M. ADORNI (BEL V), S. BAND (GRS), L. CARENINI, P. CHEVRIER (IRSN), G. MAZZINI*1(CVR)
1.2. SEVERE ACCIDENT PHENOMENOLOGY
*1Work performed under the project "Development of the knowledge base in the field of nuclear safety assessment", CZ.1.07/2.3.00/30.0026, Ministry of Education, Youth and Sports, CR
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CONTENTS
1 INTRODUCTION
2 IN-VESSEL
3 HYDROGEN IN THE CONTAINMENT
4 FISSION PRODUCT TRANSPORT AND RETENTION
5 VESSEL FAILURE
6 DIRECT CONTAINMENT HEATING (DCH)
7 STEAM EXPLOSION
8 MOLTEN-CORE CONCRETE INTERACTION (MCCI)
9 CONCLUSION
10 REFERENCES
INTRODUCTION
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1.1FOREWORD
Severe accidents (SA) have the potential to cause large releases of radioactive material and thus pose a risk to public health and safety as well as the environment.
The processes involved in the progression of severe accidents are very complex, requiring experimental data to support the development of models, to determine design and procedural requirements and to prevent and/or mitigate the consequences of such accidents.
Severe accident is an active field of research and several issues are still open.
Severe accidents have to be considered when the core is in the reactor, but also when the whole core or a large part of the core is unloaded and stored in the fuel pool*.
*Spent fuel pool accidents not included in the current presentation.
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1.2 DEFINITION OF SEVERE ACCIDENT
Accident conditions more severe than a design basis accident and involving significant core degradation.
IAEA Satefy Glossary, 2007 Edition
Due to multiple failures and/or operator errors, safety systems fail to perform one or more of their safety functions leading to significant core or fuel damage that challenges the integrity of the remaining barriers to prevent the release of radioactive material from the plant.
ETSON, Technical Safety Assessment Guide on “Deterministic Severe Accidents Analysis”
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1.3 OBJECTIVE OF THE PRESENTATION
To document the advances achieved in the field of severe accident in light water reactors (focus on PWR), highlighting the phenomena understanding.
Does not include all the knowledge about SA, but aims more at transferring the understanding and knowledge about the complexity of the SA. For more details it should refer to specific literature on the subject.
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1.4 SEVERE ACCIDENT ACTIVITIES IN BELGIUM
Severe accident issue is dealt in Belgium within the Periodic Safety Reviews (PSR) since 90’s. After Fukushima accident, additional commitment comes from the Belgian stress tests. Severe accidents are addressed in the Belgian law (“Royal Decree of November 30, 2011”).
All 7 Belgian units were fitted with PARs between 1995-1998.
The first PSA level 2 models for Doel 1/2 and Tihange 1 and the assessment of Doel 3 and Tihange 2 containment performances against severe accident phenomena was developed before 2004.
Severe Accident Management Guidance (SAMG) was implemented (and continuously updated) for all the plants. The validation of the SAMG for Doel and Tihange units were performed after 2004 with the aim of verifying the ability of the Technical Support Center (TSC). Several SA scenarios including shutdown states were tested.
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1.5SEVERE ACCIDENT ACTIVITIES IN FRANCE
Most of French SA measures were taken after TMI-2 accident (ultimate operating procedures, emergency plans, R&D programs…)
Specific SA measures (emergency filtered containment venting system, PARs, RPV failure detection…)
Guidelines:
Severe Accident Response Guideline: defines some specific actions in order to get the lowest radioactive releases
Severe accident safety standard: defines safety requirement in SA
All recent efforts done in France on SA aim at:
Identifying the sequences of events that may lead to large radioactive release (use of PSA level 2, deterministic analysis and R&D activities)
Improving the operating procedures or the plant design to make these sequences of events almost impossible
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1.6SEVERE ACCIDENT ACTIVITIES IN CZECH REPUBLIC
NRI Experience:
Support to SUJB
CSARP – NRC SA Codes
PHEBUS FP Program
5th FWP / SARNET – 6th FWP / SARNET2 – 7th FWP
Project and Facility of OECD/NEA/CSNI
RCRez Experience:
Fukushima Daiichi NPP Project - Analyses of stress tests of Czech nuclear power plants
CSARP – NRC SA Codes
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1.7SEVERE ACCIDENT ACTIVITIES IN GERMANY
SAM activities in Germany
Preventive and mitigative (Filtered Venting (late `80s), PARs orInertisation (90s), Upgrade of emergency power supplies (bunkered)
Completion of European “Stress Test” & Definition of findings and NPP improvement needs
Different experimental facilities (THAI (aerosol behavior, hydrogen (stratification combustion and recombination), HDR, PANDA, Battelle, DISCO)
Analysis of the Fukushima Daiichi NPP accident (Validation of GRS-computer code ATHLET-CD in the BSAF Project)
Participation in OECD-NEA-CSNI Projects
Experiments and Analysis about sump clogging (e.g. in a Siemens test facility Erlangen)
Retrofitting partly of existing NPPs hardware and procedures (local government)
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IN-VESSEL PHENOMENA
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2.1IN VESSEL PHENOMENA: EVENT INITIATORS
For Light Water Reactors (LWRs) we can distinguish severe accident initiators into two broad classes:
Reactivity Insertion Accidents (RIAs)
Core Uncovery Accidents (CUAs)
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2.2CORE DAMAGE AS A FUNCTION OF TEMPERATURE
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2.3CORE DEGRADATION PHENOMENA AS FUNCTION OF TIME (BASE ON PHEBUS)
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2.4CORE DEGRADATION PHENOMENA AND HYDROGEN PRODUCTION
What occurred between intact corestate and final degradation ?•Formation of upper cavity•Solid debris bed•In-core molten pool•Surrounding crust•Corium relocation•Molten materials in lower plenum•Final corium cooling
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2.4CORE DEGRADATION PHENOMENA AND HYDROGEN PRODUCTION
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2.4CORE DEGRADATION PHENOMENA AND HYDROGEN PRODUCTION
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2.4CORE DEGRADATION PHENOMENA AND HYDROGEN PRODUCTION
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2.5RELAP/SCDAPSIM EXAMPLE (BASE ON UNIT3 –FUKUSHIMA DAIICHI NPP)
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151.000 s
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2.5RELAP/SCDAPSIM EXAMPLE (BASE ON UNIT3 –FUKUSHIMA DAIICHI NPP)
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160.000 s
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2.5RELAP/SCDAPSIM EXAMPLE (BASE ON UNIT3 –FUKUSHIMA DAIICHI NPP)
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163.000 s
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2.5RELAP/SCDAPSIM EXAMPLE (BASE ON UNIT3 –FUKUSHIMA DAIICHI NPP)
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170.000 s
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2.5RELAP/SCDAPSIM EXAMPLE (BASE ON UNIT3 –FUKUSHIMA DAIICHI NPP)
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Fukushima Daiichi NPP Hydrogen Production
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2.6CORIUM BEHAVIOR IN THE LOWER PLENUM
Corium jet fragmentation in water debris bed formation
Debris bed coolability ? : residual power / debris bed porosity / water reinjection
Molten corium pool formation Possible oxide-metal phases separation focusing effect :
influenced by zirconium oxidation degree (~25%-80%),
crust and debris bed positions and characteristics
heat flux distribution on vessel wall
Main issue for :
In-vessel retention strategy,
Characteristics of corium ejected at vessel failure (DCH, MCCI studies…)
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HYDROGEN INSIDE THE CONTAINMENT
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Hydrogen concentration inside the containment is dependent on scenario and mostly non-uniform.
3.1HYDROGEN INSIDE THE CONTAINMENT
Hydrogen Combustion
April 12, 2012
Scenario SBO, release via pressurizer / relief tank
Medium Break LOCA, Primary Cooling active till switch to sump recirculation, Secondary Side Shutdown with 100K/h
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3.2HYDROGEN INSIDE THE CONTAINMENT
Hydrogen Combustion
Hydrogen combustion:
Air > 22%-27%
H2 (or other burnable gases) > 4%
Inert gases < 62% (less for steam)
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3.3HYDROGEN INSIDE THE CONTAINMENT
Hydrogen Combustion
Velocity of Flame Propagation is depended on
concentration of burnable gases
flow velocity just before ignition
installed equipment / junctions that could cause turbulence
steam limits velocity, high temperature increases velocity
Laminarv≈1m/s, Ma << 1
Turbulentv ≈ 300m/s, Ma ≈ 1
(quasi-)DetonationV>1000m/s, Ma > 1
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3.4HYDROGEN INSIDE THE CONTAINMENT
Hydrogen Combustion
Combustion Experiment L.VIEW AnlageNEA/CSNI/R(2000)7
10.5 Vol.% H2
combustion starts in the first room
slow combustion and compression of unburned gas upstream flame front
flame propagation in second room and immediate combustion
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3.5HYDROGEN INSIDE THE CONTAINMENT
countermeasures
There are two possible means to deal with hydrogen:
Passive Autocatalytic Recombiner:
passive system
H2 and CO are recombined, rate increases with concentration / temperature
Continuous energy production
Short peak injection cannot be dealt with very fast.
Experiments in the THAI facility showed hydrogen ignition in the outlet (concentration 8-12 Vol.%)
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chimney
Platin / palladium plates
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3.6HYDROGEN INSIDE THE CONTAINMENT
countermeasures
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Igniters:
Active system
Not operative when the gas mixture is not burnable (H2 < 4.2 Vol.%, steam > 55Vol.%)
Very high temperature, strong pressure increase
Can cope with high release rates very fast.
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3.7HYDROGEN INSIDE THE CONTAINMENT
Experimental data:
Forschungszentrum Jülich REKO-3 Experiment (3 Plates, gas composition well defined, temperature profile measured)
THAI Facility: HR-.. Experiments (PAR behavior with different boundary conditions), HD-.. Experiments Hydrogen Deflagration)
Pro-Sience / KIT experiments for deflagration in stratified atmosphere
Also experiments in the Battelle and HDR Facility, although the data basis isn’t ideal.
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FP TRANSPORT AND RETATION
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4.1MAIN PHENOMENA INFLUENCIG SOURCE TERM
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4.1MAIN PHENOMENA INFLUENCIG SOURCE TERM
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4.2SYSTEMS FOR RETATION OF SOURCE TERM
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4.2SYSTEMS FOR RETATION OF SOURCE TERM
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4.2SYSTEMS FOR RETATION OF SOURCE TERM
VENTING VENTING SYSTEMS SYSTEMS
IN SWEDISH IN SWEDISH PWRPWR
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4.3AEROSOL TRANSPORT EXAMPLE (FPT1 CONTAINMENT)
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0
10
20
30
40
50
60
70
80
90
0 5000 10000 15000 20000 25000 30000
Time (s)
Susp
. Mas
s (g)
MELCORASTECExperimental
Total suspended aerosols mass
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4.3AEROSOL TRANSPORT EXAMPLE (FPT1 CONTAINMENT)
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0
1
2
3
4
5
6
7
8
9
10
0 5000 10000 15000 20000 25000 30000
Time (s)
Susp
. Mas
s (g) MELCOR
ASTECExperimental
Uranium aerosol suspended mass
VESSEL FAILURE
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5.1THREATEN OF THE VESSEL INTEGRITY
■ Heat flow and temperature field from corium pool into vessel wall
■ Internal pressure■ Deadweight of RPV and melt
pool■ Thermo-mechanical attack of
corium ■ Steam explosion
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Lower head failure: final stage of the in-vessel accident progression
MOLTEN POOL
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5.2MECHANISMS HAVING THE POTENTIAL TO CAUSE LOWER HEAD FAILURE
■ Penetration Tube Heatup and Rupture■ Penetration Tube Ejection■ Lower head Global Rupture■ Localized Effects Jet Impingement■ Steam explosion
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5.3PREDICTING LOWER HEAD FAILURE AND MAIN RPV FAILURE EXPERIMENTS
■ LHF (SNL, 1994-1999) ■ OLHF (SNL, 1999-2002)■ FOREVER (KTH, 1998-2002)
■ Lower head failure prediction Experiments revealed a variability of material behavior around 1000°C (fragile or ductile failure) and that consequently the final breach size is strongly impacted by material composition.The computational models (simplified or finite element) have demonstrated their capacity to determine the lower head failure time and the zone in which the crack is initiated. For a more precise investigation of crack propagation and final breach size, the only option at present is a 3D finite element model.
MELCOR LH nodalization
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DIRECT CONTAINMENT HEATING (DCH)
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6.1PHENOMENA DURING SEVERE ACCIDENTS
Direct Containment Heating
Melt Entrainment in gas jet can disperse melt to areas in the containment where controlled cooling is impossible.
Increased surface and distribution leads to an increase in heat transfer and boosts interaction of steam with not oxidized melt, so Hydrogen production is increased.
Melt can ignite Hydrogen.
Consequences are very much plant specific due to different cavity designs.
Figure 1: DCH aspects
Direct Containment Heating (DCH) by dispersed melt is a consequence of a failure of the RPV under elevated/high pressure
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6.2PHENOMENA DURING SEVERE ACCIDENTS
Direct Containment Heating
April 12, 2012
Typical debris dispersal fractions in different reactor designs:
GA 301
0,0
0,2
0,4
0,6
0,8
1,0
ZION CalCliffs EPR P'4 VVER
DEB
RIS
FR
AC
TIO
N
Containment
Compartments
Bottom access
Pit
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6.3PHENOMENA DURING SEVERE ACCIDENTS
DCH Experiments
April 12, 2012
Figure 1: DISCO Facility
“Melt Dispersion and Direct Containment Heating (DCH) Experiments in the DISCO-H Test Facility” L. Meyer, G. Albrecht, M. Kirstahler, M. Schwall, E. Wachter, G. Wörner, May 2004
A couple of experiments were performed at the Containment Technology Test Facility (CTTF) at Sandia National Laboratories (SNL) between 1990 and 1994.
“DCH Dispersal and Entrainment Experiment in a Scaled Annular Cavity” M. Lopez de Bertodano, A.Becker, A.Sharon, R. Schnider, Nuclear Engineering and Design, Vol 164
“Experiments to Investigate Direct Containment Heating Phenomena with Scaled Modesla of the Calvert Cliffs Nuclear Power Plant” T.K. Blanchat et al., NUREG/CR-6469, SANDA96-2289, SNL
“Direct Containment Heating Experiments at Low Reactor Coolant System Pressure in the Surtsey Test Facility”T.K. Blanchat et al., NUREG/CR-5746, SANDA99-1634, SNL
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6.4PHENOMENA DURING SEVERE ACCIDENTS
Direct Containment Heating
April 12, 2012
Figure 1: (top view) from the DISCO-H06 test (2,2 bar, direct opening to Containment, driving gas: steam, hole size: 28mm)
STEAM EXPLOSION
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7.1STEAM EXPLOSION PHENOMENOLOGY (1/3)
Steam explosion may occur when two fluids are brought into contact and the temperature of one (e.g. corium) is much higher than the boiling temperature of the other (e.g. coolant): it’s one of possible effects of fuel coolant interaction (FCI)
T(corium) > Tboiling(coolant)
Different contact modes susceptible to induce a steam explosion:
Corium pouring into water in the vessellower plenum (in-vessel explosion)
Corium pouring into water in the reactor pitafter vessel failure (ex-vessel explosion)
Water pouring onto coriumin case of reflooding (stratified explosion)
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7.1STEAM EXPLOSION PHENOMENOLOGY (2/3)Steam explosion can be decoupled in 2 steps (with different space and time scales):
Premixing: rough and slow mixing of corium and coolant(~ 1 mm ; ~ 1 s)
Explosion: fast mixing with fast energy release leading to shockwaves build-up(~ 100 µm ; ~ 1 ms)
The second step needs a triggering event to be initiated
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7.1STEAM EXPLOSION PHENOMENOLOGY (3/3)
Steam explosion is caused by unstable interactions:
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Necessary condition for an explosion:
Characteristic time for heat transfer
Characteristic time for pressure relief<
Strength of explosion strongly depends on initial conditions, given by the mixing process (premixing phase)
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7.2STEAM EXPLOSION: EXPERIMENTAL FACILITIES
Integral experiments:
KROTOS (CEA - France)
TROI (KAERI - Korea)
OECD program: SERENA
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Analytical tests:
DROPS (IKE - Germany)
MISTEE (KTH - Sweden)
DISCO-FCI(KIT – Germany and IRSN – France)
KROTOS TROI
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7.3STEAM EXPLOSION: CODES
MC3D
IRSN, CEA and EDF - France
JEMI/DEMO
GRS and IKE - Germany
JASMINE
JNES - Japan
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7.4STEAM EXPLOSION EFFECTS
In-vessel explosion: explosive interaction should not bring a failure of the vessel, not accounting for potential effects of thermal loads and ageing (lesson from SERENA-1 project)
Ex-vessel explosion: very high loading on reactor pit walls (20 to 50 MPa)
Low probability for direct containment failuredue to over-pressurization
Deformation of reactor pit ⇒ Moving of connectedfloors ⇒ Risk of containment damage
Stratified explosion (without premixing phase):Far less energetic (~ 1 MPa)
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7.5PERSPECTIVES
Major uncertainties to be reduced
Corium fragmentation and particularly drop size
Impact of corium oxidation on premixing phase
Pressurization processes during the explosion
Corium solidification physical processes and impacts on fragmentation (premixing and explosion)
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MOLTEN-CORE CONCRETE INTERACTION (MCCI)
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8.1MCCI phenomenologyMCCI occurs after vessel failure leading to corium pouring into reactor pit
Decay power of fission products (FPs) and heat from oxidation reactions into the corium cause concrete wall ablation
Concrete is mainly composed of: SiO2, CaCO3 and H2O
Concrete decomposition (by corium) produces:
Condensed phases: SiO2 and CaO
Gases: H2O and CO2
Thus, corium pool is composed of:
Heavy oxides from core: UO2 and ZrO2
Light oxides from concrete: SiO2 and CaO
Metals: Fe, Cr, Ni and ZrAugust 26, 2013
corium pool
concrete
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8.2MCCI EFFECTSOxidation reactions between metals and oxides would occur, producing new gas species: H2 (hydrogen hazard), CO and SiO
Basemat melt-through, due to concrete wall ablation by corium, leading to FPs transfer to environment (including into soil-water table)
Containment pressurization due to:
Gases produced during MCCI (H2O, CO2, CO, H2 and SiO)
Steam production (including vaporization of sump water in case of melt-through up to the sump)
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sumpcorium
Vesselbreach
corium/concrete interaction
containment pressurisation
Basemat melt-through
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8.3R&D STRATEGY ON MCCIMain question to be solved:
How is heat flux distributed along the pool interfaces?
Might pool stratification in oxide and metal layers occur and speed-up axial ablation?
Integral experiments (real material):
MCCI-OECD (oxide pool)
VULCANO (oxide/metal pool)
Analytical tests (simulant):
ARTEMIS
CLARA (oxide pool)
BALISE, GREENE, ABI (oxide/metal pool)
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8.4LESSON FROM EXPERIMENTS (1/2)From MCCI-OECD (oxide pool, real material)
Limestone common sand (LCS) concrete: rather isotropic ablation
Siliceous concrete: prevailing lateral ablation
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LCS concrete Siliceous concrete
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8.4LESSON FROM EXPERIMENTS (2/2)From VULCANO (oxide/metal pool, real material)
Segregation observed between oxide and metal phases, but without real stratification: stratification is in principle possible but still not demonstrated in realistic conditions
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presence of a metallic phase incurved towards the top on sides
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8.5(SOME) MCCI CODES
ASTEC/MEDICIS
Developed by IRSN (France) and GRS (Germany)
TOLBIAC
Developed by CEA (France), funded by EDF (France)
COSACO
Developed by Areva (France)
MELCOR/CORCON
Developed by SNL (USA)
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8.6PERSPECTIVES
Determine if metal/oxide stratification really occurs in the reactor case
Investigate the MCCI mitigation issue using existing or new devices involving top and/or bottom water injection
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top quenching bottom quenching
CONCLUSIVE REMARKS
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9.1 CONCLUSIVE REMARKS
The current understanding of severe accident phenomena can be summarized as follow.
In-vessel stage of the accident. There is good understanding and prediction for the early phase, and large but acceptable difference in late phase, from melt generation and relocation to vessel lower head. Quenching phenomena are not well modeled, especially in late phase. Experiments and theoretical efforts are ongoing.
Ex-vessel stage of the accident. MCCI still needs experimental programs, DCH models are not enough accurate.
FP behavior. The release from core can be considered accurate in most cases, uncertainty is still large in iodine behavior in RCS.
Containment behavior. Thermal hydraulic and aerosol models can be considered globally acceptable.
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REFERENCES
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10.1REFERENCES
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Rapport IRSN-2007/83, Rapport CEA-2007/351 “Research and development with regard to severe accidents in pressurised water reactors: Summary and outlook”, 2007
Rapport IRSN 2008-98 "Accidents graves des réacteurs à eau de production d'électricité ”, 2008
B. R. Sehgal et al. “Nuclear Safety in Light Water Reactors”, 2012
“Achievements and Status of Research Activities in the Containment Area”, L. Meyer, H. Wilkening, I. Kljenak, D. Magallon, SARNET
„Melt Dispersion and Direct Containment Heating (DCH) Experiments in the DISCO-H Test Facility“, L. Meyer et al., FZKA 6988, Forschungszentrum Karlsruhe
„Nachweis der Wirksamkeit von H2-Rekombinatoren auf der Basis ergänzender analytischer Untersuchungen mit COCOSYS für die Referenzanlage GKN-2“, S. Band et al., GRS-A-3652
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10.2REFERENCES
G. Mazzini, “Severe Accident Phenomenology Analyses and Fission Gas Release in Advanced Nuclear Reactors”, Ph. D. Final paper, University of Pisa 03/05/2012.
Luis E. Herranz, Mònica Vela-Garcìa, Joan Fontanet, Claudia Lòpez del Prà, “Experimental interpretation and code validation based on the PHEBUS-FP programme: Lessons learnt from the analysis of the containment scenario of FPT1 and FPT2 tests”, Nuclear Engineering and Design 237 (2007) 2210–2218, 22 March 2007.
F. L. Venturi, G. Mazzini, M. Mazzini, "Investigation on Thermal-Hydraulics and Core Degradation Issues of Fukushima Accident by RELAP /SCDAPSIM Code", ANS 2012 winter meeting, San Diego, Ca (US), 2012.
SARNET2/ENENIII, “Materials of Short Course on Severe Accident analyses Phenomenology”, January 2011.
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10.3REFERENCES
NUREG/CR-5642 “Light Water Reactor Lower Head Failure Analysis”, EGG-2618, 1993
NUREG/CR-5582 “Lower Head Failure Experiments and Analyses”, SAND98-2047,Sandia National Laboratories, Albuquerque, NM, 1998
NEA/CSNI/R(2002)27 “OECD Lower Head Failure Project (1999-2002) Final Project Report OECD/NRC/NERI Performed at Sandia National Laboratories”, 2002
Theofanous T. G. et al., “Lower head integrity under steam explosion loads”, Nuclear Engineering and Design, Vol. 189, Issue 1, 1999
Saito M. et al., “Melting attack of solid plates by high temperature liquid jet –effect of crust formation”, Nuclear Engineering and Design, Vol. 121, Issues 1–3, 1990
B. R. Sehgal et al. "Assessment of reactor vessel integrity (ARVI)", Nuclear Engineering and Design, Volume 221, Issues 1–3, 2003
B. R. Sehgal et al. "Assessment of reactor vessel integrity (ARVI)", Nuclear Engineering and Design, Volume 235, Issues 2–4, 2005
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