Development of Integrated Analytical Tools for Level-2 PSA of LMFBR

Transcription

1 FR09, Dec.7-11, 2009, Kyoto Development of Integrated Analytical Tools for Level-2 PSA of LMFBR December 8, 2009 K. Haga, H. Endo, T. Nakajima, T. Ishizu Nuclear Energy System Safety Division Japan Nuclear Energy Safety Organization (JNES) H. Iizuka MHI Nuclear Engineering Company 1 1

2 Contents 1. Background and Objectives 2. Computer codes 2.1 For the plant response phase NALAP-II ACTOR 2.2 For the core disruptive phase ARCADIAN-FBR APK 2.2 For the Containment vessel AZORES 3. Phenomenological Relation Diagram : PRD 4. Summary 2 2

3 1.Background and Objectives JNES is engaged in the technical support for the regulatory activity for Monju, the Nuclear and Industrial Safety Agency (NISA). The major areas of the supportive work are Inspection and Safety analysis. To the safety analysis, JNES has been developing the own tools for several years. Our reset big activity is the effectiveness evaluation of the proposed accident management (AM) of Monju by using PSA. The developed tools for the PSA are not only computer codes but also the decision method for the balancing points of event trees. This presentation presents the outline of each tool. 3

4 Process of severe accidents of LMFBR and JNES s Analysis tools Plant response phase Core disruption phase CV response phase Sodium Boundary Sodium Core Re- FP Na-concrete temp. & failure boiling disruptio criticali release reaction press. n ty.debris-conc. NALAP-II increase ARCADIAN-FBR APK reaction Plant kinetics Neutronics Super-prompt Na spray burning criticality H2 burning Coolant temp. Keff and reactivity Aerosol release *System pressure coeffs. of degraded FP release core ACTOR Power and reactivity FP behavior transient of Super in PHTS prompt criticality ABAQUS FP evap. from fuel Structure Analysis FP transport in Na, from Na to cover gas, Cover gas heating by FP Hydrogen behavior due to Large deformation within cover gas space Na-concrete and debris and inelastic FP deposition on piping wall reactions deformation analysis Hydrogen burning Aerosol and FP transports AZORES Containment Vessel responses & FP transport 4 EFD *Phenomenological ET expansion PRD Quantification pf phenomenological ET Containment failure Frequency CFF 4

5 2. Computer codes 2.1 For the plant response phase(1) NALAP-II code Functions: Plant kinetics of the primary and secondary loops Sales point: A function to evaluate the degree of hightemperature creep of the location by SCDF (Structural cumulative damage factor, D c ). D c =Σ(Δt i /t r ), (1) where, Δt i : computation time step (s), t r : creep failure time (s). In eq. (1), t r depend on the material, temperature and pressure. It is calculated by using the Larson-Miller Parameter. It was considered that the location will fail when D c =0.2~1.0 5

6 The thermo-hydraulic behavior analyzed by NALAP-II to PLOHS -Reactor inlet and outlet temperatures, cover gas pressures, and SCDF of major locations- 6 The RV outlet temperature exceeded 800 at 51hr. PHTS pressure began to increase at 22 hr due to the sodium volume expansion. The SCDF is evaporator cylindrical wall made of 2.25Cr-1Mo steel firstly exceeded the line of 0.2. Another locations exceeded the line of SCDF=0.2 after about 10 hr.

7 2.1 For the plant response phase(2) Released to inside of reactor containment vessel due to a failure in the deck shield [ACTOR code] Functions: FP behavior from fuel to cover gas or flowing sodium Adsorption behavior on wall surface Transferred into cover gas by N percent Adsorption behavior on wall surface Deck shield Cover gas space Recirculation Air-liquid diffusion process Primary system piping Cover gas Breakup into Z bubbles After 50 cm ascension Fuel Bubble Sodium Primary loop Gas plenum Gas plenum Bubble Bubble Sales point: Analysis of FP gas bubble behavior with other FP nuclides Fuel pin In sodium Release behavior from fuel pin In the event of fuel failure In the event of cladding breach Atomic FP FP associated with bubble 7 FP on wall surfaces

8 Calculations by ASFOR Amount 移行量 of iodine / 初期体積 transferred/initial [mol mol/m3] volume [mol/m 3 ] 試験 Test (1mol%) 試験 Test (4mol%) 試験 (1mol%) 試験 (4mol%) 試験 Test (8mol%) 試験 Test (20mol%) 試験 (8mol%) 試験 (20mol%) 試験 Test (40mol%) 浸透モデル Permeation model (1mol%) 試験 (40mol%) 浸透モデル (1mol%) 浸透モデル Permeation 浸透モデル model (4mol%) (4mol%) 浸透モデル Permeation 浸透モデル model (8mol%) (8mol%) 浸透モデル Permeation 浸透モデル model (20mol%) (20mol%) 浸透モデル Permeation model (40mol%) (40mol%) バルクモデル Bulk model バルクモデル (1mol%) (1mol%) (1mol%) バルクモデル Bulk model (4mol%) (4mol%) バルクモデル Bulk model バルクモデル (8mol%) (8mol%) (8mol%) バルクモデル Bulk model (20mol%) (20mol%) バルクモデル Bulk model (40mol%) (40mol%) Distance [m] 距離 [m] Amount of iodine transferred per bubble volume: Experiment & Calculation Transferred iodine (mol/m 3 ) Percentage of Nuclide Groups to Initial Inventories[-] 1.E+00 1.E-01 1.E-02 1.E-03 Fu e l p i n s C o o l a n t Zones i n reactor Z on es W a l l ( C o o l a n t ) C o v e r g a s W a l l ( C o v e r g a s ) Ni b b l e g a s e s H a l og e n Al ka l i n e m e ta l s Su l f u r g r ou p Al ka l i n e e a r th Ce r i u m g r ou p P r e c i ou s m e ta l s N u c l i d e G r N u c l i d e G r o u p s La n th a n oi d Distribution of FP nuclides in reactor 100s after cladding failure in PLOHS

9 2.2 For the core disruption phase (1) ARCADIAN-FBR code Functions: Neutronics for degraded core Sales point: The combination of (1)a deterministic calculation part to provide the various reactivity coefficients and those spatial distributions and (2) a Monte Carlo calculation part to provide reference results against the deterministic calculations and to evaluate approximation effects employed in the deterministic calculations 9

10 Constitution of ARCADIAN-FBR Deterministic Calculation Part JFS-3-J3.3 J3.3 Monte Carlo Calculation Part JENDL ENDL-3.3 Cell Code SLAROM 70-group Effective Cross Sections σ eff eff Diffusion Code DIF3D Transport Code VARIANT Monte Carlo Code GMVP Monte Carlo Code MVP k eff eff φ φ * k eff eff φ φ * k eff eff φ φ * k eff eff φ Burnup Cal. Code REBUS Perturbation Code PERKY Kinetics Parameter Calculation Module Burnup Cal. Code MVP-BURN N ρ l β l β N 10

11 2.2 For the core disruption phase (2) APK code Functions: Reactivity evaluation by solving 6-group point reactor kinetics equations to debris beds of some temperature Sales point: Simple using one point temperature Validation: Comparison with SAS4A POWER-APK POWER-sas4a 1.50 NET-APK NET-sas4a 3500 TEMP(PK) TEP-sas4a FPS-APK FPS-sas4a P o w e r ( P 0 ) 1000 r e a c t iv it y ($ ) T e m p e r a t u r e (K ) e n e r g y ( F P S ) time(s) time(s) time(s) time(s) (a) Power history (b) Net reactivity (c) Average fuel temp. (d) Input energy 11

12 2.3 For the CV response phase AZORES code Functions: To analyze various reactions caused by the sodium and core materials leaked through a failed coolant boundary. Sales point: Radioactive materials behaviours to the environment are also solved with the thermo-chemical reactions. Nitrogen atmosphere 窒素雰囲気 HVA 格納容 Containment Vessel FP FP aerosol HVAC Reactor 炉 vessel 容器 Core 炉心 Core-concrete コアコンクリート反応 reactions Hydrogen burning 水素燃焼 PHT Cooling system 冷却系 Na 燃焼 Na コンクリート反応 Air atmosphere 空気雰囲 Sodium burning FP SHTS Sodium-concrete reactions Auxiliary 補助建家 building Sodium burning Na 燃焼 N 12

13 Example of CV response calculated by AZORES 13 Core temperature (K) Time (hour) Fig. 7-22(1) Core Temperature (Case 15: PLOHS, Non-CVBP, PL-NSL-Hγ Pressure (Pa) 1.0E E E E E E E E E E E+00 Reactor chamber PHTS chamber: AC Loop HVAC room: AC Loop PHTS chamber: B Loop HVAC room: B Loop On the containment vessel floor Inside the Reactor vessel/the primary system pipe Inside the secondary system pipe Annulus. SHTS chamber Environment Time (hour) Fig. 7-22(2). Pressure (Case 15: PLOHS, Non-CVBP PL-NRL-Hγ) Hydrogen mol fraction (-) Time (hour) Reactor chamber PHTS chamber: AC Loop HVAC room: AC Loop PHTS chamber: B Loop HVAC room: B Loop On the containment vessel floor Inside the Reactor vessel/the primary system pipe Inside the secondary system pipe Annulus SHTS chamber Environment Fig. 7-22(3). Hydrogen Mol Fraction (Case 15: PLOHS, Non-CVBP, PL-NRL- Fig. 7-23(3) Ce Gr Initial Inventory Ratio (Case 15:PLOHS, Non-CVBP PL-NRL-Hγ) H-γ) Xe (-) 1E+00 1E-01 1E-02 1E-03 1E-04 1E-05 1E-06 1E-07 1E-08 1E-09 1E Time (hour) Reactor chamber PHTS chamber: AC Loop HVAC room: AC Loop PHTS chamber: B Loop HVAC room: B Loop On the containment vessel floor Inside the Reactor vessel/the primary system pipe Inside the secondary system pipe Annulus. SHTS chamber Environment Fig. 7-23(1) Xe Gr Initial Inventory Ratio (Case 15: PLOHS, Non-CVBP PL- NRL-H-γ) I (-) 1E+00 1E-01 1E-02 1E-03 1E-04 1E-05 1E-06 1E-07 1E-08 1E-09 1E-10 Reactor chamber PHTS chamber: AC Loop HVAC room: AC Loop PHTS chamber: B Loop HVAC room: B Loop On the containment vessel floor Inside the Reactor vessel/the primary system pipe Inside the secondary system pipe Annulus. SHTS chamber Environment Time (hour) Fig. 7-23(2) I Gr Initial Inventory Ratio (Case 15, PLOHS, Non-CVBP PL- NRL-H-γ) Ce (-) 1E+00 1E-01 1E-02 1E-03 1E-04 1E-05 1E-06 1E-07 1E-08 1E-09 1E Time (hour) Reactor chamber PHTS chamber: AC Loop HVAC room: AC Loop PHTS chamber: B Loop HVAC room: B Loop On the containment vessel floor Inside the Reactor vessel/the primary system pipe Inside the secondary system pipe Annulus. SHTS chamber Environment 13

14 14 Japan Nuclear Energy Safety Organization 3. Phenomenological Relation Diagram : PRD Probability decision method for the blanching points in event trees (ET) Objectives of PRD: To eliminate the subjective decision for the probability of blanching points in ETs. Procedure: (i) Constructs an event tree from the top event (an event considered at a branching point of an phenomenological event tree) to lower events. (2) Function gate is settled to calculate by a certain function with lower events. (3) To lower events the probabilities are to be give by any method as much as possible. The probability distributions are transferred to the upper events.

15 -PRD Appoach for the top event of acceleration - Top event [ Acceleration: a] Incidence probability -The top event that will be evaluated by the PRD approach. -[Acceleration distribution a incidence probability distribution] Yes Conditional branch No Top event Probability distributions are given to events that are considered to have a margin of uncertainties. Distributions to be given are defined by the probability distribution function. -Events are composed of several events. This case represents that events are composed of elementary events1 to 3 Events (outputs from the function gate) Events [Force: F] Function gate [Motion equation: F=ma] Sub-PRD [Mass: m] To be calculated by a certain function with lower events as inputs. The details of calculation are indicated in the adjoining blue fames. The probability distribution function to be given to a certain variable: Example) U[1,10]: A uniform distribution between 1 and 10 N[10,5]: A normal distribution with the average value of 10 and the standard deviation of 5 [L,U]=[0,100]: Lower limit 0; Upper limit ] Probability distribution function Elementary event 1 [F1] Elementary event 2 [F2] Elementary event 3 [F3] To be expanded in detail in a lower PRD Bottom event 15

16 Evaluation procedures of ROAAM used LWR and PRD ROAAM RO AAM PRD PRD (1) Selection of the dominant factors and development of diagram regarding (1) 着目現象に対する支配要因の抽出とダイヤグラムの展開 the concerned phenomenon 着目現象の支配要因を抽出して下位事象へと分解し 相互に関連付ける Select the dominant factors of the concerned phenomenon, break them down into low-order events, and relate them to each other. Develop the reciprocal relations among the lower-order events into a diagram 下位事象の相互関係をダイヤグラムに展開 (2) Establishment of parameters of the dominant factors and introduction of probability distribution (2) 支配要因のパラメータ設定と確率分布の導入 着目現象を記述するパラメータを Determ inistic(d) Probabilistic(P) intangible variables(i) の 3 種に分類 Parameters, which describe the concerned phenomenon, are classified into three categories including deterministic (D), probabilistic (P) and intangible variables (I). 展開した下位事象の P パラメータについて確率分布を設定 Establishment of the probability distribution regarding the P parameter of the developed lower-order events Establishment of the value based on engineering judgment regarding the I parameter I パラメータについて工学的判断による値を設定 (3) Supplementary use of deterministic findings (3 ) 決定論的知見の援用 下位事象の代表値 (D パラメータ ) を解析コードによる解析 単純な相関式などを援用して設定 Establishment of the representative value of the lower-order events (D parameter) by using analysis based on the analytical codes, supported by a simple correlation equation 下位事象の代表値 (D パラメータ ) を実験的知見を援用して設定 Establishment of the representative value of the lower-order events (D parameter) supported by the empirical findings (4) Evaluation of branching probability regarding the concerned phenomenon (4 ) 着目現象に対する分岐確率の評価 Evaluation of the branching probability of the concerned phenomenon, tracing the functional relation モンテカルロ法などで関数関係を上位に辿り着目現象の分岐確率を評価 upwards using the Monte Carlo method and others Fig.D1-1 Evaluation Procedures of ROAAM and PRD Both methods seem 図 - 1 basically RO AAM とPRD における評価手順 identical. 16

17 Summary With the developed integrated analytical tools, one-through evaluation for severe accidents including FP release ratio became possible to LMFBR. The tools were used for the effective evaluation of AM for Monju. Now, further efforts are being made to make analyses more realistic for the Monju with an advance core and the Japanese demonstration LMFBR. These improvements are very helpful in constructing data-bases of the Emergency Response Support System (ERSS) for Monju by conducting more reliable analyses to the conceivable scenarios after initiating events. 17

18 Many thanks for your audience. 18

19 Characteristics of PLOHS After the reactor shutdown, the plant temperature increases slowly but monotonously both in the primary cooling system (PHTS) and in the secondary cooling system (SHTS). Many plant locations will fail due to the hightemperature creep. Among several conceivable accident scenarios, the containment vessel by-bass (CVBP) would be most important event due to the high probability and the high potential of large scale FP release to the environment. We evaluated the frequency of PLOHS/CVBP in this study. 19

20 FP leak passes in PLOHS/CVBP Japan Nuclear Energy Safety Organization CV SHTS Na burn aerosol generation pressure increase Annulus ac SG Non-CVBP case P PHTS RV IHX Na leakage Na/Concrete reaction Hydrogen generation Pressure increase CVBP case 20 In the case of PLOHS/CVBP, some early failure in SHTS is postulated. When some interface between PHTS and SHTS failed, FP from the molten fuel will run through PHTS and SHTS piping.

21 Boundaries of IHX that have a high potential of first failure in PHTS Diameter: 700mm Thickness: 1.5, 2.0 mm Material: SUS3316 IHX bellows Cover gas bellows Bellows SHTS coolant outlet PHTS coolant inlet Bottom head IHX Bottom Head (interface between PHTS/SHTS) Diameter: 2190mm Thickness: 25mm Material: SUS304 Heat transfer tubes Heat transfer tube Bottom head 21 PHTS coolant outlet

22 An vent tree of PLOHS Power supplies Leakage from the shield plug of RV Eearlier secondary system failure Failure in some areas of promary system yes no 0.25 (a) (b) PLOHS (c) 0.96 (a) 0.75 (d) (b) CVBP The probabilities of branching point (a), (b), (c),(d) are obtained through present analyses. 22

23 2. Analyses to obtain the probability of the branching points Failu re mechanism Objects for analysis Analysis code High-temperature creep Buckling Tensile (or breaking) stress Vessel, Piping Bellows and bottom head of IHX NALAP-II ABAQUS FINAS Thermal-hydraulic Analysis Structual Analysis 23

24 2.1 Thermal-hydraulic Analysis FBR plant thermal-hydraulic analysis code NALAP-II NALAP-II has been developed in JNES. For the present study a function was added to evaluate the degree of high-temperature creep of the location: SCDF (Structural cumulative damage factor, D c ). D c =Σ(Δt i /t r ), (1) where, Δt i : time step (s), t r : creep failure time (s). 24 In eq. (1), t r depend on the material and it is calculated by using the Larson-Miller Parameter. It was considered that the location will fail when D c =0.2~1.0.

25 Specifications of proto-type LMFBR plant and locations where high-temperature creep was evaluated (a) Spec. of LMFBR (b) Locations of plant evaluated high-temperature creep Thermal power 714 MW Number of loops 3 Primary heat transfer system (PHTS) Temperature at reactor inlet/outlet 397/529 Flow rate 5100 t/h Pressure at reactor inlet/outlet 0.8/0.1 MPa Main circulatiion pump One unit per loop IHX Number of unit 3 Type Vertical parallel flow with no sodium surface Secondary heat transfer system (SHTS) 325/ Temperature at IHX inlet/outlet 3700 t/h Flow rate 0.8/0.1 MPa Main circulatiion pump One unit per loop By-pass of the secondary Auxiliary cooling system cooling system with one air cooler per loop D (inner dia.) t (thickness) Location (mm) (mm) D/t RV cylindrical wall Reactor outlet piping Reactor inlet piping IH X vessel cylindrical wall IHX heat transfer tube IHX SHTS- side outlet piping EV cylindrical wall Air-cooler heat transfer tube Fuel pin cladding Material 142 SUS304 stainless steel 72 SUS304 stainless steel 62 SUS304 stainless steel 98 SUS304 stainless steel 16 SUS304 stainless steel 56 SUS304 stainless steel Cr-1Mo steel 16 SUS304 stainless steel 20% cold-worked modified 316 stainless steel The locations with large value of D(diameter)/t(thickness) were chosen for the present calculation of SCDF. 25

26 Critical stress 限界応力 (MPa) Japan Nuclear Energy Safety Organization 2.2 Structual analysis ABAQUS 3-d analysis to IHX bottom head - Buckling critical stress and fracture critical stress - The buckling of IHX bottom head became marked at the end of calculation: Temperature: ; Pressure: 0.89MPa) IHX 下部鏡 (SUS SUS ) の座屈限界応力座屈限界応力と破断限界応力 Buckling critical stress and fracture critical stress in IHX bottom head (SUS304) Stress-pressure relationship equation in IHX bottom head IHX 下部鏡の応力応力と圧力圧力の関係式 σ(mpa)=θ a (P-0.1) σ(mpa)=θ a (P-0.1) a=89.89, a= P(MPa): Primary P(MPa MPa):1 system 次系圧力 pressure θ: 近似式の不確定幅 θ: A margin of 確率変数 uncertainties ( ~ ) in an approximate expression Probability variable (0.95~1.05) Fracture critical stress (σ F ) ( s MPa MPa) 破断限界応力 (σf) Buckling 座屈限界応力 critical (σb) stress (σ B ) ( s MPa MPa) Results of ABAQUS-based analyses (buckling) Results :ABAQUS の結果 of the ( 座屈 ABAQUS-based ) analyses (backling) IHX exit temperature 1 次系 Na の( ) IHX 出口温度 ( ) 26 The obtained stress-pressure relation curve encountered with the buckling critical stress line and the fractional critical stress line at one point, respectively. The IHX bottom head is expected to fail at the condition between the two points.

27 The results of FINAS 2-d analysis to IHX bellows - Yield stress and tensile stress - Deformed IHX bellows (Temperature: 750 ; Pressure: MPa) (The results of the FINAS 3-d calculation Critical stress (MPa) 限界応力 (MPa MPa) Yield stress 降伏応力 IHX ベローズ (SUS SUS ) の座屈座屈 破断限界応力 IHX bellows (SUS316) buckling critical stress and fracture critical stress IHX bellows stress-pressure relationship equation σ(mpa)=φ a (P-0.1) IHX ベローズ応力 a= σ(mpa)=φ a (P-0.1) Tensile (or breaking) P(MPa): a= Primary system P(MPa MPa):1 pressure 次系圧力 (MPa MPa) stress 引張 (MPa) ( 破断 ) 応力 φ: 応力近似式の不確定幅 Φ:A margin 確率変数 of uncertainties ( ~ ) in an approximate expression (0.95~1.05) ( s MPa MPa) (198000s MPa) ( s MPa MPa) (189000s MPa) 次系 Na の IHX 出口温度 ( ) IHX outlet temperature of primary-system sodium The obtained stress-pressure relation curve encountered with the yield stress line and the tensile stress line at one point, respectively. The IHX bellows is expected to fail at the condition between the two points.

28 3. Discussion The Range of Primary System Component Fracture Pressures Pressure range for failure of PHTS components 3 3.8MPa MPa 2.6MPa Pressure(MPa) MPa 1.117MPa 1.928MPa 0.961MPa 1.075MPa 1.165MPa 0.928MPa 0.768M RV cylindrical wall RV outlet piping IHX cylindrical Names of components wall RV inlet piping IHX bellows IHX bottom head 28 Five locations (RV cylindrical wall, RV outlet piping, IHX cylindrical wall, IHX bellows and IHX bottom head) fail at the pressure between 0.77 to 2.3 MPa).

29 Time range for failure of PHTS components 29 The five locations fail at a narrow time period (between 52hr to 59hr). Thus, it will be hard to say the failure order in these locations. In PSA, it will be reasonable to assign the equal probability of first failure to these locations, that is 0.2.

30 4. Conclusion Japan Nuclear Energy Safety Organization (i) A model to analyze the high temperature creep progression by introducing the calculation function of SCDF was added to the LMFBR plant thermal-hydraulics code, NALAP-II. (ii)the model was applied to the PLOHS event of the typical proto-type LMFBR. The results indicated that the evaporator made of 2.1/4Cr-1Mo steel in SHTS firstly failed when the system temperature exceeded 800. The main plant components and piping made of SUS 304 stainless steel failed when the system temperature exceeded 870. (iii) In addition, detailed structural analyses were performed by using ABAQUS and FINAS codes and the temperature and pressure histories to the locations where the buckling and the tensile stresses are the causes of failure. (iv) Comparing the failure time information of each location, it was concluded that the probability of CVBP was 0.2 to the plant. 30

31 Appendix additional information These numbers were assigned from present study Power Leakage from the Eearlier secondary Failure in some areas supplies shield plug of RV system failure of promary system (a) yes no 0.25 (b) PLOHS CVBP CFF Frequency (1/RY) 1.00E E E E E E E E-14 Loss of annulus function RV isolaiton failure RV failure by hydrogen burning RV failure by sodium burning CVBP Retained RV integrity Terminal conditions of containment response process (a) Evaluated provability of the branching points (b) Results of trial evaluation of CFF to each level-2 scenario Present trial evaluation shows that the frequency of CVBP is one order smaller than that of retained RV integrity scenario. However, the frequency is more than one order larger than that of other scenarios. 31

32 32 Japan Nuclear Energy Safety Organization

33 Flow network of the NALAP-II code for RV and PHTS Maintenance cooling system Outlet plenum UCS Cover gas H/L Hot leg IHX IHX M/L Middle leg Cold C/L leg Flow hole Primary loop Maintenance cooling system Inner vessel Core channel Inner core Ch. Outer core Ch, Radial blanket Ch, Shield Ch, Bypass Ch, Lower plenum Primary loop R/V R/V PHTS flow network Pump PUMP RV flow network R/V Fig.5 Flow network of the NALAP-II code for RV and PHTS 33

34 IHX inlet and outlet temperatures, and PHTS flow rate IHX-1 次入口温度 IHX-1 次出口温度 IHX-2 次入口温度 IHX-2 次出口温度 1 次流量定格比 温度 ( ) 時間 ( 10,000 秒 ) 流量比 (%) 34

35 Temperature and the Mises stress behavior of IHX bellows (Stress with the maximal element of equivalent creep strain: the center of plate pressure) (The results of the FINAS-based two-dimensional elasto-plastic large=scale deformation and creep analysis) Str100.0 ess (N/ 80.0 mm2 ) 応力 (N/m m 2 ) Mises stress Circumferential ミーゼス応力 stress Stress 周方向応力 in the meridian direction 子午線方向応力 Temperature 温度 Temp eratu re ( ) 温度 ( ) Time (s) 時間 ( 秒 ) 35

36 Temperature and the Mises stress behavior of IHX bellows (Stress with the maximal element of equivalent creep strain: the center of plate pressure) (The results of the FINAS-based two-dimensional elasto-plastic large=scale deformation and creep analysis) Str100.0 ess (N/ mm2 ) 80.0 応力 (N /m m 2 ) Mises stress Circumferential ミーゼス応力 stress Stress 周方向応力 in the meridian direction 子午線方向応力 Temperature 温度 Temp eratu re ( ) 温度 ( ) Time (s) 時間 ( 秒 ) 36

37 Comparison of initiating events frequency between JAEA and JNES 37 37

38 4. Results of PSA Japan Nuclear Energy Safety Organization 4.1 Effects of AM measures to keep the RV coolant level Core Dam age Freq uenc ies / Reac tor Year 炉心損傷頻度 ( / 炉年 ) Before AM Measures After AM Measures :AM 前 :AM 後 Primary main coolant system circulation pump 1 次主冷却系 trip failure 循環ポンプトリップ失敗 Reactor covergas feeding Isolation cut failure of argon supply line to RV 原子炉カハ ーカ ス供給停止失敗 A failure to Loss of guard Keeping 1 次ナトリウム Siphon isolate the 1 次メンテナンス break vessel leakage from the sodium カ ート ヘ ッセルの原子炉 primary integrity of temperature 冷却系漏えい maintenance cooling system の健全性容器への high maintenance 箇所の隔離喪失汲み上げ cooling 失敗失敗 system Double leak (in an area below SsL) 1 following 次主冷却系 leakage from the primary 漏えい後の二重漏えい (SsLより下端) main coolant system All loss of reactor level (LORL) event LORL- 全体

39 4.2 Effects of initiating events frequency Frequency (/ry) 1.0E-06 Loss of reactor shutdown functions Loss of functions to maintain a proper reactor coolant level Loss of decay heat removal functions All relevant events 1.0E E E-09 Basic Conditions Reference Analysis 5 Basic Conditions Reference Analysis 5 JNES JAEA JNES before AM after AM JAEA

40 4.3 Effects of choosing common cause failure (In the PSA of JAEA, common cause failure is considered between the main reactor shutdown system and the backup reactor shutdown system) Frequency (/ry) 1.0E E-07 Loss of reactor shutdown functions Loss of functions to maintain a proper reactor coolant level Loss of decay heat removal functions All relevant event LORL All ATWS 2.7 times LORL All All No AM Measures for ATWS 3.8 times All PLOHS ATWS increases PLOHS 1.0E E-09 ATWS Basic Conditions Reference Analysis 7 Basic Conditions Reference Analysis 7 JNES JAEA JNES Before before AM After AM after AM JAEA

41 4.4 Contribution of initiating events to CDF (b) after AM (by JNES) 1.0E E-05 ULOF 0% UTOP 0% ULOHS 0% ULOPI 0% UTOP/ULOF 0% Core Damage Frequencies/Reactor Year 1.0E E E E E E E-07 LORL 76% PLOHS 24% 1.0E E-13 PLOHS LORL ULOF UTOP ULOHS ULOPI UTOP/ULOF Full core damage frequencies Full core damage frequency : /reactor year ATWS 41

42 1.0E-06 Japan Nuclear Energy Safety Organization 4.5 Comparison of estimated CDF between JNES and JAEA 原子炉停止機能喪失原子炉液位確保機能喪失崩壊熱除去機能喪失全体 Frequency 1.0E-07 (/ry) LORL 全体 Total 0.6 倍 times LORL 全体 Total 全体 Total 倍 times 全体 Total PLOHS PLOHS 1.0E E-09 ATWS basal condition -befoream ATWS basal condition -after AM JAEA-before-AM JAEA--after AM JNES JAEA 42 42