PSA for New Nuclear Power Plants in India: Review Insights ... · PSA for New Nuclear Power Plants in India: Review Insights and Some Relevant Issues AvinashJ.Gaikwad October, 2012

PSA for New Nuclear Power Plants in India: Review Insights and Some

Relevant Issues

Avinash J.Gaikwad

October, 2012 IAEA-TM, Vienna,

Atomic Energy Regulatory Board, INDIA

PSA of Indian Operating & New NPPs

• Level-1 PSA (internal events, full power) is performed for all NPPs. Thisincludes two Boling Water Reactors (BWRs) and eighteen PressurisedHeavy water Reactors (PHWRs).

• NEW REACTORS: Currently, two 700 MWe PHWRs, two PressurisedWater Reactors (PWRs) of VVER type design and two Prototype FastBreeder Reactors (PFBRs) are under construction/proposed.

• Regulatory review ofVVER type PWRs is also completed.

• Decay Heat Removal & S/D reliability analysis review andThe PSA review for 500 MWe PFBRs is in progress.

• PSA review for 700 MWe PHWRs is not yet taken up forreview

2October, 2012 IAEA-TM, Vienna,

3

Steam Generator

Primary Pumps

Headers Headers

Calandria

Moderator Pump Moderator

HX

Pre

ssu

rise

r

Steam line

PDHRPDHR

Steam Generator

Primary Pumps

Headers Headers

Calandria


HX

Pre

ssu

rise

r

Steam line

PDHRPDHR

PHWR

PSA in Indian NPP

• PSA insights are used for demonstration of safety and identification ofdesign vulnerabilities for NPPs, if any.

• During PSA review, the Technical adequacy of the PSA elements such asinitiating event analysis, system modeling, human reliability analysis,common cause failure analysis and data assessment (sample) are checked.

• While the current PSA methodology can be used for new reactors, thereare some aspects where new methods, models and computational toolsare required.

• New reactor designs employ many passive safety features and digitalcontrol and instrumentation in safety systems, which require differentapproach for estimating their reliability.

• Carrying out Level-1 PSA for internal events at full power is a mandatoryrequirement on PSA

• Level-1 PSAs are performed as per the procedure given in IAEA-SS-50-P4,AERB Doc on PSA, HRA & Components Data.


� 1000 MWe VVER type PWR is a light water cooled, light watermoderated reactor. It consists of reactor pressure vessel (RPV),primary and secondary circuits and auxiliary systems.

� Reactor consists of both active and passive safety systems/features.Some of these safety features are engineered first of its kind forVVER such as passive quick boron injection system, Air cooledpassive heat removal system, core catcher, etc.

� Systems such as residual heat removal system, high-pressure andlow-pressure ECC injection systems, first stage and second stagehydro-accumulators are incorporated to cool the reactor core incase of loss of coolant accidents.

� In case of reactor emergency protection system failure, passivequick boron injection system and active emergency boron injectionsystem are incorporated to achieve the sub-criticality and tomaintain it.These provision cater to ATWS.

5

Introduction: KK VVER1000


March 26, 2007 Severe Accident Analysis SADD, AERB 6

Introduction

KK-NPP: Two Units, Each of 1000 MWe Pressurized Water Reactors

�Light water coolant and light water moderator reactor.

�Four primary and secondary loops. Each loop consisting of one RCP andone SG.

�Pressuriser is connected to one of the hot legs. Two safety PSDs & onecontrol PSD are mounted on pressuriser.

�Coolant with the help of reactor coolant pumps is supplied into the reactorthrough the inlet nozzles then it goes down along the annulus between thevessel and the core barrel and from below through the perforated ellipticbottom of the core barrel and the holes in the supporting tubes it entersFAs.

�Light water under pressure of 15.7 MPa containing boron,which is neutronabsorber, is used as coolant and moderator in the core.

�Engineered Safety Features (ESFs) –To Address

�Design Basis Accidents (DBA)

�Beyond Design Basis Accidents (BDBA)

�Anticipated Transients Without Scram (ATWS)


Fuel Assembly Cartogram

158 159 160 161 162 163

149 150 151 152 153 154 156 157155

139 140 141 142 143 144 146 147145 148

129 130 131 132 133 134 136 137135 138129

116 117 118 119 120 121 123 124122 125 126 127

103 104 105 106 107 108 110 111109 112 114113 115

89 90 91 92 93 94 96 9795 98 10099 101 102

76 77 78 79 80 81 83 8482 85 8786 88

62 63 64 65 66 67 69 7068 71 7372 74 75

49 50 51 52 53 54 56 5755 58 6059 61

26 27 28 29 30 31 32 34 3533 36

38 39 40 41 42 43 45 4644 47 4837

16 17 18 19 20 21 23 2422 25

7 8 9 10 11 12 14 1513

1 2 3 4 5 6

FAs: Hexagonal, 163

Fuel elements: Triangular, 311

Number of guide tubes: 18

Number of spacing grids: 15

Instrumentation channel: 1

Central channel: 1

A part of FAs contains absorber rods of CPS or BARs.

All fuel rods and guiding channels of FA are kept apart by spacing and end grids.

C/S


Introudction-VVER

� The reactor represents a vertical high pressurevessel inside of which core barrel with corebaffle, protective tube unit, fuel assembliesforming the core and control rods connectedwith extension shafts of drive movement unitsare arranged.

� Elliptical bottom

� Core barrel is made in the form of weldedcylindrical shell.

� The perforated elliptic bottom of the core barreltogether with 163 perforated supporting tubesand the spacing grid form a structure forsupporting and spacing FAs.

� The core barrel at the level of the core has acore baffle. The core baffle fulfils function of aprotective shield.

Active Passive Safety Systems-VVER

Active safety systems Passive safety systems

1. CPSAR: control and protection system.

2. High pressure emergency injection system (JND 10-40) (Set point Pprim < 7.8 MPa).

3. Low pressure emergency injection and long term recirculation (JNA 10-40) (Set point Pprim < 2.5 MPa).

4. SG emergency cooling and blow down system (ECD) (JNB 10-40).

5. Emergency boron injection system (ATWS and SG tube failure)

1. Hydro Accumulators –stage 1 (JNG10-40) (Set point Pprim<5.89 MPa). 2. Hydro accumulators –stage 2 (JNG 50-80) (Set point Pprim<1.5 MPa). 3. Passive Heat Removal System

(PHRS). It cools the core for 24 hrs in severe accident condition.

4. Quick boron injection system. It operates during ATWS.

5. Ex-vessel core catcher keeps the molten pool and debris of core cool for unlimited period of time. 6. PARs for hydrogen management

10

1 Reactor2 Steam generator3 Reactor circulatingpump4 Pressuriser5 Pressuriser relieftank6 Accumulator

REVIEW INSIGHTS: KK VVER1000


11

KK VVER1000 Schematic

Loop 4 Loop 1

RCPRCP

RCP

Hot Leg

Cold Leg

Cold Leg

Loop 2

Hot Leg

Cold Leg

Loop 3

Hot Leg

Cold Leg

Hot Leg

JAA10 JND

20

JND60

JND

40

JND80 JND50

JND

30

JND70

JNG

40

RCP

JNG80

JNG

30JNG70

JNG

20

JNG60

JNG

10

JNG50

JND

10

SG

SG SG

SG

PRZ

JNA

20

JNA

40

JNA30

JNA10

JNA10

JNA30

IAEA Technical Meeting, Vienna, October 01-05, 2012

� Initiating events are selected using the approachessuggested by IAEA-50-P-4. The completeness of the PSAin terms of exhaustiveness of the IEs considered isverified with the ‘IE list used for VVERs- IAEATECDOC-719’

� The events having similar plant response (i.e. relevantmitigating systems, operator actions) and similar successcriteria are grouped together. The operating experienceis used for estimation of initiating event frequency.

� The events such as (i) Rupture of the reactor vessel (ii)Rupture of the steam generator header and (iii) Ruptureof the steam line with rupture of one of the SG tubesare considered as beyond design basis events. The totalcontribution form all these IEs amounts to 53% of theoverall CDF

12

IE : KK VVER1000-IE


13

REVIEW INSIGHTS: PSA FOR NEW REACTORS


1 Small LOCA2 SGTubeRupture3 Loss of offsitePower4 ServiceWater Syst5 SafetyInjection LOCA6 large LOCA7 Gen Transient8 lossOfNormalHeat Removal9 Pulsed SafetyDivice Stuck open

Leakage of the steam line in a section isolated from the steam

generator

Leakage of the steam line in a

section not isolated from the steam

generator

Other IE

Loss of off-site power with reactor

depressurised

Small leakage of the primary circuit inside containment

Small leakage of the primary circuit

outside containment

Large leakage of the primary circuit

Leakage from the primary circuit to the secondary

circuit, conventional

diameter is 13 mm

Leakage from the primary circuit to the secondary

circuit. Equivalent diameter is 100

mm

IE contribution to CDF : PSAR KK-VVER1000

15

LOOP-SSWS

GT

LOOP

LNHR

NSL

SGCL

SGTR

SILOCA-L

IPSV

SL2

SL1

OTHER Ies

IEs contribution to CDF

BDBIE

SGTR

SL2

SL1

LOOP

LOOP-S

SWSSI

OTHER IEs

IE contribution to CDF with BDBEs

REVIEW INSIGHTS: FSAR KK-VVER1000

1 Small LOCA2 SGTube Rupture3 Loss of offsite Power4 Service Water Syst5 Safety Injection LOCA6 large LOCA7 Gen Transient8 loss of NormalHeat Removal9 Pulsed SafetyDevice Stuck open

� Small LOCA IE contributes more towards CDF dueto the fact that there are common parts for highpressure and low pressure ECCS and the operatorerror in valving in passive heat removal system(PHRS) in cooling down mode. More no of smallerpipes & small breaks in large pipes. Event frequency.

� Steam generator tube rupture (SGTR) event, it wasconservatively assumed to result in core damage ifthere are subsequent failures in emergency cooldown (EHRS) and PHRS.

� Due to mutually redundant active and passivesystems for heat removal through secondary circuit,the contribution to core damage from generaltransient is quite low.

16



� The necessary safety functions and relevantmitigation systems/support systems,operator actions are identified.

� The accident sequence modeling is based ondevelopment of functional event trees. Theoperator actions are also modeled in theevent trees wherever applicable.

� The mission time for the safety systemcomponents considered to be of 24 hours.

� The recovery actions are not credited in theaccident sequence analysis

17

Features of PSA FOR NEW REACTORS


� The fault tree models are developed such thatthey reflect the ‘as-built’ and ‘as-operated’systems using the ‘immediate cause concept’ asrecommended in NUREG-0492

� The system success criteria and mission timesare derived based on the plant-specific thermal-hydraulic analyses.

� The test and maintenance related unavailabilityare accounted appropriately and are consistentwith technical specification requirements.

18

Features : PSA FOR NEW REACTORS


19

FusselVesely Syst Importance KK-VVER PSA


1 Passive Heat Removal2 JND ECCS3 Power Supply4 JNB SG ECoolDBlowDown5 QKB essential chilled watersystem6 PE Service Water Syst7 Reactor Protection Syst8 SAC01 Ventilation Syst7 JNA Low P ECCS8 LBA Isolation of Steam & Feed Water QKB

� The most important safety systems are PHRS andhigh pressure ECCS borated water injectionsystem (JND).

� The higher importance of essential power supplysystem (PS) is attributed to its supportive role tomany frontline safety systems (DG).

� The relatively high importance of steam generatorsblow down and emergency cool down system(JNB) is due to the fact that this is the normal heatremoval path for most of the general transientsand accident conditions.


FusselVesely Syst Importance KK-VVER PSA

� Intra-system common cause failures (CCFs) areconsidered using the α-factor models

� α-factors are taken from the generic databaseprovided in NUREG/CR-5497 and NUREG/CR-5485

21

CCF : PSA FOR NEW REACTORS


22

FV of CCF Groups : PSA FOR NEW REACTORS


1 PS Power Supply2 CIV containment isolation3 JND H P ECCS4 Damper C&I5 LBA steam isolation

� The important safety systems are PHRS, JND and PS as indicated in figure , hence, the CCF contributions are also expected to be from the component associated with these systems.

23



� The import human actions are identified throughsystematic process. The guidelines suggested in IAEASS 50-P-10 have been followed. For predictinghuman error probabilities (HEPs) for pre-initiatorhuman actions, a Technique for Human Error RatePrediction (THERP) is used. The HEPs for pre-initiator human actions were found to be riskinsignificant and not considered in fault trees.

� Human error probabilities are represented as thescreening values obtained from THERP, AccidentSequence Evaluation Program (ASEP) which areconservative.

24



25

Fussel-Vesely Importance Measure for Human Actions KK-VVER


Switching onPHRS in

cooldownmode

ConnectingHA2 to

primary inshutdown

mode

Leak isolation Dianosis of theevent and

action

Switching offJND system

Leak isolationby closing PRZ

PSD

Switching onJNA30 atshutdown

mode

Swithing onJNA40 atshutdown

mode

Fussel-Vesely Importance Measure for Human Actions

7.25MPa SGP

LOCA+ECCSFailure No P

1.5MPa HAP

Primary to SG leakisolation

Primary LeakOutsideContainmentInstuLINEsHookupS/Dcoolingline

S/D cooling trainsFailothers2BvalvedINMannuallyThen HA2

� The component failure data used in the PSA arebased on theVVER operating experience.

� Wherever such data are not available, theprobabilistic fracture mechanics and fault treeapproach has been used for parameterestimations.

� For example, reactor pressure vessel rupture orlarge and medium LOCA initiating eventfrequencies were estimated based oncharacteristics of materials, number and length ofwelds, quality assurance, dynamic loads, etc.

26

Data IE & component failure : PSA KK VVER


� PFBR is а 500 MWe, sodium cooled, pool type,mixed oxide (МОХ) fuelled reactor with twosecondary loops.

� Entire primary sodium circuit consisting of core,primary sodium pumps, and intermediate heatexchangers (IHX) and primary pipe connecting thepumps and the grid plate is contained in а singlelarge diameter vessel called main vessel.

� The hot primary sodium is radioactive and is notused directly to produce steam, but transfers theheat to secondary sodium through four IHX. Thenon radioactive secondary sodium is circulatedthrough two independent secondary loops, eachhaving а sodium pump, two IHX and four SG.

27

PFBR : PSA FOR NEW REACTORS


� Decay heat removal under normal conditions is doneusing the operation grade decay heat removal system(OGDHRS). In case of off-site power failure or non-availability of steam-water system, the decay heat isremoved by passive safety grade decay heat removal(SGDHR) circuit consisting of four independentloops.

� Each SGDHR loop consists of а heat exchanger(DНХ) immersed in the hot pool, one sodium/airheat exchanger (АНХ), associated sodium piping,tanks and air dampers. Diversity is provided forDHX, АНХ and dampers. The circulation of sodiumand air is by natural convection.

28



29

� Thermal power (MWt) 1250

� Electric output (MWe) 500

� Core height (mm) 1000

� Core diameter (mm) 1900

� Fuel PuO2–UO2

� Fuel pin outer diameter (mm) 6.6

� Pins per fuel subassembly 217, 181 Fuel sub assemblies

� Fuel clad material 20% CW D9 (450micron)

� Diameter of main vessel (mm) 12900

� Primary circuit layout Pool

� Primary inlet/outlet temp (◦C) 397/547

� Steam temperature (◦C) 490

� Steam pressure (MPa) 16.6

� Plant life (years) 40

� No. of shutdown systems 2

� No. of decay heat removal systems 2


30

� PFBR is a 500MWe, sodium cooled, pool type, mixed oxide(MOX) fuelled reactor having two secondary loops

� A homogenous core concept with two fissile enrichmentzones of 21/28% PuO2 is adopted for power flattening

� The active core where most of the nuclear heat is generatedconsists of 181 fuel subassemblies. Each fuel subassemblycontains 217 helium bonded pins of 6.6mm diameter. Each pinhas 1000mm column of annular MOX fuel pellets and 300mmeach of upper and lower blanket columns

� The clad material used is 20%CW 15Ni–14Cr–2Mo + Si + Ti(D9).The maximum linear power in the fuel pin is 450 W/cm


31

� Twelve absorber rods, i.e., nine control and safety rods (CSR)and three diverse safety rods (DSR) are arranged in two rings

� Two independent and diverse shut down systems areprovided for ensuring safe shut down of the reactor evenwhen one system is not available

� Both the systems are designed to shutdown the reactor inless than 1 s

� The vessel has no penetrations and is welded at the top tothe roof slab

� The main vessel is cooled by cold sodium to enhance itsstructural integrity


32

� The core subassemblies are supported on the grid plate,which in turn is supported on the core support structure. Acore catcher provided below the core support structure, isdesigned to take care of melt down of seven subassembliesand prevents the core debris from coming in contact with themain vessel.

� The main vessel is surrounded by the safety vessel, closelyfollowing the shape of the main vessel, with a nominal gap of300mm to permit robotic and ultrasonic inspection of thevessels


33


34



� The core subassemblies are supported on thegrid plate, which in turn is supported on thecore support structure. А core catcherprovided below the core support structure, isdesigned to take care of melt down of sevensubassemblies and prevents the core debrisfrom coming in contact with the main vessel.

� Regulatory review of Level-1 PSA is in progress.The review insights with respect to systemreliability analysis are presented in this paper.

35



36

REVIEW INSIGHTS: PSA FOR PFBR


Unav

aila

bili

ty

� The system success criteria are derived based on the deterministic thermalhydraulic analysis and design documents. The CCF analysis is done using theβ-factor model. The human error probabilities are estimated using TheHuman Error Rate Prediction (THERP) (latent & dynamic errors) and ASEP(Accident Sequence Evaluation program).

� The component failure data has been obtained from detailed partsstress/count method. Generic sources such as IAEA-TECDOC-478 andIEEE-500 std. are used as generic sources.

� System unavailability is less than the design target of 1.0E-03/demand. Theemergency power supply system unavailability is 2.5E-06/demand, whichshows the improvements over the existing emergency system unavailabilityestimates of 1.0E-04 to 1E-05/demand in other NPPs. Further, it is notedthat the passive decay heat removal system unavailability estimate is 7.3E-08/demand, which is well below the design target value.

37

REVIEW INSIGHTS: PSA PFBR


� The core damage frequency for VVER type ofPWRs is well below the INSAG-12 recommendedestimate of 1E-04/year. It is even below thebroadly acceptable design targets for the newNPPs of 1E-05/year.

� System unavailability in PFBRs are well belowdesign target unavailability estimates.

� The current thinking of the PSA experts in Indiasuggests that different approaches are needed fortreatment of passive systems and digital C&Isystems of new NPPs in PSA.

38



� The passive systems are being used in new NPPsin India such as VVER-1000 MWe, 500 MWe PFBRsand 700 MWe PHWRs. The literature surveyshows that the systematic procedure is notevolved yet for the reliability assessment of passivesystems. Some studies have been carried out usingReliability Evaluation of Passive Safety Systems(REPAS) methodology worldwide.

� A pilot study has been carried out for a typicalpassive isolation condenser system using REPASprocedure

39

Pilot study on Passive System Reliability analysis


� The system consists of Steam Drum (SD), heatexchanger, also known as Isolation Condenser (IC),a discharge valve in the return path of IC andassociated piping. The IC is immersed in a largewater pool, also known as Gravity Driven WaterPool (GDWP). On reactor trip signal, the dischargevalve gets actuated and opens, which activates the T-H passive system. Thereafter, the system operateson two-phase natural circulation principle. TheGDWP and IC are at higher elevation than thepower source (i.e. SD). The objective of the systemis to reject the core decay heat produced afterreactor shutdown to the heat sink (GDWP) bycondensing the primary fluid (steam) into the heatexchanger tube bundles (IC tubes).

40

Pilot study on Passive system reliability analysis


41

Passive system reliability analysis for Large PHWR


Steam Generator

Primary Pumps

Headers Headers

Calandria


HX

Pre

ssu

rise

r

Steam line

PDHRPDHR

Steam Generator

Primary Pumps

Headers Headers

Calandria


HX

Pre

ssu

rise

r

Steam line

PDHRPDHR

42

42

Introduction to Passive systems

� Passive systems should, by definition, be able to carry out

their mission with minimum or no reliance on external

sources of energy, instrumentation or control signals and

should operate only on the basis of fundamental natural

physical laws, such as gravity, etc.

� IAEA definition, “a passive system is either a system which

is composed of passive components and structures or a

system which uses active components in a very limited

way to initiate subsequent passive operation”

43

43

Introduction to Passive systems

Category A Category B Category C Category D

Input SignalExternal Power SourcesForces

No No No Yes

Moving Mechanical Parts

No No Yes Yes

Moving Working Fluid

No Yes N/A N/A

Example Core cooling systems relyingonly on heat radiation/conduction.Physical barriers against therelease of fission products,such as nuclear fuel cladding& pressure boundary systems

Reactor shutdown/emergencycooling systemsbased on injection ofborated from anexternal water pool.Reactor emergencycooling systemsbased on air orwater naturalcirculation in heatexchangersimmersed in waterpools.

Emergencyinjection systemsconsisting ofaccumulators orstorage tanks anddischarge linesequipped withcheck valves.Mechanicalactuators such ascheck valves andspring loadedrelief valves

Emergencycore coolingsystems basedon gravitydriven flow ofwateractivated byvalves whichbreak open ondemand

44

44

Passive reliability methodology

�Best Estimate (BE) code is used for uncertainties. As a

conservative approach can underestimate margins.

�There are no well established methods to evaluate the

reliability of passive systems at present.

�The methodology is evolving with process parameter

properties distribution knowledge base development.

�The reliability of various passive systems of SBWR and

AP-600 has been evaluated by REPAS.

45

45

Methodology adopted

46

46

PRESSURISER

System description

47

47

Nodalization for PHT, SG & PDHR

10010

1

10

1

102 103

104

109

110

112

107106 108

11

1

10

11

13

114 115

116

119118 120

122

121

12

3

117

Pump Pump

S.G. S.G.

Outlet Feeder

Outlet Feeder

Inlet Feeder

Inlet Feeder

Inlet Header

Inlet Header

Outlet Header

105

Outlet Header

S.V. S.V.

48

48

Pressuriser & Surge Line

SV

SV

ROH

ROH

TRIP V/V

SV

PREQ.1500000

6500000 68500006830000

6510000 6870000

68800006520000

6860000

102

602

TRIP V/V

12

1112

3

PIPE

NODALISATION OF PHT SYSTEM SURGE LINES

Sensitive Process Parameters� Initial PHT system Inventory

� Initial SG Inventory

� Initial PDHR Inventory

� Decay Power

� Non-Condensibles in PHT

� Initial PHT system Pressure

� Initial SG Pressure, Pressure Drops

� SG & Primary Coolant Flow Rate, 2-phase flow model

� Various Heat Transfer Coefficients

� Thermal Conductivities of various material involved

� Thermodynamic Properties

� Reactor Power (RRS reactivity)

� Fouling on heat transport surfaces

� SG Feed water flow & temperature

� Initial Pressuriser Heater State

� Initial Various Control Valve Positions

� Non-Condensable in SGs

� PHT system Loop Isolation

50

700 MWe PHWR Decay Removal during SBO.

� Design Basis Base Case Analysis

� Effect of Inventory� PHT system inventory� SG inventory� PDHR inventory

� Effect of Decay Power� 20%� 10%

� Effect of Non-Condensable� In the PHT 7.5%, 3.5%� In the SGs 5%, 2.5%

� Loop Configuration

� With/without Loop Isolation

� Single Failure

� With only 3 PDHRs available

Parameters Normal Mid Dist Minimum PHT Inv less (tons) 0 -7 -15 PEQU level (m) 5.5 3.6 1.71 SG Inv (m) 16.4 13.4 11.4 PDHR Inv (tons) 0 -10 -20 PDHR Level (m) 3 2.5 2 Decay Heat (%) more 0 10 20 PHT NonCon (%) 0 3.75 7.5 SG NonCon (%) 0 2.5 5

S.No. Key Parameters Range Distribution

123456

PHT Inventory (tons)SG Level (m)PDHR tank level (m)Decay heat (%)Non condensables in PHT (%)Non condensable gases inSG secondary side (%)

0 to -1511.4-16.4

2-30-200-7.50-5

NormalNormalNormalNormalNormalNormal

51

Results & Discussions

0 6000 12000 18000 24000 30000 36000 42000 48000 5400010

20

30

40

50

60

70

80

90

100

110

120 Station Blsckout with PDHR

IC In

ven

tory

(To

n)

Time (s)

288 (cntrlvar) 188 (cntrlvar) 388 (cntrlvar) 488 (cntrlvar)

0 6 000 12000 1 800 0 240 00 30000 36 000 42000 4800 0 5 40 0008

1624324048566472808896

104112120

Station Blackou t w ith P DHR

LO O P-1 Pressu rise r Iso la ted onLow Level of 1 .7m

Hea

der

Prs

sure

(bar

)

T ime (s)

150010000 (p) 102010000 (p) 122010000 (p) 110010000 (p) 114010000 (p)

0 6000 12000 18000 24000 30000 36000 42000 48000 54000369

1215182124273033363942454851

SG

Pre

ssu

re (

bar

)

Time (s)

24 (cntrlvar) Set P 197100000 (p) SG1 497100000 (p) SG2 697100000 (p) SG3 997100000 (p) SG4

0 6000 12000 18000 24000 30000 36000 42000 48000 54000-200

0

200

400

600

800

1000

1200

1400

1600

1800

2000

2200

Cor

e Fl

ow O

ne P

ass

(Kg

/sec

)

Time (s)

100010000 (mflowj) Loop1 112010000 (mflowj) Loop1 600010000 (mflowj) Loop2 612010000 (mflowj) Loop2

52


53


0 6000 12000 18000 24000 30000 36000 42000 48000 54000-0.6-0.5-0.4-0.3-0.2-0.10.00.10.20.30.40.50.60.70.80.91.0

IC In

let/

Ou

tlet

Qu

ality

Time (s)

333010000 (quale) 333160000 (quale)

0 6000 12000 18000 24000 30000 36000 42000 48000 54000

-0.10.00.10.20.30.40.50.60.70.80.91.01.1

SG H

eate

d R

iser

Exi

t Q

ualit

y

Time (s)

691110000 (quale) 691010000 (quale) 691020000 (quale)

54


0 6000 12000 18000 24000 30000 36000 42000 48000 54000-12

-8

-4

0

4

8

12

16

20

24

28

32IC

Flo

w (k

g/s)

Time (s)

140000000 (mflowj) 143000000 (mflowj)

Sensitivity Analysis for Passive System Reliability

Case

Clad Temperature (oC) Failure Time (Hr)

5Hrs 6Hrs 10Hrs

G0 171.958 173.186 215.614 17.02 G1 214.926 239.542 316.219 10.49 G2 174.039 187.128 243.005 13.88 G3 189.231 FAIL FAIL 05.22 G4 188.27 195.799 260.927 14.75 G5 270.124 271.895 FAIL 07.42 G6 268.744 268.618 FAIL 08.41 G7 224.169 246.277 397.771 10.02 G8 195.551 203.601 267.759 14.72 G9 262.733 273.721 FAIL 08.93 G10 272.241 273.909 FAIL 08.49 G11 276.681 314.675 FAIL 06.20 G12 247.976 269.43 FAIL 08.73 G13 306.377 910.49 FAIL FAIL 05.51 G14 308.129 1016.77 FAIL FAIL 05.47 G15 185.873 190.514 268.849 14.57 G16 234.054 248.653 281.199 11.70 G17 <400 <400 <400 08.92 G18 242.102 254.653 282.377 11.83 G19 <400 <400 <400 17.02 G20 <400 <400 <400 10.49 G21 269.67 269.458 FAIL 08.5 G22 224.219 275.614 244.657 05.22 H1 205.574 221.115 311.154 14.75 H2 203.061 209.469 261.403 07.42

Sr. No.

PHT Inv PRL (m)

SG Lev (m)

PDHR Lev(m)

+ Decay Power (%)

PHT NonC (%)

SG NonC (%)

0 3.6m 16.4 3.0m 0 0 0

1 3.6m 16.4 2.5m 20 3.75 0

1a 3.6m 16.4 2.5m 16.5 3.75 0

2 5.5m 16.4 3.0m 0 0 2.5

3 3.6m 13.4 2.0m 10 7.5 5

4 3.6m 13.4 2.5m 10 0 5

5 1.71m 13.4 0.0m 0 7.5 2.5

6 1.71m 13.4 0.0m 0 0 0

7 1.71m 13.4 2.0m 0 7.5 5

8 5.5m 13.4 2.0m 20 7.5 2.5

9 5.5m 13.4 2.0m 20 0 2.5

10 5.5m 11.4 2.5m 0 7.5 2.5

11 1.71m 14.4 2.5m 0 0 0

12 1.71m 16.4 2.0m 0 3.75 0

13 1.71m 11.4 2.0m 20 7.5 0

14 1.71m 16.4 2.0m 0 3.75 2.5

15 5.5m 13.4 2.0m 10 7.5 5

16 1.71m 13.4 3.0m 0 3.75 2.5

17 1.71m 16.4 2.5m 20 3.75 0

18 1.71m 16.4 2.5m 0 0 0

19 -15 15.4 0 20 7.5 2.5

20 -15 12.4 0 0 7.5 2.5

21 -7 15.4 -10 10 3.75 0

22 -15 15.4 -10 10 0 2.5

23 0 15.4 -10 0 3.75 5

24 -7 15.4 -20 0 3.75 0

25 0 15.4 -20 0 3.75 2.5

26 -7 15.4 -20 10 7.5 5

27 -7 15.4 0 0 7.5 0

28 -15 10.4 0 10 0 0

29 -15 15.4 -20 20 7.5 2.5

30 -7 12.4 -20 20 3.75 5

31 -15 10.4 0 20 3.75 0

32 0 10.4 -20 20 0 0

33 5.5m 13.4 2.5m 0 0 0

34 3.6m 16.4 3.0m 0 3.75 5

35 -15 15.4 -20 10 3.75 5

36 -7 10.4 -20 10 3.75 5

37 -7 10.4 -10 10 7.5 5

38 -15 12.4 -20 10 7.5 5

39 1.71m 16.4 2.5m 0 3.75 5

40 3.6m 13.4 3.0m 20 0 2.5

56

266

255

244

233

222

211

6655443322110

xbxbxbxbxbxb

xaxaxaxaxaxaaT

++++++

++++++=

Coefficient Value

a0a1

a2a3

a4a5a6

b1b2

b3b4b5

b6

-5960.3416-13.953464842.27824127.302479-6.3074330-46.79045267.05214618.5671719-28.324671-64.8436510.97715686.5101249-8.5379061

•Monte Carlo simulation technique

•Estimated failure probability 2.302x10-3

•Fuzzy 7.513×10−4

57

1. Simulation model which includes all the three heat transport systems has been used for the assessment of passive reliability for decay heat removal.

2. Sensitivity studies were carried out to obtain the critical process parameters affecting the decay heat removal.

3. The clad temperature has to remain below 400C for the specified period (6 hours considered here) following a SBO.

4. Based on the thermal hydraulic analysis the response surface has been developed.

5. Using the Monte Carlo simulation technique failure probability has been estimated.

6. For the preliminary analysis the estimated failure probability obtained is 2.302x10-3 Due large variation in decay heat & non condensable quantities assumed.

7. Fuzzy Mote Carlo Simulation give 7.513x10-4

� There are some open issues, which require further attention inorder to get the confidence in the analysis results.

� There is a need to develop a thorough understanding of differentdegradation mechanisms, which can affect the passive systemperformance. Based on this understanding, the analysts need toestablish proper system ‘success criteria’ as it has a significantimpact on the reliability assessment of passive systems.

� Currently, in absence of the ‘adequate’ operating experience, theresearchers ‘assumed’ probability density function for uncertaintycharacterization of the input parameters. There is a need toreduce the ‘subjectivity’ in choosing these probability densityfunctions through experimental evidence.

58

Pilot study on Passive system reliability analysis


� With the increasing use of digital C&I systems in new NPPs, therehas been growing debate over the issue of whether probabilistictechniques can be applied to digital systems for reliabilityassessment.

� The discussion includes the issue of whether the software failurescan be modeled probabilistically. The treatment of software in thequantitative reliability assessment has been the source of difficultyin the absence of acceptable quantitative methods.

� The traditional FT analysis is limited in its ability to model some ofthe failure modes associated with digital C&I systems, especiallytime-dependent, sequential failures and failure mode thatincorporate fault tolerance. These difficulties can be handled byintroducing some special dynamic fault tree gates

59

Digital System Reliability


Open issues

� The mechanisms through which software and hardwarecomponents fail are quite different. Hardware failures occurgenerally due to ageing or the occurrence of randomexternal “shocks”. Software on the other hand, does notwear out and its behavior is deterministic. i.e. for a given setof inputs, it is simply either “correct” or “incorrect”. Therehas been growing debate over the issue of whether softwarefailures can be modeled probabilistically. In absence ofacceptable quantitative models, most analysts hesitate toinclude software into fault-tree models.

60



Some of the open issues in software reliability are:

� Identification of failure modes and modeling of dynamicinteractions between software and hardwarecomponents

� Quantitative reliability assessment for softwarecomponent

� Effect of digital C&I systems on human factors andhuman error probability assessment

61



THANK YOU


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