmentation of Natural

Convection Heat Transfer by a Forced Jet in a Large Enclosure”, Journal of Heat Transfer, Vol. 124,

pp. 660-666, 2002.

Friedman, P.D. and Katz, J. “Rise Height for Negatively Buoyant Fountains and Depth of Penetration

for Negatively Buoyant Jets Impinging an Interface”, J. Fluids Engineering, Vol. 122, 779-782, 2000

Deri, E., Cariteau, B., Abdo, D., “Air fountains in the erosion of gaseous stratifications”, Int. J. Heat

Fluid Flow, Vol. 31, pp.935-941, 2010.


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3.1.15 P1-15 - Buoyancy Induced Mixing in Gases

Description:

Gas/vapour can be mixed by buoyancy induced motion due to pressure gradients created by local

gas density differences in a gravitational field. These density differences are due to composition

differences and/or temperature differences, induced by local mass and/or heat transport processes

(e.g., gas injection, convection and condensation heat removal at surfaces). As for the momentum

controlled mixing, two cases can be considered: mixing of the buoyant fluid within an ambient

atmosphere, and mixing of a stratified ambient atmosphere due to buoyancy sources.

Buoyant fluid injection: At some distance from a positively buoyant injection, the flow is

buoyancy controlled, and the entrainment of the fluid from the ambient dilutes the injected

fluid. For fully developed plumes, the gas distribution within the enclosure can be

conceptualized by imagining the lighter fluid floating toward the ceiling and forming a layer

that becomes thicker over time, though instead of a distinct interface, there would be

progressive gradient in the concentration where the lighter components are more concentrated

at the top. This process can also be described by top-down propagation of a stratification

“front” (defined at a specific concentration). For long durations of the injection eventually

the front will reach the bottom of the enclosure. For a negatively buoyant injection, a bottom-

up propagation of the stratification front will occur.

Mixing can be accelerated by re-evaporation of liquid films running down the structures. An

important phenomenon affecting the relocation of hydrogen is the variation of the effect of

condensation on the buoyancy of a steam/air/hydrogen mixture with its composition. If the

concentration of air is small, condensation will produce a lighter mixture than tends to rise.

However, if both steam and air are in large concentrations, condensation will produce a

mixture heavier than the ambient, and air will provoke the hydrogen to propagate towards the

bottom of the vessel. This phenomenon can thus result in higher hydrogen concentrations in

the lower part of certain compartments.

Mixing in an ambient atmosphere with pre-existing stratification: The stratification can

be destroyed by the injection of buoyant fluid due to the acceleration of the fluid, which

produces perturbation and deformation of the density interface. Depending on the relative

differences between the densities of the injection and densities of the ambient layers, as well

as the density gradients in the ambient, various modes of interaction of the plume with the

density interface will occur. These all result in a slow erosion of the stratified environment.

Stratification can also be destroyed by thermal plumes originated by heat sources, such as hot

surfaces, or descending, negatively buoyant plumes produced by condensation

References:

Gebhart, B., Yaluria, Y., Mahajan, R. and Sammakia, B.,” Buoyancy-Induced Flows and Transport”,

Hemisphere Publishing Corp., p. 786-788, 1988

Mott, R.W. and Woods, A.W., “On the mixing of a confined stratified fluid by a turbulent buoyant

plume”, J. Fluid Mech., Vol. 623, pp. 149-165, 2009


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3.1.16 P1-16 - Pressure Wave Propagation in Water

Description:

Rupture of a CANDU fuel channel at operating conditions (10 MPa, 300C) will generate a

pressure pulse in the surrounding heavy water moderator (0 MPa, 80C). Due to the relatively

incompressibility of the heavy water moderator, the shock and the pressure wave are co-incident. The

high pressure/temperature escaping fluid would generate a high velocity pressure wave in the

moderator that would interact with the surrounding fuel channels. The wave could cause the colder,

thin-walled calandria tube to collapse onto the concentric, hot pressurised pressure tube. The pressure

tube would then experience impact forces and asymmetrical thermal conditions from the calandria

tube. This could induce failure in that pressure tube which, in turn, amplify the original pressure

pulse. A single rupturing fuel channel would generate a pressure pulse that could interact with

approximately 20 of its nearest fuel channels which could then induce a cascading fuel channel

rupture sequence. Depending on the physical location of the bursting channel, i.e., in the center

location or near the containment vessel component, the containment vessel could experience high

impact forces.

Blowdown of a BWR, steam-gas mixture from its reactor pressure vessel or drywell, into its

pressure suppression pool, involve large pressure differences between the two volumes along with

rapid steam condensation and mixing of the water in the suppression pool. This can generate large

pressure waves that will impact against the pipes and walls of the suppression pool.

References:

Leitch, B.W., Shewfelt, R.S.W. and Godin, D.P., “Two-phase Fluid/structure interactions in a

bursting tube”, AECL Report AECL RC-1711, COG Report COG-96-486, 1997

Shewfelt, R.S.W., Leitch, B.W. and Godin, D.P., “Guillotine failure of fixed-end pipes, pressurised

with hot water”, AECL Report AECL-10948, 1994

Shewfelt, R.S.W. and Godin, D.P, “Small-scale burst tests in air and water”, AECL Report RC-1454,

COG Report COG-95-356, 1995

Group of Experts of the NEA/CSNI, “Pressure Suppression System Containments – A State-of-the-

Art Report”, CSNI Report 126, 1986 October

Prepared by: B.W. Leitch (AECL) and Y.S. Chin (AECL)

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3.1.17 P1-17 - Mixing in Water Pools

Description:

The temperature distribution in a body of water where energy is deposited depends on the mixing

induced by the flow produced by mass, momentum and energy sources. Three main issues are

associated with thermal mixing include the formation of hot fluid layers at the pool free surface, the

uptake of warm water for emergency cooling with reduced heat removal capability, and the local

subcooling of the fluid adjacent to a heated surface:

Thermal stratification in the suppression pool of a BWR leads to high temperatures at the pool

surface, which in turn leads to increased partial pressure of steam in the gas space and thus to

higher containment pressure. Immediately after the first seconds of a LOCA, the vent flow is

composed essentially of steam, which condenses within a certain length. The steam jet

consists of the vapour core and the two-phase mixing region. Depending on the mass flux

(which decreases with time), the liquid flow originating from the condensation of the steam

jet produces a liquid jet or a plume. For high water temperatures, a two-phase plume will rise

to the surface and partial steam bypass can occur. The mixing produced by these flow

structures depends on the initial momentum, the geometry of the vent, the penetration length

of the steam jet, the entrainment of liquid into the plume/jet, and the interaction of the flow

with the pool structures. For horizontal injections, the initially horizontal jet will rise to the

surface as a wall jet after impinging on the side walls. Pressure oscillations will also affect

the prevailing flows. Similar phenomena are also relevant for the discharge of steam through

SRV in the suppression pool and in the IRWST of advanced reactors trough vent or

depressurisation lines. For low steam mass fluxes, the capability of the plume to steer the

water pool is strongly affected by the presence of non-condensable gases. Due to the large

entrainment produced by the two-phase plume, less than 1% non-condensable mass fraction

produces efficient mixing above the vent elevation. The hydrodynamics of the two-phase

plume is extremely complex, including various forces acting on the bubbles (drag, lift, virtual

mass, dispersion, etc.), bubble break-up and calescence, two-way coupling between bubbles

and liquid, two-phase turbulence, 3-D oscillations of the plume, etc.

In some designs, where water from pools is used for emergency cooling of the containment,

stratification is beneficial to ensure that the water taken from the bottom of the pool remains

cold and thus permits efficient cooling. In the case of internal condenser fed by water

extracted from the bottom of a pool, the flow returning to the pool is in single phase for a long

time. During this period, the mixing is controlled by the circulation produced by a liquid

jet/plume. With increasing time, the flow in the return line becomes two-phase (for some

time because of flashing) and similar phenomena as at point 1 prevail. During the single-

phase period, turbulence plays a very important role in the propagation of the thermal front.

Boiling heat transfer and CHF are generally affected by the subcooling of the fluid in the

vicinity of the heated wall. In open pools (inside or outside the containment) where energy is

deposited by heat transfer from a thermal structure, the bulk subcooling is mostly determined

by the gravitational head. Convective mixing induced by the boiling process, however, tends

to homogenize the temperature of the pool, and, consequently to reduce the local subcooling.

This mixing affects boiling heat transfer on the secondary side of immerged condensers, and

should be considered in the evaluation of the effectiveness of ex-vessel cooling because the

reduction of subcooling reduces the local value of the CHF.

In the case of spent fuel tank, the circulation in the water pool can also be produced by a pump.

References:

Gamble, R.E., Nguyen Thuy T., Shiralkar, B.S., Peterson P.F., Greif, R. and Tabata, H., “Pressure

Suppression Pool Mixing in Passive Advanced BWR Plants”, Nuclear Engineering and Design, Vol.

204, pp. 321–336, 2001

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Li, H., Kudinov, P. and Villanueva, W., “Modeling of Condensation, Stratification, and Mixing

Phenomena in a Pool of Water”, Nordic Nuclear Safety Research, Report NKS-225, ISBN976-87-

7893-295-2, 2010 December

Bestion, D., Anglart, H., Mahaffy, J., Lucas, D., Song, C.H., Scheuerer, M., Zigh, G., Andreani, M.,

Kasahara, F., Heitsch, M., Komen, E., Moretti, F., Morii, T., Mühlbauer, P., Smith, B.L. and

Watanabe, T., “Extension of CFD Codes Application to Two-Phase Flow Safety Problems - Phase 2”,

Report NEA/CSNI/R(2010)2, pp. 74-87, 2010 July

Andreani, M. and Coddington, P., “SBWR PCCS Vent Phenomena and Suppression Pool Mixing”, in

Proc. 7th Int. Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-7), Sept. 10-15,

Saratoga Springs, NY, USA, NUREG/CP-0142, American Nuclear Society, Vol. 2, pp. 1249-1271,

1995

T.L. Norman et al., “Jet-plume condensation of steam-air mixtures in sub-cooled water, Part 1:

Experiments”, Nuclear Engineering and Design, Vol. 240, pp. 524-532, 2010

Simiano, M., Zboray, R., de Cachard, F., Lakehal, D. and Yadigaroglu, G., “Comprehensive

Experimental Investigation of the Hydrodynamics of Large-scale, 3D, Oscillating Bubble Plumes”,

Int. J. Multiphase Flow, Vol. 32, pp. 1160-1181, 2006

Zoran V., Stosic, Z.V., Brettschuh, W., Stoll, “Boiling Water Reactor with Innovative Safety Concept:

the Generation III+ SWR-1000”, Nuclear Engineering and Design 238 (2008), 1863–1901, 2008


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3.1.18 P1-18 - Mass Diffusion in Vapour

Description:

It is the relative motion of species in a vapor/gas mixture due to the presence of concentration

gradients. The concentration gradient in the vapor/gas mixture provides the driving potential for

transport of that species. Mass diffusion will tend to reduce the concentration gradients to result in

uniform concentrations (well mixed) conditions. The diffusion mass transfer rate depends on the gas

component diffusion coefficients for the multi-component mixture.

The common way of estimating diffusion mass fluxes (J) in binary mixture is adopting the Fick's

law. In a binary mixture, the diffusion mass flux is proportional to the gradient of the selected species

(Y mass fraction) via a diffusion coefficient D between the two chemical species involved. In a

mixture of species i and j, the diffusion mass flux of i species is therefore given by:

Ji = -ρ Dij grad(Yi)

In multi-component mixtures, the simple solution based on effective binary diffusion (Dim)

approximation is usually adopted instead of a full multi-component diffusion model based on kinetic

theory of gases:

where X is the molar fraction.

The kinetic theory of gases can be used to derive formulas for the binary diffusion coefficients

Dij. Marrero et al. (1972) have also provided simple empirical relations taking into account pressure

and temperature effects.

The mass diffusion coefficients are used to compute the diffusion mass fluxes of gaseous species

in CFD codes. Generally, turbulent diffusion is dominant. Steam diffusion coefficient in gaseous

mixture is also usually used in the calculation of mass transfer coefficient in wall condensation

models based on heat and mass transfer analogy.

References:

Bird R.B., Stewart W.E. and Lightfoot E.N., “Transport phenomena”, Wiley, 1965

Marrero T.R. and Mason E.A., “Gaseous diffusion coefficients”, Journal Physical Chemistry

Reference Data, 1:3–118, 1972


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3.1.19 P1-19 - Laminar Flow (No Experiments)

Description:

Laminar flow occurs when a fluid flows in parallel layers without disruption of these layers. In

containment thermal-hydraulics there are only few situations where laminar flow occurs. Close to a

wall, there is a thin layer (called viscous sublayer) where the flow is laminar. In CFD computer code

models, this zone is usually not computed but modeled by the use of wall functions. Low Reynolds

turbulence models address also this zone but they are presently not used at reactor scale. Experiments

usually do not address this zone.

Another situation where laminar flow exists corresponds to the transition between convection

dominant zone and stratified zone (i.e., zone without velocity where temperature and/or gas

concentration gradients can exist). So, there is a buffer layer where the flow is laminar before the

flow stops. This transition from turbulent to laminar flow is usually a challenging modeling issue.

Large scale experiments performed in the PANDA and MISTRA facilities have addressed this

complicated situation with simplified boundary conditions. Also, in some dead-end zones of large

scale experiments, this transition has certainly occurred.

The theory of laminar flow is well established and documented in the open literature. Thus, no

experimental results are needed. CFD codes are generally checked against Hagen-Poiseuille law and

laminar round jet self similar solutions, given in Schlichting (2000). In the containment thermal

hydraulics, these two experiments PANDA ST7_1 and MISTRA LOWMA3 tests may have zones

where laminar flow occurs.

References:

G.K. Batchelor, “An introduction to fluid dynamics”, Cambridge University Press, 2000

P.S. Sutera et al., “The history of Poiseuille's law”, Annual Review of Fluid Mechanics, Vol. 25, pp.

1-19, 1993

H. Schlichting and K. Gersten, “Boundary layer theory”, 8th Edition, Springer, 2000


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3.1.20 P1-20 - Turbulent Flow

Description:

A turbulent flow is a fluid flow that includes rapid variations in the velocity and pressure in time

and space, and generally has stoichastic components. Turbulence involves eddy formation at many

different length scales, with the largest related to geometry and the smallest to viscosity. It influences

effective diffusion of heat, mass and momentum and the deposition and agglomeration of drops.

In containment thermal hydraulics, turbulence is important in many different phenomena like gas

mixing, wall condensation, droplet behavior, etc.

In LP codes, turbulence is inherently part of the correlations used to model the different

phenomena. In CFD models, turbulence is generally modeled via two transport equations for the

turbulent kinetic energy and its dissipation (k-e models). These two scales are required to compute

the turbulent viscosity. Fully turbulent flows can reasonably be predicted by this type of models.

Weakly turbulent or transition from turbulent to laminar flow, that can for example occur in gas

mixing with stratification, are more challenging and predictive capabilities have not yet been

confirmed. Another area of concern is the wall treatment, where wall boundary layers are modeled

with universal log-law profiles that are mainly valid for forced convection flows. Extensions to other

flow conditions with heat and mass transfer need careful investigations. Recently, low Reynolds

turbulence models have been used for benchmarking in wall condensation phenomena but these

models are presently not relevant to reactor scale analysis due to the requirement to have very fine

mesh close to the walls.

References:

Batchelor G.K., “An introduction to fluid dynamics”, Cambridge University Press, 2000

Wilcox, D.C., “Turbulence modeling for CFD”, DCW industries, 1998

Allelein, H.J. et al., “International Standard Problem ISP47 on containment thermalhydraulics - Final

report”, NEA/CSNI/R(2007)10


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3.1.21 P1-21 - Critical Flow (Choked Flow)

Description:

The discharge rate from a higher pressure reservoir or vessel to a lower pressure vessel or

reservoir is governed by such factors as flow area, pressure difference, wall friction, exit losses, state

of the fluid and upstream flow regime.

As the fluid flows to the break, there may be a physical point in the system at which a critical

pressure ratio - upstream to downstream, is reached. Any subsequent reduction to the downstream

pressure will not lead to an increase in the fluid velocity. When this happens, the flow is said to be

“critical”. The point where the flow becomes critical may be internal - at valves, orifices or other

obstructions to the flow, or it may be at the break plane itself. Pressure ratios below that required to

cause critical flow are called “sub-critical”.

Mass discharge rates are significantly affected by the state of the fluid (single phase liquid, a

two-phase mixture or single-phase steam). This occurs partly because of the large density differences

between the phases and partly because of the large difference in sonic velocities between the phases.

For the case of single-phase flow, an area change or some other obstruction in the flow path may

cause the fluid velocity to become equal to the local sound speed. At this point, information cannot

be transmitted upstream and the flow is said to be choked. When two phases are present, the picture

is not as clear. The discharged fluid can be single phase liquid, a two-phase mixture or single-phase

gas, and this has a larger effect on the mass discharge rates. This strong dependence on the state of

the fluid occurs partly because of the large density differences between the phases and partly because

of the large difference in sonic velocities between the phases.

The structure of the fluid flow near the break plane can also impact on the discharge rate as in the

case of stratified flow. In this case, the location of the break relative to the steam water interface will

determine whether liquid, vapour or both phases are discharged.

References:

I. Brittain et al., “Critical Flow Modelling in Nuclear Safety, A State of the Art Report”, NEA-OECD,

1982

E. Elias and G.S. Lellouche, “Two-Phase Critical Flow”, Int. J. Multiphase Flow, Vol. 20, Suppl. pp.

91-168, 1994

R.E. Henry and H.K Fauske, “The Two-Phase Critical Flow of One Component Mixtures in Nozzles,

Orifices, and Short Tubes”, J. Heat Transfer, 93, 179-187, 1971

R.E. Henry, “The Two-Phase Critical Discharge of Initially Saturated or Subcooled Water”, Nucl. Sci.

Engng., 41, p. 336, 1970

F.J. Moody, “Maximum Two-Phase Vessel Blowdown from Pipes”, Trans. ASME, J. Heat Transfer,

Vol. 87, pp. 285-295, 1966

F.J. Moody, “Maximum Flow Rate of a Single Component, Two Phase Mixture”, Journal of Heat

Transfer, Trans ASME, 87, No. 1, 1965 February

K.H. Ardron and R.A. Furness, “A Study of the Critical Flow Models Used in Reactor Blowdown

Analysis”, Nuclear Design and Engineering, Vol. 39, pp. 257-266, 1976

K.H. Ardron, “A Two-Fluid Model for Critical Vapour-Liquid Flow”, Int. Journal of Multiphase

Flow, Vol. 4, pp. 323-327, 1978

H.J. Richter, “Separated Two-Phase Flow Model: Application to Critical Two-Phase Flow”, Int.

Journal of Multiphase Flow, Vol. 9, pp. 511-530, 1983

F. Dobran, “Nonequilibrium Modelling of Two-Phase Critical Flows in Tubes”, Journal of Heat

Transfer, 109, pp. 731-738, 1987

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“The Marviken Full Scale Jet Impingement Tests”, Fourth Series, Test 10 Results, MXD-210, 1982

March

A.R. Edwards and T.P. O’Brian, “Studies of Phenomena Connected With the Depressurization of

Water Reactors”, J. of British Nuclear Society, 9, 1970 No. 2

P.J. Ingham, G.R. McGee and V.S. Krishnan, “LOCA Assessment Experiments in a Full-Elevation,

CANDU-Typical Test Facility”, Proc. of the 3rd

Int. Topical Meeting on Nuclear Power Plant Thermal

Hydraulics and Operations, Seoul, Korea, 1988 November. Also, published in Nuclear Engineering

and Design, Vol. 122, pp. 401-412, 1990


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3.1.22 P1-22 - Laminar/Turbulent Leakage Flow

Description:

Gas, vapour and aerosols can leak through the containment concrete walls. The leak paths can

be cracks through the wall thickness or gaps around constructed wall penetrations. Noble gases leak

more readily than molecular gases. The gaps may also be plugged by condensate or aerosol

deposition. The flow can creep through the smallest flow passages, even if there are only very low

pressure differences. The flow is laminar for very low Reynolds numbers (<100), it is transitional up

to a Reynolds number of 1000 and it is turbulent above.

Nuclear reactor containment concrete walls are generally a prestressed concrete structure, which

is reinforced with prestressed tendons and surface reinforcements. When the containment internal

pressure exceeds the sum of the pre-stress and the tensile stress of the concrete, cracking will occur.

In the initial phase of an accident, the containment pressurization is mainly due to steam build-up.

However, as the accident progress, other processes (hydrogen combustion, direct containment heating

and MCCI) can occur which can significantly increase the containment pressure. This will increase

cracking of the concrete containment and lead to increased leakage flows.

References:

Blejwas, T.E., “Containment Integrity Program Recent Results and Plans,” Proc. Int. Meeting on

LWR Severe Accident Evaluation, Vol. 2, Paper TS-10.7, Cambridge, MA, 1983 August 28 to

September 01

Rizkalla, S.H., S.H. Simmonds and J.G. MacGregor, “Prestressed Concrete Containment Model,”

Structural Engineering, 110:(4) 730-743, 1984 April

Baker, P.H., S. Sharples and I.C. Ward, “Air Flow Through Cracks,” Building and Environment,

22:(4) 293-304, 1987

Suzuki, T., K. Takiguchi and H. Hotta, “Leakage of Gas through Concrete Cracks,” Nuclear

Engineering and Design, 133:(1) 121-130, 1992 February

Lau, B.L., “Leakage of Pressurized Gases through Cracks in Reinforced Concrete Structures,” M.Sc.

Thesis, Dept. of Civil Engineering, University of Manitoba, 1982

Rizkalla, S.H., S.H. Simmonds and J.G. MacGregor, “Leakage Tests of Wall Segments of Reactor

Containments,” Structural Engineering Report No. 80, Dept. of Civil Engineering, University of

Alberta, Edmonton, 1979 October

Hindy, A. and A. Danay, “Assessing Leakage through Cracked Pressurized Reinforced Concrete

Containment Structures,” Trans. 11th SMIRT, Vol. H, Paper H08/4, Tokyo, 1991 August 18-23

Mivelaz P., Thèse no 1539, Ecole Polytechnique Fédérale de Lausanne; 1996

Riva P., L. Brusa, P. Contri and L. Imperato, “Prediction of air and steam leak rate through cracked

reinforced concrete panels”, Nuc. Eng. Des., 192 (1999), pp. 13–30

Gelain T., J. Vendel, “Research works on contamination transfers through cracked concrete walls,”

Nuclear Engineering and Design, Vol. 238(4), pp. 1159–1165, 2008

Gelain T., An original method to assess leakage through cracked reinforced concrete walls,

Engineering Structures, Vol. 38, 2012, pp. 11–20

Granger L. and P. Labbe et al., “A mock-up near Civaux nuclear power plant for containment

evaluation under severe accident—the CESA Project”, Proceedings of the FISA-97 Symposium on

EU Research on Severe Accidents, Luxembourg (1997), pp. 293-302, 1997

Greiner U. and W. Ramm, Air leakage characteristics in cracked concrete, Nucl Eng Des, 156 (1995)

Prepared by: Y.S. Chin (AECL) and T. Gelain (IRSN)

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3.1.23 P1-23 - Vent Clearing

Description:

During the blow-down associated with a LOCA in a BWR the drywell pressure increases, and

the water in the main vents is accelerated and flows into the suppression pool. Within seconds, the

vents would clear of liquid and air (or nitrogen in inerted containments), followed by a two-phase

mixture of gas, steam, and suspended water would flow into the suppression pool. This flow would

initially create a gas bubble at the downstream end of the vent, which would then cause level swell

(phenomenon P1-24) before eventually breaking through to the pool surface.

The bubble dynamics that would be involved include the initial acceleration of the liquid

surrounding the bubble, the level swell in the suppression pool, the steam condensation within the

bubbles, and finally the release of the gas and steam to the wetwell atmosphere after bubbles break

through the suppression pool surface. This sequence of events is referred to as the vent-clearing

transient.

The short-term peak drywell pressure (and drywell-wetwell pressure difference), is to a large

extent, controlled by the vent clearing time, i.e., the time required for the gas to penetrate to the far

end of the vents on the wetwell side. The vent clearing is influenced by the inertia of the liquid in the

vent path, vent hydraulic resistance and choking. Moreover, the drywell pressure is also controlled by

the pressure rise in the wetwell and the hydrostatic head of the water above the bubble (phenomenon

P1-24). Large dynamical loads in the suppression pool are associated with vent clearing

(phenomenon P1-24). The basic phenomena are the same for all designs, though the geometry,

submergence and orientation of the vents (vertical or horizontal), all play significant roles.

Additionally, for the MARK III containment, the three rows of vents at different elevation clear at

different times.

During the initial phase of the blow-down large vent flow rates produce high-velocity jets and

large condensation rates at the vent exit. In the period following the first seconds, the flow decreases,

and the steam jet breaks in bubbly-flow. When the flow is further reduced, chugging regime occurs

(see phenomenon P1-13). These processes are mostly oscillatory, and, especially in the case of

chugging large mechanical loads on the structure are produced. Several flow regime maps have been

proposed for characterising the various steam discharge modes. The main parameters are pool

temperature and steam mass flux. For small holes (such as one of a sparger), also the diameter of the

jet has to be considered.

Vent clearing is also relevant for depressurisation transients following the opening of SRVs.

Mechanical loads associated with chugging can be reduced by appropriate design of the sparger.

Vent clearing is also to be considered in the analysis of the long-term cooling of passive BWR

containments, where additional, low-submergence vent paths exist between the drywell and the

suppression pool.

Steam discharge in a pool is also of interest for the design of emergency core cooling system in

advanced designs (e.g., the IRWST of the AP1000) and evaluation of their behaviour under accident

conditions. Also for these cases, two issues are considered: mechanical loads and thermal mixing. To

some extent mechanical loads are coupled to fluid mixing, because the oscillatory behaviour of the jet

is affected by the presence of hot fluid spots. The effect of multiple discharging has also to be

considered, because the regime map is slightly different from that of the single nozzle case.

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References:

Karwat, H., Lewis, M.J., Mazzini, M. and Sandervaag, O., “Pressure suppression systems

Containments”, A State-of-the-Art Report by a Group of Experts of the NEA CSNI, CSNI Report

126, p. 10, 1986 October

Gamble, R.E., Nguyen, T.T., Shiralkar, B.S., Peterson, P.F., Greif, R. and Tabata, H., “Pressure

suppression pool mixing in passive advanced BWR plants”, Nucl. Eng. Design, Vol. 204, pp. 321-

336, 2001

Petrovic de With A., Calay, R.K. and de With. G., “Three-dimensional Condensation Regime

Diagram for Direct Contact Condensation of Steam Injected into Water”, Int. Journal of Heat and

Mass Transfer, Vol. 50, pp. 1762–1770, 2007


Kasahara, F., M. Heitsch, M., Komen, E., Moretti, F., T. Morii, T., Mühlbauer, P., Smith, B.L. and

Watanabe, T., “Extension of CFD Codes Application to Two-Phase Flow Safety Problems - Phase 2”,

Report NEA/CSNI/R(2010)2, pp. 74-87, 2010 July


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3.1.24 P1-24 - Pool Swell / Air Injection

Description:

During this early phase of a LOCA (vent clearing period, see phenomenon P1-23) a gas bubble

or bubbles form in the wetwell and raise to the pool surface. The pressure in the bubble depends on

the pressures in the drywell and gas space above the pool, the hydrostatic head of water above the

bubble and the inertia of the water that must be accelerated to make room for the bubble. As the

drywell pressure continues to rise, more gas is forced into the wetwell and the bubble continues to

grow, forcing the pool surface higher. The bubble starts to rise relative to the rising water above the

bubble due to buoyancy forces.

The liquid slug above the bubble thins as some of the water above the bubble moves laterally and

returns to the lower part of the pool. As the liquid slug rises, the gas space volume is reduced and the

gas space pressure increase opposes the lifting force and eventually decelerates the rising slug. As the

slug slows the bubble continues to rise and breaks through the slug, forming a froth region that rises

above the breakthrough level.

Equipment that is located between the initial pool surface and the maximum slug height will be

subjected to impact loads, followed by drag loads from the rising slug. In addition to the normal drag

load from a steady velocity field, there will be an additional load due to the accelerating fluid moving

past the equipment. Equipment in the froth region will experience impact and drag loads although

they will be lower than those in the slug regions because the mixture density of the froth is

substantially lower.

After the bubbles have cleared the pool surface and the vent flow becomes predominately steam,

the equipment in the slug and froth regions will be subjected to reverse drag loads as the water falls

back to the pool. These loads will typically be substantially smaller than those due to the rising water

because the fall back velocities are lower. Inertia, momentum transport, drag effects and gas

compression are the controlling physical effects in pool swell. During the pool swell period, the pool

walls, floor and ceiling will be subjected to increased loads due to the high bubble pressure, gas space

pressure and the hydrostatic pressure.

Four parameters are important for characterising hydrodynamics loads: maximum swell height,

maximum velocity of the rising water slug, maximum bubble pressure during the pool swell phase,

and maximum gas space pressure. The main phenomena are well understood and have been

addressed experimentally in the ISP17.

The continuous release of steam-gas mixtures from a low submergence vent produces a two-

phase plume in the pool, with the formation of a “spout” on the surface of the pool, which can reach a

height of the order of 10 cm. The hydrodynamic conditions in the surface breaking plume lead to

high void fractions in the spout with the formation of a dispersed droplet flow. The large interfacial

area results in very efficient condensation of the steam which could not be condensed in the pool due

to the short travel time.

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References:

Toshiba Corporation “Post LOCA Suppression Pool Swell Analysis for ABWR Containment

Design”, U7-C-STP-NRC-090142, Attachment 4, UTLR-0005-NP Rev.0, p. 3, September 2009

(available on the Web)

Notafrancesco, A., Esmaili, H., Lee, B. and Tills, J.L., “Application of the MELCOR Code to Design

Basis BWR Containment Analysis”, USNRC Report RES/FSTB 2011-01, p. 2, 2011 May

Widener, S. K., “Analytical simulation of Boiling Water Reactor Pressure Suppression Pool Swell”,

Nuclear Technology, Vol. 72, pp. 34-38, 1986

Marklund, J.E., “Preliminary Data Comparison Report For ISP17 - An international containment

standard problem based on the Marviken full scale experiment Blowdown No 18”, Studsvik Report

SD-84/43, NR-84/423, 1984 June

Andreani M. and Tokuhiro, A “Condensation in the Spout Region of a Gas-Vapour Plume Rising in a

Subcooled Water Pool”, in Proc. of the 2nd

Int. Conf. on Multiphase Flow `95 Kyoto, April 3-7,

Kyoto, Japan Society of Multiphase Flow, Vol. 2, pp. PC2-17 to PC2-24, 1995


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3.1.25 P1-25 - Interfacial Drag (No Experiments)

Description:

Interfacial drag occurs on the fluid interface between two fluid phases. It depends on relative

velocity between the two phases, density of the continuous phase, the volume fraction of the dispersed

phase, the characteristic (equivalent) diameter of the dispersed phase and the interfacial drag

coefficient (i.e., friction factor). The friction factor depends on the shape of the interface, and

therefore on the flow regime. Three main flow regimes are of interest for containment thermal-

hydraulics: bubbly flow (dispersed bubbles in liquid); dispersed droplet flow (droplets in gas), and

separated flow regime. Other complex flow regimes originate from the interaction of the corium melt

jet with the water in the reactor pressure vessel cavity, which are not considered here.

Bubbly flow is mostly of interest for the venting of gas-steam mixtures in liquid pools, where the

interfacial drag controls the residence time of the gas and consequently affects the interfacial heat and

mass transfer. For sufficiently low flow rate conditions, the bubbles are produced by the

fragmentation of the primary bubble. During blow-down, however, the primary bubble expands and

produces pool swelling before any fragmentation occurs. Under these conditions, the bubbles produce

a separated flow. In principle, a comprehensive theory of bubbly flow and interfacial drag should

include these two regimes and transitions, and therefore address separately small and large bubbles.

In this section, however, only interfacial drag for small bubbles is considered, whereas the dynamics

of large bubble is described in the section addressing vent clearing. Bubbly flow is also of interest for

boiling processes (relevant, e.g., for pool condensers and ex-vessel cooling), where the interfacial

drag controls phase velocities and void fractions, and consequently the natural circulation flow

associated with boiling.

Dispersed droplet flow is produced by the operation of spray systems or by entrainment from

liquid films. Although the interfacial drag for dense sprays should account for the liquid fraction

effect (collective effects), at some distance from the source droplets can be assumed to behave as

isolated particles.

Separated flow is mostly associated with the interaction of gas flows with thin liquid films.

Interfacial drag on thin film is responsible, for instance, for slowing down the downward motion of

liquid along structures (e.g., condensate along walls, external water film cooling the metallic

containment of the AP1000, etc.) and droplet entrainment.

The theory for isolated particles (bubble and droplets) is well known and documented in the

literature. As for separated flow, most information is related to annular flow in tubes and horizontal

stratified flow, but the main results as concerns interfacial shear are applicable to any continuous,

deformable gas-liquid interface. The friction factor correlations for each of the flow regimes are also

available in literature. Thus no experiments are needed for validation.

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References:

Dispersed bubble regime:

R. Clift, and W.H. Gauvin, Motion of Entrained Particles in a Gas Stream, Canadian Journal of

Chemical Engineering, Vol. 49, pp. 439–448, 1971

Slag regime:

There is no specific reference for this regime. Most of the computer codes assume some smooth

transition between the two dispersed flow regimes.

Dispersed droplet regime:

R. Clift and W.H. Gauvin, Motion of Entrained Particles in a Gas Stream, Canadian Journal of

Chemical Engineering, Vol. 49, pp. 439–448, 1971

Separated flow regime:

N.K. Popov and U.S. Rohatgi, Effect of Interfacial Shear and Entrainment Models on Flooding

Predictions, AIChE Symposium Series, Heat Transfer – Seattle, Vol. 79, No. 225, pp. 190-203, 1983

T. Fukano and T. Furukawa, Prediction of the Effects of Liquid Viscosity on Interfacial Shear Stress

and Frictional Pressure Drop in Vertical Upward Gas-Liquid Annular Flow, Int. J. Multiphase Flow,

Vol. 24, No. 4, pp. 587-603, 1998

J.E. Kowalski, Wall and Interfacial Shear Stress in Stratified Flow in a Horizontal Pipe, AIChE

Journal, Vol. 33, No. 2, pp. 274–281, 1987

Prepared by: A. Vasić (AECL) and M. Andreani (PSI)

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3.1.26 P1-26 - Liquid Film Flow

Description:

During the course of loss of coolant accident, steam and water flow into the containment through

the break. The steam is partially condensed along the cold steel and concrete walls, while the water

pools, generally at the bottom of the containment, where it can be pumped away through a safety

system.

Given enough fluid, the liquid condensate from the walls or overflow from a pool can run

downwards along the walls, and this can affect the heat transfer processes. Lumped parameter

containment codes have specific models for such situations (e.g., drain wall model in COCOSYS).

Interestingly, in some PANDA tests, it was found that condensation produced a liquid film on the

walls in the region above the injection and CFD simulations have shown that the key phenomenon

was the re-vaporization of the condensate film.

References:

Andreani M., Paladino D. and George T., “On the unexpectedly large effect of the re-vaporization of

the condensate liquid film in two tests in the PANDA facility revealed by simulations with the

GOTHIC code”, XCFD4NRS conference, Grenoble, France, 2008 September 10-12

Klein-Hessling W., “COCOSYS short description”, GRS report, 2008 May


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3.1.27 P1-27 - Gas Dissolved in Water (No Experiments)

Description:

Gaseous products can be dissolved in water, and for dilute solutions, the amount of gas that can

be dissolved is governed by the Henry’s Law, which states that the (equilibrium) solubility of a gas in

a liquid at a particular temperature is proportional to the partial pressure of that gas above the liquid.

The proportionality constant (Henry’s constant) varies with temperature. Therefore, the amount of

gas that can stay in solution depends on temperature and pressure. At low temperatures, gas solubility

decreases with temperature. However, at higher temperatures (above 100°C for nitrogen, oxygen, and

hydrogen), the solubility starts to increase. Water temperature increases and boiling cause the release

of gases to the gas space.

Under normal operating conditions, two main sources of gases dissolved in the primary water

exist. Radiolysis of the water in the reactor core is always producing some elemental hydrogen and

oxygen. In order to limit corrosion, in PWR it is common practice to operate with an excess of

dissolved hydrogen in the primary coolant, which has the effect of scavenging the oxygen produced

by radiolysis. Moreover, a small fraction of the gaseous fission products (I, Xe, Kr) is released to the

coolant, due to diffusion through the fuel matrix and microscopic cracks in the fuel cladding.

Additionally, short half-life nitrogen is produced by transmutation of oxygen.

During a LOCA, the release of dissolved gases (from the sump water) has a minimal effect on

containment thermal-hydraulics. In case of severe accidents, a variety of sources exist for production

of gases. The gases dissolved in the corium melt are likely to play a role in the behavior of the melt

stream during the high pressure melt ejection scenario. If the melt is highly gas supersaturated,

vigorous gas effervescence - when the melt is ejected into low-pressure environment - could

significantly contribute to melt disintegration and possibly production of aerosols. In consideration of

the events in the power station of Fukushima, the release of hydrogen produced by radiolysis in the

spent fuel could lead to increased interest for the effect of gases dissolved in water.

The effect of gases on the performance of internal or external condensers in passive

containments is discussed in phenomenon P6-5. For this safety equipment, the continuous

accumulation of small amount of non-condensable gases inside or around the tubes impairs

condensation and therefore the long-term cooling of the containment. For these new designs, sources,

transport and distribution of gases within the containment have to be evaluated more carefully than for

current light water reactors.

Values of the Henry’s Law coefficient can be determined through laboratory experiments under

generic conditions. No specific experiments are therefore required for the conditions relevant for

containment simulation. The quality of simulation performed with the codes thus only depends on the

accuracy of the Henry's constant for the range of temperatures of interest.

References:

Atkins, P. and de Paula, J., Atkin’s Physical Chemistry, pp. 143-147, 8th Edition, Oxford University

Press, 2006

Pray, H.A., Schweickert, C.E. and Minnich, B.H., “Solubility of Hydrogen, Oxygen, Nitrogen, and

Helium in Water at Elevated Temperatures”, Industrial and Engineering Chemistry Vol. 44, No. 5, pp.

1146-1151, 1952

Frid, W., “Containment Severe Accident Thermalhydraulic Phenomena”, Report Rama III 89-04,

RAMA III Final Report, Studsvik Stockholm, 1991 August


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3.1.28 P1-28 - Gas Entrainment by Spray Droplets (Dousing)

Description:

Entrainment of the gaseous mixture in the containment due to spray activation – concern mainly

the local characterisation of the gas entrainment.

References:

J. Malet, Gas entrainment by one single PWR spray, SARNET-2 Elementary benchmark - Results

report, IRSN/ PSN-RES/SCA/LEMAC/2012-11, 2012


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3.1.29 P1-29 - Heat and Mass Transfer of Spray Droplets (Dousing)

Description:

Heat and mass transfer (condensation, evaporation) of droplets due to spray activation.

References:

Malet J., SARNET-2 Droplet heat and mass transfer elementary benchmark, comparison report, IRSN

Technical Report DSU/SERAC/LEMAC/11-04, 2011

K.V. Beard and H.R. Pruppacher, “A wind tunnel investigation of the rate of evaporation of small

water drops falling at terminal velocity in air”, J. Atmos. Sci, Vol. 28, pp. 1455-1465, 1971

B. Abramzon and W. Sirignano, “Droplet vaporization model for spray combustion calculations”, Int.

J. Heat Mass Transfer, 32, 1605–1618. 16, 19, 20, 21, 1989


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3.1.30 P1-30 - Droplet Interaction (Dousing)

Description:

Droplet interaction (coalescence, bouncing, splashing) due to spray activation.

References:

Rabe C., Malet J. and Feuillebois F., “Experimental investigation of water droplet binary collisions

and description of outcomes with a symmetric Weber number”, Physics of fluids, Vol. 22, 2010

Foissac A., “Modélisation des interactions entre gouttes en environnement hostile”, Thèse de Doctorat

de l’Université Paris VI, 2011


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3.1.31 P1-31 - Mixing by Sprays

Description:

This phenomenon deals with mixing of a gas (and or vapour) by the activation of spray system.

This includes light gas (i.e., hydrogen), which concerns mainly a global gas mixing by the activation

of spray.

References:

Malet J. and Vizet J., SARNET spray benchmark, dynamic part: TOSQAN test 113, code-experiment

comparison, IRSN Technical Report DSU/SERAC/LEMAC/08-04, 2008

Erkan N., Kapulla R., Mignot G., Zboray R. and Paladino D., Experimental investigation of spray

induced gas stratification break-up and mixing in two interconnected vessels, Nucl. Eng. Des. 241,

2011


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3.1.32 P1-32 - Turbulence Induced by Sprays

Description:

Turbulence is induced when sprays are activated. Fine or large dispersed droplets in the flow

attenuate or enhance the gas phase turbulence. This result depends on different parameters, droplets

size, flow characteristic size, etc.

References:

This field is very large in the literature, mainly for engine sprays. Only one reference is given here.

A. Sadiki, M. Chrigui, J. Janicka and M.R. Maneshkarim, Modeling and Simulation of Effects of

Turbulence on Vaporization, Mixing and Combustion of Liquid-Fuel Sprays; Flow, Turbulence and

Combustion 2005 75: 105–130 C 2005


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3.2 Hydrogen Behaviour (Combustion, Mitigation and Generation) Phenomena

3.2.1 P2-1 - Deflagration

Description:

Deflagration deals with combustion with flame speeds on the order of several meters per second

to several hundred meters per second. The burning rate can be affected by initial conditions (mixture

compositions, pressure, and temperature), geometry of the confinement, location of ignition, and

turbulence level. For slow flames, the maximum deflagration pressure is bounded by the adiabatic

isochoric complete combustion (AICC) pressure.

In an accident involving a deflagration of a pre-mixed cloud of hydrogen, the general process is

as follows. A weak source, like a spark, produces the ignition of the reactive mixture. The flame

starts out as a slow flame with a velocity between several centimeters to several meters per second. In

the absence of turbulence and confinement, the flame will not suffer strong accelerations, and the

overpressure generated will be small. Under those circumstances, the main reasons for flame

acceleration are the flame instabilities, turbulence generated by the flame itself, as well as interactions

with surfaces.

In real buildings, industrial facilities, etc., however, there would likely be some internal structure

that would provide some sort of obstruction and confinement to the flame. Rooms, closed spaces,

equipment, and pipes are all examples of obstructions that could provide this. The expansion of the

gas generates a turbulent flow field, and feedback from this in turn increases the effective burning

rate, as well as the rate of expansion in turn. This mechanism results in flame acceleration, and under

certain circumstances can even lead to a transition to detonation. The strength of an explosion

depends on many different factors, but generally, the effects of mixture composition, mixture non-

uniformities can, at least, be characterized by its influence on the laminar flame speed.

In general, the main causes of a pressure build-up would be the degree of confinement, and the

propagation speed of the flame mainly driven by the turbulence field. In a situation where there are

numerous obstructions, the flame would be very likely to accelerate to velocities on the order of

several hundreds of meters per second. This increased burning rate is caused by the wrinkling of the

flame front by large eddies in the turbulent flow. Additionally, increased heat and mass transfer rates

occur at the reaction front, resulting in even higher rates of combustion.

In a realistic environment, turbulence is generated by the obstacles existing in the structure. As

the flame consumes the unburnt gas, the products expand, pushing the flame ahead and generating

turbulence. When the flame propagates past obstacles, which increase the intensity of the turbulent

flow field, the burning rate increases dramatically, which then increases both the flow velocity and

turbulence ahead of the flame. With an increased burning velocity comes increased pressure in the

flame front, and the acceleration of a flame due to the interaction with repeated obstacles constitutes a

strong positive feed-back loop.

Two competing mechanisms govern the pressure increments in a partially confined explosion.

Flame acceleration appears due to enhanced burning rate. This increased rate is created by the

turbulence generated when the flow overcomes the obstacles. The acceleration of the flame produces

a very significant increment of the pressure. On the other hand, venting provides some pressure relief

reducing the feedback mechanism described in the previous paragraph. The interaction between both

mechanisms is complex, and fluid dynamics calculations (CFD) are generally required to resolve

them. The ultimate effects of the explosion would be dependent on the balance between these two

mechanisms.

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References:

Ya.B. Zeldovich et al., The mathematical theory of combustion and explosions. Consultants bureau,

New York, London

David A. Frank-Kamenetskii, “Diffusion and Heat Transfer in Chemical Kinetics”, Russian editions:

Moscow-Leningrad: USSR Academy of Science Press, 1947; 1967; NAUKA Press (updated and

extended edition), 1987

Prepared by: J. Yanez (KIT)

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3.2.2 P2-2 - Hydrogen Flame Acceleration (FA)

Description:

The process immediately following a weak ignition in a combustible gas mixture is characterized

as deflagration, where the combustion propagates at subsonic speed into the unburned mixture (see

phenomenon P2-1). The initially smooth flame surface can be wrinkled due to the Landau-Darrieus

instability, which can be stabilized or destabilized by the thermal-diffusion effects. This can result in

the formation of a cellular flame leading to an increase of the flame surface and the acceleration of the

flow generated by the expansion of the combustion products. In addition, turbulence and the

obstacles located along of the flame path can cause further increase of the flow velocity, the flame

speed relative to a fixed observer and the flame surface. Depending on the mixture properties and

boundary conditions, the interaction of the flame with turbulence in the unburned gas can lead to

either weak flame acceleration within relatively slow, unstable, turbulent flame regimes, or to strong

flame acceleration resulting in fast flames that propagate at supersonic speeds. The distinction

between weak and strong flame acceleration regimes had been investigated by several authors (listed

in the references below). An important conclusion from this research was that as the mixture

expansion ratio is the key parameter that defines the border between weak and strong flame

acceleration. A sufficiently large expansion ratio was found necessary for the development of fast

flames. Within a sufficiently large round-up distance, supersonic combustion regimes can be

developed.

References:

S. Dorofeev, “Flame acceleration and explosion safety applications”, Proceeding of the combustion

institute 33, 2161-2175, 2011

S.B. Dorofeev, Kuznetsov M.S., Alekseev V.I., Efimenko A.A. and Breitung W., Evaluation of limits

for effective flame acceleration in hydrogen mixtures, J. Loss Prevent., Vol. 14, pp. 583-589, 2001

N. Chaumeix et al., H2 Gradient Effect on Premixed Flame Propagation in a Vertical Facility:

ENACCEF, Proceedings of the 20th Int. Colloquium on the Dynamics of Explosions and Reactive

Systems, Montréal, Canada, 2005

H. Cheikhravat et al., Influence of Hydrogen Distribution on Flame Acceleration, ECM 2007

Prepared by: A. Bentaib (IRSN)

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3.2.3 P2-3 - Deflagration-to-Detonation Transition (DDT)

Description:

Within a sufficiently large round-up distance, supersonic combustion regimes can be developed.

In this case, the initiation of detonation could only occur if the physical size of mixture volume were

to be sufficiently large compared to the detonation cell size, which represents the reactivity length

scale.

References:

S. Dorofeev, “Flame acceleration and explosion safety applications”, Proceeding of the combustion

institute 33 (2011) 2161-2175

S.B. Dorofeev, Kuznetsov M.S., Alekseev V.I., Efimenko A.A. and Breitung W., Evaluation of limits

for effective flame acceleration in hydrogen mixtures, J. Loss Prevent., Vol. 14, pp. 583-589, 2001

N. Chaumeix et al., H2 Gradient Effect on Premixed Flame Propagation in a Vertical Facility:

ENACCEF, Proceedings of the 20th Int’l Colloquium on the Dynamics of Explosions and Reactive

Systems, Montréal, Canada, 2005

H. Cheikhravat et al., Influence of Hydrogen Distribution on Flame Acceleration, ECM 2007

G. Ciccarelli and Dorofeev, S., Flame acceleration and transition to detonation in ducts. Progress in

Energy and Combustion Science 34, pp. 499-550, 2008


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3.2.4 P2-4 - Hydrogen Detonation

Description:

Detonation is an extremely fast and energetic combustion process with a leading shock wave

having a typical velocity on the order of a few kilometers per second. The bounding detonation

pressure can be evaluated using calculations based on the Chapman-Jouguet (C-J) equilibrium

detonation model. A detonation wave can pose a threat to equipment and the integrity of structures.

In an accident situation, direct initiation of detonation is very unlikely, as it would have to be

initiated from a strong shock source (e.g., high explosive detonation). However, transition from

deflagration to detonation (DDT) may still occur if a certain set of conditions are present. Whether

DDT occurs, however, depends on both the initial conditions (such as mixture composition, pressure,

temperature) and the boundary conditions (such as size of the enclosure, obstacle configuration and

obstacle spacing etc.).

References:

J.H.S. Lee, The detonation phenomenon, Cambridge University press, 2008

W. Fickett & W.C. Davis, Detonation, University of California press, 1979

H.I. Lee and D.S. Stewart, “Calculation of linear detonation instability”, J. Fluid Mech., Vol. 216, pp.

103-132, 1990


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3.2.5 P2-5 - Quenching of Detonations by Geometrical Constrains

Description:

The detonability limits of a reactive mixture are the critical conditions for the propagation of

self-sustained detonation. The critical conditions denote both the initial and boundary conditions of

the explosive mixture. If self-sustained detonation propagation is not possible, then the deflagration

to detonation transition could not be achieved.

The limiting tube diameter for stable detonation propagation in a cylindrical smooth-walled tube

can be estimated as λ/π, where λ is the size of the detonation cell. For wide planar channels with

height much less than width, the channel width must be at least as large as one detonation cell size for

the stable propagation of detonation to be achieved.

References:

Chapter 3, “OECD-SOAR: Flame Acceleration and DDT”, 1999


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3.2.6 P2-6 - Quenching

Description:

Flame quenching can occur for a wide spectrum of flame propagation regimes including laminar,

wrinkled and turbulent flames. Beyond a certain range of mixture composition, continued

propagation of the reaction front would no longer possible due to heat losses to walls and low burned

gas temperature. This composition limit is commonly known as the flammability limit and this

process can be defined as quenching of laminar flames.

Turbulent mixing processes that produce flame acceleration can also result in local or global

quenching. Local quenching is important to the flame acceleration process since it can lead to violent

secondary explosions and DDT.

References:

Chapter 2, “OECD-SOAR: Flame Acceleration and DDT”, 1999

Bradley, D., Gaskell, P.H., Gu, X.J. and Sedaghat, A., Premixed flamelet modeling: Factors

influencing the turbulent heat release rate source term and the turbulent burning velocity. Combustion

and flame 143,227, 2005

Bradley, D., Lau, A.K. and Lawes, M., Flame stretch rate as a determinant of turbulent burning

velocity. Phil. Trans. R. Soc. Lond. A, 338, 359-387, 1992

Bradley, D., Lawes, M., Kexin Liu and Woolley, R., The quenching of pre-mixed turbulent flames of

iso-octane, methane and hydrogen at high pressures, Proceedings of the combustion institute 31,

1393-1400, 2007

Prepared by: A. Bentaib (IRSN) and J. Yanez (KIT)

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3.2.7 P2-7 - Hydrogen Diffusion Flame (Standing Flame)

Description:

A standing flame occurs when a jet or plume of fuel emerging into an enclosure filled with air

ignites, and then continues to burn. Ignition does not always result in a stable standing flame unless

the hot combustion products can continuously ignite the fuel jet. The stability of a standing flame

depends on the balance between the heat generated by chemical reaction and the quenching due either

to flame stretching in a non uniform flow field, or rapid mixing of the unburned mixture with the

surrounding air.

References:

Shepherd, J.E., “Hydrogen-Steam Jet Flame Facility and Experiments”, NUREG/CR-3638/SAND84-

0060, 1984 October

Shepherd, J.E., “Analysis of Diffusion Flame Tests”, NUREG/CR-4534/SAND86-0419, 1987 August

Prepared by: Z. Liang (AECL)

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3.2.8 P2-8 - Hydrogen Mitigation - Passive Autocatalytic Recombiners

Description:

Passive Auto-catalytic Recombiners (PARs) are used to avoid an excessive hydrogen

accumulation inside the reactor containment in case of severe accident. Most of them are constructed

using catalytic materials (bed of beads or row of vertical plates with platinum and palladium on

ceramic washcoat) and housed in a metallic structure. Their behaviour is based on the exothermic

recombination of hydrogen into steam in presence of oxygen. The reaction rate is diffusion limited

(i.e., slower than any active ignition system) and its total recombination rate is nominally 0.3 g/s for a

medium-sized recombiner in an atmosphere containing 4% hydrogen (no O2 restriction). This surface

mechanism leads to an overheating of the catalytic plates and activates natural convection driven

circulation of gases in contact with the catalyst. PARs efficiency has been investigated in several

experiments performed both by manufacturer (such as AREVA, AECL/EACL, etc.) or research

institutes (such as FZJ, IRSN, CEA, etc.)

Recently, dedicated experiments had been performed in frame of the OECD/THAI project to

investigate the possible ignition by PARs and to study the effect of oxygen starvation of PARs

efficiency. Nevertheless, efforts are needed to investigate topics as:

the effect of carbon monoxide on hydrogen recombination,

PARs interaction with external convection flows,

PARs efficiency at reduced oxygen concentrations,

PARs interaction with CsI aerosols, and

unsteady PARs efficiency.

References:

D. Leteinturier et al., “Essais H2PAR: période mi-98 à fin-2000, Synthèse des essais, Conclusions du

programme”, Technical Report, IRSN, DPEA/DIR/02/01, (2002).

P. Rongier et al., “Studies of catalytic recombiner performances in H2PAR facility”, Proc. CSARP,

Bethesda, USA, 1997 May 5-8

O. Braillard, “Test of passive catalytic recombiners (PARs) for combustible gas control in nuclear

power plants”, Proc. 2nd

Int. Topical Meeting on Advanced Reactor Safety ARS, Vol. 97, pp. 541-

548, 1997

M. Sonnenkalb and G. Poss, “The International Test Programme in the THAI Facility and its Use for

Code Validation”, Proc. EUROSAFE, Brussels, Belgium, 2009 November 2-3

N. Meynet and A. Bentaib, “Numerical Study of Hydrogen Ignition by Passive Auto-catalytic

Recombiners”, Proc. 2nd

Int. Meeting of the Safety and Technology of Nuclear Hydrogen Production,

Control and Management, American Nuclear Society, San Diego, USA, 2010 June 13-16

E.A. Reinecke et al., “Open issues in the applicability of recombiner experiments and modelling to

reactor simulations”, Progress in Nuclear Energy, 52, pp. 136-147, 2010

Prepared by: N. Meynet and A. Bentaib (IRSN)

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3.2.9 P2-9 - Hydrogen Ignition by PARs (Weak Ignition)

Description:

A passive autocatalytic recombiner (PAR) is employed in nuclear reactors as a hydrogen

mitigation device (see Section P2-8) to prevent accumulation of high hydrogen concentrations. The

recombination of hydrogen is an exothermic reaction and as the hydrogen concentration increases, the

catalyst plate temperatures and exhaust gas temperatures also increase. Thus, PARs have a potential

to act as an unintended ignition source for hydrogen. Tests performed with various commercial PARs

have shown that the minimum hydrogen concentration needed for PAR ignition is dependent on the

PAR design.

References:

OECD, “Implementation of hydrogen mitigation techniques during severe accidents in nuclear power

plants”, NEA/CSNI/R(96)/27, OECD/GD(96)195, 1996

OECD, “SOAR on containment thermal-hydraulics and hydrogen distribution”, NEA/CSNI/R(99)16,

pp. 22-29, 1999 June

Kanzleiter, T., Multiple hydrogen-recombiner experiments performed in the BMC. Battelle

Ingenieurtechnik, Eschborn, Report BF-V68.405-02, European Commission, Draft Report CONT-

VOASM(97)-D005, 1997

Blanchat, T.K. and Malliakos, A., Performance testing of passive autocatalytic recombiners. In: Proc.

of the Int. Cooperative Exchange Meeting on Hydrogen in Reactor Safety, Paper 4.2, 1997

Bachellerie, E., Arnould, F., Auglaire, M., de Boeck, B., Braillard, O., Eckardt, B., Ferroni, F. and

Moffett, R., Generic approach for designing and implementing a passive autocatalytic recombiner

PAR-system in nuclear power plant containments. Nucl. Eng. Des. 221, 151–165, 2003

Reinecke, E.A., I.M. Tragsdorf and K. Gierling, Studies on innovative hydrogen recombiners as

safety devices in the containments of light water reactors. Nucl. Eng. Des. 230, 49-59, 2004

Fineschi, F., M. Bazzichi and M. Carcassi, A study on the hydrogen recombination rates of catalytic

recombiners and deliberate ignition. Nucl. Eng. Des. 166, 481–494, 1996

Prepared by: A. Bentaib (IRSN), J. Fontanet(CIEMAT), L.E. Herranz (CIEMAT) and J. Yanez

(KIT)

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3.2.10 P2-10 - Hydrogen Mitigation by Hydrogen Ignitors (Mild Ignition)

Description:

The deliberate hydrogen combustion is used for mitigating of hydrogen deflagration or

explosions effects. Igniters initiate combustion in containment volumes in mixtures near the

flammability limit by diffusion flames and/or slow deflagrations. Thus, the use of ignitors prevents

the accumulation of hydrogen and the formation of dangerously rich mixtures that would be able to

produce energetic modes of combustion. Several types of ignitors (spark plugs, glow plugs) should be

optimally placed in the containment so that the flame would not propagate to regions of higher

concentration and produce damaging effects. Other important problems related with the effectiveness

of the igniters is that the atmosphere surrounding igniters remains poor in hydrogen and in the

absence of fast local mixing mechanism the igniter may fail in generating a flame while hydrogen

continues to accumulate elsewhere in the compartment. Effects of gas composition, as well as the

thermodynamic conditions can be characterized by the laminar flame velocity for which numerous

data is available, e.g., Szabo et al. (2012).

References:

OECD, “Implementation of hydrogen mitigation techniques during severe accidents in nuclear power

plants”, NEA/CSNI/R(96)/27, OECD/GD(96)195, 1996

OECD, “SOAR on containment thermal-hydraulics and hydrogen distribution”, NEA/CSNI/R(99)16,

pp. 22-29, 1999 June

Szabo, Yanez, Kotchourko, Kuznetsov, Jordan, Parameterization of laminar flame speed dependence

on pressure and temperature in hydrogen-air-steam mixtures, Combustion science and technology,

posted online, 2012 June 29

Prepared by: J. Fontanet (CIEMAT), L.E. Herranz (CIEMAT) and J. Yanez (KIT)

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3.2.11 P2-11 - Strong Ignition of Hydrogen

Description:

Strong ignition of hydrogen can occur through a strong spark, electric arc, high explosive

detonation, and/or ignition in shock reflections. This initiation mechanism provides an enhanced

overpressure, and under certain circumstances, these ignition sources can trigger a detonation.

References:

Dorofeev S.B., Sidorov V.P., Velmakin S.M. et al., Large Scale Hydrogen-Air Detonation

Experiments. The effect of Ignition Location and Hydrogen Concentration on Load. Laboratory of

induced Chemical Reactions. Russian Research Center “Kurchatov Institute”. Report number RRCKI-

80-05/59 1993


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3.2.12 P2-12 - Jet Ignition of Hydrogen

Description:

In a situation where a burnable mixture (uniform, or non uniform) is divided in two

interconnected chambers. Ignition in the first chamber will result in a pressure rise that will force a jet

of fresh combustible mixture, followed by hot combustion products into the second chamber.

Depending on the nature of the situation, such as the diameter of the connecting duct, it would be

possible that the mixture in the second chamber could be ignited as well. Ignition can occur due to

transmission of turbulent flame through the duct (if it is of sufficiently large diameter), or due to

combustion revival (with time delay) between flame quenching in the duct and flame re-ignition in the

turbulent jet of combustion products.

There is a complex relationship that exists between chemical kinetics, gas dynamics, and mass,

momentum, and energy transfer processes in jet ignition. An initial analysis of the phenomenon can

be performed based on the approximation of the induction time (Zeldovich et al. (1985).

References:

Ya.B. Zeldovich et al., “The mathematical theory of combustion and explosions”, Plenum Publishing

Corporation, 1985 January

S.B. Dorofeev, V.P. Sidorov, S.M. Velmakin et al., Large Scale Hydrogen-Air Detonation

Experiments. The effect of Ignition Location and Hydrogen Concentration on Load. Laboratory of

induced Chemical Reactions. Russian Research Center “Kurchatov Institute”, Report number RRCKI-

80-05/59, 1993

Carnasciali, J.H. Lee, R. Knytautas and F. Fineschi, Turbulent Jet Initiation of Detonation”,

Combustion and Flame, Vol. 84, 170, 1991

D.J. MacKay, S.B. Murray, I.O. Moen and P.A. Thibault, Flame-Jet Ignition of Large Fuel-Air

Clouds, 22nd

Symposium (Int.) on Combustion, The Combustion Institute, 1339-1353, 1988

R. Knystautas, J.H. Lee, I. Moen and H.G. Wagner, Direct Initiation of Spherical Detonation by a Hot

Turbulent Gas Jet, 17th Symposium (Int.) on Combustion, The Combustion Institute, Pittsburgh, 1235-

1245, 1979

F. Mayinger, Jordan M., Eder A., Zaslonko I.S., Karpov V. P. and Frolov S. M. Flame-Jet Ignition of

Fuel-Air Mixtures. Experimental Findings and Modeling. Proc. 11th ONR Propulsion Meeting. FSU,

Tallahassee, 1998


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3.2.13 P2-13 - Radiolysis (Hydrogen Production by Water Radiolysis)

Description:

When water absorbs ionizing radiation (e.g., alpha, beta, gamma and neutron), chemical

reactions occur and a set of free radical (e.g., hydrated electrons, hydrogen atom, hydroxyl radicals)

and molecular species (e.g., hydrogen, hydrogen peroxide, oxygen) can be created. With continuous

radiation, steady state concentrations of these species are established.

References:

J.W.T. Spinks and R.J. Woods. “An Introduction to Radiation Chemistry”, John Wiley & Sons, Inc.,

Toronto, 1990.

Allen, A.O., Hochanadel, C.J., Ghormley, J.A., and Davis, T.W., 1952. Decomposition of Water and

Aqueous Solutions under Mixed Fast Neutron and Gamma Radiation. Phys. Chem., 56, 575- 586.

Hochanadel, C.J., 1952. Effects of Cobalt γ-Radiation on Water and Aqueous Solutions. J. Phys.

Chem., 56, 587- 593.

Hochanadel, C.J., 1955. Réaction de H2 et O2 Dans L’Eau Sous Irradiation; In Proceedings of the Int.

Conf. on the Peaceful Uses of Atomic Energy, United Nations. 7, 739.

Kabakchi, S.A., Shubin, V.N., Dolin, P.I., 1965. Steady States in the Radiolysis of Neutral Solutions

of Oxygen. Doklady Akademii Nauk SSSR, 165, 601-603.

Kabakchi, S.A., Shubin, V.N., Dolin, P.I., 1967. Influence of pH on Stationary Concentrations of the

Radiolysis Products of Aqueous Oxygen Solutions. High Energy Chemistry, 1, 127-131.

Schwarz, H.A., 1962. A Determination of Some Rate Constants for the Radical Processes in the

Radiation Chemistry of Water. J. Phys. Chem., 66, 255-266.

Prepared by: G.A. Glowa (AECL)

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3.2.14 P2-14 - Effect of Droplets on Hydrogen Combustion

Description:

Hydrogen flame propagates initially in subsonic regime and accelerates under turbulence effect

to reach supersonic regime. Most of the available literature concerns hydrogen flame propagation in

gaseous mixture within air, hydrogen and water steam. However, the analysis of severe accident

scenarios shows that the reactor containment atmosphere could enclose water droplets issued either

form bulk condensation or from spray system activation.

Recently, the effect of those water droplets had been investigated experimentally by Cheikhravat

(2010). These results show that the ignition of initial inert gas mixture is possible when water spray is

activated. Moreover, and according to the hydrogen concentration and to the water droplet size,

different behaviours have been identified:

For large droplets (d>250 µm) and regarding to the ratio, s, between flame and droplet velocities,

two situations have been distinguished:

i) For high values of s (s>>1) no significant effects have been observed.

ii) For small values of s (s~1), the evaporation of water droplets occur and the wrinkling of the

flame front is observed.

For small droplets, the flame front becomes wrinkled due to the turbulence induced by water

droplet, especially at low hydrogen concentrations.

Even if the use of spray enhances the turbulence, the spray-premixed flame interactions lead to

low pressure values. On the other hand, due to the induced turbulence, it has been observed that

sprays increase the pressure slope for lean hydrogen-air mixtures.

References:

Cheikhravat H., “Etude expérimentale de la combustion de l’hydrogène dans une atmosphère

inflammable en présence de gouttes d’eau”, PhD thesis Orléans University, 2010.

H. Cheikhravat et al., “Evaluation of the Water Spray Impact on Premixed Hydrogen-Air-Steam

Flames Propagation”, Proceeding American Nuclear Society conference, San Diego, 2010.

Bjerketvedt D. and Bjørkhaug M., “Experimental investigation: Effect of waterspray on gas

explosions”, Report prepared by the Christian Michelsen Institute, Bergen, Norway, for the UK

Department of Energy, OTH 90 316, HMSO, 1991.


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3.3 Aerosol and Fission Product Behaviour Phenomena

3.3.1 P3-1 - Aerosol Formation in a Flashing Jet

Description:

The discharge of a water jet from high pressure, high temperature reservoir conditions to

conditions that are near atmospheric results in a rapid production of steam throughout the bulk water.

This causes an energetic fragmentation of the liquid water into small droplets over a short time scale.

The associated acceleration of the jet provides a second fragmentation mechanism, since any water

droplets experiencing conditions under which they exceed a critical Weber number (ratio of inertial

over viscous forces acting on the droplet) will undergo further breakup.

References:

E. Hervieu T. Veneau Experimental determination of the droplet size and velocity distributions at the

exit of the bottom discharge pipe of a liquefied propane storage tank during a sudden blowdown J.

Loss Prev. Ind. Ç, 6, 423-455, 1996

S. Vandroux-Koenig, G. Berthoud Modelling of a two-phase flow Momentum jet close to a breach in

the containment vessel of a liquefied gas, J. Loss Prev. Ind., 10, 1, 17-29, 1997



Aerosols”, NEA/CSNI/R(2009)5, pp. 127, 2009 December

G. Polancoa, A. E. Holdøb, G. Mundayc, General review of flashing jet studies, Journal of Hazardous

Materials, Vol. 173, Issues 1–3, pp. 2–18, 2010 January 15


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3.3.2 P3-2 - Aerosol Formation in a Steam Jet

Description:

Release of high temperature steam into a lower temperature environment can result in the

formation of large numbers of water droplets due to steam condensation. The presence of seed

particles (e.g., dust, fission product aerosols) in the steam jet prior to this condensation can alter the

result of the condensation process, producing an aerosol population having different characteristics

than would have resulted from steam condensation alone.

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 127, 2009 December

Polancoa, G., A.E. Holdøb and G. Mundayc, General review of flashing jet studies, Journal of

Hazardous Materials, Vol. 173, Issues 1–3, pp. 2–18, 2010 January 15

Thomas K. Lesniewski and Sheldon K. Friedlander, Particle nucleation and growth in a free turbulent

jet, Proc. R. Soc. Lond. A, 1998 September

Girshick, S.L., Chiu, C.P., Time dependent aerosol models and homogeneous, nucleation rates.

Aerosol Science and Technology 13, 465, 1990

Martin, F., La nucléation homogène: étude des intéractions vapeurs-aérosols, dans le circuit primaire

d’un réacteur nucléaire lors d’un accident grave. Ph.D., Thesis, Universite´ de Provence/Aix-

Marseille I., 1997


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3.3.3 P3-3 - Aerosol Impaction (Jet Impingement)

Description:

A two-phase flashing jet impinging vigorously onto one or more vessel surfaces results in very

complex flow patterns (large velocity gradients, very sharp changes in flow direction and large rates

of energy dissipation through turbulence). As well, it is likely that virtually all aerosol formation and

removal process are occurring. Therefore the term “jet impingement”, used in the sense of an aerosol

removal mechanism, is a convenient catch all term, encompassing a number of very complex aerosol

removal mechanisms that are recognized separately, such as turbulent deposition, impaction, inertial,

diffusional, phoretic deposition, and others.

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 25, 69, 79, 2009 December

Giralt, F., Chia, C.J. and Trass, O., Characterization of the impingement region in an axisymmetric

turbulent jet. Ind. Eng. Chem. Fundam. 16, 21-28, 1977

Martin, H., Heat and mass transfer between impinging gas jets and solids surfaces. In Advances in

Heat Transfer, Vol. 13, (Edited by Hartnett, J. P. and Irvine, T. F. Jr.,) Academic Press, New York,

pp. 1-60, 1977

Mercer, T.T. and Stafford, R.G., Impaction from round jets. Ann. Occup. Hyg. 12, 41-48, 1969

Kastner, W. and R. Rippel, Jet impingement forces on structures — Experiments and empirical

calculation methods, Nuclear Engineering and Design, Vol. 105, Issue 3, pp. 269–284, 1988 January 2


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3.3.4 P3-4 - Thermophoresis

Description:

Aerosol droplets or particles are subject to continuous agitation by the thermal motion of

molecules in the gas that contains these droplets or particles. The presence of a temperature gradient

in the gas will also result in a gradient in the measure of this thermal agitation. Any particles or

droplets present in such a gradient will see a net force tending to move them down the gradient.

Under these conditions, there will be a flux of particles or droplets towards lower temperature.

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 35, 45-46, 68, December 2009

L. Talbot et al., Thermophoresis of particles in a heated boundary layer, Journal of Fluid Mechanics,

101, pp. 737_758 (1980)

A. Zoulalian, T. Albiol, Evaluation of aerosol deposition by thermo and diffusiophoresis during flow

in a circular duct - application to the experimental programme 'Tuba diffusiophoresis', Canadian

Journal of Chemical Engineering, Vol. 76, Issue 4, August 1998, pp. 799-805

V. Saldo, E. Verloo, A. Zoulalian, Study on aerosol deposition in the PITEAS vessel by settling,

thermophoresis and diffusiophoresis phenomena, J. Aerosol Science, vol 29, suppl.1, pp. S1173-

S1174 (1998)

Romay, F. J., Takagaki, S. S., Pui, D. H. Y., and Liu, B. H. Y. (1998). Thermophoretic deposition of

aerosol particles in turbulent pipe flow. J. Aerosol Sci. 29: 943-959.

Tsai, C. J., Lin, J. S., Aggarwal, G., and Chen, D.R. (2004). Thermophoretic deposition of particles in

laminar and turbulent tube flows. Aerosol Science of Technology 38: 131-139

Sagot, B., Antonini, G., and Buron, F. (2009). Annular flow configuration with high deposition

efficiency for the experimental determination of thermophoretic diffusion coefficients. Journal of

Aerosol Science 40: 1030-1049

Healy, D. P., and Young, J. B. (2010). An experimental and theoretical study of particle deposition

due to thermophoresis and turbulence in an annular flow, Int. Journal of Multiphase Flow 36: 870-

881

Prepared by: J. Malet, N. Michielsen, E. Brugiere, B. Clement (IRSN)

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3.3.5 P3-5 - Diffusiophoresis

Description:

The transfer of particles due to a concentration gradient.

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 32, 68, December 2009

M. Missirlian, Modélisation des dépôts d'aérosols par diffusiophorèse dans un écoulement.

Application aux réacteurs à eau sous pression en situation accidentelle, Thèse Université de Provence

/ Aix-Marseille 1 (1999)

A. Zoulalian, T. Albiol, Evaluation of aerosol deposition by thermo and diffusiophoresis during flow

in a circular duct - application to the experimental programme 'Tuba diffusiophoresis', Canadian

Journal of Chemical Engineering, Vol. 76, Issue 4, August 1998, pp. 799-805

V. Saldo, E. Verloo, A. Zoulalian, Study on aerosol deposition in the PITEAS vessel by settling,

thermophoresis and diffusiophoresis phenomena, J. Aerosol Science, Vol. 29, Suppl.1, pp. S1173-

S1174 (1998)

Prepared by: J. Malet, B. Clement (IRSN)

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3.3.6 P3-6 - Liquid Aerosol Evaporation

Description:

The evaporation of water vapour from airborne aerosol particles decreases the sizes of these

particles. The evaporation rate depends on the drop diameter, relative velocity of the drops in the

vapor phase, drop and vapor temperature, vapor phase viscosity and density, steam diffusivity in the

vapor phase and the steam concentration in the vapor phase. This also includes mist depletion (Small

drop depletion rate due to super heat in the vapor phase or high mist concentration.)

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 28, 29, 66, December 2009

Chang, R., and E. J. Davis, 1976: Knudsen aerosol evaporation. J. Colloid Interface Sci., 54, 352–

363.Fuchs, N. A., 1959: Evaporation and Droplet Growth in Gaseous Media. Pergamon, 72 pp.


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3.3.7 P3-7 - Condensation on Aerosols

Description:

This includes the condensation of steam on water droplets and the formation of water droplets in

a super saturated atmosphere. The condensation of water vapour onto airborne aerosol particles

increases the sizes of these particles. The condensation rate depends on the drop diameter, relative

velocity of the drops in the vapor phase, drop and vapor temperature, vapor phase viscosity and

density, steam diffusivity in the vapor phase and the steam concentration in the vapor phase. This

also includes mist generation (small drop formation due to super saturation in the vapour phase), as

well as hygroscopic effects.

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 28, 29, 66, 2009 December

Gelbard F., Modeling multicomponent aerosol particle growth by vapour condensation, Aerosol

Science and Technology, 12:399-412, 1990

Michael Mozurkewich, “Aerosol Growth and the Condensation Coefficient for Water: A Review”,

Aerosol Science and Technology, Vol. 5, Issue 2, 1986, pp. 223-236

Nadykto, A. B., E. R. Shchukin, M. Kulmala, K. E. J. Lehtinen, and A. Laaksonen, 2003: Evaporation

and condensational growth of liquid droplets in nonisothermal gas mixtures. Aerosol Sci. Technol.,

37, 315–324

Qu, X., and E. J. Davis, 2001: Droplet evaporation and condensation, in the near-continuum regime. J.

Aerosol Sci., 32, 861– 875

Qu, X., and E. J. Davis and B. D. Swanson, 2001: Non-isothermal droplet evaporation and

condensation in the near-continuum regime. J. Aerosol Sci., 32, 1315–1339.

Vesala, T., M. Kulmala, R. Rudolf, A. Vrtala, and P. E. Wagner, 1997: Models for condensational

growth of binary aerosol particles. J. Aerosol Sci., 28, 565–598


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3.3.8 P3-8 - Gravitational Agglomeration

Description:

When aerosol particles settle under the action of gravity, they quickly acquire a constant settling

velocity due to a balance between the gravitational force acting on the particle and the drag force

exerted by the surrounding fluid on the particle. Particles of different sizes will have different settling

velocities. Thus when particles settle, the faster ones have the possibility of colliding with slower

ones, causing agglomeration.

References:




Clement et al., 1995, “Charge Distributions and Coagulation of Radioactive Aerosols”, J. Aerosol Sci.

26 1207 – 1225.

Tuttle R F and Loyalka S K, “Gravitational Collision Efficiency of Non-spherical LMFBR Aerosols”,

Trans. American Nucl. Soc, Vol. 32, No. 40, 1981.

Tuttle R F and Loyalka S K, “Gravitational Collision Efficiency of Nonspherical Aerosols I:

Definitions of Shape Factors”, Nuclear Technology, Vol. 69, p. 319, 1985.

Tuttle R F and Loyalka S K, “Gravitational Collision Efficiency of Nonspherical Aerosols II: Motion

of an Oblate Spheroid in a viscous Fluid”, Nuclear Technology, Vol. 69, p. 327, 1985.

Tuttle R F and Loyalka S K, “Gravitational Collision Efficiency of Nonspherical Aerosols III:

Computer Program NGCEFF and Calculation of Shape Factors”, Nuclear Technology, Vol. 69, p.

337, 1985.


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3.3.9 P3-9 - Diffusional Agglomeration

Description:

The small particles suspended in a gas can move randomly because of the thermal motion, this

phenomenon is called Brownian diffusion. The simultaneous random walk of a large number of

particles can cause inevitably collisions and agglomerations. This phenomenon is called (Brownian)

diffusional agglomeration.

References:




Clement et al., 1995, “Charge Distributions and Coagulation of Radioactive Aerosols”, J. Aerosol Sci.

26 1207 – 1225.

Loyalka SA, 1976, “Brownian Coagulation of Aerosols”, J. Colloid and Interface Science, Vol. 57,

No. 578, 1976.


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3.3.10 P3-10 - Turbulent Agglomeration of Aerosols

Description:

Turbulent agglomeration can be further divided into two phenomena, turbulent inertial

agglomeration and turbulent shear agglomeration. In turbulent flow, the fluid is in a state of random

motion containing eddies of varying sizes. When there is a large difference in the densities of the

particles and the fluid, the particles are not fully entrained by the turbulent eddies, and an inertial

effect forces particles out of one eddy into another. These particles can have large velocities relative

to those of other particles encountered, which can lead to agglomeration. The magnitude of this

process depends mainly on the particle sizes involved, turbulent energy dissipation rate, and particle

Reynolds number (based on the relative settling velocities of the particles).

References:




Clement et al., “Charge Distributions and Coagulation of Radioactive Aerosols”, J. Aerosol Sci. 26

1207 – 1225, 1995


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3.3.11 P3-11 - Drop Breakup

Description:

Drops can be broken up into smaller drops due to hydrodynamic forces exceeding surface

tension forces in a high speed water jet. Drops are also generated by the flashing of a high

temperature water jet. Drops can also be broken up into smaller drops due to drops interactions at

higher Weber numbers.

Experiments exist with drops constituted of various fluids, but very few experiments have been

conducted on the collision of water droplets at high Weber numbers. First experiments were

performed by Roth et al. (2008), where the splashing regime of water drops was observed. This field

is completely open to research.

References:

Roth, N., Rabe, C., Weigand, B., Feuillebois, F., Malet, J., 2007, Droplet Collision at High Weber

Number, ILASS-Europe, Mugla, Turkey

Rabe, C., 2009. Etude de la coalescence dans les rampes de spray: application au système d'aspersion

des Réacteurs à Eau Pressurisée, PhD Thesis, University Paris VI, France


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3.3.12 P3-12 - Gravitational Settling (Drop Settling)

Description:

All solid particles or liquid droplets suspended in a gas are subject to settling because of gravity.

Gravitational settling rates are determined by particle or droplet size, material density, temperature,

pressure, gas viscosity and mixing assumptions.

References:




James G. Crump, John H. Seinfeld, Turbulent deposition and gravitational sedimentation of an aerosol

in a vessel of arbitrary shape, Journal of Aerosol Science, Vol. 12, Issue 5, pp. 405–415, 1981

Nian-Sheng Cheng, “Comparison of formulas for drag coefficient and settling velocity of spherical

particles”, Powder Technology, Vol. 189, Issue 3, pp. 395-398, 2009 February 13


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3.3.13 P3-13 - Diffusional Deposition

Description:

When aerosol particles with a carrier gas enter a system of a given geometry, the Brownian

motion of the particles can cause the deposition of the particles to the wall surfaces. Such a process is

usually important for particles with size smaller than 1 micrometer.

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 32, 45, 69, December 2009

J.F. van der Vate, Investigations into the dynamics of Aerosols in Enclosures Used for air pollution

studies, ECN-86, Netherlands Energy Research Foundation, Petten, The Netherlands, July 1980

Lai, A. C. K. (2002), “Particle deposition indoors: a review”, Indoor air 12, 211-214.

Chen, F., Tu, S. et Lai, A. C. K. (2006) Modeling particle distribution and deposition in indoor,

environments with a new drift-flux model. Atmospheric Environment 40, 357-367


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3.3.14 P3-14 - Inertial Deposition of Aerosols (Also called Impaction)

Description:

When aerosol particles in a carrier gas enter a system of a given geometry, the movement of the

particles will be affected by the flow field of carrier gas in the geometry. If the fluid streamlines are

curvilinear, the path of the particles may deviate from the fluid streamlines of the carrier gas because

of the particle inertia (or momentum). The deviation of the particles from fluid streamlines can cause

the deposition of the particles onto the wall surfaces. Such a process is called inertial deposition and

is usually important for particles with size larger than 1 micrometer.

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 33, 69, 76-79, December 2009

B.Y.H. Liu and J.K. Agarwal, J. Aerosol Science, 5 (1974) 145

L.A. Hahn, J.J. Stukel, K.H. Leong and P.K. Hopke, “Turbulent Deposition of Submicron Particles on

Rough Walls”, J. Aerosol Science, 16 (1985) 81.

Shimada, M., Okuyama, K. and Kousaka, Y. (1989) Influence of particle inertia on aerosol deposition

in a stirred turbulent flow field, J. Aerosol Sci. 20,419-429.

Lai, A. C. K. (2002) “Particle deposition indoors: a review”, Indoor air 12, 211-214.


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3.3.15 P3-15 - Turbulent Deposition of Aerosols

Description:

When aerosol particles are transported by the carrier gas flowing inside a conduit or over a plate,

the particles can acquire transverse (radial) velocities caused by turbulent eddy diffusion. Such

transverse (radial) velocities can cause the particles to cross the viscous sublayer and deposit on the

wall surface.

References:



Aerosols”, NEA/CSNI/R(2009)5. December 2009

Friedlander, S. K. et Johnstone, H. F. (1957) Deposition of suspended particles from turbulent gas

streams. Ind. Eng. Chem. 49, 1151-1156.

Brooke, J. W., Kontomaris, K., Hanratty, T. J. et McLaughlin, J. B. (1992) Turbulent deposition and

trapping aerosols at a wall. Phys. Fluids 4, 825-834.

Guha, A. (1997) A unified eulerian theory of turbulent deposition to smooth and rough surfaces. J.

Aerosol Sci. 28, 1517-1537.

Liu, B. Y. H. et Agarwal J. K. (1974) Experimental observation of aerosol deposition in turbulent,

flow. Aerosol Science 5, 145-155.

Liu, B. Y. H. et Hori, T. A. (1974) Aerosol deposition in turbulent pipe flow. Environmental Science,

and Technology 8, 351-356.

Sehmel, G. A. (1970) Particle deposition from turbulent air flow. J. Geophys. Res. 75, 1766, 1781.

Sippola, M. R. and Nazaroff, W. W. (2002) Particle deposition from turbulent flow: review of

published research and its applicability to ventilation ducts in commercial buildings. Lawrence

Berkeley National Laboratory Report.

Lai, A. C. K. (2002), “Particle deposition indoors: a review”, Indoor air 12, 211-214.

Nerisson P., Modélisation des transferts des aerosols dans un local ventilé, Thèse IRSN-2009/112,

2009


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3.3.16 P3-16 - Re-volatilisation

Description:

Revolatilisation of a fission-product compound on a structure surface occurs when its

equilibrium vapour pressure at the surface exceeds the partial pressure of the compound in the bulk

gas phase. Incongruent revaporisation (i.e., a chemical reaction with the surface, a deposited material,

or an atmospheric constituent, giving the vaporisation of a different chemical compound) can also

occur. The equilibrium concentrations that are associated with these reactions can be described using

thermodynamic data for the appropriate compounds. An unusual case of re-volatilisation may occur

when a deposited fission product nuclide decays from a lower-volatility chemical element to a higher

volatility element; the main decay of interest in this context is the decay of Te-132 to I-132.

The equilibrium vapour pressure is affected by the surface temperature, as well as the chemical

speciation and physical form (e.g., solution or segregated phases) of the deposit. Chemical speciation

probably has the strongest effect on the volatilities of fission products. The physical form will also

affect the equilibrium vapour pressure; dissolved compounds have lower equilibrium vapour pressures

than when they are present as segregated material. The pressure and chemical composition of the gas

will affect the chemical speciation and mass-transfer rates of the gas phase and deposited material.

The transport of vapour to and from the surface will also be affected by mass-transfer considerations

such as fluid temperature, pressure and composition, and flow regime (turbulent or laminar). The

surface temperature may be affected by decay heating from the deposited fission products, as well as

by thermalhydraulic conditions.

The fission product elements that are most likely to undergo revolatilisation are iodine, cesium,

tellurium and ruthenium. Revolatilisation of iodine has not been extensively studied, because it is

soluble in water and the condensing environment is generally considered to wash it into pools.

Cesium deposits undergo some re-volatilisation at high temperatures (about 600°C), but this is limited

in some cases by trapping by small silicate inclusions in surface oxides. The deposition and re-

volatilisation chemistry of ruthenium is very complex.

References:

D. Bottomley, R.S. Dickson, T. Routamo, J. Dienstbier, A. Auvinen, N Girault, “Revaporisation

Issues: An Overview”, Conference ERMSAR-07, FZ Karlsruhe, Germany, 12-14 June 2007,

SARNET-ST-C23, 2007 June

C. Mun, L. Cantrel and C. Madic, “Review of Literature on Ruthenium Behaviour in Nuclear Power

Plant Severe Accidents”, Nucl. Tech. 156, 332, 2006

Prepared by: R. Dickson (AECL)

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3.3.17 P3-17 - Aerosol Removal in Leakage Paths

Description:

Aerosol and gas can leak from a post-accident pressurized containment through a variety of

paths. Gas flow hydraulics affects drastically the aerosol transport so that leak pathways have been

classified accordingly: short pathways with sudden changes in flow cross-section area (i.e., valves and

seals); tortuous and relatively long pathways (i.e., concrete joints, cracks and penetration gaps); and

small diameter, long channels with high flow resistance (i.e., pores in intact concrete).

There are experimental and theoretical evidences of strong retention of particles in leak paths.

The main working aerosol removal mechanisms are Brownian diffusion, gravitational sedimentation

and, occasionally, inertial impaction as deposition mechanisms. If particle deposition is large enough,

pathway plugging may occur. Nonetheless, under turbulent flows deposition may not be permanent

and particles can bounce off surfaces they impact and/or resuspend from deposits due to changes in

gas flow over deposits, to particle impact and/or to substrate vibration. A key condition for aerosol

transport through leakage paths is steam content of gas.

References:



Aerosols”, NEA/CSNI/R(2009)5, pp. 130-136, December 2009

Gelain T., J. Vendel, “Research works on contamination transfers through cracked concrete walls”,

Nuclear Engineering and Design, Vol. 238(4), 2008, pp. 1159–1165

D. Mitrakosa, S. Chatzidakisa, E.P. Hinis, L.E. Herranz, F. Parozzi, C. Housiadas, “A simple

mechanistic model for particle penetration and plugging in tubes and cracks”, Nuclear Engineering

and Design, Vol. 238 (12) December 2008, pp. 3370–3378

Williams M.M.R., “Particle deposition and plugging in tubes and cracks (with special reference to

fission product retention” Prog. Nuclear Energy, 28 (1994), pp. 1–60

M.M.R. Williams, “A model for the transport of vapour, gas and aerosol droplets through tubes and

cracks” Progress in Nuclear Energy, Volume 30, Issue 4, 1996, pp. 333–416, 1995

Parozzi F., Chatzidakis, S., Gelain, T., Nahas, G., Plumecocq, W., Vendel, J., Herranz, L.E., Hinis, E.,

Housiadas, C., Journeau, C., Piluso, P., Malgarida, E., “Investigations on aerosol transport in

containment cracks”, Int. Conf. on Nuclear Energy for New Europe, 2005.

Parozzi, F., Caracciolo, E., Herranz, L.E., Housiadas, C., Mitrakos, D., Journeau, C. and Piluso, P.

(2008), “Investigation on aerosol leaks through containment cracks in nuclear severe accidents using

prototypic materials”. 2008 European Aerosol Science Conference (EAC2008), Thessaloniki, 2008

August 24-29.

T. Gelain, F. Gensdarmes, J. Vendel “Experimental study on aerosol penetration through cracked,

concrete wall”, Congress EAC 2004 Budapest, 2004 September 6-10

Prepared by: J. Fontanet (CIEMAT), L.E. Herranz (CIEMAT) and T. Gelain (IRSN)

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3.3.18 P3-18 - Pool Scrubbing of Aerosols

Description:

Pool scrubbing, or wet scrubbing, removes aerosol particles in gas bubbles rising in a water pool.

Several severe accident scenarios involve the transport paths of aerosols which include passages

through stagnant pools of water: BWR suppression pools, safety injection piping directly into a water

tank, steam generator tube rupture (SGTR) accidents concurrent with the stuck-open safety relief

valve, and molten core concrete interaction phenomena with corium covered by water. Additionally,

some containment venting systems employ water pools to clean the gas coming from the containment.

Several fundamental processes take place during aerosol pool scrubbing: diffusiophoresis,

thermophoresis, inertial impaction at the nearby of gas injection, gravity settling, centrifugal

deposition and diffusion during bubbles rise, Brownian diffusion, etc. Aerosol characteristics (i.e.,

size, hygroscopicity, etc.) are key factors for the effectiveness of these removal processes. Gas

hydrodynamics plays an essential role determining key variables for pool scrubbing such as bubbles

size and surface/volume ratio. In addition, other parameters like pool depth water sub-cooling, carrier

gas composition and temperature and velocity, injection mode, water composition, etc., heavily

influence individual pool scrubbing processes. Particular attention should be given to removal of

aerosols during formation and subsequent disintegration and coalescence of bubbles, and also to the

effects of submerged structures and contaminants (surfactants).

The retention of aerosols in the pool shows an inverted Gaussian type of trend as a function of

particle diameter with a minimum at about 0.1 μm. The effect of the pool scrubbing on the particle

size distribution is narrowing it towards the particle size yielding the minimum efficiency.

References:




A. Dehbi, , D. Suckow, S. Guentay, Aerosol retention in low-subcooling pools under realistic accident

conditions, Nuclear Engineering and Design, Volume 203, Issues 2–3, 2001 January 2, pp. 229–241

Prepared by: J. Fontanet (CIEMAT) and L.E. Herranz (CIEMAT)

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3.3.19 P3-19 - Radionuclide Transport

Description:

Transport of radionuclide (in the form of aerosols as well as in the form of gases) by the flowing

vapor phase. This also includes aerosol transport (The transport of small solid or liquid particles

carried by the gas).

References:

Allelein, H.J., Auvinen, A., Ball, J., Güntay, S., Herranz, L.E., Hidaka, A, Jones, A., Kissane, M.,

Powers, D., Weber, G., State-of-the Art Report on Nuclear Aerosols, OECD Report,

NEA/CSNI/R(2009)5, 2009.


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3.3.20 P3-20 - Radionuclide Decay Heat (No Experiments)

Description:

Radioactive decay is a process in which an unstable radionuclide transforms into a more stable

state, losing energy by emitting ionizing radiation in the process. When the radiation is absorbed by

matter, much of the energy is eventually converted into heat. This energy is normally contained

within the reactor core (or in the irradiated fuel bays) where the energy is utilized to make electricity

or removed as waste heat. However, after an accident, the radionuclides could be dispersed outside of

the core and may reach containment where they could contribute to the containment heat load.

A suite of codes would be required to calculate the decay heat load in containment. The total

recoverable decay heat would be estimated using physics codes that calculate the inventory and power

of fuel. Releases of activation products in the coolant may be a significant fraction of the releases for

events with few, if any, fuel failures. The oxidation of the fuel cladding and subsequent release of

various elements from the fuel and transport to the break location are the next phase of the

calculation. The location and quantity of heat adsorption will depend upon the type of radiation

released (alpha, beta, gamma etc.), its physical and chemical form within containment (gas phase,

aqueous phase or adsorbed on surfaces), particle size and state of aerosols, material densities,

shielding, distances etc. It is not likely that anything more than very simple heat adsorption

calculations are being used at this time. There are numerous sources of decay data, including ICRP-

107, and ENDF-VII.1.

References:

Int. Committee on Radiation Protection, “Nuclear Decay Data for Dosimetric Calculations”, ICRP-

107, Ann. ICRP 38 (3), 2008

M.B. Chadwick, M. Herman, P. Obložinský, M.E. Dunn, Y. Danon, A.C. Kahler, D.L. Smith, B.

Pritychenko, G. Arbanas, R. Arcilla, R. Brewer, D.A. Brown, R. Capote, A.D. Carlson, Y.S. Cho, H.

Derrien, K. Guber, G.M. Hale, S. Hoblit, S. Holloway et al., “ENDF/B-VII.1 Nuclear Data for

Science and Technology: Cross Sections, Covariances, Fission product Yields and Decay Data”,

Nuclear Data Sheets, 112 (12), pp. 2887-2996, 2008

Prepared by: G.A. Glowa (AECL) and D.H. Barber (AECL)

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3.3.21 P3-21 - Release Rate Change Due to Oxidizing Environment

Description:

Releases of many fission products from fuel or corium will increase in an oxidizing environment

(such as in containment or during air ingress into a vessel). For example, release rates of volatile

fission products from otherwise intact fuel increase during oxidation because of increases in diffusion

coefficients due to oxidation. Oxidation increases the volatility of some other fission products

(notably ruthenium and molybdenum) because their oxides are much more volatile than the parent

metal. Air oxidation particularly increases the release rate of ruthenium due to the formation of

ruthenium tetraoxide (RuO4), a highly volatile oxide. Some of the chemical reactions occurring

during P5-13 - Molten Core Concrete Interaction will also increase the oxidation state of the corium

and thereby increase the release rates of some fission products.

References:

J. McFarlane, J.C. Wren and Lemire, R.J., “Chemical Speciation of Iodine Source Term to

Containment” Nucl. Tech.., 138,162, 2002

C. Mun, L. Cantrel and C. Madic, “Review of Literature on Ruthenium Behaviour in Nuclear Power

Plant Severe Accidents” Nucl. Tech., 156, 332, 2006.

L.W. Dickson and R.S. Dickson, “Fission-Product Releases from CANDU Fuel at 1650°C: The

HCE4 Experiment”, 7th Int. Conf. on CANDU Fuel, Kingston, Ontario, 2001 September 23-27.

Lewis, B.J., Dickson R., Iglesias, F.C., Ducros, G., and Kudo, T.. Overview of Experimental

Programs on Core Melt Progression and Fission Product Behaviour. J. Nucl. Mat., 380, 126 (2008).

Kärkelä, T., Backman, U., Auvinen, A., Zilliacus, R., Lipponen, M., Kekki, T., Tapper, U. and

Jokiniemi, J., 2007. Experiments on the behaviour of ruthenium in air ingress accidents – Final

Report. Severe Accident Network of Excellence Document, SARNET-ST-P58,

VTT-R-01252-07.

Matus, L., Prokopiev, O., Alföldy, B., Pintér, A. and Hózer, Z., 2002. Oxidation and Release of

Ruthenium in High Temperature Air. KFKI Atomic Energy Research Institute Report, PHEBUS PF:

HU-02-1.

Matus, L., Nagy, I., Windberg, P., Vér, N., Kunstár, M., Alföldy, B., Pintér, A. and Hózer, Z., 2004.

Oxidation and Release of Ruthenium from Short Fuel Rods in High Temperature Air. KFKI Atomic

Energy Research Institute Report, AEKI-FRL-2004-111-01/01.


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3.3.22 P3-22 - Containment Chemistry Impact on Source Term

Description:

This comprises all of the chemical reactions that affect the iodine and ruthenium chemistry in the

containment without reacting directly with iodine or ruthenium (see Aqueous Phase Oxidation and

Reduction of Iodine Species (phenomenon P4-1) and Ruthenium Volatility and Behaviour in

Containment (phenomenon P3-23)). These reactions include: aqueous phase radiolytic oxidation of

organic materials to produce acids, which affect pH; air radiolysis, which produces nitric acid (affects

pH), as well as ozone and other air radiolysis products (increase volatility of deposited ruthenium);

dissolution of CO2 produced by core-concrete interaction in the water pool (affects pH).

References:

J.W.T. Spinks and R.J. Woods. An Introduction to Radiation Chemistry. John Wiley & Sons, Inc.,

Toronto, 1990.

J.C. Wren, J.M Ball and G.A. Glowa, “The Interaction of Iodine with Organic Material in

Containment”, Nucl. Tech, 125, 337, (1999).

J.C. Wren, J.M Ball and G.A Glowa, “The Chemistry of Iodine in Containment”, Nucl. Tech., 129,

297 (2000).


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3.3.23 P3-23 - Ruthenium Volatility and Behaviour in Containment

Description:

After release from the fuel, ruthenium in some chemical forms in the containment may become

volatile in the form of ruthenium tetraoxide (RuO4) (boiling point 40°C). If RuO4 is in the

containment, it is volatile and slightly soluble in water pools. The RuO4- and RuO4

2- chemical forms

may become volatile in the radiation field. The volatilities of these species are higher in acidic

solutions, and lower in basic solutions. RuO2·nH2O does not become volatile in solution, even under

irradiation. Deposited RuO2 may become volatile by reacting with air radiolysis products from the

gas phase (see phenomenon P3-16 - Re-volatilisation). These are the phenomena that have been

studied to date, and models based on these mechanisms will produce conservative results. However,

RuO4 and some of the other forms may become less volatile by reacting with organic and inorganic

materials present in the reactor sump, which would tend to reduce the impact of ruthenium volatility.

References:

Mun, C., Cantrel, L. and Madic, C., “Review of Literature on Ruthenium Behaviour in Nuclear Power

Plant Severe Accidents”, Nuclear Technology, 156, 332 (2006).

Mun, C., Cantrel, L. and Madic, C., “Study of RuO4 Decomposition in Dry and Moist Air”,

Radiochimica Acta. 95(11), pp. 643-656 (2007).

Mun, C., Cantrel, L. and Madic, C., “Oxidation of Ruthenium Oxide Deposits by Ozone”,

Radiochimica Acta. 96, pp. 375-384 (2008).

Mun, C., Cantrel, L. and Madic, C., “Radiolytic Oxidation of Ruthenium Oxide Deposits”, Nuclear

Technology 164(2), pp. 245-254 (2008).


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3.3.24 P3-24 - Aerosol Removal by Sprays (Dousing)

Description:

Aerosol (and fission-product) wash-out by spray droplets during spray activation.

References:

Porcheron, E., Lemaitre, P., Marchand, D., Plumecocq, W., Nuboer, A. and Vendel, J. Experimental

and numerical approaches of aerosol removal in spray conditions for containment application. Nuclear

Engineering and Design, Vol. 240, pp. 336-343, 2010.

Porcheron, E., Lemaitre, P., Nuboer, A., Vendel, J. Heat, mass and aerosol transfers in spray

conditions for containment application. Journal of Power and Energy Systems, Vol. 2, N°2, pp. 633-

647, 2008.

Porcheron E., Lemaitre P., Marchand D., Aerosol Removal by Emergency Spray in PWR

Containment, Journal of Energy and Power Engineering 5 (2011).

Firnhaber, M., Kanzleiter, T.F., Schwarz, S., Weber, G., “International Standard Problem ISP-37.

VANAM M3 – A, multi compartment aerosol depletion test with hygroscopic; aerosol material”,

Comparison Report, OCDE/GD(97)16, ; December 1996.

Prepared by: J. Malet, E. Porcheron, P. Lemaitre (IRSN)

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3.3.25 P3-25 - Re-suspension (Dry)

Description:

Dry aerosols deposited in the containment could be resuspended because of hydrogen

deflagration, steam explosion or fast depressurization due to containment failure or venting. Both

experimental and theoretical studies have been carried out in the past, so that several resuspension

models for it are already available in the literature.

References:



Aerosols”, NEA/CSNI/R(2009)5

A.H. Ibrahim, P.F. Dunn: Microparticle detachment from surfaces expose to turbulent air flow:

Effects of flow and particle deposition characteristics, J. Aerosol Science, vol 35, 2004, 805-821.

M.W. Reeks, D. Hall: Kinetic models for particle resuspension in turbulent flows: theory and

measurement, J. Aerosol Science, vol 31, 2001, 1-31.

Prepared by: J. Malet, S. Peillon (IRSN)

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3.3.26 P3-26 - Re-entrainment (Wet)

Description:

Re-entrainment of aerosols from boiling water pools (see OECD SOAR on Aerosols, 2009)

References:




Prepared by: M. Sonnenkalb (GRS)

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3.3.27 P3-27 - Aerosol De-agglomeration

Description:

Particle agglomerates can break-up for different reasons: large shear stresses caused by the

particle-gas relative motion at particle surface, collisions with structural surfaces and/or other

particles, etc. Large shear forces in actual reactor conditions can be generated by very high (up to

sonic) velocity in the steam generator tubes or in a sonic front when the aerosol laden gas is

discharged from a break into the secondary side from the primary (which pressure could be at least 2

times higher than in the secondary side). Collisions against structural components could occur in a

cross-flow configuration in which gas throws particles at high velocity against obstacles; an example

might be, also, the particle ejection into a dry secondary side of the steam generator during a

meltdown SGTR sequence.

The physics of de-agglomeration is not currently well understood. The relevance of de-

agglomeration to the SGTR accident is that it can shift the aerosol size distribution towards smaller

sizes significantly as the aerosols enter into the secondary side from the break. As an example,

ARTIST experiments show that de-agglomeration has caused a reduction of the aerodynamic mass

median diameter from an initial value of 3-4 μm to about 2 μm.

References:





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3.4 Iodine Chemistry Phenomena

3.4.1 P4-1 - Aqueous Phase Oxidation and Reduction of Iodine Species

Description:

Various oxidation and reduction reactions exist that interconvert iodine between many iodine

species (e.g., I-, I2 and IO3

-). Iodine speciation is a function of oxidation or reduction of iodine

species, and depends on the presence of radiolysis products of water, dose rate, solution pH,

impurities, etc. The chemical and physical nature of the species is quite different. The species I- and

IO3- are non-volatile whereas the dissolved I2 gas is volatile and reactive.

References:

Clement, B., Cantrel, L., Ducros, G., Funke, F., Herranz, L., Rydl, A., Weber, G., and Wren, C.,

“State of the ART Report on Iodine Chemistry”, NEA/CSNI/R(2007)1


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3.4.2 P4-2 - Inorganic Iodine Hydrolysis

Description:

Iodine hydrolysis is one of the key reactions that govern the conversion of volatile I2 into non-

volatile iodine species (I-, HOI) in the water phase. In general, equilibrium between the I2 and the

non-volatile species is achieved rapidly by hydrolysis. This equilibrium depends strongly on the pH

of the sump. At a high pH the equilibrium shifts towards the non-volatile species and the I2

concentration becomes small.

References:




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3.4.3 P4-3 - Inorganic Iodine Radiolysis in Water Phase

Description:

Radiolysis in the water phase is the most important conversion of non-volatile iodine into

volatile I2. In the presence of radioactive radiation (dose rate) water radiolysis forms reactive radicals,

which react with the iodine species in a complex way. I2 is formed radiolytically mainly from iodide

(I-) and iodate (IO3

-). Both reactions depend on the dose rate and the pH of the sump.

References:




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3.4.4 P4-4 - Homogeneous Organic Reactions in Water Phase

Description:

In sumps volatile organic iodides are formed from reactions of dissolved iodine with various

organic materials like oil, solvents, and destruction products leached from painted surfaces. Organic

iodides (e.g., methyl iodide, CH3I, or higher molecular weight organic iodides, HMWI) are mainly

produced by reactions of I2 and HOI with organic material. A part of the organic iodides is

decomposed radiolytically.

References:




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3.4.5 P4-5 - Iodine Reactions with Surfaces in the Water Phase

Description:

Adsorption is the deposition of dissolved volatile iodine (I2) onto a submerged surface. The

surface can be bare material (steel, concrete) or material coated with paint or other protective

covering. Desorption is the release of adsorbed iodine from a surface back to the aqueous phase.

References:




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3.4.6 P4-6 - Iodine reactions with surfaces in the gas phase

Description:

Adsorption is the deposition of gaseous iodine (I2) onto a surface. The surface can be bare

material (steel, concrete) or material coated with paint or other protective covering. The freshly

deposited iodine is loosely bound to the surface and can be easily desorbed, e.g. by a temperature

increase. This physisorbed iodine may react with the surface material to form iodine compounds,

which are more strongly bound to the surface (chemisorbed). In the gas phase I2 adsorption on

painted surfaces is a major sink for I2. The iodine deposited on paint may serve as a source for

organic iodides. Desorption is the release of physisorbed iodine from a surface back to the gas phase

or to the aqueous phase. Adsorption and desorption affect the I2 gas phase concentration.

References:




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3.4.7 P4-7 - Silver Iodine Reactions in the Water Phase

Description:

In the sump silver reacts with dissolved I2 and I- to the insoluble AgI, which accumulates on the

sump bottom. The silver originates from molten control rods and would be released into the

containment in aerosol form. It would be deposited mainly on the containment floors and washed

down into the sump. The total silver mass in the containment is much higher than the iodine mass and

therefore silver has the potential to bind large amounts of iodine. Elemental silver reacts with I2 and

oxidized (AgOx) silver reacts with I-. The reactions occur on the surface of the Ag or AgOx particles

in the sump. As a consequence of the iodine trapping in the sump the I2 concentration in the gas

phase is also reduced significantly.

References:




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3.4.8 P4-8 - Gas Phase Radiolytic Oxidation of Molecular Iodine (I2) (Iodine/Ozone Reaction)

Description:

Molecular iodine (I2) is oxidized by air radiolysis products (e.g., ozone) to form an aerosol (IOx)

by a gas-to-particle conversion. These particles are generally submicron and have the potential to stay

airborne in the atmosphere for long periods of time. In the presence of a nuclear aerosol the IOx

aerosol particles agglomerate with these bigger particles. The IOx aerosol is an important contributor

to the iodine source term.

References:




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3.4.9 P4-9 - Homogeneous Organic Iodine Reactions in Gas Phase

Description:

Organic iodine generation refers to the process by which organic iodides are created. This

process occurs on painted surfaces (dry or submerged) that have been in contact with I2. Organic

iodides are carbon based molecules that contain iodine. These compounds have a range of volatilities.

The most common example is methyl iodide (CH3I). Organic iodides are harder to trap on filters than

molecular iodine (I2).

References:




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3.4.10 P4-10 - RI (Organic Iodine) Radiolytic Destruction

Description:

Radiolytic iodine (RI) destruction is the decomposition of organic iodides (e.g., methyl iodide)

due to radiation in the gas phase. The reaction is the most important destruction mechanism for

organic iodide in the gas phase. Parts of the formed I2 react rapidly with air radiolysis products to

generate IOx aerosol.

References:




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3.4.11 P4-11 - Interfacial Mass Transfer

Description:

Mass transfer describes the exchange of volatile species between gas and water phases by

transport processes (diffusion, convection) through the interfacial surface areas, e.g., sump surface.

These processes are important for I2 and organic iodides, which are produced in the water phase

(sump) and transported from there into the gas phase. On the contrary, if the gas phase concentration

is higher, the volatile iodine species are transferred from the gas phase into the water. The partition

coefficient is the ratio of the concentration in the water phase over the concentration in the gas phase

at equilibrium. It is different for individual iodine species (e.g., I2 or CH3I). It mainly depends on the

temperature.

References:




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3.4.12 P4-12 - Decomposition of Iodides (CsI) by Heat-up in PARs

Description:

Metal iodide aerosols like CsI and CdI2 passing through a passive autocatalytic recombiner

(PAR) are exposed to high temperature, and may decompose thermally, leading to an additional

source of volatile I2 in the containment. Since in general several dozens of PARs are located in

containment, this additional I2 source may be significant.

References:




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3.4.13 P4-13 - Iodine Filtration

Description:

According to the different physical and chemical nature of the iodine species, different sorption

materials are used for iodine filtration. Particulate iodine (CsI, IOx) would be trapped efficiently in

common aerosol filters, like fibrous filters, venturi scrubbers, or gravel bed filters. Filtration of

gaseous iodine (I2 and organic iodides) needs special sorption materials, like carbon impregnated with

TEDA (triethylene di-amine), or special additives in the washing solution of scrubbers, e.g., NaOH,

Na2S2O3. In general the filtration efficiency is significantly lower for gaseous iodine than for iodine

aerosols.

References:




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3.4.14 P4-14 - Volatile Iodine Trapping by Airborne Droplets

Description:

Gaseous iodine (I2, RI) is transferred to droplets until equilibrium is reached. The droplets may

originate from volume condensation (for formation) or from a spray system. Iodine will equilibrate

with the iodine transferred into the water. Because of the large surface/volume ration of the droplet

iodine trapping is relatively fast. Operational filters are designed for long residence times and

relatively low material concentrations to be filtered while in venting filters the expected

concentrations are high.

References:




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3.4.15 P4-15 - Iodine Retention in Leakage Paths

Description:

When gaseous (I2) and particulate iodine (CsI, IOx) are transported by a gas flow through a

narrow leak they partially deposit on the surfaces of the path. I2 is adsorbed and the particles are

deposited by different processes (impaction, diffusion, sedimentation and diffusiophoresis). The

retention of organic iodides will be negligible in most cases. No experiments have been performed to

examine retention of the different forms of iodine in leakage paths.

References:




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3.4.16 P4-16 - I2 Interaction with Aerosols

Description:

Gaseous I2 is adsorbed on the surface of aerosol particles. Freshly deposited I2 is loosely bound

to the particle surface (physisorbed) and may be desorbed with changing conditions (increasing

temperature or decreasing I2 concentration in the gas phase). This physisorbed I2 will partly react

with the aerosol material to form chemisorbed iodine, which is strongly bound to the particles. I2

deposited on an aerosol is transported like an aerosol, and is subject to other effects like

agglomeration and deposition.

References:




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3.4.17 P4-17 - Iodine Wash-down

Description:

I2, particulate iodine (CsI, IOx), silver particles, and other aerosol materials deposited on walls

and floors of containment can be transported with the draining condensate (wash-down) into elevated

flat pools, and from there finally into the sump. In general soluble materials (e.g., CsI, CsOH) are

washed down more completely than insoluble materials (e.g., Ag). Deposited iodine and other fission

products can react with the surface material (mostly paint) to form species, which are attached to the

surface and can hardly be washed off. Soluble and insoluble fission products are also retained

partially in elevated pools by an incomplete dissolution respectively by the settling of insoluble

particles.

Wash-down processes determine the fraction of iodine and Ag reaching the sump. Many of the

iodine reactions depend essentially on these concentrations. Moreover wash-down governs the

distribution of decay heat of iodine and other fission products (FP) in the containment. A high

amount of iodine and other fission products in the sump may cause sump boiling.

References:




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3.4.18 P4-18 - Pool Scrubbing of Iodine

Description:

If an iodine loaded (I2, CsI-aerosol) gas is discharged through an orifice or nozzles into a water

pool, the iodine in the rising bubbles will be partly trapped in the water. Different deposition and

transfer mechanisms and the bubble size influence the efficiency of this trapping process. Pool

scrubbing may occur in connection with a steam generator tube rupture.

References:




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3.4.19 P4-19 - Iodine Release from Flashing Pool or Flashing Jet

Description:

If an iodine loaded (I2, I-, etc.), thermally saturated water pool undergoes a sudden

depressurization, the pool water would partly evaporate, perhaps violently and become dispersed into

small droplets. This flashing pool can release iodine. By transfer mechanisms, some of the dissolved

I2 would be released directly into the gas phase. The remaining portion would become part of the

airborne droplets, or would stay in the coolant. The water in the droplets would possibly evaporate

away, and species (I-, HOI, IO3-) dissolved inside the droplet would form much smaller, dry aerosol

particles. A flashing pool could occur, for instance, if there were a sudden leak in containment.

If an iodine loaded (I2, I-, etc.) coolant, under high pressure were to be discharged into a volume

with low pressure, the supersaturated jet would evaporate and disperse into small droplets. The iodine

release mechanisms into the gas phase are principally the same as for the flashing pool. A flashing jet

may occur with a steam generator tube rupture.

References:




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3.5 Core Melt Distribution and Behaviour in Containment Phenomena

3.5.1 P5-1 - Corium Release from Failed Dry Reactor Pressure Vessel

Description:

For a light water reactor (with a dry cavity on the outside of the RPV), there are two possible

failure modes for the RPV. The first failure mode is with the RPV at high pressure. This high

pressure can be due solely due to core heat-up, or because of dynamic failure caused by in-vessel fuel

coolant interaction. This can result in single phase ejection of corium melt or two phase steam-corium

melt ejection.

The second failure mode involves a RPV at low pressure (depressurized primary system) with

the corium melting through the vessel walls or failure of a penetration nozzle. This will result in a

“pouring” of the corium melt into the reactor cavity/pit (which may be dry concrete, water pool on top

of the concrete floor or a core catcher). There can be one or several opening(s) and pour(s).

The mass flow of the melt would depend on the breach location, the amount and composition of

the poured corium, the pool configuration in the lower head/plenum and the timing of the release

episode(s). For rapid ejection sequences, flowrate is also of major importance.

Depending on break location relative to the melt in the RPV and the break pressure, the release

of corium can be either an:

ejection of single phase corium liquid, or

ejection of a corium/steam two phase jet.

For a LWR with a wet cavity (VVER 440 Loviisa, AP1000) the vessel could fail if in-vessel

retention fails. A low pressure scenario is expected, since SAMG are designed to depressurize the

vessel for the IVR phase. The phenomena should be close to that of the dry low pressure failure

except that the pour will occur into a water filled cavity.

For a CANDU, ex-vessel represents the case when the corium leaves the calandria vessel. The

calandria vessel atmosphere is connected to the containment by rupture disks, so the calandria vessel

is always at containment pressure, meaning there would be no energetic ejection of the melt out of the

calandria vessel. As well, the calandria vessel is located in a water filled shield tank. So, in order for

the calandria vessel to fail, the water in the shield tank should be low enough (below the level of the

corium pool inside the calandria vessel) so that the calandria vessel wall heats up and fails at low

water level.

Chemical thermodynamic data are needed to predict the corium physical state and the nature of

its phases. Thermophysical properties such as density, viscosity, surface tension, thermal

conductivity are of high importance for modelling of corium behaviour. Typically corium release is

simplified for LP and Integral codes.

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References:

H.G. Willschutz, E. Altstadt, B.R. Sehgal, F.P. Weiss, Simulation of creep tests with French or

German RPV-steel and investigation of a RPV- support against failure, Annals of Nuclear Energy,

Volume 30, 2003, pp. 1033-1063.

V. Koundy, F. Fichot, H.G. Willschuetz, E. Altstadt, L. Nicolas, J.S. Lamy, L. Flandi, Progress on

PWR lower head failure predictive models, Nuclear Engineering and Design, Volume 238, 2008, pp.

2420-2429.

B.R. Sehgal, A. Theerthan, A. Giri, A. Karbojian, H.G. Willschutz, O. Kymalainen, S. Vandroux, J.

M. Bonnet, J. M. Seiler, K. Ikkonen, R. Sairanen, S. Bhandari, M. Burger, M. Buck, W. Widmann, J.

Dienstbier, Z. Techy, P. Kostka, R. Taubner, T. Theofanous, T.N. Dinh, Assessment of reactor vessel

integrity (ARVI), Nuclear Engineering and Design, Volume 221, 2003, pp. 23-53.

Prepared by: C. Journeau (CEA)

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3.5.2 P5-2 - Corium Entrainment Out of the Reactor Primary Vessel with Lateral Breaches

Description:

The melt is ejected at high or intermediate pressures, and consists of a multi-phase steam/corium

mixture, depending on the relative position of the hole to the corium level. Most SAMGs tend to

decrease the primary circuit pressure to prevent high pressure ejection. As discussed under

phenomenon P5-1, melt ejections at high pressure is not anticipated for CANDUs.

References:

M.M. Pilch, H. Yan, T.G. Theophanous, The probability of containment failure by direct containment

heating at Zion, Nucl. Eng. Des., 164 (1996) 1-36.

H. Yan, T.G. Theophanous, The prediction of direct containment heating, Nucl. Eng. Des. (1996) 95-

116.

L. Meyer, G. Albrecht, C. Caroli, I. Ivanov, Direct containment heating integral effect experiments in

geometries of European nuclear power plants, Nucl. Eng. Des. 239 (2009) 2070-2084.


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3.5.3 P5-3 - Corium Particles Generation from the Corium Pool

Description:

Corium particles can be generated from the corium pool by interaction between the corium pool

and the high speed surrounding gas. This interaction can lead to further oxidation and hydrogen

generation. This phenomenon is only modelled in specialized codes.

References:

Q. Wu, G. Zhang, M. Ishii, R. Lee, Modelling of corium dispersion in DCH accidents, Nucl. Eng.

Des., 164 (1996) 211-235.


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3.5.4 P5-4 - Corium Particles Generation from the Two Phase Jet

Description:

The corium may be fragmented inside the corium/steam two-phase jet.

References:

L Meyer, C. Caroli, Direct Containment Heating in Nuclear Safety in Light Water Reactors: Severe

Accident Phenomenology, B.R. Sehgal, ed., Academic Press, Waltham, MA, 2012.

L. Meyer, M. G. Gargallo, Low pressure corium dispersion experiments with simulant fluids in a

scaled annular cavity, Nucl. Technol., 141, 257-274, 2003.


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3.5.5 P5-5 - Corium Particles Entrainment

Description:

Entrainment of the corium particles along the reactor pit can occur, and depends on the cavity

geometry (rather horizontal or vertical entrainment in case of an annular space around the vessel).

References:

S.B. Kim, M-K Chung, H-Y Lee, M-H Kim, A parametric study of geometric effect on the debris

dispersal from a reactor cavity during high pressure melt ejection, Int. Comm. Heat Mass transfer, 22

(1995) 25-34.




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3.5.6 P5-6 - Corium Particles Trapping

Description:

Deposition of corium particles due to flow and geometrical aspects.

References:

R. Meignien, S. Mikasser, C. Spengler, A. Bretault, Synthesis of analytical activities for Direct

Containment heating, ERMSAR07, Karlsruhe, 2007.




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3.5.7 P5-7 - Direct Containment Heating

Description:

High pressure ejection of molten core material into the containment atmosphere would lead to

direct containment heating by the release of thermal and chemical energy of the debris.

References:

OECD/NEA/CSNI, High Pressure Melt Ejection (HPME) and Direct Containment Heating (DCH),

State of the Art Report, OCDE/GD(96)194, 1996.

L Meyer, C. Caroli, Direct Containment Heating in Nuclear Safety in Light Water Reactors: Severe

Accident Phenomenology, B.R. Sehgal, ed., Academic Press, Waltham, MA, 2012.


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3.5.8 P5-8 - Corium Jet Break-up in Water Pool

Description:

Complete fragmentation could occur for deep water pools (> 4 m deep) depending on melt

jet/stream conditions. This is similar to in-vessel situation, but under sub cooling, deeper water pools

and low pressure. Likewise, incomplete fragmentation would occur for shallow pools (< 4 m deep),

leaving a coherent jet.

References:

D. Magallon, I. Huhtiniemi, H. Hohmann, Lessons learnt from FARO/TERMOS corium melt

quenching experiments, Nucl. Eng. Des., 189 (1999) 223-238.

T.N. Dinh, V.A. Bui, R.R. Nourgaliev, J.A. Green, B.R. Sehgal, Experimental and analytical studies

of melt jet coolant interactions: a synthesis, Nucl. Eng. Des.189 (1999) 299-327

M. Bürger, Particulate debris formation by breakup of melt jets: Main objectives and solution

perspectives, Nucl. Eng. Des. 236 (2006) 1991-1997.


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3.5.9 P5-9 - FCI and Steam Explosion - Melt into Water Ex-Vessel (Melt Quenching)

Description:

In many BWRs (and some PWRs), a water pool is located under the reactor pressure vessel

(likewise the CANDU has a shield tank/reactor vault filled with water on the outside of the calandria

vessel), either by design or as a means to manage a severe accident. The release of molten corium

from the reactor pressure vessel into this external body of water could result in fuel coolant interaction

(FCI) including possible ex-vessel steam explosion.

FCI is rapid coolant vaporization due to molten fuel coolant contact, and generates fragmented

droplets. An FCI-induced steam explosion is a phenomenon in which molten fuel rapidly fragments

and transfers its energy to the coolant, resulting in steam generation, shock waves and possible

structural (including containment failure) damage. Steam explosions would also have a dramatic

effect on the debris bed granulometry, and thus coolability, even if it may not lead to structural

damages.

References:

Theofanous, T.G., “The Study of Steam Explosions in Nuclear Systems”, Nuclear Engineering and

Design, Vol. 155, Issue 1-2, pp. 1-26, April 1995.

Prepared by: P.M Mathew (AECL)

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3.5.10 P5-10 - Pressure Load on Corium Retention Devices

Description:

This phenomenon is a consequence of the pressure load due to FCI or steam explosion on the

core catcher. Consequences would be minor if the core catcher is well designed. This is only

modelled by specialized codes.

References:

I. Szabo, P. Richard, Y. bergamaschi, J.M. Seiler, A multi-crucible core-catcher concept: Design

considerations and basic results, Nucl. Eng. Des., 157, 417-435, 1995.

V.N. Mineev, F.A. Akopov, A.S. Vlasov, Yu. A. Zeigarnik, O.M Traktuev, Optimization of the

materials composition of the external core catchers for nuclear reactors, Atom. Ener. 93, 872-879,

2002.


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3.5.11 P5-11 - Particulate Debris Bed Formation

Description:

Once the melt has been cooled/solidified by the water pool, it will fall down through the water

pool and form a particulate debris bed on the floor of the water pool. This phenomena is similar to an

in-vessel case, but pools are deeper and at lower pressures for the ex-vessel case.

References:

Burger, M., Cho, S.H., Berg, E.v., and Schatz, A., “Breakup of melt jets as pre-condition for

premixing: Modeling and experimental verification”, Nuclear Engineering and Design, Vol. 155,

Issue 1-2, pp. 215-251, April 1995.

Prepared by: P.M. Mathew (AECL)

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3.5.12 P5-12 - Corium Debris (Solid) Heat Transfer

Description:

Heat transfer within the solid corium and also heat transfer at the solid boundary to vapour,

liquid or solid.

References:

V. Dauvois, S. Goldstein, C. Gueneau, K. Froment and J.M. Seiler, “Boundary Conditions for Liquid

Corium in Thermalhydraulic Steady State and Experimental Validation”, Proc. of ICONE 8 April 2–

6, 2000, Baltimore, USA (2000).


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3.5.13 P5-13 - Molten Core Concrete Interaction

Description:

Molten Core Concrete Interaction (MCCI) is the chemical interaction between concrete and the

molten core. It includes compound reactions, heat of reaction, concrete ablation and gas generation.

The molten core will interact with the concrete, breaking up the concrete into gases (i.e., hydrogen,

carbon monoxide), liquids and solid components. MCCI generates combustible gases, and also

attacks the containment basemat.

The gases that are produced move upwards through the melt (as a sparging gas). This sparging

gas can affect:

the melt stratification,

induce convective currents in the melt,

heat transfer at the liquid-liquid interface of miscibility gaps,

entrainment of melt as the gas flows through openings in upper crust,

an attack of the cavity walls could also lead to mechanical damages that may threaten

containment.

oxidation of metallic melt and generation of flammable gases

References:

H. Alsmeyer et al., Molten corium/concrete interaction and corium coolability – A state of the art

report, Report EUR 16649, European Commission, 1995.

MT Farmer, S Lomperski, S. Basu, Results of Reactor Materials Experiments Investigating 2-D Core-

Concrete Interaction and Debris Coolability, ICAPP04, Pittsburgh, 2004.

C. Journeau, P. Piluso, JF Haquet, E Boccaccio, V Saldo, JM Bonnet, S Malaval, L. Carénini, L.

Brissonneau, Two-dimensional interaction of prototypic corium with concretes: The VULCANO VB

Test series, Ann. Nucl. Energy, 36(2009) 1597-1613.

M. Cranga, B. Spindler, E. Dufour, D. Dimov, K. Atkhen, J. Foit, M. Garcia-Martin, T. Sevon, W.

Schmidt, C. Spengler, Simulation of corium concrete interaction in 2D geometry, Progr. Nucl.

Energy, 52 (2010) 76-83.


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3.5.14 P5-14 - Corium Melt Stratification

Description:

The phenomenon deals with separation of corium into an oxidic and metallic layer, resulting in

stratification of melt layers. This may result from corium, containing iron, interacting with concrete

and separating under gravity.

References:

G.A. Greene, J.C. Chen, M.T. Colin, Onset of entrainment between immiscible liquid layers due to

rising gas bubbles, Int J Heat mass transfer, 31 (1988) 1309-1317.

J.C. Casas, M.L. Corradini, Study of void fraction and mixing of immiscible liquids in a pool

configuration by an upward gas flow, Nucl. Technol., 99 (1992) 104-119

B. Tourniaire, J. M. Seiler, J. M. Bonnet: Study of the Mixing of Immiscible Liquids: Results of the

BALISE Experiments, NURETH-10, Seoul, Korea, 2003.


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3.5.15 P5-15 - Corium Spreading

Description:

Spreading of corium (ex-vessel) into nearby compartments, which are possibly filled with water.

Note that for BWRs, within the cavity there are numerous structures (e.g., CRDMs, structural beams,

etc.) that could impede corium transport after the lower RPV heads fails.

Additional information and experiments on corium spreading can be found in Sehgal (2012).

References:

J.J. Sienicki, M.T. Farmer, B.W. Spencer, Spreading of molten corium in Mk1 geometry following

vessel melt through, Joint meeting of the European Nuclear Society and the American Nuclear

Society, Washington, DC, USA, 1988.

G. Cognet et al., Corium Spreading and Coolability: the CSC project, Nucl. Eng. Des. 209 (2001)

127-138.

C. Journeau et al., Ex-vessel corium spreading: results from the VULCANO spreading tests, Nucl.

Eng. Des. 223 (2003) 75-102.

B.R. Sehgal (Editor), “Nuclear Safety in Light Water Reactors”, Elsevier, Chapter 4.2, 2012


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3.5.16 P5-16 - Molten Corium Heat Transfer

Description:

Heat transfer within the melt and also heat transfer from the melt to the boundaries (includes the

gas phase above the melt pool, and potentially a second liquid layer).

References:

Kutateladze and I.G. Malenkov, Hydrodynamic analogy between heat transfer and nucleate boiling

crisis in boiling and bubbling—experimental data, Heat Transfer Soviet Res. 16 (1984), pp. 1–46.

G.A. Greene and T.F. Irvine, “Heat transfer between stratified immiscible liquid layers driven by gas

bubbling across the interface”, ANS Proceedings of the National Heat Transfer Conference, Houston,

TX, July 24-27 1988.

J.M. Seiler, Phase segregation model and molten pool thermal-hydraulics during molten core-concrete

interaction, Nucl. Eng. Des. 166 (1996), pp. 259–267.

J.M. Bonnet, Thermal hydraulic phenomena in corium pools for ex-vessel situations: the Bali

experiment, ICONE 8 Baltimore (2000).


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3.5.17 P5-17 - Corium Evaporation/Vaporization

Description:

In a large corium pool, heat-up due to decay heat could lead to significant vaporization of metals

and/or fuels. This could have an impact on the fission product source term.

References:

B.W. Spencer, WH Gunther, DR Armstrong, DH Thompson, MG Chasanov, BR Sehgal, EPRI/ANL

Investigations of MCCI Phenomena and aerosol release, OECD Spec Mtg Core debris Concrete

Interaction Phenomena, Palo Alto 1986.


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3.5.18 P5-18 - Corium Solidification/Crust Formation

Description:

This effect involved the formation of a solid layer of corium resulting from cooling processes.

References:

M.T. Farmer et al., “Corium Coolability Under Ex-Vessel Accident Conditions for LWRs”, Nuclear

Engineering and Technology, Vol. 41, No. 5, June 2009.


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3.5.19 P5-19 - Cracking (Crust)

Description:

Crack formation could occur in the upper crust due to the thermal constraints applied. Water

may penetrate the cracks and improve the heat transfer.

References:

M.T. Farmer et al., “Corium Coolability Under Ex-Vessel Accident Conditions for LWRs”, Nuclear

Engineering and Technology, Vol. 41, No. 5, June 2009.


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3.5.20 P5-20 - Ex-Vessel Corium Coolability, Top Flooding

Description:

This issue concerns the ex-vessel corium coolability in the cavity by top water flooding and all

the encountered phenomena dealing with heat transfer, crust formation, sparging gas and cracking of

the crust.

Top flooding is not expected to arrest MCCI, except in some particular cases (thin corium layer

and early flooding), even if a recent MCCI2 test suggests it may be more efficient than previously

expected.

References:

H. Alsmeyer et al., Molten corium/concrete interaction and corium coolability – A state of the art

report, Report EUR 16649, European Commission, 1995.

S. Lomperski, MT Farmer, Experimental evaluation of the water ingression mechanism for corium

cooling, Nucl. Eng. Des., 237(2007)905-917.

KR Robb, ML Corradini, Towards understanding Melt Eruption Phenomena during Molten Corium

Concrete Interaction, ICONE18, Xi’an, 2010.

M.T. Farmer, R. Aeschlimann, DJ Kilsdonk, S Lomperski, The CCI-6 large scale Core-Concrete

Interaction experiment Examining Debris Coolability under Early Cavity Flooding Conditions,

OECD/NEA MCCI Seminar, Cadarache, 2010.


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3.5.21 P5-21 - Ex-Vessel Corium Catcher - Coolability and Water Bottom Injection

Description:

This phenomenon deals with water bottom injection to cool corium pool and its impact on

containment pressurization. The issue concerns the corium coolability by bottom water injection in

the corium catcher. Both items might affect the containment integrity.

Water bottom injection would result in:

fragmentation and mixing between melt and water

rapid quenching and solidification

strong steam production

References:

W. Widmann, M Bürger, W. Tromm, H Alsmeyer, Experimental and theoretical investigation on the

COMET concept for ex-vessel core melt retention, Nucl. Eng. Des., 236 (2006) 2304-2327.

D Paladino, SA Theerthan, BR Sehgal, DECOBI: Investigation of melt coolability with bottom

coolant injection, Progr. Nucl. Ener. 40 (2002) 161-206.

D.H. Cho, R.J. Page, S.H. Abdulla, M.H. Anderson, H.B. Klockow, M.L. Corradini, Melt quenching

and coolability by water injection from below: Co-injection of water and non-condensable gases, Nucl

Eng Des 236 (2006) 2296-2303.

S. Lomperski, R Aeschlimann, MT Farmer, D Kilsdonk, SSWICS Bottom Water Injection Systems to

Enhance Melt Cooling Rate, OECD/NEA MCCI Seminar, Cadarache, 2010


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3.5.22 P5-22 - Ex-Vessel Corium Catcher - Corium-Ceramics Interaction and Properties

Description:

Corium interactions occur with the corium catcher materials. For light water reactors, the corium

oxidic layer sinks to the bottom and interacts with the ceramic core catcher. The phenomenon is

related to stratified pool configuration, where the density ratio between metallic and oxidic phases

depends on previous phase, and includes effect of oxygen potential on the dissolution mechanism.

References:

J.M Seiler, K Froment, Material effects on multiphase phenomena in late phases of severe accidents

of nuclear reactors, Multiphase Sci. technol. 12 (2000) 117-257.

V.G. Asmolov et al., Choice of Buffer Material for the Containment Trap for VVER-1000 Core Melt,

Atom. Ener. 92 (2002) 5-14.

C. Journeau, C. Jégou, J. Monerris, P. Piluso, K. Frolov, Y.B. Petrov, R. Rybka, Phase

Macrosegregation during the slow solidification of prototypic corium, NURETH-10, Seoul, 2003.


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3.5.23 P5-23 - Effect of Non Homogeneous Ablation on Gate Ablation

Description:

Crust instability may introduce heterogeneity in concrete ablation above the gate. Concept EPR,

depends also on gate material. The EPR concrete has been tested. The last remaining issue is the

ablation of the bottom concrete when side refractory material is reached if ablation is faster sidewards.

References:

Christophe Journeau, Lionel Ferry, Pascal Piluso, José Monerris, Michel Breton, Gérald Fritz, Tuomo

Sevon, Two EU-funded tests in VULCANO to assess the effects of concrete nature on its ablation by

molten corium, 4th European Review Meeting on Severe Accident Research (ERMSAR-2010),

Bologna-Italy, 11-12 May 2010.


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3.5.24 P5-24 - Crust Anchorage

Description:

The melt ejection mechanism is different if the upper crust is floating or if it is anchored to the

reactor pit wall. In the latter case, the presence of a cavity between the pool and the crust may prevent

any ejection and limit upwards heat transfer.

References:

B.W. Spencer, M.T. Farmer, D.R. Armstrong, D.J. Kilsdonk, R.W. Aeschlimann, M. Fischer, Results

of MACE Tests M0 and M1, OECD/CSNI Spec Mtg Core debris Concrete Interaction, Karlsruhe,

1992.

H. Alsmeyer, T. Cron, G. Messemer, W. Haefner, ECOKATS-2: A Large Scale Experiment on Melt

Spreading and Subsequent Cooling by Top Flooding, ICAPP 04, Pittsburgh, 2004.

S. Lomperski, M.T. Farmer, Corium crust strength measurements, Nucl. Eng. Des. 239, (2009) 2551-

2561.


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3.5.25 P5-25 - Radionuclide Release from MCCI and Core Catchers

Description:

During its interaction with concrete or core catcher materials, corium composition varies. So

there may be some new volatile species that could transport the fission products out of the pool. The

presence of an oxidizing atmosphere can also modify the releases. This affects not only the

radiological source term but also the pool decay heat.

References:

M. Mignanelli, MCCI Chemistry and Properties, ACEX-TR-C22, EPRI 1998

S.V. Bechta et al., Influence of corium oxidation on fission product release from molten pool, Nucl

Eng Des 240, 1229-1241, 2010


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3.5.26 P5-26 - Core Catchers with External Cooling

Description:

In this case, the corium is collected in a core-catcher that is cooled at its boundaries, through a

vessel, and possibly by top-flooding. The major issue with this method is to guarantee that critical

heat fluxes will not be attained, which requires both an optimization of the external heat transfer and a

knowledge of the heat fluxes from the melt pool.

References:

M. Fischer, O. Herbst and H. Schmidt, Demonstration of the heat removing capabilities of the EPR

core catcher, Nucl Eng Des, 235 (2005) 1189-1200.

M. Farmer, R. Aeschlimann, D.J. Kilsdonk and S. Lomperski, Water Cooled Basemat (WCB-1) test

Investigating Core Melt Stabilization using a Water Cooled Surface, OECD/NEA MCCI Seminar,

Cadarache 2010


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3.5.27 P5-27 - Oxidation of Corium

Description:

During melt relocation (ejection from the vessel), the oxidation for both zirconium and iron (and

likely other metals) can be very strong and nearly complete depending on the melt ejection conditions

(with or without water in the pit). Once frozen, only zirconium oxidation should lead to a possible re-

escalation in case of refolding.

References:

Tsurikov, D., “MASCA2 Project Major Studies and Results, MASCA2 Seminar 2007 Proceedings,

Cadarache, France, 11-12 Oct 2007.

D. H. Cho, D. R Armstrong, W. H. Gunther, “Experiments on Interactions Between Zirconium-

Containing Melt and Water”, October 1998, NUREG/CR-5372

Prepared by: R. Meignen (IRSN) and P.M. Mathew (AECL)

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3.5.28 P5-28 - Corium Attack of Metallic Liner

Description:

In BWRs, there is a metallic liner at the bottom of the dry well (usually covered by some

concrete). In the case of vessel melt through, the core melt could spread until reaching the vertical

liner and interact with it.

References:

W.H. Amarasooriya, H. Yan, U. Ratnam, T.G. Theophanous, The probability of liner failure in a

Mark-1 containment, Part III: corium/concrete interactions and liner attack phenomena, Nucl.

Technol, 101 (1993) 354-384


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3.5.29 P5-29 - Corium Release from Failed Flooded Reactor Pressure Vessel

Description:

In case of an external vessel cooling strategy, low probability scenarios exist in which the in-

vessel core retention fails and corium pours out of the vessel into a flooded cavity.

FCI experiments done to date have not taken this configuration into account (no jet in air before

arrival in water), especially in the case of a vessel rupture near the lower head equator (small space

between leak and pit walls, presence of thermal shields), or at a tubing joint (presence of a bundle of

tubes).

References:

H. Esmaili, M. Khatib-Rahbar, Analysis of likelihood of lower head failure and ex-vessel fuel coolant

interaction energetics for AP1000, Nucl. Eng. Des., 235, 1583-1605, 2005.


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3.6 Systems Phenomena

3.6.1 P6-1 - Ventilation Systems

Description:

Simulation of flow through air ventilation systems is an important topic. A detailed modelling of

ventilation systems installed between different rooms is needed for detailed containment analyses and

analyses predicting the source term. Often the systems contain filters for aerosol or iodine retention.

Active components (fans) are typically installed in air ventilation systems of the containment or

the annulus. The fan start-up or coast-down and imparted momentum (head) to the gas/atmosphere

may be important. It is important to look at the relative humidity and temperature in the ventilation

system as this may lead to a failure of the fan or the filters. The interest is on experiments related to:

Behaviour of active ventilation systems

Flow through (fan off) passive systems including buoyant flow through the stack

Failure of ventilation ducts and leakage from ventilation system

Behaviour including failure criteria for fans, back flaps and filters

Fission product transport through ventilation systems

References:

German PSA guidelines published by the Federal Office for Radiation Protection (BfS):

Bundesamt für Strahlenschutz (BfS), Methoden zur probabilistischen Sicherheitsanalyse für

Kernkraftwerke, Facharbeitskreis Probabilistische Sicherheitsanalyse für Kernkraftwerke, BfS-

SCHR-37/05, August 2005


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3.6.2 P6-2 - Behaviour of Doors, Burst Membranes, Rupture Discs etc.

Description:

Component behaviour is important for containment scenario calculation. A detailed modelling

of the behaviour of doors, burst membranes and rupture discs (for instance, installed between different

rooms) is needed for detailed containment analyses and analyses predicting the source term. The

expected information from experiments is mainly the failure pressure and time to fully open (if

possible) the flow path area in case of a failure of the components mentioned.

References:

German PSA guidelines published by the Federal Office for Radiation Protection (BfS):

Bundesamt für Strahlenschutz (BfS), Methoden zur probabilistischen Sicherheitsanalyse für

Kernkraftwerke, Facharbeitskreis Probabilistische Sicherheitsanalyse für Kernkraftwerke, BfS-

SCHR-37/05, August 2005


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3.6.3 P6-3 - Air Cooler (Fan Cooler) Heat Transfer

Description:

A local air cooler (LAC, sometimes referred to as air cooler unit or fan cooler) is a gas / vapour

mixture to liquid water heat exchanger. An air/steam mixture is drawn through the heat exchanger by

a fan. The air/steam mixture may enter and leave the LAC either through a duct, or without a duct.

Heat and mass (steam) removal from the air/steam mixture by a LAC is governed by several factors,

namely heat exchanger design, cooling water supply temperature and flow rate, steam concentration,

temperature, pressure and flow conditions of the air steam mixture, and fouling on the heat exchanger.

References:

Not required


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3.6.4 P6-4 - Pump Performance including Sump Clogging (No Experiments)

Description:

Pumps are typically installed in emergency core cooling systems of the reactor system. Such

pumps typically take the water out of storage tanks or pools which are part of the containment (for

instance, PWR containment sump or BWR wetwell). The pump start-up or coast-down and imparted

momentum (head) to the fluid is of importance for the RCS. Pump performance is influenced by

vapour and temperature in the flow. Conditions stronger than the pump design my cause cavitation of

pumps.

Fibres from insulation material solved in the water may cause pump failure as well by sump

clogging phenomena. The phenomena includes the blockage of filters in the sump upstream the pump

system so that the pressure difference across the filters increase.

For the pool inside the containment drawing of water from a water pool will create a vortex

(whirlpool). If the pool is not deep enough, pump can draw in gases. As well often pump bypass

flows (in BWR) are going back by a spray flow into the pool. Information related to phenomena

described may result from various experiments.

This issue is being addressed by a separate CSNI task group activity on preparation of a “State-

of-the-Art Report on the Knowledge Base for Emergency Core Cooling System Recirculation

Reliability”. This work will provide a more detailed description of the problem and the research that

has been carried out. Thus, this CSNI-CCVM report does not include any experiments suitable for

validation of this phenomenon.

References:

Not required


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3.6.5 P6-5 - Passive Cooling by Internal and External Condensers

Description:

Certain reactors are equipped with internal condensers. New passive reactor designs are also

equipped with internal or external condensers. External condensers are immerged in water pools.

The heat removal capability of these units during the long-term containment response is controlled by

phenomena that are less relevant for the containment thermal-hydraulics of conventional reactors:

External (immerged) condensers: Performance is affected by small amounts of non-

condensable gases flowing into the tubes. In fact, due to the small volumes of the pipes, a

very small concentration of gases in the drywell could lead in principle to a large gas fraction

in a section of the tubes, with the reduction of the active heat transfer area. In extreme cases,

the gases could even block the flow in the tubes with complete loss of the heat removal

capability.

Although tests in large-scale facilities suggest that this cannot occur in the more advanced

ESBWR design, in general these phenomena should always be considered. Return of gases

from the wetwell (in the case of vacuum breaker opening) and release of gas trapped in other

compartments has therefore to be carefully evaluated. A special feature of these units is that

the heat transfer rate is auto-regulating in the presence of air, in the sense that the active heat

transfer area adjusts itself to the decay heat. The function of the condensers is also affected

by the process of vent clearing of additional low-submergence vent paths between the drywell

and the suppression pool, which must clear for smaller pressure differences than those forcing

the flow through the main vents, to avoid steam bypass. In the ESBWR, for instance, the

venting of the non-condensable gases through the vent pipe of the passive condensers plays

an important role in the performance of this system.

During the long-term cooling period, depending on various conditions, gas is vented

continuously or intermittently to the suppression pool whenever the pressure difference

between drywell and suppression pool is higher than the hydrostatic head between the vent

exit and the water pool surface. By design, the vented fluid is a steam-lean mixture. Under

certain conditions (condenser overload or in presence of hydrogen during a BDBA), however,

also large amounts of steam could be vented to the pool (steam bypass).

Condensation rates and pool mixing are affected by the composition of the vented mixture.

Stratification builds-up only when the amount of non-condensables in the vented mixture is

small. Moreover, in presence of a light gas (hydrogen in a BDBA scenario), a complex

internal circulation of gases between the tubes could lead to severe deterioration of the global

heat removal capability, and dumping of steam to the wetwell (condenser bypass). This

circulation (with reverse gas flow in some tubes) is driven by small density differences and

accumulation of light gas due to condensation.

Heat transfer modes usually not considered for containment thermal-hydraulics exist on both

sides of the tubes. On the primary side, condensation in pipes with non-condensables is

somewhat different from condensation on structures in large volumes and requires specific

models. On the secondary side, boiling heat transfer under natural circulation conditions

controls the performance of the condenser. CHF (phenomenon P1-4) is a limiting mechanism

for heat transfer. In the case of prolonged heat removal from the containment, bulk boiling

occurs in the water reservoirs and post-dryout heat transfer has to be considered on the dry

portion of the pipes during the boil-off transient.

Internal condensers: In this case, water is flowing inside the pipes. However, if during the

accident pressure (and temperature) inside the drywell becomes sufficiently high, boiling

occurs inside tubes, and CHF is the limiting mechanism of heat transfer. Large instabilities

NEA/CSNI/R(2014)3

203

can develop in the circuits coupled to water reservoirs and strong condensation loads can

develop. For intermediate thermal loads, flashing occurs in the upper section of the return

line, which triggers an unstable behaviour. If the tubes are inclined, additional effects

influence heat transfer. On the secondary side (drywell), condensation processes on the outer

wall of the pipes is very complex. In fact, the flow field inside and around the tube bundle is

strongly coupled with temperature and concentration fields, and the global condensation rate

depends on the local gas concentration. Under these conditions, for horizontal or inclined

condensers, a sharp stratification front can develop at some elevation between tube rows,

producing inactive heat transfer areas. In the case of BDBA, the operation of the condenser is

aided (and further complicated) by the venting of hydrogen to the suppression pool.

Additionally, the condensate liquid is in the form of droplets, films and bridges (water

curtains) between adjacent tube rows. Finally, in some designs the use of finned tubes further

complicates the phenomena controlling heat transfer.

References:


Kasahara, F., M. Heitsch, M., Komen, E., Moretti, F., T. Morii, T., Mühlbauer, P., Smith, B.L., .

Watanabe, T. “Extension of CFD Codes Application to Two-Phase Flow Safety Problems - Phase 2”,

Report NEA/CSNI/R(2010)2, pp. 65-73, July 2010.

Alamgir, MD, Marquino, W. Jesus Diaz-Quiroz, J..and Tucker, L “ESBWR Long Term Containment

Response to Loss of Coolant Accidents”, Paper 10370, Proceedings of ICAPP ’10, San Diego, CA,

USA, 2010 June 13-17.

J. Hart, J., W.J.M. Slegers, W.J.M., de Boer, S.L., Huggenberger, M., Lopez Jimenez, J. Munoz-Cobo

Gonzalez, J.M., Reventos Puigjaner, F. “TEPSS - Technology Enhancement for Passive Safety

Systems”, Nuclear Engineering and Design, Vol. 209, pp. 243–252, 2001.

Bandurski, T., Huggenberger, M., Dreier, J., Aubert, C., Putz, F., Gamble, R.E., Yadigaroglu, G.,

“Influence of the distribution of noncondensibles on passive containment condenser performance in

PANDA”, Nucl. Eng. Design, Vol. 204, pp. 285-294, 2001.

Dreier, J., Paladino, D. Huggenberger, M., Andreani, M., Yadigaroglu, G., “PANDA: a Large Scale

Multi-Purpose Test Facility for LWR Safety Research”, 8th Int. Conf. on the Physics of Reactors

(PHYSOR’08), Paper 517, Interlaken, Switzerland, 2008 September 14-19.

Paladino, D., Auban, O., Huggenberger, M., Dreier, J. “A PANDA Integral Test On The Effect Of

Light Gas On A Passive Containment Cooling System (PCCS)”, Nuclear Engineering and Design ,

Volume 241, Issue 11, November 2011, pp. 4551-4561.

Leyer, S., Maisberger, F., Herbst, V., Doll, M., Wich, M., Wagner, T., “Status of the full scale

component testing of the KERENA Emergency Condenser and Containment Cooling Condenser”

Paper 10257, Proceedings of ICAPP ’10, San Diego, CA, USA, June 13-17, 2010.

D. Paladino and J. Dreier, “PCCS response with DGRS activated during a postulated LOCA”,

Nuclear Engineering and Design, Volume 241, Issue 9, September 2011, pp. 3925-3934.

D. Paladino and J. Dreier, “Passive Containment Cooling System (PCCS) response with Drywell Gas

Recirculation System (DGRS) activated during a severe accident scenario with release of non-

condensable gas”, Nuclear Engineering and Design, Volume 247, June 2012, pp. 212-220.


NEA/CSNI/R(2014)3

204

3.6.6 P6-6 - Aerosol Removal in EFADS

Description:

A final protection against radioactive material released from reactors under accident conditions is

to filter the gas stream. Three types of filters are commonly used: fibre filters, venturi scrubbers, and

gravel bed filters.

Fibre filters remove particles by trapping them and the overall collection efficiency depends

strongly on the particle size. The characteristics of the filter, like collection efficiency and pressure

drop, change as mass is collected and reduces the porosity of the filter. Venturi scrubbers

precondition the aerosol by injecting water (about a liter of water per cubic meter of gas) along with

gas through a constricting throat (gas velocity reaching velocities as high as 120 m/s). The liquid

water is “atomised” into small droplets at the high velocities. Water droplets can coagulate with the

aerosol or otherwise capture the aerosol particles. Gravel beds remove particles by deposition onto

the large surface area provided by the gravel. Submerged gravel beds use water to wash away

deposited materials from the gravel.

References:





NEA/CSNI/R(2014)3

205

4 EXPERIMENTS

A basic set of experiments were identified by the CSNI-CCVM participants to cover most of the

containment phenomena. No qualification regarding the suitability of the experiments for validation

is given. It is the responsibility of the validator to assess the suitability of the experiment for

validation purposes.

An experimental synopsis was written for each test. The information in the synopsis is described

in Table 4-1. The information is provided in a combination of the following tables and a separate

section for each test (Sections 4.1 to 4.6):

Table 4-2 - Containment Thermalhydraulics Experiments

Table 4-3 - Hydrogen Behaviour (Combustion, Mitigation and Generation) Experiments

Table 4-4 - Aerosol and Fission Product Behaviour Experiments

Table 4-5 - Iodine Chemistry Experiments

Table 4-6 - Core Melt Distribution and Behaviour in Containment Experiments

Table 4-7 - Systems Experiments

Table 4-1

List of Information Provided for Each Experiment

Test Number and

Name

Test Number and Name (used as an identifier in this CNSI-CCVM

report)

Availability Availability of the experimental data:

Open: Available in open literature

OECD: Available to OECD members

Closed: Available by bi-lateral agreement

N/A: Not available

OTHER: Comments are provided to define availability of

experiments.

Phenomena covered in

Experiment

Phenomena that occur within the test.

Type of test The type of tests:

SE: Separate effect test

COM: combined effects test (more than one phenomena)

INT: integral test (entire system)

DBA Does this test cover DBA conditions? (Yes or No)

SA Does this test cover SA/BDBA conditions? (Yes or No)

NEA/CSNI/R(2014)3

206

Test Number and

Name

Test Number and Name (used as an identifier in this CNSI-CCVM

report)

3D Identify if the code is suitable for validation of:

CFD

Lumped

Codes.

The main criterion for an experiment to be suitable for CFD validation is

if there is sufficient instrumentation to capture the spatial variation of the

important parameters.

Test Facility Name of the Test Facility

Owner Organization Identify the organization that owns the experimental data

Experimental

Description

A brief concise description of the experiment. This is limited to a single

test (each test on a separate line).

References for

Experiment

References to documents that will provide more detailed information

regarding the experiment (include both internal and/or open reports).

Range of Key

Experimental Parameters

List of key parameters and the ranges covered in the test.

Year Tests Performed Year the test was performed

Repeatability Check Was the repeatability of the experiment demonstrated?

Yes,

No, or

N/A (not applicable)

Past Code Validation/

Benchmarks

Identify any past code validation or benchmarks performed with this test.

Where possible, references to validation/benchmark reports are also

provided.

Prepared by Name of the person (and their organization) that prepared this

experimental synopsis.

NEA/CSNI/R(2014)3

207

Table 4-2

Containment Thermalhydraulics Experiments

Test Number and Name Availability Language Phenomena Covered in Experiment Type of

Test DBA

SA/

BDBA 3D

E1-1 - Flow through Interconnected

Vessels

OECD English /

Electronic

P1-20 - Turbulent Flow

P1-21 - Critical Flow (Choked Flow)

COM No No Lump

E1-2 - Bruce LAC Test in Air, Test

No. 50

OECD English /

Electronic

P6-3 - Air Cooler (Fan Cooler) Heat Transfer

SE Yes No Lump

E1-3 - LSGMF GMBT001 OECD English P1-1 - Stratification



COM No No CFD

E1-4 - LSGMF GMUS001 OECD English P1-1 - Stratification



COM No No CFD

E1-5 - AECL-SP Dousing Test No.

1

OECD English P1-11 - Heat Removal by Dousing

SE Yes No CFD

E1-6 - FIPLOC F2 Closed

(available on

bilateral

agreement)

English P1-1 - Stratification

P1-5 - Heat Conduction in Solids

P1-6 - Convection Heat Transfer (Natural and

Forced)

P1-9 - Condensation on Surfaces


INT Yes Yes CFD

NEA/CSNI/R(2014)3

208

Table 4-2



Test DBA

SA/

BDBA 3D

E1-7 - VANAM M3 (ISP-37) OECD English P1-1 - Stratification


Forced)



P3-7 - Condensation on Aerosols

P3-12 - Gravitational Settling (Drop Settling)

P3-13 - Diffusional Deposition

P3-19 - Radionuclide Transport

INT Yes Yes Lump

E1-8 - EREC LB LOCA Test 1 EU, EREC English P1-5 - Heat Conduction in Solids


Forced)


P1-10 - Pool Surface Evaporation and Condensation

P1-13 - Direct Contact Condensation


P1-23 - Vent Clearing

P1-24 - Pool Swell / Air Injection

P1-26 - Liquid Film Flow

P1-29 - Heat and Mass Transfer of Spray Droplets

(Dousing)

P6-2 - Behaviour of Doors, Burst Membranes,

Rupture Discs etc.

INT Yes No Lump

NEA/CSNI/R(2014)3

209

Table 4-2



Test DBA

SA/

BDBA 3D

E1-9 - EREC LB LOCA Test 5 EU, EREC English P1-5 - Heat Conduction in Solids


Forced)









(Dousing)


Rupture Discs etc.

INT Yes No Lump

E1-10 – EREC MSLB Test 7 EU, EREC English P1-5 - Heat Conduction in Solids


Forced)







Rupture Discs etc.

INT Yes No Lump

NEA/CSNI/R(2014)3

210

Table 4-2



Test DBA

SA/

BDBA 3D

E1-11 - EREC MSLB Test 9 EU, EREC English P1-5 - Heat Conduction in Solids


Forced)







Rupture Discs etc.

INT Yes No Lump

E1-12 - EREC SLB G02 EREC & GRS English P1-5 - Heat Conduction in Solids


Forced)







Rupture Discs etc.

INT Yes No Lump

E1-13 - HDR V44 (ISP-16) Open

(ISP-16)

English P1-2 - Flashing (Flashing Discharge)


Forced)


INT Yes No Lump

NEA/CSNI/R(2014)3

211

Table 4-2



Test DBA

SA/

BDBA 3D

E1-14 - HDR T31.5 (ISP-23) Open

(ISP-23)



Forced)

INT Yes Yes Lump

E1-15 - HDR E11.2 (ISP-29) Open

(ISP-29)

English /

German

P1-1 - Stratification


Forced)



INT Yes Yes CFD

E1-16 - HDR E11.4 OPEN English /

German


Forced)



INT Yes Yes CFD

E1-17 - GKSS M1 Closed German P1-5 - Heat Conduction in Solids


Forced)






INT Yes No Lump

NEA/CSNI/R(2014)3

212

Table 4-2



Test DBA

SA/

BDBA 3D

E1-18 - MISTRA ISP-47 OECD English P1-6 - Convection Heat Transfer (Natural and

Forced)




COM Yes Yes CFD

E1-19 - MISTRA M7 Closed

(Avail to

SARNET

participants)

English P1-6 - Convection Heat Transfer (Natural and

Forced)




COM Yes Yes CFD

E1-20 - MISTRA-M8 Closed

(Avail to

SARNET

participants)


Forced)




COM Yes Yes CFD

E1-21 - MISTRA-MASP Closed English P1-29 - Heat and Mass Transfer of Spray Droplets

(Dousing)

COM Yes Yes CFD

E1-22 - NUPEC M-7-1 (ISP-35) Open

(ISP-35)

English P1-5 - Heat Conduction in Solids


Forced)



INT No Yes CFD

NEA/CSNI/R(2014)3

213

Table 4-2



Test DBA

SA/

BDBA 3D

E1-23 - NUPEC M-8-2 Open English P1-14 - Momentum Induced Mixing in Gases


INT No Yes Lump

E1-24 - PANDA ISP-42, Phase A OECD English P1-1 - Stratification

P1-3 - Boiling Heat and Mass Transfer





INT Yes No CFD

E1-25 - PANDA ISP-42, Phase C OECD English P1-3 - Boiling Heat and Mass Transfer




INT Yes No CFD

E1-26 - PANDA ISP-42, Phase E OECD English P1-3 - Boiling Heat and Mass Transfer





INT N/A Yes CFD

E1-27 - PANDA ISP-42, Phase F OECD English P1-1 - Stratification






INT N/A Yes CFD

NEA/CSNI/R(2014)3

214

Table 4-2



Test DBA

SA/

BDBA 3D

E1-28 - PANDA BC4 OECD English P1-1 - Stratification




P6-5 - Passive Cooling by Internal and External

Condensers

INT N/A Major CFD

E1-29 - SVUSS G02 Closed English P1-5 - Heat Conduction in Solids


Forced)






INT Yes NO Lump

E1-30 - THAI TH1 Closed English P1-6 - Convection Heat Transfer (Natural and

Forced)


SE /

COM

Yes Yes CFD

E1-31 - THAI TH2 Closed English P1-1 - Stratification


Forced)



INT Yes Yes CFD

NEA/CSNI/R(2014)3

215

Table 4-2



Test DBA

SA/

BDBA 3D



Forced)


INT Yes Yes CFD



Forced)


INT Yes Yes CFD

E1-34 - THAI TH13 (ISP-47) Open

(ISP-47)



Forced)



INT Yes Yes CFD

E1-35 - THAI HM2 OECD English P1-1 - Stratification


Forced)



INT Yes Yes CFD

E1-36 - TOSQAN ISP-47 OECD: All data

Open: some

data published

in NED


Forced)



COM Yes Yes CFD

NEA/CSNI/R(2014)3

216

Table 4-2



Test DBA

SA/

BDBA 3D

E1-37 - TOSQAN Condensation

Tests

Closed French P1-6 - Convection Heat Transfer (Natural and

Forced)



COM Yes Yes CFD

E1-38 - TOSQAN Test 113 Closed

(Available to

SARNET

participants)

English and

French

P1-28 - Gas Entrainment by Spray Droplets

(Dousing)

P1-31 - Mixing by Sprays

SE Yes Yes CFD

E1-39 - TOSQAN Spray Tests Closed

(Available by

bilateral

agreement)

Test TOSQAN-

101 is available

to SARNET

members

French and

some in

English


(Dousing)


SE Yes Yes CFD

E1-40 - University of Wisconsin

Flat Plate Condensation Tests

OPEN English P1-9 - Condensation on Surfaces

SE Yes Yes CFD

E1-41 - CONAN SARNET

Benchmark No. 1

Open to

SARNET

members

English P1-9 - Condensation on Surfaces

SE Yes Yes CFD

E1-42 - CONAN SARNET2

Benchmark No. 2

Open to

SARNET2

members

English P1-9 - Condensation on Surfaces

SE Yes Yes CFD

NEA/CSNI/R(2014)3

217

Table 4-2



Test DBA

SA/

BDBA 3D

E1-43 - CSTF Tests OTHER

(signatories to

the LACE

consortium)

English P3-4 - Thermophoresis

P3-5 - Diffusiophoresis


P3-8 - Gravitational Agglomeration

P3-9 - Diffusional Agglomeration

INT Yes Yes Lump

E1-44 - Marviken Test 18 OECD

(ISP-17)








COM Yes Yes Lump

E1-45 - CARAIDAS EVAP and

COND tests

9 tests (from 36)

made available

within

SARNET-2 and

are available in

open literature

English for

the 9 tests,

French for

the others


(Dousing)

SE Yes Yes CFD

E1-46 - TOSQAN sump tests Closed

Some tests

published

partially

French for

the test

reports.

Several

publications

in English





COM Yes Yes CFD

NEA/CSNI/R(2014)3

218

Table 4-2



Test DBA

SA/

BDBA 3D

E1-47 - CALIST PWR spray test Closed

(1 test could be

made available

within

SARNET-2)

French for a

PhD thesis,

English for

SARNET-2

reports


(Dousing)

SE Yes Yes CFD

E1-48 - MISTRA LOWMA OECD English P1-14 - Momentum Induced Mixing in Gases


SE No

Yes CFD

E1-49 - PANDA OECD/SETH

tests

OECD English P1-1 - Stratification


Forced)




COM Yes Yes CFD

E1-50 - PANDA OECD/SETH-2 Closed

(Open to SETH-

2 participants)



Forced)




P1-18 - Mass Diffusion in Vapour


(Dousing)


(Dousing)


COM Yes Yes CFD

NEA/CSNI/R(2014)3

219

Table 4-2



Test DBA

SA/

BDBA 3D

E1-51 - CYBL Boiling Tests Papers: open

Report:

Unknown

Data: Unknown

English /

Papers and

Reports


P1-4 - Critical Heat Flux (CHF)


Forced)


P1-2 - Flashing (Flashing Discharge)5


COM No Yes Not

Assessed

E1-52 - ULPU CHF Tests Papers: open

Test Reports

and Data:

Other6

English




Forced)


P1-2 - Flashing (Flashing Discharge)

COM /

INT

No Yes CFD

E1-53 - SULTAN CHF Tests Closed

English/

papers



SE No Yes CFD

5 This phenomenon may not be covered in this experiment.

6 ULPU CHF test reports and data can be obtained by contacting T. G. Theofanous, Professor of Chemical Engireering, Professor of Mechanical Engineering,

University of California Santa Barbara, Director, Center for Risk Studies and Safety.

NEA/CSNI/R(2014)3

220

Table 4-2



Test DBA

SA/

BDBA 3D

E1-54 - SBLB Boiling Tests Papers: open

Reports: some

difficult to

obtain

Information

looks scattered,

incomplete and

confusing. Test

matrix not

available

English/

papers and

one report




Forced)


COM No Yes Lump

E1-55 – Small Scale Burst Test

Experiments

Closed English P1-16 - Pressure Wave Propagation in Water

COM /

INT

No Yes CFD

NEA/CSNI/R(2014)3

221

Table 4-3

Hydrogen Behaviour (Combustion, Mitigation and Generation) Experiments


Test DBA

SA/

BDBA 3D

E2-1 - LSVCTF S01 OECD English P2-1 - Deflagration

SE Yes Yes CFD

E2-2 - LSVCTF S03 OECD English P2-1 - Deflagration

SE Yes Yes CFD

E2-3 - BMC Hx series Closed English /

German

P2-1 - Deflagration

P2-2 - Hydrogen Flame Acceleration (FA)

P2-10 - Hydrogen Mitigation by Hydrogen Ignitors

COM Yes Yes Lump

E2-4 - BMC Ix series Closed English P2-1 - Deflagration


COM Yes Yes Lump

E2-5 - BMC Gx Series Closed

(available on

bilateral

agreement)

English P2-1 - Deflagration

P2-8 - Hydrogen Mitigation - Passive Autocatalytic

Recombiners


COM Yes Yes Lump

E2-6 - BMC Kx Series Closed

(available on

bilateral

agreement)



COM Yes Yes CFD

E2-7 - BMC Ex Series Closed

(available on

bilateral

agreement)



COM Yes Yes CFD

E2-8 - ENACEFF SARNET2

Tests

OECD

(SARNET2)



SE No Yes CFD

NEA/CSNI/R(2014)3

222

Table 4-3



Test DBA

SA/

BDBA 3D

E2-9 - ENACEFF SARNET Test

(Run 703)

OECD

(SARNET)

French P2-1 - Deflagration


SE No Yes CFD

E2-10 - ENACEFF SARNET Test

(Run 717)

OECD

(SARNET)

French P2-1 - Deflagration


SE No Yes CFD

E2-11 - ENACEFF Run 765 (ISP-

49)

OECD (ISP-49) English P2-1 - Deflagration


SE No Yes CFD


49)



P2-6 - Quenching

SE No Yes CFD


49)



SE No Yes CFD

E2-14 - DRIVER HYCOM MC

003

Open English P2-1 - Deflagration


P2-6 - Quenching

COM No Yes CFD


012



COM No Yes CFD

E2-16 - FZK R 0498_09 Open English P2-1 - Deflagration


P2-4 - Hydrogen Detonation

COM No Yes CFD

NEA/CSNI/R(2014)3

223

Table 4-3



Test DBA

SA/

BDBA 3D


043



COM No Yes CFD

E2-18 - DRIVER HYCOM HC

020



COM No Yes CFD

E2-19 - DRIVER HYCOM-

HC027

Close English P2-1 - Deflagration


COM No Yes CFD

E2-20 - RUT HYC01 Open English P2-1 - Deflagration


COM No Yes CFD

E2-21 - RUT HYC12 Close English P2-1 - Deflagration


COM No Yes CFD

E2-22 - RUT HYC14 Open English P2-1 - Deflagration


COM No Yes CFD

E2-23 - VGES Tests Open English P2-1 - Deflagration


Yes Yes Lump

E2-24 - NTS Tests Open English P2-1 - Deflagration


P2-14 - Effect of Droplets on Hydrogen

Combustion

Yes Yes Lump

E2-25 - PET Tubes Open English P2-1 - Deflagration


SE Yes Yes CFD

NEA/CSNI/R(2014)3

224

Table 4-3



Test DBA

SA/

BDBA 3D

E2-26 - THAI HD Series

(Combustion Tests)

OECD English P2-1 - Deflagration SE Yes Yes CFD

E2-27 - THAI HR Series (PAR

Tests)

OECD English P2-1 - Deflagration

P2-8 - Hydrogen Mitigation - Passive Autocatalytic

Recombiners

P2-9 - Hydrogen Ignition by PARs

P4-12 - Decomposition of Iodides (CsI) by Heat-up

in PARs

COM Yes Yes CFD

E2-28 - THAI Hydrogen

Combustion During Spray

Operation

BMWi / OECD English P2-14 - Effect of Droplets on Hydrogen

Combustion

SE No Yes CFD

E2-29 - DFF SFSER01 Closed English P2-7 - Hydrogen Diffusion Flame (Standing Flame)

SE Yes Yes CFD

E2-30 - LSVCTF S02 Closed English P2-1 - Deflagration

SE Yes Yes CFD

E2-31 - LSVCTF DC Closed


P2-12 - Jet Ignition of Hydrogen

SE Yes Yes CFD

E2-32 - LSVCTF 3C Closed



SE Yes Yes CFD

E2-33 - LSVCTF CIC Closed

English P2-7 - Hydrogen Diffusion Flame (Standing Flame)


SE Yes Yes CFD

E2-34 - Gammacell Radiolysis

Tests

Closed English P2-13 - Radiolysis (Hydrogen Production by Water

Radiolysis)

SE Yes Yes Lump

NEA/CSNI/R(2014)3

225

Table 4-3



Test DBA

SA/

BDBA 3D

E2-35 - LACOMECO UFPE2 Open for

SARNET2

community



SE No Yes CFD

E2-36 - LACOMECO

HYGRADE10

Open for

SARNET

community



SE Yes Yes CFD

E2-37 - LACOMECO

HYGRADE09

Open for

SARNET

community



SE Yes Yes CFD

E2-38 - LACOMECO

HYGRADE03

Open for

SARNET

community



SE Yes Yes CFD

E2-39 - LACOMECO HYDET06 Open for

SARNET

community




SE No Yes CFD

E2-40 - LACOMECO HYDET07 Open for

SARNET

community




SE No Yes CFD

E2-41 - H2PAR E 12 OECD English P2-8 - Hydrogen Mitigation - Passive Autocatalytic

Recombiners

P2-9 - Hydrogen Ignition by PARs

SE No Yes CFD

E2-42 - H2PAR E 13 OECD French P2-8 - Hydrogen Mitigation - Passive Autocatalytic

Recombiners

SE No Yes CFD

NEA/CSNI/R(2014)3

226

Table 4-3



Test DBA

SA/

BDBA 3D

E2-43 - H2PAR E 3 Closed French P2-8 - Hydrogen Mitigation - Passive Autocatalytic

Recombiners

SE No Yes CFD

E2-44 – KIT DDT Tests in

CHANNEL Facility

Open English P2-3 - Deflagration-to-Detonation Transition (DDT) SE No No CFD

E2-45 – KIT Jet Ignition Tests in

HPHR Facility

Open English P2-12 - Jet Ignition of Hydrogen SE No Yes CFD

E2-46 – KIT Geometric

Quenching of Detonation Tests in

the HYKA-A1 Facility

Open English P2-5 - Quenching of Detonations by Geometrical

Constrains

SE No Yes CFD

E2-47 – Cheikhravat Experiments

on Effect of Spray on Hydrogen

Combustion

Open French

(PhD

Thesis)

and

English

(conf

paper

P2-14 - Effect of Droplets on Hydrogen

Combustion

COM No Yes CFD

E2-48 – Bjerketvedt Experiments

on Effect of Spray on Hydrogen

Combustion

Unknown Unknown P2-14 - Effect of Droplets on Hydrogen

Combustion

Not

Assessed

Not

Assessed

Not

Assessed

Not

Assessed

NEA/CSNI/R(2014)3

227

Table 4-4

Aerosol and Fission Product Behaviour Experiments


Test DBA

SA/

BDBA 3D

E3-1 - AHMED OECD

benchmark

Rough data can be

extracted from the

Report

NEA/CSNI/R(95)23

More precise data

can be obtained

through a bilateral

agreement

English P3-7 - Condensation on Aerosols


COM No Yes Lump /

CFD (?)

E3-2 - KAEVER CsI series Closed (available on

bilateral agreement)

Test K123 was used

in ISP-44

German P1-6 - Convection Heat Transfer (Natural and

Forced)




SE /

COM

No Yes Lump

E3-3 - KAEVER K187 (ISP-44) OECD

(ISP-44)


Forced)




SE /

COM

No Yes Lump


(ISP-44)


Forced)




SE /

COM

No Yes Lump

NEA/CSNI/R(2014)3

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Table 4-4



Test DBA

SA/

BDBA 3D


(ISP-44)


Forced)




SE /

COM

No Yes Lump

E3-6 - LACE LA2 OTHER

SAND94-2166

(available)

LACE (TR-004,

007, 009, 010) –

needs to ask EPRI

English P3-5 - Diffusiophoresis




P3-10 - Turbulent Agglomeration of Aerosols


INT No Yes Lump

E3-7 - LACE LA4 OTHER

LACE TR-025 can

be obtained from

EPRI

English P3-5 - Diffusiophoresis






INT No Yes Lump

E3-8 – LACE LA5 and LA6 Unknown English P3-26 - Re-entrainment (Wet)

SE No Yes Lump

NEA/CSNI/R(2014)3

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Table 4-4



Test DBA

SA/

BDBA 3D

E3-9 - Phebus FPT-1 (ISP-46) OECD (ISP-46) English P1-9 - Condensation on Surfaces






P3-22 - Containment Chemistry Impact on Source

Term

P4-7 - Silver Iodine Reactions in the Water Phase

INT No Yes Lump

E3-10 - POSEIDON PA10 Closed English P3-18 - Pool Scrubbing

COM No Yes Lump

E3-11 - BMC VANAM M2 Closed English P1-1 - Stratification

P1-5 - Heat Conduction in Solids


Forced)




P3-19 - Radionuclide Transport

INT No Yes CFD

E3-12 - VICTORIA test 58 Reports including

data reports have

been requested by

NEA from VTT, EC

being informed.

English P3-7 - Condensation on Aerosols

COM No Yes Lump

NEA/CSNI/R(2014)3

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Table 4-4



Test DBA

SA/

BDBA 3D

E3-13 - CSTF ABCOVE Tests Open English P3-5 - Diffusiophoresis






INT No Yes Lump

E3-14 - CSTF ACE Other

(signatories to the

ACE Consortium)


P3-4 - Thermophoresis



P3-6 - Liquid Aerosol Evaporation


P3-15 - Turbulent Deposition of Aerosols

INT No Yes Lump

E3-15 - CARAIDAS Aerosol

washout by single droplet tests

several tests could

be made available

French P3-24 - Aerosol Removal by Sprays (Dousing) SE No Yes Lump

E3-16 - Whiteshell Flashing Jet

Tests

Closed English P3-1 - Aerosol Formation in a Flashing Jet




COM Yes No Lump

E3-17 - Clarkson College

Brownian Agglomeration

Open

English P3-9 - Diffusional Agglomeration

SE Yes No Lump

E3-18 - JAERI Thermophoresis

Tests

Open English P3-4 - Thermophoresis



SE Yes Yes Lump

NEA/CSNI/R(2014)3

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Table 4-4



Test DBA

SA/

BDBA 3D

E3-19 - PITEAS Diffusiophoresis

Tests (PDI 08, PDI 09, PDI 11 and

PDI 12)

Closed English P3-4 - Thermophoresis




COM Yes Yes Not

Assessed

E3-20 - PITEAS Aerosol

Condensation Tests (PCON 01 to

PCON 05)

Closed English P3-5 - Diffusiophoresis



COM Yes Yes Not

Assessed

E3-21 - Aerosol Deposition in

Turbulent Vertical Conduits

(Sehmel)

Open English P3-15 - Turbulent Deposition of Aerosols





P3-14 - Inertial Deposition of Aerosols (Also

called Impaction)

SE Yes ? Lump



(Forney)

Open English P1-20 - Turbulent Flow



COM No Yes

(but

some

test

variables

are out

of

typical

ranges)

Lump

NEA/CSNI/R(2014)3

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Table 4-4



Test DBA

SA/

BDBA 3D



(Friedlander)




COM No Yes Lump


Turbulent Vertical Conduits (Liu)




COM No Yes Lump



(Wells)





COM No Yes Lump

E3-26 - CSE Fission Product

Transport Tests

Open English P3-4 - Thermophoresis







COM No Yes Lump

E3-27 - CSE Aerosol Removal

Tests

Open English P3-17 - Aerosol Removal in Leakage Paths

P4-15 - Iodine Retention in Leakage Paths

SE NO YES Lump

E3-28 - LASS-SGTR Closed English /

electron

files


called Impaction)


P3-27 - Aerosol De-agglomeration

COM No Yes Lump

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233

Table 4-4



Test DBA

SA/

BDBA 3D

E3-29 - MCE, UCE and HCE

Tests

Closed

(limited data

available in open

literature)

English P3-21 - Release Rate Change Due to Oxidizing

Environment

SE Yes Yes Lump

E3-30 - GBI Tests Closed

(limited data

available in open

literature)

English P3-21 - Release Rate Change Due to Oxidizing

Environment

SE Yes Yes Lump

E3-31 - Aerosol Trapping Effects

in Containment Penetration (A.

Watanabe)

OECD English P3-17 - Aerosol Removal in Leakage Paths


SE NO YES Lump

E3-32 - Aerosol transfer through

cracked concrete walls

Open English P3-17 - Aerosol Removal in Leakage Paths SE No Yes Lump

E3-33 - Whiteshell Steam Jet

Experiments

Closed English P3-2 - Aerosol Formation in a Steam Jet SE Yes Yes Lump

E3-34 - WALE Closed English P3-3 - Aerosol Impaction




P3-1 - Aerosol Formation in a Flashing Jet


called Impaction)

COM Yes Yes Lump

E3-35 – AEREST (Aerosol

resuspension shock tube)

Unknown English P3-25 - Re-suspension (Dry) SE No Yes Lump

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Table 4-4



Test DBA

SA/

BDBA 3D

E3-36 – VANAM-M4 Closed German P3-12 - Gravitational Settling (Drop Settling)

P3-25 - Re-suspension (Dry)

COM No Yes Lump

E3-37 – THAI Aer-1, Aer-3 and

Aer-4 tests

Closed English P3-25 - Re-suspension (Dry) SE No Yes Lump

E3-38 – Phebus FPT4

Revaporization

Closed

(available to Phebus

FP participants)

English P3-16 - Re-volatilisation

INT No Yes Lump

E3-39 – Ruthenium

Revolatilisation Studies at VTT

Open English P3-16 - Re-volatilisation

P3-21 - Release Rate Change Due to Oxidizing

Environment

P3-23 - Ruthenium Volatility and Behaviour in

Containment

SE No Yes Lump




P3-21 - Release Rate Change Due to Oxidizing

Environment


Containment

SE No Yes Lump

E3-41 – Ruthenium deposition

studies at Chalmers University



Containment

SE No Yes Lump

E3-42 – Ruthenium

Revolatilisation Studies at IRSN



Containment

SE No Yes Lump

NEA/CSNI/R(2014)3

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Table 4-5

Iodine Chemistry Experiments


Test DBA

SA/

BDBA 3D

E4-1 - CFTF Charcoal Filter Test Open English P4-13 - Iodine Filtration

SE Yes Yes Lump

E4-2 - RTF P9T3 Open to

SARNET

members

English P4-1 - Aqueous Phase Oxidation and Reduction of

Iodine Species

P4-2 - Inorganic Iodine Hydrolysis

P4-3 - Inorganic Iodine Radiolysis in Water Phase

P4-4 - Homogeneous Organic Reactions in Water

Phase

P4-5 - Iodine Reactions with Surfaces in the Water

Phase

P4-6 - Iodine reactions with surfaces in the gas phase

P4-10 - RI (Organic Iodine) Radiolytic Destruction

P4-11 - Interfacial Mass Transfer

COM Yes Yes Lump

E4-3 - RTF P9T1 OECD (BIP)

Available to

OECD

members March

2014


Iodine Species




Phase


Phase




COM Yes Yes Lump

NEA/CSNI/R(2014)3

236

Table 4-5



Test DBA

SA/

BDBA 3D


Available to

OECD

members March

2014


Iodine Species




Phase


Phase




COM Yes Yes Lump


Available to

OECD

members March

2014


Iodine Species




Phase


Phase




COM Yes Yes Lump

NEA/CSNI/R(2014)3

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Table 4-5



Test DBA

SA/

BDBA 3D


Available to

OECD

members March

2014

P4-1 - Aqueous Phase Oxidation and Reduction of

Iodine Species




Phase


Phase




COM Yes Yes Lump


Available to

OECD

members March

2014

P4-1 - Aqueous Phase Oxidation and Reduction of

Iodine Species




Phase


Phase




COM Yes Yes Lump

NEA/CSNI/R(2014)3

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Table 4-5



Test DBA

SA/

BDBA 3D

E4-8 - RTF P0T2 OECD

(ISP-41)


Iodine Species




Phase


Phase




COM Yes Yes Lump

E4-9 - RTF P10T1 OECD

(ISP-41)


Iodine Species




Phase


Phase




COM Yes Yes Lump

NEA/CSNI/R(2014)3

239

Table 4-5



Test DBA

SA/

BDBA 3D

E4-10 - RTF PHEBUS RTF1 Open to

PHEBUS FP

members,

OECD

(ISP-41)


Iodine Species




Phase


Phase




COM Yes Yes Lump

E4-11 - EPICUR Test Series S1, S2

and S3

Closed

(available to

ISTP partners)

(data could be

opened to

others under

conditions)

English P4-3 - Inorganic Iodine Radiolysis in Water Phase


Phase


Phase


P4-9 - Homogeneous Organic Iodine Reactions in

Gas Phase


SE No Yes Lump

E4-12 - THAI Iod-09 Open for

SARNET

members

English P4-6 - Iodine reactions with surfaces in the gas phase


COM No Yes Lump


SARNET2

members

English P3-19 - Radionuclide Transport


INT No Yes Lump

NEA/CSNI/R(2014)3

240

Table 4-5



Test DBA

SA/

BDBA 3D


SARNET2

members

English P3-19 - Radionuclide Transport


INT No Yes Lump


SARNET2

members

English P4-8 - Gas Phase Radiolytic Oxidation of Molecular

Iodine (I2) (Iodine/Ozone Reaction)

COM No Yes Lump


SARNET2

members

English P4-8 - Gas Phase Radiolytic Oxidation of Molecular

Iodine (I2) (Iodine/Ozone Reaction)

COM No Yes Lump

E4-17 - THAI Iod-25 OECD

(THAI Project)

English P4-16 - I2 Interaction with Aerosols

COM No Yes Lump

E4-18 - THAI Iod-26 OECD

(THAI Project)

English P4-16 - I2 Interaction with Aerosols

COM No Yes Lump

E4-19 - THAI AW OECD

(THAI Project)

English P4-17 - Iodine Wash-down

SET No Yes CFD

E4-20 - THAI HR31 OECD

(THAI Project)

English P4-12 - Decomposition of Iodides (CsI) by Heat-up

in PARs

SET No Yes CFD

E4-21 - THAI HR32 OECD

(THAI Project)

English P2-8 - Hydrogen Mitigation - Passive Autocatalytic

Recombiners

SET No Yes CFD

E4-22 - LASS-GIRS DABASCO Closed English P4-1 - Aqueous Phase Oxidation and Reduction of

Iodine Species


P4-14 - Volatile Iodine Trapping by Airborne

Droplets

COM No Yes Lump

NEA/CSNI/R(2014)3

241

Table 4-5



Test DBA

SA/

BDBA 3D

E4-23 - OECD-THAI2 Gaseous

Iodine Release from Flashing Jet

Test

OECD

(THAI2

Project)

English P4-19 - Iodine Release from Flashing Pool or

Flashing Jet

INT Yes Yes CFD

E4-24 - CAIMAN 97/02 test OECD English P4-1 - Aqueous Phase Oxidation and Reduction of

Iodine Species




Phase


Phase



Gas Phase



INT No Yes Lump

NEA/CSNI/R(2014)3

242

Table 4-5



Test DBA

SA/

BDBA 3D

E4-25 - CAIMAN 2001/01 Test OECD English P4-1 - Aqueous Phase Oxidation and Reduction of

Iodine Species




Phase


Phase



Gas Phase



INT No Yes Lump

E4-26 – Iodine Clean-Up in a

Steam Suppression System

Open English P4-18 - Pool Scrubbing of Iodine INT No Yes Lump

NEA/CSNI/R(2014)3

243

Table 4-6

Core Melt Distribution and Behaviour in Containment Experiments


Test DBA

SA/

BDBA 3D

E5-1 - IET Experiments - Zion

Geometry

Open English P1-9 - Condensation on Surfaces



P2-1 - Deflagration

P5-4 - Corium Particles Generation from the Two

Phase Jet

P5-7 - Direct Containment Heating

P5-16 - Molten Corium Heat Transfer

P5-27 - Oxidation of Corium

COM No Yes Lump

E5-2 - IET Experiments - Surry

Geometry

Open English P1-9 - Condensation on Surfaces



P2-1 - Deflagration


Phase Jet




INT

No Yes Lump

NEA/CSNI/R(2014)3

244

Table 4-6



Test DBA

SA/

BDBA 3D

E5-3 - FARO Tests Other

(FARO-Test

L-14 is ISP-

39 and

therefore

OPEN.

Other

experiments

are available

to countries

within

European

Union;

other

countries

need

agreements)

English P5-8 - Corium Jet Break-up in Water Pool

P5-9 - FCI and Steam Explosion - Melt into Water

Ex-Vessel (Melt Quenching)

P5-10 - Pressure Load on Corium Retention

Devices

P5-11 - Particulate Debris Bed Formation

P5-15 - Corium Spreading

COM No Yes Lump

E5-4 - DISCO-C Tests Closed English P1-12 - Liquid Re-Entrainment (Resuspension)

P5-1 - Corium Release from Failed Dry Reactor

Pressure Vessel

P5-2 - Corium Entrainment Out of the Reactor

Primary Vessel with Lateral Breaches

P5-3 - Corium Particles Generation from the

Corium Pool


Phase Jet

P5-5 - Corium Particles Entrainment


COM No Yes Lump

NEA/CSNI/R(2014)3

245

Table 4-6



Test DBA

SA/

BDBA 3D

E5-5 - DISCO-H Tests Closed English P5-1 - Corium Release from Failed Dry Reactor

Pressure Vessel

P5-2 - Corium Entrainment Out of the Reactor

Primary Vessel with Lateral Breaches

P5-3 - Corium Particles Generation from the

Corium Pool


Phase Jet


P5-6 - Corium Particles Trapping


COM No Yes Lump

E5-6 - DISCO-A2 Open English P2-1 - Deflagration


P2-7 - Hydrogen Diffusion Flame (Standing Flame)


COM No Yes CFD

E5-7 - KROTOS JRC Tests Other

(experiments

are available

to countries

within

European

Union; other

countries

need

agreements)





Devices


COM No Yes Lump

NEA/CSNI/R(2014)3

246

Table 4-6



Test DBA

SA/

BDBA 3D

E5-8 - SERENA-2 KROTOS and

TROI Commissioning Tests

Open English P5-8 - Corium Jet Break-up in Water Pool




Devices


P5-12 - Corium Debris (Solid) Heat Transfer

COM No Yes Lump

E5-9: SERENA-2 KROTOS and

TROI Tests

SERENA

OECD

partners





Devices


P5-12 - Corium Debris (Solid) Heat Transfer

COM No Yes Lump

E5-10 - MCCI-1 Tests CCI Tests

1-3; SSWICS tests 1-7

Open English P5-13 - Molten Core Concrete Interaction


P5-17 - Corium Evaporation/Vaporization

P5-18 - Corium Solidification/Crust Formation

P5-19 - Cracking (Crust)

P5-20 - Ex-Vessel Corium Coolability, Top

Flooding

P5-21 - Ex-Vessel Corium Catcher - Coolability

and Water Bottom Injection

P5-24 - Crust Anchorage


INT / SE No Yes CFD/Lump

(CCI tests were

2-D)

NEA/CSNI/R(2014)3

247

Table 4-6



Test DBA

SA/

BDBA 3D

E5-11 - MCCI-2 Tests CCI Tests

4-6; SSWICS tests 8-13; WCB-1

Closed

(Data can be

released 12-

2014)

English P5-13 - Molten Core Concrete Interaction






Flooding




P5-26 - Core Catchers with External Cooling


INT / SE No Yes CFD/Lump

(CCI tests were

2-D)

E5-12 - ECO Tests Open English P5-9 - FCI and Steam Explosion - Melt into Water


P1-16 - Pressure Wave Propagation in Water


SE No Yes CFD

E5-13 - BALI Ex-Vessel Tests Closed English P5-13 - Molten Core Concrete Interaction


SE No Yes Lump

E5-14 - BALISE Tests Closed English P5-14 - Corium Melt Stratification

SE No Yes Lump


concrete)




P5-23 - Effect of Non Homogeneous Ablation on

Gate Ablation

COM No Yes Lump

NEA/CSNI/R(2014)3

248

Table 4-6



Test DBA

SA/

BDBA 3D

E5-16 - VULCANO VW-U1

(COMET bottom flooding)





Gate Ablation



COM No Yes Lump

E5-17 - VULCANO VE-U7 Other

(Published

More details

subject to

discussion)




Gate Ablation

COM No Yes Lump

E5-18 – SURC-1 and SURC-2 Open English P5-13 - Molten Core Concrete Interaction

P5-14 - Corium Melt Stratification




P5-25 - Radionuclide Release from MCCI and Core

Catchers


COM No Yes Lump

NEA/CSNI/R(2014)3

249

Table 4-6



Test DBA

SA/

BDBA 3D

E5-19 - SURC-3 Open English P5-13 - Molten Core Concrete Interaction






Catchers


SE No Yes CFD (1D)

E5-20 - SURC-3A Open English P5-13 - Molten Core Concrete Interaction






Catchers


SE No Yes CFD

(axisymmetric)

E5-21 - SURC-4 Closed

(Data not

available at

the NEA

Database.

Availability

to the NEA

Database to

be discussed.

(ISP-24))







Catchers


COM No Yes CFD (?)

NEA/CSNI/R(2014)3

250

Table 4-6



Test DBA

SA/

BDBA 3D

E5-22 - BETA V5.1 Open

ISP-30


SE No Yes Lump

E5-23 - ACE Phase C Tests L1,

L2, L4, L5, L6, and L7

Closed

(ACE Phase

C program

participants)







Catchers


INT No Yes Lump

E5-24 - MACE Tests M0, M1b,

M3b, M4, and MSET-1

Closed

(MACE

program

participants)







Flooding



Catchers


INT/SE No Yes CFD (1 test)

E5-25 - COLIMA CA-U4 Open English P3-14 - Inertial Deposition of Aerosols (Also called

Impaction)

P3-17 - Aerosol Removal in Leakage Paths

P3-18 - Pool Scrubbing


Catchers

INT No Yes Lump

NEA/CSNI/R(2014)3

251

Table 4-6



Test DBA

SA/

BDBA 3D

E5-26 - BURN-1 Open English P5-13 - Molten Core Concrete Interaction



SE No Yes CFD

(axisymmetric)

E5-27 – SWISS-1 and SWISS-2 Open English P5-20 - Ex-Vessel Corium Coolability, Top

Flooding

SE No Yes CFD (1D)

E5-28 – HSS-1 and HSS-3 Open English P5-12 - Corium Debris (Solid) Heat Transfer





Flooding



Catchers

SE No Yes CFD (1D)

E5-29 - TURC1T and TURC1SS Open English P5-13 - Molten Core Concrete Interaction






Catchers


SE No Yes CFD (1D)

NEA/CSNI/R(2014)3

252

Table 4-6



Test DBA

SA/

BDBA 3D

E5-30 – TURC2 and TURC3 Open English P5-13 - Molten Core Concrete Interaction






Catchers


INT No Yes CFD (1D)

E5-31 - LSL-1,2,3 Open English P5-13 - Molten Core Concrete Interaction






Catchers


INT No Yes CFD (3D)

E5-32 - LBL-1,2,3 Open English P5-13 - Molten Core Concrete Interaction






Catchers


INT No Yes CFD (3D)

NEA/CSNI/R(2014)3

253

Table 4-6



Test DBA

SA/

BDBA 3D

E5-33 - LSCRBR-1,2,3 Open English P5-13 - Molten Core Concrete Interaction






Catchers


INT No Yes CFD (3D)

E5-34 - COIL-1 Open English P5-13 - Molten Core Concrete Interaction






Catchers


INT No Yes CFD

(axisymmetric)

E5-35 - WETCOR-1 Open English P5-14 - Corium Melt Stratification


Flooding



Catchers

SE No Yes CFD (1D)

E5-36 - FRAG Open English P5-12 - Corium Debris (Solid) Heat Transfer



Catchers

SE No Yes CFD

(axisymmetric)

NEA/CSNI/R(2014)3

254

Table 4-6



Test DBA

SA/

BDBA 3D

E5-37 - 1DHtFlx Open English P5-16 - Molten Corium Heat Transfer

SE No Yes CFD (1D)

E5-38 – MC Tests Open English P5-28 - Corium Attack of Metallic Liner

SE No Yes CFD (1D)

E5-39 – Plate Tests Open English P5-28 - Corium Attack of Metallic Liner

SE No Yes CFD (1D)

NEA/CSNI/R(2014)3

255

Table 4-7

Systems Experiments


Test DBA

SA/

BDBA 3D

E6-1 - CSE EFADS Tests Open7 English P6-6 - Aerosol Removal in EFADS

SE No Yes Lump

E6-2 - ACE-CSTF EFADS Tests Closed (obtain

from EPRI)

English P6-6 - Aerosol Removal in EFADS

SE No Yes Lump

E6-3 - ACE-LSFF EFADS Tests Closed (obtain

from EPRI)

English P6-6 - Aerosol Removal in EFADS

SE No Yes Lump

7 Reference (BNWL-1587) available at http://www.osti.gov/energycitations/product.biblio.jsp?osti_id=4746863

NEA/CSNI/R(2014)3

256

4.1 Containment Thermalhydraulics Experiments

4.1.1 E1-1 - Flow through Interconnected Vessels

Test Facility: AECL-Interconnected Vessels

Owner Organization: AECL

Experiment Description:

Investigation of transient compressible flows in systems of long ducts and interconnected volumes.

The system consists of a 7.9 m3 reservoir of 791 kPa dry air, blowing down into two 1.6 m

3 pressure

vessels (connected in series) and 15-cm smooth ABS piping (up to 91 m long) with tees, elbows and

branch ducts of 10 and 15 cm diameters.

Figure 4.1.1-1 Layout of AECL Interconnected Vessels

NEA/CSNI/R(2014)3

257

Figure 4.1.1-2 Elbows used in AECL Interconnected Vessels

Table 4.1.1-1

Test Matrix for AECL Interconnected Vessels Tests

Test No.

Reservoir

Nozzle Dia.

(mm)

Main Duct (0.15 m dia) Branch Duct**

90° Elbow Throat Dia

(m)

Duct

Length (m) Tee*

1 96.39 None 0.15 43

2 96.39 None 0.15 89

3 96.39 None 0.102 89

4a 60.73 None 0.102 89

4b 38.1 None 0.102 89

5a 96.39 Standard 0.15(?) 89

5b 96.39 Mitre 0.15(?) 89

6 60.73 Mitre 0.15(?) 89

7a 60.73 None 0.15 89 transition

7b 96.39 None 0.15 89 transition

7c 96.39 None 0.15 89 standard

11a 96.39

11b 60.73

*Transition tee - convert from round to square duct cross-section

** 25.9 m branch duct (0.15 m dia) located at 45.7 m along main duct

Only Test 11a and 11b used two1.6 m3 pressure vessels

NEA/CSNI/R(2014)3

258

References for Experiment:

Brimley, W.J.G., “An experimental investigation of transient compressible flows in systems of long

ducts and interconnected volumes, AECL Report TDVI-354, January 1979.

Range of Key Experimental Parameters:

Initial reservoir pressure = 274 kPa to 791 kPa of dry air

Initial Temperature: ≈ 25°C

Pressure Vessel pressure: up to 450 kPa (a)

Temperatures in Pressure vessel and duct: up to 60°C

Wall temperature: 43 to -27°C

duct exit velocity: max of 275 m/s

Year Tests Performed: 1979

Repeatability Check: Yes (for case 2)

Past Code Validation/Benchmarks: None

Prepared By: Y.S. Chin (AECL)

NEA/CSNI/R(2014)3

259

4.1.2 E1-2 - Bruce LAC Test in Air, Test No. 50

Test Facility: None



An experiment was performed on a full sized Bruce Nuclear Generation Station local air cooler to

determine its heat removal/cooling capability.


Conrath, J.J., “Steam Condensation Heat Transfer Rates in the Presence of Air”, AECL Report, TDVI-314,

November 1973


Air mass fraction: 0.874

Mixture Velocity: 0.85 m/s

Inlet Mixture Temperature: 60.84°C

Water Flow: 7.67 kg/s

Inlet Cooling Water T: 16.4°C

Outlet Cooling Water T: 38.33°C

Heat Removal: 704 kW


Repeatability Check: No

Past Code Validation/Benchmarks:

AECL used this experiment in its GOTHIC validation exercise.

M. Krause et al., “Validation of GOTHIC-IST 6.1a for Modeling Air Cooler Heat Transfer in

CANDU Containment Analysis”, AECL Report RC-2575, Rev. 0, 2001.


NEA/CSNI/R(2014)3

260

4.1.3 E1-3 - LSGMF GMBT001

Test Facility: AECL-LSGMF



Vertical injection of helium from the bottom centre of a 1000 m3 room (8.2 by 10.3 by 10.95 m high).

Tests were performed at essentially isothermal conditions.

Figure 4.1.3-1 AECL Large Scale Gas Mixing Facility

NEA/CSNI/R(2014)3

261


Jones, S.C., and Chan, C.K., “Buoyancy Induced Gas Mixing in a Large Enclosure Data Report:

Experiments Performed in the Large Scale Gas Mixing Facility, Series GMSR01”, AECL Technical Note,

CAB-TN-085, January 1997.


Helium injection: 2.97 g/s for 600 s

Jet Diameter: 5.1 cm

Jet Velocity: 8.6 m/s

He injection T: 16.5C

Pressure: 100 kPa

Initial air T: 18C

Max helium concentration near ceiling: 2%


Repeatability Check: Yes (GMBT007 is a repeat and shows similar hydrogen concentrations)


PSI tested the standard k- ε turbulence model in GOTHIC using this experiment.

Andreani, M., and Smith, B., “On the Use of the Standard k-ε Turbulence Model in GOTHIC to Simulate

Buoyant Flows with Light Gases”, The 10th Int. Topical Meeting on Nuclear Reactor Thermal Hydraulics

(NURETH-10), Seoul, Korea, 2003 October 5-9


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4.1.4 E1-4 - LSGMF GMUS001

Test Facility: AECL-LSGMF



Sideways injection of helium at a height of 8.22 m into a 1000 m3 room (8.2 by 10.33 by 10.95 m

high). Tests were performed at essentially isothermal conditions.

Figure of LSGMF shown in description of test E1-3 - LSGMF GMBT001.


Jones, S.C., and Chan, C.K., “Buoyancy Induced Gas Mixing in a Large Enclosure Data Report:

Experiments Performed in the Large Scale Gas Mixing Facility, Series GMSR01”, AECL Technical Note,

CAB-TN-085, January 1997.


Helium injection: 2.97 g/s for 600 s

Jet Diameter: 5.1 cm

Jet Velocity: 8.6 m/s

He injection T: 16.5C

Pressure: 100 kPa

Initial air T: 18C

Max helium concentration near ceiling: 6%


Repeatability Check: Yes (GMUS002 is a repeat and shows similar hydrogen concentrations, with a

maximum absolute difference of 1% hydrogen)



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4.1.5 E1-5 - AECL-SP Dousing Test No. 1

Test Facility: N/A



A series of 30 experiments to determine the thermal utilization of dousing sprays as a function of

spray droplet size and the air steam ratio in the containing vessel. In particular the experiments were

performed to assess the degree of heat transfer associated with droplet characteristics and containment

conditions for a dousing process under steady state conditions. Test conditions are at steady-state. Droplet

temperatures measured at three elevations within the 3.35 m vessel height. A photographic method was

employed to determine the droplet size of the dispersed spray pattern.

Thermal utilization is defined as the ratio of the actual heat absorbed by a droplet, falling through an

atmosphere maintained at uniform temperature and pressure, to the maximum heat that could be absorbed

under the specified conditions.


Koroyannakis D., Salij S., Hendrie: G., “An Experimental Study of the Thermal Utilization of Dousing

Sprays” AECL Report IR-452, Sept 1983.

Aivaliotis, S.K., “Thermal Utilization of Dousing Spray Droplets in 600 MW CANDU Reactors”, AECL

Report TTR-24, May 1982.

Krause, M., et. al., “Validation of GOTHIC-IST 6.1b for Modeling Heat Removal by Dousing Water in

CANDU Containment Analysis”, AECL Report RC-2576, Rev. 0, 2001.


Drop Diameter: 2.362 mm (sauter mean)

Dousing flow: 0.114 L/s

Vessel pressure: 172.4 kPa

Initial Drop Temperature: 6.1°C

Average Environment Temperature: 70.6°C

Steam/air mass ratio: 12%



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GOTHIC 6.1b validation performed by AECL.

Krause, M., Ramachandran, S., Collins, W.M. and Nguyen, T., “Validation of GOTHIC-IST 6.1b for

Modeling Heat Removal by Dousing Water in CANDU Containment Analysis”, AECL Report No. RC-

2576, Rev. 0, 2001.


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4.1.6 E1-6 - FIPLOC F2

Test Facility: BMC

Owner Organization: BMWi


The experiment has been performed in the multi-compartment Battelle Model Containment (BMC).

BMC was built from reinforced concrete, had a free volume of 640 m³. It was designed to be a 1/64

representation of the Biblis B containment. In Phase 1 of the experiment (50 h) the containment was

heated up by steam injection, then special emphasis was placed on the study of natural convection

phenomena in the multi-compartment geometry affected by variations of steam -, air- and heat injections

into various compartments.

There is an uncertainty remaining related to the leakage from the vessel.


Fischer et al., “CEC Thermal-Hydraulic Benchmark Exercise on FIPLOC Verification Experiment F2 in

BMC, Specification for Phase 1” Battelle Institute e.V. Frankfurt, Germany, September 1989


BMC, Specification for Phases 2,3 and 4” Battelle Institute e.V. Frankfurt, Germany, July 1990


BMC, Long-Term Heat Up Phase, Results for Phase 1” Battelle Institute e.V. Frankfurt, Germany, Report

BF-R-67.249-1, September 1990


BMC, Experimental Phases 2, 3 and 4, Results of comparisons” Commission of the European

Communities, ISBN 92-826-6454-6, Luxembourg


Pressure 1 to 3.2 bar

Atmospheric temperature 20 to 135°C

Atmospheric velocity 0 to 2 m/s


Repeatability Check: N/A

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BMC, Experimental Phases 2, 3 and 4, Results of comparisons” Commission of the European

Communities, ISBN 92-826-6454-6, Luxembourg

Prepared By: M. Sonnenkalb (GRS)

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4.1.7 E1-7 - VANAM M3 (ISP-37)

Test Facility: BMC




BMC was built from reinforced concrete, had a free volume of 640 m³. It was designed to be a 1/64

representation of the Biblis B containment. In Phase 1 of the experiment (15 h) the containment was

heated up by steam injection into the upper internal compartment R5. Then the hygroscopic NaOH aerosol

was injected at the same position using air as carrier gas. The aerosol depletion under superheated “dry”

conditions was studied. After a phase without any injection, a second aerosol injection was performed

followed by the steam injection. The position was changed temporarily to the lower room R3 and then

switched back to R5. The aerosol depletion under (super-) saturated “wet” conditions was studied.

There was an inhomogeneous aerosol distribution in the containment with regard to the dead end

room connected to R9 only. The other rooms have been well mixed.


Firnhaber et al., “Specification of the ISP37, VANAM M3 - A Multicompartment Aerosol Depletion Test

with Hygroscopic Aerosol Material”, GRS, Köln, 1995

Firnhaber et al.,” ISP37, VANAM M3 - A Multicompartment Aerosol Depletion Test with Hygroscopic

Aerosol Material, Comparison Report”, NEA/CSNI/R(96)26 December 1996


Aerosol concentration up to 10 g/m³

Relative humidity 80 to 100%

Temporarily high fog concentrations

Pressure 1 to 2 bar


Atmospheric velocity 0 to 0.75 m/s




Firnhaber et al., “ISP37, VANAM M3 - A Multicompartment Aerosol Depletion Test with Hygroscopic

Aerosol Material, Comparison Report”, NEA/CSNI/R(96)26 December 1996


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4.1.8 E1-8 - EREC LB LOCA Test 1

Test Facility: EREC BC-V-213

Owner Organization: EU


The test facility BC V-213 had been constructed in 1998-1999 at the Electrogorsk Research and

Engineering Centre in the framework of the PH 2.13/95 “Bubble Condenser Experimental Qualification”

Project according to the PHARE and TACIS programs under contract with the European Commission. It

is a large scale integral facility designed to carry out thermal hydraulic experimental studies on the

behaviour of the bubble condenser (BC) in NPPs with WWER-440/V-213 under design basis accident

conditions. The test facility consists of reinforced concrete boxes which model the hermetic compartments

of Paks NPP Bubble Condenser Containment (BCC) 1:100 scaled according to the volume. It includes an

original fragment of the BC with 18 full-scale gap/cap systems. A schematic view of the BC V-213 can be

found at http://base.erec.ru/Specific/BC/BC.htm.

The BC V-213 is designed for a maximum pressure of 300 kPa and a minimum pressure of 80 kPa. A

high pressure system consisting of 5 vessels is designed for preparing the mass and energy release into the

steam generator box at one of three possible break locations. The BC V-213 consists of 5 hermetic

compartments: dead end volume, two steam generator boxes, BC shaft with a full-scale BC fragment and

air trap. The total volume of the BCC model including the BC water is about 510 m3. The BC section is

located in the BC shaft at a concrete pedestal. The BC shaft volume is diminished by volume displacers to

meet the volume scale ratio. The BC gas room is connected with the air trap via one check valve with a

diameter of 173 mm and with the BC shaft by one relief valve with upstream orifice of 122 mm diameter.

The thickness of the hermetic compartments walls is 0.8 m. The internal surfaces of all boxes are lined

with 6 mm thick carbon steel plates. Main elements of the BC model are made of stainless steel with a

thickness of 3 mm. To make the ratio of the inner surface area of compartment walls to their volumes in

the test facility equal to that of Paks NPP, some walls of the hermetic compartments are insulated with

wooden plates of 12 mm thickness. There are about 250 gages for pressure, pressure differences and

temperatures as well as water level, humidity, flow velocity, air concentration, check valve displacement

and steel wall displacement.

The LB-LOCA Test 1 simulates a large break LOCA 2F DN500. It is considered to be representative

for the maximum pressure and temperature loads in the BCC.

http://base.erec.ru/Specific/BC/BC.htm

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1 - air trap; 2 - water treatment installation vessels; 3 - steam generator box No.2;

4 - bubble condenser; 5 - BC shaft; 6 - steam generator box No.1; 7 - dead

volume

Figure 4.1.8-1 EREC BC-V-213 Facility


“Detailed Description of the Thermal-hydraulic Test Facility”, Project PH 2.13/95: Bubble Condenser

Experimental Qualification, BC-D-ER-SI-0002, Revision 2, Deliverable 2.1, Part 1, July 1998

“Detailed Description of the Data Acquisition and the Control Systems of the Thermal-hydraulic Test

Facility”, Project PH 2.13/95: Bubble Condenser Experimental Qualification, BC-D-ER-SI-0004, Revision

2, Deliverable 2.1, Part 3, November 1998

“EREC Test Facility Instrumentation”, Project PH 2.13/95: Bubble Condenser Experimental Qualification,

BC-D-ER-SI-0007, Revision 2, Deliverable 2.1, Part 4, April 1999

“Test procedure for test facility BC V-213”, Project PH 2.13/95: Bubble Condenser Experimental

Qualification, BC-D-ER-SI-0015, Revision 0, Deliverable 2.1, Part 2, July 1999

“Quick Look Report. Test No.: 1”, Project PH 2.13/95: Bubble Condenser Experimental Qualification,

BC-D-ER-SI-0025, 1999

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Initial conditions:

o p: 0.99 bar

o Tatm: 47 - 55°C

o TBC water: 29 - 31°C

o humidity: 60 - 100%

o water level in BC trays: 0.49 m

Water/steam injection (LB LOCA) at middle break location to SG box 1

Experiment range:

o p: up to 2.1 bar

o Δp BC walls: up to 12 kPa

o Tatm: up to 122°C

o TBC water: up to 63°C


Repeatability Check: Yes, (re-mobilisation test in 2003)


Phare PH 2.13/95 – DRASYS code

“Final thermal-hydraulic test report”, Project PH 2.13/95: Bubble Condenser Experimental Qualification,

BC-D-SI-EC-0028, November 1999

Phare PR/TS/17 – COCOSYS

“Performance of independent post-test calculations”, Phare Contract N° 02-0025, RISKAUDIT Report N°

576, Rev. 4, December 2003

Prepared By: M. Sonnenkalb and S. Arndt (GRS)

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4.1.9 E1-9 - EREC LB LOCA Test 5




Test facility description and basic documentation see LB LOCA Test 1 (Test E1-8).

The LB LOCA Test 5 represents loads characteristic for the short term of an LB LOCA in the Bubble

Condenser Containment of NPP with WWER-440/V-213, i.e., maximum pressure difference across the BC

walls.


“Quick Look Report. Test No.: 5”, Project PH 2.13/95: Bubble Condenser Experimental Qualification,

BC-D-ER-SI-0024, 1999


Initial conditions:

o p: 0.98 bar

o Tatm: 17 - 45°C




Water/steam injection (LB LOCA) to SG box 1 at break location far from BC

Experiment range:

o p: up to 2.75 bar

o Δp BC walls: up to 20 kPa




Repeatability Check: Yes (with slightly different conditions for Kola-3 in 2003)

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Phare PH 2.13/95 – DRASYS code

“Final thermal-hydraulic test report”, Project PH 2.13/95: Bubble Condenser Experimental Qualification,

BC-D-SI-EC-0028, November 1999


“Performance of independent post-test calculations”, Phare Contract N° 02-0025, RISKAUDIT Report N°

576, Rev. 4, December 2003

other projects – CONTAIN, TRACO

“Answers to Remaining Questions on Bubbler-Condenser”, Activity report of the OECD NEA Bubbler-

Condenser Steering Group, OECD report NEA/CSNI/R(2003)12, January 2003


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4.1.10 E1-10 – EREC MSLB Test 7





The MSLB tests represents loads characteristic for a main steam line break scenario in the Bubble

Condenser Containment of NPP with WWER-440/V-213. The reference plant is the NPP Kola, unit 3.


Osokin G., Melikhov, V., “Results of MSLB Test 7”, Experimental Studies on: A. Bubble Condenser Test

Facility (R2.01/99). B. TKR Test Facility (R2.02/99) at EREC-Electrogorsk. Final Report, BC-TR-11E,

January 2004

Osokin G., “Results of MSLB Test 7”, Experimental Studies on: A. Bubble Condenser Test Facility

(R2.01/99). B. TKR Test Facility (R2.02/99) at EREC-Electrogorsk. Technical Report BC-TR-07E,

October 2003


Initial conditions:

o p: 0.98 bar

o Tatm: 22 - 52°C



o water level in BC trays: 0.482 - 0.489 m

Steam injection (MSLB) to SG box 1 at break location far from BC

Experiment range:

o p: up to 1.84 bar

o Δp BC walls: up to 7.8 kPa




Repeatability Check: not done

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Past Code Validation/Benchmarks: TACIS project R2.01/99 – COCOSYS

Osokin G., V. Melikhov, “Results of MSLB Test 7”, Experimental Studies on: A. Bubble Condenser Test

Facility (R2.01/99). B. TKR Test Facility (R2.02/99) at EREC-Electrogorsk. Final Report, BC-TR-11E,

January 2004


(R2.01/99). B. TKR Test Facility (R2.02/99) at EREC-Electrogorsk. Technical Report BC-TR-07E,

October 2003


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4.1.11 E1-11 - EREC MSLB Test 9




Test facility description and basic documentation see LB-LOCA Test 1 (Test E1-8).

The MSLB tests represents loads characteristic for a main steam line break scenario in the Bubble

Condenser Containment of NPP with WWER-440/V-213. The reference plant is the NPP Kola, unit 3.

Notes:

Detailed descriptions of facility, measuring system, experimental procedure and results

Experimental data electronically available, video records available

Weak points:

o wooden wall surface insulation causes uncertainties in heat transfer to walls,

o volume displacers (polyurethane plates in BC shaft)



(R2.01/99). B. TKR Test Facility (R2.02/99) at EREC-Electrogorsk, Technical Report BC-TR-08E,

October 2003


Initial conditions:

o p: 0.98 bar

o Tatm: 22-52°C

o TBC water: 32-38°C

o humidity: 32-73%

o water level in BC trays: 0.482-0.489 m

Steam injection (MSLB) to SG box 1 at break location near BC

Experiment range:

o p: up to 1.84 bar





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Repeatability Check: not done

Past Code Validation/Benchmarks: TACIS project R2.01/99 – COCOSYS


(R2.01/99). B. TKR Test Facility (R2.02/99) at EREC-Electrogorsk, Technical Report BC-TR-08E,

October 2003


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4.1.12 E1-12 - EREC SLB G02


Owner Organization: EREC & GRS



The SLB test G02 represents loads characteristic for a steam line break scenario in the Bubble

Condenser Containment of NPP with WWER-440/V-213 (first 30 min). A special boundary condition is

the “cold” initial test facility state (without pre-heating by steam to reduce uncertainties from initial wall

temperatures and atmosphere humidity) as well as failure of active spray system.


Melikhov O.I., Osokin, G.V., Melikhov, V.I., Sokolin, A.V., “Two SLB tests at BC V-213 test facility”,

Quick Look Report, Bilateral Russian-German Project INT 9142, EREC, Electrogorsk, 2003


Initial conditions:

o p: 0.98 bar

o Tatm: 15 - 34°C

o TBC water: 18°C



Steam injection to SG box 1 at break location far from BC

Experiment range:

o p: up to 1.8 bar





Repeatability Check: Yes - with pre-heating in test SLB G01

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o “Performance of independent post-test calculations”, Phare Contract N° 02-0025,

RISKAUDIT Report N° 576, Rev. 4, December 2003

Bilateral Russian-German Project INT 9142 – CCOSYS

o Wolff, H., Arndt, S. and Steinborn, J., “Pre- and post-test calculations of the EREC

experiment SLB G02”, Report GRS-V-INT 9142 - 6/2003, GRS Berlin, May 2004


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4.1.13 E1-13 - HDR V44 (ISP-16)

Test Facility: HDR



Test V44 was a full scale experiment conducted in the containment of the HDR – nuclear power plant

in Karlstein, Germany. The containment had a height of 60 m and a diameter of 20 m. It was separated

into 60 compartments with a dome on top. A rupture of a steam line leading to an early two-phase flow

was investigated. The blowdown lasted about 50 s.

The objectives of the experiment were to determine the loads upon the containment, particularly with

regard to the buildup of pressures and temperatures in the containment compartments during a LOCA in

three different time phases. The mass flow through the vent flow openings and the heat transfer to the

structures has also been of importance.

Due to some problems with the measurement techniques, uncertainty exists related to the time history

of the break energy flow.


Schall: “Design report for the HDR-Containment experiments V21.1, V42, V44”, PHDR-Report No.

3.280/82, January 1982

M. Firnhaber, International standard problem ISP 16: rupture of a steam line within the HDR-containment

leading to an early two-phase-flow: results of post-test analyses: final comparison report, 1985, NEA-

CSNI-112, Vol. 1


Pressure up to 2.5 bars (at 25 s)

Pressure differences up to 0.75 bars




International standard problem ISP 16: rupture of a steam line within the HDR-containment leading to an

early two-phase-flow: results of post-test analyses: final comparison report, 1985, NEA-CSNI-112, Vol. 1


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4.1.14 E1-14 - HDR T31.5 (ISP-23)

Test Facility: HDR



Test T31.5 was full scale experiments conducted in the containment of the HDR – nuclear power

plant in Karlstein, Germany. The containment had a height of 60 m and a diameter of 20 m. It was

separated into 60 compartments with a dome on top. Test T31.5 investigated a large break Blowdown

which lasted about 50 s. Data were also analysed in the so called long term period (up to 1200 s).


Wenzel: “Versuchsprotokoll, Blowdown und Wasserstoffverteilungsversuchsgruppe CON Versuch

T31.5”, PHDR-Arbeitsbericht Nr. 3.520/88 (7.12.87)


Pressure up to 2.5 bars

Pressure differences up to 0.75 bars

Temperatures up to 130°C




Karwat: “International Standard Problem ISP-23, Rupture of a Large-Diameter Pipe within the HDR-

Containment”, Gesellschaft für Anlagen- und Reaktorsicherheit, GRS-A-1622, Oktober 1989

International standard problem: ISP 23: “Rupture of a large diameter pipe within the HDR-containment”:

final comparison report, 1989. Vol. 1, NEA/CSNI-160


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4.1.15 E1-15 - HDR E11.2 (ISP-29)

Test Facility: HDR



The E11 tests are full scale experiments conducted in the containment of the HDR – nuclear power


separated into 60 compartments with a dome on top.

During the first 12 hours of test E11.2, steam was released into the upper compartment 1805,

afterwards a mixture of hydrogen and helium was released at the same position. This established an

atmospheric stratification consisting of a mixture of steam, air, hydrogen, and helium in the upper part of

the containment while the lower part remained filled with air. A steam injection into the lower

compartment 1405 could not dissolve the stratification. An outside spraying of the steel shell of the dome

resulted in a partial dissolution of the stratification.

There are uncertainties on:

heat losses and leakages from the steel containment into the surrounding compartments

heat removal by external spraying of the steel shell


Karwat: OECD Standard Problem OECD-CSNI-ISP29: Distribution of Hydrogen within the HDR-

Containment under Severe Accident Conditions” Final Comparison Report, NEA/CSNI/R(93) 4, February

1993

Datenzusammenstellung, HDR-Containment, PHDR-Arbeitsbericht Nr.3.143/79, Januar 1982

HDR Sicherheitsprogramm, Untersuchungen zur Wasserstoffverteilung in einem Reaktorcontainment,

Quick Look Report Versuchsgruppe E11, Techn. Fachbericht PHDR 111-92, März 1993

HDR Sicherheitsprogramm, Auswertung der experimentellen und analytischen Ergebnisse der HDR-

Wasserstoffverteilungsversuche, Auswertbericht Versuchsgruppe E11, Versuche E11.0-6, Techn.

Fachbericht PHDR 117-94, Februar 1994

GRS has a documentation of the year 1989 about the flow paths between the compartments and the

concrete and steel structures.

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Maximum values:

Pressure 2 bar

Atmospheric temperature 125°C

Light gas concentration (H2 + He) 30 vol.%




Karwat: OECD Standard Problem OECD-CSNI-ISP-29: Distribution of Hydrogen within the HDR-

Containment under Severe Accident Conditions” Final Comparison Report, NEA/CSNI/R(93) 4, February

1993


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4.1.16 E1-16 - HDR E11.4

Test Facility: HDR



The E11 tests are full scale experiments conducted in the containment of the HDR – nuclear power


separated into 60 compartments with a dome on top. In test E11.4 steam was released for 34 h into the

lower compartment 1405, then a mixture of hydrogen and helium gas was released into this room. The

containment atmosphere was well mixed, except for the rooms in the level below room 1405. Later on the

steel shell of the dome was sprayed from the outside.


References, see test E1-15 - HDR E11.2 (ISP-29).


Maximum values:

Pressure 2 bar

Atmospheric temperature 100°C

Light gas concentration (H2 + He) 10 vol.%





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4.1.17 E1-17 - GKSS M1

Test Facility: GKSS



The GKSS test facility was originally constructed as a height and volume scaled quarterly section of a

pressure suppression system (PSS) intended to be used for nuclear-powered ships. Later on the facility

was modified for the performance of fundamental investigations for PSS of BWR type nuclear power

plants with one or three real sized vent pipes (condensation pipes).

Figure 4.1.17-1 Schema of the GKSS Test Facility

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The principal facility construction is characterised by serial connected vessels: pressure vessel,

drywell vessel, wetwell vessel and expansion room. The total volume is 240 m³ and the maximum design

pressure is 6 bars. The pressure vessel simulates the reactor circuit and provides the mass and energy flow

rates for the simulation of LOCAs. The drywell vessel, which simulates the containment space upstream

the PSS, is linked with the wetwell by 3 vent pipes with real diameter of 610 mm and typical pipe

submergence depth of 2.8 m in the PSS water pool. The wetwell is connected with the expansion vessel to

receive the non-condensable gases escaping from the pool. The vessels walls consist of carbon or stainless

steel with different wall thicknesses. There is no outer heat insulation. The measuring technique was

designed mainly for recording of the fast running pressure changes in the PSS. About 230 gauges were

installed. The sampling frequency of the gauges was 10 ms and for selected measurements 0.15 ms. High

speed cameras were used for visual recording of processes in the wetwell.

The main objective of the GKSS M1 experiment is the simulation of a LB LOCA in a German BWR,

and covers the short-term high dynamic loading of the PSS - the vent clearing and pool swell phases - and

the late phase consisting of air lean and air free steam condensation phase (condensation oscillations and

chugging after about 120 s in this experiment).


Aust E., Bültemann, A., Niemann, H.R., Sattler, P., Schwan, H. and Vollbrandt, J., “DAS-Experimente der

GKSS – Bericht über den Hauptversuch M1”, Versuchsbericht Nr. 23 11 AR E 04, GKSS-

Forschungszentrum Geesthacht GmbH, 1981


Initial conditions:

o p: 1.02 bar

o Tatm: 25 - 31°C

o TPSS water: 26.5°C

o humidity: 100%

o water level in PSS pool: 3.8 m

Water/steam mixture injection

Exp. range:

o p: up to 2.45 bar

o Δpdrywell-wetwell: up to 95 kPa


o TPSS water: up to 59°C



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GRS – DRASYS code:

Schwinges B., “Theoretische Untersuchungen zum Verhalten eines Druckabbausystems bei

Störfällen”, GRS-A-969, GRS, Juni 1984

GRS – COCOSYS, ASTEC/CPA:

Nowack H. and Arndt, S., “Post-calculation of the GKSS M1 test with COCOSYS”, GRS-A-3390,

Oktober 2007

Arndt S., “Implementation of a fast running model for the simulation of vent pipes into ASTEC”,

BMWi Project RS 1159, TN-ARN-1/2007 rev. 1, GRS Berlin, December 2007


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4.1.18 E1-18 - MISTRA ISP-47

Test Facility: MISTRA

Owner Organization: CEA


The MISTRA facility (~100 m3 free volume) was in the empty configuration with a bottom centered

vertical injection device. This test series corresponds to two steady-states between superheated steam

injection and wall condensation (Phase A: air-steam mixture and Phase B: Air-steam-helium mixture).

Between the two steady-states, helium was added to the main steam injection and transient flows during

helium injection phase were observed.

Figure 4.1.18-1 MISTRA Facility for ISP-47 Test

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Steam injection:

location: lower center

o R = 0 mm

o height = 1285 mm)

nozzle diameter: 200 mm

superheated steam at: 130 g/s, - 200°C

Helium Injection:

flow: 10.6 g/s during 1850 s

Condensers:

3 condensers temperature at 115°C during both phases A and B


D. Abdo et al., “ISP47 – Phase A – MISTRA experimental results”, CEA Internal Report

SFME/LTMF/RT/03-011/A, 2003.

D. Abdo et al., “ISP47 – Phase B – MISTRA experimental results”, CEA Internal Report

SFME/LTMF/RT/05-010/A, 2005.


Initial Conditions: Pressure and temperature at room conditions (around 20°C and 1.01 bar) before

preheating phase

Superheated steam injection:

o Qsteam=130 g/s at 200°C during all the test

Helium injection:

o 10.6 g/s during 1850 seconds between the two phases A and B

Steady-state results (mean values)

o Phase A

Tmeam=124°C

Pmean=3.3 bar

o Phase B

Tmeam=128°C

Pmean=5.4 bar

Vertical profiles of gas temperature and gas concentration have been recorded

One radial profile of velocity has been recorded with LDA technique

Condensation distribution along the three condensers is also an interesting result and especially the

behavior during the transient helium injection

Year Tests Performed: 2002-2004

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Repeatability Check: Yes (these tests have been successfully repeated several times in order to get all the

profiles)


These tests are part of the ISP47 exercise.

E. Studer et al., “International standard problem on containment thermal-hydraulics ISP47: Step 1 - Results

from the MISTRA exercise “, Nuclear Engineering and Design, Vol. 237, pp. 536-551, 2007.

H.J. Allelein et al., “International Standard Problem ISP47 on containment thermalhydraulics- Final

report”, NEA/CSNI/R(200)710

Prepared By: E. Studer (CEA)

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4.1.19 E1-19 - MISTRA M7




The MISTRA facility was in the compartmented configuration for these test series. The objective is

to assess the effect of compartments and injection location (upper off-centered) on the steady-state results.

The other boundary conditions were exactly the same as for the MISTRA M series.

Figure 4.1.19-1 MISTRA Facility for M7 Test

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The test was divided into two phases with steady-state in air-steam mixture and air-steam-helium

mixture. Between the two steady-states, helium was added to the main steam mass flow and transient

flows were recorded during the helium injection phase.

Free volume ~100 m3 (with compartment)

Steam injection:

o location: high off centred injection (chimney)

o radius = 1352 mm

o height = 3660 mm)

o nozzle diameter: 72 mm

o superheated steam at: 140 g/s, -200°C during all the phases

Helium Injection:

o flow: 2.4 g/s during 5430 s

Condensers:

o 3 condensers temperature at 100°C during all the phases


D. Abdo et al., “M7 – MISTRA Experimental results”, CEA Internal Report SFME/LTMF/RT/07-009/A,

2007.


Initial Conditions: T and P at room conditions (around 20°C and 1.01 bar)

Steam injection

o Di=72 mm

o Ri=1352 mm

o Hi=3660 mm

o Qsteam=140 g/s at 200°C

Helium injection

o Di=72 mm

o Ri=1352 mm

o Hi=3660 mm

o QHe=2.4 g/s at 200°C

o duration 5430 s

3 Condensers at T = 100°C

Phase A: Pmean=2.6 bar

Phase B: Pmean=3.84 bar


Repeatability Check: Yes (test has been performed two times)

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M7 has been used by a PhD student of JSI Slovenia to validate the CFD part of TONUS code. This has

been done within the framework of the SARNET project

M. Babic, “Simulation of the MISTRA M7 experiment with the TONUS CFD code”, CEA internal report,

SFME/LTMF/RT/07-052/A, 2007.

Prepared By: E. Studer and I. Tkatschenko (CEA)

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4.1.20 E1-20 - MISTRA-M8




The MISTRA facility was in the compartmented configuration for these test series. The objective is

to assess the effect of compartments and injection location (lower off-centered) on the steady-state results.

The other boundary conditions were exactly the same as for the MISTRA M series.

Figure 4.1.20-1 MISTRA Facility for M8 Test

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The test was only a steady-state in air-steam mixture. LDA radial velocity profiles were recorded

during the steady state.

Free volume ~100 m3 (with compartment)

Steam injection:

o location: lower off centred injection

o radius = 1352 mm

o height = 3660 mm)


o superheated steam at: 140 g/s, - 200°C during all the phases

3 condensers temperature at 100°C during all the phases


D. Abdo et al., “M8 – MISTRA Experimental results”, CEA Internal Report SFME/LTMF/RT/06-050/A,

2006.


Initial Conditions: T and P at room conditions (around 20°C and 1.01 bar)

Steam injection:

o Di=72 mm

o Ri=1352 mm

o Hi=1279 mm

o Qsteam=140 g/s at 200°C

Phase A:

o Tmeam=124°C

o Pmean=2.64 bar

Year Tests Performed: 2005 – 2006

Repeatability Check: Yes (the test was run three times)


M8 has been used by a PhD student of University of Pisa to validate the CFD part of TONUS code. This

has been done within the framework of the SARNET project

M. Bucci, “Analysis of the M8 MISTRA test with the TONUS code”, CEA internal report,

SFME/LTMF/RT/07-047/A, 2007.


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4.1.21 E1-21 - MISTRA-MASP




MASP test corresponds to spray tests in the MISTRA facility without compartments after a steady-

state condition in air/steam mixtures. Different spray temperatures have been tested and the natural decay

without any spray injection except heat losses has also been recorded for comparison. The initial

conditions are complicated due to the different temperatures of each condenser.

Air-steam steady state at initial conditions

Free volume = 100 m3 (no compartment)

Steam injection:

o location: lower centre

o radius = 0 mm

o height = 1265 mm)


o superheated steam at: 140 g/s, -200°C

3 condensers at different temperature

o 140°C (top and medium condenser) and

o 80°C (lower condenser)


D. Abdo et al., “M5-MASP MISTRA experimental results”, CEA internal report, SFME/LTMF/RT/06-

006/A, 2006


Spray angle: 30°, full cone

Nozzle location: top centered

MASP0: natural decay with only heat losses (injection was stopped prior to the test)

MASP1: 0.87 kg/s of spray at 40°C

MASP2: 0.87 kg/s of spray at 60°C



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MASP tests have been used in a Spray benchmark organised in the SARNET project

L. Blumenfeld et al., “SARNET spray benchmarks: MISTRA thermalhydraulic part, comparison report”,

CEA internal report, SFME/LTMF/RT07-015/A, 2007


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4.1.22 E1-22 - NUPEC M-7-1 (ISP-35)

Test Facility: NUPEC

Owner Organization: JNES


The NUPEC test facility represents a ¼ scaled Japanese PWR with a volume size of 1312 m3 and 25

compartments. The test studied the distribution of hydrogen (simulated by helium) under the influence of a

spray system.


OECD/NEA/CSNI, “Final Comparison Report on ISP-35: NUPEC Hydrogen Mixing and Distribution

Test (Test M-7-1)”, NEA/CSNI/R(94)29, 1994.


He injection: 0 to 0.03 kg/s

Steam injection: 0.08 to 0.03 kg/s

Sprays: 19.4 kg/s

Initial (= max)T: 70°C

Initial (= max) pressure: 140 kPa

Injection Location: Low

Max He-concentration (16 Vol.%)

Year Tests Performed:



OECD/NEA/CSNI, “Final Comparison Report on ISP-35: NUPEC Hydrogen Mixing and Distribution

Test (Test M-7-1)”, NEA/CSNI/R(94)29, 1994.


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4.1.23 E1-23 - NUPEC M-8-2

Test Facility: NUPEC

Owner Organization: NUPEC


The NUPEC facility is a domed cylinder approximately 10.8 m in diameter, 17.4 m high, and 1310 m3

in volume. The facility contains 28 compartments, of which only 25 are interconnected. The dome

volume constitutes approximately 71% of the total containment volume. The containment is constructed

entirely of carbon steel. The containment shell and floors are 12 mm thick except for the first floor, which

is 16 mm thick. The compartment walls are 4.5 mm thick. The outside of the containment is covered with

a layer of insulation, which is covered by a thin metal sheet to protect from weather damage.

Figure 4.1.23-1 NUPEC Facility

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The NUPEC mixing tests were conducted in a large, 1/4-scale simulated containment. The tests

explored the containment response to steam injection and containment spray actuation. Helium gas was

introduced into the containment as a surrogate for hydrogen. Test M-8-2 included steam injection, which

was injected into the lower steam generator foundation compartment along with the helium. The

containment sprays also operated for the duration of the test. The spray water was cooler than the initial

gas and structure temperatures, and despite the addition of hot steam, was the primary cause of the

temperature changes in the test. This particular test was identified as testing three broad areas: (1)

hydrogen mixing; (2) the temperature distribution and stratification; and (3) the containment spray

performance.


Nuclear Power Engineering Corporation (NUPEC), System Safety Department, “Specification of ISP-35 -

NUPEC's Hydrogen Mixing and Distribution Test, Test M-5-5”, ISP35-027, Revision 1, NUPEC,

November 3-4, 1993.


He (used as surrogate for hydrogen): 0 to 0.3 kg/s

Steam: 0.03 to 0.33 kg/s

Steam temperature: 90 K to 115 K

Containment spray: 0 to 19.4 m3/s



Past Code Validation/Benchmarks: MELCOR code benchmark (part of MELCOR set of benchmark

experiments)

Prepared By: R. Lee (NRC)

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4.1.24 E1-24 - PANDA ISP-42, Phase A

Test Facility: PANDA

Owner Organization: PSI


PANDA is a large-scale thermal-hydraulics facility which was originally scaled to the Simplified

Boiling Water Reactor (SBWR) design that is 1:1 in height and 1:25 for volume and power. The total

height of PANDA facility is 25 m and the maximum operating conditions are 10 bars and 200°C. With

respect to the ESBWR, PANDA has scaling factor of approximately 1:1 in height and about 1:40 for

volume and power. A schematic of the ESBWR design including passive safety systems and the

corresponding PANDA facility configuration is shown in the figure below.

PANDA has a modular structure with six cylindrical pressure vessels representing the relevant

volumes: RPV, DW, WW and GDCS. The total volume of PANDA vessels is 460 m3. A system of lines

and valves allows for interconnecting or isolating the vessels individually for specific test requirements.

The two DW vessels are linked by one large diameter interconnecting pipe (IP) and allow studying gas

distribution in multi-compartment geometries. The two WW vessels are also interconnected but with two

large pipes, one in the gas space region and one in the suppression pool region. Four rectangular pools,

open to the atmosphere, contain four condensers: the Isolation Condenser (IC) is directly connected to the

RPV, while one PCC can be connected to DW1 and two PCCs can be connected to DW2. The large

ESBWR Main Vent (MV) lines are represented in PANDA by two pipes, one from each DW vessel to the

suppression pool of the corresponding WW vessel.

Each MV line is submerged to a depth equivalent to the top of the uppermost ESBWR main LOCA

vent. The MV lines allow flow from the DW to the WW when the corresponding pressure difference

(between DW and WW) exceeds the hydrostatic head corresponding to the main LOCA vent submersion

depth.

Also each of the three PCC condensers has a vent line to the suppression pool. The submersion depth

of the PCC vent lines is lower compared to the MV lines. Two Vacuum Breakers (VB1 and VB2) are

available between the WW and the DW vessels. The vacuum breakers allow a gas flow from the WW gas

space to the DW if the DW pressure becomes lower by a prescribed amount than the WW pressure (for

example in case the PCC units have a condensation rate higher than the steam production rate in the RPV).

A total number of 115 electrical heater elements, with a total power of 1.5 MW, are installed in the lower

part of the RPV. The electrical power can be programmed to follow the specification of the test.

Phase A of the ISP-42 investigated the PCC start-up. The objective of this test was to investigate the

start up of a passive cooling system when steam is injected into a cold vessel filled with air and to observe

the resulting gas mixing and system behaviour. The PANDA facility configuration in this test was

representing the main containment features of the Generation III+ SBWR/ESBWR.


ISP-42 (PANDA tests) Blind Phase Comparison Report, NEA/CSNI/R(2003)6, May 2003

ISP42 (PANDA Tests) Open Phase Comparison Report, NEA/CSNI/R(2003)7, May 2003

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Figure 4.1.24-1 Comparison Between ESBWR and PANDA Facility


Containment pressure up to ~3.2 bar

Test duration: ~5400 s,






Prepared By: D. Paladino (PSI)

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4.1.25 E1-25 - PANDA ISP-42, Phase C




The Phase C of the ISP-42 addressed the normal PCCs operation in case of LOCA. The test simulates

the long-term decay heat removal from the containment with three PCCs operating. PCC operation and

self-adjustment are functions of decay heat curve decrease.

The PANDA facility configuration in this test was representing the main containment features of the

Generation III+ SBWR/ESBWR





Drywell (containment) pressure up to ~2.9 bar

Test duration ~2 hours







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4.1.26 E1-26 - PANDA ISP-42, Phase E




The Phase E of the ISP-42 addressed the release of trapped air. The test objective is to investigate the

effect of release of hidden air into a dead-end drywell compartment (DW1).







Drywell pressure up to ~3.3 bar

Time duration: ~5000 s







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4.1.27 E1-27 - PANDA ISP-42, Phase F




The Phase F of the ISP-42 was performed with the objective is to investigate the release of helium

into the RPV and to observe the stratification in the drywell and the resulting effect on the system pressure.







Drywell pressure up to ~5 bar

Time duration: ~5000 s







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4.1.28 E1-28 - PANDA BC4




The SWR-1000 BWR (the actual acronym is KERENA) design passively transfers decay heat from

the reactor pressure vessel to core flooding pools, which are located in the upper part of the DW and

gradually start heating the DW with steam. Containment cooling condensers installed above the core

flooding pools condense steam from the gaseous mixture present within the DW. Schematically how the

SWR-1000 building condenser is represented in PANDA is shown in the figure. These condensers consist

of inclined finned tube. The low and high ends of their secondary sides are connected to the bottom and to

the higher level of the dryer/separator storage pool, which is situated on top of the containment. The

containment cooling condenser system of the KERENA reactor is represented in PANDA at about 1:26

scale, except for the dryer/separator storage pool, which is much smaller, leading to a time compression

factor of about 8 for the transient system tests. Different type of tests were performed, based on different

small, medium and large-break LOCAs with and without core degradation, in the former case with helium

injected to simulate the release of hydrogen. The test series as a whole identified important containment

phenomena and successfully demonstrated the robustness of the KERENA containment cooling system.

The PANDA BC4 test has been performed to investigate small break loss of coolant accident with

core overheat. The system pressurization under the mitigation effects of the building condenser, as well as

venting to the wetwell through the hydrogen overflow pipe, are addressed in this test. The test is directly

related to the Generation III+ KERENA (formerly identified as SWR1000).


J. Dreier et al., “The PANDA tests for the SWR1000 passive containment cooling system”, ICONE-7316,

Tokyo, Japan, 1999


Containment pressure: Start/end: ~1.9 - 4.5 bar

Drywell temperature: ~110°C - 150°C

Test duration: ~23 hours



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Figure 4.1.28-1 Layout of the PANDA Facility



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4.1.29 E1-29 - SVUSS G02

Test Facility: SVUSS

Owner Organization: SVUSS & GRS


The facility at SVUSS Bechovice (Czech Republic) was designed to conduct experiments with

regards to the thermal-hydraulic loads applied to a bubble condenser containment, of a VVER-440/V-213

reactor, during design basis accidents. There were two test facility configurations; here only the vertical

vessel configuration (valid for the G0) test is described. The facility is a small-scaled, integral test facility

with a total containment model volume of ~21 m³, i.e., 1:2200 scaled to NPP (average volume scale).

A reactor pressure vessel model serves as source of hot pressurised water with parameters (pressure

and temperature) corresponding to real nuclear power plant coolant parameters. The containment model

comprises the following parts:

discharge tube from the reactor pressure vessel with quick opening starting device;

horizontal vessel modelling the hermetic compartments;

the vertical vessel containing a Bubble Condenser (BC) model, the gas shaft and a part of the air

trap model including a model of the DN500 check valve connecting the BC and the air trap;

pressure vessel located outside the laboratory which complements the air trap volume.

The BC model contains one gap/cap system, a little shorter in length, but with original height

parameters. Its scaling was determined according to water volumes in the model and in the real BC. The

facility walls are made from steel with a thickness between 3 and 12 mm. A part of the horizontal vessel

inner surface is covered by rubber insulation.


Wolff H., Arndt S. (GRS Berlin), Suchanek M. (SVUSS Prag-Bechovice), “DRASYS Post-test Analysis

for the SVUSS Experiment G02”, BMBF-contract RS 1045, Bilateral German-Czech

cooperation, Technical Note TN-WFF-1/97, Berlin/Prague, September 1997


Initial conditions:

o P: 1.0 bar

o Tatm: 6 - 28°C

o TBC water: 28°C

Water/steam mixture injection

Experiment range:

o p: up to 2.3 bar




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Figure 4.1.29-1 SVUSS Vertical vessel (~11 m³) with Bubble Condenser




Bilateral German-Czech cooperation – DRASYS code:

Wolff H., Arndt S. (GRS Berlin), Suchanek M. (SVUSS Prag-Bechovice), “DRASYS Post-test Analysis

for the SVUSS Experiment G02”, BMBF-contract RS 1045, Technical Note TN-WFF-1/97, Berlin/Prague,

September 1997


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4.1.30 E1-30 - THAI TH1

Test Facility: THAI



The THAI facility is a cylindrical containment with a height of 9.2 m, a diameter of 3.2 m and a

volume of 60 m3. It is well insulated by 0.12 m rockwool. TH1 was a commissioning test with

homogeneously mixed atmosphere. In the first phase the vessel was heated up by steam injection. There

were also phases where pressure and temperatures were held constant, without and with operating cooler

mantels and a phase of natural cool-down. Especially the first phase is suited to check the mass and energy

balance of computer codes.


Kanzleiter et al., “Experimental Facility and Program for the Investigation of Open Questions on Fission

Product Behaviour in the Containment (THAI) Part 1”, Becker Technologies GmbH, Report No. 1501218

– S1, October 2003


Pressure up to 3.5 bar

Atmospheric temperatures up to 125°C





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Figure 4.1.30-1 THAI Facility - General Layout with Removable Internals.

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4.1.31 E1-31 - THAI TH2

Test Facility: THAI




volume of 60 m3. It is well insulated by 0.12 m rockwool. TH2 was a commissioning test. It started with

a 2.5 h steam injection at an elevated position establishing an atmospheric stratification with high steam

content in the upper part of the vessel and pure air in the lower part. The stratification was dissolved from

its bottom, first by producing steam by sump heating, then by steam injection near the bottom. This

process lasted 10 h and was well predicted by 3 independent COCOSYS calculations, which were

performed one year before the commissioning of THAI.


Same as for test E1-30 - THAI TH1.







Schwarz et al., “Blind COCOSYS calculations for the Containment Experiment ThAI TH2”, Annual

Meeting on Nuclear Technology 2002, Proceedings page 95

Prepared By: M. Sonnenkalb

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4.1.32 E1-32 - THAI TH7

Test Facility: THAI




volume of 60 m3. It is well insulated by 0.12 m rockwool. During the first 2000 s of test TH7 steam was

injected at an elevated position establishing an atmospheric stratification. Then steam was injected near

the bottom against an impingement plate, rapidly dissolving the stratification. This was followed by a

phase without injection, and a phase with wall heating.









Fischer, Rastogi: “Containment Code Benchmark Abschlussbericht zum Vorhaben 1501232”, Becker

Technologies GmbH Eschborn (August 2003)

Fischer et al., “Containment Code Comparison Exercise on Experiment ThAI TH7”, NURETH 10, Seoul,

Korea, October 5-9 2003


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4.1.33 E1-33 - THAI TH10

Test Facility: THAI




volume of 60 m3. It is well insulated by 0.12 m rockwool. TH10 was a repetition of the HDR E11.2 test in

a small scale. It started with a steam release at an elevated position, followed by a helium release at the

same position. Then the steam injection was switched towards the bottom of the facility. The outside

spraying of test E11.2 was simulated in TH10 by cooling with the upper cooling mantle. Similar

atmospheric conditions and the same phenomena as in E11.2 where achieved in TH10. First an

atmospheric stratification with steam, air and helium in the upper part of the vessel and almost pure air in

the lower part was established. The stratification was not dissolved by the steam injection near the bottom,

but it was partially dissolved by the outside cooling at the upper part of the facility.






Helium concentration up to 22 Vol.%





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4.1.34 E1-34 - THAI TH13 (ISP-47)

Test Facility: THAI




volume of 60 m3. It is well insulated by 0.12 m rockwool. TH13 was the base of Step 2 of the ISP-47.

The test consisted of four subsequent phases. First helium, then steam were injected at an elevated

position, establishing a steam-air-helium cloud in the upper part of the facility, while there was almost pure

air in the lower part. In the 3rd

phase a horizontal steam jet was injected near the bottom of the vessel.

Because of its low velocity, after a short distance, the jet changed into a buoyant plume, rising in upward

direction. Up to the end of the 3rd

phase the plume had partially dissolved the steam-air-helium cloud from

its bottom. During the 4th phase, the remaining stratification was conserved. Observations made over the

course of TH13 were difficult to predict through the use of computer codes, because differences between

the simulation and the experiment in preceding phases had an impact in later phases, and because it was

especially difficult to simulate the transition of the horizontal jet into a buoyant plume, which then rose,

partially into the inner cylinder, and partially into the annulus of the facility, where it could be partially

blocked by the horizontal condensate trays.


Kanzleiter et al., “Experimental Facility and Program for the Investigation of Open Questions on Fission

Product Behaviour in the Containment (ThAI Phase II) Part 1”, Becker Technologies GmbH, Report No.

1501272 – S1, March 2007






Repeatability Check: Test TH12 confirmed the results of TH13


“International Standard Problem ISP-47 on Containment Thermalhydraulics,” Final Report,

NEA/CSNI/R(2007) 10, (Sept. 2007)


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4.1.35 E1-35 - THAI HM2

Test Facility: THAI




volume of 60 m3. HM2 was the base of an international code benchmark. It was designed to investigate

open questions of the ISP-47. Compared to THAI - TH13, it had a simplified test arrangement and test

procedure, along with more gas concentration measurements (especially in the upper part of the vessel). It

consisted of two phases, with the first involving hydrogen injection, and the establishment of a hydrogen–

air cloud in the upper part of the containment, and second, a vertically directed steam injection near the

bottom of the chamber. Since the injection velocity was low, as it was in test TH13, a buoyant plume rose

upward and gradually dissolved the hydrogen-air cloud from its bottom. The test was conducted until

complete dissolution of the cloud was achieved.


Kanzleiter, Fischer: “Quick Look Report, Helium/Hydrogen Material Scaling Test HM-2”, Becker

Technologies GmbH, Eschborn, Germany, Report No. 150 1326 – HM-2 QLR Rev. 3, March 2008






Repeatability Check: Yes with Helium instead of H2


Schwarz et al., “Benchmark on Hydrogen Distribution in a Containment Based on the OECD-NEA THAI

HM-2 Experiment”, Nuclear Technology, Vol. 175, September 2011


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4.1.36 E1-36 - TOSQAN ISP-47

Test Facility: TOSQAN

Owner Organization: IRSN


Steam, air, and helium injections in a closed vessel with controlled wall condensation.

The TOSQAN facility is a closed cylindrical vessel (7 m3 volume, 4.8 m height, 1.5 m internal

diameter), having walls that are thermostatically controlled through the use of heated oil circulation in a

double stainless steel shell. The envelope is divided into two zones, in order to fix two different values of

the wall temperature. The top and bottom parts of the vessel, called the ‘hot wall’ or ‘non condensing

wall’, have the same temperature. On the middle zone, called the ‘cold wall’ or the ‘condensing wall’,

which constitutes the condensation zone, different, colder temperatures are imposed.

The TOSQAN ISP-47 test sequence is composed of a succession of different steady-states obtained

by varying the injection conditions in the test vessel, leading to three steady-states of the air-steam mixture

at two different pressure levels, and one steady-state of the air-steam-helium mixture. Each steady-state is

reached naturally by keeping a constant steam injection flow rate, as well as a constant wall condensation

mass flow-rate.

The available instrumentation in the TOSQAN condensation tests includes injection mass flow-rates

(steam and/or non condensable gas), gas temperature (over 90 thermocouples in the bulk and 20 located on

the walls), steam and/or helium/non condensable gas volume fraction measured by mass spectrometry (54

points), vessel total pressure and wall condensation mass flow-rate. The facility provides a number of

different optical diagnostics at 4 different levels with 14 viewing windows, as well, allowing steam

concentration (through spontaneous Raman spectrometry), velocities, and turbulence measurements to be

made.


J. Malet, E. Porcherona and J. Vendel, “OECD International Standard Problem ISP-47 on containment

thermal-hydraulics—Conclusions of the TOSQAN part”, Nuclear Engineering and Design, Volume 240,

Issue 10, October 2010, pp. 3209-3220

Cornet P., Malet J., Porcheron E., Vendel J., Studer E., Caron-Charles M. - ISP 47 – International Standard

Problem on Containment Thermal-Hydraulics – Step 1:TOSQAN – MISTRA – Specification of the

calculations – IRSN/DPEA/SERAC Technical Report 02-44 June 2002.

Malet J., Porcheron E., Cornet P., Brun P., Norvez O. and Menet B., Thause L., ISP 47 – International

Standard Problem on Containment Thermal-Hydraulics – Step 1: TOSQAN – MISTRA - Phase A: air -

steam mixtures, TOSQAN experimental results of air-steam phases Technical report,

IRSN/DPEA/SERAC/LPMAC/02-45, 2002.

Porcheron E., Nuboer A., Brun P., Cornet P., Malet J., Menet B., Thause L., Vendel J., ISP 47 –

International Standard Problem on Containment Thermal-Hydraulics – Step 1: TOSQAN – MISTRA –

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Phase B: air-helium steam mixtures – TOSQAN experimental results of the air-helium-steam phase, Rev.

0, 2003.


Steam injection 1 to 12 g/s

Helium injection: 1 g/s

Gas temperature: around 115° C

Wall temperature: 107 and 123°C

Pressure: from 1 to 3 bar


Repeatability Check: Yes (more than 20 times)


J. Malet, E. Porcherona, J. Vendel, “OECD International Standard Problem ISP-47 on containment

thermal-hydraulics—Conclusions of the TOSQAN part”, Nuclear Engineering and Design, Volume 240,

Issue 10, October 2010, pp. 3209-3220

Prepared By: J. Malet (IRSN)

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4.1.37 E1-37 - TOSQAN Condensation Tests




A series of 7 TOSQAN condensation tests performed on the same basis as the TOSQAN ISP-47 test

(E1-36), but with different steam mass-flow-rate, and condensing and non-condensing wall temperatures.

The available instrumentation in the TOSQAN condensation tests includes measurements of the

injection mass flow-rates (steam and/or non condensable gas), the gas temperature (over 90 thermocouples

in the bulk and 20 located on the walls), the steam and/or helium/non condensable gas volume fraction,

measured by mass spectrometry (54 points), the total pressure in the vessel, and the wall condensation

mass flow-rate. The facility also provides a number of locations at which to perform optical diagnostics at

4 different levels with 14 viewing windows, which allows steam concentration (through spontaneous

Raman spectrometry), velocity, and turbulence measurements to be made.


J. Malet, E. Porcheron, F. Dumay and J. Vendel, Code-experiment comparison on wall condensation tests

in the presence of non-condensable gases – numerical calculations for containment studies, under

submission in Nuclear Engineering and Design, 2012

Internal reports

Porcheron E., Malet J., Expérience TOSQAN: essai d’injection de vapeur dans de l’air en présence de

condensation en paroi - Rapport d’essai Ref. 1, IRSN/DPEA/SERAC n°03-18, 2003.

Malet J., Projet TOSQAN: essai d’injection de vapeur dans de l’air en présence de condensation en paroi -

Rapport d’essai n°2, IRSN/DPEA/SERAC n°03-02, 2003.

Malet J., Porcheron E., Projet TOSQAN: essai d’injection de vapeur dans de l’air en présence de

condensation en paroi - Rapport d’essai n°3, IRSN/DPEA/SERAC n°03-03, 2003.








condensation en paroi - Rapport d’essai n°9b, IRSN/DPEA/SERAC n°03-07, 2003.

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Steam injection 1 to 12 g/s

Helium injection: 1 g/s

Gas temperature: around 115°C

Wall temperature: 107 and 123°C

Pressure: from 1 to 3 bar


Repeatability Check: Yes (more than 20 times)



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4.1.38 E1-38 - TOSQAN Test 113




This experiment involved cold spray injection in a stratified layer of helium and dry air. The

TOSQAN facility is a closed cylindrical vessel (7 m3 volume, 4 m high, 1.5 m internal diameter). The

vessel walls are thermostatically controlled by heated oil circulation. The inner spray system is located 65

cm from the top of the enclosure on the vertical axis. It is composed of a single nozzle producing a full-

cone water spray. This nozzle can be moved along the vertical axis in order to perform measurements at

different distances from the spray nozzles under steady-state conditions. In the lower part of the vessel, the

water impacting the sump is removed to avoid water accumulation and to limit evaporation.

The TOSQAN 113 spray test sequence consists on upper radial injections of helium at a given flow-

rate (around 1 g/s) in a dry air closed vessel at ambient pressure. When the vessel relative pressure reaches

1 bar, helium injection is stopped. A delay of 400 s is applied before spray activation. During this time,

mass spectrometry measurements are performed in order to characterize initial stratification of helium and

to check the repeatability of this stratification. For this test, the walls are not thermostatically controlled

and no laser diagnostics are performed.

The available instrumentation on TOSQAN spray tests includes measurements of the mass flow-rate,

temperature, and pressure of the water spray injected, the mass flow-rate and temperature of the water

removed (or drained) to the sump, the mass flow-rate of injected steam and helium, the gas composition

measurements, temperature measured by protected thermocouples, volume fraction measured by mass

spectrometry and Raman spectroscopy, and vessel total pressure. Gas temperature and volume fraction are

also measured in the spray zone. For droplet measurements, available techniques are droplet velocity

measured by PIV, droplet size, as measured by out-of-focus visualization, and droplet temperature.


Malet et al., Spray in containment: final results of the SARNET spray benchmark, Nuclear Engineering

and Design, Volume 241, 2011, pp. 2162-2171

Internal report

Lemaitre P., Nuboer A., Porcheron E., Poulizac A., Rochas V. TOSQAN Experimental programme.

Spray test N°113. Rapport DSU / SERAC / LECEV / 05-22, 2005.

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Helium volume fraction around 40%

Ambient pressure and temperature

Droplet between 100 and 200 µm


Repeatability Check: Yes




Malet J., Vizet J., SARNET spray benchmark, dynamic part: TOSQAN test 113, code-experiment

comparison, IRSN Technical Report DSU/SERAC/LEMAC/08-04, 2008.


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4.1.39 E1-39 - TOSQAN Spray Tests




The TOSQAN facility is a closed cylindrical vessel (7 m3 volume, 4 m high, 1.5 m internal diameter).

The vessel walls are thermostatically controlled by heated oil circulation. The inner spray system is

located 65 cm from the top of the enclosure on the vertical axis. It is composed of a single nozzle

producing a full-cone water spray. This nozzle can be moved along the vertical axis in order to perform

measurements at different distances from the spray nozzles under steady-state conditions. In the lower part

of the vessel, the water impacting the sump is removed to avoid water accumulation and to limit

evaporation.

These 12 TOSQAN spray tests are based on the following sequence. An initial pressurization (under

dry air initial conditions) is performed in the vessel with superheated steam up to around 2.5 bars. Then,

steam injection is stopped and spraying starts simultaneously at a given water temperature (around 25°C to

60°C) and water mass flow-rate (around 10 to 50 g/s). The transient state of depressurization starts and

continues until an equilibrium phase. Some of the tests include helium injection as well.

The available instrumentation on TOSQAN spray tests includes measurements of the mass flow-rate,

temperature and pressure of the water spray injected, the mass flow-rate and temperature of the water

removed (or drained) to the sump, the mass flow-rate of injected steam and helium, gas composition

measurements, temperature measured by protected thermocouples, volume fraction measured by mass

spectrometry and Raman spectroscopy, and vessel total pressure. Gas temperature and volume fraction are

also measured in the spray zone. For droplet measurements, available techniques are droplet velocity

measured by PIV, droplet size, as measured by out-of-focus visualization, and droplet temperature.




Internal reports

Lemaitre P., Nuboer A., Porcheron E. TOSQAN Experimental Programme. Spray test N°101. Rapport

DSU / SERAC / LECEV / 05-11 anglais, 2005.

Lemaitre P., Narbonne G., Nuboer A., Porcheron E., Rochas V. Programme expérimental TOSQAN.

Essais Aspersion N°103 et 104. Rapport DSU / SERAC / LECEV / 05-35, 2005.

Lemaitre P., Nuboer A., Porcheron E., Rochas V. Programme expérimental TOSQAN. Essai Aspersion

N°102 à buse centrée. Rapport DSU / SERAC / LECEV / 06-21, 2006.

Lemaitre P., Nuboer A., Porcheron E., Rochas V. Programme expérimental TOSQAN. Essais Aspersion

N°105 et 106 à buse centrée. Rapport DSU / SERAC / LECEV / 06-28, 2006.

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N°111 à buse centrée. Rapport DSU / SERAC / LECEV / 06-31, 2006.


N°101H à buse centrée. Rapport DSU / SERAC / LECEV / 06-39, 2006.


Helium volume fraction around 40%

Ambient pressure and temperature

Droplet between 100 and 200 µm







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4.1.40 E1-40 - University of Wisconsin Flat Plate Condensation Tests

Test Facility: University of Wisconsin Flat Plate

Owner Organization: University of Wisconsin-Madison


Forced convection steam condensation tests on a ~2 m long flat plate (vertical, horizontal upward

facing, inclined downward facing surface).


Barry, J.J., “Effects of Interfacial Structure on Film Condensation”, Ph.D. Thesis, University of Wisconsin-

Madison, 1987

Huhtiniemi, I.K., “Condensation in the Presence of Noncondensable Gas: Effect of Surface Orientation”,

Ph.D. Thesis, Nuclear Engineering and Engineering Physics, University of Wisconsin, Madison, WI, 1991.


Vapour Temperature: 50 to 99.2°C

Pressure: ~101 kPa

Vapour Velocity: 0.6 to 7 m/s

steam concentration: 13 - 100% (saturated)

Wall Temperature: 25 - 40°C

Average Heat Transfer Coefficient: 89 to 6,208 W/m2-K

Year Tests Performed: 1987 & 1991



Chin, Y.S., and M. Krause, “Validation of GOTHIC-IST 6.1 and 6.1b for Modeling Condensation

Heat/Mass Transfer in CANDU Containment Analysis”, AECL Report No. RC-2574, Rev. 0, 2001


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4.1.41 E1-41 - CONAN SARNET Benchmark No. 1

Test Facility: CONAN

Owner Organization: University of Pisa


The CONAN facility (CONdensation with Aerosols and Noncondensable gases) is installed at the

Scalbatraio Laboratory of the Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione of the

University of Pisa.

The test section of the CONAN facility is a vertical square cross-section channel, with a height of 2.0

m and sides of 0.34 m (see figure below). The mixture of steam and air enters the channel in downward

flow at the channel top at atmospheric pressure. The walls of the channel, except of the cooling plate, are

nearly adiabatic; so, negligible heat and mass transfer occurs on them. A secondary water flow, nearly at

atmospheric pressure, cools the back side of the 0.045 m thick 5083 aluminium cooling plate, flowing in

upward motion through a 0.005 m deep, 0.35 m wide channel.

Secondary Coolant in

Patm, Wsec, Tsec,in

Secondary Coolant out

Tsec,out

Primary Steam + Air in Patm, Vmixt, RH=100%,

Tavg,chann,

Primary Steam + Air out Condensate Flow

?

Surface Temperature and Heat Flux

along the plate centreline ?

z

y

PRIMARY TEST CHANNEL GEOMETRY Square cross-section

Length = 2 m

Sides = 0.34 m

SECONDARY COOLANT CHANNEL GEOMETRY

Rectangular cross-section Length = 2 m

Width = 0.35 m

Depth = 0.005 m

PRIMARY TEST CHANNEL

OUTER SURFACE

COOLED PLATE

SECONDARY

COOLANT CHANNEL

Proposed 2D domain

Figure 4.1.41-1 Layout of CONAN SARNET Benchmark No. 1

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This test facility was used to perform 5 tests (used in the SARNET Condensation Benchmark 1

exercise) on forced convection condensation. The test conditions are listed below in the “Range of Key

Experimental Parameters”.


Ambrosini, W., Bucci, M., Forgione, N., Oriolo, F. and Paci, S., “Data for a Numerical Benchmark on

Condensation Modelling in the frame of SARnet CAM Working Package”, University of Pisa, 2007

December 20.

Ambrosini, W., Bucci, M., Forgione, N., Oriolo, F., and Paci, S., “Quick Look on SARnet Condensation

Benchmark-1 Results, Step 1 – 10 kW Heating Power Exercise”, University of Pisa, 2008 February 22.


Inlet Velocity: 1.5 to 3.6 m/s

Inlet Relative Humidity: 87% to 100%

Inlet Gas Temperature: 75 to 82.7C

Pressure: atmospheric

Cooling Water Temperature (at inlet): 31C

Cooling Water Flowrate: 1.2 kg/s

Year Tests Performed: 2000s


Past Code Validation/Benchmarks: SARNET Benchmark (see references above)

Prepared By: Y.S. Chin (AECL) and W. Ambrosini (University of Pisa)

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4.1.42 E1-42 - CONAN SARNET2 Benchmark No. 2

Test Facility: CONAN

Owner Organization: University of Pisa


A description of the test facility is provided in test E1-41 - CONAN SARNET Benchmark No. 1.

This benchmark involved 10 tests and the conditions of those tests are provided in the “range of key

experimental parameters” below.


Ambrosini, W., Bucci, M., Forgione, N., Oriolo, F. and Paci, S., “SARnet-2 Condensation Benchmark No.

2, Data for a Numerical Benchmark on Condensation Modelling Proposed in the Frame of the SARnet-2

NoE”, University of Pisa, DIMNP RL1225(2009) - Rev. 1, 2009 November 16.

Ambrosini, W., Bucci, M., Forgione, N., Oriolo, F. and Paci, S., “Quick Look Report on SARnet-2

Condensation Benchmark-2 Results”, University of Pisa, DIMNP RL 1252(2010), 2010 June 15.


Inlet Velocity: 2.6 m/s

Inlet Relative Humidity: 100%

Inlet Gas Temperature: 75.6 – 97.1C


Cooling Water Temperature (at inlet): 30.4 – 42.3C

Cooling Water Flowrate: 1.3 – 1.8 kg/s



Past Code Validation/Benchmarks: SARNET2 Benchmark No. 2 exercise (see references above)

Prepared By: Y.S. Chin (AECL) and W. Ambrosini (University of Pisa)

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4.1.43 E1-43 - CSTF Tests

Test Facility: CSTF

Owner Organization: Hanford Engineering Development Laboratory (HEDL)


The Hanford Engineering Development Laboratory (HEDL) Containment Systems Test Facility

(CSTF) was used to investigate hydrogen concentration and mixing for ice condenser containment

configurations. These experiments investigated the degree of mixing and the potential for either

“pocketing” of hydrogen or stratification of hydrogen-rich mixtures. Hydrogen-steam and helium-steam

mixtures were investigated on a scaled basis for releases from a small pipe break in the reactor coolant

system (RCS), or from the pressurizer relief tank through a failed rupture disk. For all but one of these

experiments, helium was used as a simulant for hydrogen. In the other test, hydrogen was used with a

nitrogen atmosphere in the CSTF vessel. Experiments were performed with and without forced circulation

flow to investigate the potential for hydrogen accumulation for different accident scenarios.

References for Experiment: Not provided

Range of Key Experimental Parameters: Not provided

Year Tests Performed: 1982 - 1984


Past Code Validation/Benchmarks: MELCOR 1.8.2

Prepared By: R. Lee (NRC) and M. Salay (NRC)

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4.1.44 E1-44 - Marviken Test 18

Test Facility: Marviken

Owner Organization: Studsvik


The MARVIKEN station was originally built to operate as a boiling heavy-water direct-cycle reactor

with natural circulation. The station had been completed up to acceptance testing including preoperational

light water tests. The main parts of the facility are the pressure vessel and the containment. Discharge

flow from the pressure vessel was fed into the containment through a discharge pipe. The containment

(commercial containment) involves multi-compartment drywell part (1,970 m3) connected to:

i) the wet well (2,149 m3, 556 m3 water pool) by channel-type vent system and header;

ii) 58 vent pipes 5.5 m long, 0.3 m diameter, and 30 vent pipes blocked during the experiment No. 18;

iii) as well as a total vent flow area, activated during the experiment (1.95 m2), and with the initial

submergence of vent pipes of 2.8 m.

Pressure vessel was filled with 278 tons of water, which was heated and pressurized to approximately

4.6 MPa. In all the blow-down tests from No. 18 and beyond, a temperature stratification was established

with about 25° sub-cooling in the lower region, and with saturation temperature in the upper region in

order to prolong the period of sub-cooled single-phase liquid flow into the containment and facilitate the

measurements of the discharge flow rate in the early phase of the blow-down.

Measured parameters:

Pressure and differential pressure in the vessel

Temperature

Mass flow rates, including in the main discharge pipe to measure discharge flow rate in periods of

sub-cooled liquid flow

Liquid level and phase boundary

Impact load

Data accuracy:

Discharge flow rate error: between ±7% and ±20% (for rapid transients)

Specific enthalpy: between -1% and +3%

Flow rate of the air into the wet well: 5-6% (although larger errors exist for short periods)

Steam flow rate into the wet well: 10-15% for stable flow rates and 30-40% for strong flow

variations.

Water flow rate into the wet well: no reasonable accuracy can be claimed

Gas velocity in downcomers: ±5%

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The following MXB reports are available at the OECD/NEA:

The Marviken full scale Containment Experiments. Description of the test facility, MXB-101

The Marviken full scale Containment Experiments. Measurement system, MXB-102

The Marviken full scale Containment Experiments. Data accuracy, MXB-105

The Marviken full scale Containment Experiments. Blowdown 18 results, MXB-218

The Marviken full scale Containment Experiments. Appendix to Blowdown 18 results, MXB-218 App

The Marviken full scale Containment Experiments. Summary report, MXB-301

The Marviken full scale Containment Experiments. Report abstracts, MXB-401

J-E Marklund, Summary Report for ISP-17, International Standard Problem for Containment Codes on

Blowdown No. 18 in the Marviken Full Scale Experiments, STUDSVIK/NR-84/464 (1984)

Jan-Erik Marklund, Preliminary Data Comparison Report for ISP-17 - An international containment

standard problem based on the Marviken full scale experiment Blowdown No 18,

STUDSVIK/NR-84/423 (84-06-04), referenced also as CSNI Report No 103.


Initial Conditions:

o Vessel, see experiment description

o Containment:

dry well (P= 0.105 MPa, T= 37.5°C)

wet well (P= 0.105 MPa, T= 16°C)

Final conditions

o Dry well pressure ~0.175 MPa


Repeatability Check: Yes (tests of the same type have been performed)


The Experiment No. 18 has been used for the CSNI ISP-17 where participants utilized 5 computer codes

(ARIANN-1 and CONTEMPT-LT/26 for Italy, CONTEMPT-LT/28 for Finland, COPTA-7 for Sweden

and ZOCO-V for the Netherlands); see the CSNI Report No 103 mentioned above.

Prepared By: A. Amri (OECD)

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4.1.45 E1-45 - CARAIDAS EVAP and COND tests

Test Facility: CARAIDAS

Owner Organization: IRSN (with EDF partial finding)


This experiment investigated single droplets falling down in a pressurized homogeneous air-steam

mixture, looking at the diameter droplet evolution at two levels. The IRSN CARAIDAS experimental set-

up was used to study drop evolution under representative conditions of post-accident atmosphere. The

cylindrical enclosure is of 5 m high and 0.6 m inner diameter. Homogeneous conditions are obtained, and

gas temperatures set from 20 to 160°C, absolute pressures set from 1 to 8 bar, and relative humidities set

from 3% up to 95%. The drop generator is located at the top of the vessel in order to keep it at the same

temperature as the surrounding vessel. It produces monodisperse water droplets at sizes that can be set

from 200 to 700 µm in diameter. Drop injection temperature is set between 20 and 80°C by an electric

heater. Initial droplet size, velocity, and temperature are determined experimentally for each test. So-

called ‘evaporation’ and ‘condensation’ tests are performed.

Drop diameter optical measurements are performed at 3 elevations.

Evaporation tests EVAP3, EVAP13, EVAP18, EVAP21 and EVAP24

Condensation tests: COND1, COND2, COND7 and COND10


J. Malet, SARNET-2: Droplet heat and mass transfer elementary benchmark: comparison report, IRSN

report IRSN/DSU/SERAC/LEMAC/11-04, February 2011

Internal reports

J.B. Coreau, D. Ducret, D. Roblot, J. Vendel, Programme Aspersion Rapport D’Essais de la Campagne

Evaporation Sur Caraidas Technical report, IRSN/DPEA/SERAC/LPMC/97-07

D. Ducret, D. Roblot, J. Vendel, Programme Aspersion Campagne D’Essais Condensation Sur Caraidas

Technical report, IRSN/DPEA/SERAC/LPMC/98-15


Gas temperature: 20 to 140°C

Pressure: 1 to 5 bar

Droplet size: 200 to 600 µm


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Internal IRSN validation of ACACIA, CASPER and ASTEC

SARNET-2: Droplet heat and mass transfer elementary benchmark: comparison report

J. MALET, IRSN report IRSN/DSU/SERAC/LEMAC/11-04, February 2011


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4.1.46 E1-46 - TOSQAN sump tests




These tests examined evaporation of water over a water sump surface in atmospheres of pressurized

air-steam, air-steam-helium, air-steam-CO2, and air-steam-SF6 mixtures. There were 6 tests under steady-

states evaporative conditions, 2 transient depressurisation tests with sump. In these experiments, wall

condensation also occurs.

The TOSQAN facility is a closed cylindrical vessel (7 m3 volume, 4.8 m height, 1.5 m internal

diameter), having thermostatically controlled walls by heated oil circulation in a double stainless steel

shell. The TOSQAN sump is a small vessel of 350 litres, with an internal diameter of 684 mm, connected

on the basement of the main vessel.

The TOSQAN sump tests are mainly based on the following test sequence: the vessel is initially

closed with dry air at rest at a given thermal equilibrium obtained from the imposed wall temperatures:

‘condensing wall’: middle wall, of 2 m height, at 2.6 m from the bottom of the vessel

‘non-condensing wall’: the remaining walls, including the upper, the lower and the sump walls

At a given time, steam is injected in the vessel, and an equilibrium state is obtained. The total

pressure is constant, and the steam injection mass flow-rate is equal to the wall condensation mass flow-

rate. Following this equilibrium, steam injection is stopped, vessel depressurization occurs and a second

thermal equilibrium state is reached. The water is then injected into the sump. After a given time, the

sump water is heated with a power supply, so that a third steady-state is reached where the sump

evaporation mass flow-rate reaches the value of the middle wall condensation mass flow-rate.

The available instrumentation in the TOSQAN main volume concerns injection mass flow-rates

(steam and/or non condensable gas), gas temperature (over 150 thermocouples), steam and/or helium/non

condensable gas volume fraction measured by mass spectrometry (54 points) and vessel total pressure.

The facility provides also numerous possibilities of optical diagnostics at 4 different levels with 14 viewing

windows.

In the TOSQAN sump, a sampling manifold for mass spectrometry is available above the sump

interface, for gas concentration measurements, as well as three dense bundles of 32 thermocouples

allowing the recording of two horizontal and one vertical temperature profiles.

This is a series of 8 tests (TOSQAN sump tests 201, 202, 203, 204a, 204b, 205, 206 and 207)

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J. Malet, , M. Bessiron, and C. Perrotina, “Modelling of water sump evaporation in a CFD code for

nuclear containment studies”, Nuclear Engineering and Design, Volume 241, Issue 5, May 2011, pp. 1726-

1735

Internal reports

Lemaitre P., Nuboer A., Porcheron E., Rochas V. Programme expérimental TOSQAN. Essais Puisard

N°201, 206 et 207. Rapport DSU / SERAC / LECEV / 08-20, 2008.


N°202 et 203. Rapport DSU / SERAC / LECEV / 08-14, 2008.


N°204A, 204B et 205. Rapport DSU / SERAC / LECEV / 08-19, 2008.


Gas temperature: around 110°C

Pressure: 2-3 bar




Internal IRSN ASTEC/CPA and TONUS-CFD validation

Validation published: J. Malet, M. Bessiron and C. Perrotina, “Modelling of water sump evaporation in a

CFD code for nuclear containment studies”, Nuclear Engineering and Design, Volume 241, Issue 5, May

2011, pp. 1726-1735


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4.1.47 E1-47 - CALIST PWR spray test

Test Facility: CALIST

Owner Organization: IRSN (with EDF partial finding)


This experiment studied entrainment effects with a real PWR spray nozzle. Exact design and

reference of the nozzles used in many French PWRs (and in some US NPPs). The hollow cone spray had a

flow rate of 1 kg/s at 3.5 bar nozzle pressure.

The CALIST facility (Characterization and Application of Large and Industrial Spray Transfer)

allows global and local spray characterization. The set-up is composed of a hydraulic circuit supplying, for

those experiments, a single spray nozzle with a flow-rate of 1 L/s at a relative pressure of 3.5 bars. The

pulverized water is collected in a 5 m3 pool. The axial position of the spray nozzle can be changed using a

monitored carriage.

For the PWR spray nozzle gas entrainment tests, all measurements were performed in air under

atmospheric conditions. Experimental measurements were performed on a single spray nozzle which is

routinely set up in many Pressurized Water Reactors. This nozzle is generally used with water at a relative

pressure supply of 3.5 bars, producing a mass flow rate of approximately 1 kg/s. Measurements of the

velocities of the fog droplet give some information on the important entrainment of the gas occurring

around this high momentum spray.

Radial profiles of droplet sizes and 3D velocities can be measured at all positions in the spray using

Phase-Doppler Interferometry (PDI) with large focal length lenses. The CALIST facility allows the

estimation of the gas velocity using fog sprays of tiny droplets entrained by the gas.

Experimental results are obtained at several different heights, in terms f radial profiles in four

different directions.


Foissac A., Modélisation des interactions entre gouttes en environnement hostile, Thèse de Doctorat de

l’Université Paris VI, 2011.

Malet J., Gas entrainment by one single PWR spray, SARNET-2 Elementary benchmark - Results report,

IRSN/ PSN-RES/SCA/LEMAC/2012-11, 2012.

Foissac A., Malet J., Vetrano M.R., Buchlin J.M., Mimouni S., Feuillebois F., Simonin O., Droplet size

and velocity measurements at the outlet of a hollow-cone spray nozzle, Atomisation and Sprays, Vol. 21,

pp. 893-905, 2011.

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Gas temperature: ambient

Pressure: ambient





IRSN-EDF NEPTUNE-CFD and ANSYS validations

Foissac A., Malet J., Mimouni S., Ruyer P., Feuillebois F., Simonin O., Eulerian simulation of interacting

PWR sprays including droplet collisions, accepted for publication in Nuclear technology, 2012.

Prepared By: J. Malet

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4.1.48 E1-48 - MISTRA LOWMA




A helium rich layer is formed at the top of the MISTRA facility. Then, upper off-centered injection

of air is performed in order to erode the light gas stratified layer. These tests are part of the OECD SETH

II project and a counterpart experiments with almost the same conditions has been performed in the

PANDA facility for scaling issues. The evolution of the helium stratified layer without any air injection

has also been recorded for comparison.

Figure 4.1.48-1 MISTRA Facility for LOWMA Test

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Free volume = 100 m3 (with compartment)

Air injection:

location: high off centred injection (chimney)

o radius = 1352 mm

o height = 3660 mm

nozzle diameter: 72 mm

different air mass flow-rate at room temperature


J. Brinster et al., “OECD-SETHII project: Synthesis report for MISTRA INITIALA/LOWMA tests”, CEA

internal report SFME/LEEF/RT/09-007/A, 2009

E. Studer et al., “Interaction of a light gas stratified layer with an air jet coming from below: large scale

experiments and scaling issues”, CFD for Nuclear Reactor Safety Applications (CFD4NRS-3) Workshop,

Bethesda, 2010 September 14-16


MISTRA is initially at room pressure and temperature

A rich Helium layer at the top is created by the use of the four upper radial injector (about 40 vol%

of helium in air)

INITIALA test: evolution of the Helium stratified layer is recorded without air injection

LOWMA4: air injection of 50 g/s

LOWMA3: air injection of 15 g/s

LOWMA2: air injection of 4.5 g/s

LOWMA1: air injection of 1.5 g/s


Repeatability Check: Yes (LOWMA3 has been repeated six times)


These tests have been submitted to a benchmark exercise organised within the SETH-II project


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4.1.49 E1-49 - PANDA OECD/SETH tests




Two PANDA vessels have been used in the SETH project to represent nuclear containment

compartments. These two vessels, having each a diameter of about 4 m and height of 8 m (the total

volume of the two vessels being about 183 m3), are interconnected by a pipe (IP) of about 1 m diameter.

For the SETH tests, it was necessary to upgrade the auxiliary systems. This included the implementation

of components for reaching the specified vessel wall temperature, and the fluid temperature and

composition. Also, additional components have been included for obtaining the specified injection fluid

flow rate, temperature and composition and components to control the fluid pressure by regulating the

venting flow rate.

Figure 4.1.49-1 PANDA Facility for OECD/SETH Tests

The PANDA instrumentation allows for the measurement of fluid and wall temperatures, absolute and

differential pressures, flow rates, heater power, gas concentrations and flow velocities. The sensors are

implemented in all the compartments of the facility, in the system lines, and in the auxiliary systems. For

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the SETH tests, the measurement grids in the two main vessels have been refined in consideration of the

spatial resolution required in each test, in order to obtain experimental data with a spatial resolution

suitable for advanced code validation.

These are three main series of tests (with a total of 25 tests) addressing with CFD-grade

instrumentation, for a broad range of thermal hydraulic and geometrical conditions basic flow structure,

e.g., wall plume, free plume (the figure shows one of the configuration for free plume with injection line

located in the center of one of the two vessels), horizontal jet. Main phenomena are captured at a large-

scale and in a multi-compartment facility. The main phenomena addressed in the tests, allowed identifying

main parameters leading to gas stratification build-up into the containment. Gas stratification build-up has

high safety relevance for LWR containment because, during a severe accident, it can lead to elevated

hydrogen concentrations, and possibly even to a detonation or explosions.


D. Paladino, M. Andreani, R. Zboray and J. Dreier, “Toward a CFD quality database addressing LWR

containment phenomena”, Nuclear Engineering and Design, doi:10.1016/j.nucengdes.2011.08.064

O. Auban, R. Zboray and D. Paladino, “Investigation of Large-Scale Gas Mixing and Stratification

Phenomena related to LWR Containment Studies in the PANDA Facility”, Nuclear Engineering and

Design, Volume 237, Issue 4, pp. 409-419, February 2007.

D. Paladino, M. Andreani, R. Zboray and J. Dreier, “Flow transport and mixing induced by Horizontal jet

impinging on a vertical wall of the multi-compartment PANDA facility”, Nuclear Engineering and Design,

240 (2010) 2054-2065

D. Paladino, R. Zboray and O. Auban, “The PANDA Tests 9 and 9bis investigating gas mixing and

stratification triggered by low momentum plumes”, Nuclear Engineering and Design, Volume 240, Issue 5,

May 2010, pp. 1262-1270

D. Paladino, R. Zboray, P. Benz and M. Andreani, “Three-gas-mixture plume inducing mixing and

stratification in a multi-compartment containment”, Nuclear Engineering and Design, Vol. 240, Issue 2, pp.

210-220 (2010).

R. Zboray and D. Paladino, “Experiments on basic thermalhydraulic phenomena relevant for LWR-

Containments: gas mixing and transport induced by buoyant jets in a multi-compartment geometry”,

Nuclear Engineering and Design, Volume 240, Issue 10, October 2010, pp. 3158-3169


Two (air-steam, helium-steam) and three gases (helium, steam, air)

Pressures: 1.3 up to 3 bar

Temperature: 76-150°C



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OECD/NEA SETH Seminar 2007 at IRSN.

Two PANDA tests were used within the EU 5TH

FWP ECORA project for code benchmark


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4.1.50 E1-50 - PANDA OECD/SETH-2




Two PANDA vessels have been used in the SETH-2 project to represent nuclear containment

compartments. These two vessels, having each a diameter of about 4 m and height of 8 m (the total

volume of the two vessels being about 183 m3), are interconnected by a pipe (IP) of about 1 m diameter.

Figure 4.1.50-1 PANDA Facility for OECD/SETH-2 Tests

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Within the SETH-2 project, six series of PANDA tests (with a total of 24 tests) were performed. The

tests addressed at large scale with CFD-grade instrumentation, hydrogen (simulated using helium)

stratification break-up in a containment induced by:

Basic phenomena such as negatively buoyant vertical (series ST1) or horizontal (series ST2)

jet/plume.

Operation of safety components such as containment spray (series ST3), containment cooler

(series ST4), heat source simulating thermal effect of the recombiner (series ST5). The figure

shows the two PANDA vessels in one of the ST5 tests with the heat source located in the lower

region of Vessel 1. At the beginning of the test, a helium-rich layer existed in the upper part of

Vessel 1. Then, due to the effects of the flow induced by the heat source, the helium-rich layer

mixed with the containment atmosphere beneath it.

In a dedicated test series (ST6) the effect on gas stratification build-up/break-up after opening of

hatches in containment (series ST6) has been investigated.

The experimental results provide a unique experimental database for the assessment and validation of

the most advanced computational tools used in nuclear safety. This is crucial to increase reliability and

confidence in safety analysis. Moreover with these tests has been possible to analyze the beneficial effect

of mass and energy sources in mixing the containment atmosphere. Component tests allow any possible

undesired effects taking place during the activation of safety systems during postulated severe accident to

be identified. This encourages further analysis which could lead to a refining/improvement of severe

accident management procedures.


“OECD/SETH-2 project, PANDA AND MISTRA experiments: investigations of key issues for the

simulation of thermal-hydraulic conditions in water reactor containment”, Final Summary Report

submitted to CSNI/PRG 19 October 2011

OECD/NEA SETH-2 Horizontal Fluid Release Tests, Test Series Report, R. Zboray, D. Paladino, G.

Mignot, N. Erkan, R. Kapulla, M. Ritterath, M. Fehlmann, C. Wellauer, TM-42-10-05, September 2010.

OECD/NEA SETH-2 Containment spray test report, R. Zboray, N. Erkan, G. Mignot, R. Kapulla, M.

Fehlmann, C. Wellauer, M. Ritterath, D. Paladino, TM-42-10-33, December 2010.

N. Erkan, G. Mignot, R. Kapulla, R. Zboray, D. Paladino, “Experimental investigations of spray induced

gas stratification break-up and mixing in two interconnected PANDA vessels”, Nuclear Engineering and

Design, Volume 241, Issue 9, September 2011, pp. 3935-3944

R. Kapulla, G. Mignot, D. Paladino, “Large-Scale Containment Cooler Performance Experiments under

Accident Conditions”, Science and Technology of Nuclear Installations, Volume 2012 (2012)

G. Mignot, R. Kapulla, D. Paladino, R. Zboray, “Experiments on Large Scale Plume Interaction with a

Stratified Gas Environment Resembling the Thermal Activity of Autocatalytic Recombiner”, 2012

International Congress on Advances in Nuclear Power Plants (ICAPP 2012), 2012 June 24-28, Chicago,

USA.


Two (helium-steam, air-steam, helium-air) and three gases (helium, steam, air)

Pressures: 1 up to 3 bar

Temperature: 76-150°C

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OECD/NEA SETH-2 Seminar 2011


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4.1.51 E1-51 - CYBL Boiling Tests

Test Facility: CYBL (precursor experimental rig)

Owner Organization: SNL (Sandia National Laboratories)


CYBL is a reactor-scale ex-vessel boiling experimental facility designed for assessing the flooded

cavity design of the new production heavy water reactor. Results were also used for demonstrating that the

heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-

600 advanced light water reactor, can be met with the flooded cavity design. The CYBL facility has a

tank-within-a-tank design, where the inner tank simulates the reactor pressure vessel and the outer tank

simulates the reactor cavity. The inner tank is 3.7 m in diameter and 6.8 m high. The outer tank is 5.1 m

in diameter and 8.4 m high. The inner test tank has a torispherical head. The energy deposition on the

bottom head is accomplished with an array of 20 radiant lamp panels.

For all the experiments performed, the water level is 5 m above the bottom center of the test vessel.

Because of the 5 m of gravity head, the bulk condition near the bottom head area was subcooled and

therefore boiling outside the reactor vessel was subcooled nucleate boiling. The boiling process exhibits a

cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth,

coalescence, and vapor mass dispersion. By adjusting the power input to each zone, the density of lamps

of each heating panel and the three-dimensional configuration of the panel array, the heat flux distribution

can be customized.

Nearly 300 data channels are used to monitor vessel and surface temperatures, as well as water

temperatures. Temperature gradients from in-depth and surface temperature measurements are used to

calculate local heat fluxes. Results are presented for distribution of wall temperature excess (Tw-Tsat)

along the wall surface and associated heat transfer coefficients. The CYBL facility experiments

demonstrated that the ex-vessel boiling process is dominated by spatial structures on a scale of meters.

There is general agreement between the heat transfer results of the quenching experiments (see below) and

the CYBL experiments.

Boling curves were obtained in quenching experiments, where an initially hot aluminium mass was

plunged in water. Two 61 cm diameter specimens were used, the first with flat bottom and the second with

curved bottom. The CHF was found to be essentially the same, with an average value of approximately

500 kW/m2.

Ten downward-facing boiling tests (CHF during 2 quenching experiments)


T.Y. Chu, J.H. Bentz, and R.B. Simpson “Observations of the Boiling Process from a Downward-Facing

Torispherical Surface: Confirmatory Testing of the Heavy Water New Production Reactor Flooded Cavity

Design”, 30th National Heat Transfer Conference, Portland, Oregon, August 5-9, 1995

T. Y. Chu, et al., “Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity

concept for in-vessel core retention Part II: Reactor-scale boiling experiments of the flooded cavity concept

for in-vessel core retention”, Nuclear Engineering and Design, Vol. 169, pp. 89-99, 1997.

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T. Y. Chu, et al., “Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity

concept for in-vessel core retention Part I: Observation of quenching of downward-facing surfaces”,

Nuclear Engineering and Design, Vol. 169, pp. 77-88, 1997.


Pressure (free surface): 1 bar:

Subcooling (nominal maximum): 12 K

Table 4.1.51-1

CYBL Boiling Test Matrix

Test Heat Flux (W/cm2) Flux Distribution

NPR-A 16 Uniform

NPR-B 16 Edge-peaked

NE1-UA 16 Uniform

NE1-UB 18

NE1-UC 20

NE2-A 8 Edge-peaked

NE2-B 16

NE2-C 17

NE2-D 18

NE2-E 20 Uniform




Prepared By: M. Andreani (PSI)

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4.1.52 E1-52 - ULPU CHF Tests

Test Facility: ULPU

Owner Organization: UCSB (University of California Santa Barbara)


ULPU is a facility for prototypic simulation of boiling heat transfer and CHF under external reactor

vessel flooding conditions for the Loviisa reactor (Finland), the AP-600 (Configurations I to III) and the

AP-1000 (configurations IV and V). The facility consists of a full-scale, full-length, 2D slice (15-cm slice

of the rector) simulating one half of the curved portion of the hemispherical lower head at full scale. The

curved portion contains heaters supplying specified power to the slice test section, whose outer surface is

cooled by water. The water flows in a natural circulation circuit with appropriate flow resistances to

represent the actual situation. The heaters in the facility shape the power to match the expected thermal

load distribution from melt configuration in the RPV and were designed to provide heat fluxes of up to

2000 kW/m2 for Configuration I to IV, and 2400 kW/m

2 for configuration V.

The more recent configurations of the facility included a baffle to channel the external coolant. It was

therefore possible to examine the potential of achieving higher values of CHF with an “organised” flow of

the coolant, and thus optimize the effect of the geometry on the coolability limits. Special attention has

also been given to the flow regimes with and without full simulation of the natural circulation. For the last

series of tests, extensive information is available on the hydrodynamics of the natural circulation loop

(mass flow rates, pressure drops, and flashing-induced instability). Using configuration V, also the effect

of water chemistry on CHF was investigated.

Large number of tests on CHF under natural circulation conditions in a 2-D geometry and with 5

configurations.


T.G. Theofanous, S. Syri, T. Salmassi, O. Kymäläinen, H. Tuomisto “Critical heat flux through curved,

downward facing, thick walls”, Nuclear Engineering and Design, Vol. 151, pp. 247-258, 1994.

T.G. Theofanous, S. Syri “The coolability limits of a reactor pressure vessel lower head”, Nuclear

Engineering and Design, Vol. 169, pp. 59-76, 1997.

T.G. Theofanous, C. Liu, S. Additon, S. Angelini, O. Kymäläinen and T. Salmassi “In-vessel coolability

and retention of a core”, Nuclear Engineering and Design, Vol. 169, pp. 1-48, 1997

T.G. Theofanous, J.P. Tu, T. Salmassi and T.N. Dinh, “Quantification of Limits to Coolability in ULPU-

2000 Configuration IV”, CRSS Technical Report 02.05.3, 2002 May

T-N. Dinh, J.P. Tu, T. Salmassi, T.G. Theofanous “Limits of Coolability in the AP 1000-Related ULPU-

2400 Configuration V Facility”, Center for Risk Studies and Safety University of California, Santa

Barbara, Report CRSS-03/06, June 2003.

T-N. Dinh, J.P. Tu, T. Salmassi, T.G. Theofanous “Limits of Coolability in the AP 1000-Related ULPU-

2400 Configuration V Facility”, Center for Risk Studies and Safety University of California, Santa

Barbara, Report CRSS-03/06, June 2003.

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T. N. Dinh, et al., “Limits of coolability in the AP1000-related ULPU-2400 configuration V facility,” in

The 10th Int. Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), Seoul, Korea, 2003.

T. N. Dinh, et al., “ Two-Phase Natural Circulation Flow in AP-1000 In-Vessel Retention-Related ULPU-

V Facility Experiments”, in Proceedings of ICAPP ’04, Paper 4242, Pittsburgh, PA USA, June 13-17,

2004,


Heated section: 30 and 90°

Boiling: pool and natural flow loop conditions

Baffle position:

o Configuration I to III: not streamlined

o Configuration IV: streamlined, distance form vessel wall: 2.5, 5, 7 and 9 inches

o Configuration V: streamlined, distance from vessel wall: 3 and 6 inches and variable gap

size from 3 to 6 inches. In this series, the riser exit zone is modified to reflect the AP1000

geometry

Heat flux: various shapes with peak inlet fluxes up to 2,400 kW/m2

Pressure (free surface): 1 bar at the top and prototypic gravity head at the bottom.

Subcooling (maximum, nominal): 13 K (due to gravity head).

Flow rate: 0.5 to 0.7 m3/min



Past Code Validation/Benchmarks: N/A


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4.1.53 E1-53 - SULTAN CHF Tests

Test Facility: SULTAN



The SULTAN facility was designed to study the main characteristics of two-dimensional, two-phase

flow along a heated plate, and the limits of the critical heat flux for a wide range of thermohydraulic and

geometric parameters, including gap size and inclination angle. It was designed to provide data for code

development and validation. Therefore, the system effect (natural circulation controlling the flow) is not

represented, and forced convection is used. The representative range of mass flow rate is determined by

pre-calculations. Experimental data include: pressure drop differential pressure, CHF limits, local profiles

of temperature and void fraction in the gap, and visualizations.


S. Rougé, “SULTAN test facility for large-scale vessel coolability in natural convection at low pressure”,

Nuclear Engineering and Design, Vol. 169, pp. 185-195, 1997

S. Rouge, I. Dor and G. Geffray, “Reactor vessel external cooling for corium retention SULTAN

experimental program and modeling with CATHARE code,” Workshop on In-Vessel Core Debris

Retention and Coolability, Garching, Germany, 1998 March 3-6

S. Rougé, A. Liégeois and A. Giacomelli, “Descriptif de la boucle SULTAN”, STR/LETC/94-224, 1994

Décembre

S. Rough and A. Carenza, “SULTAN TEST REPORT, Fourth Campaign, Inclination: 45 - Fluid depth:

15 cm”, SETEX/LTEMI98-97, 1998 September


Inclination: vertical (90°) and 10°

Gap size: 0.03 and 0.15 m

Outlet absolute pressure: 0.1 to 1 MPa;

Inlet temperature: 50 to 180°C;

Mass flux: 5 to 5000 kg/s-m2;

Heat flux: uniform, 0.1 to 1 MW/m2.



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Data intended to be used for the validation of the CATHARE code


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4.1.54 E1-54 - SBLB Boiling Tests

Test Facility: SBLB

Owner Organization: PSU (Penn State University)


Boiling experiments using test vessels with and without coatings were conducted in the PSU Subscale

Boundary Layer Boiling (SBLB) facility for the cases with and without an enhanced thermal insulation to

investigate the separate effect as well as the integral effect of the enhanced insulation design and vessel

coatings. The SBLB test facility consists of a water tank (with a diameter of 1.22 m and a height of 1.14

m) with a condenser assembly, a heated hemispherical test vessel with or without an insulation simulator, a

data acquisition system, a photographic system, and a power control system. For the steady-state tests, the

test vessel was comprised of two main parts made of aluminum: a segmented, heated lower hemispherical

vessel and a non-heated upper cylindrical portion. Cartridge heaters, 31.75 mm long and 9.52 mm thick,

were employed to provide independent heating of the segments. Heat flux levels of up to 1.2 MW/m2

could be achieved in a first phase of the work (Cheung, Haddad and Liu, 1997). Later, the range of heat

flux was increased (Cheng and Liu, 1998). The upper cylindrical portion had an outside diameter of 0.3 m

and a wall thickness of 12.7 mm. The tank was equipped with three immersion heaters with a total power

of 36 kW for preheating the water in the tank.

Various quenching and steady-state tests with a 3-D geometry were employed, and with various

subcoolings, and different geometries.


Cheung, F.B., Haddad, K. and Liu, Y.C., 1997, Critical Heat Flux (CHF) Phenomena on a Downward

Facing Curved Surface, NUREG/CR-6507, U.S. Nuclear Regulatory Commission, Washington, D.C.

Cheung, F. B. and Liu, Y. C., 1998, Critical Heat Flux (CHF) Phenomenon on a Downward Facing Curved

Surface: Effects of Thermal Insulation, NUREG/CR-5534, U.S. Nuclear Regulatory Commission,

Washington, D.C.

Cheung, F. B. and Liu, Y. C., 2001, Critical Heat Flux Experiments to Support In-Vessel Retention

Feasibility Study for an Evolutionary Advanced Light Water Reactor Design, EPRI Technical Report-

1003101

F. B. Cheung, “Limiting Factors for External Reactor Vessel Cooling”, Nuclear Technology, Vol. 152, pp.

145-161, 2005

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Geometries:

o 1) without insulation,

o 2) insulation based on the AP600 design, and

o 3) insulation based on the APR1400 design.

Pressure (free surface): 1 bar:

Water temperature: 90 to 100°C

Heat Fluxes: up to 1.2 MW/m2 (first phase, for which accessible documentation exists)

Heat flux distribution: not provided





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4.1.55 E1-55 – Small Scale Burst Test Experiments

Test Facility: SSBT



Small Scale Tube Burst Facility (SSBT) has been constructed to measure the effect of the pressure

wave initiated by an internally pressurised tube containing saturated fluid rupturing into cooler surrounding

water. The escaping fluid will flash to a two-phase steam-vapour mixture generating a high velocity

pressure wave that will interact with the surrounding structures and the containment vessel. The pressure

wave will produce severe damage in the immediate fuel channels and possibly initiate subsequent

pressurised tube ruptures. Three dimensional data collection of the pressure wave propagation is recorded

with strain measurements on the containment vessel and end plates. The small tube diameter experimental

data is used to provide validation of computer models which in turn can then be used to analyse the full

sized components.

Figure 4.1.55-1 Schematic of the Small Scale Burst Test Facility

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Leitch, B.W., Shewfelt, R.S.W. and Godin, D.P., “Two-phase Fluid/structure interactions in a bursting

tube”, AECL Report AECL RC-1711, COG Report COG-96-486, 1997.

Shewfelt, R.S.W., Leitch, B.W., and Godin, D.P., “Guillotine failure of fixed-end pipes, pressurised with

hot water”, AECL Conference AECL-10948, Int. Journal of Pressure Vessel and Piping, Vol. 57, pp. 211-

221, 1994

Shewfelt, R.S.W. and Godin, D.P., “Small-scale burst tests in air and water”, AECL Report RC-1454,

COG Report COG-95-356, 1995.


Pressure – 18 MPa maximum

Temperature – maximum tube 340ºC

Temperature – maximum moderator 90ºC

Year Tests Performed: 2004 to 2012



Prepared By: B.W. Leitch (AECL)

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4.2 Hydrogen Behaviour (Combustion, Mitigation and Generation) Experiments

4.2.1 E2-1 - LSVCTF S01

Test Facility: AECL-LSVCTF

Owner Organization: AECL / COG


A series of experiments were performed in AECL’s Large Scale Vented Combustion Test Facility

(LSVCTF). The LSVCTF is a 10 m long, 4 m wide and 3 m high rectangular enclosure with an internal

volume of 120 m3. Eight hydraulic fans are installed on the side walls to provide well-mixed gas mixtures

or produce initial turbulence.

The purpose of the test series S01 was to investigate the vented combustion behavior of dry hydrogen-

air mixtures. In this test series, the facility was configured with a single chamber. A TAYCO glow plug

was mounted in the centre of the vessel. Rectangular steel panels were removed from the front wall (wall

surface of 4 m by 3 m) to form the desired vent opening. The vents were covered with aluminum foil

during gas addition but they were easy to rupture (~1 kPa overpressure).

The initial hydrogen concentration was analyzed with a process Mass Spectrometer. Six dynamic

pressure transducers were used to record the combustion pressure. Thirty S-type fast response

thermocouples were mounted along three directions (toward/away the front cent, sideways, and

upward/downward) to detect the flame front. A total of 17 tests were performed under quiescent (mixing

fans turned off prior to ignition) conditions with hydrogen-air mixtures at room temperatures and

pressures.


Kumar, R.K., Loesel Sitar, J., Dewit, W.A., Bowles, E.M. and Thomas, B., “Experiments in the Large-

Scale Vented Combustion Test Facility: Series S01 – Quiescent Vented Combustion Tests with Central

Ignition in Hydrogen-Air Mixtures in the Full-Volume Geometry”, COG Report COG-96-578, 1997.

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Vent Area: 0.55, 1.1 2.2 m2

Initial Conditions:

o Hydrogen: 8 to 12%

o Temperature: 23-29°C

o Pressure: ~100 kPa

o Quiescent (mixing fans off)

Combustion overpressure: 1 to 40 kPa(g)




AECL performed GOTHIC validation using some of the tests:

Chan, C.K. and Chin, Y.S., “Validation of GOTHIC 7.2a for Modeling Combustion of Near-Flammability-

Limit Hydrogen-Air Mixtures in Closed and Vented Vessels”, COG Report ISTR-07-5036, 2008.

Prepared By: Z. Liang (AECL)

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4.2.2 E2-2 - LSVCTF S03




The purpose of the test series S03 was to investigate the vented combustion behavior of hydrogen-air-

steam mixtures. In this test series, the facility was also configured with a single chamber. One TAYCO

glow plug mounted in the centre of the chamber, or four spark ignitors mounted symmetrically about the

centre, were used. The vents configuration was the same as S01.

Gas composition and dynamic pressure measurements were the same as the S01 series. Eighty S-type

fast response thermocouples were mounted in the chamber. A total of 29 tests were performed under

quiescent or turbulent (mixing fans remained on during ignition) conditions with hydrogen-air-steam

mixtures at elevated temperatures and room pressures


Loesel Sitar, J.V., Dewit, W.A., Bowles, E.M., and Thomas, B., “Experiments in the Large-Scale Vented

Combustion Test Facility: Series S03-Vented Combustion Tests at 100C in Hydrogen-Air-Steam

Mixtures in the Full Volume Geometry”, COG Report COG-99-135, 2003.


Vent Area: 0.55 and 2.2 m2

Initial Conditions:


o Steam: 0, 10, 20, 30%

o Temperature: 75 to 100°C

o Pressure: atmospheric

o Quiescent and turbulent

Overpressure: 1.5 to 90 kPa(g)





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4.2.3 E2-3 - BMC Hx series

Test Facility: BMC



The experiment has been performed in the multicompartment Battelle Model Containment (BMC).

BMC was built from reinforced concrete, had a free volume of 640 m³, a height of 10 m and a diameter of

12 m. It was designed to be a 1/64 representation of the Biblis B containment.

42 hydrogen deflagration experiments have been performed in the Hx series. In these experiments,

the effects that compartment and vent opening geometry, initial hydrogen concentrations, and local steam

contents have on the combustion process were investigated. Different positions of ignitors have been

defined for the tests.


T. Kanzleiter, “Hydrogen Deflagration Experiments in a Multi-compartment Containment Geometry”,

Final Report, BIeV-R66.985-0, BMWi Germany, Nov. 1992

T. Kanzleiter et. all, “Wasserstoff-Deflagrationsversuche im Modellcontainment”, Fachberichte: BIeV-

R66.985-301 to -305

Kanzleiter, T.F. and Fisher, K.O., 1994, Multi-compartment hydrogen deflagration experiments and model

development, Nuclear Engineering and Design, 146, 417-426


Opening Area: 0.3 to 1.8 m2

Initial Conditions differs between ignition compartment and others:

o Hydrogen: 7.4 to 14%

o Steam: mostly 0%, some up to ~25%

o Temperature: atmospheric


Pressure peak: 200 – 300 kPa



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Figure 4.2.3-1 Hx-, Ix- and Gx-Test Geometries A to E

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Figure 4.2.3-2 Hx-, Ix- and Gx-Test Geometries G to K

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Table 4.2.3-1

H2-Deflagration Tests Performed (“Hx Tests”)

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Past Code Validation/Benchmarks: These experiments have been analyzed (pre- and post-test) using the

CONTAIN and BASSIM code. The later one was used to support the combustion model in CONTAIN.

RALOC was used in addition by others.


Final Report, BIeV-R66.985-0, BMWi Germany, November 1992


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4.2.4 E2-4 - BMC Ix series

Test Facility: BMC






25 ignitor tests have been performed in the Ix series for the German nuclear industry. In these

experiments the following effects have been studied: different positions of ignitors, influence of steam

content on ignition and flame propagation, stratified atmosphere conditions in ignition compartment, and

completeness of combustion. Experiments have been performed to check ignitor performance for

controlled ignition at about 10 vol.% hydrogen.

Information on room configuration can be found in Section 4.2.3 (E2-3 - BMC Hx series).





Opening Area: 0.3 to 1.8 m2


o Hydrogen: 9.7 to 13.8%

o Steam: mostly 0%, some up to ~40%



Pressure peak: up to 320 kPa

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Table 4.2.4-1

H2 Igniter Tests Performed (Utilities’ Program, “Ix Tests”)

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Past Code Validation/Benchmarks: These experiments have been analyzed (pre- and post-test) using the

CONTAIN and BASSIM code. The later one was used to support the combustion model in CONTAIN.

RALOC was used in addition by others.




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4.2.5 E2-5 - BMC Gx Series

Test Facility: BMC






8 different sub-series have been performed where a combination of PARs and ignitors under different

conditions have been tested. In these experiments the following effects have been studied: different

positions of PARs and ignitors, effect of early ignition at low hydrogen and efficiency of “dual concept”,

using PARs and ignitors together. The sub-series are structured as follows:

Gx2: 2 tests with one ignitor in each room

Gx3: 6 tests with different ignitor numbers

Ix11: 4 tests with different number of ignitors

Gx4: 2 tests with one PAR

Gx5: 2 tests with one PAR

Gx6: 1 test with one PAR

Gx7: 4 tests with one PAR and one ignitor

Gx8: 3 tests with one PAR and one ignitor

Information on room configuration can be found in Section 4.2.3 (E2-3 - BMC Hx series).

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Table 4.2.5-1

H2-Mitigation Tests Performed (“Gx Tests” and VGB Test Ix11)

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Final Report, BIeV-R66.985-0, BMWi Germany, Nov. 1992

T. Kanzleiter et. all, “Versuche zur Wirksamkeit von Wasserstoff-Gegenmaßnahmen in einer Mehrraum-

Containment-Geometrie”, Abschlussbericht: BIeV-R67.036-01 to -02, Nov. 1991


One room combination with 5 rooms was used for all tests


o Hydrogen injection rate up to 0.48 g/s

o Steam: up to 60%, but different in each room





Past Code Validation/Benchmarks: BASSIM code was used for analyses.


Final Report, BF-R68.145, BMWi Germany, 1994


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4.2.6 E2-6 - BMC Kx Series

Test Facility: BMC



Two experimental series have been performed in parallel, Kx series in an 8 m long deflagration tube

of 400 mm diameter and Ex series in a 10 m long BMC compartment of 4 m² area. In both test facilities

different obstacles have been placed to study its influence on the combustion processes. The same

experimental procedure and identically instrumentation and evaluation methods have been applied for

many experiments that were carried out. In addition, the Kx series aimed to study the effects prior to other

experiments in BMC facility, and therefore, more data are available from the Kx experiments than are

available from the Ex experiments.

Surprisingly, different results were obtained in the smaller scale Kx trials compared to the larger scale

Ex trials. The strength of the deflagration increases with the size and the three-dimensional character of

the combustion, but it is possible that there is a scaling effect that can be partially attributed to the different

heat losses to the structures, the differences in the turbulences of the unburned gas, and to the enlarged

reaction zone. Jet ignition behind obstacles and openings was studied as well in other tests in BMC.

Kx tests: about 50 different test have been performed. Tests have been performed without obstacles,

and with 5 different obstacles (grid, multiple rods, Blende, etc.)

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Table 4.2.6-1

Test Matrix for BMC Kx Series Tests

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Table 4.2.6-2

Initial Conditions for BMC Kx Series Tests

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Table 4.2.6-3

Initial Conditions for BMC Kx Series Tests (continued)

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Initial Conditions:

Hydrogen/steam concentration: 9% - 14% without steam; up to ~17% with steam of ~40%

Temperature: atmospheric or saturation



Repeatability Check: Yes partially





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4.2.7 E2-7 - BMC Ex Series

Test Facility: BMC



Two experimental series have been performed in parallel, Kx series in an 8 m long deflagration tube

of 400 mm diameter, and Ex series in a 10 m long BMC compartment of 4 m² area. In both test facilities,

different obstacles have been placed to study its influence on the combustion processes. The same

experimental procedure and identically instrumentation and evaluation methods have been applied for

many experiments.

Surprisingly, different results were obtained in the smaller scale Kx trials compared to the larger scale

Ex trials. The strength of the deflagration increases with the size and the three-dimensional character of

the combustion, but it is possible that there is a scaling effect that can be partially attributed to the different

heat losses to the structures, the differences in the turbulences of the unburned gas, and to the enlarged

reaction zone. Jet ignition behind obstacles and openings was studied as well in other tests in BMC.

Ex tests: about 29 different test have been performed. Tests have been performed without obstacles,

and with different obstacles (grid, multiple rods, vertical or horizontal cylinder, etc.)





Initial Conditions:

Hydrogen/steam concentration: 9% - 11% without steam;

Temperature: atmospheric


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Table 4.2.7-1

Test Matrix for BMC Ex Series Tests


Repeatability Check: Yes partially





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4.2.8 E2-8 - ENACEFF SARNET2 Tests

Test Facility: ENACCEF

Owner Organization: IRSN and CNRS


ENACCEF is a 5 m high, vertical facility, divided into 2 parts, and can be equipped with repeated

obstacles in the bottom portion. It is divided in 2 parts:

The acceleration tube (3.2 m long and 154 mm i.d.), is equipped at its bottom-end with 2 tungsten

electrodes as a low energy ignition device. At a distance of 1.9 m from the ignition point, 3

rectangular quartz windows (40 mm x 300 mm optical path) are mounted flush with the inner

surface, 2 of them are opposed to each other the third one being perpendicular to the others. These

windows allow the recording of the flame front during its propagation along the tube using either a

shadowgraph or a tomography system. The tube is also equipped with 11 small quartz windows

(optical diameter: 8 mm, thickness: 3 mm) distributed along its length,

The dome (1.7 m long, 738 mm i.d.) is connected to the upper part of the acceleration tube via a

flange. This part of the facility is also equipped with 3 silica windows (optical path: 170 mm,

thickness: 40 mm), perpendicular to each other, 2 by 2. Through these windows, the arrival of the

flame can be recorded via a Schlieren or a tomography system.

The ENACEFF tests provided to SARNET2 are test RUN 153; RUN 158 and RUN 160.


N. Chaumeix, A. Bentaib, SARNET2 project- hydrogen deflagration benchmark, specification report,

IRSN/DSR 102, 2010.


The initial conditions are:

o Temperature = 23°C

o Pressure = 1 bar

o Initial hydrogen concentration = 13%

o Initial air concentration = 87%

Blockage ratio:

o RUN 153 : BR =0.63

o RUN 158 : BR =0.33

o RUN 160 : BR =0



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A. Bentaib, SARNET2 project- hydrogen deflagration benchmark, final report, IRSN/DSR

Prepared By: A. Bentaib (IRSN)

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4.2.9 E2-9 - ENACEFF SARNET Test (Run 703)




A description of the experimental facility is provided in Section 4.2.8 (E2-8 - ENACEFF SARNET2

Tests). The test conditions are described below in the “Range of Key Experimental Parameters”.


N. Chaumeix et al., “Mesures des Vitesses Spatiales de Propagation de Flamme dans une ENceinte

d'ACCElération de Flamme. Influence des inhomogénéités de concentration d'hydrogène sur la

propagation d'une flamme en présence d'obstacles”, rapport IRSN-CNRS




o Pressure = 1 bar

o Initial hydrogen concentration = linear gradient from 13% to 10.5%

o Initial air concentration = from 87% to 89.5%

Blockage ratio: BR =0.63


Repeatability Check: Yes (at least 3 times)


A. Bentaib et al., Hydrogen combustion with concentration gradients in experiments and simulations:

preliminary results of ENACCEF Benchmark, ERMSAR-2007


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4.2.10 E2-10 - ENACEFF SARNET Test (Run 717)







N. Chaumeix et al., Mesures des Vitesses Spatiales de Propagation de Flamme dans une ENceinte

d'ACCElération de Flamme. Influence des inhomogénéités de concentration d'hydrogène sur la

propagation d'une flamme en présence d'obstacles, rapport IRSN-CNRS




o Pressure = 1 bar

o Initial hydrogen concentration = linear gradient from 10.45% to 13.03%

Blockage ratio: BR = 0.63




A. Bentaib et al., Hydrogen combustion with concentration gradients in experiments and simulations:

preliminary results of ENACCEF Benchmark, ERMSAR-2007


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4.2.11 E2-11 - ENACEFF Run 765 (ISP-49)







N. Chaumeix and A. Bentaib, ISP 49 – Specification of ENACCEF test Flame Propagation in a Hydrogen

Gradient, rapport IRSN/DSR 40




o Pressure = 1 bar






Kotchourko et al., “ISP-49 on Hydrogen Deflagration”, Final Report, NEA/CSNI/R/(2011)9, 2011.


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4.2.12 E2-12 - ENACEFF Run 736 (ISP-49)











o Temperature = 23°C,

o Pressure = 1 bar






Kotchourko et al., “ISP-49 on Hydrogen Deflagration”, Final Report, NEA/CSNI/R/(2011)9, 2011


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4.2.13 E2-13 - ENACEFF Run 733 (ISP-49)












o Pressure = 1 bar

o Initial hydrogen concentration = linear gradient from 5.7% to 12%







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4.2.14 E2-14 - DRIVER HYCOM MC 003

Test Facility: DRIVER facility

Owner Organization: EC


Combustion experiments were carried out in an obstructed tube with 174 mm internal diameter and

12.2 m length (DRIVER facility). Repeatable obstacles with blockage ratio (BR) of 0.6, and at distances

equal to the internal diameter. Hydrogen/air mixture with concentration of 10% hydrogen was tested.


Nuclear Engineering and Design Volume 235, Issues 2-4, February 2005, pp. 253-270


Initial Pressure = 1 bar

Temperature = 23°C

Initial %H2 = 10%

BR=0.6


Repeatability Check: Yes (each 2 times)


Prepared By: A. Kotchourko (KIT)

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12.2 m length (DRIVER facility). Repeatable obstacles with blockage ratio (BR) of 0.6, and at distances

equal to the internal diameter. Hydrogen/air mixture with concentration of 13% hydrogen was tested.


Integral large scale experiments on hydrogen combustion for severe accident code validation-HYCOM,




Temperature = 23°C

Initial %H2 = 13%

BR = 0.6


Repeatability Check: Yes (each 2 times)



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4.2.16 E2-16 - FZK R 0498_09

Test Facility: FZK 12 m tube

Owner Organization: KIT


Combustion experiments were carried out in an obstructed tube with 350 mm internal diameter and 12

m length (FZK Tube). Repeatable obstacles with blockage ratio (BR) of 0.3, and at distances equal to the

internal diameter. Hydrogen/air mixture with concentration of 15% hydrogen.

Data available by request from KIT representative


A. Kotchourko, W. Breitung, A. Veser, S.B. Dorofeev Tube Experiments and Numerical Simulation on

Turbulent Hydrogen-Air Combustion 21st Int. Symposium on Shock Waves, Great Keppel Island,

Australia, July 20-25, 1997, p. 82



Temperature = 23°C

Initial %H2 = 15%

BR = 0.3




Similar experiment: A. Kotchourko et al., Tube Experiments and Numerical Simulation on Turbulent

Hydrogen-Air Combustion. 21st Int. Symp. on Shock Waves, Great Keppel Island, Australia, July 20-25,

1997


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12.2 m length (DRIVER facility). Repeatable obstacles were present at distances equal to the internal

diameter. The experimental tube was divided in two equal parts by thin polyethylene membrane (1 μm)

with different blockage ratios and different hydrogen concentrations. Critical pressure for membrane

breaking was found to be about 5 Torr. Hydrogen concentration/(BR) for Part 1 was 13% / 0.6, and for

Part 2 was 10% / 0.3. Ignition was carried out at the tube end in part 1.

Flame acceleration was observed in the presence of non-uniform initial conditions, accounting for:

Effect of concentration gradients (transient of accelerated flame from rich to lean mixture)

Effect of blockage ratio changes (transient of accelerated flame to the area with lower degree of

obstruction)






Temperature = 20°C

Initial %H2 = 13% - 15%

BR = 0.3 - 0.6





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4.2.18 E2-18 - DRIVER HYCOM HC 020




Combustion experiments were carried out in non-uniformly obstructed tube with a 12.4 m total

length, and constructed of two parts with internal diameters of 174 and 520 mm respectively. Obstacles

were spaced at the tube diameter, with the diameter / (BR) for Part 1 being 174 mm / 0.6, and for Part 2

being 520 mm / 0.3. The combustion of a uniform test mixture with 10% hydrogen in air was investigated,

and the ignition point was in I2 (part 2).

Hydrogen deflagration was observed in this case, accounting for:

Effect of tube diameter change (flame propagating from higher to less tube size)

Effect of blockage ratio changes (flame propagating to the area with more degree of obstruction)

Effect of pre-compression in non-uniform geometry






Temperature = 20°C

Initial %H2 = 10%

BR = 0.3 - 0.6





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4.2.19 E2-19 - DRIVER HYCOM-HC027




Combustion experiments were carried out in non-uniformly obstructed tube with a 12.4 m total

length, and constructed of two parts with internal diameters of 174 and 520 mm respectively. Obstacles

were spaced at the tube diameter, with the diameter / BR for Part 1 being 174 mm / 0.3, and for Part 2

being 520 mm / 0.6. The sonic combustion of a uniform test mixture with 13% of hydrogen in air was

investigated, and the ignition point was in I1 (Part 1)

Hydrogen deflagration was observed in this case, accounting for:

Effect of tube diameter change (flame propagating from less to higher tube size)

Effect of blockage ratio changes (flame propagating to the area with lower degree of obstruction)

Flame propagation in the presence of non-uniform geometry


Integral large scale experiments on hydrogen combustion for severe accident code validation, EU Project

HYCOM, Final report, 2003



Temperature = 20°C

Initial %H2 = 13%

BR = 0.3 - 0.6





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4.2.20 E2-20 - RUT HYC01

Test Facility: RUT



Combustion experiments were carried out in a large scale multi-compartment geometry, consisting of

a curved channel (2.3x2.5 m cross-section and 15.5 m length) and a canyon (6.3x2.5x16.4 m). Four

repeatable obstacles with blockage ratio BR = 0.3 were installed in the channel, and two obstacles were

installed in the bottom part of canyon. Uniform hydrogen/air mixture with a concentration of 10%

hydrogen was tested, and the ignition was in the channel (I1).






Temperature = 18°C

Initial %H2 = 10%

BR = 0.3







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4.2.21 E2-21 - RUT HYC12

Test Facility: RUT





repeatable obstacles with blockage ratios BR = 0.3 were installed in the channel. The canyon has been

divided into four separate rooms, and connected with orifices. A uniform hydrogen/air mixture with a

concentration of 11.5% hydrogen was tested, and ignition was in the channel (I1).


Integral large scale experiments on hydrogen combustion for severe accident code validation, EU Project

HYCOM, Final report, 2003



Temperature = 18°C

Initial %H2 = 11.5%

BR = 0.3





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4.2.22 E2-22 - RUT HYC14

Test Facility: RUT





repeatable obstacles with blockage ratios BR = 0.3 were installed in the channel. The canyon has been

divided into four separate rooms, and connected with orifices. A uniform hydrogen/air mixture with a

concentration of 11.5% hydrogen was tested, and ignition was in the channel (I1).






Temperature = 18°C

Initial %H2 = 11.5%

BR = 0.3







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4.2.23 E2-23 - VGES Tests

Test Facility: VGES

Owner Organization: Sandia National Laboratories


The Variable Geometry Experimental System (VGES), located at Sandia National Laboratories, was

used to conduct research into the behaviour and control of hydrogen during accidents at nuclear power

plants. Eleven test series, consisting of over 100 experiments, were performed in the 5 m3 Burn Tank

(cylindrical pressure vessel). The overall objectives were to evaluate:

Combustion of low and high hydrogen concentrations,

Effect of raising glowplug ignitor location,

Effect of ignitor type,

Effect of initial pressure on the combustion overpressure,

Effect of nitrogen concentration and steam (simulated with carbon dioxide),

Equipment survivability, and

Hydrogen burns in aqueous foam environment.

The experimental procedure for each test was as follows:

After the tank was sealed, the fans were turned on for about 10 min to eliminate any thermal

stratification of the air.

Nitrogen, carbon dioxide, or hydrogen was admitted to the tank until a predetermined pressure was

reached. For the tests with nitrogen and CO2, these gases were admitted to the tank before the

hydrogen.

The fans were left on for another 10 min to ensure complete mixing of the tank atmosphere. Then

the pre-burn gas sample was taken.

For the quiescent burns, the fans were turned off about 10 min prior to ignition. Although the fan

blades were constructed of black plastic, very little melting or deformation was observed after the

quiescent burns. The same fans were used for all those quiescent tests.

The turbulent or “fans on” burns were ignited while the fans were still running. Unlike the

quiescent burn tests, the heat transfer was so greatly enhanced during the turbulent burns that the

fan blades were melted after a single shot and the fans had to be replaced after nearly every burn.

For the foam tests, the hydrogen and air concentrations were obtained using the procedure

previously discussed. With the fan still running, the surfactant-water mixture contained in a small

pressurized tank was added to the foam generator. Foam was generated until a visual inspection of

the tank, through the top flanged Lexan window, indicated the tank was full of foam. The fan and

foam generator were then turned off prior to ignition.


Bendick, W.B., Cummings, J.C. and Prassinos, P.G., “Combustion of Hydrogen: Air Mixtures in the

VGES Cylindrical Tank”, NUREG/CR-3273, 1984.

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Parameters and Variables (total min-max from experiments):

Burn Tank Vol: 5 m3

Hydrogen conc.: 3.8-23.4%

Nitrogen conc.: 27.4-82.7%

CO2 conc: 0-56%

Igniter type: 300-W photoflood lamp, 14-V glow plug (GM7G) and raised spark-gap

Igniter location: Figure 4 of NUREG/CR-3273

Pre-combustion gas motion

Pre-combustion gas pressure

Presence of aqueous foam

Year Tests Performed: ~1980s




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4.2.24 E2-24 - NTS Tests

Test Facility: NTS

Owner Organization: US DOE


A series of premixed combustion experiments were performed at the Nevada Test Site (NTS). The

tests were performed in a 2048 m3 vessel with hydrogen concentrations ranging from 5 to 13% (by

volume) and steam ranging from 4 to 40%. The objective was to study the combustion process in a large-

scale vessel and to evaluate associated safety-related equipment response to the resulting thermal

environments. (24 tests conducted)


A.C. Ratzel, “Data Analyses for Nevada Test Site (NTS) Premixed Combustion Tests,” NUREG/CR-4138,

May 1985.


Hydrogen concentration: 5 to 13 vol.%

Steam concentration: 4 to 40 vol.%

Spray or/and fan (2.4 m3/s or 5000 cfm)




A.C. Ratzel, “Data Analyses for Nevada Test Site (NTS) Premixed Combustion Tests,” NUREG/CR-4138,

May 1985.


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4.2.25 E2-25 - PET Tubes

Test Facility: PET Tube



Studies of gaseous explosions in vented tubes with large vent ratios can be considered as a bridge

between cases of explosions in closed tubes and cases of unconfined explosions in congested areas. The

critical conditions for flame acceleration and DDT in the latter situation are much less understood

compared to those in closed systems.

The PET-Tube (see Figure 4.2.25-1) allows experiments on the deflagration of different homogeneous

hydrogen-air or hydrocarbon-air mixtures (CH4, C3H8) in a tube with adjustable transverse venting varying

from 0% to 40%, which enables a comparison of terminal flame speed and flame acceleration conditions

for such mixtures. It is designed for studying flame propagation and transition to detonation in:

1. closed tubes,

2. vented tubes surrounded by air, and

3. vented tubes surrounded by combustible mixture.

The data generated in such experiments can be used for the validation of computer codes for

simulation of gaseous explosions in semi-confined geometries.

The facility consists of an explosion tube with movable brackets to adjust vent ratio, a support

construction, a control system, and a measurement system. A schematic of the facility in Configuration 1

(vented tube with combustible gas surrounded by air) is shown in Figure 4.2.25-2.

The main part of the facility is a steel explosion tube of 100 mm i.d., which has a length of 7 m. It

consists of three main sections, each 2.1 m long (see Figure 4.2.25-2) and two additional sections, 0.22 m

length, at each end. Each main section of the tube has 16 rectangular openings of variable size. The two

end sections represent pieces of closed tubes. Circular orifice plate obstacles are installed along the entire

length of the tube. The distance between the obstacles is equal to one tube diameter. Different sets of the

orifice plates are available with blockage ratios (BR) equal to 0.3, 0.45, and 0.6.

A schematic of the facility in Configuration 2 (vented tube inside the combustible gas) is shown in

Figure 4.2.25-3. In this configuration, the vented explosion tube with fixed vent ratio is placed into a

cylindrical plastic bag. The diameter of the bag is about 400 mm. The supports of the tube are provided

with two discs at the ends, which are used to fix the plastic bag hermetically. The discs are equipped with

hermetic penetrations for the measuring cables.

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2) Explosions in vented tubes

1) Confined explosions in tubes

3) Explosions vented into combustible gas

Figure 4.2.25-1 Schematic Illustration of the 3 Cases to be Investigated in the PET-tube

Figure 4.2.25-2 Schematic of PET Facility in Configuration 1

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Figure 4.2.25-3 Schematic of PET Facility in Configuration 2

The measurement system consists of a subsystem for mixture composition measurements and a

subsystem for measurements of explosion parameters. To measure the uniformity of the mixture

composition, a Rosemount-Fischer MLT4-gas-analyzer is used. Relative accuracy for differential

measurements is about 1% of the mean value. Measurements of explosion parameters include 16

collimated time-of-arrival photodiodes and 16 pressure transducers located in the closed parts of the tube.

In Configuration 2, additional pressure transducers are installed outside of the tube at distances from 1 m to

15 m (Figure 4.2.25-3). They are located at the same height (1 m above the ground level) as the axis of the

main tube. These pressure transducers are used to record parameters of the air blast wave generated by

semi-confined explosions.


Sergey Dorofeev, Anke Veser, Ulrich Bielert, Wolfgang Breitung, Alexei Kotchourko, Report on partially

vented explosion tube (PET) European Integrated Hydrogen Project –Phase II; CONTRACT N°: ENK6-

CT2000-00442


Fuel concentration: 0 to 100%

BR: 0.3, 0.45 and 0.6

vent ratio: 0% to 40%


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Past Code Validation/Benchmarks: No

Prepared By: A. Veser (Pro-Science)

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4.2.26 E2-26 - THAI HD Series (Combustion Tests)

Test Facility: THAI




volume of 60 m3. For the hydrogen deflagration test the vessel had a free volume (one without internal

structures). The upward and downward propagation of flames in premixed and in stratified air-steam-

hydrogen atmospheres was investigated. The test matrix provides a systematic variation of initial pressure,

initial temperature, steam content, and spatial gas distribution. The influence on pressure built up,

temperature development, flame front propagation and completeness of combustions was quantified.

The test results provide additional data for an improved understanding of hydrogen combustion

phenomena (since hydrogen deflagrations cannot be ruled out completely even by use of PAR) and for the

further development and validation of containment system codes. They have enlarged the existing

database because the test conditions are typical for severe accidents and the facility is relatively large,

which allows combustion in upward and downward direction, since experimental data have existed mainly

for horizontal flame propagation and/or small geometries, which provide non-conservative data. In large

geometries, deflagrations occur faster than in small geometries due to things like increased turbulence

generation, and therefore produce higher loads for the vessel and its internals.

The experimental results of some selected tests have been used for open and blind post-test

calculations by various institutions of the member countries.

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Figure 4.2.26-1 THAI HD-tests Instrumentation

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Kanzleiter, Langer: “Hydrogen Deflagration Tests in the THAI Test Facility”, Becker Technologies

GmbH, Germany, Report No. 150 1326-HD-2, January 2010

“OECD/NEA THAI Project Hydrogen and Fission Product Issues Relevant for Containment Safety

Assessment under Severe Accident Conditions” Final Report, June 2010, NEA/CSNI/R(2010)3


Hydrogen concentration 6 – 12 Vol.%

Initial pressure 1 – 1.5 bar

Initial Temperature 25 – 140°C

Steam concentration 0 – 50 Vol.%

Mixed and stratified atmosphere

Upward and downward Flame propagation


Repeatability Check: Some test repeated




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4.2.27 E2-27 - THAI HR Series (PAR Tests)

Test Facility: THAI




volume of 60 m3. For the PAR (passive autocatalytic recombiner) – tests it contained only the lower part

of the inner cylinder. On the outside of the inner cylinder the PAR was attached, with its inlet at about 2 m

elevation. At 1.26 m hydrogen was injected through a ring feed line. With increasing hydrogen

concentration the onset of recombination was reached. By continuing the injection the hydrogen

concentrations increased further. After stopping the injection the PAR performance could be studied under

decreasing hydrogen and oxygen concentrations. In several tests deflagrations occurred.

The tests have provided additional data about the PAR behaviour under accident typical conditions.

Since the experimental data of past investigations have been partially proprietary, i.e., not publically

available, experiments to fill this gap have been agreed upon. Commercial PAR types (provided by the

companies AREVA, AECL, NIS) have been tested with focus on:

conditions for the start of catalytic reaction (onset of recombination),

recombination rate, depending on hydrogen concentration, pressure, steam content,

PAR operation under lack of oxygen (oxygen starvation),

conditions for ignition by PAR, resulting in hydrogen deflagration in the vessel volume.

The obtained data improved the understanding of PAR behaviour in general and also of the

characteristics of the different PAR types. Furthermore, the data were used as a basis for modelling the

PAR behaviour and PAR-initiated deflagrations.

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Figure 4.2.27-1 THAI HR-tests: General Experimental Set-up

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H2 concentration 0 – 9 Vol.%

Initial Pressure 1 – 3 bar

Steam concentration 0 – 60 Vol.%

Initial Temperature 25 - 117°C


Repeatability Check: Some test were repeated

Past Code Validation/Benchmarks: Used at GRS for COCOSYS and CFX PAR model validation


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4.2.28 E2-28 - THAI Hydrogen Combustion During Spray Operation

Test Facility: THAI

Owner Organization: BMWi / OECD


Several test runs are planned within the current OECD THAI2 program to investigate the effect of

containment spray operation on vertical hydrogen burns, by comparing the results to corresponding

deflagration tests without spray available from the first OECD-THAI program. The experimental data will

support model development and validation in the areas “containment spray” and “hydrogen combustion”.

The THAI test vessel (V = 60 m3, H = 9.2 m, D = 3.2 m, Pmax = 15 bar) is equipped with spray

nozzle(s) at its top, spark igniters at its bottom and its top, and with supply systems for air, steam,

hydrogen and spray water, including flow and temperature measurements. A fan is available to provide

homogeneous gas and steam distribution in the preconditioning phase. Droplet size spectrum is provided

by the spray nozzle manufacturer.

Initial local hydrogen concentration is measured by a total of 15 sampling systems with heat-

conductivity and hydrogen sensors operating in parallel. Flame front development is detected by a matrix

of 43 “fast” thermocouples at 12 elevations. Six more thermocouples installed in small drain pans monitor

spray water temperature at different elevations. Two “slow” and three “fast” pressure transducers measure

initial pressure, pressure transient and possible pressure wave effects.

A test matrix consisting of three test runs is proposed in the table below.

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Figure 4.2.28-1 THAI Hydrogen Combustion During Spray Operation

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Tests are part of ongoing OECD THAI 2 project. Report not yet provided.


Test ID Pressure Temperature Steam

content

H2 content Spray characteristics

Water flow

rate

Water

temperature

HD 30 1.5 bar 20C 0 vol.% 10 vol.% 1 kg/s 20C




Repeatability Check: preparatory tests done



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4.2.29 E2-29 - DFF SFSER01

Test Facility: AECL-Diffusion Flame Facility

Owner Organization: AECL/COG


A series of experiments were performed in AECL’s Diffusion Flame Facility (DFF) to examine the

structure of horizontal hydrogen-steam diffusion flames. Five orifice break configurations were used to

create flames of different sizes and shapes. Three were circular in shape, with diameters of 6.35, 12.7 and

19.0 mm respectively. The other configurations were a slit (3 mm x 51 mm) and a cat-eye-shaped opening

(1.8 mm x 61 mm). Experiments were conducted both in ambient air (dry) and in steam-air (wet)

atmospheric conditions. The steam composition in the jet mixture was varied from 0% to 90% by volume,

with jet velocities ranging from 100 to 500 m/s. Stability regimes, penetration depth (flame lengths),

transverse and longitudinal temperature profiles and peak time-average flame temperatures were

determined for various jet compositions and jet velocities.

Measurements include jet velocity/diameter and gas composition, flame temperature along the centre

line of the flame axis and along the transverse line perpendicular to the axis, and flame length by camera.


Liang, Z., “A Consolidation Report on the Hydrogen-Steam Diffusion Flame Experimental Program at the

Whiteshell Laboratories”, COG Report COG-00-245-R1, AECL Report 153-126510-COG-011, October

2009.

Guerrero, A.M., and Chan, C.K., “The Structure of Horizontal Hydrogen-Steam Diffusion Flame Data

Report for Experiment Series DFSER01”, COG Report COG-97-030, 2000 (draft).


Hydrogen: 8 to 12%

Ambient temperature: ~25°C

Ambient pressure: ~100 kPa

Steam concentration of the jet: 0-90 vol.%

H2 concentration of the jet: 10-100%

Jet speed: 100, 200, 300, 400, 500 m/s

Jet diameter: 6.35, 12.7 and 19.0 mm



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4.2.30 E2-30 - LSVCTF S02




Experiments were performed in AECL’s Large Scale vented Combustion Test Facility (LSVCTF) in

the 120 m3 volume. The purpose of the test series S02 was to investigate the effect of ignitor location

relative to the vent (near- and far-vent) on vented combustion behavior of dry hydrogen- air mixtures.

Measurements included dynamic pressure and flame front location (with “fast” temperature sensors). A

total of 19 tests were performed under quiescent conditions with hydrogen-air mixtures at room

temperatures and pressures.


Kumar, R.K., “Experiments in the Large-Scale Vented Combustion Test Facility: Effects of Igniter

Location on the Combustion Behaviour in Quiescent Hydrogen-Air Mixtures” COG Report COG-97-226,

2007.


Vent Area: 0.55, 1.1 2.2 m2

Initial Conditions:




o Quiescent (mixing fans off)

Combustion overpressure: 1 to 60 kPa(g)





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4.2.31 E2-31 - LSVCTF DC




Experiments were performed in AECL’s Large Scale vented Combustion Test Facility (LSVCTF) in a

two-volume configuration (~60 m3 each). The purpose of the test series DC was to investigate the vented

combustion behavior in two interconnected volumes. Measurements included dynamic pressure and flame

front location (with “fast” temperature sensors). Over 46 tests were performed under quiescent/turbulent

conditions with hydrogen-air steam mixtures at room (or elevated) temperatures and pressures.


Loesel Sitar, J.V., and Chan, C.K., “Hydrogen-Air-Steam Large-Scale Vented Combustion Tests in the

Double-Chamber Geometry”, COG Report COG-00-244, 2003


Vent Area: 0.38 m2 (internal), 0.55/1.1 m

2 (external)

Initial Conditions:

o Hydrogen: 6-10% (downstream), 0-10% (upstream)

o Steam: 0-30%



o Quiescent/turbulent

Combustion overpressure: up to 130 kPa(g)



Past Code Validation/Benchmarks: AECL performed GOTHIC validation with some tests:

Liang, Z., “GOTHIC 7.2a Guidelines for Multi-Volume Vented Combustion of Near-Flammability Limit

H2-Air Mixture”, COG Report ISTR-09-5007, 2010.


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4.2.32 E2-32 - LSVCTF 3C




Experiments were performed in AECL’s Large Scale vented Combustion Test Facility (LSVCTF) in a

three-volume configuration (~25, 28, 60 m3 each). The purpose of the test series DC was to investigate the

vented combustion behavior in three interconnected volumes. Measurements included dynamic pressure

and flame front location (with “fast” temperature sensors). A total of 24 tests were performed under

quiescent conditions with hydrogen-air mixtures at room temperatures and pressures by placing the ignitor

in different rooms.


McIlwain, H., and Loesel Sitar, J.V., “Experiments in the Large-scale Vented Combustion Test Facility:

Data Report for Quiescent Hydrogen-Air Mixtures in Three-Chamber Geometry”, COG Report COG-02-

2028, 2003.


Vent Area: 0.38-0.4 m2 (internal), 1.1 m

2 (external)

Initial Conditions:

o Hydrogen: 7, 8 9%



o Quiescent

Combustion overpressure: up to 130 kPa(g)



Past Code Validation/Benchmarks: AECL performed GOTHIC validation using some tests:

Liang, Z., “GOTHIC 7.2a Guidelines for Multi-Volume Vented Combustion of Near-Flammability Limit

H2-Air Mixture”, COG Report ISTR-09-5007, 2010.


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4.2.33 E2-33 - LSVCTF CIC




Experiments were performed in AECL’s Large Scale vented Combustion Test Facility (LSVCTF) in

the 60 m3 volume (front half of the total chamber) with an operating ignitor and continuous hydrogen gas

injection. The parameters that were varied during testing include injection rate/location, ignitor location,

and turbulence level. Measurements included dynamic pressure and flame front location (with “fast”

temperature sensors). A total of 24 tests were performed under quiescent conditions with hydrogen-air

mixtures at room temperatures and pressures by placing the ignitor in different rooms.


Loesel Sitar, J.V., “Continuous Injection of Hydrogen into a Test Chamber with Operating Ignitors”, COG

Report COG-01-206, 2003.


Vent Area:2.2 m2

Initial Conditions:

o Injection rate: 10-23 g/min

o Injection location: bottom, top. Mid

o Ignitor location: top/mid

o Temperature: 50°C


o Quiescent/turbulent

o Max. H2 in the chamber: 6-17%





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4.2.34 E2-34 - Gammacell Radiolysis Tests

Test Facility: AECL Gamamcell



The purpose of the tests was to measure the accumulation of hydrogen in the airspace of an irradiated

glass cell containing room temperature air and water. The radiation source was a Co-60 Gammacell that

provided 4-5 kGy/h. Tests had durations ranging from about 20 to 300 hours. Gas samples were

periodically removed using a gas-tight syringe and the hydrogen concentration was determined by gas

chromatography. The results show that nitric acid formation in the airspace affects the accumulation of

hydrogen. The tests explore the effect of dose rate, geometry, pH, and various other additives on hydrogen

production. The results can be used to develop and verify a radiolytic hydrogen formation model.


Glowa, G.A., Wren, J.C., and Mitchell, J.R.D., 2004. Modelling Radiolytic Hydrogen Formation: Analysis

of Bench Scale Data. CANDU Owners Group Inc. Report, COG-03-2067.

G.A. Glowa, J.C. Wren and J.D. Mitchell, 2003. Experimental Studies on Radiolytic Hydrogen Production,

CANDU Owners Group Report, COG-02-2134.

Glowa, G.A., Wren, J.C., and Mitchell, J.R.D., 2004. Modelling Radiolytic Hydrogen Formation: Analysis

of Bench Scale Data. CANDU Owners Group Inc. Report, COG-03-2067.

McCracken, D.R., Stuart, C.R., Ouellette, D.C., Shultz, C.M., 1998. The Radiolysis of Oxygenated Water:

Benchmark Experiments 1. Atomic Energy of Canada Limited Report, RC-2145.

McCracken, D.R., Shultz, C.M., 1998. The Radiolysis of Water Dosed with Hydrogen and Oxygen:

Benchmark Experiments 2. Atomic Energy of Canada Limited Report, RC-2152.

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Room Temperature

Dose Rate: ~4 kGy/h

Pure water, and water with various additives (CsI, nitrate, metal ions, LiOH, organic impurities,

O2)


Repeatability Check: Some


Prepared By: G.A. Glowa (AECL)

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4.2.35 E2-35 - LACOMECO UFPE2

Test Facility: HYKA-A2



Large scale combustion experiment was performed in a vertical cylindrical volume of 220 m3 with

aspect ratio of H/D =1.5 (H = 9 m, D = 6 m). A uniform hydrogen-air-steam mixture with 10% hydrogen

was axially ignited from the bottom (hign = 1.5 m above the floor). Flame propagation regime and effects

of instabilities on flame development in an unobstructed volume were investigated using high speed

imaging, thermocouples and pressure sensors. Scaling-down of hydrogen combustion phenomena in a

containment of nuclear reactor for numerical code validations was the main goal of the test.

Figure 4.2.35-1 HYKA-A2 Facility

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A. Miassoedov, M. Kuznetsov, M. Steinbrück, S. Kudriakov, Z. Hózer, I. Kljenak, R. Meignen, J.M.

Seiler, A. Teodorczyk, “Experiments of the LACOMECO Project at KIT”, Proc. of the 5th European

Review Meeting on Severe Accident Research (ERMSAR-2012), Cologne (Germany), March 21-23, 2012


Initial Pressure = 1.5 bar

Temperature = 90°C

Hydrogen concentration: 10 vol.%

Steam concentration: 25 vol.%




Prepared By: M. Kuznetsov (KIT)

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4.2.36 E2-36 - LACOMECO HYGRADE10




A large scale combustion experiment with hydrogen concentration gradient in a vertical cylindrical

volume of 33 m3 with aspect ratio of H/D =3.4 (H = 8 m, D = 2.35 m) was performed. A non-uniform

hydrogen-air mixture with a linear vertical hydrogen concentration gradient, ranging from 5% (at the

bottom) to 13% hydrogen (at the top), in air was axially ignited from the top (hign = 7 m). An obstacle

array with a blockage ratio BR of 0.5 was installed inside the volume.







Temperature = 17°C

Hydrogen concentration gradient: 5-13 vol.% H2Ignition at the top (hign = 7 m)





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Figure 4.2.36-1 HYKA-A3 Facility

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A large scale combustion experiment with hydrogen concentration gradient in a vertical cylindrical

volume of 33 m3 with aspect ratio of H/D =3.4 (H = 8 m, D = 2.35 m) was performed. A non-uniform

hydrogen-air mixture with a linear vertical hydrogen concentration gradient from 5% (at the bottom) to

13% hydrogen (at the top) in air was axially ignited from the bottom (ignition height of 1.2 m). An

obstacle array with blockage ratio BR of 0.5 was installed in the volume. The same test facility used as for

the test E2-36 - LACOMECO HYGRADE10.







Temperature = 13°C

Hydrogen concentration gradient: 5-13 vol.% H2

Ignition at the bottom (hign = 1.2 m)





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A large scale combustion experiment with uniform hydrogen-air mixture in a vertical cylindrical

volume of 33 m3, with aspect ratio of H/D =3.4 (H = 8 m, D = 2.35m), was performed. A uniform

hydrogen-air mixture was axially ignited from the bottom (hign = 1.2 m). An obstacle array with blockage

ratio BR = 0.5 was installed in the volume. The same test facility used as for the test E2-36 -

LACOMECO HYGRADE10.







Temperature = 5°C

Hydrogen concentration 9 vol.% H2Ignition at the bottom (hign = 1.2 m)





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4.2.39 E2-39 - LACOMECO HYDET06




Detonation propagation experiments in a stratified semi-confined layer of hydrogen-air mixture were

performed. A horizontal rectangular box of 9x3x0.6 m for hydrogen-air mixture was installed inside of

safety volume A1 (V=100 m3). A non-uniform hydrogen-air mixture, with a vertical concentration

gradient of 1.1% H2/cm, and 26% hydrogen in air at the top boundary, was used in the test. A driver

section was used in order to initiate the detonation in the non-uniform mixture. The main goal was to

experimentally find the critical layer thickness and maximum hydrogen concentration for detonation

propagation in partially confined layer of stratified hydrogen-air mixture.

This work is to be published at the 34th Int. Combustion Symposium, Warsaw, Poland, 2012

Figure 4.2.39-1 Test Section for LACOMECO HYDET06 Test


J. Grune, M. Kuznetsov, R. Porowski, W. Rudy,, K. Sempert, A. Teodorczyk, Critical conditions of

hydrogen-air detonation in partially confined geometry”, to be published at the 34th Int. Combustion

Symposium, Warsaw, Poland, 2012






Temperature = 20°C

Layer geometry: horizontal 9x3x0.3 m open at the bottom

Hydrogen concentration profile: CH2 = -1.1·h[cm] + 25.9[%H2]

Test layer 30 cm

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4.2.40 E2-40 - LACOMECO HYDET07




Detonation propagation experiments in a stratified semi-confined layer of hydrogen-air mixture were

performed. A horizontal rectangular box of 9x3x0.6 m containing the hydrogen-air mixture was installed

inside of safety volume A1 (V=100 m3). A non-uniform hydrogen-air mixture, with a vertical

concentration gradient 1.1%H2/cm, and 25% hydrogen in air at the top boundary, was used in the test. A

driver section was used in order to initiate the detonation in the non-uniform mixture. The main goal was

to experimentally find the critical layer thickness and maximum hydrogen concentration for detonation

propagation in partially confined layer of stratified hydrogen-air mixture. The same test facility used as for

the test E2-39 - LACOMECO HYDET06.


J. Grune, M. Kuznetsov, R. Porowski, W. Rudy, K. Sempert, A. Teodorczyk, Critical conditions of

hydrogen-air detonation in partially confined geometry”, to be published at the 34th Int. Combustion

Symposium, Warsaw, Poland, 2012






Temperature = 20°C

Layer geometry: horizontal 9x3x0.3 m open at the bottom

Hydrogen concentration profile: CH2 = -1.1·h[cm] + 24.75[%H2]





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4.2.41 E2-41 - H2PAR E 12

Test Facility: H2PAR

Owner Organization: IRSN and EdF


The internal volume of H2PAR facility is equal to 7.6 m3 with a diameter of 2 m. The inner

containment confines the gaseous mixture and the aerosols, whereas the outer containment ensures the

thermal insulation of the system. A heated 50 litres water volume inside the inner containment allows

control of the atmospheric steam content.


P. Rongier et al., M., Rapport d’expérience – H2PAR essais E12 et E12BIS, internal report SERE 98/017


Gas and catalytic sheets temperature, hydrogen concentration and pressure are measured. For this

experiment, SIEMENS recombiner FR90/1-150 had been used. The initial conditions are:

Temperature = 85°C

Pressure = 1 bar

Steam concentration = 0%

Injection of hydrogen in dry air during 100 s within flow rate of 0.48 g/s




W. Plumecocq, V. D. Layly, A. Bentaib, “Modelling of the containment mitigation measures in the

ASTEC code, focusing on Spray and Hydrogen”, Nureth 11, Avignon, October 2-6, 2005.

Reinecke, E.A. Bentaib, A., Kelm, S., Jahn, W., Meynet, N., Caroli, C., “Open issues in the applicability of

recombiner experiments and modelling to reactor simulations”, Progress in Nuclear Energy, Volume 52,

Issue 1, pp. 136-147, 2010 January


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4.2.42 E2-42 - H2PAR E 13


Owner Organization: IRSN and EdF







P. Rongier et al., M., Rapport d’expérience – H2PAR essais E13 et E19, internal report SERE 98/021




Temperature = 85°C

Pressure = 1 bar

Steam concentration = 57.7%

Injection of hydrogen in dry air during 100 s within flow rate of 0.48 g/s




W. Plumecocq, V. D. Layly, A. Bentaib, “Modelling of the containment mitigation measures in the

ASTEC code, focusing on Spray and Hydrogen”, Nureth 11, Avignon, October 2-6, 2005

Reinecke, E.-A. Bentaib, A. , Kelm, S., Jahn, W. , Meynet, N , Caroli, C., “Open issues in the applicability

of recombiner experiments and modelling to reactor simulations, Progress in Nuclear Energy, Volume 52,

Issue 1, January 2010, pp. 136-147


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4.2.43 E2-43 - H2PAR E 3


Owner Organization: IRSN and EDF







S. Grandgeorge-Poulain et. al., Rapport d’expérience – H2PAR essais E3, E3S et E3S6, internal report

SERE 97/034




Temperature = 85°C,

Pressure = 1 bar

Saturated atmosphere with aerosols (steam concentration 58 vol.%)

Hydrogen concentration 10 vol.% in dry air





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4.2.44 E2-44 – KIT DDT Tests in CHANNEL Facility

Test Facility: CHANNEL



Flame acceleration and detonation transition experiments in a smooth channel filled with

stoichiometric hydrogen-oxygen mixture were performed. A horizontal rectangular channel of (3000-

6000) x 50 x 50 mm with transparent quartz window for optical observations was used. The deflagration-

to-detonation mechanism as well as the effect of boundary layer and different mixture reactivity (by

changing the initial pressure) on run-up-distance for detonation onset were experimentally investigated by

using a high speed camera combined with a Schlieren system.

1

3 4 44 4 4

2

Figure 4.2.44-1 Schematic of CHANNEL Test Facility


Kuznetsov M., Alekseev V., Matsukov I., Dorofeev S. DDT in a Smooth Tube filled with Hydrogen-

Oxygen Mixtures. Shock Waves, Vol. 14, No. 3, pp. 205 - 215 (2005)

Kuznetsov, M., Liberman, M., Matsukov, I. Experimental Study of the Preheat Zone Formation and

Deflagration to Detonation Transition. Combustion Science and Technology, 182, 11, pp. 1628- 1644

(2010)

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Initial Pressure = 0.1-1 bar

Temperature = 20°C

Channel geometry: horizontal (3000-6000)x50x50 mm closed

Stoichiometric hydrogen – oxygen mixture





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4.2.45 E2-45 – KIT Jet Ignition Tests in HPHR Facility

Test Facility: HPHR



Spontaneous ignition processes due to the high-pressure hydrogen releases into the air are

investigated. High-pressure hydrogen releases in the range of initial pressures from 20 to 275 bar and with

nozzle diameters of 0.5 – 4 mm have been investigated. Glass tubes of different length were used for

experimental study of self-ignition process using a high-speed CCD camera. The minimum initial pressure

of 25 bar leading to the self-ignition of hydrogen with air was measured in the tests. A probability of the

ignition of hydrogen-air cloud depending on the bulk pressure of high pressure tank, nozzle diameter and

release tube length was also investigated in the tests.

Figure 4.2.45-1 HPHR Test Section


Grune, J., Sempert, K., Kuznetsov, M., Jordan, T., Experimental Study of Ignited Unsteady Hydrogen

Releases From a High Pressure Reservoir. Proc. of the 4th Int. Conf. on Hydrogen Safety (ICHS2011),

2011 September 12-14, San Francisco, USA, Paper #133, pp. 1-11.

Grune J., Kuznetsov M., Lelyakin A., Jordan T. Spontaneous Ignition Processes due to High-Pressure

Hydrogen Release in Air. Proc. of the 4th International Conference on Hydrogen Safety (ICHS2011), 2011

September 12-14, San Francisco, USA, Paper #132, pp. 1-11.

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Initial Pressure = 25 - 250 bar

Temperature = 20°C

High pressure volume = 370 cm3

Hydrogen released in air or into hydrogen-air mixture





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4.2.46 E2-46 – KIT Geometric Quenching of Detonation Tests in the HYKA-A1 Facility




An experimental investigation of the deflagration and deflagration-to-detonation transition (DDT) in

an obstructed semi-confined flat layer filled with uniform hydrogen-air mixtures was performed. Effects

of mixture reactivity and flat layer thickness on the flame propagation regimes in order to evaluate critical

conditions for sonic flame propagation and detonation onset were investigated. The experiments were

performed in a large rectangular box with dimensions 9 x 3 x 0.6 m opened from below and filled with

obstacles with a blockage ratio of BR=0.5. The hydrogen concentration in its mixtures with air was varied

in the range of 13-28 %vol. The detonation onset in a semi-confined mixture layer occurred if the layer

thickness, h, was 13-14 times the detonation cell width. It also was found that the detonation cannot

propagate or fails just beyond an obstacle if the orifice size, b, in obstructions is less than 3 times the

detonation cell width. In those cases with the critical orifice size, we could visualise a zone of failed

detonation on sooted plates installed just after the obstacle then a detonation re-initiation zone with refined

cellular structure, followed by re-establishment to normal detonation cell size.

Figure 4.2.46-1 Schematic of HYKA-A1 Facility

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M. Kuznetsov, J. Grune, A. Friedrich, K. Sempert, W. Breitung and T. Jordan, Hydrogen-Air Deflagrations

and Detonations in a Semi-Confined Flat Layer, In: Fire and Explosion Hazards, Proceedings of the 6th Int.

Seminar (Edited by D. Bradley, G. Makhviladze and V. Molkov), 2011, pp. 125-136, ISBN: 978-981-08-

7724-8, doi:10.3850/978-981-08-7724-8_02-05.

J. Grune, K. Sempert, H. Haberstroh, M. Kuznetsov, T. Jordan, Experimental Investigation of Hydrogen-

Air Deflagrations and Detonations in Semi-Confined Flat Layers, Journal of Loss Prevention in the

Process Industries, Available online 8 October 2011, ISSN 0950-4230, 10.1016/j.jlp.2011.09.008.



Temperature = 20°C

Total volume = 100 m3

Hydrogen-air cloud volume = 9 x 3 x (0.15-0.6) m

Hydrogen-air mixture = 15 – 28 %H2





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4.2.47 E2-47 – Cheikhravat Experiments on Effect of Spray on Hydrogen Combustion


Owner Organization: IRSN and CNRS/ICARE


ENACCEF is a vertical facility of 5 m high and can be equipped with repeated obstacles in the bottom

part. It is divided in 2 parts:

The acceleration tube (3.2 m long and 154 mm i.d.), is equipped at its bottom-end with 2 tungsten

electrodes as a low energy ignition device. At a distance of 1.9 m from the ignition point, 3

rectangular quartz windows (40 mmx300 mm optical path) are mounted flush with the inner

surface, 2 of them are opposed to each other the third one being perpendicular to the others. These

windows allow the recording of the flame front during its propagation along the tube using either a

shadowgraph or a tomography system. The tube is also equipped with 11 small quartz windows

(optical diameter: 8 mm, thickness: 3 mm) distributed along it,

The dome (1.7 m long, 738 i.d.) is connected to the upper part of the acceleration tube via a flange.

This part of the facility is also equipped with 3 silica windows (optical path: 170 mm, thickness:

40 mm), perpendicular to each other 2 by 2 (see figure1, left). Through these windows, the arrival

of the flame can be recorded via a Schlieren or a tomography system.

Before each test, the chamber was vacuumed and the residual pressure was lower than 3 Pa. The gases

were introduced using the partial pressure method. The synthetic air consisted of 21 vol. % O2 + 79 vol. %

N2. The initial conditions are:

Temperature: between 20C and 25C,

Pressure: 1 bar

Hydrogen concentration: from 10.5% to 13.06%


Spray: hollow cone with rate of 8.6 l/min.

Water temperature: between 13C and 17C

After ignition (ignition point is located in the lower part of ENACCEF), spray is activated

simultaneously or within a time delay (sensitive parameter).

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Cheikhravat H., “Etude expérimentale de la combustion de l’hydrogène dans une atmosphère inflammable

en présence de gouttes d’eau”, PhD thesis Orléans University, 2010.

Cheikhravat H. et al., “Evaluation of the Water Spray Impact on Premixed Hydrogen-Air-Steam Flames

Propagation”, Proceeding American Nuclear Society conference, San Diego, 2010.


Gas Temperature between 20C and 25C,

hydrogen concentration from 10.5% to 13.06%

Water temperature between 13C and 17C





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4.2.48 E2-48 – Bjerketvedt Experiments on Effect of Spray on Hydrogen Combustion

Test Facility: Not available

Owner Organization: Not available

Experiment Description: Not available


Bjerketvedt D. ; Bjørkhaug M., “Experimental investigation: Effect of waterspray on gas explosions”,

Report prepared by the Christian Michelsen Institute, Bergen, Norway, for the UK Department of Energy,

OTH 90 316, HMSO, 1991.

Range of Key Experimental Parameters: Not available

Year Tests Performed: Not available

Repeatability Check: Not available

Past Code Validation/Benchmarks: Not available

Prepared By: A. Bentaib (IRSN) provided reference

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4.3 Aerosol and Fission Product Behaviour Experiments

4.3.1 E3-1 - AHMED OECD benchmark

Test Facility: AHMED

Owner Organization: VTT (Finland)


The AHMED (Aerosol and Heat Transfer Measurement Device) was constructed by VTT to study the

effect of thermalhydraulics on hygroscopic and inert aerosol behaviour in containment during a

hypothetical severe accident. Hygroscopic NaOH, CsOH and CsI and inert Ag aerosol behaviour at

different temperatures and relative humidities (RH) was studied in the instrumented and controlled

AHMED vessel, where homogeneous thermal-hydraulic conditions for aerosol measurement were

achieved. The vessel and input line pressures and steam and air flows were also continuously monitored.

Input air flow was filtered and dried. The vessel surface temperature was controlled using computer

controlled heating cables. The input gas temperature was regulated using a heat exchanger. The aerosol

number and mass concentration were measured continuously during the experiments using a condensation

nucleus counter and a tapered element oscillating microbalance. The particle size distribution and

chemical composition in the test conditions were measured by Berner low pressure impactors.

At first the behaviour of the aerosols of different chemical species was studied separately in the

AHMED facility. In a second step, both Ag and CsOH aerosol particles were generated and injected in the

vessel simultaneously to study the behaviour of multi-component aerosol. The temperature in the

experiments was varied between 17°C and 51°C, and the RH between 7.3% and 97%. The experiments

have shown that in the case of NaOH, the ratio of aerosol mass concentration half lives at low to high RH

experiments was about 4 while for CsI and CsOH this ratio was about 2. This difference is due to the

density effect: CsOH and CsI have higher density than NaOH and thus during condensation their

aerodynamic sizes do not increase as much as the AMMD (equilibrium diameter of the particles).

The AHMED system has a 1.81 m3 total free volume, 63.5 cm vessel radius, 142.5 cm vessel height,

12,700 cm2 sedimentation area, operated at atmospheric pressure, and had a 2.6 L/min sampling rate.


Jokiniemi, J. The effect of airborne hygroscopic matter on aerosol behaviour in severe nuclear power plant

accident, technical Research Centre of Finland, publications 59 (Dissertation)

AHMED Code Comparison Exercise – Comparison report, NEA/CSNI/R(95)23

J.M. Mäkynen et al., AHMED experiments on hygroscopic and inert aerosol behaviour in LWR

containment conditions: experimental results, Nuclear Engineering and Design, 178 (1997) 45-59.

Some information may also be found in NEA/CSNI/R(2009)5

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Temperature from 17 to 51°C.

RH from 7.3 to 97%

AMMD 2.1 to 2.7 μm

Mass concentration: 60 to 638 mg/cm3

Number concentration: 48,000 to 137,000 cm-3

Year Tests Performed: ~1995

Repeatability Check: Yes (several tests have been performed with the same aerosol in different

conditions)


A series of AHMED experiments was used for computer code benchmark at the NEA. Benchmark

included MELCOR 1.8.3, MELCOR 1.8.2, IDRA 4.1, CONTAIN 1.12 FIPLOC-MI 2.0, MACRES and

NAUAHYGROS 1.1 and involved 7 organisations from 5 countries.


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4.3.2 E3-2 - KAEVER CsI series

Test Facility: KAEVER



The KAEVER experiments investigated the aerosol depletion in a 10 m³ steel vessel under

homogeneously mixed atmospheric conditions. In the CsI-series the following conditions were

established:

Test K100A had a relative humidity (RH) of 0%,

Test K102A had a RH of 0% ,

Test K106A had a RH of 85 – 90%,

Test K108A had a RH of 95%,

Test K110A had a RH of 95%,

Test K123A had a RH of 100% and weak fog formation,

Test K159A had strong fog formation.


Poss, Weber: “Versuche zum Verhalten von Kernschmelzaerosolen im LWR-Containment -KAEVER

Abschlussbericht Teil I” Battelle-Ingenieurtechnik GmbH, BF-R-67863, Mai 1997

Poss, Weber: “Versuche zum Verhalten von Kernschmelzaerosolen im LWR-Containment,

Datensammlung” Battelle-Ingenieurtechnik GmbH, BF-R-67863, Mai 1997


Maximum concentration of aerosol material ~4 g/m³

Year Tests Performed: 1993- 1995



Firnhaber et al., “ISP-44-KAEVER Experiments of the Behaviour of Core-melt Aerosols in a LWR

Containment” NEA/CSNI/R(2003)5 August 2002


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4.3.3 E3-3 - KAEVER K187 (ISP-44)





homogeneously mixed atmospheric conditions. In test K187 the aerosol mixture of Ag, CsOH and CsI was

injected. The relative humidity was 100% and there was a slight fog concentration.


Firnhaber et al., “Draft Specification of the ISP No. 44; KAEVER Experiments of the Behaviour of Core-

melt Aerosols in a LWR Containment”, GRS July 1999




Maximum concentration of aerosol material < 1g/m³

Maximum concentration of aerosol material plus condensed water (wet concentration) = 10 g/m³)







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4.3.4 E3-4 - KAEVER K148 (ISP-44)





homogeneously mixed atmospheric conditions. In test K148 Ag aerosol was injected. The relative

humidity was 100% and there was a slight fog concentration.







Maximum concentration of aerosol material = 1 g/m³

Maximum concentration of aerosol material plus condensed water (wet concentration) ~10 g/m³







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4.3.5 E3-5 - KAEVER K188 (ISP-44)





homogeneously mixed atmospheric conditions. In test K188 CsOH aerosol was injected. The relative

humidity was 100% and there was a slight fog concentration.







Maximum concentration of aerosol material ~0.2 g/m³

Maximum concentration of aerosol material plus condensed water (wet concentration ~20 g/m³







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4.3.6 E3-6 - LACE LA2

Test Facility: LACE



Objective of this test was to determine retention and behaviour of aerosols in a containment system

with two pre-existing leak paths. This is used to represent a failure to isolate the containment building.

Results will be used to benchmark computer codes used to predict T/H and aerosol transport behaviour in

containment. The test used a modified CSTF from experiments AB5, AB6, AB7 tests.

The test was performed in four consecutive thermalhydraulics stages:

1. A rapid heat up phase in which steam was injected upward along the vessel centerline from a

point in the lower part of the CSTF. Heated nitrogen was added through the aerosols delivery line

at a low rate.

2. Aerosol release phase where aerosols, steam, and non condensable gases were added to the

containment system through the aerosol delivery line. Steam was also added at a reduced rate

through the vertical steam pipe in the lower part of the CSFT.

3. A slow cooldown phase in which steam and nitrogen ware added to the containment at a reduced

rate through the vertical steam pipe and aerosol delivery pipe respectively,

4. A cooldown period in which nitrogen addition continued at a low rate through the aerosol delivery

line but steam injection was discontinued.


Souto, F.J., Haskin, F. E and Kmetyk, L. N.,”MELCOR 1.8.2 Assessment: Aerosol Experiments ABCOVE

AB5, AB6, AB7 and LACE LA2”, SAND94-2166, Oct. 1994.

Hilliard R.K., Muhlestein L. D., Albiol T. J., “Final report of experimental results of LACE test LA2 –

Failure to isolate containment”, LACE TR-007, June 1987

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Experiment Parameters (LA2 Test):

Rep. nonradioactive aerosol:

o CsOH, rep. water soluble species

o MnO, rep. insoluble species

Suspended mass concentration (max; p8):

o CsOH, 1.69 g/m3

o MnO, 2.06 g/m3

Note: See Table 7 (Pg. 15 of SAND94-2166) for more details




Souto, F.J., Haskin, F. E and Kmetyk, L. N., “MELCOR 1.8.2 Assessment: Aerosol Experiments

ABCOVE AB5, AB6, AB7 and LACE LA2,” SAND94-2166, Oct. 1994.

LACE (TR-004 & TR-009) ORNL post test report (Owned by EPRI)

LACE (TR-010) Intermountain Technologies, Inc., Idaho Falls, ID (Owned by EPRI)


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4.3.7 E3-7 - LACE LA4

Test Facility: LACE



This is a large vessel test with mixed aerosol (CsOH and MnO). Aerosol injected into humid vessel

with measurement of airborne concentrations and depletion rate. Testing was performed in multi

thermalhydraulic and aerosols injection stages into the CSTF vessel, and the behaviour of aerosols were

measured.


McCormack & Hilliard, R.K. (WHC) and Salgado, J.M. (TECHATOM), “Final Report of Experimental

Results of LACE Test LA4 – Late containment Failure with Overlapping Aerosol Injection Periods,”

LACE-TR-025, 1987


CSTF steel containment vessel: 852 m3

pre-existing vent path: sharp edge orifice plate of 34mm ID in a 6m length of duct

Seven consecutive T/H periods (steam, CsOH, CsOH+MnO, MnO, low steam, vent, cool down)

Average mass flow rate of CsOH=0.929g/s; MnO=0.757g/s


Repeatability Check:


Wilson, J.H., and Arwood, P.C., “Comparison of (Posttest) Predictions of Aerosol Codes with

Measurements in LWR Aerosol Containment Experiment (LACE) LA4”, Oak Ridge National Laboratory,

LACE TR-084, ORNL/M-991, 1990.


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4.3.8 E3-8 – LACE LA5 and LA6

Test Facility: CSTF (Containment Systems Test Facility)



The LA5 and LA6 experiments were devoted on aerosol behaviour during rapid containment

depressurization transient and on aerosol re-entrainment from a flashing pool.

The experiments were performed in the CSTF (Containment Systems Test Facility) of the Handford

Engineering Development Laboratory (HEDL). This facility mainly consists of the containment steel

vessel (a 852 m3 vessel) in which bottom an internal pool was filled with up to 3 m of water. In order to

maximize pool flashing, the water was heated to its boiling point before depressurization. To achieve a

rapid depressurization to generate a sufficient re-entrainment of aerosols from the flashing pool, a

discharge pipe (444-mm diameter) was installed at the middle level of the vessel and a full-port butterfly

valve was opened by a fast-acting pneumatic actuator.

The LA5 and LA6 experiments consisted of the following periods:

Rapid vessel heating by injecting hot steam at high flow rate (70 min for LA5 and 150 min for

LA6).

Slow heating by injecting intermediate steam and nitrogen flows. (50 min).

Pressurization al low steam and nitrogen flows up to a final pressure of 4.4 bar at 124ºC (370 min

for LA5 and 400 min for LA6).

Rapid depressurization to atmospheric pressure with pool flashing (1 min.).

Cooling of the vented vessel (1600 min).

CsOH and MnO aerosols were added in the LA6 experiment during the slow heating phase whereas in

LA5 no aerosols were injected. Both aerosol materials were generated by vaporization/condensation

processes and mixed in the Aerosol Mixing Vessel (AMV) to provide some agglomeration before being

carried to the containment vessel.

The amount of liquid entrained from the flashing was successfully measured using Li2SO4 previously

dissolved into the water pool. Significant differences in Li2SO4 concentration were found depending on

the location: 5 to 20 mg/m3 (1.2 m above the top of the internal tank) and 1 to 2 mg/m

3 at locations away

from the pool. Size distribution was around 4 μm. CsOH and MnO concentration were reduced by a

factor from 4 to 7 at the time of the depressurization. This reduction can be explained by fog formation,

condensation on particles and by the subsequent rain-out of large drops.


D.R. Dickinson, D.C. Mecham and D.C. Slaughterbeck, 1988. “Final Report of Experimental Results of

LACE Tests LA5 and LA6 – Rapid Containment Depressurization” U.S. Department of Energy, LACE

TR-026, September 1988

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Pressure: 4.5·105 Pa to 1.0·10

5 Pa.

Aerosol size:

o inlet: AMMD = ~2.5 μm, GSD = ~1.8

o vessel: AMMD = 1.7 – 6.0 μm, GSD = 1.4 – 2.1

o Li2SO4: AMMD: ~4 μm

Aerosol concentration (before the depressurization):

o CsOH: 2·10-3

g/m3

o MnO: 7·10-3

g/m3

o Li2SO4: 2·10-2

g/m3


Repeatability Check: Yes. Two similar tests performed.

Past Code Validation/Benchmarks: Unknown

Prepared By: J. Fontanet (CIEMAT) and L.E. Herranz (CIEMAT)

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4.3.9 E3-9 - Phebus FPT-1 (ISP-46)

Test Facility: Phébus



Phébus FPT-1 is an integral experiment that involved the degradation of a bundle made of twenty 1-m

long irradiated fuel rods including an Ag-In-Cd control rod, the release of fission product and structural

material in a model reactor cooling system , their behaviour in a model containment and iodine chemistry

in the containment. FPT61 was used for the International Standard Problem 46 (ISP-46). ISP-46 was

divided into 4 phases:

1. fuel degradation and material release from the bundle,

2. fission product and structural material transport in the reactor cooling system,

3. thermal-hydraulics and aerosol behaviour in the containment, and,

4. iodine chemistry in the containment.

Phases 3 and 4 are relevant to the CCVM.


Haste, T, “Specification of International Standard Problem ISP-46 (Phebus FPT1) Revision2”, IRSN Note

technique SEMAR 03/05

Jacquemain, D. Bourdon, S. De Breamecker, A. Barrachin, M., “PHEBUS FPT1 Final Report”, IPSN

Report SEA 1/00, December 2000

Clément, B., Haste, T, “ISP-46 – PHEBUS FPT1, Integral Expeiment on Reactor Severe Accident –

Comparison Report”, Report NEA/CSNI/R(2004)18, August 2004


The test was performed in a 10 m3 vessel, including gas phase, water phase and cooled surfaces, in

which steam, hydrogen and a mixture of fission product and structural material was injected during few

hours. The relative humidity was moderate and the pressure was about 2 bars. After having stopped the

injection, the containment was isolated and aerosol depletion measured. Finally, iodine chemistry was

studied during several days with a radiation field coming from injected radioactive material.

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Repeatability Check: No repeats, but other tests from the Phebus FP series (mainly FPT-2 and FPT-3)

can be used for comparison


Clément, B., Haste, T, “ISP-46 – PHEBUS FPT1, Integral Expeiment on Reactor Severe Accident –

Comparison Report”, Report NEA/CSNI/R(2004)18, August 2004

Prepared By: B. Clément (IRSN) and A. Bentaib (IRNS)

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4.3.10 E3-10 - POSEIDON PA10

Test Facility: POSEIDON



The experiments are meant to provide a thorough investigation of the pool scrubbing phenomena.

The experiment aimed at determining the dependence of hot pool Decontamination Factor (DF) on water

height at different carrier gas steam mass fractions with a relatively constant inlet aerosol diameter. The

DF increases exponentially with height. It also increases with carries gas steam mass fraction as

condensation at the injection point is an efficient aerosol removal mechanism.


A. Dehbi, D. Suckow, S. Guentay, The Effect of Liquid Temperature on Pool Scrubbing of Aerosols, J.

Aerosol Sci. Vol. 28, 1997


Injection Flow rate: 125 kg/hours

Gas temperature: 250°C

Pool temperature: 85°C

Pool height: 4 m

Test duration: ~1 hour





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4.3.11 E3-11 - BMC VANAM M2

Test Facility: BMC




BMC was built from reinforced concrete, had a free volume of 640 m³, and was designed to be a 1/64

representation of the Biblis B containment. The test-procedure of M2 was similar as the procedure of test

M3 (ISP-37), but insoluble instead of soluble aerosol was used. In Phase 1 of the experiment (17 h) the

containment was heated up by steam injection into the upper internal compartment R5. Then the insoluble

SnO2 aerosol was injected at the same position using air as carrier gas. After a phase without any injection,

a second aerosol injection was performed. The steam injection position was changed temporarily to the

lower room R3 and then switched back to R5.


Kanzleiter: “VANAM Multi-compartment Aerosol Depletion Test M2* with Insoluble Aerosol Material”,

Battelle-Institut e.V. Frankfurt, Technical Report BleV-R67.098-303, July 1993


Aerosol concentration up to 10 g/m³

Relative humidity 80 to 100%

Temporarily high fog concentrations

Pressure 1 to 2 bar


Atmospheric velocity 0 to 0.75 m/s





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4.3.12 E3-12 - VICTORIA test 58

Test Facility: VICTORIA

Owner Organization: VTT (Finland), EC


The VICTORIA facility is a scale model of the ice condenser containment of Loviisa NPP with linear

scale of 1:15 and volume scale of 1:3375. The facility experiments were aimed at validating the

containment aerosol models used in the computer codes, in particular models related to radioactive

hygroscopic and non-hygroscopic aerosol behaviour in non-homogeneous multi-compartment

containments. Two research programmes have been carried out. In the earlier research programme (1990-

1995) focus was put on the T-H behaviour and hydrogen distribution in severe accident conditions. During

1996-1997 period, the VICTORIA facility was used in a modified geometry for aerosol experiments in a

CEC 4th Framework programme project.

The aerosols were generated by two aerosol generators. Water-soluble aerosol, either NaOH or

CsOH, was generated with two opposite jet atomisers. The dry particle AMMD was ~2.3 μm and the feed

rate was approximately 3 mg/s. A high temperature entrainment flow reactor was used for the generation

of silver particles. The piping from the aerosol generators (diameter 20 mm) was connected into one

injection line which was directed into the upper or lower compartment of the containment. The injection

pipe was made out of acid-resistant stainless steel. The temperature of the injection line was kept at

slightly over 100°C and it was adjusted by a separate controller.

In both 58 and 59 experiments, aerosol was injected into the lower compartment of the facility, while

the main difference was that in experiment 58, the aerosol material was CsOH, whereas in experiment 59,

CsOH was mixed with silver. In experiments 61 and 62 the particles were injected in the upper part of the

facility. In experiment 61, CsOH was used as aerosol material, while in experiment 62, the material was

silver.

The following were the measured parameters:

Aerosol number and concentration (condensation nucleus counter and tapered element oscillating

microbalance)

Particle mass and chemical composition distributions (11-stage multi-jet type low pressure

impactors)

Dry number size distribution (Electrical low pressure impactor (ELPI) and a differential mobility

analyser (DMA))

Temperature

Humidity

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Estimated data accuracy is as follows:

Measurement size range of the impactors: 0;03 – 15 μm

Size range of the ELPI: 0.01 – 5 μm

Size range of the DMA: 0.02 – 0.8 μm

Temperature range: -40 to +180°C

Relative humidity in range of 0 to 100%


E. Heikkilä, Technical description of VICTORIA Facility, E.C. Report ST-APC(96)-P07 (1996)

J.M. Mäkynen, J.K. Jokiniemi, E.I. Kauppinen, H. Tuomisto, T. Routamo, LWR Containment Aerosol

Experiments at Victoria Facility – Data Report 1/96 – E.C. Report ST-APC(96) –P08 (1996)

J.M. Mäkynen, J.K. Jokiniemi, E.I. Kauppinen, H. Tuomisto, T. Routamo, LWR Containment Aerosol

Experiments at Victoria Facility – Final Report – E.C. Report ST-APC(98) –P19 (1998)

J.M. Mäkynen, J.K. Jokiniemi, E.I. Kauppinen, A. Slide, S. Outa, T. Routamo, H. Tuomisto, Experimental

and Modelling Studies on Containment Aerosol Behaviour in the Victoria Facility (1998)

J.M. Mäkynen, T. Routamo, LWR Containment Aerosol Experiments with the Victoria Facility – Data

Report Experiment 62 (1999).


Dry particle AMMD ~2.3 μm

Particle feed rate: 3 mg/s

Pressure: 1 bar

Maximum capacity of the steam generator: 25g/s

Year Tests Performed: 1996 – 1997



Results of experiment 61 were compared with calculations using FIPLOC and CONTAIN computer codes.


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4.3.13 E3-13 - CSTF ABCOVE Tests

Test Facility: CSTF



The Containment System Test Facility (CSTF) vessel, located at the Hanford Engineering

Development Laboratories, USA, was used to perform the ABCOVE tests. The tests covered a series of

dry aerosol experiments under hypothetical severe accident conditions in a Liquid Metal Fast Breeder

Reactor (LMFBR). Tests AB5, AB6 and AB7 studied behaviour of sodium or sodium iodide aerosols

before, during and after a sodium fire.

The objectives of these tests were to:

provide experimental data on behaviour of aerosols generated by a sodium spray fire, and

demonstrate co-agglomeration behaviour of two aerosol species.


Souto, F.J., Haskin, F. E and Kmetyk, L. N.,”MELCOR 1.8.2 Assessment: Aerosol Experiments ABCOVE

AB5, AB6, AB7 and LACE LA2,” SAND94-2166, Oct. 1994.


Experiment Parameters (AB5 Test):

CSTF containment vessel: 852m3

Interior surfaces: modified phenolic paint

Exterior surface: 25.4mm fibre glass insulation

Spray:

o nozzle loc.: 5.15m elevation

o 223 kg of sodium over a period of 872 s, with

o all the sodium converted to a 60% Na2O2 and 40% NaOH aerosol

Containment seal time: 5.136x105 s

Containment pressure (max): 214 kPa

Mean atmospheric temp:553.15K (local temp max at 843.15K)

Suspended mass conc.

o Max, 170 g/m3 @ 383 s

o SS, 110±17 g/m3

Note: More details available in Tables 1 thru 7 (p9-16 of SAND94-2166) for AB5, AB6, AB7 and LA2

tests

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Souto, F.J., Haskin, F. E and Kmetyk, L. N., “MELCOR 1.8.2 Assessment: Aerosol Experiments

ABCOVE AB5, AB6, AB7 and LACE LA2,” SAND94-2166, Oct. 1994.


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4.3.14 E3-14 - CSTF ACE

Test Facility: CSTF



The Advanced Containment Experiment (ACE) facility in the Containment System Test Facility was

used to investigate iodine behaviour in the presence of aerosols. The aerosols consisted of iodine (non-

radioactive), manganese and cesium.

References for Experiment Description:

Wall, I. And Merilo, M., 1992, “Advanced Containment Experiments (ACE) Project: Summary Report,”

EPRI TR-100346s


Pressure: 242 and 250 kPa

Temperature: ~100°C

Steam fraction: ~43%

Steam injection: 28 g/s

Aerosol sizes:

Cs: 2.77 and 3.42 µm

Mn: 3.93 and 3.39 µm



Past Code Validation/Benchmarks: MAAP code


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4.3.15 E3-15 - CARAIDAS Aerosol washout by single droplet tests

Test Facility: CARAIDAS

Owner Organization: IRSN (with EDF partial funding)


The IRSN CARAIDAS experimental set-up was used to study drop evolution under representative

conditions of post-accident atmosphere. The cylindrical enclosure is of 5 m high and 0.6 m inner diameter.

Homogeneous conditions are obtained with gas temperatures from 20 to 160°C, absolute pressures from 1

to 8 bar and relative humidities from a 3 up to 95%. The drop generator is located at the top of the vessel

in order to keep it at a constant temperature, independent of the vessel temperature. It can produce

monodisperse water droplets from 200 to 700 µm in diameter. Drop injection temperature is set between

20°C and 80°C by an electric heater. Initial droplet size, velocity, and temperature are determined

experimentally for each test. Drop diameter optical measurements are performed at 3 elevations. So-

called ‘evaporation’ and ‘condensation’ tests are performed.

Aerosol generation is based on mechanical spraying, by a rotating disk, of caesium iodide solution

tagged by soda fluorescein. With this specific generator, aerosols can be produced at specified

temperatures (20 to 160°C) and pressure levels (1 to 7 bar) in the vessel. The aerosol diameter ranges

between 0.5 and 5 µm, with a geometric standard deviation lower than 2. The aerosol mass concentration

is 0.1 g/L. Homogeneity of concentration and particles size distribution is checked. Aerosols

concentrations are measured by sampling on 25 mm diameter fibreglass filters during one or two minutes

with a 1 l/min flow rate. The particle size distribution measurements are given by inertial impactors at

vessel pressure and temperature conditions. The CARAIDAS facility allows to measure and the mass of

aerosols collected by falling drops as a function of different experimental conditions representative of

severe accident scenarios.


D. Ducret, Y. Billarand, D. Roblot, J. Vendel Study on collection efficiency of fission products by spray:

experimental device and modelling 24th DOE/NRC Nuclear Air Cleaning and Treatment Conference,

Portland, USA, 15-18 July 1996, NUREG/CP-0153.2, 1996

D. Ducret et al., Etude expérimentale et modélisation du rabattement des aérosols par des systèmes

d’aspersion 14e Congrès Français sur les Aérosols, Dec. 1998, Paris, France, 1998

Hans-Josef Allelein, Ari Auvinen, Joanne Ball , Salih Güntay, Luis Enrique Herranz, Akihide Hidaka ,

Alain V. Jones, Martin Kissane, Dana Powers , Gunter Weber , “State-of-the-Art Report on Nuclear


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Gas temperature: 20 to 160°C

Pressure: 1 to 8 bar

Relative humidity: 3 to 95%


Droplet Injection T: 20 to 80°C

Aerosol size: 0.5 to 5 µm

Aerosol mass concentration: 0.1 g/l

Year Tests Performed: Early 2000's




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4.3.16 E3-16 - Whiteshell Flashing Jet Tests

Test Facility: Whiteshell Flashing Jet Facility



Flashing jets experiments were performed at AECL’s Whiteshell laboratories using high-temperature

(166 to 285°C), high-pressure (1 to 10 MPa) subcooled water discharged through various nozzles. The

nozzle assemblies consisted of simple round-hole nozzles, conical nozzles and nozzles fitted with

extension pipes. Nozzle diameters varied from 0.061 to 0.24 cm, and the L/D ratios varied from 0.5 to 200,

depending on the nozzle type. At any point in the flow field of the flashing jet, the aerosol size distribution

and velocity were simultaneously and non-intrusively measured with a Phase Doppler Anemometer.


Mulpuru, S.R., R. Balachandar and M. Hogeveen Ungurian, “Phase Doppler Anemometer -

Commissioning tests for Measurement of Water Aerosol Sizes and Velocities in Flashing Jets”, The 3rd

Int.

Conf. on Containment Design and Operation, Conference Proc., Vol. 2, 1994 October 19-21, Toronto,

1994.

S.R. Mulpuru, R. Balachandar and M.H. Ungurian, “Phase-Doppler Anemometer- Commissioning tests for

Measurement of water Aerosol Sizes and Velocities in Flashing Jets”, COG-93-395 , 1994

R. Balachandar, S.R. Mulpuru, and M.H. Ungurian, “Droplet Size and Velocity Measurements in Flashing

Water Jets Generated by using a Custom Made Nozzle (Diameter = 0.061 cm and Throat Length = 0.122

cm)”, COG-94-543 , 1995

R. Balachandar, S.R. Mulpuru and M.H. Ungurian, “Droplet Size and Velocity Measurements in Flashing

Water Jets Generated by using a Custom Made Nozzle Fitted with Pipe Extension (Diameter = 0.061 cm

and L/D = 200)”, COG-94-557 , 1995



cm)”, COG-95-008, 1995


Water Jets Generated by a WALE Nozzle (Diameter = 0.061 cm)”, COG-95-86, 1995



cm)”, COG-95-195, 1995

S.R. Mulpuru and M.H. Ungurian, “Droplet Size and Velocity Measurements in Flashing Water Jets

Generated by using a Custom Made Nozzle (Diameter = 0.061 cm and Throat Length = 0.61 cm)”, COG-

96-29, 1996


Generated by using a Custom Made Nozzle (Diameter = 0.24 cm and Throat Length = 12 cm)”, COG-96-

108, 1996

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Generated by using a Custom Made Nozzle (Diameter = 0.122 cm and Throat Length = 0.61 cm)”, COG-

96-164, 1996

S.R. Mulpuru, M.H. Ungurian and M.D. Pellow, “Droplet Size and Velocity Measurements in Flashing

Water Jets Generated by using a Custom Made Nozzle (Diameter = 0.24 cm and Throat Length = 2.4 cm)”,

COG-96-167, 1996


Water Jets Generated by a WALE Nozzle (Diameter = 0.122 cm)”, COG-96-166, 1997



cm)”, COG-96-247, 1997


Upstream Water Temperature: 166 to 285°C

Upstream Water Pressure: 1 to 10 MPa

Downstream Conditions: Normal room pressure and temperatures

Nozzle Diameter: 0.061 to 0.24 cm

L/D Ratio: 0.5 to 200





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4.3.17 E3-17 - Clarkson College Brownian Agglomeration

Test Facility: N/A

Owner Organization: Clarkson College


Aerosol particles were generated by condensing dibutyl phthalate (DBP) vapour onto NaCl nuclei

(condensed from NaCl vapours). The DBP aerosol particles were passed through a cylindrical tube used as

an agglomeration chamber. A light-scattering photometer was used to measure the size distribution of

aerosols at both entrance and exit of the agglomeration chamber.


Huang, C.M., M. Kerker, E. Matijevic and D.D. Cooke, “Aerosol Studies by Light Scattering, VII

Preparation and Particle Size Distribution of Linolenic Acid Aerosols”, J. Colloid and Interface Science,

33, 244, 1970.

Nicolaon, G., D.D. Cooke, M. Kerker and E. Matijevic, “A New Liquid Aerosol Generator”, J. Colloid and

Interface Science, 34, 534, 1970.

Nicolaon, G., D.D. Cooke, E.J. Davis, M. Kerker and E. Matijevic, “A New Liquid Aerosol Generator, II

The Effect of Reheating and Studies on the Condensation Zone”, J. Colloid and Interface Science, 35, 490,

1971.

Nicolaon, G., M. Kerker, D.D. Cooke and E. Matijevic, “Brownian Coagulation in a Submicron Aerosol:

Comparison of Experiment with Theory”, J. Colloid and Interface Science, 38, 460, 1972.


Initial aerosol size:

o mean = 0.237 µm

o std dev = 0.10

Agglomeration time: 41 and 110 s


Repeatability Check: Unknown



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4.3.18 E3-18 - JAERI Thermophoresis Tests

Test Facility: N/A

Owner Organization: JAERI


Experiments studied thermophoretic deposition of aerosols on the pipe walls of a heat exchanger in a

temperature gradient along its length. Aerosol was sodium oxide particles. The deposited sodium atoms

were measured by atomic adsorption spectrometry. The amount of aerosol deposition due to

thermophoresis was obtained by taking the difference between tests performed with and without a

temperature gradient.


Nishio, G., S. Kitani and K. Takahashi, “Thermophoretic Deposition of Aerosol Particles in a

Heat-Exchanger Pipe”, Ind. Eng. Chem., Process Des. Develop., 13, 408, 1974.


Inlet conditions:

o Gas flow: 1.45 to 41.5 L/min

o Aerosol concentration: 0.00685 to 1.39 mg/l

o Gas temperature: 75 to 21ºC

Outlet conditions:

o Gas temperature: 22 to 51ºC



Past Code Validation/Benchmarks: Not provided


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4.3.19 E3-19 - PITEAS Diffusiophoresis Tests (PDI 08, PDI 09, PDI 11 and PDI 12)

Test Facility: PITEAS



Examination of diffusiophoresis (of CsI aerosols) to the vessel walls where steam condensation is

occurring.


Albiol, T. and C. Lefol, “Piteas Programme, Diffusiophoresis of Aerosol Particles during Steam

Condensation PDI Experiments”, Note Technique SREAS/LEA 93/134, 1993.

V. Saldo, E. Verloo, A. Zoulalian: Study on aerosol deposition in the PITEAS vessel by settling,

thermophoresis and diffusiophoresis phenomena, J. Aerosol Science, vol 29, suppl.1, pp. S1173-S1174

(1998)


Aerosol size: 1 to 5

Vessel pressure: 3.4 to 4.5 bar

Vessel temperature: 121°C

Relative humidity: ~95%





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4.3.20 E3-20 - PITEAS Aerosol Condensation Tests (PCON 01 to PCON 05)

Test Facility: PITEAS



Starting from ambient conditions, the Piteas vessel containing 50 litres of water is heated up to 120C

and the containment pressure reaches ~3.3 bar. When the pressure is stabilized, the vessel is vented to

atmospheric pressure and closed. After several hours, when the pressure has reached a new equilibrium

value (~2.1 bar), the bottom vessel water is extracted. The CsI aerosol injection can start and is stopped for

a vessel pressure of ~4 bar. Some minutes later, the thermofluid temperature setpoint is decreased to

~100C in order to obtain condensation and diffusiophoresis.


Sabathier, F., “Piteas Programme, the Growth of Aerosol Particles by Steam Condensation PCON

Experiments”, Note Technique SREAS/LEA 92/093, 1992.


Vessel pressure: 4 bar

Vessel temperature: ~120°C

Relative humidity: 52, 90, 95 and 100%





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4.3.21 E3-21 - Aerosol Deposition in Turbulent Vertical Conduits (Sehmel)

Test Facility: N/A

Owner Organization: Pacific Northwest Laboratory


Tests to measure turbulent deposition of aerosols (uranine/uranine-methylene blue particles) in a

vertical aluminium tube (3.65 to 15.24 m in length and inside diameters of 0.53, 1.57, 2.93 and 7.14 cm).


Sehmel, A., “Aerosol Deposition from Turbulent Airstreams in Vertical Conduits”, Pacific Northwest

Laboratory Report BNWL-578, 1968.


Aerosol size: 1 to 28 µm

Reynold’s number: 100 to 60,000

Deposition velocity: below 10 cm/s





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4.3.22 E3-22 - Aerosol Deposition in Turbulent Vertical Conduits (Forney)

Test Facility: N/A

Owner Organization: Harvard University


The experiments deal with the aerosol deposition on the inner walls of vertical circular tubes under

fully developed turbulent flow. Deposition occurs in a removable test section consisting of a glass tube

(150 cm in length) having inside diameters ranging from 1.3 to 4.4 cm. Perfect sticking was achieved by

coating internal surfaces with a mixture of paraffin (75%) and petroleum jelly (25%). Fully developed

turbulent flow was assured for all test conditions using an entrance tube section that was large enough.

The aerosols that were used (Lycopodium spores, Ragweed and Pecan pollen and Polystyrene spheres)

were dispersed into the air flow by an aspirator. The deposition time was typically 4 min and afterwards

the test section was removed and the collected particles were examined in 30 different locations using an

optical microscope.


Forney, L.J. and Spielman L.A., “Deposition of coarse aerosols from turbulent flow”. Aerosol science,

Vol. 5. pp. 257-271 (1974)


Aerosol size: 19.5 – 48.5 μm

Reynolds number: 4×103 – 6×10

4

Air velocity: 7 – 30.2 m/s

Tube diameter: 1.3×10-2

– 4.4×10-2

m.

Normalized stopping distance: 22.5 -715


Repeatability Check: Detailed information about repeatability is not available, but it was reported that

“measurements reproducibility is about 25-50 per cent” (Forney and Spielman, 1974).



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4.3.23 E3-23 - Aerosol Deposition in Turbulent Vertical Conduits (Friedlander)

Test Facility: N/A

Owner Organization: University of Illinois


The experiment studies the deposition rate of dust particles on the walls tubes in a turbulent stream.

Three types of particles were used in the experiment: iron powder, aluminium powder and lycopodium

spores. The aerosols were dispersed using an atomizing nozzle and then the aerosol stream was mixed

with a secondary air stream to adjust the desirable flow. The observation tubes have a diameter ranging

from 0.54 to 2.5 cm and were made of either glass or brass. To avoid re-entrainment the test tubes were

coated with “pressure-sensitive” Scotch tape or glycerol jelly. In general the duration of the tests was

between 5 and 30 minutes and the deposition was measured by counting particles in a microscope.


Friedlander, S.K. and Johnstone, H.F., “Deposition of suspended particles from turbulent gas streams”.

Industrial and Engineering Chemistry, Vol. 49, No. 7, pp. 1151-1156 (1957)


Aerosol size: 3, 5 and 30 μm

Reynolds number: 7×103 – 5×10

4


– 2.5×10-2

m


Repeatability Check: Yes, “Experiment with the same conditions but using different tube and adhesive

materials gave similar results” (Friedlander and Johnstone, 1957)



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4.3.24 E3-24 - Aerosol Deposition in Turbulent Vertical Conduits (Liu)

Test Facility: N/A

Owner Organization: University of Minnesota


The experiment measured the deposition of aerosols inside a vertical tube with turbulent flow. A

vibrating-orifice monodisperse aerosol generator was used as source of the test aerosols. The aerosols

were transported by an air flow towards the tube test section: a 103 cm long glass pipe of 1.27 cm of

diameter. Downstream the test section was a fibreglass filter. The aerosol particles that were used were

uniform, spherical droplets of olive oil containing less than 10 per cent of uranine (used as fluorescent

tracer). Electrical charge was neutralized with a 85

Kr source before entering the test section. Since liquid

aerosols were used, adhesion of particles to the wall pipe was not considered a problem and no adhesive

coating was used.

The tests run for a period ranging from 15 to 60 minutes, and at the end of the test, the material

deposited inside the glass pipe was washed, and the concentration of uranine in the wash liquid was

determined by a fluorometer. Similarly the amount of uranine collected in the filter was also measured.


Liu, B.Y.H. and Agarwal J.K., “Experimental observation of aerosol deposition in turbulent flow”. Aerosol

Science Vol. 5, pp. 145-155 (1974)


Aerosol size: 1.4 – 21 μm (monodisperse particles) Reynolds number: 1×10

4 – 5×10

4

Dimensionless particle relaxation time: 0.21 – 774


m


Repeatability Check: Reproducibility of the experiments was checked in several tests. The results

showed that the maximum difference between duplicate tests was only 3 per cent (Liu and Agarwal, 1974).



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4.3.25 E3-25 - Aerosol Deposition in Turbulent Vertical Conduits (Wells)

Test Facility: N/A

Owner Organization: Atomic Energy Research Establishment


Radioactive particles were used to study deposition from airstreams on vertical surfaces. Three

different types of particles were used as an aerosol source: monodisperse droplets of tri-cresyl-phosphate

(tagged with 32

P), polystyrene spheres (tagged with 51

Cr), and Aitken nuclei (containing 212

Pb). The size

range of the particles that were used allowed Brownian deposition and impaction depletion mechanisms to

be investigated. Aerosols were transported by an air flow to the test section (0.5 m long), which consisted

of a 1.27 cm brass rod placed axially in a copper tube of 3.81 cm diameter.

The quantity of deposited particles on the central rod was measured at the end of the test from the

activity on the surface, measured by a scintillation counter. For this purpose a demountable, 2.54 cm long,

section was inserted in the test section.


Wells, A.C. and Chamberlain A.C., “Transport of small particles to vertical surfaces”. Brit. J. Appl. Phys.,

Vol. 18, pp. 1793-1799 (1967)


Aerosol size: 0.17 – 5 μm (monodisperse and polydisperse particles) Reynolds number: 1×10

3 – 5×10

4

Prandtl number: 105 - 10

6

Stopping distance: 1.8 – 80 μm


m

Rod diameter: 1.27×10-2

m


Repeatability Check: No information available



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4.3.26 E3-26 - CSE Fission Product Transport Tests

Test Facility: CSE

Owner Organization: Battelle Northwest


Six Containment Systems Experiment (CSE) tests (D-1, D-2, A-1, A-2, A-5 and A-11) were

performed to study fission product transport (using a simulant) within containment. In those experiments,

no engineered safety systems (i.e., spray systems) were provided and all fission product transport occurred

solely by natural, passive processes. The experiments were performed in two different sizes of

containment vessels (2,286 and 26,500 ft3). The largest vessel was about 1/5 linear scale of a large PWR

containment building. This large vessel had a total surface area to volume ratio (most important parameter

for iodine deposition) of approximately twice that of a typical large PWR.

After representative post accident conditions were established in the vessel, fission product stimulants

were injected in essentially an instantaneous manner. The initial iodine concentration exceeded 100 mg/m3

in three of the tests and was about 1 mg/m3 in the other three. Initial particle concentrations ranged from

0.1 to 10 mg/m3 for three types of particles – cesium, uranium (oxide) and particulate iodine. Time-

dependent measurements were made for mass concentrations in the vapour space at many locations, in the

condensate film on the vessel walls, and in the liquid pools which accumulated due to steam condensation.

After 1 to 2 days aging, the vessel was decontaminated to determine the final distribution of stimulant

materials between gas, liquid and paint.

A major parameter was changed in each experiment. Two vessel sizes were used (2,286 and 26,500

ft3). Temperatures of 180 and 250

ºF were investigated. Two initial iodine concentrations were used (1 and

100 mg/m3). The effect of heat transfer rate was measured by performing tests with both uninsulated and

insulated walls. Finally, the effect of unsteady temperature and pressure was determined by one test in

which temperature and pressure decayed with time.


Hilliard, R.K., and Coleman, L.F., “Natural Transport Effects of Fission Product Behaviour in the

Containment Systems Experiments”, Battelle Northwest Report BNWL-1457, Dec 1970.

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Parameters Varied between tests:

Vessel size: 2,286 and 26,500 ft 3

Temperature range: 180°F and 250°F

Atmosphere pressure varied

Paint age: 1-2 days

Iodine conc.: 3 tests >100 mg/m3, 3 tests ~1mg/m

3

Particle conc. for Cs, U oxide and particulate Iodine range from 0.1-10 mg/m3

Release period restricted to 10 min

The time dependence of iodine, methyl iodide, cesium, ruthenium and uranium concentrations was

measured at various locations in the vapour space and in the steam condensate.



Past Code Validation/Benchmarks: MELCOR 1.8.3


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4.3.27 E3-27 - CSE Aerosol Removal Tests

Test Facility: CSE

Owner Organization: BNWL & AEC


The CSE facility was sized to represent one-fifth linear model of a typical 1000 MWe PWR. The

vessel has an 870 m3 volume (7.6 m diameter and 20.4 m height) and it is divided into three compartments:

main room, middle room and lower room. Its nominal pressure was 0.52 MPa and its nominal leak rate

0.1%/day at the design pressure. Steam and fission product simulants were injected into the lower part of

the main room where convective flow was created. Stable isotopes of four classes of fission product

elements were used with radiochemical tracing isotopes. Iodine, caesium and UO2 were used in all the

tests, whereas tellurium, barium, ruthenium and xenon were used in selected tests. Initial airborne iodine

concentrations were varied, but averaged at ~150 mg/m3. Particle aerosol concentration (caesium and

UO2) were significantly lower (~10 mg/m3). Atmosphere in the vessel was sampled and the fission

products were characterized by maypacks as particulate (caesium, UO2 and iodine, etc.) and gas (elemental

iodine, methyl iodine and xenon).

Three tests series were performed. One of them was devoted to investigate natural aerosol transport

inside the vessel, with several tests aimed to determine the effect of atmosphere composition and

temperature on leakage rate (a variety of penetration and deliberate leak pathways existed in the vessel.

Three different types of atmosphere were studied: ambient temperature air, hot air, and steam-air mixtures.

The other test series were conducted to assess airborne fission product removal by sprays and filtration

systems from containment atmospheres.


Hilliard, R.K., Postma, A.K. 1981. “Large-Scale Fission Product Containment Tests” Nuclear Technology,

Vol. 53, n. 2, pp. 163-175, 1981

Postma, A.K. and Johnson, B.M., 1971. “Containment system experiment. Final program summary”,

BNWL-1592. Battelle Pacific Northwest Laboratories (July 1971)


Pressure: 0.52 MPa

Temperature: 83 - 121°C

Atmospheres: ambient air, hot air and steam-air mixture

Aerosol concentration: ~150 mg/m3 (iodine), ~10 mg/m

3 (Cs, U)



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4.3.28 E3-28 - LASS-SGTR

Test Facility: LASS-SGTR

Owner Organization: CIEMAT


The experiments were focused on aerosol deposition within the break stage of a failed and dry (i.e., no

water present) vertical steam generator, under conditions expected in SGTR severe accident sequences.

The facility consists of three main sections: the aerosol generator (based on the fluidization principle), the

injection line (through which particles reach the tube breach) and the tube bundle (a tube matrix of 11 x 11

tube which geometry is the real one in a steam generator). Suitable instrumentation is used at the inlet and

outlet of the bundle to characterize incoming and outgoing particles (TiO2). A mass balance was intended

by measuring the aerosol mass deposited in the bundle.

The variables studied were: the inlet gas mass flow rate (75 – 250 kg/h), the breach type (either

guillotine or fish mouth), and the gas flow orientation in the fish mouth tests (facing tube or diagonal). The

results were reported in terms of collection efficiency and showed low and moderate values (ranging from

5 to 30%, approximately) under all the conditions explored. Measurements of the aerosol size distribution

in the outlet (of the break) and inlet (of the pipe) gas are also available.

Note: LASS stands for Laboratory for Analysis of Safety Systems and is used to house several

experimental facilities (GIRS was one of them).


Herranz, L.E., Velasco, F.J.S. and del Prá C.L. “Aerosol retention near the tube breach during steam

generator tube rupture sequences”. Nuclear Technology, Vol. 154 (2006)

Peyres, V., Polo, J. and Herranz L.E. “SGTR project: Separate effect studies for vertical steam generators”

Informes Técnicos CIEMAT 1016 (march, 2003)

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Inlet gas flow: 70 – 250 kg/h (1.9×10-2

– 6.9×10-2

kg/s)

Particle size:

o AMMD: 3.4 – 7.4 μm

o GSD 1.5

Break type:

o guillotine

o fish-mouth 0.5D

o fish-mouth 1D





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4.3.29 E3-29 - MCE, UCE and HCE Tests

Test Facility: AECL Chalk River Hot Cells



The experiments were focused on fission-product release from fuel samples that were irradiated under

CANDU heavy water reactor conditions. The samples were inserted in horizontal or vertical tube furnaces,

heated to test temperature and subjected to air environment. Other fuel samples were also tested in the

same experimental apparatus, but in steam and inert gas test environments.

Twenty-one tests were performed on samples of unsheathed UO2 fuel at test temperatures ranging

from 1200°C to 2080°C [1-13, 17]. Releases of Kr, Xe, I, Cs, Ru, Ba, La, Nb, Zr and Ce isotopes were

measured using direct-viewing and gas-measurement gamma spectrometry techniques. The releases of La,

Nb, Zr and Ce were less than the fraction of the sample that was vaporized [10, 13]. Releases in oxidizing

environments (including air) were usually significantly more rapid than releases in inert environments.

Thirteen tests were performed on clad fuel samples at test temperatures ranging from 1350°C to

1890°C [12, 14-16, 18-22]. In most of the tests, the sample was a segment of a fuel element, with Zircaloy

end-caps press-fitted onto the ends of the segment; four tests were done on fuel samples that were only

fitted with one end-cap. Releases of Kr, Xe, I, Cs, Te, Ru and Ba isotopes were measured using direct-

viewing, gas-measurement and post-test scanning gamma spectrometry techniques. Release rates

(particularly at low temperatures) increased significantly after the cladding wall had been oxidized. There

was a notable delay (~3000 s) between release of volatile fission products (Kr, Xe, I and Cs) and release of

ruthenium for samples that had two end-caps.

Two tests were done on clad segments of PWR fuel with press-fitted end-caps at test temperatures of

1350°C and 1650°C [14, 20]. These samples were subjected to a long decay period before testing, so that

only Kr and Cs releases could be measured.


Open References

[1] Dickson, R.S., Peplinskie, R.T. and Gauthier, M.D., “Release of Fission Products From CANDU

Fuel in Air Environment”, 10th Int. Conf. on CANDU Fuel, 2008 October 5-8, Ottawa, Ontario,

Canada, AECL Report CW-126320-CONF-003, and references therein.

[2] Hunt, C.E.L., Iglesias, F.C., Cox, D.S., Keller, N.A., Barrand, R.D., Mitchell, J.R. and O'Connor,

R.F., “The release and transport of fission products during oxidation of UO2 in air,” Proc.

Symposium on Chemical Phenomena associated with Radioactivity Releases during Severe

Nuclear Plan Accidents, Anaheim, California, 1986 September 9-12, pp. 2-51 – 2-63

[3] Hunt, C.E.L., Iglesias, F.C., Cox, D.S., Keller, N.A., Barrand, R.D., Mitchell, J.R. and O’Connor,

R.F., “Fission product release during UO2 oxidation”, Proc. Int. Conf. on CANDU Fuel, Chalk

River, Ontario, 1986 October 6-8, pp. 508-526

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[4] Iglesias, F.C., Hunt, C.E.L., Cox, D.S., Keller, N.A., Barrand, R.D., O'Connor, R.F. and

Mitchell, J.R., “UO2 oxidation and fission product release,” Proc. Chemical Reactivity of Oxide

Fuel and Fission Product Release, Berkeley, Gloucestershire, England, 1987 April 7-9

[5] Cox, D.S., Iglesias, F.C., Hunt, C.E.L., Keller, N.A., Barrand, R.D., O'Connor, R.F. and

Mitchell, J.R., “Fission product release from UO2 in air during temperature ramps,” Proc. 8th

Annual Conference of Canadian Nuclear Society, St. John, New Brunswick, 1987 June 14-17,

pp. 58-64

[6] Hunt, C.E.L., Iglesias, F.C., Cox, D.S., Keller, N.A., Barrand, R.D., Mitchell, J.R. and O'Connor,

R.F., “UO2 oxidation in air or steam-release or retention of the fission products Ru, Ba, Ce, Eu,

Sb and Nb”, Proc. 8th Annual Conf. of Canadian Nuclear Society, St. John, New Brunswick,

1987 June 14-17, pp. 49-57

[7] Iglesias, F.C., Hunt, C.E.L., Garisto, F., Cox, D.S., Keller, N.A., Barrand, R.D., Mitchell, J.R.

and O'Connor, R.F., “Measured release kinetics of ruthenium from uranium oxides in air and

steam”, Proc. Int. Conf. on Thermal Reactor Safety, Avignon, France, 1988 October 2-7, paper

143

[8] Iglesias, F.C., Hunt, C.E.L., Garisto, F. and Cox, D.S., “Ruthenium release kinetics from uranium

oxides,” Proc. Fission Product Transport Processes in Reactor Accidents, Dubrovnik,

Yugoslavia, 1989 May 22-26

[9] Hunt, C.E.L., Cox, D.S., Liu, Z., Keller, N.A., Barrand, R.D., O'Connor, R.F. and Iglesias, F.C.,

“Ruthenium release in air,” Proc. 12th Annual Conference of the Canadian Nuclear Society,

Saskatoon, Saskatchewan, 1991 June 9-12, pp. 290-295

[10] Cox, D.S., Hunt, C.E.L., Liu, Z., Keller, N.A., Barrand, R.D., O'Connor, R.F. and Iglesias, F.C.,

“Fission-product releases from UO2 in air and inert conditions at 1700 2350 K: analysis of the

MCE-1 experiment,” Proc. Safety of Thermal Reactors Conference, American Nuclear Society,

Portland, Oregon, U.S., 1991 July 21-25, AECL Report AECL-10438, 1991 July

[11] Hunt, C.E.L., Cox, D.S., Liu, Z., Keller, N.A., Barrand, R.D., O'Connor, R.F. and Iglesias, F.C.,

“Xenon and ruthenium release from UO2 in air,” Proc. Safety of Thermal Reactors Conference,

Portland, Oregon, U.S., 1991 July 21-25

[12] Dickson, R.S., Liu, Z., Cox, D.S., Keller, N.A., O'Connor, R.F. and Barrand, R.D., “Cesium

release from CANDU fuel in argon, steam and air: the UCE12 experiment,” Proc. 15th Annual

Conference of the Canadian Nuclear Society, Montreal, Quebec, 1994 June 5-8, Session 3C,

AECL report AECL-CONF-00085

[13] Liu, Z., Cox, D.S., Dickson, R.S. and Elder, P.H., “Release of semi- and low-volatile fission

products from bare UO2 samples during post-irradiation annealing,” Proc. 15th Annual

Conference of the Canadian Nuclear Society, Montreal, Quebec, Canada, 1994 June 5-8, Session

5A, AECL report AECL-CONF-00087

[14] Cox, D.S., Liu, Z., Dickson, R.S. and Elder, P.H., “Fission-product releases during post-

irradiation annealing of high-burnup CANDU fuel”, Proc. 3rd

Int. Conf. on CANDU Fuel, Chalk

River, Ontario, 1992 October 4-8, pp. 4-61 to 4-73

[15] Barrand, R.D., Dickson, R.S., Liu, Z. and Semeniuk, D.D., “Release of fission products from

CANDU fuel in air, steam and argon atmospheres at 1500 1900°C: the HCE3 experiment”, Proc.

6th International Conference on CANDU Fuel, Niagara Falls, Ontario, 1999 September 26-30,

pp. 271-280

[16] Dickson, L.W. and Dickson, R.S., “Fission-product releases from CANDU fuel at 1650°C: the

HCE4 experiment”, Proc. 7th Int. Conf. on CANDU Fuel, Kingston, Ontario, Canada, 2001

September 23-27, pp. 3B-21 – 3B-30

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Closed References

[17] Liu, Z., Cox, D.S., Dickson, R.S. and Elder, P.H., “A Summary of CRL Fission Product Release

Measurements from UO2 Samples During Post Irradiation Annealing (1983-1992),” CANDU

Owners Group R&D report COG-92-377, 1994 May and references therein.

[18] Dickson, R.S., Lee, C.Y., Liu, Z., Hunt, C.E.L., Cox, D.S., Keller, N.A., O'Connor, R.F. and

Barrand, R.D., “Krypton, Cesium, Ruthenium and Antimony Release from UO2 Fragment

Specimens in Air, Steam and Argon Atmospheres at 1100°C-1625°C: UCE12 Data and Analysis

Report”, COG R&D report COG-92-330, 1994 February.

[19] Liu, Z., Hunt, C.E.L., Cox, D.S., Keller, N.A., O’Connor, R.F., Barrand, R.D., “Cesium and

Ruthenium Release from UO2 Mini-Element Specimens in Air, Steam and Argon (2% H2) at

1500 and 1600°C: HCE1 Data Report”, COG R&D Report COG-91-82, 1991 December

[20] Liu, Z., Hunt, C.E.L., Cox, D.S., Keller, N.A., Elder, P., O’Connor R.F., Barrand, R.D., Wood,

G., “Cesium Release from Irradiated CANDU and LWR Fuels in Steam and Air at 1350-1650°C:

HCE2 Quick-Look Report”, COG R&D Report COG-92-28, 1992 January

[21] R.S. Dickson, Z. Liu, R.D. Barrand, D.D. Semeniuk, D.S. Cox and P.H. Elder, “Release of

Xenon, Krypton, Iodine, Cesium, Tellurium and Ruthenium from Segments of CANDU Fuel in

Air, Steam, and Argon Atmospheres at 1500 1900°C: HCE3 Data Report Part I,” COG R&D

Report COG 95 357, 1998 June.

[22] L.W. Dickson, R.S. Dickson, Z. Liu, R.D. Barrand and D.D. Semeniuk, “Effect of Atmosphere

and Heating Rate on Fission Product Release from CANDU Fuel Heated to 1650°C: HCE4 Data

Report Part III,” COG R&D report COG-99-216, 2000 March


1200°C to 2080°C




Barber, D.H., Dickson, L.W., Dickson, R.S. and Audette-Stuart, M., “SOURCE IST 2.0 Validation

Approach”, 7th Int. Conf. on CANDU Fuel, Honey Harbour, Kingston, Ontario, 2001 September 23-27.

Barber, D.H., Parlatan, Y., Dickson, L.W., Corse, B., Kaye, M.H., Lewis, B.J., Thompson, W., Colins, K.,

Dickson, R.S., Hoang, Y., Lemire, R.J., McLean, C.G., Muir, W.C., Popescu, A., Szpunar, B. and Yatabe,

S., “SOURCE IST 2.0: Fission Product Release Code”, 9th Int. Conf. on CANDU Fuel, Belleville, Ontario,

2005 September 18-21.

Plumecocq, W.; Kissane, M.P.; Manenc, H.; Giordano, P., “Fission-product release modelling in the

ASTEC integral code: the status of the ELSA module”, 8th Int. Conf. on CANDU Fuel, Honey Harbour,

Ontario, 2003 September 21-24, pp. 540-550

Prepared By: R. Dickson (AECL)

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4.3.30 E3-30 - GBI Tests

Test Facility: AECL Chalk River Hot Cells



The experiments were focused on fission-product release from unsheathed irradiated UO2 fuel

samples in air at temperatures up to 1100°C. In addition to measuring the grain-boundary inventory (GBI)

[1-8], these conditions are relevant to the CANDU reactor end-fitting failure accident scenario, in which

fuel is ejected into containment. The samples were inserted in vertical tube furnaces, oxidized to powder

in air at 500°C, and then heated to the final test temperature. Releases of Kr, Xe, I, Cs, Te, Ru, Mo and Tc

isotopes were measured in 23 tests on fuel samples from well-characterized radial locations using direct-

viewing and gas-measurement gamma spectrometry techniques. The samples from were from four fuel

elements with burnups between 115 MWh/kgU and 540 MWh/kgU and peak linear powers between 26

kW/m and 58 kW/m. Many other tests were performed in which only Kr and Xe releases were measured

[5].


Open References

[1] Dickson, R.S., Peplinskie, R.T., and Gauthier, M.D., “Release of Fission Products From CANDU

Fuel in Air Environment”, 10th Int. Conf. on CANDU Fuel, 2008 October 5-8, Ottawa, Ontario,

Canada, AECL Report CW-126320-CONF-003.

[2] Hunt C.E.L., Iglesias F.C., Cox D.S., Keller N.A., Barrand R.D., O'Connor R.F., Mitchell J.R.,

Wood G.W. and Mikuch R., “Fission product grain-boundary inventory”, Proc. 10th Annual

Conf. of the Canadian Nuclear Society, Ottawa, ON, 1989 June, AECL Report AECL-10036.

[3] Elder P.H., Cox D.S., Dickson L.W. and Murphy R.V., “New post-irradiation examination

techniques at Chalk River Laboratories: gamma tomography and grain-boundary-inventory

measurements on irradiated fuel”, Proc Recent Developments on Post-Irradiation Examination

Techniques for Water Reactor Fuel, Cadarache, France, 1994 October 17-21

[4] Elder P.H., Cox D.S., Liu Z., Dickson R.S. and Bilanovic Z., “Measurement of krypton grain-

boundary inventories in CANDU fuel”, Proc. 4th Int. Conf. on CANDU Fuel, Pembroke, Ontario,

1995 October 1-4, pp. 6B-48 – 6B-57

[5] Dickson R.S., O’Connor R.F. and Semeniuk D.D., “Grain-boundary inventories of krypton in

CANDU fuel,” Proc. Fission Gas Behaviour in Water Reactor Fuels, Cadarache, France, 2000

September, pp. 337-346, AECL Report AECL-CONF-00145

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Closed references

[6] Liu, Z., Cox, D.S., Dickson, R.S. and Elder, P.H., “A Summary of CRL Fission Product Release

Measurements from UO2 Samples During Post Irradiation Annealing (1983-1992),” CANDU

Owners Group R&D report COG 92 377, 1994 May.

[7] Elder, P.H., O'Connor, R.F., Semeniuk, D.D., Liu, Z., Dickson, R.S., Keller, N.A.. Kunkel, T.J.,

Barrand, R.D., Shields, D.F., Wood, G.W. and Cox, D.S., “Measurement of the Krypton Grain

Boundary Inventory in Irradiated CANDU Fuel: The GBI2 Experiment Analysis Report Part I,”

COG R&D report COG-93-186, 1997 March.

[8]. Elder, P.H., O'Connor, R.F., Cox, D.S. and Dickson, R.S., “Krypton Grain-Boundary Inventories

in Irradiated CANDU Fuel: The GBI3 Experiment Data Report,” COG R&D report COG-95-

188, 1997 March.


500°C to 1100°C




Barber, D.H., Dickson, L.W., Dickson, R.S. and Audette-Stuart, M., “SOURCE IST 2.0 Validation

Approach”, 7th Int. Conf. on CANDU Fuel, Honey Harbour, Kingston, Ontario, 2001 September 23-27.

Barber, D.H., Parlatan, Y., Dickson, L.W., Corse, B., Kaye, M.H., Lewis, B.J., Thompson, W., Colins, K.,

Dickson, R.S., Hoang, Y., Lemire, R.J., McLean, C.G., Muir, W.C., Popescu, A., Szpunar, B. and Yatabe,

S., “SOURCE IST 2.0: Fission Product Release Code”, 9th Int. Conf. on CANDU Fuel, Belleville, Ontario,

2005 September 18-21.

Plumecocq, W.; Kissane, M.P.; Manenc, H.; Giordano, P., “Fission-product release modelling in the

ASTEC integral code: the status of the ELSA module”, 8th Int. Conf. on CANDU Fuel, Honey Harbour,

Ontario, 2003 September 21-24, pp. 540-550

Prepared By: R. Dickson (AECL)

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4.3.31 E3-31 - Aerosol Trapping Effects in Containment Penetration (A. Watanabe)

Test Facility: N/A

Owner Organization: NUPEC


These tests investigated containment penetrations behavior under accident conditions. Two aspects

were addressed: failure temperature and fission product trapping along the leakage paths. Tests were

conducted using actual containment penetrations of a BWR plant (electrical penetration assemblies and

hatch flanges). Most of test pieces were heated up to 400 ºC and pressure was increased up to a maximum

value of 1 MPa. Test pieces were also irradiated to 800 kGy simulating the maximum accumulated dose in

containment.

Caesium iodide (CsI) was injected as a representative fission product aerosol. An atomizing type of

aerosol generator was used and a constant aerosol concentration could be maintained over 10 hours. The

maximum aerosol concentration generated by this method is more than 1 g/m3. Real time aerosol

concentrations in the inlet and outlet test pieces as well as the particle size distribution were obtained with

an optical particle counter. Detectable range of aerosol particle size was from 0.5 to 15 μm aerodynamic

mass median diameter. Representative aerosol concentrations and their diameters were selected to be

around 100 mg/m3 and 1 μm respectively. An electrical steam boiler and a gas compressor were used to

supply superheated steam-air mixture. A mixing chamber was used to mix supplied CsI aerosol and the

superheated steam-air uniformly. Online measurements were: inlet and outlet gas temperature, piece

surface temperatures, inlet and outlet pressures and gas flow rate through the leakage path.

Three types of tests were carried out: integrity tests, failure criteria and aerosol trapping tests. For the

integrity test two types of heating modes were used: steady heating at ~200ºC for 20h and cyclic one, with

temperature varying between 200 ºC and 120 ºC and pressure between 0.8 MPa and 0.1 MPa during 30 to

50 hours. For the failure test conditions the heating of the test section was kept until leakage occurrence.


A. Watanabe, T. Hashimoto and M. Osaki. 1998. “Fission Product Aerosol Trapping Effects in the

Leakage Path of Containment Penetration under Severe Accident Conditions” in Proc. 3rd OECD

Specialist Meeting on Nuclear Aerosols in Reactor Safety, Cologne, Germany, 1998

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Aerosol concentration: ~100 mg/m3

Particle size: ~1 μm

Temperature: <200ºC

Pressure: 1.1×105 – 5.5×10

5 Pa





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4.3.32 E3-32 - Aerosol transfer through cracked concrete walls

Test Facility: N/A



The objectives of this experiment are to determine the contamination (gas and aerosol) transfer

models through cracked concrete walls. The experiments are performed on concrete walls (128 cm in

width, 75 cm in height and 10 cm in thickness) cracked by shear stresses.

The experimental device is divided into three parts. In the centre is the wall with the crack network

isolated by two boxes. The upstream part contains an aerosol generator that produces a soda fluorescent

aerosol with a controlled diameter and a known mass concentration. The downstream part enables

controlling the mass flow rate and the pressure drop between upstream and downstream sides. A High-

Efficiency Particulate Air (HEPA) filter collects the penetrating aerosols in order to calculate the

downstream aerosol mass concentration

The first part of the experiment consists of measuring the gas flow through three cracked concrete

walls with and without shear stresses. The measurements are performed on three walls subjected to a

vertical stress that is representative of the building weight (150 kN) and to different horizontal shear

stresses of alternate directions up to 645 kN that are representative of a seismic activity. The second part

consists of studying the behaviour of aerosols in a crack network. For this, some experiments are

performed with different aerosol sizes to reach different kinds of deposition (diffusion, sedimentation and

impaction at the crack entrance). Four different aerosol sizes were used: 0.06, 0.8, 1.1 and 4.1 μm in

aerodynamic mass median diameter.

Finally, an additional experiment was performed to determine the crack network volume, depending

on the crack characteristics. Helium was injected as a tracer gas into the upstream part of the cracked wall

and to measure its transfer time to the downstream part.


T. Gelain and J. Vendel. 2008. “Research work on contamination transfers through cracked concrete

walls”. Nuclear Engineering and Design 238 (2008), pp. 1159-1165

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AMMD: 0.06, 0.8, 1.1 and 4.1 μm.

Gas flow rate: ~5×10-5

– 6×10-4

m3/s





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4.3.33 E3-33 - Whiteshell Steam Jet Experiments

Test Facility: Whiteshell Flashing Jet Facility



The boiler used in the flashing jet study at WL (E3-16 - Whiteshell Flashing Jet Tests) was modified

for aerosol characterization in steam jets. The boiler vessel could be heated to 311C at a pressure of 10

MPa.

In the steam-jet test, the water was heated to about 130C to remove any undesirable gases, mostly

air. As boiling occurred, the steam was allowed to escape for 15 minutes. The temperature and pressure

were set to the desired operating conditions. Once the water reached saturation temperature and pressure, a

steam blow-down was initiated from the head-space inside the boiler. The steam was discharged through a

0.305-cm diameter nozzle fitted at the downstream end of the steam-discharge pipe connected to the boiler

head.

A DANTEC phase doppler anemometer was used to measure the aerosol size distribution and velocity

(60 cm away from the nozzle). The jet temperature profiles were measured using thermocouples

positioned in the jet as it expanded from the nozzle.


M.H. Ungurian and J. McFarlane, “A Facility to Measure Droplet Characteristics in a High-Temperature

High-Pressure Steam Jet”, RC-2187, AECL report, 2000

J. McFarlane and M.H. Ungurian, “Characterization of Droplet Formation in a Condensing Steam Jet. Part

I: Measurements along the Jet Axis”, RC-2485, AECL Report, 2000

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Nozzle Diameter: 0.305 cm

o Upstream Pressure: 1 to 10 MPa

o Upstream Steam Temperature: 1 to 6oC superheat

Nozzle Diameters: 0.089and 0.48 cm

o Upstream Pressure: 0.1 to 1.3 MPa

o Upstream Steam Temperature: 113 to 207oC





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4.3.34 E3-34 - WALE

Test Facility: WALE



The WALE (Water Aerosol Leakage Experiment) test facility consists of a 19 m3 cylindrical pressure

vessel, 2.24 m outside diameter, and 5.23 m high, with dished heads and a steam heating jacket. A vent

line, nozzle extends 500 mm into the vessel, discharges de-ionized water at high pressure and temperature

into the vessel. A chemical tracer (CsNO3) solution is added upstream of the nozzle. The flashing water

jet is directed either horizontally towards an impingement plate, or vertically upwards. The distance

between the nozzle and the impingement plate is varied.

Figure 4.3.34-1 Schematic of the WALE Test Facility

Steam and liquid aerosol exit from the containment vessel via a vent line. The entrance to the vent

line is sized such that the steam exit velocity is about 10 m/s. This flow is selected to permit the orifice to

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be large enough to pass the larger aerosols, yet small enough to keep the flow rate sufficient to reduce

steam condensation.

Multiple troughs located throughout the facility are used to collect the falling liquid and permit

aerosol mass measurements to be made for different aerosol removal conditions. A forward scattering

spectrometer probe (FSSP) is used to measure particle size distributions at various locations in the vessel

and the vent line. The FSSP uses laser light scattering to count and size particles between 1 and 95m

in diameter.

A drop sampler probe was used to measure droplet size distributions in some of the tests. This device,

located inside the WALE vessel, delivers a pneumatically fired “bullet” at a controlled velocity across the

sample field and intercepts any droplets in its path. The leading face of the “bullet” is glass with a coating

which maintains an imprint of drops impinging the surface. The imprinted information on the bullet is

characterized to determine number density and droplet sizes using correlation factors relating the imprint to

the size of the intercepted droplet.


Koziak, W.W., C.F. Forrest and R.J. Fluke, “Water Aerosol Leakage Experiments: Objectives, Test

Matrix, Facility Description”, OH-DD-88463, COG-88-158, 1988.

R.J. Fluke, K.R. Weaver, G.L. Ogram, L.N. Rogers and C.F. Forrest, “The Water Aerosol Leakage

Experiments: Programme Description and Preliminary Results”, 2nd

Int. Conf. on Containment Design and

Operation, Toronto, 1990 October

Forrest, C.F., “Water Aerosol Leakage Experiments – Results of High Flow Tests”, Stern Laboratories

Report SL-037, COG Report, COG-91-123, 1992

C.F. Forrest, “Water Aerosol Leakage Experiments – December 1989 Test Results”, Stern Laboratories

Inc., SL-021, COG-91-122, 1993 January

R.J. Fluke, G.L. Ogram, L.N. Rogers and K.R. Weaver, “Aerosol Behaviour in the Water Aerosol Leakage

Experiments”, 2nd

Int. Conf. on Containment Design and Operation, Toronto, 1990 October

C.F. Forrest, “Water Aerosol Leakage Experiments – Decommissioning Report”, Stern Laboratories

Report SL-098, COG Report, COG-97-339, 1998

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Discharge Line Pressure: 3, 4.8, 10 MPa

Discharge Line Temperature: Saturation to 65oC subcooled

Discharge Flowrate: 0.03 - 0.8 kg/s

Containment Pressure: 102 kPa

Containment Temperature: 100oC

Nozzle Diameter: 0.025, 0.1, 0.907, 1.45, 2.11, 3.07 mm

Plate Separation: 0.037 m to 1.38 m




SMART validation: Carlson, P.A., “Validation Exercises for SMART-IST VER-0.300 – Jet

Impingement”, COG Report COG-01-035, Nov 2011.


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4.3.35 E3-35 – AEREST (Aerosol resuspension shock tube)

Test Facility: AEREST

Owner Organization: Technical University of Munich (TU München)


The aerosol test section is located in a 7 m long pipe with a pressure vessel in an end (up to 160 bar).

Viscosity and density of the fluid are controlled by the initial temperature of the vessel which is equipped

with an electric heater. The experiment is started by an abrupt expansion of the vessel content by means of

a high speed ball valve and/or a rupture disk. The initial pressure, initial temperature and ball valve end

position determine the duration and intensity of the released flow wave. Within a flow conditioning

section the leading shock wave is broken by means of built-ins and the flow is parallelised by honeycombs.

Behind the conditioning section the expansion wave is transmitted to the optically accessible test section

with the aerosol deposition plate (prepared in an extra sedimentation vessel) located at the bottom. The

different geometric configurations of the aerosol deposition plate allow different angles of attack to the

deposits layer.

An aerosol layer ranging from 0.1 to 5 mg was deposited on the plate in a separated vessel. The

aerosol materials used are silver particles of different shapes and tin oxide (TiO2).

To detect the total mass of deposit during resuspension, a light scattering method has been developed.

The aerosol layer, deposited on a transparent glass plate, is illuminated by a constant-light source. By

means of an avalanche photo diode and an amplifier, the scattered light originating from the deposed

particles is detected. The amount of resuspended particles has been detected locally by means of a laser-

extinction method which takes advantage of the increasing absorption coefficient with an increasing

particle density. The particle movement has also been recorded by high speed particle tracking (high speed

CCD camera).





N. Ardey and F. Mayinger. 1998. Aerosol Resuspension by Highly Transient Flow – Insights by means of

a Laser Optical Methods. Kerntechnik (1998) 68-75.

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Aerosol deposits layer: 0.1 – 5 mg





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4.3.36 E3-36 – VANAM-M4

Test Facility: Battelle Model Containment (BMC)



The experiment was performed in the multi-compartment Battelle Model Containment (BMC). The

BMC was built from reinforced concrete, had a free volume of 640 m³ (an inner diameter of 11.2 m and an

internal height of 9.8 m). It was designed to be a 1/64 representation of the Biblis B containment.

The VANAM M4 experiment investigated the influence of hydrogen deflagration on the aerosol

airborne concentration. In a first step of the experiment, hygroscopic sodium hydroxide (NaOH) aerosols

and insoluble tin dioxide (SnO2) aerosols were injected into the inner compartments of the BMC and were

left to settle. In the second step hydrogen was injected into the same compartment as the aerosols. The

hydrogen was ignited in the inner rooms and the expanding atmosphere spread out into the outer

containment rooms, separated from the inner rooms by rupture foils. Although the deflagration ran very

mildly, the generated air currents were capable of resuspending around 1% of the initially injected aerosol

mass.





M. Bendiab. 2007. Erweiterung des Containment Codes COCOSYS zur Quelltermbewertung der

trockenen Resuspension infolge transienter Strömungen. PhD Thesis. Ruhr-Universität Bochum


Hygroscopic (NaOH) and insoluble (SnO2) aerosols.





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4.3.37 E3-37 – THAI Aer-1, Aer-3 and Aer-4 tests

Test Facility: THAI



The THAI facility is a cylindrical steel vessel with a volume of 60 m3 (9.2 m height and 3.2 m

diameter). Three tests were carried out to investigate the aerosol resuspension: Aer-1, Aer-3 and Aer-4.

For this purpose, a vertical deflagration tube was installed inside the facility. In the first phase of the

experiment, CsI aerosol was injected into the vessel (dry atmosphere) and was left to settle for a period of

25 h on the surfaces inside the vessel. In a second phase the deflagration tube was filled with hydrogen

and the resulting hydrogen-air mixture was ignited at a lower point of the tube. The expanding hot

atmosphere was released through a 2x50 cm2 large nozzle. The air flow ran over a deposition plate. This

yielded resuspension of aerosol material into the atmosphere. The hydrogen load inside the deflagration

tube was varied in the experiments. Thus different airflow velocities in the range of 17 m/s up to 67 m/s

over the deposition plate could be adjusted. This experimental setup was chosen to gain an air flow with

relatively well defined velocities in front of the nozzle.

The CsI aerosol was characterized by means of low pressure impactors, filters, sedimentation coupons

and others during the CsI settling and resuspension phases. The surface loading was measured before each

deflagration.





H. Nowack and H.J. Allelein. 2007. Dry aerosol resuspension after a hydrogen deflagration in the

containment. Nuclear Energy for New Europe 2007. September 10-13. Portoroz (Slovenia)

M. Sonnenkalb and G. Poss. The international test programme in the THAI facility and its use for code

validation. EURSAFE document.


Gas velocity: 17 – 67 m/s

Maximum aerosol concentration: ~10-3 kg/m3


Repeatability Check: Three tests were done with similar conditions



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4.3.38 E3-38 – Phebus FPT4 Revaporization

Test Facility: Phebus FP

Owner Organization: IRSN (Phebus FP Project)


The samples were from the in-reactor experiment Phebus FPT4, in which fission products and

actinides were released from a fuel debris bed (near full oxidation) without control rod material, to look at

late phase core degradation and the releases of low volatile fission products and actinides. The substrate of

the samples was thoria (usually chemically passive) about 1 m above the fuel debris bed. The re-

volatilisation studies were performed at temperatures up to 1000°C, in steam and reducing environments.

The re-volatilised materials included Cs, Ag, Cd, Sn, and U.


P.D.W. Bottomley, P. Carbol , J-P. Glatz, D. Knoche, D. Papaioannou, D .Solatie, S. Van Winckel, A-C.

Gregoire, G. Gregoire, D. Jacquemain, “Fission product and actinide release from the debris bed test

Phebus FPT4: Synthesis of the post test analyses and of the revaporisation testing of the plenum samples”,

Nuclear Engineering and Technology 38 (2006) 163-174.

P.D.W. Bottomley, T. Gouder, F. Huber, D. Papaioannou, D. Pellottiero, Mikrochimica Acta 145 (2004) 3-

12.


Maximum temperature 1000°C

Re-volatilisation environments were steam, steam-H2 and H2-N2 mixtures



Past Code Validation/Benchmarks: FPT4 releases have been used for validation, but the re-

volatilization data are not known to be used for validation

Prepared By: R.S. Dickson (AECL)

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4.3.39 E3-39 – Ruthenium Revolatilisation Studies at VTT

Test Facility:

Owner Organization: VTT


The formation and transport of volatile ruthenium oxides was studied by exposing RuO2 powder to

diverse oxidising atmospheres at a relatively high temperature. Transport of gaseous RuO4 was further

investigated by injecting it into the facility in similar conditions. The sample environments were slightly

moist air and air-steam mixtures. Upon cooling of the gas flow, RuO2 aerosol particles were formed in the

system. Higher sample temperatures favoured the formation of RuO2 aerosol particles.


T. Kärkelä, U. Backman, A. Auvinen, R. Zilliacus, M. Lipponen, T. Kekki, U. Tapper and J. Jokiniemi,

“Experiments on the behaviour of ruthenium in air ingress accidents,” Nordic Nuclear Safety Research

Report NKS-151, VTT report VTT-R-01252-07, 2007 March

A. Auvinen, U. Backman, J. Jokiniemi, M. Lipponen, R. Zilliacus; M. Kissane; I. Nagy, M. Kunstár, N.

Vér, “Investigation on Ruthenium Transport in Highly Oxidising Conditions,” Conference paper/ PU,

SARNET-ST-P39

P. Giordano, A. Auvinen, G. Brillant, J. Colombani, N. Davidovich, R. Dickson, T. Haste, T. Kärkelä, J.S.

Lamy, C. Mun, D. Ohai, Y. Pontillon, M. Steinbrück, and N. Vér, “Ruthenium behaviour under air ingress

conditions: main achievements in the SARNET project,” Proceedings of ERMSAR 2008 Meeting,

Nesseber, Bulgaria, 23-25 September 2008

A. Auvinen, G. Brillant, N. Davidovich, R. Dickson, G. Ducros, Y. Dutheillet, P Giordano, M. Kunstar, T.

Kärkelä, M. Mladin, Y. Pontillon, C. Séropian and N. Vér, Progress on Ruthenium Release and Transport

under Air Ingress Conditions”, Nucl. Eng. and Design, Volume 238, Issue 12, December 2008, pp. 3418-

3428


Vaporization temperatures between 730°C and 1430°C

Sample environments: slightly moist air and air-steam mixtures





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4.3.40 E3-40 – Ruthenium Transport and Revolatilisation Studies at KFKI

Test Facility: RUSET

Owner Organization: AEKI and KFKI (Hungary)


In separate effect tests at 1000-1200°C, the oxidation rate of Ru and the amount of Ru in outlet air

flow were studied in the presence of other fission product elements to determine the effect of these other

elements on Ru oxidation and transport. In a decreasing temperature section (1100-600C) most of the

RuO3 and RuO4 (~95%) decomposed and formed RuO2 crystals; the partial pressure of RuO4 in the outlet

air was in the range of 10-6

bar. The re-evaporation of deposited RuO2 resulted in about 10-6

bar partial

pressure in the outlet gas as well. Measurements demonstrated the importance of surface quality in the

decreasing temperature area on the heterogeneous phase decomposition of ruthenium oxides to RuO2.

Steam or molybdenum oxide vapour in air decreased the surface catalyzed decomposition of RuOx to RuO2

and increased RuO4 concentration in the outlet air. High temperature reaction with caesium changed the

form of the released ruthenium and caused a time delay in appearance of maximum concentration of

ruthenium oxides in the ambient temperature outlet gas, while reaction with Ba and Nd/Ce oxides

increased Ru escape from the high temperature area as Ba and rare earth ruthenates.


N. Vér, L. Matus, M. Kunstár, J. Osán, Z. Hózer, and A. Pintér, “Influence of fission products on

ruthenium oxidation and transport in air ingress nuclear accidents”, J. Nucl. Mater. 396 (2010) 208-217

A. Auvinen, G. Brillant, N. Davidovich, R. Dickson, G. Ducros, Y. Dutheillet, P Giordano, M. Kunstar, T.

Kärkelä, M. Mladin, Y. Pontillon, C. Séropian and N. Vér, Progress on Ruthenium Release and Transport

under Air Ingress Conditions”, Nucl. Eng. and Design, Volume 238, Issue 12, December 2008, pp. 3418-

3428


Sample temperatures: 1000°C to 1200°C

Sample environment: air and air-steam mixtures


Repeatability Check: Limited



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4.3.41 E3-41 – Ruthenium deposition studies at Chalmers University

Test Facility:

Owner Organization: Chalmers University (Goteborg, Sweden)


RuO4(g) interactions with metal substrates were studied at 20°C to 50°C in glass apparatus in air and

nitrogen environments (the deposition did not change for these environments). Deposits of RuO2·nH2O(s)

were found on glass surfaces (and to a much greater extent on zinc surfaces) and copper had both

Cu(RuO2(OH)4) and RuO2 deposits. Significantly less deposition occurred on aluminium surfaces.


J. Holm, H. Glänneskog and C. Ekberg “Interactions of RuO4(g) with different surfaces in nuclear reactor

containments,” Nordic Nuclear Safety Research Report NKS-166, 2008 July

J. Holm, H. Glänneskog and C. Ekberg “Deposition of RuO4(g) on various surfaces in a nuclear reactor

containment,” J. Nucl. Mater. 392 (2009) 55-62


Sample temperatures: 20°C to 50°C

Sample environments: air and nitrogen (saturated with water vapour)





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4.3.42 E3-42 – Ruthenium Revolatilisation Studies at IRSN

Test Facility: EPICUR



The decomposition of RuO4 in air on stainless steel and paint substrates was found to depend on

temperature and the presence of steam. The deposited ruthenium (in the form of RuO2·nH2O) did not

interact with the substrate. Oxidation of the deposited ruthenium by ozone or air radiolysis products

yielded RuO4 again. Air radiolysis products were more effective than ozone in re-volatilising the Ru

deposits. In studies in the EPICUR facility, a large fraction of RuO4 was thermally volatilised from the

water, RuO4- was significantly less volatile and required irradiation to be volatilised, and RuO2·nH2O was

not volatile.


Mun, C., Cantrel, L. and Madic, C., “Study of RuO4 Decomposition in Dry and Moist Air”, Radiochimica

Acta. 95(11), pp. 643-656 (2007).

C. Mun, J.J. Ehrhardt, J. Lambert; and C. Madic, “XPS Investigations of Ruthenium Deposited onto

Representative Inner Surfaces of Nuclear Reactor Containment Buildings”, Applied Surface Science

253(2007) 7613-7621

C. Mun, L. Cantrel; and C. Madic, “Oxidation of Ruthenium Oxide Deposits by Ozone”, Radiochimica

Acta 96 (2008) 375-384

C. Mun, L. Cantrel; and C. Madic, “Radiolytic Oxidation of Ruthenium Oxide Deposits”, Nucl.

Technology 164 (2008) 245-254


Test temperatures: 40°C to 90°C

Sample environments: dry air, moist air, and air-steam mixtures

Some tests in radiation environment



Past Code Validation/Benchmarks: Used in model development in ASTEC 2.0


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4.4 Iodine Chemistry Experiments

4.4.1 E4-1 - CFTF Charcoal Filter Test

Test Facility: CFTF



The Charcoal Filter Test Facility (CFTF) is designed to determine the efficiency of TEDA

(triethylenediamine) impregnated charcoal for trapping CH3I under postulated accident conditions.


A.C. Vikis, J.C. Wren, C.J. Moore and R.J. Fluke, “Long-Term Desorption of 131

I from KI-Impregnated

Charcoals Loaded with CH3I Under Simulated Post-LOCA Conditions”, Proc. of the 18th DOE Nuclear

Airborne Waste Management and Air Cleaning Conf., Baltimore, Maryland, August 12-16, 1984, CONF-

840806, p. 65, U.S. Department of Energy (1985).

J.C. Wren and C.J. Moore, “Long-Term Desorption of CH3I From a TEDA-Impregnated Charcoal Bed

Under Post-LOCA Conditions”, Proc. of the 20th DOE/NRC Nuclear Air Cleaning Conf., Boston,

Massachusetts, August 22-25, 1988, p. 1117, U.S. Department of Energy (1989).

J.C. Wren, C.J. Moore, A.C. Vikis and R.J. Fluke, “A Study of the Performance of Charcoals Filters Under

Post-LOCA Conditions”, Proc. of the 20th DOE/NRC Nuclear Air Cleaning Conf., Boston, Massachusetts,

August 22-25, 1988, p. 786, U.S. Department of Energy (1988).

J.C. Wren and C.J. Moore, “The Effect of Weathering on Charcoal Filter Performance.

I: The Adsorption and Desorption Behaviour of Contaminants”, Nucl. Technol. 94, 242 (1991).

J.C. Wren and C.J. Moore, “The Effect of Weathering on Charcoal Filter Performance.

II: The Effect of Contaminants on the CH3I Removal Efficiency of TEDA Charcoal”, Nucl. Technol., 94,

252 (1991).

J.C. Wren, W. Long, C.J. Moore and K.R. Weaver, “Modelling the Removal and Retention of Radioiodine

by TEDA-Impregnated Charcoal Under Reactor Accident Conditions”, Nucl. Technol., 125, 13 (1999).

J.C. Wren, C.J. Moore, M.T. Rasmussen and K.R. Weaver, “Methyl Iodide Trapping Efficiency of Aged

Charcoal Samples from Bruce A EFADS”, Nucl. Technol., 125, 28 (1999).

J.C. Wren, C.J. Moore, and Z. Qin, “Dynamic Adsorption of CH3I from Flowing Airstreams on Activated

TEDA Charcoal”, Int. J. of Materials Eng. and Tech. 3:1-32 (2010).

Z. Qin, J.C. Wren, and C.J. Moore, “A Temperature Model for an EFADS Charcoal Filter Exposed to a

Dry Airflow”, 55th Canadian Chem. Eng. Conf., Toronto, ON, Oct. 15-18, 2005.

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Temperature: 20 to 110°C.

Charcoal bed depth: 2.5 to 20 cm

Charcoal bed diameter: 5 cm

Relative humidity: 0 to 95% RH

CH3I challenge concentration: 5e-6 to 1 g/l

Measured Parameters

Decontamination Factor

Dynamic Adsorption capacity

Bed Distribution

Temperature Transient




J.C. Wren, W. Long, C.J. Moore and K.R. Weaver, “Modelling the Removal and Retention of Radioiodine

by TEDA-Impregnated Charcoal Under Reactor Accident Conditions”, Nucl. Technol., 125, 13 (1999).

Z. Qin, J. C. Wren and C. J. Moore, “Modeling the Iodine Removal Efficiency and Temperature Behaviour

for an FADS Charcoal Filter by FEMLAB”, Proc. of FEMLAB – COMSOL Multiphysics Conference,

Oct. 23-25, 2005, Boston, MA, edited by J. Hiller, 119-123 (2005).

Prepared By: G. Glowa (AECL)

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4.4.2 E4-2 - RTF P9T3

Test Facility: RTF



The Radioiodine Test Facility (RTF) studied iodine behaviour in containment.


Glowa, G.A. and Ball, J.M., 2007. The Radioiodine Test Facility Program. Severe Accident Research

Network Report, SARNET-ST-P60.

Glowa, G.A. and Ball, J.M., 2007. Radioiodine Test Facility P9T3 Data. Severe Accident Research

Network Report, SARNET-ST-P61.


Dose Rate (1-2 kGy/h)

Temperature (25-90C)

pH (4-10)

Surfaces (steel and paint)

Total gas phase iodine (<1e-8 M)

Total aqueous iodine (<1e-5 M )





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4.4.3 E4-3 - RTF P9T1

Test Facility: RTF



The Radioiodine Test Facility (RTF) studied iodine behaviour in containment


Glowa, G.A. and Ball, J.M., 2010. Radioiodine Test Facility P9T1 Test Report. Atomic Energy of Canada

Limited Report, 153-126530-440-008, Rev. 0.




pH (4-10)








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4.4.4 E4-4 - RTF P9T2

Test Facility: RTF





Glowa, G.A., 2010. Radioiodine Test Facility P9T2 Test Report. Atomic Energy of Canada Limited

Report, 153-126530-440-012, Rev. 0.




pH (4-10)








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4.4.5 E4-5 - RTF P10T2

Test Facility: RTF










pH (4-10)








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4.4.6 E4-6 - RTF P10T3

Test Facility: RTF










pH (4-10)








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4.4.7 E4-7 - RTF P11T1

Test Facility: RTF





Glowa, G.A., 2007. Radioiodine Test Facility P11T1 Test Report. Atomic Energy of Canada Limited

Report, 153-126530-440-002.




pH (4-10)








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4.4.8 E4-8 - RTF P0T2

Test Facility: RTF





Ball, J.M., Hnatiw, J.B., Palson, A., Portman, R., and Sanipelli, G.G., 1993. Radioiodine Test Facility

Phase 0 – Test 2 (Stainless Steel, 60

Co Source) Data Report. Atomic Energy of Canada Limited Report,

COG-93-49.

Ball J.M., 2000. ISP 41 Containment Iodine Computer Code Exercise Based on a Radioiodine Test Facility

(RTF) Experiment. NEA/CSNI Report, R(2000)6/ Volumes 1 and 2.




pH (4-10)






Past Code Validation/Benchmarks: ISP-41

Ball J.M., 2000. ISP 41 Containment Iodine Computer Code Exercise Based on a Radioiodine Test Facility

(RTF) Experiment. NEA/CSNI Report, R(2000)6/ Volumes 1 and 2.


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4.4.9 E4-9 - RTF P10T1

Test Facility: RTF





G.A. Glowa, J.M. Ball, J. Merritt, R. Portman and G.G. Sanipelli “Radioiodine Test Facility Phase 10 Test

1 Data Report”, Atomic Energy of Canada Limited Report, RC-2050, (2000).

ISP-41 Follow-up Exercise Phase 2 (RTF and CAIMAN Experiments) - Final Comparison and

Interpretation Report NEA/CSNI/R(2004)16



Temperature (25-90 C)

pH 4-10










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4.4.10 E4-10 - RTF PHEBUS RTF1

Test Facility: RTF





Ball, J.M., Chuaqui, C.A, Merritt, J.A., Portman, R, Sanipelli, G.G., and Wren, J.C., 1996. Results from

PHEBUS RTF1 Test: Final Report. Atomic Energy of Canada Limited Report, COG-95-53, COG-

PHEBUS-FP-04.



Temperature (25-90 C)

pH 4-10










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4.4.11 E4-11 - EPICUR Test Series S1, S2 and S3

Test Facility: EPICUR

Owner Organization: IRSN, CEA, EDF + ISTP partners


The EPICUR (Experimental Programme of Iodine Chemistry Under Radiation) tests address the

iodine chemistry under severe accident conditions. The test data will help reduce the uncertainties on the

release of radioactive iodine during a severe accident. The tests are part of the International Source Term

Programme (ISTP) with the partnership of European Commission, AECL, PSI, KINS, and USNRC.

All the experiments are performed under radiation and use labeled iodine. 19 tests in 3 series S1, S2

and S3, were performed.


Guilbert, S. t al;, “Radiolytic oxidation of iodine in the containment at high dose rate”, Nuclear Energy for

New Europe 200, Portoroz, Slovenia, September 1-13, 2007

S. Guilbert et al., “Formation of organic iodide in the containment in case of a severe accident”, American

Nuclear society Meeting 2008, June 8-13, Anaheim (CA)

J. Colombani et al., “Experimental Study of organic iodide formation on painted surfaces in the

containment during a severe accident”, Proceedings of ICAPP 2011, May 2-6, 2011, Nice (France)


The tests are performed in an irradiation tank with a volume of a few liters at temperatures ranging

from 80 to 120°C, including or not including a liquid phase, and in most cases with painted coupons. The

volatile gaseous iodine species are continuously swept towards as selective iodine filter separating

aerosols, inorganic and organic iodides, with on-line measurements by gamma spectrometry.


Repeatability Check: Yes for some tests


Prepared By: B. Clement (IRSN) and A. Bentaib (IRSN)

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4.4.12 E4-12 - THAI Iod-09

Test Facility: THAI




volume of 60 m3. For Test Iod-09 the facility had two sumps, the main sump at the bottom and the

condensate pans, which were filled with water. At the beginning the main sump was stratified, after four

hours it was homogenized by a pump. For the first 25 h, the atmosphere was dry, and then steam was

injected. Iod-09 investigated the mass-transfer of iodine between atmosphere and sumps, along with the

deposition of gaseous iodine on steel surfaces under dry and condensing atmospheric conditions. The dry

surfaces acted as iodine storage.


Funke, Kanzleiter: “Test facility and program to investigate open questions on fission product behaviour in

the containment – THAI phase II”, AREVA NP and Becker Technologies, Germany, August 2006, NTR-

G/2007/de/0233A


Iodine concentrations in

-gas phase 10-7

to 10-5

g/l

-main sump (max) 7x10-5

g/l

-condensate pans (max) 3x10-4

g/l

-wall condensate (max) 10-2

g/l





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4.4.13 E4-13 - THAI Iod-11

Test Facility: THAI




volume of 60 m3. For Test Iod-11 the facility was subdivided into 5 compartments (by placing an inner

cylinder and some horizontal plates/trays). At the beginning the atmosphere was stratified with 80°C near

the top and 25°C near the bottom. The humidity was about 80% and the walls were dry. The iodine was

injected at the top and stayed at elevations at and above 7 m. At 4 h an atmosphere mixing was started,

first by heating the lower vessel walls, then by helium injection and finally by steam injection. Although

the atmosphere was homogenized by these measures, at 25 h the gaseous iodine concentrations in the upper

part of the facility were still a factor of 10 higher than those ones in the lower part. At the end of the test

steam was injected to wash the iodine from the walls in to the sump.




G/2007/de/0233A


Iodine concentrations

gas phase 10-7

to 3x10-5

g/l




Weber: “Specification of the Sarnet-2 WP8 THAI Benchmark, Part 1: Multicompartment Iodine Test Iod-

11”, GRS, July 8th, 2010


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4.4.14 E4-14 - THAI Iod-12

Test Facility: THAI




volume of 60 m3. For Test Iod-12 the facility had the same configuration as for Iod-11, it was subdivided

into 5 compartments (by placing an inner cylinder and some horizontal plates/trays). Also the first phase

of Iod-12 was similar as Iod-11; the atmosphere was stratified with high temperatures at the top and low

temperatures near the bottom; the iodine was injected at the top and stayed at elevations at and above 7 m.

At 3 h steam was injected into the inner cylinder which was closed at its top. After heating its walls the

steam left the inner cylinder through its bottom, starting to mix the atmosphere of the facility. At 7 h the

atmosphere was mixed, however the iodine concentrations near the top stayed a factor of 5 higher than

near the bottom. In parallel to the steam injection the middle and the lower vessel walls were cooled, in

order to achieve wet iodine deposition (in Iod-11 there was only dry deposition). The facility was left over

night without any action; the differences in the iodine concentrations did not change. Then iodine was

resuspended by heating the upper vessel walls. At the end of the test steam was injected to wash the iodine

from the walls in to the sump.




G/2007/de/0233A


General data

1.4 bar

Air, steam, helium atmosphere

Temperature 95C / 30C


gas phase 10-7

to 10-4

g/l



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Weber: “Specification of the Sarnet-2 WP8 THAI Benchmark, Part 2: Multicompartment Iodine Test Iod-

12”, GRS, May 30th, 2010


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4.4.15 E4-15 - THAI Iod-13

Test Facility: THAI



Iod-13 was carried out in the homogeneously mixed THAI vessel (60m³) at about 100°C (fluid and

wall temperatures) and at a relative humidity in the range of 50 to 60%. There were no sumps and the

walls were dry.

After the initial injection of 0.68 g I2 in the upper dome area, the atmospheric I2 fall exponentially

from a theoretical initial concentration of 1.1×10-5

g/l during an initial 3-hour testing phase. After the first

relatively low O3 injection of 0.53 g oxidation of the I2 to iodine oxide (IOx) took place. The finely

dispersed IOx aerosol that was produced had a particle diameter of 0.1 to 0.2 µm. Then a second, high O3

injection (4.3 g) was made oxidizing parts of the remaining I2. The IOx aerosol concentration decreased

slowly after this, mainly by diffusive deposition onto the vessel walls.


F. Funke, G. Langrock, T. Kanzleiter, G. Poß, K. Fischer, G. Langrock,,G. Weber, H.-J. Allelein: Test

facility and program to investigate open questions on fission product behaviour in the containment - THAI

phase II- Part 2 Iodine Tests (August 2006), Becker Technologies, AREVA NP, GRS, (Excerpt of report

released to SARNET2)


General data

1.5 bar

Air, Steam atmosphere

Temperature 100C superheated


gas phase max 1.1×10-5

g/l



Past Code Validation/Benchmarks: Validation work in SARNET2 is ongoing


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4.4.16 E4-16 - THAI Iod-14

Test Facility: THAI



Iod-14 was carried out in the THAI vessel (60m³) without any internal structures under stationary

thermalhydraulic conditions (100°C) with 60 – 65% relative humidity. At the beginning a high ozone

amount was injected, exceeding the stoichiometric ratio to I2 largely. The total of 1.6 g I2 reacted quickly

to an IOx aerosol. The mean particle size of the IOx aerosol grow due to agglomeration and the

concentration depleted slowly mainly by diffusive deposition onto the vessel walls.


F. Funke, G. Langrock, T. Kanzleiter, G. Poß, K. Fischer, G. Langrock,,G. Weber, H.-J. Allelein: Test

facility and program to investigate open questions on fission product behaviour in the containment - THAI

phase II- Part 2 Iodine Tests (August 2006), Becker Technologies, AREVA NP, GRS, (Excerpt of report

released to SARNET2)


General data

1.5 bar

Air, steam atmosphere

Temperature 100C superheated



Past Code Validation/Benchmarks: Validation work in SARNET2 is ongoing


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4.4.17 E4-17 - THAI Iod-25

Test Facility: THAI



For the Iod-25 test, standard internal structures (inner cylinder and four lateral condensate trays) had

been removed to allow free unobstructed environment for aerosol interaction with the molecular iodine in

the gas phase. The interaction of gaseous I2 with a non-reactive aerosol was measured. First about 1g of I2

was injected into the vessel (60 m³), which was at 70°C. Afterwards a non-reactive SnO2 aerosol (max. 2

g/ m³) was injected. A part of the gaseous aerosol was adsorbed by the aerosol particles. This aerosol-

bound iodine settled with the SnO2 particles on the floor of the vessel.


F. Funke, S. Gupta, B. Balewski, OECD-NEA THAI2 Project, Final Report, Deposition of molecular

iodine on aerosol particles Iod-25 (Test using non-reactive aerosol SnO2) Report No. 150 1420–Iod-25-

FR, April 2012


General data

1.3 bar

Air atmosphere

Temperature 70C





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4.4.18 E4-18 - THAI Iod-26

Test Facility: THAI



Same test vessel configuration and test procedure as in Iod-25 (E4-17 - THAI Iod-25), but with an

aerosol that is reactive toward iodine (Ag). The amount of aerosol adsorbed on the Ag particles was much

higher because of chemisorption reactions than on the non-reactive SnO2 particles.


Planned in OECD THAI2 project in September 2012


General data

1.3 bar

Air atmosphere

Temperature 70C

Year Tests Performed: 2012 (planned)




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4.4.19 E4-19 - THAI AW

Test Facility: THAI



In the lower part of the THAI vessel, a stainless steel deposition surface for aerosol was installed that

had a downward gradient of about 2°. Part of the surface could form a puddle, while the other part (plate

section) had no condensate retention. In addition, laboratory tests were performed with the plate sections.

In THAI, first of the soluble CsI aerosol was injected, and after it had become deposited, steam was

injected that went on to condense on the vertical walls and drain over the surfaces with the deposited

aerosols. The puddle acted as an intermediate storage of the aerosol material, and delayed considerably the

wash down transport. The laboratory tests showed that in case of “high” deposition loads and “weak”

water flows, the aerosol material was washed down only in the area of rivulets, while it was washed down

more completely in case of “low” loads and “strong” flows.


Gupta. Langer: “Technical Report (Quick Look Report) Aerosol Wash-down Test (AW), Becker

Technologies GmbH, Germany, December 2009, Report No. 1501326-AW-QLR


Horizontal surface area

THAI plate section: 6 m2

THAI puddle: 1.5 m2

laboratory tests: 0.75 m2

Aerosol load of horizontal surface

THAI-test: 80 g/m2

laboratory test 1: 180 g/m2

laboratory test 2: 91 g/m2

Water flow-rate

THAI plate section: 2.5 g/s

THAI puddle: 1.8 g/s

laboratory test 1: 0.75 g/s

laboratory test 2: 1.5 g/s

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4.4.20 E4-20 - THAI HR31

Test Facility: THAI



These tests are part of the THAI HR series described above. Two tests (HR31.2, HR31.3) were

performed with AREVA PARs to investigate the conversion of CsI aerosol particles to gaseous iodine

while passing through an operating PAR under realistic accident boundary conditions. The conversion

rates were evaluated from the CsI aerosol concentration at the PAR inlet and the gaseous iodine

concentration at the PAR outlet; they were in the range of 1% to 3%.




Gupta et al., “OECD/NEA THAI Project, Quick Look Report, Hydrogen Recombiner Test HR-31, CsI-

PAR interaction Test”, Becker Technologies GmbH, Germany, September 2009, Report No. 1501326-HR-

QLR5


Hydrogen concentration 8 – 9 Vol.%

Steam concentration 60 Vol.%

Catalytic surface temperature 800°C

CsI particle size 0.5 to 0.7 µm


Repeatability Check: HR31.2 and HR31.3



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4.4.21 E4-21 - THAI HR32

Test Facility: THAI



This test is part of the THAI HR series described above. Test HR32 considers an early phase of a

reactor accident where PARs are exposed to saturated steam favouring formation of a condensate layer and

deposition of aerosols on catalyst foils, along with exposing of the foils to gaseous iodine. Tests performed

with AREVA PARs. After preconditioning, hydrogen was injected from 0 to 1000 s. From 1000 to 2600 s

hydrogen, aerosols, and iodine were injected. The onset of recombination was at1500 s at the same

hydrogen inlet conditions as for comparable HR tests. The recombination efficiency remained in the range

of 50 to 60%, which is comparable to other tests of the HR series, with the same thermalhydraulic

conditions, but without aerosols and iodine. This has shown that even under these challenging conditions,

the effect of aerosol and iodine on the PAR performance is not significant.





Hydrogen concentrations

o onset of recombination 3.7 Vol.%

o maximum 4.8 Vol.%

o end of test close to zero

Steam concentration 40 Vol.%

Aerosol concentrations 1.5 – 2.5g/m3

SnO2 (insoluble)and LiO3 (soluble)

Gaseous iodine 0.5 10-3

g/m3





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4.4.22 E4-22 - LASS-GIRS DABASCO

Test Facility: LASS-GIRS

Owner Organization: CIEMAT


The objective of the experiments was to quantify the effect of pH, temperature and droplets size on

the removal efficiency of inorganic gaseous iodine by spray systems. The experiments were performed in

the GIRS facility. A prismatic transparent vessel of 0.5x0.5 m base, containing I2 at specific temperature,

pressure and humidity conditions, was sprayed with water through nozzles. By sampling the vessel

atmosphere and the water collected at the bottom of the vessel, the iodine concentration evolution was

recorded. Water and gas were at the same temperature, and relative humidity was set between 50 and 90%,

depending on the experiment.


Herrero B., Artigao, A., Álvarez, M.T. and Jimenez J.L., “Final report: GIRS project” DFN technical note

DFN/SN-02/IF-00 (March 2000)


Spray temperature:

o Low temp.: 15 – 30°C

o High temp.: 80 -100°C

Droplet diameter: 0.25 and 0.8 mm

pH: 4 , 7 and 9

Water flowrate:

o Large nozzle tests: 2.5×10-3

– 3.0×10-3

L/s

o Small nozzle tests: 0.8×10-3

– 1.6×10-3

L/s





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4.4.23 E4-23 - OECD-THAI2 Gaseous Iodine Release from Flashing Jet Test

Test Facility: THAI



The OECD-NEA THAI-2 Project includes plans to perform experiments on volatile iodine formation

during flashing of water. A high pressure vessel (about 1 m3 volume) will be used as the break source,

containing water at high pressure (~40 bars) and temperature (~250°C, saturation conditions). Its water

will contain dissolved molecular iodine. A radioactive iodine tracer may be added in order to support

iodine concentration measurements. The high pressure vessel will be connected to the THAI facility (as

shown in the figure below), which will be at about 1 bar, 100°C superheated air-steam environment. The

walls of the THAI vessel are heated to minimize iodine deposition and steam condensation.

Figure 4.4.23-1 THAI Facility for Gaseous Iodine Release from Flashing Jet Test

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The jet from the pipe end is directed to a droplet separator to remove the droplets (simulate the jet

droplet entrainment on internal structures). The gaseous part of the flashing jet is released to the rest of the

THAI vessel atmosphere. The iodine concentrations in the atmosphere are measured by a gas scrubber

technique. The thermalhydraulic state of the THAI vessel is measured by standard instrumentation

(thermocouples and pressure transducers).


Not available yet


Break Source:

o Saturated water

o Pressure: 40 bars

o Temperature: 250oC (saturation)

Containment (THAI) atmosphere

o superheated steam-air

o Pressure: 1 bar

o Temperature: 100oC

Break Discharge Flowrate: ~2.5 kg/s

Year Tests Performed: ~2012



Experiment will be used by the OECD THAI-2 participants in a benchmark exercise.


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4.4.24 E4-24 - CAIMAN 97/02 test

Test Facility: CAIMAN



The CAIMAN Facility studied iodine behaviour in containment. The 97/02 test is representative of

selected severe accidents conditions with pH=5, 90°C in the sump, dose rate of 1 kGy/h and presence of

painted surfaces in both the gas and aqueous phases.

Figure 4.4.24-1 Layout of the CAIMAN Test Facility

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J. Ball , C. Marchand , H. Allelein , L. Cantrel , R. Cripps, G. Glowa, L. Herranz, A. Rincon, J. Royen, A.

Rydl, P. Schindler, G. Weber and J. Wren, 2006. International Standard Problem ISP-41, Follow-up

exercise (Phase II): Iodine Code Comparison Exercise against CAIMAN and RTF Experiments Results

from PHEBUS RTF1 Test: Final Report. CSNI/R(2004)16


Dose Rate: 1 kGy/h in aqueous phase

Temperature: 90-110C

pH: 5 (starting)

Surfaces: steel and painted coupons in both gas and aqueous phases

Starting concentration of iodide: 10-5

mol/dm3







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4.4.25 E4-25 - CAIMAN 2001/01 Test

Test Facility: CAIMAN



The CAIMAN Facility studied iodine behaviour in containment. The 2001/01 test is representative of

selected severe accidents conditions and is suitable to validate organic iodine formation sub-models.

See E4-24 - CAIMAN 97/02 test for drawing of the test facility.


J. Ball , C. Marchand , H. Allelein , L. Cantrel , R. Cripps, G. Glowa, L. Herranz, A. Rincon, J. Royen, A.

Rydl, P. Schindler, G. Weber and J. Wren, 2006. International Standard Problem ISP-41, Follow-up

exercise (Phase II): Iodine Code Comparison Exercise against CAIMAN and RTF Experiments Results

from PHEBUS RTF1 Test: Final Report. CSNI/R(2004)16


Dose Rates: 3.2 kGy/h in aqueous phase and 2.0 kGy/h in gas phase

Temperatures: 80-110C

pH: 5.2 (starting)

Surfaces: steel and epoxy polyamide paint in aqueous and gas phases

Starting concentration of iodide: 4×10-5

mol/dm3







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4.4.26 E4-26 – Iodine Clean-Up in a Steam Suppression System

Test Facility: N/A

Owner Organization: United Kingdom Atomic Energy Research Establishment, Harwell, England.

Experiment Description: (taken from Diffey et al. (1965))

These experiments studied the removal of fission products by a suppression pool. Simulated fission

products carried in an air-steam flow at 34 m/s have been fed through lutes (a lute is a liquid-sealed tube)

of 3 mm and 50 mm diameter into ponds of cold water under steady state conditions. The depth of

immersion was 20 and 10 lute diameters, respectively. Measurements have been made of the

decontamination factors and of the proportion of the trapped activity released when air is subsequently

bubbled through the water.

Elemental iodine, methyl iodide and hydrogen iodide, labelled with radioactive Iodine 131 or Iodine

132, were used in the experiments to simulate the behaviour of gaseous forms of iodine likely to be present

in a release of fission products from nuclear fuel. To simulate the behaviour of iodine attached to

particulate material, an aerosol with a mean size of about 0.06 microns was used. This aerosol was

labelled with ThB (212

Pb).


H.R. Diffey et al., “Iodine Clean-up in a Steam Suppression System”, International Symposium on Fission

Product Release and Transport Under Accident Conditions, CONF-650407, Vol. 2, Oak Ridge National

Laboratory, 1965


Small lute:

o 90% dry saturated steam

o Steam flow: about 10 g/min

o Vapour velocity in the flute: 34 m/s

o Pool temperature: about 28C

Large lute:

o Steam flow: up to 4.5 kg/min

o Water flow: up to 9 kg/min

o Air flow: up to 1.4 m3/min

o Vapour velocity in the flute: 34 m/s

o Pool temperature: 50C

Year Tests Performed: ~1960s


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4.5 Core Melt Distribution and Behaviour in Containment Experiments

4.5.1 E5-1 - IET Experiments - Zion Geometry

Test Facility: Surtsey Vessel



The IET tests were designed to investigate the phenomena associated with Direct Containment

Heating (DCH). Figure 4.5.1-1 shows the Surtsey vessel, the high-pressure melt ejection system, and the

subcompartment structures used in the IET experiments. The figure also shows instrumentation location

by channel number. In these tests, high-temperature, chemically reactive melt was ejected from a melt

generator by high-pressure steam into a scale model of a reactor cavity, geometrically typical of the Zion

and Surry nuclear power plants. Debris was entrained by the steam blowdown into a large test vessel

simulating a reactor containment building.

High-pressure ejection of molten core material into the containment atmosphere would lead to direct

containment heating by the release of thermal and chemical energy of the debris. Direct containment

heating is applicable mainly to large dry PWR containments. BWRs have automatic depressurization

systems, which depressurize the primary system following a loss of emergency cooling and prior to severe

fuel damage.

Table 4.5.1-1

IET Experiments in Zion Geometry

IET Test No. Date Description (geometry, key variations)

1, 1R 9/91, 2/92 Zion, N2 inert atmosphere

3 12/91 Zion, reactive atmosphere

4 3/92 Zion, reactive atmos. + wet basement

5 5/92 Zion, CO2 inerted, reactive atmos. + wet basement

6 7/92 Zion, reactive + pre-existing H2

7 7/92 Zion, reactive + pre-existing H2 + wet basement

8A, 8B 8/92 Zion, cavity half filled with H2O

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Figure 4.5.1-1 Surtsey Vessel


Allen, M.D., Pilch, M.M., Blanchat, T.K., Griffith, R.O. and Nichols, R.T., “Experiments to Investigate

Direct Containment Heating Phenomena with Scaled Models of the Zion Nuclear Power Plant in the

SURTSEY Test Facility”, Sandia National Laboratory, NUREG/CR-6044, SAND93-1049, 1994 May


High pressure (~12 MPa)

Steam containment atmosphere

Iron-alumina thermite


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Repeatability Check: two tests repeated

Past Code Validation/Benchmarks: MELCOR 1.8.5, 2.1 and CONTAIN


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4.5.2 E5-2 - IET Experiments - Surry Geometry

Test Facility: CTTF and Surtsey



The IET tests were designed to investigate the phenomena associated with Direct Containment

Heating (DCH). In these tests, high-temperature, chemically reactive melt was ejected from a melt

generator by high-pressure steam into a scale model of a reactor cavity, geometrically typical of the Zion

and Surry nuclear power plants. Debris was entrained by the steam blowdown into a large test vessel

simulating a reactor containment building.

High-pressure ejection of molten core material into the containment atmosphere would lead to direct

containment heating by the release of thermal and chemical energy of the debris. Direct containment

heating is applicable mainly to large dry PWR containments. BWRs have automatic depressurization

systems, which depressurize the primary system following a loss of emergency cooling and prior to severe

fuel damage.

The CTTF (Containment Technology Test Facility), is operated by Sandia National Labs. Three tests

were conducted in the CCTF and 1 test in Surtsey.

Table 4.5.2-1

IET Experiments in Surry Geometry

IET Date Description (geometry, key variations)

9 11/92 Surry (CTTF), w/- annular gap, air/steam/H2 atmos.

10 11/92 Surry (Surtsey), w/-annular gap, air/steam/H2 atmos.

11 12/92 Surry (CTTF), w/o-annular gap, air/steam/H2 atmos.

12 12/92 Surry(Surtsey), w/o-annular gap, air/steam/H2 atmos.

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Figure 4.5.2-1 Models of Surry Structures in the Containment Technology Test Facility

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Figure 4.5.2-2 Side-View of the Experiment Setup used in the IET/CTTF Tests (IET-9, 10 and 11)

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Figure 4.5.2-3 Model of the Surry Bottom Head Used in the IED Experiments

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Figure 4.5.2-4 Models of Surry Structures in the Surtsey Test Facility

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Figure 4.5.2-5 Side-View of the Experiment Setup used in the IET/Surtsey Test (IET-12)


Blanchat, T.K., Allen, M.D., Pilch, M.M. and Nichols, R.T., “Quick-Look Report on the Tenth Integral

Effects Test (IET-10) in the Containment Technology Test Facility”, Sandia National Laboratories, 1993

August

Blanchat, T.K., et. al., 1994 “Experiments to Investigate Direct Containment Heating Phenomena with

Scaled Models of the Surry Nuclear Power Plant,” SAND93-2519, NUREG/CR-6152

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CCTF (1/5.75 linear scale)

RPV hole size – ablated from 7.0cm to (7.0 cm-9.8 cm) [in CCTF]

RPV steam pressure at plug failure: 12.1 MPa to 13.2 MPa [in CCTF]

Containment initial gas composition: inert (steam/CO2) to reactive

Wet basement (condensate) none to 700 kg

Initial melt simulant: 158 kg

Initial Containment Pressure: 0.14 MPa to 0.2 MPa

Initial Containment Temperature: 400 K

Containment pre-existing hydrogen:

2 mol%

Surtsey (1/10 linear scale)

RPV hole size – ablated from 5.6 cm to (5.6 cm)

RPV steam pressure at plug failure: 11.2 MPa

Containment initial gas composition: inert/steam

Wet basement (condensate) none

Initial melt simulant: 30kg

Initial Containment Pressure: 0.16MPa

Initial Containment Temperature: 407K

Containment pre-existing hydrogen: 6 mol%



Past Code Validation/Benchmarks: MELCOR codes and CONTAIN


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4.5.3 E5-3 - FARO Tests

Test Facility: FARO

Owner Organization: Joint Research Centre of the European Commission, Ispra, Italy (JRC-Ispra)


The JRC-Ispra FARO (Fuel Melt and Release Oven) plant is a large multi-purpose test facility in

which reactor severe accidents could be simulated by out-of-pile experiments. A quantity in the order of

up to 200 kg of oxide fuel type melts (up to 3000°C) could be produced in the FARO furnace, possibly

mixed with metallic components, and delivered to a test section containing a water pool at an initial

pressure up to 5.0 MPa. The reference scenario of the current test series is relevant to a postulated in-

vessel core melt down accident when jets of molten corium penetrate into the lower plenum water pool,

fragment and settle on the lower head. The release vessel is designed for a pressure of 10 MPa at 570 K

combined with a debris catcher on the lower part of the vessel. FARO test L-14 (it became ISP 39) run

with 125 kg of 80% UO2 and 20% of ZrO2 released by gravity in 2 m deep pool of saturated water at 5.1

MPa.

The objectives were to investigate basic phenomenologies relevant to the progression of severe

accidents in water cooled reactors with particular emphasis on the interaction of molten fuel with coolant

and/or structures under both in-vessel and ex-vessel postulated severe accident conditions.


Information website: http://stresa.jrc.ec.europa.eu/sarnet/DataBase/index.asp

L-06: Discharge of 18 kg of UO2/ZrO2 in saturated water at 50 bar

D. Magallon - 'FARO LWR Programme Scoping Test L-06 Data Report' - I.92.135 - 1/1/1992

H.U. Wider, A. Benuzzi, H. Hohmann, D. Magallon, A. Yerkess - 'The FARO/LWR Experimental

Programme Quick Look Report on the Scoping Test L-06' - I.92.139 - 1/1/1992


Magallon - 'FARO LWR Programme Quenching Test-2 Data Report' - I.93.154 - 1/1/1993

H.U. Wider - 'Test Analysis Report on FARO Test L-08' - I.96.40 - 1/1/1996

L-11: Discharge of 151 kg of UO2/ZrO2/Zr in saturated water at 50 bar

M. Dehn, D. Magallon - 'FARO LWR Programme Base Case Test L-11 Data Report' - I.94.147 -

1/1/1994

A. Benuzzi, D. Magallon - 'FARO LWR Programme Base Case Test L-11 Quick Look Report' -

I.94.55 - 1/1/1994

L-14: Discharge of 125 kg of UO2/ZrO2 in saturated water at 50 bar (OECD/CSNI ISP-39)

D. Magallon, G. Leva - 'FARO LWR Programme Test L-14 Data Report' - I.96.25 - 1/1/1996

A. Benuzzi, D. Magallon - 'FARO Test L-14 Quick Look Report' - I.94.171 - 1/1/1994

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A. Annunziato, C. Addabbo, A. Yerkess, R. Silverii, W. Brewka, G. Leva - 'OECD/CSNI

International Standard Problem 39 on FARO Test L-14 on Fuel Coolant Interaction and Quenching -

Comparison Report, Volume I: Analysis of the Results' - NEA/CSNI/R(97)31 - 2/4/2000

A. Annunziato, C. Addabbo, A. Yerkess, R. Silverii, W. Brewka, G. Leva - 'OECD/CSNI

International Standard Problem 39 on FARO Test L-14 on Fuel Coolant Interaction and Quenching -

Comparison Report, Volume II: Participants Appendices' - NEA/CSNI/R(97)31 - 2/4/2000


R. Silverii, D. Magallon - 'FARO LWR Programme Test L-19 Data Report' - - 9/15/2000

D. Magallon, A. Annunziato - 'FARO Test L-19 Quick Look Report' - I.96.27 - 1/1/1996


R. Silverii, D. Magallon - 'FARO LWR Programme Test L-20 Data Report' - I.00.94 - 1/1/1994

A. Annunziato, C. Addabbo, D. Magallon - 'FARO Test L-20 Quick Look Report' - I.96.163 -

1/1/1996



C. Addabbo, A. Annunziato, D. Magallon - 'FARO Test L-24 Quick Look Report' - I.97.185 -

1/1/1997




1/1/1998



A. Romor, A. Annunziato, D. Magallon - 'Addendum to QLR on FARO Test L-28' - - 5/22/2001


1/1/1999

L-29: Discharge of 39 kg of UO2/ZrO2 in subcooled water at 2 bar


'FARO Test L-29 Quick Look Test'




1/1/1999


R. Silverii, D. Magallon - 'FARO LWR Programme Test L-33 Data Report' - TN I.00.124 -

1/1/2000

A. Romor, A. Annunziato, D. Magallon - 'Addendum to QLR on FARO Test L-33' - - 5/22/2001

A. Annunziato, C. Addabbo, D. Magallon - 'Test L-33 Quick Look Report' - I.00.111 - 1/1/2000

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QUENCHING

o 5-50 bar

o 18-175 kg of UO2 melt at T > 3000°C

SPREADING

o Dry and 1 cm water core catcher


Repeatability Check: Yes, 12 available experiments


FARO Test L-14 is ISP-39 (8 different codes or code versions participated in the validation: COMETA,

IFCI, IVA, JASMINE, MC3d, TEXAS, THIRMAL, VAPEX)

A. Annunziato, C. Addabbo, A. Yerkess, R. Silverii, W. Brewka, G. Leva, “OECD/CSNI International

Standard Problem 39 on FARO Test L-14 on Fuel Coolant Interaction and Quenching”,

NEA/CSNI/R(97)31, 1996

A. Annunziato, A. Yerkess and C. Addabbo, “FARO and KROTOS code simulation and analysis at JRC

Ispra”, Nuclear Engineering and Design, Vol. 189, No. 1-3, 1999.

Prepared By: M. Sangiorgi (ENEA)

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4.5.4 E5-4 - DISCO-C Tests

Test Facility: DISCO-C



In a LWR, a failure of the reactor pressure vessel, followed by melt expulsion and blowdown of the

reactor cooling system might disperse molten core debris out of the reactor pit. The mechanisms of

efficient debris-to-gas heat transfer, exothermic metal/oxygen reactions and hydrogen combustion may

cause a rapid increase in the pressure and temperature in the reactor containment. The test facility DISCO-

C (DIspersion of Simulant COrium – Cold) models the annular reactor cavity and the sub-compartments of

a large European reactor in a scale 1:18. The fluid dynamics of the dispersion process was studied using

model fluids, water, gallium-indium-tin or bismuth alloy instead of corium, and nitrogen or helium instead

of steam. The objective of the tests was to study the effect of different breach sizes and locations and

different failure pressures on the dispersion, specifically by testing central holes, lateral holes, horizontal

rips, and complete ripping of the bottom head.


C. Caroli, Proposition of supplementary DCH tests to be carried out using the DISCO facility of FZK and

representative of the French reactor pit geometry, Note Technique IRSN DPEA/SEAC/2002-040.

L. Meyer, D. Plassart, DCH test campaign in the modified DISCO-C facility using water as corium

simulant, Note Technique IRSN DSR/SAGR/2005-59.

L. Meyer, D. Plassart, DCH test campaign in the modified DISCO-C facility using a gallium-indium-tin

alloy as corium simulant, Note Technique IRSN DSR/SAGR/NT/2006-107.

L. Meyer, M. Gargallo et al., Low Pressure Corium Dispersion Experiments in the DISCO-C Tests Facility

with Cold Simulant Fluids, Report FZKA 6591, Forschungszentrum Karlsruhe, 2006.

Low pressure corium dispersion experiments in the DISCO test facility with cold simulant fluids / L.

Meyer, Karlsruhe: Forschungszentrum Karlsruhe, 2006. Report-Nr.: FZKA 6591.


Breach diameter: 0.011 - 0.1 m

Position and shape of the breach: angle of 45, horizontal slot, and unzipping and tilting of the

lower head

Initial pressure in the RPV: 0.25 - 2 MPa

Fluids used: water, woods-metal, gallium-indium-tin alloy

RPV/Pit geometries: Konvoi, French P’4


Repeatability Check: Yes, about 40 available experiments

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WILHELM, D, “Transient Code Models for Low Pressure Corium Dispersion”, OECD Workshop on Ex-

Vessel Debris Coolability, Karlsruhe, Germany, 15-18 November 1999.

PLASSART D., RUPUICUV calculations for DISCO-L1, Contribution of IRSN to the Code Benchmark

with SARNET, SARNET-CONT-P08, 2005.

BRETAULT A., MAAP calculations for DISCO-L1, Contribution of EDF to the Code Benchmark with

SARNET, SARNET-CONT-P06, 2005.

SPENGLER C., CONTAIN calculations for DISCO-L1, Contribution of GRS to the Code Benchmark with

SARNET, SARNET-CONT-P07, 2005.

MIKASSER S., MEIGNEN R., Computation and analysis of the Direct Containment Heating dispersion

process with the multiphase flow software MC3D, proceedings of ICAPP-2007, Nice, France, May 13-18,

2007.

Prepared By: G. Albrecht (KIT)

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4.5.5 E5-5 - DISCO-H Tests

Test Facility: DISCO-H



The test facility DISCO-H (DIspersion of Simulant COrium – Hot) was set up to perform scaled

experiments that simulate core melt ejection scenarios under low system pressure in severe accidents in

Pressurized Water Reactors (PWR). The main components of the facility are scaled about 1:18 linearly to

large European PWR. The experiments are designed to investigate the fluid-dynamic, thermal, and

chemical processes during melt ejection out of a breach in the lower head of a PWR pressure vessel at

pressures below 2 MPa with an iron-alumina melt and steam. Also some experiments with water in the

reactor pit were performed to investigate the ex-vessel fuel coolant interaction and debris formation.


Melt dispersion and direct containment heating (DCH) experiments in the DISCO-H test facility / L.

Meyer. Forschungszentrum Karlsruhe, 2004. Report-Nr.: FZKA 6988.

L. Meyer, G. Albrecht, M. Kirstahler, M. Schwall, E. Wachter, G. Wörner, Melt Dispersion and Direct

Containment Heating (DCH) experiments in the DISCO-H test facility”, Report FZKA 6988,

Forschungszentrum Karlsruhe, 2004.

L. Meyer, A. Kotchourko, “Separate Effects Tests on Hydrogen Combustion during Direct Containment

Heating Events in European Reactors”, Proceedings of SMiRT-19, Toronto, Canada, 2007 August 12-17.

L. Meyer, G. Albrecht, M. Kirstahler, M. Schwall, E. Wachter, “Separate Effects Tests on Hydrogen

Combustion during Direct Containment Heating Events”, Forschungszentrum Karlsruhe, Report FZKA

7379, 2008.

L. Meyer, G. Albrecht, C. Caroli, I. Ivanov, “Direct containment heating integral effects tests in geometries

of European nuclear power plants”, Nuclear Engineering and Design, Vol. 239, 2009, pp. 2070-2084.

Melt dispersion and direct containment heating (DCH) experiments for KONVOI reactors by L. Meyer.

Karlsruhe: KIT Scientific Publishing, 2011. ISBN: 978-3-86644-579-6, Report-Nr.: KIT-SR 7567.


Breach diameter: 0.028 - 0.06 m

H2 concentrations: 0 - 8%

Containment atmosphere: air (1 bar abs.), air/steam (2 bar abs.)

Initial pressure in the RPV: 0.5 - 2.2 MPa

Melts used: iron-alumina

Amount of melt: 10.6 - 16 kg

RPV/Pit geometries: Konvoi, French P’4, VVER1000

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Year Tests Performed: 1998 – present


A large number of tests with different geometries of the cavity (EPR, KONVOI, French P’4, VVER-1000)

and parameters were done in the DISCO-H facility.


WILHELM, D., “Chemical Reaction Models in a Code of the SIMMER-Family”, Joint IAEA/NEA

Technical Meeting on the Use of Computational Fluid Dynamic Codes for Safety Analysis of Reactor

Systems, Including Containment, Pisa, Italy, 11-13 November, 2002

D. WILHELM, “Recalculation of Corium Dispersion Experiments at Low System Pressure,” NURETH-

10, Seoul, Korea, October 5-9 (2003)

R. MEIGNEN, D. PLASSART, C. CAROLI, L. MEYER, D. WILHELM, “Direct Containment Heating at

Low Primary Pressure: Experimental Investigation and Multi-dimensional Modeling”, NURETH-11, paper

164, Avignon, France, (2005).

D.WILHELM, The Challenge of Simplifying DCH Modeling, The first European Review Meeting on

Severe Accident Research (ERMSAR-2005), Aix-en-Provence, France, 14-16 November 2005.

L. MEYER, A. KOTCHOURKO, “Separate Effects Tests on Hydrogen Combustion during Direct

Containment Heating Events in European Reactors”, SMiRT 19, Toronto, Canada, August 12-17, (2007)

R. MEIGNEN, S. MIKASSER, C. SPENGLER, A. BRETAULT, Synthesis of Analytical Activities for

Direct Containment Heating, The second European Review Meeting on Severe Accident Research

(ERMSAR-2007), Forschungszentrum Karlsruhe GmbH (FZK), Germany, 12-14 June 2007.

SPENGLER, C.; MEYER, L.; MEIGNEN, R., Investigations of direct containment heating (DCH) in

European reactors: database of integral tests and progress in modeling. 3rd European Review Meeting on

Severe Accident Research (ERMSAR-2008), Nesseber, BG, September 23-25, 2008.


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4.5.6 E5-6 - DISCO-A2

Test Facility: DISCO-A2



Hydrogen combustion tests were performed in two different size facilities, DISCO-H (linear scale

1:18) and DISCO-A2 (linear scale 1:7) to reproduce hydrogen effects during a severe accident with high

pressure melt ejection and direct containment heating. The hydrogen was blown out of a pressure vessel

into a constrained compartment, modeling the reactor pressure vessel and the reactor pit, respectively, and

from there into a large vessel, modeling the containment. A number of distributed igniters simulated hot

melt particles. Tests with and without steam and with concentrations of pre-existing hydrogen in the

containment atmosphere between 0 and 8% were conducted.


L. Meyer, G. Albrecht, C. Caroli, I. Ivanov, “Direct containment heating integral effects tests in geometries

of European nuclear power plants”, Nuclear Engineering and Design, Vol. 239, 2009, pp. 2070-2084.

L. Meyer, A. Kotchourko, “Separate Effects Tests on Hydrogen Combustion during Direct Containment

Heating Events in European Reactors”, Proceedings of SMiRT-19, Toronto, Canada, 2007 August 12-17.

L. Meyer, G. Albrecht, M. Kirstahler, M. Schwall, E. Wachter, “Separate Effects Tests on Hydrogen

Combustion during Direct Containment Heating Events”, Forschungszentrum Karlsruhe, Report FZKA

7379, 2008.

L. Meyer, G. Albrecht, M. Kirstahler, M. Schwall, E. Wachter, “Large Scale Separate Effects Tests on

Hydrogen Combustion during Direct Containment Heating Events”, Karlsruhe Institute of Technology,

KIT-Report to be published.

L. Meyer, G. Albrecht, “Experimental Study of Hydrogen Combustion during DCH Events in two different

Scales”. Proceedings of NURETH 14, Toronto, Canada, September 25-30, 2011.

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Small facility

- Containment volume: 14 m3

- Breach diameter: 0.025 m

- Initial pressure in the RPV: 1.9-2.7 MPa

- Steam concentration (Containment): 0-38.6 mol%

- Initial H2 in Containment: 0-64 mol (Concentration: 0-7 mol%)

- RPV blow down H2: 24-53 mol

Large facility

- Containment volume: 227 m3

- Breach diameter: 0.0625 m

- Initial pressure in the RPV: 0 and 1.8 MPa

- Steam concentration (Containment): 0-48.3 mol%

- Initial H2 in Containment: 0-1198 mol (Concentration: 0-7.8 mol%)

- RPV blow down H2: 327-671 mol


Repeatability Check: Yes, 8 (small scale) respectively 9 (large scale) available experiments


R. Meignen, D. Plassart, C. Caroli, L. Meyer, D. Wilhelm, “Direct Containment Heating at Low Primary

Pressure: Experimental Investigation and Multi-dimensional Modeling”, NURETH-11, Avignon, France,

2005.


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4.5.7 E5-7 - KROTOS JRC Tests

Test Facility: KROTOS JRC

Owner Organization:

Institute for Systems, Informatics and Safety (ISIS) of the Joint Research Centre of the European

Commission, Ispra site, Italy (JRC-Ispra) with participation from US-NRC.


The KROTOS test facility is a relatively small scale experimental installation dedicated to the study

of: (a) molten fuel–coolant pre-mixing either with prototypic reactor melts or simulants such as alumina up

to 5 kg; (b) progression and energetics of spontaneous and triggered fuel–coolant interactions (vapor

explosions).

The main components of the facility are: the radiation furnace, the test section and pressure vessel.

The furnace, maximum electric power 130 kW, consists of a cylindrical tungsten heater, which encloses

the tungsten or molybdenum crucible containing the melt material. Depending on the crucible design and

material, melt masses from 1 to 10 kg can be heated up to 3270 K. The furnace is covered with a bell-

shaped, water-cooled lid designed to withstand a 0.25 MPa over-pressure of cover gas (Ar, He) or vacuum.

Having reached the desired stable melt temperature the crucible is released from the furnace and falls by

gravity through a 4 m long release tube. Half-way down this tube, a rapid-acting slide valve closes

immediately after the crucible has passed in order to isolate and protect the furnace from any energetic

events experienced in the test section below.

During its fall, the crucible breaks a copper wire that generates the zero time signal for the data

acquisition system. Finally, the crucible impacts onto a retainer ring at the bottom of the release tube

where a conical shaped metal puncher breaks the bottom of the crucible and allows the melt to pour into

the test section. The initial diameter of the melt-pour (jet) is defined by guiding the melt through a funnel

of high refractory material with an exit diameter of 30 mm.

The lower part of the KROTOS facility consists of a stainless steel test section bolted to lugs welded

on the inner side walls of a stainless steel pressure vessel. The cylindrical pressure vessel, inner diameter

0.4 m, height, 2.21 m, has a thick flat bottom and a flanged flat upper head and is designed to withstand a

static pressure of 2.5 MPa at 493 K. The cylindrical test section, inner diameter 200 mm, outer diameter

240 mm, closed at the bottom by either a flat plate or with a gas trigger device, can contain water up to a

height of about 1.27 m (about 40 L).

The gas trigger, chamber volume 15 cm3, can be charged to a pressure of up to 20 MPa (Ar) and is

closed by a 0.1–0.25 mm thick steel membrane. After melt penetration down to the lower region of the test

section, the mechanical destruction of the membrane causes a pressure pulse to propagate vertically

upwards through the water column that contains varying concentrations of melt and steam. The gas trigger

device is activated either by a specific thermocouple signal or by a backup time delay circuit.

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Figure 4.5.7-1 Schematic of the KROTOS JRC Facility

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Information website: http://stresa.jrc.ec.europa.eu/sarnet/DataBase/index.asp

K1-TT4 11/19/1991 Trigger test in pure water, Test section 10 cm

'Digital Data 0 - 3 ms' - - 2/3/2000

K-26 6/5/1991 Melt Al2O3, Mass 1.4kg - Temp. 2573 K Pressure 1 bar, sub-cooling 40C

H. Hohmann, M. Field, K. Klein, H. Schins, A. Yerkess - 'KROTOS 26 to KROTOS 30: Experimental

Data Collection' - I.92.115 - 2/9/2000

KROTOS digital data - 'Base time range 0.0 0.05 s' - - 3/25/1997




Digital Data - 'Digital data (digitized from EDR)' - - 1/17/2000











Huhtiniemi, H. Hohmann, R. Faraoni, M. Field, G. Gambaretti, K. Klein - 'KROTOS 38 to KROTOS

44: Data Report' - I.96.37 - 1/25/2000

'Digital Data 0 - 30 ms' - - 6/7/1999

'Digital Data 0 - 5.2 ms' - - 6/13/1996

K2-TT6 1/26/1998 Trigger test in pure water, Test section with plastic windows

I. Huhtiniemi, D. Magallon, M. Field, R. Gambaretti, K. Klein - 'KROTOS Data Report: KROTOS 58 (

KT-2 )' - I.98.186 - 1/19/2000

'Digital Data -1 - 10 ms' - - 2/3/2000

K2-TT7-1/26/1998 Trigger test in pure water, Test section with metal plates


KT-2 )' - I.98.186 - 1/19/2000

Digital Data -1 - 10 ms - - 2/3/2000

K2-TT8 8/11/1999 Trigger test in pure water, 20 cm test section with plastic charge

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I. Huhtiniemi, A. Romor, R. Gambaretti, G. Nicol - 'KROTOS KT-3 Data Report (K-63)' - I.99.198 -

1/28/2000

'Digital Data 0 - 3 ms' - - 2/4/2000





K-32 9/1/1993 Melt UO2/ZrO2, Mass 3kg - Temp. 3063 K Pressure 1 bar, sub-cooling 22C

I. Huhtiniemi, H. Hohmann, R. Faraoni, M. Field, G. Gambaretti, K. Klein - 'KROTOS 32 to KROTOS

36 Data Report' - I.95.128 - 1/26/2000

KROTOS digital data - 'Base time range -0.7 4.3 s' - - 8/20/1996

KROTOS digital data - 'Range time base -5.2 45. S' - - 8/20/1996

K-33 10/21/1993 Melt UO2/ZrO2, Mass 3.2kg - Temp. 3063 K Pressure 1 bar, sub-cooling 75C


36 Data Report' - I.95.128 - 1/26/2000





36 Data Report' - I.95.128 - 1/26/2000



K-36 6/2/1994 Melt UO2/ZrO2, Mass 3kg - Temp. 3025 K Pressure 1 bar, sub-cooling 79C


36 Data Report' - I.95.128 - 1/26/2000





I. Huhtiniemi, H. Hohmann, D. Magallon, M. Field, R. Gambaretti, K. Klein - 'KROTOS Corium Tests

37, 45-48, 52-54' - I.97.177 - 2/7/2000



44: Data Report' - I.96.37 - 1/25/2000



44: Data Report' - I.96.37 - 1/25/2000


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44: Data Report' - I.96.37 - 1/25/2000



44: Data Report' - I.96.37 - 1/25/2000

K-43 7/7/1995 Melt Al2O3, Mass 1.5kg - Temp. 2625 K Pressure 2.1 bar, sub-cooling 100C


44: Data Report' - I.96.37 - 1/25/2000



44: Data Report' - I.96.37 - 1/25/2000



37, 45-48, 52-54' - I.97.177 - 2/7/2000



37, 45-48, 52-54' - I.97.177 - 2/7/2000



37, 45-48, 52-54' - I.97.177 - 2/7/2000

K-48 6/11/1996 Melt UO2/ZrO2, Mass 5.kg - melting test


37, 45-48, 52-54' - I.97.177 - 2/7/2000

K-49 7/12/1996 Melt Al2O3, Mass 1.5kg - Temp. 2688 K Pressure 3.7 bar, sub-cooling 120C

I. Huhtiniemi, D. Magallon, M. Field, R. Gambaretti, K. Klein - 'KROTOS Alumina Tests: K-49, K-

50, K-51, K-57' - I.00.21 - 3/10/2011



50, K-51, K-57' - I.00.21 - 3/10/2011



50, K-51, K-57' - I.00.21 - 3/10/2011



37, 45-48, 52-54' - I.97.177 - 2/7/2000

K-53 3/25/1997 Melt UO2/ZrO2, Mass 3.6kg - Temp. 3129 K Pressure 3.6 bar, sub-cooling 122C


37, 45-48, 52-54' - I.97.177 - 2/7/2000

K-54 5/15/1997 Melt UO2/ZrO2, Mass 5.5kg - melting test

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37, 45-48, 52-54' - I.97.177 - 2/7/2000


I. Huhtiniemi, D. Magallon, M. Field, R. Gambaretti, K. Klein - 'KROTOS Data Report: Test KT-1

(KROTOS 56)' - I.98.97 - 1/19/2000

'EKTAPRO camera of the jet in the water. Film rotated 90 degrees

I. Huhtiniemi - 'Film of the test section at the entrance in the water (7 Mbytes)' - - 11/14/1997

I. Huhtiniemi - 'Film of the test section from two viewing angles at the same axial position in the water

(2 Mbytes)' - - 11/14/1997

I. Huhtiniemi - 'Film with NAC camera of the jet entrance into the water (5 Mbytes)' - - 2/17/2000



50, K-51, K-57' - I.00.21 - 3/10/2011

I. Huhtiniemi - 'EKTAPRO camera of the jet in the water. The left side is rotate 180 degrees (6.3

Mbytes)' - - 2/17/2000

I. Huhtiniemi - 'Film from Canon camera located xx m below/above the water (0.4 Mbytes)' - -

12/17/1997

I. Huhtiniemi - 'SONY camera in the water (2 Mbytes)' - - 2/17/2000



KT-2 )' - I.98.186 - 1/19/2000

K-63 7/27/1999 Melt UO2/ZrO2, Mass 4.5kg - Temp. n.a., Pressure 2.1 bar, sub-cooling 99C

I. Huhtiniemi, A. Romor, R. Gambaretti, G. Nicol - 'KROTOS KT-3 Data Report (K-63)' - I.99.198 -

1/28/2000

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MELT

o Al2O3 or UO2/ZrO2 1.4-5.4 kg

o 2473-3077 K

WATER

o 1-3.7 bar

o 12-125°C


Repeatability Check: Yes (many repeats were attempted)


A. Annunziato, A. Yerkess and C. Addabbo, “FARO and KROTOS code simulation and analysis at JRC

Ispra”, Nuclear Engineering and Design, Vol. 189, No. 1-3, 1999.

Prepared By: M. Sangiorgi (ENEA)

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4.5.8 E5-8 - SERENA-2 KROTOS and TROI Commissioning Tests

Test Facility: KROTOS (CEA) and TROI (KAERI)

Owner Organization: CEA and KAERI


The objective of the experimental programme SERENA-2 is threefold:

1. Provide experimental data to clarify the explosion behaviour of prototypic corium melts,

2. Provide innovative experimental data for validation of explosion models for prototypic materials,

including spatial distribution of fuel and void during the premixing and at the time of explosion,

and explosion dynamics,

3. Provide experimental data for the steam explosion in more reactor-like situations to verify the

geometrical extrapolation capabilities of the codes.

The KROTOS facility features rather one-dimensional behaviour of mixing and explosion

propagation. This allows a clear characterisation of mixing behaviour (melt and void distribution) and

escalation and propagation behaviour (given path starting from bottom triggering), with the respective

possibilities of direct checking of code results. Six complementary tests in each facility are planned.

The effect of the fuel material properties will be investigated with the use of 4 different compositions.

The basic oxidic corium will be 70%UO2-30%ZrO2, as it revealed to induce spontaneous explosions more

energetic than with 80%UO2-20%ZrO2 in TROI conditions. Tests will be performed with standard ex-

vessel conditions, i.e., a pressure of 0.2 MPa and a subcooling of 50 K.

KROTOS tests have been performed with the following common conditions:

Corium melt mass ~5 kg

Pool depth ~1.1 m

Pool diameter 200 mm

Free fall 50 cm

New release mechanism (produce circular jet) and X-Ray radioscopy were added (to the KROTOS

JRC facility)

A specific device has been used and developed to characterize the fragmentation/premixing phase of

the corium in the water (Linatron: high X-ray energy). This device is able to distinguish the water, the

void and the corium. Despite the wide pool of knowledge gained during the last three decades of research

and development on fuel-coolant interaction, the simulation of the premixing still remains to be a

challenging task, whereas the conditions of the premixing impose directly the conditions of the explosion.

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KROTOS Facility

Figure 4.5.8-1 Schematic of the KROTOS CEA Facility

The gap in the understanding of the premixing mechanisms is coming from the experimental

limitations. Having no direct way to observe the spatial distribution of phases, experimentalists and code

developers are limited to indirect methods such as tracing the signals of sacrificial thermocouples, global

void fraction measurements, void front detection, fast video recording of the jet propagation within the

coolant and post test debris analysis.

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Following these analyses, Experts involved in SERENA Phase-1 have identified the following

remaining uncertainties:

void fraction and its distribution in the interaction zone

the effect of corium properties on initial conditions of premixing breakup

limiting effect of the crust formation on the corium fragments

fine fragmentation of melt drops by pressure waves.

The absence up to now of the fine characterization of the premixing phases, i.e., corium, so-called

“void”, and water, is one key parameter to answer to the steam explosion fine quantification, needed for

reactor case calculation.

In the frame of OECD/SERENA-Phase 2 project, it has been planned to use a unique tool, a high

energy X-Ray beam, to characterize the premixing phases with prototypical corium on KROTOS test

facility at CEA-Cadarache. Contrary to the traditionally applied methods of video recording, X-Ray

radioscopy is not affected by vapor formation or melts radiation and distinguishes phases according to their

X-Ray attenuation properties and density.

TROI tests have been performed with the following common conditions

Corium melt mass ~20 kg

Pool depth 0.7 – 1.3m

Pool diameter 600 mm

Free fall 50 cm – 100 cm with an intermediate catcher.

O

392 878

248

200

368

14°

1

2 3

4

5

6

7

8

9

1 - X-Ray source; 2 - Lead collimator; 3 - Test section: Fortal (Al -90.1%, Zn – 5.6%); 4 - Scinscillator:

Ta (0.5mm) Gadox (1.4 mm); 5 – Mirror; 6 - Opaque box; 7 - Lead screen; 8 - High sensitivity CCD

Camera; Camera acquisition/control block

Figure 4.5.8-2 X-Ray Radioscopy for the KROTOS CEA Facility

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Figure 4.5.8-3 Schematic of the TROI Facility

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In the frame of the SERENA-2 programme, the same conditions for the two facilities have been taken

for the corium melt composition and water of the test section.


P. Piluso, S.W. Hong, “OECD SERENA: A Fuel Coolant Interaction Programme (FCI) devoted to reactor

case”, ISAMM-2009 Paul Scherrer Institute Villigen, Switzerland, October 26-28, 2009.


materials effect (thermodynamic and thermophysical properties)

composition of corium (solidification interval, sub-stoichiometric)

overmelting (+150°C)




Prepared By: P. Piluso (CEA)

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4.5.9 E5-9: SERENA-2 KROTOS and TROI Tests

Test Facility: KROTOS (CEA) and TROI (KAERI)

Owner Organization: CEA and KAERI


The KROTOS and TROI tests used the configuration defined in the commissioning tests (E5-8 -

SERENA-2 KROTOS and TROI Commissioning Tests). Each test was duplicated in the KROTOS and

TROI facility, to provide information on the effect of scaling on this phenomenon. There were 5 tests with

the conditions shown in the test matrix below.

Table 4.5.9-1

Test Matrix for SERENA-2 KROTOS and TROI Experiments

Test P (MPa) T (K) Material Trigger (s)

KROTOS KS-1 /

TROI TS-1

0.4 301 70% UO2 – 30% ZrO2 ~0.85

KROTOS KS-2 /

TROI TS-2

0.2 334 70% UO2 – 30% ZrO2 ~0.85

KROTOS KS-3 /

TROI TS-3

0.2 331 70% UO2 – 30% ZrO2 ~0.85

KROTOS KS-4 /

TROI TS-4

0.2 333 80% UO2 – 20% ZrO2 ~1.05

KROTOS KS-5 /

TROI TS-5

0.2 337 70% UO2 – 15% ZrO2 + 15% Zr ~1.05

Other specifics about the five tests are as follows:

KROTOS KS-1/TROI TS-1: The conditions have been selected because of their bounding trends

for ex-vessel situation, expect for the temperature that has been limited by heater capacity but

nevertheless at a realistic 160 K of overheating. For the first time, the visualization of the corium

jet structure when it enters in water and starts its fragmentation has been possible with specific

high energy X-ray imaging system. The pictures give also information of steam distribution

around corium jet or fragments.

KROTOS KS-2/TROI TS-2: Comparing to KS-1 test conditions the following has been modified

in the present test: (1) the tin membrane location (below the release cone), (2) the position of the

X-ray radioscopy system (695 mm level), (3) the installation of the video record system for the

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visualization of the melt release conditions, (4) the initial mass of the corium (4.9 kg), and (5) the

overheating of the corium melt (about +180°C for a calculated liquidus temperature of 2560°C).

KROTOS KS-3/TROI TS-3: For test KS-3, the explosion has not been triggered in this test due

to the unsuccessful crucible release, as well as no coherent corium jet had been obtained.

KROTOS KS-4/TROI TS-4: 3.2 kg of corium at 2690°C has been released into 40.5 kg water

heated at 60°C. The triggered steam explosion provided maximum pressure pick of 447 Bar.

KROTOS KS-5/TROI TS-5: 2.4 kg of prototypic corium (80.1wt.% UO2 – 11.4wt.% ZrO2 –

8.5wt.% Zr) at 2587°C have been released into 34.5 kg of water at 53°C. The mixture has been

successfully triggered. Nevertheless, no global steam explosion occurred.

References for Experiment Description: Not provided


materials effect (thermodynamic and thermophysical properties)

composition of corium (solidification interval, sub-stoichiometric)

overmelting (+150°C)



Past Code Validation/Benchmarks: Performed with MC3D

Prepared By: P. Piluso (CEA)

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4.5.10 E5-10 - MCCI-1 Tests CCI Tests 1-3; SSWICS tests 1-7

Test Facility: MCCI

Owner Organization: ANL


This program consisted of both separate effect and integral reactor material experiments that

investigated multi-dimensional core-concrete interaction under dry cavity conditions, as well as the extent

that core debris interacting with concrete can be cooled by top water injection. The integral effect CCI

tests investigated core-concrete interaction and debris coolability in 2-D notch-type concrete cavity

configurations. Decay heat was simulated at prototypic levels using direct electrical heating. Melt

temperatures, concrete erosion depth, non-condensable gases from concrete erosion, and core debris

cooling rate were measured as core-concrete interaction progressed. Data analysis focused on evaluating

the extent that core debris could be rendered permanently coolable by melt eruption and water ingression

cooling mechanisms. Tests parameterized on concrete type. In all experiments, the cavity was flooded

after significant concrete erosion (> 20 cm) had occurred.

The SSWICS tests investigated the extent that water is able to ingress into solidifying core material to

augment what would otherwise be a conduction-limited cooling process. These were transient quench

tests; thus, the water ingression rate was evaluated by comparing the actual debris cooling rate to the

conduction-limited solution. Tests parameterized on concrete type, the amount of concrete present in the

melt, as well as the system pressure during quench.


M. T. Farmer, S. Lomperski, D. Kilsdonk, R. W. Aeschlimann, and S. Basu, “A Summary of Findings

from the Melt Coolability and Concrete Interaction (MCCI) Program,” Paper 7544, Proceedings ICAPP

’07, Nice, France, May 13-18, 2007.

M. T. Farmer, D. J. Kilsdonk, and R. W. Aeschlimann, “Corium Coolability under Ex-Vessel Accident

Conditions for LWRs,” Nuclear Eng. Technology, Vol. 41, pp. 575-602, June 2009.

S. Lomperski and M. T. Farmer, “Corium Crust Strength Measurements,” Nuclear Eng. Design, Vol. 239,

pp. 2551-2561, March 2009.

S. Lomperski and M. T. Farmer, “Experimental Evaluation of the Water Ingression Mechanism for Corium

Cooling,” Nuclear Eng. Design, Vol. 237, pp. 905-917, August 2006.

S. Lomperski, M. T. Farmer, and S. Basu, “Experimental Investigation of Corium Quenching at Elevated

Pressure,” Nuclear Eng. Design, Vol. 236, pp. 2271-2280, 2006.

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CCI Tests:

Initial melt mass: 400 kg

Initial Melt temperature: 2000-2100C

Decay Heat Level: 300 W/kg Fuel

Melt Compositions: BWR (UO2, ZrO2, and concrete oxides)

Cladding oxidation: 100%

Cavity: 2-D rectilinear: 50 cm x 50 cm initial basemat size

Concrete types: limestone/common sand and siliceous

SSWICS Tests:

Initial melt mass: 68-75 kg


Transient cooling (no fuel heating)

Melt Compositions: BWR with 8 to 23 wt.% concrete


Cavity: 1-D circular test section 30 cm ID


Test section Pressure: 1 or 4 bar




CCI-2 test was conducted as a blind pretest/postest code validation exercise among program participants

Prepared By: R. Lee (NRC) and M. Farmer (ANL)

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4.5.11 E5-11 - MCCI-2 Tests CCI Tests 4-6; SSWICS tests 8-13; WCB-1

Test Facility: MCCI



This program was a continuation of the OECD/MCCI-1 and consisted of both separate effect and

integral reactor material experiments that investigated multi-dimensional core-concrete interaction under

dry cavity conditions, as well as the extent that core debris interacting with concrete can be cooled by top

water injection. The program scope was also expanded to examine effectiveness of engineered core

catcher systems. The integral effect CCI (core-concrete interaction) tests investigated core-concrete

interaction and debris coolability in 2-D notch-type concrete cavity configurations. Decay heat was

simulated at prototypic levels using direct electrical heating. Melt temperatures, concrete erosion depth,

non-condensable gases from concrete erosion, and core debris cooling rate were measured as core-concrete

interaction progressed. Data analysis focused on evaluating the extent that core debris could be rendered

permanently coolable by melt eruption and water ingression cooling mechanisms. Tests parameterized on

concrete type. In two tests (CCI-4 and CCI-5), the cavity was flooded after significant concrete erosion (>

20 cm) had occurred, while in CCI-6 the cavity was flooded within a few minutes of melt contact with the

basemat.

The additional SSWICS tests expanded the parameter space for evaluating the extent that water is able

to ingress into solidifying core material to augment what would otherwise be a conduction-limited cooling

process. These were transient quench tests; thus, the water ingression rate was evaluated by comparing the

actual debris cooling rate to the conduction-limited solution. The additional tests looked at crust strength

issues, as well as the effect of non-condensable gas injection on the debris cooling rate.

In terms of core catcher tests, two separate effect tests were conducted in the SSWICS apparatus to

look at transient cooling and quench of core debris by bottom water injection; one test through concrete

nozzles cast in the basemat, while the second was with stainless steel nozzles that could also inject

concurrent noncondensable gas. In addition, a large scale test was conducted with a water cooled stainless

steel basemat plate covered with concrete to investigate transient erosion and core melt stabilization on a

water cooled surface.


S. Lomperski and M. T. Farmer, “Performance Testing of Engineered Corium Cooling Systems,” Nuclear

Eng. Design, Vol. 243, pp. 311-320, January 2012.

M. T. Farmer, S. Lomperski, D. J. Kilsdonk, and R. W. Aeschlimann, “OECD MCCI-2 Project Final

Report,” OECD/MCCI-2010-TR07, November 2010

S. Lomperski and M. T. Farmer, “Corium Crust Strength Measurements,” Nuclear Eng. Design, Vol. 239,

pp. 2551-2561, March 2009.

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CCI Tests:

Initial melt mass: 400 -900 kg



Melt Compositions: BWR (UO2, ZrO2, Zr, stainless, and concrete oxides)

Cladding oxidation: 70-100%

Cavity: 2-D rectilinear: 50 cm x 50 cm and 70 cm x 70 cm initial basemat sizes


SSWICS Tests:




Melt Compositions: BWR with 8 to 23 wt.% concrete




Test section Pressure: 1 or 4 bar

Core Catcher Tests: SSWICS Bottom Water Injection




Melt Compositions: BWR with 23 wt.% silicoue concrete



Bottom water injection through concrete nozzles, and stainless steel nozzles with concurrent

noncondensable gas injection

Core Catcher Test WCB-1


Initial Melt temperature: 2000C


Melt Compositions: BWR (UO2, ZrO2, and concrete oxides)

Cavity: 1-D rectilinear: 50 cm x 50 cm, water-cooled steel plate with 15 cm overlying siliceous

concrete



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CORQUENCH Code was further developed and released as part of this program

M.T. Farmer, “Modeling of Ex-Vessel Corium Coolability with the CORQUENCH Code,” Proc. 9th Int.

Conf. On Nucl. Eng., ICONE-9696, April 2001. [Note: CORQUENCH is a stand-alone model]


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4.5.12 E5-12 - ECO Tests

Test Facility: ECO

Owner Organization: FZK


The Energy Conversion (ECO) tests were conducted in Forschungszentrum Karlsruhe (FzK),

Germany to measure steam explosion pressures and the energy conversion ratio with Al2O3 melt masses

more than 10 kg. Nine tests were conducted in total, although other tests have shown that U-Zr-O melts

have lower tendency to steam explosion than the Al2O3 melts used in these tests.

The objective of the ECO tests was to obtain an experimental upper limit for the pressure and energy

conversion ratio and their dependence on initial and boundary conditions. The second objective of the tests

was to provide data for the validation of models.


Cherdron, W., Huber, F., Kaiser, A., and Schütz, W., “ECO Steam Explosion Experiments –

Documentation and Evaluation of Experimental Data”, Forschungszentrum Karlsruhe GmbH, FZKA 7011,

2005.


Release mass of Al2O3: 0.9 to 18 kg

Average Release Rate: 2.4 to 60 kg/s

Height of fall to water surface: 0.275 to 0.325 m

Pool water temperature: 293 to 369C (some with 30 to 40C temperature stratification)

Initial system pressure: 0.1 to 0.25 MPa


Repeatability Check: ECO-06 was a repeat of ECO-05



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4.5.13 E5-13 - BALI Ex-Vessel Tests

Test Facility: BALI

Owner Organization: CEA (with EDF funding)


The BALI tests investigated thermalhydraulic phenomena in corium pools for in-vessel and ex-vessel

conditions. The tests were performed at CEA Grenoble with Frammatome and EDF funding. The tests

were designed in 1993 to generate a database on heat transfer distribution at boundaries of corium pools for

in-vessel and ex-vessel geometric configurations. A rectangular slice test section was used to study the ex-

vessel thermalhydraulic phenomena.

The objective of the tests was to develop a good understanding of the heat flux distribution at corium

pool boundaries and to quantify heat transfer coefficients at pool boundaries as a result of natural

convection. The BALI tests used simulant materials and used dimensionless parameters to categorize the

various flow regimes so that results can be applied to reactor conditions


JM Bonnet, JM Seiler, Thermalhydraulic phenomena in corium pools, The BALI experiment, ICONE-7,

Tokyo, 1999.


Prandtl: 4-1000

Height: 0.4-0.5m

Thickness: 0.1 m

Length: 2.4-2.9 m

superficial gas velocity: 1-20 cm/s

Viscosity: 1-350 mPa-s

Internal Rayleigh number: 108-10

11

heat transfer coefficient: 300-5000 W/m2-K




Prepared By: C. Journeau (CEA)

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4.5.14 E5-14 - BALISE Tests

Test Facility: BALISE

Owner Organization: CEA (with EDF funding)


The BALISE tests investigated the mixing of immiscible liquids by a sparging gas. The tests were

performed at CEA Grenoble with EDF funding. One of the issues in severe accident progression in a PWR

is molten corium relocation into the reactor pit and corium concrete interaction. The concrete ablation

velocity depends on layout of the metallic and oxide layers of the corium pool and the gas evolution during

the corium-concrete interaction. From a phenomena perspective, the mixing or stratification phenomena of

immiscible liquids by a sparging gas was examined in these tests.


Tourniaire, B., Seiler, J-M., Bonnet J-M., “Study of the mixing of immiscible liquids by sparging gas;

results of the BALISE experiments”, Proceeding of NURETH-10, Seoul, Korea, Oct. 5-9, 2003


ratio of layer heights: 0.5-10

density difference: 5-70%

viscosities: 0.9-100 mPa-s

surface tensions: 10-70 mN/m

Void fraction: 0-15%

superficial gas velocities: 0-4 cm/s

Stratified/emulsioned configurations





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4.5.15 E5-15 - VULCANO VB-U7 (EPR concrete)

Test Facility: VULCANO

Owner Organization: CEA (performed with EURATOM funding)


The VULCANO facility at CEA Cadarache is devoted to the study of ex-vessel corium behaviour

with 30-60 kg of prototypic (depleted UO2 containing) corium. In the MCCI configuration, a hemi

cylindrical concrete cavity is filled by 30-50 kg of corium and decay heat is simulated by induction. In the

VB-U7 test, a feroosiliceous concrete (representing EPR reactor pit sacrificial concrete) has been ablated.

Ablation was largely anisotropic.


C. Journeau, L. Ferry, P. Piluso, J. Monerris, M. Breton, G. Fritz, T. Sevon, Two EU-funded tests in

VULCANO to assess the effects of concrete nature on its ablation by molten corium, 4th European Review

Meeting on Severe Accident Research (ERMSAR-2010), Bologna-Italy, 11-12 May 2010.


Inputs:

initial mass: 30-60 kg

temperature: 2000-2400 K

decay heat power: ~10-20 kW

Outputs:

concrete temperature: 20-1400C

ablation front position: 0-20 cm

pool temperature: 1500-2400C

surface temperature: 1000-2500C

For the COMET test:

water inlet flow (~75 g/s)

steam out flowrate (0-50 g/s)


Repeatability Check: Partial



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4.5.16 E5-16 - VULCANO VW-U1 (COMET bottom flooding)


Owner Organization: CEA (performed with EURATOM funding)



with 30-60 kg of prototypic (depleted UO2 containing) corium. In the MCCI configuration, a hemi

cylindrical concrete cavity is filled by 30-50 kg of corium and decay heat is simulated by induction. In the

VW-U1 test, a unit cell of the COMET core catcher concept (porous concrete tubes enabling bottom water

injection after melting through a thin leak tight concrete layer) has been tested with prototypical oxidic

corium and decay heat simulation.


C. Journeau, H. Alsmeyer, Validation of the COMET Bottom-Flooding Core-Catcher with Prototypic

corium, Int. Congr. Advances nuclear Power plants (ICAPP06), Reno, NV, June 4-6, 2006.


Inputs:

initial mass: 30-60 kg

temperature: 2000-2400 K

decay heat power: ~10-20 kW

Outputs:


ablation front position: 0-20 cm

pool temperature: 1500-2400C


For the COMET test,

water inlet flow (~75 g/s)

steam out flowrate (0-50 g/s)



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WABE code (University of Stuttgart/Institute for Nuclear and Energy techniques – IKE )

W. Widmann, M. Bürger, G. Lohnert, H. Alsmeyer, W. Tromm, Experimental and theoretical

investigations on the COMET concept for ex-vessel core retention, Nucl Eng. Des., 236, 2304-2327, 2006.

M. Bürger, B.R. Sehgal, Debris Formation and Coolability, In: Nuclezar Safety in Light Water Reactors,

Severe Accident Phenomenology, B.R. Sehgal, ed., Academic press, Waltham, MA, 2012.


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4.5.17 E5-17 - VULCANO VE-U7




VULCANO spreading tests


with 30-60 kg of prototypic (depleted UO2 containing) corium. In the spreading configuration, corium is

spread over a flat surface at low pouring rates (to favour immobilization processes). In the VE-U7 test,

two parallel spreading sections were considered, one in concrete, the other with an inert ceramic.


C. Journeau, J.F. Haquet, B. Spindler, C. Spengler, J. Foit, “The Vulcano VEU7 Corium Spreading

Benchmark”, Progr. Nucl. Ener., 48, 215-234 (2006).


Inputs:

corium mass: 40 kg

temperature: 2450 80 K

decay heat power: 0

Outputs:


spreading front position: 0-50 cm

corium temperature: 1500-2400C



Repeatability Check: Partial

Past Code Validation/Benchmarks: RFLOW, LAVA, THEMA


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4.5.18 E5-18 – SURC-1 and SURC-2

Test Facility: SURC

Owner Organization: SNL


The major components of the experimental apparatus for these tests include a sealed, water-cooled

containment vessel; interaction crucible; and induction coil. A containment vessel was used in the tests to

ensure that nearly all of the reaction products would pass through the instrumented exit flow piping. The

interaction crucible used in the experiments is of cylindrical geometry and is shown in the following figure.

The crucible consists of three major components: the lower crucible, upper crucible, and cover. The

annulus of the upper and lower crucible and the cover are cast using a magnesium oxide (MgO) castable

refractory material. The overall dimensions of the crucible are 60 cm diameter 100.0 cm high with a

40cm-diameter cavity 60.0 cm deep. Cast into the bottom of the lower crucible was an instrumented

limestone concrete cylinder 40.0 cm diameter 40.0 cm thick for SURC-1 experiment and a basaltic

concrete cylinder 40.0 cm diameter 40.0 cm thick for SURC-2 experiment. The two crucible sections

were assembled and sealed with Saureisen Cement No. 31

The SURC-1 test was run at local atmospheric pressure (0.83 atm) and at an ambient temperature of

25°C. Power to the apparatus was gradually increased to a level of 99 kW (gross) at 48 minutes, to a level

of 154 kW at 126 minutes, and to a level of 200 kW at 222 minutes. The charge became molten at times

after 120 minutes and concrete attack began at 135 minutes when the zirconia insulator board at the bottom

of the charge was dissolved into the melt. Aerosol samples were taken at 153 minutes as the optical

pyrometer started to indicate large amounts of aerosol production. Gas composition grab samples were

also taken at this time.

The SURC-2 experiment was a molten material/concrete interaction test designed to sustain a melt of

203.9 kg of depleted uranium oxide, zirconium metal and zirconium oxide in an MgO crucible with a

basaltic concrete bottom. The goals of the experiment were to measure in detail the gas evolution, aerosol

generation, and erosion characteristics associated with molten oxide-concrete interactions. The charge

material in SURC-2 was a mixture of 69 wt.% UO2 – 22 w/o ZrO2 – 9 wt.% Zr. Additionally, 3.4 kg of

fission product stimulants were added to the melt to study fission product release. The SURC-2 test was

run at local atmospheric pressure (0.83 atm) and at an ambient temperature of 25°C. The SURC-2 test ran

for a total of 280 minutes. A total of 35 cm of basaltic concrete was eroded during the final 150 minutes of

the experiment. This was increased to a net power of 84 kW for the final portion (between 220–280

minutes) of the test.

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Figure 4.5.18-1 Schematic of SURC Test Facility

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E. R. Copus, R. E. Blose, J. E. Brockmann, R. B. Simpson, and D. A. Lucero, Core-Concrete Interactions

Using Molten Urania With Zirconium on a Limestone Concrete Basemat, The SURC-1 Experiment,

NUREG/CR-5443, SAND90-0087. Sandia National Laboratories, Albuquerque, NM, September 1992.

E. R. Copus, R. E. Blose, J. E. Brockmann, R. B. Simpson, and D. A. Lucero. Core-Concrete Interactions

Using Molten UO2 With Zirconium on a Basaltic Basemat, The SURC-2 Experiment, NUREG/CR-5564,

SAND90-1022. Sandia National Laboratories, Albuquerque, NM, August 1992


SURC1 SURC2

Pressure (atm.) 0.83 0.83

Temperature (°C) 25 25

Power supply (kW) 50→154 50→150

Concrete use limestone basaltic




Part of suite of experiments used for MELCOR code assessment


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4.5.19 E5-19 - SURC-3

Test Facility: SURC

Owner Organization: NRC

Experiment Description (from abstract in SAND86-2638):

Four inductively sustained experiments, QT-D, QT-E, SURC-3, and SURC-3A, were performed in

order to investigate the additional effects of zirconium metal oxidation on core debris-concrete interactions

using molten stainless steel as the core debris simulant. The SURC-3 experiment had a 45 kg charge of

stainless steel to which 1.1 kg of zirconium was subsequently added. SURC-3 axially eroded 33 cm of

limestone concrete during two hours of interaction. The experiment showed in a large increase in erosion

rate, gas production, and aerosol release following the addition of Zr metal to the melt. In the SURC-3A

test the measured erosion rates increased from 14 cm/hr to 27 cm/hr, gas release increased from 50 to 100

slpm (standard litres per minute), and aerosol release increased from 0.02 g/sec to 0.04 g/sec. The effluent

gas was composed of 80% CO, 10% CO2, and 2% H2 before Zr addition and 92% CO, 4% CO2, 4% H2

during the Zr interactions which lasted 10-20 minutes. Additional measurements indicated that the melt

pool temperature ranged from 1600C-1800C and that the aerosols produced were comprised primarily of

Te and Fe oxides.


E.R. Copus et al., “Experimental Results of Core-Concrete Interactions Using Molten Steel with

Zirconium”, NUREG/CR-4794; SAND86-2638, July 1990


45 kg stainless steel melt prepared in contact with limestone concrete. 1.1 kg of metallic zirconium

added once concrete ablation had begun.



Past Code Validation/Benchmarks: Used to validate model predictions of the effects of metallic

zirconium on interactions


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4.5.20 E5-20 - SURC-3A

Test Facility: SURC



Four inductively sustained experiments, QT-D, QT-E, SURC-3, and SURC-3A, were performed in

order to investigate the additional effects of zirconium metal oxidation on core debris-concrete interactions

using molten stainless steel as the core debris simulant. The experiment, SURC-3A, eroded 25 cm of

limestone concrete axially and 9 cm radially during 90 minutes of sustained interaction. It utilized 40 kg of

stainless steel and 2.2 kg of added zirconium as the charge material. The experiment showed in a large

increase in erosion rate, gas production, and aerosol release following the addition of Zr metal to the melt.

In the SURC-3A test the measured erosion rates increased from 14 cm/hr to 27 cm/hr, gas release

increased from 50 to 100 slpm (standard litres per minute), and aerosol release increased from 0.02 g/sec to

0.04 g/sec. The effluent gas was composed of 80% CO, 10% CO2, and 2% H2 before Zr addition and 92%

CO, 4% CO2, 4% H2 during the Zr interactions which lasted 10-20 minutes. Additional measurements

indicated that the melt pool temperature ranged from 1600C-1800C and that the aerosols produced were

comprised primarily of Te and Fe oxides.


E.R. Copus et al., “Experimental Results of Core-Concrete Interactions Using Molten Steel with

Zirconium”, NUREG/CR-4794; SAND86-2638, July 1990


50 kg of melt interacted with limestone concrete. 5 kg of metallic zirconium was added in the

course of the interaction



Past Code Validation/Benchmarks: Used to validate model predictions of the effects of metallic

zirconium on interactions


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4.5.21 E5-21 - SURC-4

Test Facility: SURC



The SURC (Sustained Urania-Concrete) test series was designed to extend the existing database for

core-concrete interactions and associated aerosol source term production. The SURC experiments are

intended to provide information needed to validate three aspects of core-concrete interactions models,

including heat transfer mechanisms, gas release chemistry, and vaporization release of aerosols. SURC 1,

2, 5 and 6 are integral tests involving UO2-ZrO2, and SURC 3, 4, 7 and 8 are separate effect tests using

stainless steel.

The SURC-4 experiment was conducted at the Sandia National Laboratories in a 600 mm interaction

crucible constructed with a 400 mm diameter basaltic concrete cylinder in the base of a magnesium oxide

(MgO) annulus. A 10 mm thick, circular cover of MgO was placed on the top of the crucible. The SURC-

4 experiment used 200 kg of stainless steel, 20 kg Zr metal, 6 kg fission product stimulants, and basaltic

concrete. Duration of SURC-4 test was 162 minutes with specified periods of heating between 98 kW and

245 kW. Nothing mentioned about accuracy in the final comparison report of the ISP-24. However,

reliability and accuracy of experimental data were confirmed during post-test analyses workshop.


E.R. Copus et al., “Core-Concrete Interactions Using Molten Steel with Zirconium on a Basaltic Basemat:

The SURC-4 Experiment”, NUREG/CR-4994; SAND87-2008, April 1989

M; Lee, R.A. Bari, International Standard problem ISP-24 – SURC-4 Experiment on Core-Concrete

Interactions

Final Comparison Report – CSNI Report No. 155, Vol. 1, September 1988

Final Workshop Summary Report – CSNI Report No. 155, Vol. 2, December 1989


200 kg of stainless steel melt interacted with a one dimensional basaltic concrete plug in a

magnesia melt facility. During the interaction 20 kg of zirconium was added to the melt.

Atmospheric pressure

Heating by induction coils up to 280 kW

AMMD 2.1 to 2.7 μm

Mass concentration: 60 to 638 mg/cm3

Number concentration: 48,000 to 137,000 cm-3

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Repeatability Check: Yes (several tests have been performed with the same aerosol in different

conditions)


SURC-4 experiment was the basis for the CSNI ISP-24 where 8 organisations from 7 countries participated

with 3 Thermal-hydraulic codes (CORCON/2.02 and 2.04, WECHSL, DECOMP-DOE) and 2 aerosol

codes (VANESA, METOXA-DOE).

Prepared By: A. Amri (OECD) and R. Lee (NRC)

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4.5.22 E5-22 - BETA V5.1

Test Facility: BETA


Experiment Description: (taken from NEA/CSNI/R(92)9)

This experiment studied the ex-vessel phase of a core-melt accident, mainly in the high temperature

phase of the concrete (siliceous concrete crucible) erosion with high zirconium content of the melt. The

experiment measured the temperature of the melt, oxidation behaviour of the zirconium, erosion of the

concrete, release and composition of the gases and release of aerosols.

The melt is kept in a cylindrical, axisymmetric concrete crucible fabricated from siliceous concrete. The

internal cavity of the crucible is partly filled by the melt with a metallic melt at the bottom and an oxidic

melt on top. During the experiment the metallic part of the melt is heated electrically by the induction coil

enclosing the concrete crucible. Most of the melt is produced by a thermite reaction and poured into the

crucible under controlled conditions at the start of the experiment. In this experiment, the addition of

metallic Zircaloy to the melt was accomplished by dropping 80 kg of solid Zircaly rubble at room

temperature into the lower crucible and pouring the thermite melt onto the Zircaloy.


ISP 30 – BETA V5.1 Experiment on Melt-Concrete Interaction – Comparison Report, OECD/NEA

NEA/CSNI/R(92)9

Alsmeyer, H., and Firnhaber, M., “Specification of the International Standard Problem ISP 30: BETA

V5.1 Experiment on Melt Concrete Interaction, September 1990.


Siliceous concrete

Thermite melt had 300 kg of metal and 50 kg of oxide with a temperature of 2170 K

The thermite melt was poured onto 80 kg of Zircalloy-4

Time average downward ablation rate through the concrete crucible is 1.0 m/h




ISP 30 – BETA V5.1 Experiment on Melt-Concrete Interaction – Comparison Report, OECD/NEA

NEA/CSNI/R(92)9

Prepared By: A. Kotchourko

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4.5.23 E5-23 - ACE Phase C Tests L1, L2, L4, L5, L6, and L7

Test Facility: ACE/MCCI



This program consisted of a series of six large scale reactor material experiments that investigated

concrete erosion with concurrent fission product release under dry cavity conditions. Decay heat was

simulated at prototypic levels using direct electrical heating. Melt temperatures, concrete erosion depth,

non-condensable gases from concrete erosion, and fission product release rates were measured as core-

concrete interaction progressed. Peak cavity erosion depths ranged up to 15 cm.


J. K. Fink, D. H. Thompson, D. R. Armstrong, B. W. Spencer, and B. R. Sehgal, “Aerosol and Melt

Chemistry in the ACE Molten Core-Concrete Interaction Experiments,” High Temperature and Materials

Science, 33, pp. 51, 1995

B. R. Sehgal, B. W. Spencer, D. H. Thompson, J. K. Fink and M. T. Farmer, “ACE Project Phases C&D:

ACE/MCCI and MACE Tests”, Int. Topical Meeting on the Safety of Thermal Reactors, Portland, OR,

1991 July 21-25



Initial Melt temperature 1650-2200C

Decay Heat Level: 350-400 W/kg Fuel

Melt Compositions: BWR and PWR melt (UO2, ZrO2, Zr, and concrete oxides)


Cavity: 1-d rectilinear; 50 cm x 50 cm cross-section

Concrete types: limestone/common sand, limestone-limestone, siliceous, and serpentine





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4.5.24 E5-24 - MACE Tests M0, M1b, M3b, M4, and MSET-1

Test Facility: MACE



The CCI test facility consisted of a test apparatus, a power supply for direct electrical heating of the

corium, a water supply system, steam condensation (quench) tanks, a ventilation system to complete

filtration and exhaust the off-gases, and a data acquisition system. A schematic illustration of the overall

setup is provided in the figure below. The apparatus consisted of three rectilinear sidewall sections and a

lid. The overall structure was 3.4 m tall. The two upper sidewall sections had a square internal cross-

sectional area of 50 cm x 50 cm. The internal dimensions of the lower test section, where the core-

concrete interaction took place, was varied from 30 cm x 30 cm to 120 cm x 120 cm; resultant core melt

masses varied from 100 kg to 2000 kg. The sidewalls of the cavity also varied; in MO, all four sidewalls

were concrete (resulting in 3-D cavity erosion behavior), whereas for all other tests (M1b, M3b, and M4)

the sidewalls were made from refractory MgO (resulting in 1-D cavity erosion).

Operationally, the tests were carried out as follows. The first step was to produce the core melt in-situ

over the concrete basemat. For early tests MO and M1b, this was carried out by direct electrical heating of

the core debris, resulting in several hours of preheat time. In the subsequent tests, an exothermic chemical

reaction was used in which the melt was produced in a timeframe of 30-60 seconds (the latter approach

significantly increased the reliability of producing the melt, as well as minimizing preheating of the test

cavity through heat losses during the extended preheat phase). Once the melt was produced, heating

continued to be applied to simulate decay heat at ~2 hours in the accident sequence (equivalent to ~300

W/kg fuel). Heating was maintained until the test was terminated. Soon after the melt was produced,

cavity erosion would commence. At a predefined time (or ablation depth), the cavity was flooded using

the water supply system and the subsequent core debris cooling behavior was observed. Steam from the

interaction was vented to the quench system which provided data on the debris cooling rate. Water was

periodically added to maintain a 40-60 cm pool depth over the debris. Melt temperature and concrete

ablation rates were measured simultaneously to provide data for code validation. The tests were

terminated on the basis of two criteria:

i) maximum permissible cavity erosion depth reached, or

ii) debris was quenched and core-concrete interaction was terminated.

Following the experiment, the apparatus was disassembled and the material examined to provide

information on morphology (i.e., coherent crust material vs. fragmented debris in the form of porous crust

structure and/or particle beds formed by melt eruptions). Selected samples were analyzed to provide

information on the debris composition and phase structure.

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Figure 4.5.24-1 Schematic of MACE Test


M. T. Farmer, B. W. Spencer, J. L. Binder, and D. J. Hill, “Status and Future Direction of the Melt Attack

and Coolability Experiments (MACE) Program at Argonne National Laboratory,” Proceedings 9th Int.

Conf. on Nucl. Eng., ICONE-9697, April 8-12, 2001

M. T. Farmer, D. J. Kilsdonk, and R. W. Aeschlimann, “Corium Coolability under Ex-Vessel Accident

Conditions for LWRs,” Nuclear Eng. Technology, Vol. 41, pp. 575-602, June 2009




Decay Heat Level: 300-1000 W/kg Fuel

Melt Compositions: BWR and PWR melt (UO2, ZrO2,Zr,stainless, and concrete oxides)


Cavity: 1-d rectilinear: 50 cm x 50 cm to 120 cm x 120 cm cross-section; 3-D rectilinear: initially

30 cm x 30 cm




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M.T. Farmer, “Modeling of Ex-Vessel Corium Coolability with the CORQUENCH Code”, Proc. 9th Int.

Conf. On Nucl. Eng., ICONE-9696, April 2001.


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4.5.25 E5-25 - COLIMA CA-U4

Test Facility: COLIMA



Prototypic aerosols (having the right chemistry, but natural isotopic composition) can be generated

from corium pools in the COLIMA facility. In this test, the aerosol cloud was directed to a representative

crack in a concrete test section. The crack had been realized to mimic containment stresses and the

pressure gradient was representative of reactor case. Impactors upstream and downstream measured the

nature size and concentration of aerosols, providing insight on decontamination.


F. Parrozzi, DJ Caracciolo, C. Journeau, P. Piluso, The COLIMA Experiment on Aerosol retention in

Containment Leak paths under Severe Nuclear Accidents, Nucl. Energy New Europe, Bovec, Slovenia,

2011


Internal pressure: 0.3 MPa

External pressure: 0.1 MPa

Crack dimensions: 300 x 130 x 0.5 mm

Aerosol concentration: ~0.05 – 0.15 g/m3

Aerosol mean mass diameter = 1.1 m




Validation with the code ECART:

S. Morandi, F. Parozzi, E. Salina, C. Journeau, P. Piluso, Aerosol retention in containment leak paths:

indications for a code model in the light of COLIMA experimental results, Eur. Rev. Mtg., Severe Acc.

Res., ERMSAR 2012, Cologne, Germany, March 2012.


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4.5.26 E5-26 - BURN-1

Test Facility: Sandia National Laboratories Core Melt Test Facility



Thermitically generated melt of iron (~1.1 kg) and aluminum oxide (~0.9 kg) was formed in an 8.9

cm diameter, 14.9 cm deep cylindrical cavity within a limestone/common sand concrete crucible. The

interaction was monitored using a Linatron Model 1500 pulsed x-ray source to obtain images (melt

imaging rate of 24/s) of the melt interface with concrete and data on the periodicity of melt contact with

gas-evolving and ablating concrete. The x-ray voltage was 7.5 MeV. The dose rate was about 500

rad/minute.


D.A. Powers and F.E. Arellano, “Direct Observation of Melt Behavior During High Temperature

Melt/Concrete Interactions”, NUREG/CR-2283, SAND81-1754, January 1982


Melt of iron (~1.1 kg) and aluminum oxide (~0.9 kg)



Past Code Validation/Benchmarks: Test results were used to develop the model of melt-concrete heat

transfer used in the CORCON computer code which has been incorporated into the MELCOR accident

analysis computer code.


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4.5.27 E5-27 – SWISS-1 and SWISS-2



Experiment Description (taken from abstract in SAND85-1546):

These tests examined the effects of an overlying water pool on high temperature melt interactions

with concrete. In both tests, a melt of about 46 kg of type 304 stainless steel was formed and deposited

onto a 21.6 cm diameter disk of limestone/common sand concrete. The concrete disk was retained within a

cast MgO annulus. The molten steel was sustained at a power input of 1.3 to 1.7 Watts/gram by induction

heating. In test SWISS-1 a water pool was formed over the melt after about 12 cm of concrete had eroded.

In test SWISS-2, the water pool was formed about one minute after melt contacted the concrete and before

any significant erosion of concrete could take place. In both tests the water pool was kept below the

boiling point. Interactions were sustained for about 40 minutes in the two tests. Concrete erosion rates,

concrete temperatures, heat fluxes to the overlying water pool, gas generation rates, and evolved gas

compositions during tests SWISS-1 and SWISS-2 are reported. Aerosol generation rates are reported for

test SWISS-2.


R.E. Blose et al., SWISS: Sustained Heated Metallic Melt/Concrete Interactions with Overlying Water

Pools, SAND85-1546, 1987


Concrete type: limestone coarse aggregate and silicon dioxide fine aggregate

Melt: 46 kg type 304 stainless steel and ablated concrete

Water addition timing: delayed in test SWISS-1 to accumulate an ablated concrete melt mass;

prompt in test SWISS-2.




D.R. Bradley and J.E. Brockmann, “Analysis of Molten Fuel-Concrete interactions and Fission-Product

Release from Ex-vessel Core Debris”, 13th Water Reactor Safety Research Information Meeting, October,

1985


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4.5.28 E5-28 – HSS-1 and HSS-3




The Hot Solid program is intended to measure, model, and assess the thermal, gas evolution, and

fission product source terms produced as a consequence of hot, solid, core debris-concrete interactions.

Two preliminary experiments, HSS-1 and HSS-3, were performed in order to compare hot solid UO2-

concrete and hot solid steel-concrete interactions. The HSS-1 experiment ablated 6 cm of limestone-

common sand concrete in a little more than three hours using a 9 kg slug of 304 stainless steel at an

average debris temperature of 1350C. The HSS-3 experiment ablated 6.5 cm of limestone-common sand

concrete in four hours using a 10 kg slug of 80% UO2-20% ZrO2 at an average debris temperature of

1650C. Both experiments were inductively heated and contained in a 22 cm alumina sleeve to simulate

one-dimensional axial erosion.


E.R. Copus and D.R. Bradley, “Interaction of Hot Solid Core Debris with Concrete”, NUREG/CR-4558;

SAND85-1739, June 1986


HSS-1: 304 stainless steel slug sustained at 1350C by induction heating in contact with

limestone/common sand concrete

HSS-3: 10 kg of 80% UO2 – 20% ZrO2 sustained by induction heating in contact with

limestone/common sand concrete at 1650C.



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The HOTROX computer code model was evaluated using the results from the HSS tests. HOTROX is a 1-

D concrete ablation model that calculates transient conduction and gas release in the concrete as well as

heatup of the hot solid slug. Using the HSS-1 power input history and geometry, HOTROX calculates 6.2

cm of concrete erosion in 200 minutes. Using the HSS-3 input conditions, HOTROX predicts 6.8 cm of

erosion in 190 minutes. These results compare favorably with the experimental erosion rates. The

calculated thermal response of the concrete is also close to experimentally measured values. The

information from the Hot Solid Program will be used both to expand the post-accident phenomena data

base and to extend the range of applicability of current accident analysis computer models such as

CORCON and CONTAIN.


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4.5.29 E5-29 - TURC1T and TURC1SS




Two large scale molten debris-concrete experiments, TURC1T, a thermite-concrete interaction

experiment, and TURC1SS, a stainless steel-concrete experiment, are reported here. The experiments

consisted of teeming molten debris (>100 kg) onto limestone/common sand concrete. The molten debris

was allowed to cool naturally. The concrete ablation rate, composition of evolved gases, and aerosol data

are presented. The experimental results have been compared to CORCON calculations in order to validate

the code. This comparison showed that, while some parts of the code performed well (chemical

equilibrium model), other sections required further model development (melt-concrete heat transfer

model). An analysis of the two experiments was performed using a new analysis model. The results of the

analysis seem to suggest that the heat transfer mechanism of concrete ablation is similar to nucleate boiling

heat transfer, rather than gas film heat transfer.


J.E. Gronager et al., TURC1: Large Scale Metallic Melt-Concrete Interaction Experiments and Analysis,

NUREG/CR-4420; SAND85-0707, 1986.


TURC1T: 200 kg thermitically generated melt onto a disk of limestone / common sand concrete

TURC1SS: 200 kg type 304 stainless steel at 2350 K teemed into a magnesia crucible with a

limestone common sand concrete disk



Past Code Validation/Benchmarks: Used for CORCON validation


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4.5.30 E5-30 – TURC2 and TURC3




Two large scale UO2/ZrO2/Zr debris-concrete experiments TURC2 and TURC3 are reported here.

The experiments consisted of pouring a large quantity of molten UO2/ZrO2/Zr mixtures onto limestone-

common sand concrete. The molten material was allowed to cool naturally - no internal heating was

present. Data for concrete ablation, gas evolution including composition and flow rate, and aerosol

generation are presented. The experimental results indicate very rapid crusting with no detectable concrete

ablation. Gas reduction of H2O and CO2 to H2 and CO was found to occur even with a purely oxidic

(UO2/ZrO2) melt. Aerosol concentrations varied from 62 g/m3 to less than 1 g/m

3 in the experiments. A

thermal analysis of the experiments was performed. The analysis is consistent with the result that rapid

crusting with minimal concrete ablation occurs in both experiments.


J.E. Gronager, A.J. Suo-Anttila, and J.E. Brockmann, “TURC2 and 3: Large Scale UO2/ZrO2/Zr Melt-

Concrete Interaction Experiments and Analysis”, NUREG/CR-4521; SAND86-0318, June 1986.


TURC2: A melt of 140 kg UO2 and 60 kg of ZrO2 was prepared in the IRIS generator and poured

into a magnesia crucible with a 46.1 cm diameter bottom slug of limestone/common sand concrete.

The melt cooled naturally.

TURC3: A melt of 123.4 kg UO2 , 27.3 kg of ZrO2, and 9 kg Zr was prepared in the IRIS

generator and poured into a magnesia crucible with a 46.1 cm diameter bottom slug of

limestone/common sand concrete. The melt cooled naturally.





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4.5.31 E5-31 - LSL-1,2,3




Tests involved the formation of 220 kg of molten stainless steel heated to 1700C and teemed into

large concrete crucibles with 61 cm diameter hemispherical cavities. Solidified melt was extracted,

erosion profiles measured and the test was repeated twice to obtain two-dimensional erosion profiles.

Concrete used in the experiments was made using limestone coarse aggregate and common sand (SiO2)

fine aggregate.


D.A. Powers and F.E. Arellano, “Large-Scale, Transient Tests of the Interaction of Molten Steel with

Concrete”, NUREG/CR-2282; SAND81-1753, 1982


Melt mass and temperature: 200 kg stainless steel initially at 1700C and allowed to cool naturally

Angle of concrete surface exposed to melt: varied between 0 and 90

Hydration of concrete: varied between “as-placed” to substantially dehydrated.



Past Code Validation/Benchmarks: Results used to validate and refute elements of the model of

melt/concrete interactions used in the Reactor Safety Study (WASH-1400).


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4.5.32 E5-32 - LBL-1,2,3







Concrete used in the experiments was made using a siliceous (basalt) coarse aggregate and common sand

(SiO2) fine aggregate.





Melt mass and temperature: 220 kg stainless steel initially at 1700C and allowed to cool naturally





Past Code Validation/Benchmarks: Results used to validate and refute elements of the model of

melt/concrete interactions used in the Reactor Safety Study (WASH-1400).


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4.5.33 E5-33 - LSCRBR-1,2,3







Concrete used in the experiments was made using limestone coarse aggregate and crushed limestone fine

aggregate.





Melt mass and temperature: 200 kg stainless steel initially at 1700C and allowe







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4.5.34 E5-34 - COIL-1




205 kg type 304 stainless steel was melted and heated to 1973K. The melt was teemed into a

limestone concrete crucible cavity 38.1 cm in diameter and 38.1 cm deep. The melt was sustained by an

embedded induction heating coil.


D.A. Powers, “Sustained Molten Steel/Concrete Interactions Tests”, NUREG/CR-0166; SAND77-1423,

June 1978


Concrete type: limestone coarse and fine aggregate

Melt type: 205 kg stainless steel heated initially to 1973 K and sustained by induction heating

Melt-concrete orientation: two –dimensional so both axial and radial ablation monitored.





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4.5.35 E5-35 - WETCOR-1




The WETCOR-1 test of simultaneous interactions of a high-temperature melt with water and a

limestone/common-sand concrete is described. The test used a 34.1-kg melt of 76.8 wt.% A12O3, 16.9

wt.% CaO, and 4.0 wt.% SiO2 heated by induction using tungsten susceptors. Once quasi-steady attack on

concrete by the melt was established, an attempt was made to quench the melt at 1850 K with 295 K water

flowing at 57 liters per minute. Net power into the melt at the time of water addition was 0.61 0.19

W/cm3. The test configuration used in the WETCOR-1 test was designed to delay melt freezing to the

walls of the test fixture. This was done to test hypotheses concerning the inherent stability of crust

formation when high temperature melts are exposed to water. No instability in crust formation was

observed. The flux of heat through the crust to the water pool maintained over the melt in the test was

found to be 0.52 0.13 MW/m2. Solidified crusts were found to attenuate aerosol emissions during the

melt concrete interactions by factors of 1.3 to 3.5. The combination of a solidified crust and a 30-cm deep

subcooled water pool was found to attenuate aerosol emissions by factors of 3 to 15.


R.E. Blose et al., “Core Concrete Interactions with Overlying Water Pools – The WETCOR-1 Test”

NUREG/CR-5907; SAND92-1563, November 1993


34.1 kg of a melt composed of 76.8% alumina, 16.9% calcia, and 4% silica were sustained by

induction heating in contact with limestone common sand. Water was added at 57 L/min. The test

configuration attempted to delay crust freezing to walls of the crucible. Heat flux to the water pool

was 0.52 MW/m2.





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4.5.36 E5-36 - FRAG




Four experiments were performed to study the interactions between low-temperature core debris and

concretes typical of reactor structures. The tests addressed accident situations where the core debris is at

elevated temperature, but not molten. Concrete crucibles were formed in right-circular cylinders with 45

kg of steel spheres (approx.3-mm diameter) as the debris simulant. The debris was heated by an inductive

power supply to nominal temperatures of 1473 K to 1673 K. Two tests were performed on each of two

concrete types using either basalt or limestone aggregate. For each concrete, one test was performed with

water atop the debris while the second had no water added. The results show that low-temperature core

debris will erode either basalt or limestone-common sand concretes. Downward erosion rates of 3 to 4

cm/h were recorded for both concrete types. The limestone concrete produced a crust layer within the

debris bed that was effective in preventing the downward intrusion of water. The basalt concrete crust was

formed above the debris and consisted of numerous, convoluted, thin layers. Carbon dioxide and water

release from the decomposition of concrete were partially reduced by the metallic debris to yield carbon

monoxide and hydrogen, respectively. The overlying water pool did not effect the reduction reactions.


W.W. Tarbell et al., “Sustained Concrete Attack by Low-Temperature, Fragmented Core Debris”,

NUREG/CR-3024; SAND82-2476, July 1987


Test Name Test Conditions

FRAG-1 44.5 kg mild steel spheres inductively heated in contact with basaltic concrete

FRAG-2A 45 kg mild steel spheres inductively heated in contact with limestone/common sand

concrete

FRAG 3 45 kg mild steel spheres inductively heated in contact with limestone/common sand

concrete. A water pool was formed over the debris during the attack on concrete

FRAG 4 45.5 kg mild steel spheres inductively heated in contact with basaltic concrete. A

water pool was formed over the debris during the attack on concrete



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4.5.37 E5-37 - 1DHtFlx

Test Facility: Sandia National Laboratories High Heat Flux Test Facility



Experiments were carried out to investigate the erosion of concrete under high surface heat flux in

connection with the core-melt/concrete interaction studies. The dominate erosion mechanism was found to

be melting at the surface accompanied by chemical decomposition of the concrete beneath the melt-solid

interface. The erosion process reaches a steady state after an initial transient. The steady state is

characterized by an essentially constant erosion rate at the surface and a non-varying (with respect to the

moving melt interface) temperature distribution within the concrete. For the range of incident heat flux 64

W/cm2 to 118 W/cm

2, the corresponding steady state erosion rate varies from approximately 8 cm/h to 23

cm/h. A simple ablation/melting model is proposed for the erosion process. The model was found to be

able to correlate all temperature responses at various depths from all tests at large times and for

temperatures above approximately 250C.


T.Y. Chu, “Radiant Heat Evaluation of Concrete – A Study of the Erosion of Concrete due to Surface

Heating”, SAND77-0922, January 1978


15 cm diameter slugs of concrete exposed to heat fluxes of 64 to 118 W/cm2





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4.5.38 E5-38 – MC Tests

Test Facility: Sandia National Laboratories Melt Interactions Test Facility



Molten stainless steel (~200 kg at 1700C) was teemed into mild steel hemispheres and onto plates 0.95 to

7.62 cm thick. Times to melt penetration were monitored.


D.A. Powers, “Erosion of Steel Structures by High-Temperature Melts”, Nuclear Science and Engineering,

88 (1984) 337-368.


Melt mass: 207-220 kg

Melt composition: stainless steel

Melt temperature: 1705 to 1730C

Melt pour duration: 22.9 to 26.4 s

Structure thickness: 0.95 to 7.62 cm





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4.5.39 E5-39 – Plate Tests

Test Facility: Sandia National Laboratories Melt Interactions Test Facility



Melts of 3-5 kg were generated thermitically and drained onto steel plates. Ten tests were done with

molten iron and aluminum oxide. One test was done with a melt of uranium dioxide, zirconium dioxide

and stainless steel. Steel plates were 0.95 to 1.27 cm thick. In two tests the plates were coated with a layer

of uranium dioxide to simulate a crust. Times to melt penetration were monitored.


D.A. Powers, “Erosion of Steel Structures by High-Temperature Melts”, Nuclear Science and Engineering,

88 (1984) 337-368.


Melt Mass: 3 to 5 kg

Melt Composition: Fe/Al2O3 and UO2/ZrO2/Fe/Cr/Ni

Melt Temperature: 2400 to 2780C

Melt pour duration: 3.38 to 5.44 s

Structure thickness; 0.95 to 1.27 cm. Two tests with plates coated with 1-2 mm uranium oxide.





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4.6 Systems Experiments

4.6.1 E6-1 - CSE EFADS Tests

Test Facility: CSE

Owner Organization: BNWL and AEC


The CSE facility was sized to represent one-fifth linear model of a typical 1000 MWe PWR. The

vessel has an 870 m3 volume with 7.6 m diameter and 20.4 m height. Four compartments were arranged

inside the vessel (main room, dry well, middle room and lower room) with a total free volume of 750 m3.

The called wetwell compartment (120 m3) was sealed out and not exposed to steam and fission product

simulants in this experiment, which were injected into the lower part of the dry well compartment.

A modular filter loop was located in the main room, over the dry well. The loop components (heat

exchangers, moisture separators, prefilter, HEPA filter, activated charcoal beds) were selected to be typical

of those used in containment systems. The loop was instrumented to provide data of air flow rate,

temperature, pressure differences and fission products removal efficiency.

Four materials were released in these tests: elemental iodine, methyl iodide, caesium and UO2

particles. Iodine and caesium were traced with 131

I and 137

Cs, respectively. The bottle containing the

iodine or caesium is heated when release is desired and air is used as carrier gas to sweep the iodine and

caesium away. This air stream passes through a UO2 furnace and then injected into the vessel. Two types

of aerosol releases were used: a short term “puff” release, and a continuous, or “linear” release. The puff

release simplifies the fission product generation equipment whereas the longer release provides higher

aerosol concentration of caesium and uranium.

Twelve individual maypacks samplers were clustered together and placed at a selected position in the

containment vessel. Other samples were obtained by inserting individual maypacks into the vessel through

an airlock. Maypacks samplers were designed to characterize airborne iodine according to its chemical

identity. To measure aerosol size distribution, a cascade impactor, inserted into the vessel atmosphere,

collected airborne particles.


J.D. McCormack, R.K. Hilliard and A.K. Postma, 1971. “Removal of airborne fission products by

recirculating filter systems in the containment system experiment”. Battelle Memorial Institute Pacific

Northwest Laboratories, BNWL-1587

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Iodine concentration: 160 mg/m3

Caesium and uranium concentration: 100 mg/m3

AMMD: 0.5- 1.0 μm





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4.6.2 E6-2 - ACE-CSTF EFADS Tests

Test Facility: CSTF (Containment Systems Test Facility)

Owner Organization: BNWL, HEDL, (Sponsored by EPRI)


The objective of the Advanced Containment Experiment (ACE) program was to measure the

efficiency of different aerosol retention devices that are used in nuclear power plants: pool scrubbers, dry

sand/gravel beds, submerged gravel beds, multiventuri scrubber submerged, combined venturi scrubbers,

fibrous metallic filters and soviet filter systems.

The experiments were performed in the CSTF of the Handford Engineering Development Laboratory

(HEDL). This facility mainly consists of the containment vessel (a 852 m3 vessel) in which interior is

located the filter test vessel with the filter system to be tested. Outside the containment vessel and

upstream of the filter test vessel, there is the aerosol mixing vessel and the aerosol generators. Three

different aerosols are used in the experiment: CsOH, CsI and MnO, with aerodynamic mass median

diameters between 1 and 2 μm and geometric standard deviations around 1.8. Elemental caesium is heated

in a furnace and sweep by a low flow rate nitrogen stream into the aerosol mixing vessel. The steam

containing the aerosol mixing vessel reacts with Cs to form CsOH. Additionally, a controlled flow of HI

(hydrogen iodide) is injected into the AMV to produce CsI. The overall ratio of Cs to I in the aerosols is

about 10:1. Manganese powder is volatilized in a plasma torch and carried by nitrogen into the aerosol

mixing vessel where it reacts with steam to produce MnO. Aerosols are carried from the aerosol mixing

vessel to the filter test vessel by a mixture of nitrogen and steam (at about 0.25 m3/s). There were three

sampling stations, one upstream and two downstream of the filter test vessel in order to measure the

concentration and size distribution before and after the gas pass through the filter system.

The same thermal-hydraulics conditions were used to test the different filtration systems. For each

one, two to six 30 minute tests were performed.


M. Merilo, I.B. Wall, 1992. “Containment Filtration Systems Tests, Summary Report” Electric Power

Research Institute, ACE Phase A, TR-A22, February 1992

J.D. McCormack, D.R. Dickinson and R.T. Allemann. 1989. “Experimental results of ACE vent filtration,

pool scrubber tests”. ACE Phase A, TR-A1, January 1989


submerged graver scrubber tests”. ACE Phase A, TR-A2, July 1989


submerged multi-venturi scrubber tests”. ACE Phase A, TR-A3, September 1989

R.K. Hilliard, J.D. McCormack and A.K. Postma. 1981. “Submerged gravel scrubber. Demonstration as a

passive air cleaner for containment venting and purging with sodium aerosols. CSTF Tests AC7 – AC10”.

ACE Phase A, TR-A8, HEDL-TME 81-30. November 1981

J.D. McCormack, R.K. Hilliard and A.K. Postma. 1984. “Submerged gravel scrubber. Demonstration tests;

Performance of a large-scale unit”. ACE Phase A, TR-A9, HEDL-TME 83-120. December 1984

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J.D. McCormack, D.R. Dickinson and R.T. Allemann. 1990. “Experimental results of ACE vent filtration

tests. Heat sink gravel bed”. ACE Phase A, TR-A10, April 1990

R.T. Allemann and J.A. Bamberger. 1990. “Comparison of code results with ACE pool scrubbing tests”.

ACE Phase A, TR-A13, June 1990

D.R. Dickinson, J.D. McCormack, and R.T. Allemann. 1991. “Experimental results of ACE vent filtration:

soviet filter tests”. ACE Phase A, TR-A15, June 1991

G.L. Ogram and A. Lemyk. 1990. “Water aerosol leakage experiments, volume I: droplet size distributions

measured by FSSP”. ACE Phase A, TR-A17, February 1990

C.F. Forrest. 1990. “Water aerosol leakage experiments, volume IV: test results”. ACE Phase A, TR-A20,

February 1990


Gas flow:

nitrogen: 0.2 m3/s ;

air: 0.05 m3/s.

Aerosol size:

CsOH: AMMD=1.8 μm, GSD=1.8

CsI: AMMD=1.0 μm, GSD=1.6

MnO: AMMD=1.5 μm, GSD=1.8

Aerosol concentration (in carrier gas):

Cs: 5 – 25 g/m3

Mn: 5 – 24 g/m3

I: 0.5 – 1.7 g/m3





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4.6.3 E6-3 - ACE-LSFF EFADS Tests

Test Facility: LSFF

Owner Organization: HEDL, (Sponsored by EPRI)


A large-scale submerged gravel scrubber (SGS) was installed in the Large Sodium Fire Facility

(LSFF) at the Hanford Engineering Development Laboratory (HEDL). The large test cell is a 107 m3 (5.5

m high) rectangular vessel made on steel. The objective of the test program was to demonstrate the

effectiveness of this type of passive scrubber on a size scale of commercial interest. The test piece

consisted of a submerged gravel bed 0.61 m in depth, 3.66 m in width and either 2.1 and 3.1 m in length.

Two purposes could be served: first, test aerosols could be generated to challenge the submerged gravel

scrubber, and second, the submerged gravel scrubber could be used to clean up smoke produced during the

experiment. Hydraulic characteristics and aerosol retention efficiencies were measured in a series of 12

tests.

Test aerosols were introduced into the submerged gravel scrubber in two ways. In the first one,

sodium and lithium metal was burned in test cells inside the LSFF, and the aerosols that were produced

(which were soluble oxides) were transported to the scrubber through the normal exhaust ducts (existing

filter bank was bypassed). The second test method involved the injection of insoluble particles (hydrous

aluminium oxide and fly ash) into the scrubber inlet duct some 12 m upstream from the scrubber.

Upstream and downstream samples were withdrawn using filters and cascade impactors to determine

inlet and outlet aerosol concentration and size distribution. Typically in the order 0.5 – 1.0 g/m3 in

concentration and of 2-3 μm in AMMD (with GSD value around 2). Air flow through the test scrubber

was provided by a radial blade blower and water level inside the scrubber was controlled by two float

switches.


J.D. McCormack, R.K. Hilliard and A.K. Postma, 1984. “Submerged gravel scrubber demonstration test;

performance of a large-scale unit”. Hanford engineering Development Laboratory HEDL-TME-83-12.

ACE Phase A, TR-A9.

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Metal burned aerosol experiments:

Aerosol concentration: 0.34 -0.97 g/m3

Particle size: 2.1 – 3.2 μm (GSD 1.6 -2.4)

Gas flow rate: 4.0 – 5.5 m3/s

Insoluble aerosol experiments:

Aerosol concentration: 0.26 - 0.41 g/m3

Particle size: 1.3 – 22 μm

Gas flow rate: 4.0 – 5.6 m3/s





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5 PHENOMENA VS. EXPERIMENTS CROSS MATRIX

A phenomena vs. experiments cross matrix (not shown in this report) was generated using the

information in Table 4-2 to Table 4-7. The cross-matrix showed the experiments that can be used to

validate each phenomenon. Please note that it is still the User’s responsibility to assess the suitability of

the experiment for their code validation. The results of the cross matrix was used to generate a list of

experiments that may be used to validate each phenomenon (see Table 3-1 to Table 3-6)8. The cross-

matrix shows that the following phenomena do not have any experiments identified for validation:

P1-22 - Laminar/Turbulent Leakage Flow

P1-30 - Droplet Interaction (Dousing)

P1-32 - Turbulence Induced by Sprays

P2-11 - Strong Ignition of Hydrogen

P3-11 - Drop Breakup

P5-22 - Ex-Vessel Corium Catcher - Corium-Ceramics Interaction and Properties

P5-29 - Corium Release from Failed Flooded Reactor Pressure Vessel

P6-1 - Ventilation Systems

P6-4 - Pump Performance including Sump Clogging (No Experiments)

Thus, this containment code validation matrix is only missing experiments for 9 phenomena.

Although this CCVM appears to contain experiments to cover about 93% of the containment

phenomenon, it is a little misleading because only about half of the experiments can be used for validation

of CFD codes. The 67 phenomena lacking experiments for validation of CFD codes are shown in Table

5-1. An examination of this list of phenomena shows that the bulk of them cannot (or are not) presently

being modelled by CFD codes.

Table 5-1

List of Phenomenon without Identified Experiments for CFD Validation

P1-12 - Liquid Re-Entrainment (Resuspension)



P1-22 - Laminar/Turbulent Leakage Flow


P1-30 - Droplet Interaction (Dousing)

P1-32 - Turbulence Induced by Sprays

P2-10 - Hydrogen Mitigation by Hydrogen Ignitors (Mild Ignition)

P2-11 - Strong Ignition of Hydrogen

P2-13 - Radiolysis (Hydrogen Production by Water Radiolysis)

P3-1 - Aerosol Formation in a Flashing Jet

8 There are 6 phenomena that do not require validation (phenomenon title include the word “No Experiments”). The

reasons are given in the phenomenon descriptions.

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Table 5-1


P3-2 - Aerosol Formation in a Steam Jet

P3-3 - Aerosol Impaction (Jet Impingement)







P3-11 - Drop Breakup


P3-14 - Inertial Deposition of Aerosols (Also called Impaction)


P3-16 - Re-volatilisation

P3-17 - Aerosol Removal in Leakage Paths

P3-18 - Pool Scrubbing of Aerosols

P3-21 - Release Rate Change Due to Oxidizing Environment

P3-22 - Containment Chemistry Impact on Source Term

P3-23 - Ruthenium Volatility and Behaviour in Containment

P3-24 - Aerosol Removal by Sprays (Dousing)

P3-25 - Re-suspension (Dry)

P3-26 - Re-entrainment (Wet)

P3-27 - Aerosol De-agglomeration

P4-1 - Aqueous Phase Oxidation and Reduction of Iodine Species



P4-4 - Homogeneous Organic Reactions in Water Phase

P4-5 - Iodine Reactions with Surfaces in the Water Phase


P4-7 - Silver Iodine Reactions in the Water Phase

P4-8 - Gas Phase Radiolytic Oxidation of Molecular Iodine (I2) (Iodine/Ozone Reaction)

P4-9 - Homogeneous Organic Iodine Reactions in Gas Phase



P4-13 - Iodine Filtration

P4-14 - Volatile Iodine Trapping by Airborne Droplets


P4-16 - I2 Interaction with Aerosols

P4-18 - Pool Scrubbing of Iodine

P5-1 - Corium Release from Failed Dry Reactor Pressure Vessel

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Table 5-1


P5-2 - Corium Entrainment Out of the Reactor Primary Vessel with Lateral Breaches

P5-3 - Corium Particles Generation from the Corium Pool

P5-4 - Corium Particles Generation from the Two Phase Jet


P5-6 - Corium Particles Trapping

P5-8 - Corium Jet Break-up in Water Pool

P5-10 - Pressure Load on Corium Retention Devices


P5-15 - Corium Spreading

P5-22 - Ex-Vessel Corium Catcher - Corium-Ceramics Interaction and Properties

P5-23 - Effect of Non Homogeneous Ablation on Gate Ablation

P5-29 - Corium Release from Failed Flooded Reactor Pressure Vessel

P6-1 - Ventilation Systems

P6-2 - Behaviour of Doors, Burst Membranes, Rupture Discs etc.

P6-3 - Air Cooler (Fan Cooler) Heat Transfer

P6-4 - Pump Performance including Sump Clogging (No Experiments)

P6-6 - Aerosol Removal in EFADS

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6 SUMMARY

This containment code validation matrix has identified 127 phenomena related to both DBA and

SA/BDBA accident scenarios in western pressurised heavy water reactor (PHWR), pressurised water

reactor (PWR) and boiling water reactor (BWR), as well as Eastern European VVER reactors. It contains

a description of 213 experiments that may be suitable for validation of the 1129 of the identified

containment phenomena.

If only experiments suitable for CFD validation are considered, then only 54 containment phenomena

are covered by this CCVM. However, most of the uncovered phenomena are not presently modelled with

CFD codes.

The authors of this report do not make any claims to the suitability of the experiments for code

validation. It is the responsibility of the User to assess the suitability of the experiment for their code

validation.

It is recommended that this work be reviewed in 5 years time to include new experiments and to

attempt to close the identified experiment gaps (phenomena lacking suitable experiments for validation).

9 Of the 127 phenomena, only 121 require experiments for validation. Of the 121, only 9 phenomena do not have

experiments identified for validation.

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