development of integrated analytical tools for level-2 psa ...€¦ · development of integrated...
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Japan Nuclear Energy Safety Organization
Development of Integrated Analytical Tools for
Level-2 PSA of LMFBR
FR09, Dec.7-11, 2009, KKKKyoto
December 8, 2009
K. Haga, H. Endo, T. Nakajima, T. IshizuNuclear Energy System Safety Division
Japan Nuclear Energy Safety Organization (JNES)
H. IizukaMHI Nuclear Engineering Company
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Contents1. Background and Objectives2. Computer codes2.1 For the plant response phaseNALAP-IIACTOR
2.2 For the core disruptive phaseARCADIAN-FBRAPK
2.2 For the Containment vesselAZORES
3. Phenomenological Relation Diagram : PRD4. Summary
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1111....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.
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�Hydrogen behavior due to Na-concrete and –debris reactions�Hydrogen burning�Aerosol and FP transports
・・・・Coolant temp.*System pressure
・・・・FP evap. from fuel・・・・FP transport in Na, from Na to cover gas, ・・・・ Cover gas heating by FP within cover gas space・・・・FP deposition on piping wall
Plant response phase Core disruption phase
・・・・Large deformation and inelastic deformation analysis
CV response phase
AZORESContainment Vessel responses & FP transport
ACTORFP behavior
in PHTS
Boundary failure
FP release
ABAQUSStructure Analysis
Sodium boiling
Core disruptio
n
Re-criticali
ty
・・・・Keff and reactivity coeffs. of degraded core
Sodium temp. & press. increase APK
Super-prompt criticality
ARCADIAN-FBRNeutronics
・・・・ Na-concrete reaction.debris-conc. reaction・・・・Na spray burning・・・・H2 burning・・・・Aerosol release・・・・FP release
・・・・Power and reactivity transient of Super prompt criticality
EFD*Phenomenological ET expansion
PRDQuantification pf
phenomenological ET
Containment failure FrequencyCFF
NALAP-IIPlant kinetics
�Hydrogen behavior due to Na-concrete and –debris reactions�Hydrogen burning�Aerosol and FP transports
・・・・Coolant temp.*System pressure
・・・・FP evap. from fuel・・・・FP transport in Na, from Na to cover gas, ・・・・ Cover gas heating by FP within cover gas space・・・・FP deposition on piping wall
Plant response phase Core disruption phase
・・・・Large deformation and inelastic deformation analysis
CV response phase
AZORESContainment Vessel responses & FP transport
ACTORFP behavior
in PHTS
Boundary failure
FP release
ABAQUSStructure Analysis
Sodium boiling
Core disruptio
n
Re-criticali
ty
・・・・Keff and reactivity coeffs. of degraded core
Sodium temp. & press. increase APK
Super-prompt criticality
ARCADIAN-FBRNeutronics
・・・・ Na-concrete reaction.debris-conc. reaction・・・・Na spray burning・・・・H2 burning・・・・Aerosol release・・・・FP release
・・・・Power and reactivity transient of Super prompt criticality
EFD*Phenomenological ET expansion
PRDQuantification pf
phenomenological ET
Containment failure FrequencyCFF
NALAP-IIPlant kinetics
Process of severe accidents of LMFBR and JNES’s Analysis tools
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2. Computer codes 2.1 For the plant response phase((((1111))))NALAP-II code
Functions: Plant kinetics of the primary and secondary loopsSales point: A function to evaluate the degree of high-temperature creep of the location by SCDF (Structural cumulative damage factor, Dc ).
� Dc=Σ(Δti/tr), (1)where,
Δti: computation time step (s), tr: creep failure time (s).
In eq. (1), tr 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 Dc=0.2~1.0.
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The thermo-hydraulic behavior analyzed by NALAP-II to PLOHS-Reactor inlet and outlet temperatures, cover gas pressures, and SCDF of major locations-
•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.
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2.1 For the plant response phase((((2))))[ACTOR code]Functions: FP behavior from fuel to cover gas or flowing sodium
Cover gas
Fuel BubbleSodium Primary loop
Sales point: Analysis of FP gas bubble behavior with other FP nuclides
Released to inside of reactor containment vessel due to a failure in the deck shield
Deck shield
Adsorption behavior on wall surface Cover gas space
Transferred into cover gas by N percent
Recirculation
Adsorption behavior on wall surface
Air-liquid diffusion process
Breakup into Z bubbles
Primary system piping →→→→
After 50 cm ascension
Gas plenum
Fuel pin
BubbleBubble
Gas plenum
In sodium
In the event of fuel failure In the event of cladding breach
Release behavior from fuel pin
Atomic FP
FP associated with bubble
FP on wall surfaces
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Calculations by ASFOR
0.010.010.010.01
0.100.100.100.10
1.001.001.001.00
10.0010.0010.0010.00
0.00.00.00.0 0.50.50.50.5 1.01.01.01.0 1.51.51.51.5 2.02.02.02.0
距離 距離 距離 距離 [[[[mmmm]]]]
移行
量移
行量
移行
量移
行量
// //初
期体
積
初期
体積
初
期体
積
初期
体積
[[ [[m
ol
mol
mol
mol// //mm mm
33 33]] ]]
試験(1mol%) 試験(4mol%)
試験(8mol%) 試験(20mol%)
試験(40mol%) 浸透モデル(1mol%)
浸透モデル(4mol%) 浸透モデル(8mol%)
浸透モデル(20mol%) 浸透モデル(40mol%)
バルクモデル(1mol%) バルクモデル(4mol%)
バルクモデル(8mol%) バルクモデル(20mol%)
バルクモデル(40mol%)
Amoun
t of iod
ine tra
nsferr
ed/Init
ial vol
ume [m
ol/m3 ]
Distance [m]
試験(1mol%) 試験(4mol%)
試験(8mol%) 試験(20mol%)
試験(40mol%) 浸透モデル(1mol%)
浸透モデル(4mol%) 浸透モデル(8mol%)
浸透モデル(20mol%) 浸透モデル(40mol%)
バルクモデル(1mol%) バルクモデル(4mol%)
バルクモデル(8mol%) バルクモデル(20mol%)
バルクモデル(40mol%)
Test (1mol%) Test (8mol%) Test (40mol%) Permeation model (4mol%) Permeation model (20mol%) Bulk model (1mol%) Bulk model (8mol%) Bulk model (40mol%)
Test (4mol%) Test (20mol%) Permeation model (1mol%) Permeation model (8mol%) Permeation model (40mol%) Bulk model (4mol%) Bulk model (20mol%)
Fu e l p i n sC o o l a n t
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
P r e c i ou s m e ta l s
Ce r i u m g r ou p
La n th a n oi d
1.E-03
1.E-02
1.E-01
1.E+00
Z on es
N u c l i d e G r o u p s
Perce
ntag
e of N
uclid
e Gro
ups t
o Init
ial In
vent
ories
[-]
Amount of iodine transferredper bubble volume: Experiment & Calculation
N u c l i d e G rZones i n reactor
Tran
sfer
red
iodi
ne (m
ol/m3 )
Distribution of FP nuclides in reactor 100s after cladding failure in PLOHS
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ARCADIAN-FBR codeFunctions: Neutronics for degraded coreSales point: The combination of (1)a deterministic calculation part to provide the various reactivity coefficients and those spatial distributionsand
(2) a Monte Carlo calculation partto provide reference results against the deterministic calculations and to evaluate approximation effects employed in the deterministic calculations
2.2 For the core disruption phase (1)
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Constitution of ARCADIAN-FBR
70707070----group Effective Cross Sections group Effective Cross Sections group Effective Cross Sections group Effective Cross Sections σσσσeffeffeffeff
JFSJFSJFSJFS----3333----J3.3J3.3J3.3J3.3
Cell Code SLAROMCell Code SLAROMCell Code SLAROMCell Code SLAROM
Kinetics ParameterKinetics ParameterKinetics ParameterKinetics Parameter
Calculation ModuleCalculation ModuleCalculation ModuleCalculation Module
Diffusion CodeDiffusion CodeDiffusion CodeDiffusion Code
DIF3DDIF3DDIF3DDIF3D
Transport CodeTransport CodeTransport CodeTransport Code
VARIANTVARIANTVARIANTVARIANT
kkkkeffeffeffeff
、、、、φφφφ、、、、φφφφ
****
kkkkeffeffeffeff
、、、、φφφφ、、、、φφφφ
****
JJJJENDLENDLENDLENDL----3.33.33.33.3
Monte Carlo CodeMonte Carlo CodeMonte Carlo CodeMonte Carlo Code
MVPMVPMVPMVP
kkkkeffeffeffeff
、、、、φφφφ
Perturbation Perturbation Perturbation Perturbation CodeCodeCodeCode
PERKY PERKY PERKY PERKY
ρρρρ、、、、 、、、、ββββ l
Monte Carlo CodeMonte Carlo CodeMonte Carlo CodeMonte Carlo Code
GMVPGMVPGMVPGMVP
kkkkeffeffeffeff
、、、、φφφφ、、、、φφφφ
****
Deterministic Calculation PartDeterministic Calculation PartDeterministic Calculation PartDeterministic Calculation Part MonteMonteMonteMonte CCCCarloarloarloarlo Calculation Part Calculation Part Calculation Part Calculation Part
、、、、ββββ lNNNN
Burnup CalBurnup CalBurnup CalBurnup Cal....
CodeCodeCodeCode REBUSREBUSREBUSREBUS
NNNN
Burnup CalBurnup CalBurnup CalBurnup Cal....
CodeCodeCodeCode MVPMVPMVPMVP----BURNBURNBURNBURN
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2.2 For the core disruption phase (2)APK codeFunctions: 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
0
1000
2000
0.000 0.005 0.010 0.015 0.020 0.025 0.030
time(s)
Pow
er(P0)
POWER-APK
POWER-sas4a
-1.00
-0.50
0.00
0.50
1.00
1.50
2.00
0.000 0.005 0.010 0.015 0.020 0.025 0.030
time(s)
reactivity($)
NET-APK
NET-sas4a
1000
1500
2000
2500
3000
3500
4000
0.000 0.005 0.010 0.015 0.020 0.025 0.030
time(s)
Tem
perature(K)
TEMP(PK)
TEP-sas4a
0
5
10
0.000 0.005 0.010 0.015 0.020 0.025 0.030
time(s)
energy(FP
S)
FPS-APK
FPS-sas4a
(a) Power history (b) Net reactivity (c) Average fuel temp. (d) Input energy
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2.3 For the CV response phaseAZORES codeFunctions: 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.
窒素雰囲気窒素雰囲気窒素雰囲気窒素雰囲気
炉 容 器炉 容 器炉 容 器炉 容 器
PHT SHTSSHTSSHTSSHTS
冷 却 系冷 却 系冷 却 系冷 却 系
炉炉炉炉
N
水素燃焼水素燃焼水素燃焼水素燃焼
PHT
HVA
補助建家補助建家補助建家補助建家
空気雰囲空気雰囲空気雰囲空気雰囲
FPFPFPFP
エアロゾルエアロゾルエアロゾルエアロゾル挙動挙動挙動挙動
格 納 容格 納 容格 納 容格 納 容
コアコンクリートコアコンクリートコアコンクリートコアコンクリート反応反応反応反応 NaNaNaNaコンクリートコンクリートコンクリートコンクリート反応反応反応反応
NaNaNaNa燃焼燃焼燃焼燃焼
NaNaNaNa燃焼燃焼燃焼燃焼
FPFPFPFP
Air atmosphere
炉心
Containment VesselHydrogen burning
Cooling
vessel
system
SodiumburningHVACSodiumburning
Core-concrete reactions reactionsSodium-concrete
Nitrogen atmosphere
FPaerosol
Auxiliary building
SHTSReactor
Core
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0
500
1000
1500
2000
2500
3000
3500
4000
4500
5000
0 5 10 15 20 25 30 35 40
Time (hour)
Core tem
perature (K)
Fig. 7-22(1) Core Temperature (Case 15: PLOHS, Non-CVBP, PL-NSL-H-γ)
0.0E+00
1.0E+05
2.0E+05
3.0E+05
4.0E+05
5.0E+05
6.0E+05
7.0E+05
8.0E+05
9.0E+05
1.0E+06
0 10 20 30 40
Time (hour)
Pressure (Pa)
Reactor chamber
PHTS chamber: AC Loop
HVAC room: AC Loop゚
PHTS chamber: B Loop゚
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(2). Pressure (Case 15: PLOHS, Non-CVBP PL-NRL-H-γ)
0.0
0.1
0.2
0.3
0.4
0.5
0.6
0.7
0.8
0.9
1.0
0 10 20 30 40
Time (hour)
Hydrogen m
ol fraction (-)
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-H-γ)
1E-10
1E-09
1E-08
1E-07
1E-06
1E-05
1E-04
1E-03
1E-02
1E-01
1E+00
0 10 20 30 40
Time (hour)
Xe (-)
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)
1E-10
1E-09
1E-08
1E-07
1E-06
1E-05
1E-04
1E-03
1E-02
1E-01
1E+00
0 10 20 30 40
Time (hour)
I (-)
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(2) I Gr Initial Inventory Ratio (Case 15, PLOHS, Non-CVBP PL-NRL-H-γ)
1E-10
1E-09
1E-08
1E-07
1E-06
1E-05
1E-04
1E-03
1E-02
1E-01
1E+00
0 10 20 30 40
Time (hour)
Ce (-)
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(3) Ce Gr Initial Inventory Ratio (Case 15:PLOHS, Non-CVBP PL-NRL-H-γ)
Example of CV response calculated by AZORES
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� 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.
3. Phenomenological Relation Diagram : PRDProbability decision method for the blanching
points in event trees (ET)
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-PRDPRDPRDPRD AppoachAppoachAppoachAppoach for the top event of acceleration -
Function gate [Motion equation: F=ma]
Events (outputs from the function gate)
Top event[ Acceleration: a]
Top eventConditionalbranch
Yes
No
Events[Force: F]
Sub-PRD [Mass: m]
Probability distribution function
Elementary event ②[F2]
Elementary event ①[F1]
Elementary event ③[F3]
To be expanded in detail in a lower PRD
Bottom 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
To be calculated by a certain function with lower events as inputs. The details of calculation are indicated in the adjoining blue fames.
-The top event that will be evaluated by the PRD approach. -[Acceleration a incidence probability distribution]
The probability distribution function to be given to a certain variable:Example)U[1,10]: A uniform distribution between 1 and 10N[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 - 100 ]
Incidence probability Incidence probability Incidence probability Incidence probability
distributiondistributiondistributiondistribution
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Evaluation procedures of ROAAM used LWR and PRD
図-1 RO AAM とPRD における評価手順
(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
●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。
●下位事象の相互関係をダイヤグラムに展開
●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
(2) 支配要因のパラメータ設定と確率分布の導入
(3) 決定論的知見の援用
(4) 着目現象に対する分岐確率の評価
●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
●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
●Iパラメータについて工学的判断による値を設定
●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
●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
PRD RO AAM
Fig.D1-1 Evaluation Procedures of ROAAM and PRD
(1) Selection of the dominant factors and development of diagram regarding the concerned phenomenon
● 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
(2) Establishment of parameters of the dominant factors and introduction of probability distribution
● Parameters, which describe the concerned phenomenon, are classified into three categories including deterministic (D), probabilistic (P) and intangible variables (I).
● 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
(3) Supplementary use of deterministic findings ● 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 ● 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 ● Evaluation of the branching probability of the concerned phenomenon, tracing the functional relation
upwards using the Monte Carlo method and others
ROAAM PRD
Both methods seem basically identical.
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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.
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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 high-temperature 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.
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FP leak passes in PLOHS/CVBP
P
Annulus ac SG
Na burn → aerosol generation pressure increase
CV
SHTS
NaNaNaNa leakage↓↓↓↓
NaNaNaNa/Concrete reaction↓↓↓↓
Hydrogen generation↓↓↓↓
Pressure increase
CVBP case
Non-CVBPCVBPCVBPCVBP
case
IHX
RVPHTS
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.
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Boundaries of IHX that have a high potential of first failure in PHTS
PHTSPHTSPHTSPHTS
coolant inletcoolant inletcoolant inletcoolant inlet
Bottom headBottom headBottom headBottom head
PHTS coolant outletPHTS coolant outletPHTS coolant outletPHTS coolant outlet
BellowsBellowsBellowsBellows
SHTSSHTSSHTSSHTS
coolant outletcoolant outletcoolant outletcoolant outlet
Heat Heat Heat Heat
transfer tubestransfer tubestransfer tubestransfer tubes
IHX Bottom Head
IHX bellows
Cover gasbellows
Bottom head (interface between
PHTS/SHTS)Diameter: 2190mmThickness:::: 25252525mmmmmmmm
Material: SUS304
Diameter: 700mmThickness:::: 1.5, 2.0 mmMaterial: SUS3316
Heat transfer tube
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An vent tree of PLOHS PowerPowerPowerPower
suppliessuppliessuppliessupplies
Leakage from theLeakage from theLeakage from theLeakage from the
shield plug of RVshield plug of RVshield plug of RVshield plug of RV
Eearlier secondaryEearlier secondaryEearlier secondaryEearlier secondary
system failuresystem failuresystem failuresystem failure
Failure in some areasFailure in some areasFailure in some areasFailure in some areas
of promary systemof promary systemof promary systemof promary system
((((aaaa))))
yesyesyesyes ↑↑↑↑
nononono ↓↓↓↓
0.250.250.250.25
((((bbbb))))
(c)(c)(c)(c)
0.960.960.960.96 ((((aaaa))))
PLOHSPLOHSPLOHSPLOHS
0.750.750.750.75 (d)(d)(d)(d)
((((bbbb))))
0.0440.0440.0440.044
CVBPCVBPCVBPCVBP
The probabilities of branching point (a), (b), (c),(d) are obtained through present analyses.
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2. Analyses to obtain the probability of the branching points
Failure mechanismFailure mechanismFailure mechanismFailure mechanism Objects for analysis Objects for analysis Objects for analysis Objects for analysis Analysis codeAnalysis codeAnalysis codeAnalysis code
High-temperature creepHigh-temperature creepHigh-temperature creepHigh-temperature creep Vessel, PipingVessel, PipingVessel, PipingVessel, Piping NALAP-IINALAP-IINALAP-IINALAP-II
BucklingBucklingBucklingBuckling ABAQUSABAQUSABAQUSABAQUS
Tensile (or breaking) stressTensile (or breaking) stressTensile (or breaking) stressTensile (or breaking) stress FINASFINASFINASFINAS
Bellows and bottomBellows and bottomBellows and bottomBellows and bottom
head of IHXhead of IHXhead of IHXhead of IHX
Thermal-hydraulicThermal-hydraulicThermal-hydraulicThermal-hydraulic AnalysisAnalysisAnalysisAnalysis
StructualStructualStructualStructual AnalysisAnalysisAnalysisAnalysis
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2.1 Thermal-hydraulic AnalysisFBR 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, Dc ).
� Dc=Σ(Δti/tr), (1)where,
Δti: time step (s), tr: creep failure time (s).
In eq. (1), tr depend on the material and it is calculated by using the Larson-Miller Parameter.It was considered that the location will fail when Dc=0.2~1.0.
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Specifications of proto-type LMFBR plant and locations where high-temperature creep was evaluated
� The locations with large value of D(diameter)/t(thickness) were chosen for the present calculation of SCDF.
Thermal powerThermal powerThermal powerThermal power 714 MW714 MW714 MW714 MW
Number of loopsNumber of loopsNumber of loopsNumber of loops 3333
Primary heat transferPrimary heat transferPrimary heat transferPrimary heat transfer
system (PHTS)system (PHTS)system (PHTS)system (PHTS)
Temperature at reactor Temperature at reactor Temperature at reactor Temperature at reactor
inlet/outletinlet/outletinlet/outletinlet/outlet
397397397397////529 529 529 529 ℃℃℃℃
Flow rate Flow rate Flow rate Flow rate 5100510051005100 tttt////hhhh
Pressure at reactor Pressure at reactor Pressure at reactor Pressure at reactor
inlet/outletinlet/outletinlet/outletinlet/outlet
0.8/0.1 MPa0.8/0.1 MPa0.8/0.1 MPa0.8/0.1 MPa
Main circulatiion pump Main circulatiion pump Main circulatiion pump Main circulatiion pump One unit per loopOne unit per loopOne unit per loopOne unit per loop
IHXIHXIHXIHX
Number of unit Number of unit Number of unit Number of unit 3333
Type Type Type Type
Vertical parallel flow withVertical parallel flow withVertical parallel flow withVertical parallel flow with
no sodium surfaceno sodium surfaceno sodium surfaceno sodium surface
Secondary heat transferSecondary heat transferSecondary heat transferSecondary heat transfer
system (SHTS)system (SHTS)system (SHTS)system (SHTS)
325325325325////529 529 529 529 ℃℃℃℃
Temperature at IHX Temperature at IHX Temperature at IHX Temperature at IHX
inlet/outletinlet/outletinlet/outletinlet/outlet
3700370037003700 tttt////hhhh
Flow rate Flow rate Flow rate Flow rate 0.8/0.1 MPa0.8/0.1 MPa0.8/0.1 MPa0.8/0.1 MPa
Main circulatiion pump Main circulatiion pump Main circulatiion pump Main circulatiion pump One unit per loopOne unit per loopOne unit per loopOne unit per loop
Auxiliary cooling systemAuxiliary cooling systemAuxiliary cooling systemAuxiliary cooling system
By-pass of the secondaryBy-pass of the secondaryBy-pass of the secondaryBy-pass of the secondary
cooling system with onecooling system with onecooling system with onecooling system with one
air cooler per loopair cooler per loopair cooler per loopair cooler per loop
D (inner dia .) D ( inner dia .) D ( inner dia .) D ( inner dia .) t (thickness)t (thickness)t (thickness)t (thickness)
(mm)(mm)(mm)(mm) (mm)(mm)(mm)(mm)
RV cylindrical wallRV cylindrical wallRV cylindrical wallRV cylindrical wall 7100710071007100 50505050 142142142142 SUS304 stainless steelSUS304 stainless steelSUS304 stainless steelSUS304 stainless steel
Reactor outlet pipingReactor outlet pipingReactor outlet pipingReactor outlet piping 788788788788 11111111 72727272 SUS304 stainless steelSUS304 stainless steelSUS304 stainless steelSUS304 stainless steel
Reactor inlet pipingReactor inlet pipingReactor inlet pipingReactor inlet piping 591591591591 9.59.59.59.5 62626262 SUS304 stainless steelSUS304 stainless steelSUS304 stainless steelSUS304 stainless steel
IHIHIHIHX vessel cylindrical wallX vessel cylindrical wallX vessel cylindrical wallX vessel cylindrical wall 2940294029402940 30303030 98989898 SUS304 stainless steelSUS304 stainless steelSUS304 stainless steelSUS304 stainless steel
IHX heat transfer tubeIHX heat transfer tubeIHX heat transfer tubeIHX heat transfer tube 19.319.319.319.3 1.21.21.21.2 16161616 SUS304 stainless steelSUS304 stainless steelSUS304 stainless steelSUS304 stainless steel
IHX SHTS-side outletIHX SHTS-side outletIHX SHTS-side outletIHX SHTS-side outlet
pipingpipingpipingpiping
540540540540 9.59.59.59.5 56565656 SUS304 stainless steelSUS304 stainless steelSUS304 stainless steelSUS304 stainless steel
EV cylindrical wallEV cylindrical wallEV cylindrical wallEV cylindrical wall 2900290029002900 50505050 58585858 2.25Cr-1Mo steel2.25Cr-1Mo steel2.25Cr-1Mo steel2.25Cr-1Mo steel
Air-cooler heat transferAir-cooler heat transferAir-cooler heat transferAir-cooler heat transfer
tubetubetubetube
45454545 2.92.92.92.9 16161616 SUS304 stainless steelSUS304 stainless steelSUS304 stainless steelSUS304 stainless steel
Fuel pin claddingFuel pin claddingFuel pin claddingFuel pin cladding 5.565.565.565.56 0.470.470.470.47 11.811.811.811.8
20% cold-worked modi f ied20% cold-worked modi f ied20% cold-worked modi f ied20% cold-worked modi f ied
316 stainless steel316 stainless steel316 stainless steel316 stainless steel
LocationLocationLocationLocation MaterialMaterialMaterialMaterialD/tD/tD/tD/t
(a) Spec. of LMFBR (b) Locations of plant evaluated high-temperature creep
26
Japan Nuclear Energy Safety Organization
IHXIHXIHXIHX下部鏡下部鏡下部鏡下部鏡((((SUSSUSSUSSUS304304304304))))のののの座屈限界応力座屈限界応力座屈限界応力座屈限界応力とととと破断限界応力破断限界応力破断限界応力破断限界応力
0000
50505050
100100100100
150150150150
200200200200
250250250250
300300300300
800800800800 810810810810 820820820820 830830830830 840840840840 850850850850 860860860860 870870870870 880880880880 890890890890 900900900900
1111次系次系次系次系NaNaNaNaののののIHXIHXIHXIHX出口温度出口温度出口温度出口温度((((℃℃℃℃))))
限界応力
限界応力
限界応力
限界応力(( ((MPa
MPa
MPa
MPa)) ))
破断限界応力破断限界応力破断限界応力破断限界応力((((σσσσFFFF))))
((((200000200000200000200000ssss、、、、846846846846℃℃℃℃、、、、100100100100MPaMPaMPaMPa))))
IHXIHXIHXIHX下部鏡下部鏡下部鏡下部鏡のののの応力応力応力応力とととと圧力圧力圧力圧力のののの関係式関係式関係式関係式
σ(MPa)=θ×a×(P-0.1)σ(MPa)=θ×a×(P-0.1)σ(MPa)=θ×a×(P-0.1)σ(MPa)=θ×a×(P-0.1)
aaaa====89898989....89898989、、、、
PPPP((((MPaMPaMPaMPa))))::::1111次系圧力次系圧力次系圧力次系圧力
θθθθ::::近似式近似式近似式近似式のののの不確定幅不確定幅不確定幅不確定幅
確率変数確率変数確率変数確率変数((((0000....95959595~~~~1111....05050505))))
座屈限界応力座屈限界応力座屈限界応力座屈限界応力((((σσσσBBBB))))
::::ABAQUSABAQUSABAQUSABAQUSのののの結果結果結果結果((((座屈座屈座屈座屈))))
((((195000195000195000195000ssss、、、、838838838838℃℃℃℃、、、、80808080MPaMPaMPaMPa))))
Buckling critical stress and fracture critical stress in IHX bottom head (SUS304)Stress-pressure relationship equation in IHX bottom headσ(MPa)=θ×a×(P-0.1)a=89.89,P(MPa): Primary system pressure θ: A margin of uncertainties in an approximate expression
Probability variable (0.95~1.05)
Fracture critical stress (σF)
Buckling critical stress (σB)Critic
al str
ess (
MPa)
Results of the ABAQUS-based analyses (backling)
IHXIHXIHXIHX下部鏡下部鏡下部鏡下部鏡((((SUSSUSSUSSUS304304304304))))のののの座屈限界応力座屈限界応力座屈限界応力座屈限界応力とととと破断限界応力破断限界応力破断限界応力破断限界応力
0000
50505050
100100100100
150150150150
200200200200
250250250250
300300300300
800800800800 810810810810 820820820820 830830830830 840840840840 850850850850 860860860860 870870870870 880880880880 890890890890 900900900900
1111次系次系次系次系NaNaNaNaののののIHXIHXIHXIHX出口温度出口温度出口温度出口温度((((℃℃℃℃))))
限界応力
限界応力
限界応力
限界応力(( ((MPa
MPa
MPa
MPa)) ))
破断限界応力破断限界応力破断限界応力破断限界応力((((σσσσFFFF))))
((((200000200000200000200000ssss、、、、846846846846℃℃℃℃、、、、100100100100MPaMPaMPaMPa))))
IHXIHXIHXIHX下部鏡下部鏡下部鏡下部鏡のののの応力応力応力応力とととと圧力圧力圧力圧力のののの関係式関係式関係式関係式
σ(MPa)=θ×a×(P-0.1)σ(MPa)=θ×a×(P-0.1)σ(MPa)=θ×a×(P-0.1)σ(MPa)=θ×a×(P-0.1)
aaaa====89898989....89898989、、、、
PPPP((((MPaMPaMPaMPa))))::::1111次系圧力次系圧力次系圧力次系圧力
θθθθ::::近似式近似式近似式近似式のののの不確定幅不確定幅不確定幅不確定幅
確率変数確率変数確率変数確率変数((((0000....95959595~~~~1111....05050505))))
座屈限界応力座屈限界応力座屈限界応力座屈限界応力((((σσσσBBBB))))
::::ABAQUSABAQUSABAQUSABAQUSのののの結果結果結果結果((((座屈座屈座屈座屈))))
((((195000195000195000195000ssss、、、、838838838838℃℃℃℃、、、、80808080MPaMPaMPaMPa))))
Buckling critical stress and fracture critical stress in IHX bottom head (SUS304)Stress-pressure relationship equation in IHX bottom headσ(MPa)=θ×a×(P-0.1)a=89.89,P(MPa): Primary system pressure θ: A margin of uncertainties in an approximate expression
Probability variable (0.95~1.05)
Fracture critical stress (σF)
Buckling critical stress (σB)Critic
al str
ess (
MPa)
Results of the ABAQUS-based analyses (backling)
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: 838.7℃℃℃℃; Pressure: 0.89MPa)
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.
IHX exit temperature (℃)
Results of ABAQUS-based analyses (buckling)
27
Japan Nuclear Energy Safety Organization
The results of FINAS 2-d analysis to IHX bellows - Yield stress and tensile stress -
IHXIHXIHXIHXベローズベローズベローズベローズ((((SUSSUSSUSSUS316316316316))))のののの座屈座屈座屈座屈・・・・破断限界応力破断限界応力破断限界応力破断限界応力
0000
50505050
100100100100
150150150150
200200200200
250250250250
300300300300
350350350350
400400400400
700700700700 720720720720 740740740740 760760760760 780780780780 800800800800 820820820820 840840840840 860860860860 880880880880 900900900900
1111次系次系次系次系NaNaNaNaののののIHXIHXIHXIHX出口温度出口温度出口温度出口温度((((℃℃℃℃))))
限界
応力
限界
応力
限界
応力
限界
応力
(( ((MPa
MPa
MPa
MPa)) ))
降伏応力降伏応力降伏応力降伏応力
引張引張引張引張((((破断破断破断破断))))応力応力応力応力
((((189000189000189000189000ssss、、、、828828828828℃℃℃℃、、、、97979797....4444MPaMPaMPaMPa))))
((((198000198000198000198000ssss、、、、843843843843℃℃℃℃、、、、139139139139MPaMPaMPaMPa))))
IHXIHXIHXIHXベローズベローズベローズベローズ応力応力応力応力
σ(MPa)=φ×a×(P-0.1)σ(MPa)=φ×a×(P-0.1)σ(MPa)=φ×a×(P-0.1)σ(MPa)=φ×a×(P-0.1)
aaaa====136136136136....25252525、、、、
PPPP((((MPaMPaMPaMPa))))::::1111次系圧力次系圧力次系圧力次系圧力((((MPaMPaMPaMPa))))
φφφφ::::応力近似式応力近似式応力近似式応力近似式のののの不確定幅不確定幅不確定幅不確定幅
確率変数確率変数確率変数確率変数((((0000....95959595~~~~1111....05050505))))
Yield stress
Tensile (or breaking) stress (MPa)
Critic
al str
ess (
MPa)
IHX bellows stress-pressure σ(MPa)=φ×a×(P-0.1)a=136.25P(MPa): Primary system Φ:A margin of uncertainties
IHX bellows (SUS316) buckling critical stress and fracture critical stress
relationship equation
pressure in an approximateexpression (0.95~1.05)
(189000s、828℃、97.4MPa)
(198000s、843℃、
139MPa)
IHX outlet temperature of primary-system sodium
IHXIHXIHXIHXベローズベローズベローズベローズ((((SUSSUSSUSSUS316316316316))))のののの座屈座屈座屈座屈・・・・破断限界応力破断限界応力破断限界応力破断限界応力
0000
50505050
100100100100
150150150150
200200200200
250250250250
300300300300
350350350350
400400400400
700700700700 720720720720 740740740740 760760760760 780780780780 800800800800 820820820820 840840840840 860860860860 880880880880 900900900900
1111次系次系次系次系NaNaNaNaののののIHXIHXIHXIHX出口温度出口温度出口温度出口温度((((℃℃℃℃))))
限界
応力
限界
応力
限界
応力
限界
応力
(( ((MPa
MPa
MPa
MPa)) ))
降伏応力降伏応力降伏応力降伏応力
引張引張引張引張((((破断破断破断破断))))応力応力応力応力
((((189000189000189000189000ssss、、、、828828828828℃℃℃℃、、、、97979797....4444MPaMPaMPaMPa))))
((((198000198000198000198000ssss、、、、843843843843℃℃℃℃、、、、139139139139MPaMPaMPaMPa))))
IHXIHXIHXIHXベローズベローズベローズベローズ応力応力応力応力
σ(MPa)=φ×a×(P-0.1)σ(MPa)=φ×a×(P-0.1)σ(MPa)=φ×a×(P-0.1)σ(MPa)=φ×a×(P-0.1)
aaaa====136136136136....25252525、、、、
PPPP((((MPaMPaMPaMPa))))::::1111次系圧力次系圧力次系圧力次系圧力((((MPaMPaMPaMPa))))
φφφφ::::応力近似式応力近似式応力近似式応力近似式のののの不確定幅不確定幅不確定幅不確定幅
確率変数確率変数確率変数確率変数((((0000....95959595~~~~1111....05050505))))
Yield stress
Tensile (or breaking) stress (MPa)
Critic
al str
ess (
MPa)
IHX bellows stress-pressure σ(MPa)=φ×a×(P-0.1)a=136.25P(MPa): Primary system Φ:A margin of uncertainties
IHX bellows (SUS316) buckling critical stress and fracture critical stress
relationship equation
pressure in an approximateexpression (0.95~1.05)
(189000s、828℃、97.4MPa)
(198000s、843℃、
139MPa)
IHX outlet temperature of primary-system sodium
Deformed IHX bellows (Temperature: 750℃℃℃℃; Pressure: 0.388 MPa) (The results of the FINAS 3-d calculation
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
Japan Nuclear Energy Safety Organization
The Range of Primary System Component Fracture Pressures
0
0.5
1
1.5
2
2.5
3
0 0.2 0.4 0.6 0.8 1 1.2
Names of components
Pres
sure
(MPa
)
RV cylindricalwall
RV outlet piping IHX cylindricalwall
RV inlet piping IHX bellows IHX bottom head
0.768M
2.316MPa
1.476MPa 1.165MPa
1.928MPa
0.928MPa1.117MPa
0.768
3.8MPa
2.6MPa
0.961MPa1.075MPa
3. DiscussionPressure range for failure of PHTS components
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
Japan Nuclear Energy Safety Organization
Time range for failure of PHTS components
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
Japan Nuclear Energy Safety Organization
4. Conclusion(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, itwas concluded that the probability of CVBP was 0.2 to the plant.
31
Japan Nuclear Energy Safety Organization
Appendix –additional information
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.
PowerPowerPowerPower
suppliessuppliessuppliessupplies
Leakage from theLeakage from theLeakage from theLeakage from the
shield plug of RVshield plug of RVshield plug of RVshield plug of RV
Eearlier secondaryEearlier secondaryEearlier secondaryEearlier secondary
system failuresystem failuresystem failuresystem failure
Failure in some areasFailure in some areasFailure in some areasFailure in some areas
of promary systemof promary systemof promary systemof promary system
((((aaaa))))
yesyesyesyes ↑↑↑↑
nononono ↓↓↓↓
0.250.250.250.25
((((bbbb))))
0.80.80.80.8
0.960.960.960.96 1.01.01.01.0
PLOHSPLOHSPLOHSPLOHS
0.750.750.750.750.20.20.20.2
0.00.00.00.0
0.0440.0440.0440.044
CVBPCVBPCVBPCVBP
1.00E-14
1.00E-13
1.00E-12
1.00E-11
1.00E-10
1.00E-09
1.00E-08
1.00E-07
Loss of
annulus
function
RV isolaiton
failure
RV failure by
hydrogen
burning
RV failure by
sodium
burning
CVBP Retained RV
integrity
CFF F
requ
ency
(1/R
Y)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
These numbers were assigned from present study
33
Japan Nuclear Energy Safety Organization
Flow network of the NALAP-II code for RV and PHTS
RV flow networkRV flow networkRV flow networkRV flow network
PHTS flow networkPHTS flow networkPHTS flow networkPHTS flow network
IHX
PUMP
R/V
R/V
H/L
M/L C/L
Fig.5 Flow network of the NALAP-II code for RV and PHTS
Cold legMiddle legHot leg
Pump
R/V
R/V
IHXCover gas
Inner
vessel
Core channel
Inner core Ch.
Outer core Ch,
Radial blanket Ch,
Shield Ch,
Bypass Ch,
Primary loop
Primary
loop
Maintenance cooling
system
Maintenance cooling
system
Lower plenum
UCS
Flow hole
Outlet
plenum
RV flow networkRV flow networkRV flow networkRV flow network
PHTS flow networkPHTS flow networkPHTS flow networkPHTS flow network
IHX
PUMP
R/V
R/V
H/L
M/L C/L
PHTS flow networkPHTS flow networkPHTS flow networkPHTS flow network
IHX
PUMP
R/V
R/V
H/L
M/L C/L
PHTS flow networkPHTS flow networkPHTS flow networkPHTS flow network
IHX
PUMP
R/V
R/V
H/L
M/L C/L
IHX
PUMP
R/V
R/V
H/L
M/L C/L
Fig.5 Flow network of the NALAP-II code for RV and PHTS
Cold legMiddle legHot leg
Pump
R/V
R/V
IHXCover gas
Inner
vessel
Core channel
Inner core Ch.
Outer core Ch,
Radial blanket Ch,
Shield Ch,
Bypass Ch,
Primary loop
Primary
loop
Maintenance cooling
system
Maintenance cooling
system
Lower plenum
UCS
Flow hole
Outlet
plenum
34
Japan Nuclear Energy Safety Organization
IHX inlet and outlet temperatures, and PHTS flow rate
300
350
400
450
500
550
600
650
700
750
800
850
900
950
0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22
時間(×10,000秒)
温度(℃)
0
1
2
3
4
5
6
7
8
9
10
流量比(%)
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
35
Japan Nuclear Energy Safety Organization
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)�
-20.0
0.0
20.0
40.0
60.0
80.0
100.0
120.0
140.0
160.0
0 10000 20000 30000 40000 50000 60000
時間(秒)
応力
(N/mm
2
)
0.0
100.0
200.0
300.0
400.0
500.0
600.0
700.0
800.0
900.0
温度
(℃
)
ミーゼス応力
周方向応力
子午線方向応力
温度
Time (s)
Stress (N/㎟)
Temperature (℃)
Mises stressCircumferential stressStress in the meridiandirectionTemperature
36
Japan Nuclear Energy Safety Organization
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)
-20.0
0.0
20.0
40.0
60.0
80.0
100.0
120.0
140.0
160.0
0 10000 20000 30000 40000 50000 60000
時間(秒)
応力
(N/mm
2
)
0.0
100.0
200.0
300.0
400.0
500.0
600.0
700.0
800.0
900.0
温度
(℃
)
ミーゼス応力
周方向応力
子午線方向応力
温度
Time (s)
Stress (N/㎟)
Temperature (℃)
Mises stressCircumferential stressStress in the meridiandirectionTemperature
37
Japan Nuclear Energy Safety Organization
Comparison of initiating events frequency between JAEA and JNES
37
38
Japan Nuclear Energy Safety Organization
4. Results of PSA4.1 Effects of AM measures to keep the RV coolant level
10
-12
10
-11
10
-10
10
-9
10
-8
10
-7
10
-6
炉心
損傷
頻度
(/炉
年)
1次主冷却系
循環ポンプ
トリップ失敗
原子炉
カバーガス
供給停止
失敗
1次ナトリウム
の原子炉
容器への
汲み上げ
失敗
1次メンテナンス
冷却系漏えい
箇所の隔離
失敗
ガードベッセル
の健全性
喪失
1次主冷却系
漏えい後の
二重漏えい
(SsLより下端)
LORL-全体
■:AM前
■:AM後
Core Damage
Frequencies / Reactor
Year
Before AM MeasuresAfter AM Measures
Primary main coolant system circulation pump trip failureReactor cover-gas feeding cut failure Keeping
sodiumtemperature
high
A failure to isolate the leakage from the primary maintenance cooling system
Loss of guard vessel integrityDouble leak (in an area below SsL) following leakage from the primary main coolant system
All loss of reactor level (LORL) eventIsolation of argon supply line to RV
Siphon break of maintenance cooling system
39
Japan Nuclear Energy Safety Organization
4.2 Effects of initiating events frequency
1.0E-09
1.0E-08
1.0E-07
1.0E-06
Basic Conditions Reference Analysis 5 Basic Conditions Reference Analysis 5
Loss of reactor shutdown functions
Loss of functions to maintain a proper reactor coolant level
Loss of decay heat removal functions
All relevant events
JNES JAEA JNES JAEAbefore AM after AM
Frequency(/ry)
40
Japan Nuclear Energy Safety Organization
1.0E-09
1.0E-08
1.0E-07
1.0E-06
Basic Conditions Reference Analysis 7 Basic Conditions Reference Analysis 7
Loss of reactor shutdown functions
Loss of functions to maintain a proper reactor coolant level
Loss of decay heat removal functions
All relevant event
Before After AM
2.7 times3.8 times
ATWS
LORL
PLOHS
AllAll All
AllATWS
LORL
PLOHSATWSincreases
No AM Measures for ATWS
4.3 Effects of choosing common cause failure
JNES JAEA JNES JAEAbefore AM after AM
Frequency(/ry)
(In the PSA of JAEA, common cause failure is considered between the main reactor shutdown system and the backup reactor shutdown system)
41
Japan Nuclear Energy Safety Organization
4.4 Contribution of initiating events to CDF(b) after AM (by(by(by(by JNES))))
1.6E-07
1.0E-13
1.0E-12
1.0E-11
1.0E-10
1.0E-09
1.0E-08
1.0E-07
1.0E-06
1.0E-05
1.0E-04
PLOHS LORL ULOF UTOP ULOHS ULOPI UTOP/ULOF Full core damage
frequencies
Core
Dam
age
Freq
uenc
ies/R
eacto
r Yea
r
PLOHS24%
LORL76%
ULOF0%
UTOP0%
ULOPI0%
ULOHS0%
UTOP/ULOF0%
Full core damage frequency:1.6×10-7/reactor year
ATWS
42
Japan Nuclear Energy Safety Organization
1.0E-09
1.0E-08
1.0E-07
1.0E-06
basal condition
-beforeAM
basal condition
-after AM
JAEA-before-AM JAEA--after AM
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
JAEAJNES
0.6倍
0.3倍
ATWS
LORL
PLOHS
全体
全体
全体
全体
ATWS
LORL
PLOHS
4.5 Comparison of estimated CDF between JNES and JAEA
42
Frequency(/ry)
Total Total
Total
Total
0.6 times 0.3 times