ML20062B255

Forwards Hope Creek Generating Station Pra. Rept Summarizes Plant Level 1 Analysis.Completion Date Extension Requested

ML20062B255
Person / Time
Site:	Hope Creek
Issue date:	10/10/1990
From:	Crimmins T Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-88-20, NLR-N90191, NUDOCS 9010240190
Download: ML20062B255 (23)

Text

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Pubhc $crvice Electric and Gas

q; . Company Thomas M. Crimmins, Jr. Pubhc Service Doctnc and Gas Ccmpany P.O. Box 236. Hancocks Bridge, NJ 08'38 609-339-4700 Vace hosdent Nwcks [tgmecrerig

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OCT 101990 l NLR-N90191 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Gentlemeni 1

. GENERIC LETTER 88-20, MILESTONE COMPLETION

SUMMARY

REPORT l HOPE CREEK GENERATING STATION I DOCKET NO. 50-354 Public Service Electric and Gas Company (PSE&G) responded to Generic Letter (GL) 88-20 in a letter dated November 1, 1989.

PSE&G stated that a brief summary report would be forwarded to the NRC upon completion of each milestone.

Attached is our report summarizing the Hope Creek Level 1 analysis. This report documents the completion of the first GL >

88-20 milestone. We requested a completion date extention, to September 1990, in a letter dated July 12, 1990.

Please contact us if you have any questions regarding this transmittal.

Sincerely, L

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L ll 9010240190 901010 PDR P

ADOCK 05000354 PNV ,

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Document Control Desk 2 NLR-N90191 OCT 101MD i

C Mr. Stephen Dembek 1 Project Manager Mr. T. P. Johnson i Senior Resident Inspector j Mr. T. Martin, Administrator )

Region I Mr. Kent Tosch, Chief New Jersey Department of Environmental Protection  ;

Division of Environmental Quality Bureau of Nuclear Engineering {

CN 415 q Trenton, NJ 08625  ;

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t NLR-N90191 HOPE CREEK GENERATING STATION PROBABILISTIC RISK ASSESSMENT

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SUMMARY

OF RESULTS

I Table of contents section Enga 1.0 Introduction 1 2.0 Summary of Results of the Hope Creek PRA 1 2.1 Core Damage Frequency 1 2.2 Contribution to Core Damage by 2 Accident Class 2.3 Contribution to Core Damage by 3 Accident Subclass 2.4 Contribution to Core Damage by 3 Accident Type 1.5 Contribution to Core Damage by 3 Initiator 2.6 Contribution to Core Damage by 4 Accident Sequence 2.7 Summary of External Events Analysis 4 3.0 Conclusions and Recommendations 5

-4.0 . References 8 4

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List'of Tables i

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• Table Eggs j l

T-1 Core Damage Frequency Comparison with 9 other Studies T-2 Summary of the Core Damage Accident 10 <

Jequence Subclasses (Plant Damage Ststes for the Hope Creek PRA)

T-3 Dominant Core Damage Sequences 12 -

T Summary of External Events Quantification 13 i

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F-1 Contribution to CDF by Accident Class 14 )

l F-2 Contribution to CDF by Subclass 15 i

i-F-3 Contribution to CDF by Accident Type 16 F-4 Contribution to CDF by Initiator 17 t

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I ANALYSIS OF RESULTS

1.0 INTRODUCTION

A probabilistic risk assessment (PRA) of Hope Creek Generating Station (HCGS) was performed during the period of September 1988 through July 1990. The HCGS PRA includes a detailed analysis of core damage potential (Level 1 PRA, based on NUREG/CR-2300

-terminology) including external events. External events analysis included a comprehensive screening of all types of external and spatially-dependent internal events, and a more detailed analysis of those events not screened outt internal flooding, internal fire and seismic. The analysis-of containment response to core damage sequences and radionuclide release from the core (Level 2 PRA) will be completed at a later date.

This report presents an overview of the Hope Creek event tree quantification results. A description of the methods used in this analysis is available. The quantification results are presented as point estimates of the accident sequence frequencies. They can be characterized as the mean of the distribution of the random variables representing core vulnerable frequencies.

This report includes a hierarchical synopsis of HCGS accident sequence analysis results. The results are presented from the highest level (the core damage frequency) to the lowest level (the accident sequence frequencies) .

Insights gained from review of quantification process results are included. These insights form the basis of recommendations for HCGS plant or operation enhancements that can contribute to decreasing the core damage' frequency.

2.0

SUMMARY

OF HCGS PRA RESULTS 2.1 Core Damaae Frecuency The core damage frequency is the sum of many independent accident sequence frequencies, each one several orders of magnitude lower in-frequency than the total. The core damage frequency from internal events for HCGS is 1.42E-4 per year. This frequency compares favorably with a number of similar plant studies.

Table T-1 lists the core damage frequencies resulting from analyses of five other Mark 1 BWRs. Only minor differences exist between these analyses results and HCGS results. The core damage frequency for HCGS is lower than most of the other plant core damage frequencies.

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-- -. . - . - . -- ~ - - - - - ..- - - -. - _ - . - - .. .

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, d s 2 2.2 contribution to core Danmaa by Accident class Five accident classes were selected to represent the spectrum of possible accidents at HCGS. Accident classes extend from higher

~ frequency / lower consequence events to lower frequency / higher i consequence events. These accident classes are described below:

1 4

CLhEE DESCRIPTION j s

Cl Inadequate CoolantiInventory Makup l C2 Inadequate Decay Heat Removal j C3 LOCA with Inadequate Coolant Inventory Makeup C4 ATWS with Inadequate Containment Heat Removal i I

C5 Interfacing LOCA Figure F-1 illustrates the breakdown of total core damage  ;

frequency by accident class. This internal' events analysis does '

not include interfacing LOCA events. Interfacing system LOCAs i' will be considered in future analyses. Figure F-1 clearly shows that the class C2 accidents dominate the total core damage frequency at HCGS. Inadequate decay heat removal accidents .

. eventually result in containment pressurization (e.g., failures of the RHR system or containment vent system).

o Accident class C1 contributes about half that of class C2. Class C1. accidents involve the inability of plant systems to maintain adequate coolant inventory. Figure F-1 shows that loss.of coolant accidents and events involving a' failure to scram contribute very little to the total core damage frequency at L HCGS. Historically, loss of coolant accidents contribute little to the core damage frequency. This occurs because of plant design and the small likelihood of a LOCA. Anticipated transients without scram (ATWS) events contribute little to the total core damage at HCGS for two reasons:  ;

• The Standby Liquid-Control (SLC) system at HCGS is -f initiated automatically and.has redundant pumps. These .

features significantly reduce the likelihood of SLC '

unavailability when demanded. -l

• Emergency Operating Procedures (EOPs) at HCGS~

incorporate the latest guidance from the BWR Owners Group Emergency Procedure Guidelines. The EOPs provide the best guidance to plant operators on controlling '

reactor power under the extremely unlikely ATWS event.

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1 2.3 Contribution _to Core nammaa by Accident Subclass Accident classes are further subdivided to provide a means of identifying the status of the reactor coolant system pressure boundary, containment boundary and the operability of certain important systems (see Table T-2 for accident subclass definitions).

Figure F-2 shows the breakdown of accident subclass contribution to the total core damage frequency. Three subclasses contribute about 05% of the total core damage frequency at HCGS. Subclass C2A (loss of containment heat removal with the reactor pressure vessel initially intact, where core damage occurs after the i containment fails) contributes nearly half the core damage frequency at.HCGS. Subclass CIA (acci(wM4 ~9quences involving loss of inventory' makeup, where reactor pewsvare remains high) contributes about a third of the total cuce uamage frequency.

Subclass C28-(sequences involving a loss of containment heat removal, where the containment is successfully vented and subsequently,-the ability to provide reactor makeup is lost) contributes nearly a fifth of the total core damage frequency.

All other accident subclasses combined contribute the remaining 5% of the total.

2.4 Contribution to Core Dammaa by Accident Tvoa The process of. performing a PRA involves analyzing many different accident scenarios. It ,,s often useful to group scenarios and present the results of these collections of sequences. Figure F-3 illustrates the contributions to total core damage from these scenario groupings.

.Three quarters of the core damage frequency at HCGS is due to failures in recovery from anticipated transient effects. A fifth is due to loss of offsite electrical power. A very small fraction of the total results from failures of onsite emergency power in conjunction with a loss of offsite power. As before, LOCAs and failures to scram contribute little to the total core damage frequency.

2.5 Contribution to Core Damaae by Initiator i'

Total core damage frequency can be broken down to identify the contribution made by individual initiators. Figure F-4  !

summarizes the core damage contribution by specific initiators.

About 85% is nearly equally attibuted to four initiators.

Transients initiated by MSIV closure account for about a fourth of the total core damage frequency. Transients associated with with loss of condenser function contribute about a fifth. Loss of offsite power and loss of the safety auxiliary cooling system account for abeat a fifth each.

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-l The remaining initiating events and their contribution to' core damage frequency is listed below:

IORV- 7%

Turbine Trip 34 l Medium IDCA 2% l Loss of-Instrument Air 2%

All others combined 1%

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2.6 contribution to core n==maa by Accident samuence The dominant accident sequences identified in this analysis (having a ' frequency . greater than 1E-7 per year) are listed in Table T-3. These sequences account for greater than 99%-of the  !

total core damage frequency from internal initiators. Table T-3 1 sequences'are dominated by those involving failures of the W and W1 functions, and failures of the U and X functions. These sequences are described in more detail below..

The functional failures set' contributing most to the core damage frequency is the W and W1. The W function is containment heat l . removal by the RHR' system and the condenser. The W1 function is i

containment venting. . These function failures lead to insufficient decay. heat transfer from the containment,. inability to relieve the resultant pressure, and subsequent containment  :

overpressurization. . The containment ultimately fails. This is i assumed-to lead to severe degradation of injection systems, j through the loss of suction head resulting from the rapid  ;

containment depressurization.

The second most significant set of functional failures consists of the U and X functions. The U function is the injection of ,

high pressure coolant from the HPCI or RCIC systems.' The X- i function is.the depressurization of the reactor pressure vessel. l This set of failures leads to core damage due to insufficient reactor vessel makeup from the high pressure systems, and the ,

inability to reduce reactor pressure below low pressure system makeup capability.

2.7 Su===rv of External Events Analysig A comprehensive screening of external and spatially dependent internal events resulted in three significant contributors to risk at HCGST internal fire, flooding and seismic events. The

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following values were obtained:

Total from Fire Events 1.69E-4 Total from Flooding Events 4.59E-8 Total from Seismic Events 1.60E-6 Total from External Events 1.71E-4 4

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I, The impact of fire related events is far greater than the other external / spatially-related events evaluated. . Over 99% of the total external event core damage frequency is from fire events.

The fire event analysis contains a large amount of conservatism.

The fire analyses assumes that all equipment in an affected room is rendered unavailable. This assumption results in large sequence frequencies for those rooms in which cabling for all safety equipment passes, such as the cable. spreading room and the main control room. Fires in these rooms can force plant shutdown from the remote shutdown panel. This panel has limited control and indication capability.

, The frequencies of the fire initiating events appear realistic.

But, if a more detailed analysis was performed to identify actual.

equipment failures resulting from these fires, the frequencies would be significantly lower (probably by more than a factor of 10).

Table T-4 summarizes the results of the external events analysis.

Accident subclass contribution to the total frequency is provided for each external event type.

3.0 CONCLUSION

S AND RECOMMENDATIONS The HCGS PRA process yielded many insights into the design and operation of the plant. These insighted are described below

• The Standby Auxiliaries cooling System (SACS) supplies cooling water to the emergency. diesels generators (DG).

The SACS operating procedure specifies both pumps in a single loop in-service with the other SACS loop in standby. The system valve lineup implies that all four DGs must be aligned.to the operating SACS loop. If this loop.does not restart following a loss of offsite power, the DGs will be running without cooling. A DG trouble alarm would alert control room operators to a problem condition. Field operators must go to the local DG alarm panel to determine that the trouble is high temperature. SACS realignment to restore DG cooling is a manual evolution. The time to determine SACS failure as the problem cause and restore cooling may be longer than the DGs can operate without engine failure.

• The importance of the containment vent is eviderced by the number of accident sequences involving its failure.

The vent was modeled as currently installed, although system modification is planned. Venting presently _

involves low pressure ducting with blowout panels that rupture at 0.25 paid. Containment venting is expected to cause blowout panel failure, with steam and 5

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. atmospheric contaminant release-to the reactor

. building. The resulting environmental conditions'could 1 potentially exceed the environmental qualification of i equipment in this area. Such atmospheric conditions 1 would'also limit operator access to reactor building rooms. The containment vent issue is still under investigation with consideration of a hard pipe vent j modification.

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• The feedwater system was treated differently in this )

analysis than in most other PRAs. Feedwater.is historically assumed to have a mission time of about 30 minutes. Level is assumed to drop rapidly for a.few q seconds until the MSIVs close and then at a' slower rate "j to the top of the fuel; during this time frame it has w been assumed that feedwater restoration would:be attempted. At HCGS, experience has shown that level )

would drop rapidly to the HPCI/RCIC initiation j setpoint. If these systems failed.to start, reactor j

. pressure would be reduced to the point where secondary L condensate pumps could be used as an injection source. l If HPCI/RCIC did start, attempts to restore feedwater l would continue but are not necessary. These practices have been considered in determining feedwater i unovailability and tend to increase it.  ;

• Both the HPCI and RCIC turbine driven pumps have l

.displafed relatively poor start test performa.nce. This .;

causes higherLthan expected unavailabilities fer these ,

pumps, which results in high frequencies for sequences involviny failure of these systems; those sequences p

involving failures of the U, U1 and U2 functions.

• It'is expected that the method selected.to break fault g tree: logic loops in system models such as SACS has ,

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introduced some nonconservatisms.into the analysis.

This issue will be further. evaluated in future-analyses.

• Both the W function (containment heat removal by the; RHR system) and the W1 function (containment venting) ,

have dependencies on the service water (SW) system. i The RHR system depends on the SACS system to cool the l

RHR heat exchangers and SACS is cooled by SW. The containment vent depends upon SW to cool the instrument air compressors that provide air for vent valve -

operation. Due to'this common dependency, it would be e benificial to pre-plan recovery actions for SW-problems such as trash screen and SW strainer clogging. The SW system is extremely important to safe plant operation.

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It is required.for BOP operation, SACS operation -

which cools safety systems and the DGs - and ,

i ventilation cooling operation.

• The RCIC system is automatically isolated when the ,

exhaust pressure reaches 10 psig. Such a low isolation '

setpoint precludes system use for any accident sequences where containment pressure exceeds this value. This significantly reduces the. plant's response l capability to conditions imposed by many accident '

sequences. The existing RCIC isolation setpoint should be evaluated, and every effort made to raise the ,;

setpoint so that RCIC will remain available during  ;

accident conditions. If a.setpoint. change is not a possible, consider a modification to allow trip .i override during severe conditions. l l

• Efforts should be made to restore condenser operation whenever it is lost. The power conversion system'(PCS) ]

is the best alternative for removing containment heat  ;

in all trip scenarios. Operator training.should 1 emphasize condenser restoration through any available )

L means. Condenser. operation reduces plant reliance on_ i the.RHR system and containment vent during severe l conditions.  ;

• No credit was taken during this analysis for the reactor water cleanup system (RWCU) as a means of I removing decay heat from the reactor. Including the l

.RWCU system in the W. function analysis could 1 potentially reduce.the. frequency.of sequences involving j this function failure.

• An assumption used throughout this analysis was that o ventilation to electrical; equipment spaces was required to prevent equipment overheating. This results in many accident sequences being dominated by g

ventilation equipment failures. The current. analysis includes some simple recovery' actions for a loss of ventilation, such as' opening the affected room's doors l to promote natural circulation. Actual room J ventilation requirements should be determined by 1

,? performing quantitative analyses.- Analyses should include the actual heat load under various accident conditions and room heat-up associated with these conditions. Conservatism in important sequences could then be reduced.to obtain more realistic results. l J .

• Hope Creek has an automatic SLC system that responds i quick enough to attain the required boron concentration ,

using only one of the two installed pumps. The j

automatic initiation and' pump redundancy contribute to  :

the low ATWS sequence frequencies observed at HCGS. J 7

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• The CRD system is effectively used as an injection source at HCGS. .The procedures address its-use as a long term reactor makup source and the operators are fam:, liar with them. The automatic CST makeup system, when used in conjunction with the CRD system, establishes a good long term, low operator involvement system that should be emphasized during' operator training sessions.

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• HCGS technical staff should reconsider reestablishing PCS in the event of MSIV closure. The current requirement to reestablish initial plant conditions before reopening the MSIVs precludes PCs. usage under many-accident conditions wnere high radiation levels exist. Operator guidance on PCS use during conditions-where radiation alarms exist, but levels remain low, could prevent subsequent core damage and resultant elevated radiation levels.

• HCGS safety systems are heavily, dependent upon SACS.

Any failures that. disable SACS are quite debilitating.

SACS operation is'not completely balanced between Division I and Division II (e.g., failure of SACS A results.from failure of Chiller A which receives power

'from vital bus C). Such imbalances cause many more failures to result in SACS loss than would be expected if the systems were balanced.

4.O REFERENCES PRA Procedure Guide NUREG/CR-2300, January, 1983.

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Table T-1 CORE DAMAGE FREQUENCY COMPARISON WITH OTh?.R STUDIES

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PRA Study Core Damage Frequency t, Browns Ferry 1 2.0E-4 Cooper 2.9E-4 r

Millstone 1 3.0E-4 l Peach Bottom 2 3.0E-5 Hope Creek 1.4E-4 l

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Table T-2

SUMMARY

OF THE CORE DAMAGE ACCIDENT SEQUENCE SUBCIASSES (PIANT DAMAGE STATES) FOR THE HOPE CREEK PRA Accident Class Desianator Subclass . Definition ExamDie

'., , C1 A Accident sequences involving TQUX loss.of inventory. makeup where reactor pressure remains high.

B(x,r) Accident sequences involving TEQUV loss of off-site power and loss of coolant inventory makeup.

(Subscripts, i.e., x and uv denote system functionality at time of core damage).

C Accident sequences involving TCMC2DEP loss of coolant inventory induced by an ATWS sequence; RPV pressure high at the onset of core damage.

D Accident sequences involving TQUV loss of coolant inventory makeup where reactor pressure has been successfully reduced to 200 psi; accident sequences initiated by common mode failures disabling multiple systems (ECCS) leading to loss of coolant inventory makeup.

F Accident sequences involving TCMC2V g

loss of coolant inventory induced by an ATWS sequence; RPV pressure low at the onset of core damage.

C2 A Accident sequences involving TWW1 loss of containment heat removal with the RPV initially c intact; core damage induced post containment breach.

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Accident Class h ignator subclass Example Definition B Accident sequences involving TUV loss of containment heat removal except that the containment vent operates as designed; loss of coolant makeup occurs following vent initiation (i.e., suppression pool saturated but intact);

, core damage ensues.

C3 A Accident sequences leading to R core vulnerable ~oonditions initiated by vessel rupture where the containment integrity is not breached in the initial time phase of the accident.

B Accident sequences initiated or 81QUX resulting=in small or w.edium LOCAs where the reac+ or cannot be depressurized.

C Accident sequences initiated or -AQUV resulting in medium or large LOCAs where the reactor is at low pressure.

D Accident sequences initiated by AD a LOCA or RPV failure where the vapor. suppression system is inadequate, challenging containment integrity.

C4- Accident sequences involving TTCMC2 failure to insert negative reactivity leading to a containment; vulnerable condition

'due to high containment pressure.

C5- Un!solated LOCA outside A0V containment.

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, Table T-3 l

i DOMINANT CORE DAMAGE SEQLEhCES i

' Core Damage- Frequency Plant Damage i Sequence Par Year ggggg  !

TSA-QWW1 2.26E-05 C2B l TE-IWW1 2.25E-05 C2A '

TM-QWW1 1.71E-05 C2A ,

TM-QUX 1.54E-05 CIA i TC-QUX. 1.36E-05 CIA l TC-QWW1 1.34E-05 C2A l TI-QUlX 7.26E-06 CIA  !

=TT-QUX 3.09E-06 CIA I TSA-QUX 3.05E-06 CIA ,

TC-QUWW1 2.80E-06 C2A l S1-QUlX' 2.63E-06 C3B i TI-QWW1 2.02E-06 C2A SBO-TEDGIR12U201UV 1.94E-06 C1B

.TIA-QWW1 1.58E-06 C2A I TM-QUWW1- 1.15E-06 C2A .!

TIA-QUX 8.88E-07 CIA  ;

7 TSA-QWUV 7.95E-07 C2B '

TT-WW1 7.39E-07 C2A I SBO-TDGIR123II5WUV 6.72E-07 C2B ,

3 TE-IWUV 6.40E-07 C2B l TSA-QUWW1- 6.38E-07 C2A )

TMS-WW1 5.67E-07 C2A d TE-TMQWW1 5.54E-07 C2A I TE-IUWW1 5.28E-07 C2A ,

TT-PUV 4.95E-07 C1D 1 TM-QWUV 4.52E-07 C2B i TE-TMQUX 4.39E-07 CIA TIA-QUWW1 4.07E-07 C2A -l TC-QWUV 3.72E-07 C2B l TI-QU1UV 3.71E-07 C1D SI-QUV. 3.51E-07 C3C SBO-T3IIR2IIIR3 1.99E-07 C1B

~TF-QUX- 1.77E-07 CIA l JE SBO-TEEDGIR12U2UlX 1.72E-07 C1B

'U .TM-PQUV 1.40E-07 C1D '

. SBO-T123IIR2U2U1UV 1.39E-07 C1B TE-IUX 1.12E-07 CIA TC-PQUV 1.07E-07 C1D TIA-PQUV 1.06E-07 C1D TM-QUV 1.01E-07 C1D 1

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ii-Table T

SUMMARY

OF EXTERNAL EVENTS QUANTIFICATION 1

CIA C1B C1D C1F C2A C2B- [

FIRE  !

'4 EVENTS 1.64E-04 0.00E+00 9.66E-09 0.00E+00 5.74E-06 1.09E-07 i L

FLOOD

. EVENTS 1.40E-09 0.00E+00 3.44E-08 0.00E+00 8.74E-09 6.00E-10 l fi  ;

. SEISMIC 7

EVENTS 4.46E-07 3.89E-07 7.35E-07 1.69E-09 2.43E-08 0.00E+00  ;

N i EXTERNAL- '

L EVENTS.

TOTAL 1.64E-04 3.89E 7.79E-07 1.69E-09 5.77E-06 1.10E-07 L

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Figure F-1 CONTRIBUTION TO CDF BY ACCIDENT CIASS

-c4 (0.0%)

C1 (34.1%)

ca (1.9%)

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Figure T-3 .

1 CONTRIBUTION TO CDF BY SUBCIAss 4

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others (4.9%)  ;

,. l C2B (18.2%) . 1 i

H C2A (45.7%) ,

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v C1 A (31.2%)

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'm' CONTRIBUTION To CDF BY ACCIDE'.<T TYPE

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Figure F-4 CONTRIBUTION TO CDF BY INITIATOR l

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l others (0.6%)q Tt (3.2%)

Tsac (19.2%)

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Tm (24.3%)

Tias ~(2.2%) .f.!si:..  :

! si (2.2%)

Te (20.1%)

Tc (21.5%)

Ti (6.9%)

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