Uncertainty Analysis on Containment Failure Frequency for a Japanese PWR Plant O. KAWABATA Environmental Safety Analysis Group Safety Analysis and Evaluation Division, Japan Nuclear Energy Safety Organization (JNES), Kamiya-cho MT BLDG, 4-3-20, Toranomon, Minato-ku, Tokyo, 105-0001 Japan Presented in Uncertainty Workshop of OECD/CSNI , November 7-9 in 2005.
Contents 1. Introduction 2. Outline of the Reference PWR Plant 3. Level 1 - Level 2 Interface 4. Construction of Containment Event Trees 5. Uncertainty Analysis of Containment Failure Frequency 6. Source Terms Uncertainty Analysis 7. Conclusion
1. Introduction 1. Introduction (1) With a primary objective of estimating containment performance, the level 2 PSA by uncertainty estimate was executed for a typical Japanese 1,100 MWe PWR plant. (2) In the level 2 PSA, it is necessary to estimate the phenomenological uncertainty associated with phenomena such as steam explosions, direct containment heating, and debris cooling. (3) The evaluation methodology of probability distributions by the ROAAM method applying experiment results for simulated severe accident phenomena. (4) Quantification of Containment Event Trees was carried out considering the phenomenological probability distributions. 1
2. Outline of the Reference PWR Plant 2. Outline of the Reference PWR Plant Thermal Power 3,411 MWt Containment Design Pressure 493kPa Free Volume 73,700m 3 High Pressure Safety Injection System : 2 Train Low Pressure Safety Injection System : 2 Train Containment Spray System : 2 Train Auxiliary feed water : 3 pumps Re-circulation mode change of ECCS : Automatic Component cooling water system : Train isolation with motor-operated valves 4-Loop PWR with a Pre-stressed Concrete Containment 2
Accident Management Counter-measures Water Injection into CV Containment Vessel Raw Fire P Water Containment cooling CV Spray Ring Tank by Natural Convection Cooling Coil CCWS Forced Depressurization Refueling of the RCS Alternative Water Recirculation Storage Cooling down Pit MSRV Pressurizer CV Spray P SG Relief Valve Turbine Recirculation RV RHRP HPIP 3
Uncertainty Analysis Flow of Containment Failure Frequency Containment Failure Typical Core Damage Sequence Sequence Sequence Quantification PDS CET Quantification PDS CET - Similarity of Accident - Branch Probability - Heading Progression - ROAAM method - Containment - System Unavailability Failure Mode 1.0 0.8 Cumulative Probability Containment 0.6 Failure Modes 0.4 0.2 0.0 10 -12 10 -11 10 -10 10 -9 10 -8 10 -7 10 -6 4 Frequency (/RY)
3. Level 1 - Level 2 Interface 3. Level 1 - Level 2 Interface PDSs of a Japanese PWR Plant AE :Large&Medium LOCA/Early Core Damage /Without CV Spray AEC:Large&Medium LOCA/Early Core Damage PDS Average 5% 50% 95% /With CV Spray AL :Large&Medium LOCA/Late Core Damage TEC 3.9E-08 8.3E-10 8.1E-09 1.4E-07 /Without CV Spray SL 1.3E-08 4.2E-10 4.0E-09 5.0E-08 ALC:Large&Medium LOCA/Late Core damage /With CV Spray V 8.0E-09 3.0E-10 3.3E-09 3.0E-08 SE :Small LOCA/ Early Core Damage Total 8.8E-08 2.5E-09 2.4E-08 3.2E-07 /Without CV Spray SE’ :SBO/RC Pump Seal LOCA SE” :CCWS Failure/RC Pump Seal LOCA SEC:Small LOCA/Early Core Damage Using the WinNUPRA code, 1,000 sampling calculation /With CV Spray was performed for minimal cut sets of each PDS. SL :Small LOCA/Late Core Damage The core damage frequency as an average value was /Without CV Spray obtained to be 8.8x10 -8 /RY. SLC:Small LOCA/Late Core Damage /With CV Spray The uncertainty width of total core damage frequency TE :Transient/Early Core Damage /Without CV Spray based 95% value and 5% value was obtained to be TE’ :SBO double figures. TEC:Transient/Early Core Damage/With CV Spray G :SGTR P :Containment Failure before Core Damage V :Interface-System LOCA 5
Core Damage Frequencies with AM (4 Loop Plant) SEC TE TE' P SE" AE SE' SE 0.7% 0.2% 0.2% 0.5% 0.4% 0.3% 0.8% 1.4% G 2.2% AL 4.9% ALC 5.0% AEC 8.0% TEC 44% SLC 8.1% V The contribution fraction of TEC 9.1% in PDSs was the highest and SL 15% estimated to be about 44%. Average Core Damage Frequency 8.8E-08/RY 6
4. Construction of Containment Event Trees 4. Construction of Containment Event Trees Containment Event Trees Definition and Analysis • The containment event trees (CET) provide a systematic approach for evaluation of accident sequences that lead to containment failure in coping with severe accident. • The CET structure and nodal questions address all of the relevant issues important to severe accident progression, containment response, failure, and source terms. CET for the PWR plant • The CETs with the AMs were developed to trace the interdependent physico-chemical processes influencing severe accident progression in the reactor system and the containment. • The heading (B1, C4, C5 and D1) concern with four severe accident phenomena and are treated with the ROAAM method. 7
CET for a period from the initiation of core damage to the reactor vessel failure BM1 B1 B2 BM2 B3 B5 Accident B4 Progression Temperature Water Temperatur Hydrogen Forced In-vessel Over-pressure injection into induced e induced burning failure CV Failure Depressuri steam the SG SGTR zation LOCA before RV fail. before RV fail. Modes explosion ROAAM 1150 1150 1150 No C No No C Yes Yes Success Hydrogen Detonation Yes Steam Explosion No C Success No C Yes B Yes Hydrogen Detonation No No C No No Yes C Failure Yes No Hydrogen Detonation Yes Induced SGTR No Failure C Yes No Yes C Yes Hydrogen Detonation Yes Steam Explosion • For headings of B2, B3, and B4, probability distributions were determined with a method similar to the Zion plant evaluation in NUREG-1150. • The end points of CETs that were relevant to the integrity and retention capability of the containment were attributed to containment failure modes. 8
CET for an early phase after the reactor vessel failure C2 C7 C1 CM1 C3 C4 C5 C6 CM2 Accident Progression Release Debris Direct CV fail. Water Reactor Ex-vessel Charging Hydrogen fraction of release containment just after injection cavity steam CV Failure injection detonation core mass heating type into the CV coolant explosion RV fail. Modes 1150 ROAAM ROAAM 1150 Wet D Dispersion No D Dry No D No Yes Yes DCH No Yes D Yes Small DCH No D No No D Success Yes Yes Hydrogen Detonation Yes Steam Explosion Gravity No falling D No No D Yes Success Yes Hydrogen Detonation Yes C Steam Explosion No D Failure No No D Yes Wet Yes Hydrogen Detonation Yes Failure Steam Explosion No D Dry No D Yes Yes Hydrogen Detonation Large Same tree structure as above described • The same split probability as the point-estimate evaluation was used for top events that did not consider the probability distribution. 9
5. Uncertainty Analysis of Containment Failure Frequency 5. Uncertainty Analysis of Containment Failure Frequency 1. By using probability distributions of headings concerning mitigation AMs, and probability distributions obtained by the ROAAM method of headings, uncertainty distributions of containment failure frequency and release categories were obtained by means of the uncertainty propagation analysis of the containment event tree. 2. The probability distribution of a mode failure was calculated by 200 samplings by the PREP/SPOP code for the probability distributions of each box which constitutes the phenomenological event tree. 10
ROAAM method Containment failure probability Containment failure In-Vessel Steam Explosion F probability vs Missile energy 1 Missile energy after shield impact Missile energy after shield F impact vs Vessel head missile energy Cumulative Probability Vessel head missile energy Vessel head missile energy 0.1 F vs Net energy in vessel head Net energy in vessel head Net energy in vessel head F vs Slug energy 0.01 Upward slug energy Upward slug energy F vs Mechanical energy release Mechanical energy release 0.001 10 -9 10 -8 10 -7 10 -6 10 -5 Conversion ratio F vs Probability Thermal energy in explosion Thermal energy of melt in explosion Average : 6E-06 5% value : 7E-09 Mass of melt in premixture 50% value : 3E-08 Mass of melt in premixture F (F; stands for a function.) vs 95% value : 1E-06 Pour area Size of pour area 11 T. G. Theofaous, et al., NUREG/CR-5030
Ex-Vessel Steam Explosion ROAAM method Coarse mixing debris mass 1.00 Debris internal into the reactor cavity from energy the reactor vessel 0.80 Cumulative Probability Multiply Mechanical energy 0.60 exchange efficiency 0.40 Multiply 0.20 Mechanical energy Containment fragility 0.00 9x10 -4 1.2x10 -3 1.5x10 -3 3x10 -4 6x10 -4 0 Probability Comparison Average : 2E-04 5% value : 0.0 50% value : 0.0 Containment failure probability 95% value : 1E-03 12
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