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www.cvrez.cz Introduction to Severe Accident (SA) Phenomenology Dr. Ing. G. Mazzini, Ph. D. in Nuclear and Industrial Safety Prague, 10-11/10/2012
www.cvrez.cz
Introduction to Severe Accident (SA)
Phenomenology
Dr. Ing. G. Mazzini,Ph. D. in Nuclear and Industrial Safety
Prague, 10-11/10/2012
Outlooks (1/2)
• Mile-stones of Nuclear Safety
• Safety Functions and Defense in Depth
• Introduction to SA
• Definitions of “Beyond DBA” and a SA
2
Outlooks (2/2)
• SA Sequences
• The Most Important Nuclear Accidents
• Severe Accident Phenomena
• Example: UNIPI – Fukushima Unit 3 Analysis
3
Mile-stones of Nuclear Safety
4
Year EVENT
1954 Pubblication of Nuclear Energy Act in USA
1957 Fondation of EURATOM and IAEA Safety Metodology
USAEC of Dephense in Depth
1967 Farmer Safety Methodology, based on PSA
1974 Publication of draft WASH 1400 (Rasmussen Report)
1979 TMI 2 Accident– Research focus on Severe Accidents
1986 Chernobyl Accident
1987INSAG 3 (now INSAG 12) “Basic Safety Principles for
Nuclear Power Plants” publication by IAEA
2011 Fukushima Accident (1st for an External Event)
The Safety Functions
5
Removal of the
Decay Heat and
Full Power
Reactivity
Control under
Accident and
Operating
Conditions
Confinement of
Radioactive
Elements inside
the NPP
Reactor Safe
and Under
Control
LWR Safety Functions (WASH-1400)
6
The Defense in Depth
7
• “Defense-in-depth”
• 5 subsequent safety levels,
• including 4 physical barriers
• Accidents Prevention
• Accidents Mitigation
Introduction to SA
• SA phenomenology includes very complex
phenomena:
– Mechanics of core degradation and relocation
– Thermodynamics
– Chemistry
• Before to speak of the phenomenology, we must
understand:
– What is a Severe Accident?
– When is there a Severe Accident?8
IAEA Severe Accident Definition
• A beyond design basis accident comprises accident
conditions more severe than a design basis accident, and
may or may not involve core degradation.
• Accident conditions more severe than a design basis
accident and involving significant core degradation are
termed severe accidents.
• http://www-
pub.iaea.org/MTCD/publications/PDF/Pub1376_web.pdf9
NRC Severe Accident Definition
• Beyond design basis accident: This term is used as a
technical way to discuss accident sequences that are
possible but were not fully considered in the design process
because they were judged to be too unlikely.
• Severe accident: A type of accident that may challenge
safety systems at a level much higher than expected.
• http://www.nrc.gov/reading-rm/basic-ref/glossary/severe-
accident.html10
IRSN Severe Accident Definition• The term “severe accident” refers to an event causing
significant damage to reactor fuel and resulting from more
or less complete core meltdown.
• Such accidents are highly unlikely in light of the preventive
measures implemented by operators.
• However, they are the focus of considerable research
because the release of radioactive products into the
environment could have serious consequences.
• http://www.irsn.fr/EN/Research/publications-
documentation/Publications/DSR/SAGR/Documents/rappor
t_RetD_AG_VA.pdf11
SUJB Severe Accident Definition?• Actually the 195/99 regulatory guide has not a definition of
SA . It is under definition (draft):
– nadprojektovou nehodou havarijní podmínky, při kterých
jsou překročeny podmínky přijatelnosti pro projektové
nehody, (beyond design accident)
– těžkou havárií nadprojektová nehoda, při které došlo k
vážnému poškození a nezvratné ztrátě struktury aktivní
zóny reaktoru, nebo palivových souborů převážně v
důsledku tavení jaderného paliva, a která může vést k
radiační nehodě, nebo radiační havárii, (severe accident)
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Which are the Differences?
13Chernobyl Accident
Fukushima-Daiichi Accidents
• For Light Water Reactors (LWRs) we can distinguish severe accident initiators into two broad classes:
–Reactivity Insertion Accidents (RIAs)–Core Uncovery Accidents (CUAs)
14
In Vessel Accident Sequences and Phenomena
• In the case of TMI, core uncovery resulted from a stuck open
valve, PORV, and the operator's interference with the high
pressure safety injection.
• In the case of Chernobyl the reactivity insertion was due to a design error (initial positive reactivity insertion by the control rods at scram) and was exacerbated by the positive void reactivity of the reactor.
• In the case of Fukushima Daiichi, the facilities were damaged by a magnitude 9 earthquake and mainly by serious secondary effects followed as the tsunami. This generated a SBO sequence with the stop of about all ECCS and loss of final heat sink. 15
The Most Important Nuclear Accidents
List of the Most Important Accidents• Level 7: Chernobyl NPP, USSR (now in
Ukraine), 1986 / Fukushima Daiichi
(cumulative effect), Japan, 2011
• Level 6: Kyshtym Reprocessing
Plant, USSR (now in Russia), 1957
• Level 5: Windscale Pile, UK, 1957 /
Three Mile Island, NPP, USA, 1979 /
Fukushima Daiichi (single NPP), Japan,
2011
• Level 4: Windscale Reprocessing Plant,
UK, 1973 / Saint-Laurent NPP, France,
1980 / Buenos Aires Critical Assembly,
Argentina, 1983
• Level 3: Vandellos NPP, Spain, 1989 16
17
Severe Accident Phenomena
Melting and Release Phenomena (1/2)
Melting and Release Phenomena (1/2)
Example: UNIPI – Fukushima Unit 3 Analysis
• Based on: F. L. Venturi, G. Mazzini, M. Mazzini, “INVESTIGATION ON THERMAL-HYDRAULICS AND CORE DEGRADATION ISSUES OF FUKUSHIMA ACCIDENT BY RELAP5-SCDAP CODE”, 2012 ANS Winter Meeting and Nuclear Technology Expo,
Dan Diego, CA, November 11-15, 2012
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Plant description
� BWR- 4 + Mark I containment� Reactor Thermal Power: 2317 MWt� In commercial operation since 1976
21
• On March 11, 2011 at 14:46 (T=0), a 9 Richter scale earthquake
caused reactor scram
• By 15:38 tsunami caused flooding of auxiliary and turbine buildings,
resulting in gradual loss of AC and DC power
• Both High Pressure Coolant Injection (HPCI) and Reactor Core
Isolation Cooling (RCIC) remained available (due to steam-driven
pumps) for water injection into the reactor. RCIC operated regularly.
• At T=78000 s (an hour after RCIC trip), HPCI started
• At T=129000 s HPCI system tripped
• At T=151500 s Safety Relief Valve (SRV) was open and depressurized
the reactor
• Start water injection by fire extinguishing line
• Off-site power was restored only on March 2222
Unit-3 accident description
RELAP5\SCDAP Code
• Version (mod3.5.dg) is developed by Innovative Systems Software (ISS) Group, directed by Prof. Chris Allison;
• is used to simulate an operational/accident transient in nuclear power plants;
• solves energy/momentum/mass balance equations;
• includes many generic component models by which general systems can be simulated (pumps, valves, pipes, heat releasing or absorbing structures, reactor point kinetics, electric heaters, jet pumps,turbines, separators, accumulators, and control system components);
• with SCDAP models, is able to follow the progression of a SA scenario (melting, relocation, hydrogen production and molten pool formation);
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Nodalization of Fukushima Unit 3
Power
MTR Valve
Branch
Emergency Systems
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Results Unit-3Reactor
Pressure
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Results Unit-3 Core Level
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Results Unit-3Hydrogen
Production
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Core degradation map
151.000 s
160.000 s
163.000 s
170.000 s
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Conclusions (1/2)
• The nodalization model has been improved with some changes, in particular:
• Start with appropriate Steady-State conditions
• The simulation of the safety features involved in the accident
• Various modifications in order to reduce the mass error during the very long simulation of the sequences.
• RELAP/SCDAP is able and useful to analyze SA sequences (Unit-3)
• It can simulate the trends of the physical quantities involved in the Fukushima accident, even through some discrepancies are shown, due to accident management.
• The results are interesting in comparison with the (incomplete) data obtained from the reference reports and to be improved if new data will be available in the future.
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Conclusions (2/2)
Observations from Fukushima sequences:
• The probability of occurrence of serious earthquakes and
tsunamis should be reconsidered
• The emergency systems, components and structures in present
NPP must be improved; in particular attention shall be given to
the rapid depressurization of the Reactor coolant circuit and to
systems for coolant replacement like the Fire Extinguishing Line.
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31
Thanks for Dr. Ing. F. L. Venturi, Prof. Dr. Ing. M.
Mazzini and Prof. Dr. Ing. F. Oriolo
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Appendix
• LEVEL 1 PSA: Evaluation of the Core Damage
Frequency (CDF)
• LEVEL 2 PSA: Evaluation of the Spectrum of Releases in
Function of Cumulative Complementary Distribution
Frequency (CCDF)
• LEVEL 3 PSA: Evaluation of the Spectrum of
Consequences on Population and of Economic
Damages in Function of CCDF
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Definition of PSA Levels
• The Core Damage Frequency (CDF) requests a definition of
Core Damage.
• The Core Damage criteria are:
– If the active part of reactor core remains covered during the whole
accident phase, it may be assured that no core damage occurs.
– If the fuel cladding peak temperature does not exceed 1204°C, it
may be assumed that there aren’t fuel damages.
– If peak enthalpy deposited in the fuel pellets does not exceed 837
KJ/Kg, it may be assumed that there aren’t core damages.
• Core Damage criteria compatible with IAEA definition of
safety targets.34
Core Damage Frequency Definition (From EUR)
• Core Damage is prevented if the core remain covered.
In practice, severe Core Damage is not expected to
occur until water level decreases significantly lower
than top of active core level for a sustained period of
time.
• For cladding temperatures higher that 1300°C, the
Zircaloy oxidation rate increases rapidly, leading to a
drastic acceleration of core degradation.
• With a rather large margin, a critical temperature of
1204°C (2200°F), according to the actual Emergency
Core Cooling System (ECCS) design criteria, prevents
excessive clad oxidation. 35
Core Damage Frequency Definition (From EUR)
IAEA, EUR and NRC Probabilistic Safety Assessment
(PSA) Requirements
• International Atomic Energy Agency goals:
– <10-04/yr for Core Damage Frequency (CDF)
– <10-05/yr for Large Release Frequency (LRF)
• Nuclear Regulatory Commission goals:
– <10-04/yr for Core Damage Frequency (CDF)
– <10-06/yr for Large Release Frequency (LRF)
• European Utility Requirements:
– <10-05/yr for Core Damage Frequency (CDF)
– <10-06/yr for Large Release Frequency (LRF) 36
• At TMI, the situation was turned around by activating the safety injection and reflooding the core. The extent of damage as surmised from core bore samples and direct video inspections is illustrated in the following Figure. It is significantly more extensive than expected and predicted by prior analyses; in particular, the quantity of debris mass found in the lower plenum area is estimated at 10 to 20 tons. This accident also demonstrated the significance of the hydrogen production and burning phenomena.
• A wide variety of other types of small leaks with failure or disabling of the emergency cooling systems would lead to core uncovery and TMI-like behaviour;
• that is, severe core damage at elevated primary system pressure. 37
In Vessel Accident Sequences and Phenomena
(TMI 2 Accident)
TMI 2 Vessel Status after the Accident Sequence
38
• Blow-down: Loss of coolant through break or relief valve. May be rapid or slow and occur at high or low pressure.
• Boil-off: Liquid level gradually drops as decay heat vaporizes water and loss of coolant occurs through break or relief valve.
• Heat-up: Core becomes uncovered. – Zircalloy steam reaction produces hydrogen and heat.
– Claddings fail,
– Release of volatile Fission Products (FP) which partially deposit in primary circuit and/or escape to the containment with steam/H2.
39
In Vessel Accident Sequences and Phenomena
• Core Uncovery: The water mixture level in the reactor vessel falls below the top of the active fuel.
• Core Damage: The fuel assemblies are deformed by mechanical fracturing, or by liquefaction due to material interactions or by melting.
• Ballooning: Process connected with the change of geometry of the cladding layer, due to overpressure in the fuel element, leading to local or extended swelling of the Zry cladding tube, with consequent reduction in the section of the flow channel and change of its thermal-hydraulic characteristics; it also determines the size of the gap between the cladding and the fuel pellet and then influences the UO2-ZrO2 interaction and the kinetics of the internal oxidation. It interests also the control rod cladding tubes.. 40
In Vessel Accident Phenomena
• Early Phase: It refers to the initial stages of core damage, including cladding oxidation and the melting and relocation of mainly metallic materials of fuel bundles.
• Core Melt: The reactor core overheats and this leads to significant melting or liquefaction of the core materials.
• Degraded Core: An advanced state of core damage in which the original fuel bundle geometry has been substantially lost, due to molten material relocation.
• Late Phase: It refers to the stages of core degradation involving substantial melting and relocation of fuel materials including ceramics, comprising the displacement of materials to the vessel lower plenum and the containment, if the vessel bottom head failure occurs.
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In Vessel Accident Phenomena
• Core Melt/Slump: Bulk of core is uncovered; core melt drips/fragments and begins to fall into water pool in vessel bottom. Fission product migration through primary circuit and to the containment can be significant for more volatile fission products.
• Small scale steam explosions in vessel are probable, and large ones unlikely but possible.
• Melt through: Pressure vessel fails as molten core melts through the bottom head and drops into reactor cavity.
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In Vessel Accident Phenomena
• Core quench: The molten debris falls into the reactor cavity. This produces boiling off of whatever water is there. If sufficient water is present, the core would be cooled and solidify, and would subsequently re-melt.
• Core dispersion: Steam explosions may also occur at this time as the molten fuel falls onto water in the reactor cavity. This can disperse fuel, create radioactive aerosols, and increase the rate of heat transfer to the containment atmosphere.
• Molten Core-Concrete Interaction: The molten corium attacks the concrete basement, which proceeds to penetrate, and creates a cauldron that releases aerosols and gases; as a consequence containment pressure rises.
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Ex Vessel Accident Phenomena
