Supporting Analysis for Identification of Plant Vulnerabilities

IAEA Workshop on the Development of Severe Accident Management Guidelines

11-15 December 2017, Vienna, Austria

presented by Yasunori YAMANAKA (NRRC of CRIEPI)

Outline

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• Accident analysis– Safety assessment and analysis– Types of accident analysis– Accident analysis methods– Computer codes for accident analysis– Fission Product challenges

Safety assessment

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• Systematic process carried out during plant operation to ensure requirements are met– Provides for defense in depth– Plant equipment requirements– Plant system design requirements

• Types of analysis– Design analysis– Validation of Emergency Operating Procedures

and Plant Simulators– Probabilistic Safety Assessment– Support for Accident Management & Emergency

Planning

Safety assessment (cont.)

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• Accident analysis methods– Conservative analysis

• Goal to demonstrate safety margin– Best Estimate– Sensitivity and Uncertainty

• Codes– Integral codes

• MAAP• MELCOR• ASTEC

– Detailed codes• ATHLET-CD (TH)• ICARE/CATHARE (PWR Severe Accident)• SCDAP/RELAP5 (Severe Accident)• COCOSYS (Severe Accidents)

MAAP – Modular Accident Analysis Program

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• A severe accident code, developed after Three Mile Island (TMI) to better understand severe accident mechanisms– Fast running and flexible code– Based on 35+ years of simulation

experiments, modeling and plant severe accidents

• Used worldwide to enhance plant safety– Advanced designs– Operations and maintenance– Uprates and life extension

• Predict timing of key events during accidents using high power computing– Evaluate effects of operator actions– Predict magnitude and timing of fission

product release

MAAP is developed and maintained by EPRI and users are required to obtain a license from EPRI

BrazilCanada

ChinaFrance

HungaryJapan

MexicoNetherlands

RomaniaSlovenia

South AfricaSouth Korea

SpainSwedenTaiwan

United Arab EmiratesUnited Kingdom

United States

MAAP is Used Worldwide

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• MAAP is used in 18 countries, more than 75 organizations active in the MAAP User Group– Power plant operators– Manufacturers– Universities, research organizations, and

regulators– Nuclear plant simulator companies

MAAP membership has nearly doubled since Fukushima

MAAP Provides Support to Advanced Reactor Development and Licensing–Examples

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GEH ABWR

MHI APWR

B&W mPowerGEH ESBWR

WEC AP-1000

AREVA US EPR

TOSHIBA US-ABWR

KHNP APR-1400

VVER*

CANDU

Applications of the MAAP Code

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• Probabilistic Risk Assessment (PRA)

• License Renewal/Power Uprates

• Design & Design Certification for Advanced Light Water Reactors

• Severe Accident Guidelines (1992)

MAAP Collaboration

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Post-Fukushima Enhancement of the MAAP Code

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• Japanese MAAP5 Enhancement Project– BWR thermal

hydraulics– BWR core melt

progression– BWR lower plenum

model– Lower head

penetration modeling– Molten Core-

Concrete Interaction (MCCI)

– Containment stratification model

• EPRI funded MAAP Enhancements– MAAP validation– MAAP benchmarking– SFP modeling

updates– Generic filtered vent

models– Aerosol modeling– Hydrogen modeling

improvements– Application

Programming Interface (API)

Strong MAAP5 collaboration expands capabilities for all MAAP users

MAAP Code

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• An essential tool to enhance plant safety– Advanced designs,

operations, and life extension

• Strong collaboration– Broad applications

worldwide– Technical credibility– Japanese government funds

MAAP development• Continuing growth

MELCOR

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• MELCOR has been developed at Sandia National Laboratories for the USNRC– Started in 1982 (ongoing development of new capabilities)

• Major pieces of MELCOR referred to as “Packages”– Hydrodynamics, heat and mass transfer to structures, gas

combustion, aerosol and vapor physics– Decay heat generation, core degradation, ex-vessel phenomena

(e.g., core concrete interactions), sprays, fission product transport

– Thermodynamics, equation of state, material properties, data-handling utilities, equation solvers

• MELCOR modeling approach– T-H modeling is simple/fast-running for PRA applications– Uncertainties through sensitivity studies (substantial user

flexibility)– MELCOR is a state-of-the-art tool for source term calculations

• Most modeling is mechanistic, sometimes simplified• Evolving as a repository of our knowledge of severe accident

phenomenology

What is MELCOR?

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ASTEC

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• ASTEC has been developed jointly by IRSN and GRS – Started in 1996 (ongoing development of new capabilities)– European reference software within the EC SARNET network of

excellence.– Used in IRSN Level 2 PSA for 900 and 1300 MWe PWR and for EPR– For present/future Water-Cooled Reactors (PWR, VVER, BWR,

CANDU)

• Covers the entire phenomenology of severe accidents except– Steam Explosion, gas explosion and containment mechanical

integrity

• ASTEC modeling approach– Modeling is simple and fast-running as reasonable– Uncertainties to be dealt with through sensitivity studies

(substantial user flexibility)– ASTEC is a state-of-the-art tool for source term calculations

• Most modeling is mechanistic, only sometimes simplified,• Repository of knowledge of severe accident phenomenology

Ref : ASTEC - European severe accident integral code (Presented by Jean-Michel Bonnet, at JAEA, March,2012

ASTEC (Cont’d)

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ASTEC used for ST evaluations, in support to L2 PSA and emergency response tools

Code main features • Complete accident sequence

calculation from initiating event to releases (<12 h CPU)

• Safety systems considered • Validation and update of models

using SOA knowledge from R&DSpecific to releases (ST) calculations • Consideration of containment

failure modes, leak-paths and filtered venting (DFs for solid filter, scrubbing module for liquid filter)

• Modules dealing with FP release from fuel, transport in the RCS and behaviour in the containment – detailed modelling of physico-chemical processes, including dose effects

Ref : ASTEC - European severe accident integral code (Presented by Jean-Michel Bonnet, at JAEA, March,2012

Fission Products

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• Types of releases possible– Noble gases – Xe, Kr– Iodine – I2– Iodides – CsI, CH3I– Oxides – BaO, SrO, TeO2– Hydroxides – CsOH– Metals - Sb

Fission Products (cont’d)

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• Isotopes can be:– Gaseous – I2, CH3I– Volatile – Cs, I– Medium volatile – Ba, Ce– Non-volatile – Sr, Ru

• Fission products released from fuel matrix condense in cooler regions to form solid aerosols

Ref: NUREG/CR-6533, Code Manual for CONTAIN 2.0, 1997

Potential for Large Releases

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• Initiating event (examples)– Reactivity event– Large external event– Aircraft impact

• Accident progression– Steam explosions– Hydrogen deflagration/detonation– Direct Containment Heating

Large releases require immediate attention to protect staff and public

Large Releases Strategies

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• Habitability challenges• Desire to locate the breach and isolate

– Use of internal and external sprays– External filters– Operation of ventilation systems

• Requires close integration with Emergency Preparedness (EP) plan

Containment Bypass

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• Possible confusion with containment impairment or pre-existing failure of containment– Impairment and pre-existing failure

• Opening in containment occurs prior to core damage

• Bypass – normally refers to the creation of a direct release path from the RCS/RPV to the environment– Induced SGTR (post-core damage)– Interfacing system LOCA (pre-core damage)

High Pressure Melt Ejection

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• Requires elevated RCS/RPV pressure (e.g. > 2 Mpa)

• Spread of molten debris over large containment volume

• Debris stored heat transferred to containment atmosphere

• Short time scale

Ref: NUREG/CR-6533, Code Manual for CONTAIN 2.0, 1997

Direct Containment Heating

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• As a result of HPME, rapid heat-up of the containment atmosphere could occur with a consequential pressure spike

• DCH important factors:– Cavity geometry– De-entrainment of debris– Re-entrainment of debris

• HPME can also impact the final debris distribution and success of debris cooling

Hydrogen Generation

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• Steam oxidation of zirconium fuel cladding

• CO also generated ex-vessel due to core-concrete interactions

• Hydrogen flammable– Ignition at 4%– Flame acceleration at 8%– Deflagration-to-Detonation Transition at

14%• Steam inerting at 55%

Ref. www.world-nucler.org

Molten Core Concrete Interaction

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• Ex-vessel challenge– Basemat erosion– Sidewall erosion– H2, CO, CO2

• Occurs in dry cavity conditions– No debris cooling

• Wet cavity– May still occur for deep core

debris pools (e.g. > 10 cm)Ref: EPRI Technical Basis Report, 2012

Containment Pressurization

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• Sources of mass and energy– Accident initiator – LOCA– Discharge from RCS prior to core

damage – SRVs– Heat from reactor vessel– Steam generation ex-vessel– H2, CO, CO2 due to MCCI– H2 and CO combustion or recombination– Direct Containment Heating– Containment flooding (reduces gas

volume)

Containment Capability

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• Typical PSA includes structural analysis of the containment– Considers several potential failure locations– Includes both pressure and temperature

challenges– Looks at static and dynamic loads– Addresses penetrations and seals in addition

to structural components

Containment capability assessment is critical to planning AM strategies

Release of Fission Products to Environment

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Ref: The National Diet of Japan, The official report of The Fukushima Nuclear Accident Independent Investigation Commission, 2012

Release Consequences

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• Habitability constraints– Containment leakage– Bypass– Failure– Venting– Steam generator tube rupture– Isolation condenser tube failure– Drywell liner failure– Basemat failure– Spent fuel pool release

Spent Fuel Pool Challenges

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• Damage due to:– Initiating event (e.g. seismic event)

• Pool drain can create rapidly developing challenge– Loss of pool cooling

• Slower evolving challenge due to heat-up and boil-off

• Typically Spent Fuel Pool not inside containment, therefore, potential for unscrubbed release

QUESTIONS?

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ContactYasunori YAMANAKA+81 70-5461-9121

yasunori@criepi.denken.or.jp