An Approach to Evaluation of Uncertainties in Level 2 PSAs T. Ishigami, J. Ishikawa, K. Shintani, M. Mayumi and K. Muramatsu Japan Atomic Energy Research Institute OECD/NEA/CSNI/WGRISK Workshop, Cologne, March 29 - 31, 2004

Introduction Background PSA application study at JAERI on safety goals, emergency planning, basic technical study for legal system for compensation of nuclear damage, and so on The first phase PSA at JAERI in 1990 did not address AM, source terms for energetic events, and uncertainty Uncertainty is one of the most important issues in PSA application Purpose of This Study To develop an uncertainty analysis method for Level 2 PSA JAERI

Study on Uncertainty Evaluation at JAERI Type of Uncertainties Parameter uncertainty Model uncertainty Completeness uncertainty 　Parameter and model uncertainties are addressed in this study Study on Uncertainty Evaluation at JAERI Development and improvement of computer codes Assessment of uncertain parameters Development of uncertainty analysis method for source terms JAERI

Framework of uncertainty analysis for Level 2 PSA Analysis step Uncertain parameters Core damage frequency analysis ・Component failure rates etc. Resources ・Existing PSA 　　　　results ・Analysis method 　　(DET, ROAAM) ・Experiment 　- Steam explosion　- Release of fission 　 products from 　　 fuel SAPHIRE Accident progression analysis Uncertainty analysis ・CET branch 　　 　probabilities etc. PREP/SPOP CET ・Release rates of 　　　fission products 　　from fuel 　etc. Source term　analysis THALES2 Frequency of exceeding X 95% 50% Computer codes 　　SAPHIRE: System analysis 　　CET: Containment ET analysis 　　THALES2: Sever accident analysis 　　PREP/SPOP:Uncertainty propagation analysis 5% Source term, X Conceptual figure of uncertainty analysis results JAERI

Computer Codes SAPHIRE (USNRC) CET Analysis Code (JAERI) - Analysis of core damage frequency with ET/FT model - Capability of uncertainty analysis CET Analysis Code (JAERI) - Analysis of containment function failure probability - Object-oriented programming THALES2 (JAERI) Integrated severe accident analysis code Analysis of thermal hydraulics and fission product transport PREP/SPOP (JRC Ispra) - Analysis of parameter uncertainty propagation through a model - Monte Carlo or Latin Hypercube sampling method JAERI

Assessment of Uncertain Parameters - Core Damage Frequency and Accident Progression Analyses - Approach Component failure rates Use of the failure rates evaluated by Central Research Institute of Electric Power Industry from operation data of 49 Japanese plants for 16 years (1982-1997) CET branch probabilities for CV failure modes Overpressure Overtemperature Steam explosion Direct containment heating Hydrogen burning Survey of recent experimental and analytical research results Use of analytical methods such as DET and ROAAM - Probability of containment failure due to energetic events JAERI

Example of Assessment of Uncertain Parameter - Probability of Containment Failure due to Ex-Vessel Steam Explosion at PWR - Possible phenomena Steam explosion at cavity Load energy brought into containment Cavity wall failure Loss of containment integrity at penetration Piping tenseness Movement of reactor vessel “Load energy > Critical load energy ” 　　“Loss of containment integrity” Preceding analysis for Japanese APWR (Nuclear Safety Research Association) Point estimate with DET method Survey of research results Structural analysis JAERI

Present Analysis for 4-loop PWR Approach Use of the DET for APWR Similar physical processes at both plants Type of containment is PCCV at both plants Reevaluation of physical quantities depending on plants Flow rate of molten core at large scale RV failure Critical load energy resulting in containment failure → These quantities were scaled according to reactor powers of 　　 the two plants (APWR: 4,451MWt, PWR: 3,411MWt) New Feature Evaluation of uncertainty in containment failure probability caused by uncertainties in Branch probabilities of “reactor vessel failure mode (large/small)” and “Occurrence of triggering (Yes/No)”, and Aleatory and epistemic uncertainties were addressed JAERI

Decomposition Event Tree RV Failure Mode Melt Flow Rate Internal Energy Triggering Mass in Premixture Conversion Load Probability Small Scale Large Low High No Yes Medium E1 E2 E3 En ・・・・・・・・・・ Lower Higher JAERI

Aleatory and Epistemic Uncertainties Aleatory uncertainty : Randomness or stochastic properties Epistemic uncertainty : Lack of our knowledge, Possible to reduce Treatment in present analysis Uncertainties in the branch probabilities : Epistemic Uncertainty in the critical load energy (capacity) of the containment: Aleatory and epistemic Epistemic Critical load energy PDF Aleatory JAERI

Cumulative probability of Probability Distribution of Load Energy and Containment Failure Probability Load energy (MJ) 0.0 0.2 0.4 0.6 0.8 1.0 200 400 600 800 Cumulative probability of containment failure 5% 50% 95% Load energy (MJ) Aleatory Epistemic Scenarios with higher load energy caused from larger size of RV failure contribute to containment failure JAERI

Uncertainty Analysis Result of Containment Failure Probability (conditional probability) 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.E-07 1.E-06 1.E-05 1.E-04 1.E-03 1.E-02 1.E-01 1.E+00 Containment failure probability Cumulative probability Aleatory Epistemic (Dotted line) Aleatory+Epistemic (Solid line) Uncertainty range : 0 – 0.006 (95th) Epistemic uncertainty is dominant JAERI

Uncertainty Analysis Method for Source Terms Subject Approach A large number of accident sequences Uncertainty propagation Parameter uncertainty Model uncertainty Classification of sequences Direct use of THALES2, instead of parametric model (XSOR in NUREG-1150), with Monte Carlo simulation Survey of recent experimental and analytical research results Comparison of the model results with experimental data or different model results Repeated calculation X1 X2 Code (THALES2) ・・・ Y Input Parameters Output JAERI

Classification of sequences for uncertainty analysis (BWR Mark-II) Sequence group PRV Pressure Sequences subgroup Loss of containment heat removal function with high pressure coolant injection available High ～ Medium Transient or SB-LOCA Medium ～ Low MB- ～ LB-LOCA Low Transient with reactor depressurized by failure of SRV reclose Loss of containment heat removal function with low pressure coolant injection available Transient or SB-　～　LB-LOCA Loss of coolant injection Station blackout Anticipated transient without scram (ATWS) JAERI

Uncertain Parameters in THALES2 and Parametric Model Uncertain parameters is denoted by ＊ THALES2 ＊ ＊ ＊ ＊ -Failure pressure Input Heat transfer coefficient Release rate coefficient Deposition rate ・・ ・・・ ・・・ ・・・ -Size of rupture Thermal-Hydraulic behavior Release from fuel FP behavior in RV ・・・ Environmental release Model 　　　　Data Calculated Release fraction from fuel Pressure Temperature Release fraction from RV Release fraction to the environment Parametric Model ＊ ＊ Release fraction from fuel Release fraction from RV Release fraction to the environment × ・・・ ＝ JAERI

Comparison of the two Methods Direct Use of Code Parametric Model - Rather fundamental 　　(Some are related to 　　experimental data) - Not fundamental (To 　　be obtained from a 　　　model) Characteristics of uncertain parameters - Less dependent on 　　time or sequence 　　 - Dependent on time 　　　and sequence 　 - Individual representative 　sequence in a set of 　　　sequences classified by 　state of safety systems - Possible to compare the 　results with different 　　model results - a set of sequences classified by physical state (Zr-oxidation level, RCS pressure) - Difficult to assess the results Accident sequences to be analyzed Calculation time - Long - Short JAERI

Approach to Assessment of Uncertain Parameters Survey preceding PSAs (NUREG-1150) and comparative study of THALES2 and MELCOR to determine uncertain parameters Select parameters in THALES2 relating to the above uncertain parameters - Representative parameters are selected to reduce the number of uncertain parameters e.g. One correction factor for deposition rate in RCS for all 　　　 the deposition mechanisms Determine the uncertainties by surveying recent experimental and analytical research results as well as preceding PSA study - Experiments on FCI (FARO, COTELS, …) - Experiment on FP release from fuel (VEGA,…) etc. JAERI

Associated parameters An example of uncertain items and associated parameters in THALES2 (BWR Mark-II) Uncertain items Associated parameters Release of FPs from Fuel (in-vessel) Release rate of FPs Coolability of molten core (in-vessel) Heat transfer coefficient between molten core and coolant/or lower head wall Fraction of fragmented molten core in FCI Average particle size of molten core droplets Deposition of FPs (in-vessel) Deposition rate of FPs to the wall Deposition rate of FPs to the floor Deposition of FPs (ex-vessel) Release of FPs from molten core (ex-vessel) - Release rate of FPs Coolability of molten core (ex-vessel) - Heat transfer coefficient between molten core and coolant Integrity of containment Failure pressure by overpressure Size of rapture Pool scrubbing Size of bubble Rising velocity of bubble LOCA - Break size in LOCA JAERI

Summary Study on developing uncertainty analysis method for Level 2 PSA at JAERI includes Development and improvement of computer codes Assessment of uncertain parameters Development of uncertainty analysis method for source terms To quantify uncertain parameters - Recent experimental and analytical research results are surveyed - Analytical method such as DET is used to evaluate uncertainty in 　　　containment failure probability due to steam explosion, where 　　　　　　aleatory and epistemic uncertainties are considered Uncertainty analysis method for source terms is 　　- Direct use of THALES2 with Monte Carlo simulation, where - Uncertain parameters in THLES2 are assessed by surveying recent 　　research results JAERI