ML20056G538
| ML20056G538 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1993 |
| From: | Teh-Chiun Su NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| REF-GTECI-153, REF-GTECI-NI, TASK-153, TASK-OR NUREG-1461, NUDOCS 9309030241 | |
| Download: ML20056G538 (32) | |
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NUREG-1461 Regu~ atory Ana:ysis :for the Reso:ution 0:f Generic Issue 153:
Loss 0:? Esserr':iaL Service Water in LWRs U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research T.-M. Su f *souq U
788' 38!n = 832 1461 R PDR
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AVAILABILITY NOTICE L
Availabikty of Reference Matenals Cited in NRC Pub lications
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i Most documents cited in NRC publications will be available from one of the following f
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National Standards. from the Amencan National Standards institute, 1430 Broadway, f
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NUREG-1461 Regulatory Analysis for the Resolution of Generic Issue 153:
Loss of Essential Service Water in LWRs Manuscript Completed: July 1993 Date Published: August 1993 T.-M. Su Division of Safety Issue Resolution Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
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t AllSTRACT l
In this report, the staff of the U. S. Nuclear Itepulatory of a scoping study for GI-153, the staff recommendsthat Cornmissmo (NRC) pr ovides a'reputatory analysis for the the insights rained fiom Ihe study serve as a cornplement I
proposed resolution of Generic issue 153 (GI-153). " loss to the on-poing 1 SW performance inspection program.
of Essential Senice Water m LWRs. GI-153 deals with The staff also concludes that liSW system reliabihty is i
the concer ns pertaimng to the rehabihty of essential serv-being addr essed by various on-going i egulatory programs.
ice water (ESW) system and related probleins for all brht
'Iherefore, the staff recommends that GI-153 should be f
water reactors execpt the seven mulu-umt sites addressed consider ed " RESOLVED." 'Ihe need for iutur e action (s) by GI-130, " Essential Scrwce Water Pump Failurcs at on ESW retrabihty is expected to be determined from Multi-Unit Sites.' On the basis of the technical findings these on-poing programs.
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CONTENTS Page iii Abstract
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vii lixecutive Summary..
i Statement of the Problem.
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1.1 llackground...
I 1.2 Safety Signific;mcc 2
1.3 Objectives 3
2 Technical Findings 3
2.1 Iteview of 0perating Esperience 2.1.1 Precursor iteports 3
3 2.1.2 Operating Experience Feedback Itcports 3
2.1.3 Analysis of ESW System at Multi-Unit Sites 2.1.4 Operating Experience after Publication of NUltEG-1275 (1987-1992).
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2.2 lleview of Plant-Specific Probabilistic Itisk Assessments..
6 2.2.1 C<mtribution of liSW system to Jtisk of Core Damage 6
2.2.1.1 IlWit Plants.....
b 2.2.1.2 PWit Plants.
6 2.2.2 Discussion of Plant-Specific Probabilistic ltisk Asst ssr:ents.
7 2.2.2.1 Dominant Sequences 8
2.2.2.2 ESW System Failure Modes..
8 2.2.2.3 liffects of Water Quality on ESW System 9
2.3 Evaluation of Selected Plant-Specific PltAs 9
2.3.1 Plant 11-2...
10 2.3.1.1 Contribution of ESW system to Internal Events CDF...........
2.3.1.2 Contribution of ESW system to External Events CDF.......
10 10 i
2.3.2 Plant 11-3.
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.l 2.3.2.1 ESW System Vulnerabilities.
12 2.3.2.2 Valuellmpact Analysis........
l 2.3.2.3 Contribution of ESW to External Events CDF 12 12 2.3.3 less of ESW System as Initiator...................
12 2.3.4 Uncertainty Considerations.................................
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3 ESW-Itclated itegulatory Activities.......
15 3.1 SWS System Operational Performance Inspection Program.
15 3.2 Implementation of Generic letter 89-131tequirements.......
15 3.3 Individual Plant Examination Program.......
v NUltEG-1461
3.4 ESW.Related Generic Issues 15 3.4.1 Generic Issue 23 " Reactor Coolant INmp Seal Failures' 15 3.4.2 Generic issue 51. " Improving the Reliability of Open Cycle Senice-Water system"....
16 3.4.3 Generic Issue 130, " Essential Senice Water Pump Failures at Multi-Unit Sites"........
16 3.4.4 Generic Issue 13 32. " Ice Effects on Safety-Related Water Supplies"..............
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i 16 3.5 Main t enance R ule......................................................
f 4 Insights 17 f
4.1 Contribution of the ESW System to the Risk of Core Damage....
17 4.2 Dominant Sequences and Other Factors Influencing the ESW Sysicm Reliability..........
17 4.3 ESW-Related Risk Data from PRA Reports.
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5 Recommendation 19 l
6 Reference..
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List of Tables l
f viii E.1 ESW System's Contribution to Core Damage Frequency (Internal Events).............
2.1 Precursors to Potential Severe Core Damage Accidents Ir volving Senice Water Systems 4
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2.2 Twelve Events From NUREG-1275 Resulting in Complete loss of SW Function.........
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l 2.3 Additional Events From NUREG/CR-5526 Resulting in Complete loss of SW f
Function for PWRs...
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2.4 Recent Events Resulting in the Imss of SW Function 6
2.5 ESW System Contribution to Core Damage Frequency.13WRs (Internal Events)......
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2.6 ESW System Contribution to Core Damage Frequency, PWRs (Internal Events).............
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2.7 ESW System Failure Modes Dominant at More Than One Site........
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2.8 ESW System Vulnerabilities and Modifications............
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2.9 Valuellmpact Analysis for Individual Modification (Plant 11-3)...........................
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2.10 ESW System Contribution to Core Damage Frequency (Plant 11-3)...
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2.11 Sensitivity of Modification 2 Cost /Ilenefit Ratio to Variations in Assumptions for Plant 11-3....
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NUREG-1461 vi ft t
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EXECUTIVE SUMM ARY In this report, the statf of the U. S. Nuc! car Regulatory gree of dependency on the ESW system, the reliability of Commision (NRC) provides a regulatory analysis for the the ESW system itself and, to some extent, the ddfer-resolunon of Genenc issue 153 (GI-153), "I nss of Essen-ences in the PRAs in terms of modeling assumptions and tial Service Water in 1.WRs.' The resolution is based on scope of each PRA program.
l the technical fmdmps of a scoping stedy for GI-153.
GI-153 deals with the concerns pertaining to the reliabil-1he ESW system dominant failure modes found from the sty of essemial service water (ESW) system and related review of the 11 NRC-sponsored PRAs have some com-problems for all light water reactors except the seven mon aspects in different plants, even though the system multi-unit sites addressed under GI-130, " Essential Serv configuration for each plant reviewed is unique. For in-l see Water Pump Fadures at Multi-Unit Sites."
stance, one of the common senice water faults was the i
failure of the service water system motor-operated or I
'lhe Ichabihty of the ESW system has been a concern of air-operated iso!ation valves to open on demand f o supply the NRC and the nmlear industry for years. 'Ihe NRC coolmg water to safety-related loads. This failure mode l
concerns bas e been addressed in research reports, bulle.
w'as identified in three llWRs and two PWRs. However, j
tins. a generic letter, and generic issues. 'ihe NRC also no single common fadure mode for all the 11 PRAs was conducted a study to evaluate the operating expenence of found.
l the ESW system. In July 1989, the NRC issued Genenc I ctter (Gil) 89-13 requesting that all licensees establish A pilot plant, which is an older llWR-4, was selected from programs to improve the perf ormance of the liSW sys-the 11 PR As for a value/unpact analysis. As a result of the l
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tem. The problems with this mstem identified in GL analyss, the !!SW system vulnerabihties were identified j
89-13 include biofouhng and corrosion /crosion.
and modifications were developed to address them. By usmg the NRC-sponsored PR A (NUREG-1150), the ef-The industiv responded to the concerns about ESW sys-c3 of one m more of these mWicadons were tkn tem unavadabdity by supporting Electric Power Research mmporated into t he ser tice water fault t rees to calculate 1
ec mp m plant N-institute (EPRI) research programs to improve the per-formanceof the1
% ater % or kmg G{SW system. F PRI established a Service Some modifications could be cost effective for the pilot roup (SWWG) and Semcc % ater As' sistance I rogram (SWAP) to conduct ESW-related re-plant. For example, the addition of standby auto-actua-Don logic would reduce the potential auto-start failure of I
scarch programs, mainly to address the concerns of Gl; an emergency service water pump following a loss of 89-13. In 1990, an EPRI-sponsored study of the ES%-
.te pown accident.1 or th.is modihcaton, if imple-t o
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system reliability was completed, and the results of the study were documea ~1 in NSAC-148, " Service Water mented, the potential reduction of the plant total CDl-Systems and Nuclear PlaJ Safety." The insights gained Ontonal plus manM p'nN ws esmad to ye i
from tlus st udy were incorpoed into the GI-153 study.
1.2E-05 per reactor-year. lhe valuchmpact ratio for this modification was calculated to be $380 per person-rem i
for a 16-year remaining lifetime.1hus, this modification For the GI-153 study,11 plant-specific probabilistic risk represents a cost-effective measure for teducing risk i
assessments (PR As)w ere reviewed to evaluate the contri-based on the current 51.000 per person-rem guideline.
hunon of the ESW system to the risk of core damage. The results of the review plus data from the six addit onal On the basis of the review of all 17 plant-specific PRAs PRAs reviewed under the EPRI-sponsored NSAC-148 and a detailed valuchmpact analysis of the pilot plant, the f
st udy indicate that this contribution is significant for some staff concludes the following' plants. The contribution of the ESW system to the risk of core damage from these 17 plant-specific PRAs consider-
'Ihe contribution of the ESW system to the risk of I
ing internal events only can be divided into four groups by core damage is substantial for some plants.
separating boiling water reactors (IlWRs) from pressur-ved water reactors (PWRs)and by separating older plants In general, the c4mtribution of the ESW system to from newer plants. The calculated contributions for these 1he risk of core damage is more significant for four groups are given in Table E-1.
IlWRs than PWRs. and more for the old vintage I
than the new.
As can be seen from Table E-1. there are wide variations in the estimated contribution to CDF attributed to the 1he impact of the ESW system on plant risk varies i
ESW system. The variations indicate that the impact of widely and is plant specific, primarily because of the l
service water systems on plant risk is largely plant specific, plant-unique design of the system and the degrce of 1he reasons for the bmad range include the varying de-dependency on the system.
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vii NUREG-1461
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Table E-1 ESW System's Contribution to Core Damage Frequency (Internal Events)
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Total CDF Contribution ESW System j
(Internal (Internal i
T pe Es ents)
Events)
Contribution -
l Plant 3
J llWRs Older 4.5E-06 1.4 E-06 30
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to to to t
4.7E-04 2.7E-04 65 i
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Newer 4.1E-06 5.6E-07 14 i
to to to i
{
3.0E-04 3.0E-05 26 PWRs Older 4.0E-05 1.5E-08
<1 to to to 3.7E-04 6.7E-05 19 i
-f Newer 1.4 E-05 2.4 E-07
<1 1
to to to
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l 1.7E-04 1.6E-05 19 2
The ESW system dominant failure modes found age serve as a complement to the SWSOPI program to l
from the review of the 11 NRC-sponsored PRAs assist in such areas as determining inspection priorities.
have some common aspects in different plants, even Other GI-153 information such as dominant sequences.
l though the system configuration for each plant re-ESW failure modes, and the effects of water quality could j
viewed is unique.
also assist the program in identifying and evaluating ESW i
vulnerabilities. In addition we note that the plant-specific y
In addition to the GI-153 study, there are several on-go-PRA developed under the Individual Plant Examination i
ing ESW-related regulatory activities. These activities in-(IPE) program is also a meaningful complement to the clude the Individual Plant Examination (IPE) program, SWSOPI program to achieve one of its objective, i.e., to j
Generic Issue 23 " Reactor Coolant Pump Seal Failures,"
identify and evaluate the ESW system vulnerabilities.
t and the most recently developed Service Water System Operational Performance Inspections (SWSOPI) pro-gram (SECY-92-355). The objectives of the SWSOPI With due regard for the insights from GI-153, the ESW 4
program include: (1) to identify and evaluate ESW design system operability, availabihty and reliability can be im-vulnerabilities; (2) to assess the ESW operation, mainten-proved significantly by implementing the SWSOPI pro-ance and personnel training; and (3) to assess the ESW gram, Generic Letter 89-13 requirements, the IPE pro-unavailability due to maintenance and component failure.
gram, the GI-23 rule-making, the maimenance rule, and f
GI-23 is proposing a rule-making to require means to the industry sponsored EPRI research program results.
- mitigate the risk from failure of reactor-coolant. pump Therefore, the staff concludes that ESW system reliabil-seal cooling. One way to accomplish this could be to ity is being adequately addressed by the on-going regula-reduce the dependence on the ESW system. 'Ilese regu-tory and industry initiatives. As a result, the objective of
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latory activities will directly or indirectly affect the assess-GI-153 is achieved and the issue should be considered
. ment of the ESW system contribution to CDF.The staff
" RESOLVED."'lhe need for future action (s) on ESW recommends that the insights from GI-153 regarding the system reliability is expected to be determined from these I
contribution of the ESW system to the risk of core dam-on-going programs.
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NUREG-1461 viii j
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1 STATEMENT OF TIIE PROllLEM 1.1 Ilackground Rh. Ecsc two reports completed the Phase I study. For ihis study:
'Ihe NRC and the nuclear industry have been concerned e
Eleven NRL.-sponsored PR As were reviewed to about the r eliahihty of the essential service water (ES%c) naluate ik impatana of k N, ytem using nstem for tears. The NRC concerns have been expressed coninbution to core damare frequency (( DF) as a m rescarch reports (Rel.1 and 2), bulletins (Ref. 3 and 4).
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a penenc letter (Ref. 5), and generic issues (Ref. 6). The NRC also conducted a study to evaluate the operatmg Dominant accident sequences involving ESW Fys-e experience of the ESW system (Ref. 7).The study showed tem failures were identified and the causes of the that %0 operating events in which the ESW system has rystem's unavadabihty were examined.
f,uicJ to operate as designed ha<J occurred; 12 of these events were considered as a complete loss of the ESW A valuenmpact analysis was performed for a se-system. The causes of the system's failure and degrada-lected prototypical plant to demonstrate potential tion melude various foulmp mechamsms (deposition of improvements of the ESW system for reducing risk sediment, biofouhny, cortosion and crosion, and intru-sion of forcipn material and debris), single failures and
'lhe review of the 11 plantapecific PR As showed that other design dehciencies, floodmg. multiple equipment ESW systern vulnerabihty is a significant contributor to f ailures. and personnel and procedmal errors.
plant CDF and that cost-c!fective incasures for reducing risk are feauble.These insights from the Phase i scoping The induun responded to the concerns regardmp the study and the results of thc EPRioponsored study led the staf f to conclude that there was suf ficient basis to recom-unavadabihtv of the ESW miem Ire supporting EPRI research pr$ prams to impro've the perfoimance of the mend a r esolution ior GI-153. Consequently, the Phase II ESW syst em. EPRI established a Service Water Wor king (mor e detuded evaluations of individual plants) would not he an cf hcient use of adJed resources because the Phase I Group' to conduct ESW-related research programs, mainly to address the concerns of GE E9-13 (Ref. 5). In scopmg study had met the objectives of the issue.
1990, an EPHI-sponsored study on the ESW system reli-ability was cornpleted and tbe results w ere documented in
],2 Safety Significance NSAC-14 " Service Water Systems and Nuclear Plant Safety" (Ref. 8). I or this study, six plant-specific PRAs
'Ihe ESW system a a nuclear power plant supplies cool-ware evaluated to ascens the centnhution of the ESW ing water to transfer heat f rom vanous safety-related and system to the risk of core damare.
non-safety-related systems and equipment to the ulti-mate heat sink of the plant. It takes suction from the ultimate heat sink (e.g., the ocean, bay, river, lake, pond In 1971, the NRC sta!f comp!cted a stude pertaining to an ESW-related peneric issue G1-130, " Essential Service or cooling tow ers), removes hcat via heat exchangers from Water Pump 1 adures at Multi-Plant Sues? The insights the varie systems am! mmponents it serves, and dis-pained from the study indicated that the ESW system charpes the water back to the ultimate heat smk. It is problems were safety sipmficant and miph! not be con-known by different names at various light-water-reactor fined to the seven multi-urut sites adJressed under plants. In PWR plants, it may be refers ed to as the "essen-
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GI-130. As a result,Ihe staff concluded that the reliability tial service water system", the " emergency equipment of the ESW sistem for all other reactors (66 PWR units moling water system", the " essential raw coolmg water and 3S flWR' units) should be evaluated under a new system", the "s:dt water cooling system", the " nuclear peneric issue, that ik, GI-153, "less of Essential Senice service water symm", or others. In llWR plants,it may be Water in 1 WRs/ Under this peneric issue, all potential referred to as the " emergency equipment cooling water c:mses of the unavadabilitv of the ESW system were Ptem", the "Mandby semce water system", the " plant evaluated except that are [onsidered to be' resolved by service water systern", the " residual heat removal senice implem enting the resolutions in GL 89-13 (such as thosc water system", or others.
pertaining to biofouling).
.The design and operational characteristics of the ESW system are ddferent for PWRs and IlWRs. In addition, GI-153 was mitiated m July 1991, anJ the tasks were the design and operational characteristics differ signifi-oripmally divided into two phases: Phase ' for a scopinE cimtiv from plant to plant within each of these reactor study anJ Phase H for a penene evaluation The Phase I typei scoping study was completed and was documented in NURI.G/CR-5910 (Ref. 9). A supplemental study was The L5W syttem, which is a support system, is needed in also completed and documented in a letter report (Ref.
every phase of plant operations. Under accident condi-L 1
NUREG-14til
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l tions, it supplies adequate cooling water to systems and damage accident, posing a significant risk to the public.
l components that are important to shutdown the plant safely or for mitigating the consequences of the accident.
Under normal operating conditions, it pmvides compo.
1.3 Objectives neni and room cooling (mainly via the component cooling l
water system). During a shutdown period, it also ensures
'Ite objectives of GI-153 are to (1) assess the safety sig-that the residual heat is rem.oved from the reactor core.
nificance of the loss of ESW systems in light water reac-The ESW system may also supply makeup water to fire tors (LWRs) and the corr esponding contributions to core i
protection systems, cooling towers, and water treatment damage frequencies (CDFs), (2) perform a value/ impact systems at a plant.
analysis of a prototypical plant to demonstrate the feasi-bility ofimproving the reliabihty of the ESW system,and '
A complete loss of the ESW system could lead to a core-(3) propose a viabic resolution for the issue.
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NUllEG-141 2
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2 TECIINICAL FINDINGS The events extracted from the precursor reports 1his section presents the technical Imdmps from the e
Phase I scoping study and the supplemental study on were judged to be potential precursors to severc GI-153.1hc detaded descriptions are given in NUREG!
core damage accidents and did not necessarily in-CR-5910 and a Iciter report (Ref. 9 and 10).
volve complete loss of senice water, either poten-tially or operationally. Therefore, only limited over-lap with the results of NUREG-1275 should be 2.1 Review of. O.perat.mg Experience expecteo.
Several studies have addressed the operating experience As indicated previously, because the studies docu-related to service water system failures. The work is not e
detaded here. but the results are summarued. It should mented in the precursor reports and NUREG-1275 he noted that these studies were performed bv different were performed by different people for different people for ddferent purposes.Therefore, although some purposes, ddferent conclusions about the same independence of analyses is observed in the studies, the events are not unexpected.
results are not completely consistent.
2.1.3 Analysis of ESW System at Multi-2.1.1 Precursor Reports Unit' Sites
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Under the Accident Sequence Precursor Program. the staff at Oak Ridpc National laboratory reviews licensee lhe analysis documented in NUREG/CR-5526 (Ref.1) event reports (1 ERs) of operating eients at I WRs to show ed that the dominant failures causing partial or com-identify and categorue potential precursors to severe plete loss of the ESW system are fadure of travehng screens and common intake structure, failure of the ESW core-damage accidents. Accident sequences considered in this program are those associated with inadequate core pumps, loss of electric power to the ESW system, and coohnp. As a result, several status reports (Ref.14-21) operator error related to the ESW pumps. Degradation have been pubbshed that descobe those events that have of the ESW system was caused by sediment, corrosion, occurred as reported in LERs. A review of these for and rnechanicrd and electrical problems associated with senice water-related events showed that 24 esents were the ESW pumps. As in the other reports reviewed, there directiv related to this study. Table 2.1 lists these events are no special fadure modes that would chanpc the basic and provides a desenption of cach. As a proup, the events approach used to analy7e the pilot plant in this study.
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represent a variety of cautes.
This analysis showed that, for a!! PWRs 12 operating 2.1.2 Operating Experience Feedback Reports events from the 1970's to 1990 for all the PWRs resulted m a complete loss of the ESW system function. Of these 12 events,6 are the same events as those reported in NUREG-1275 (Ref. 7)is a comprebensne study of serv.
NUREG-1275. The other six are shown in Table 2.3.
ice water.related operating events. Of the 980 cvi:nts identified. 276 were considered to hase potential generic safety signific;mce. The causes of all the esents were 2.1.4 Operating Experience after Publication categorized as follows-of NUREG-1275 (1987-1992) foulir'g 58.3 %
In NUREG-1275, the NRC staff reported Ihe results of a e
single failures 6.5%
review of operating experience of commercial nuclear e
multiple failures 3.6 %
power plants from 1980 to 1987. Since then, licensecs o
personnel errors 16.7 %
have continued to experience operating problems with e
fhioding' 4.4 %
the ESW system (Ref. 22). 'lhe prob! cms mclude inade-e e
seistme 10.5 %
quate flow distribution, system misalignment, fouling /
biofouling, clogged screens, design deficiencies, system degradation, mamtenance or operator errors. These 1 wc!ve events were r eported as complete loss-of-service-problems have occurred in flWR as well as PWR plants.
water events.1hese events are tabulated in Tab!c 2.2 with Some of the operating events involved the complete loss simplified desenptions that demonstrate the diversity of or potentialloss of the ESW sysicm. As a result,the NRC l
the failures. Only two of the events in Table 2.2 appear in staff issued Information Notice 92-49 (Ref. 23) on J uly 2.
Table 2.1. The following reneral observations can be 1992. The operatmp events de. cussed in the information made:
notice are summari/cd m Table 2.4.
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Table 2.1 Precursons to Potential Severe Core Damage Accidents Imohing Senice Water Systems I
f Plant 11R Number Description I
Ilatch 1 II R 321/60-103 Inlet strainers partially clogged.
San Onofic 1 11R 206/80-006
'lhree salt water cooling trains failed.
l St. Lucie 1 IIR 335/80-029 RCP seal cooling lost due to inadvertent valve closure.
Calvert Cliffs 1 L1!R 317/80-027 Two service water pumps fail due to loss of compressed air, i
li Pilgrim 1 IER 293/80-070 Component cooling water lost due to maintenance and breaker trip.
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Salem 1 IfR 272/80-060 Imst SW to DG due to valve indicating open when actually closed.'
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Kewaunce LIIR 305/81-033 Operator error-two component cooling water trains unavailable.-
f San Onofre 3 LER 262/84-035 Operator error-outside limiting condition he operation.
l Surry i IJiR 280/84-011 Operator error-safety injection pump CCW supply found f
isolated.
I Salem 2 IJiR 311/85-018 Operator error-maintenance and closed valve could not be opened.
1.2Salle 1 LER 373/85-045 less 01 non-safety service water due to expansion joint failure.
Susquehanna 2 IER 388/85-014 Emergency service water failed during testing.
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4 and 85-015
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Surry 1 IIR 280/66-0291 Service water subsystem pump lost due to air binding.
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l McGuire 2 LER 370/87-016 Tnp with senice water train out for cleaning.
t and 87-017 p
i Palisades 1 E R 255/88-021 Incorrectly set relays could have resulted in loss of senice water.
Zion 1 IER 295/88-019 Potential component cooling water failure due to design f
deficiency.
[
Davis liesse L11R 346/88-007i Possible prolonged loss of instrument air would cause SW to I
isolate.
[
San Onofre IER 361/88-0102 Emergency cooling water unavailable due to low freon in chillers.
j Farlev 1 and 2 1.ER 348/88-0181 Postulated loss of senice water due to fire.
Peach Hottom 2 lER 277/89-002 Unacceptable emergency senice water performance due to 1&C problems.
I Calvert Cliffs 1 IFR 317/89-023' Potential pipe rupture could fail both senice water pumps.
t Davis 11 esse 1 IFR 346/89-004 Potential pipe rupture could fail both service water pumps.
{
Nine Mile Point 2 Ll!R 410/89-002 Potential senice water and ECCS pump failure due to flooding.
River Bend 11R 458/89-020 Senice water flooded auxiliary building impairing electric power f
and control.
'I.tsted in Ntml.G-1275 tat >Ies as a semcc uter event involving eququnent failuret
[
i NURliG-1461 4
i t
'l Table 2.2 Twelve Events l' rom NUltEG.1275 Resulting in Complete less of SW Function n
Plant LER Number Description Occmce 1 LER 269/86-11 Inadequate siphon flow to service water pumps.
t l
a Susquehanna 1 LER 387/86-21 All senice water pumps failed due to operation below design flow.
l Oyster Creek LliR 219/85-18 ileat exchanger plugged by coal tar enamel.
l Brunswick 1 LER 325/84-01 Entrapped air in suction header piping.
Palisades LER 255/84-01 less of power to senice water pumps due to operator error.
Salem 2 LER 311/83-32 Senice water bay flooded due to failed piping gasket.
l Salem 1 LER 272/82-15 tess of vital bus when 1 train of senice water out for I
maintenance.
I Brunswick 2 IJiR 324/82-05 All pumps failed to start due to low suction pressure and sediment in sensing lines.
liatch I LER 321/80-103 Inlet strainers partially cloged.
I j
San Onofre 1 LER 206/80-06 One pump shaf t sheared and valve in other train failed.
i Calvert Cliffs 2 LER 31S/82-34 Failure of common valve in discharge header.
i Catawba 1 LER 413/S5-68 Train A input vahe failed to open and train H discharge valve failed to open.
t Table 2.3 Additional Events From NUREG/CH-5526 Resulting in Complete Loss of SW Function for PWRs Plant I.ER Number Description Salern 1 NPE/PWR.2 Winter storm shut down the ESW system. Traveling screens i
VIII-110 blocked by ice.
)
(1976) i Farley 1 NPE/PWR-2 Flooding of the intake structure.
l Vill 155 (1978)
Salem 1 LER-272/84-14 Vital bus 1 A failed, bus IB in maintenance, bus 1C did not l
energi7e, loss of the ESW system.
l Crystal River LER-302/S6-02 All ESW pumps are shut down, two divers drowned.
I San Oncfre 2 & 3 LER-361/83-72 Traveling screens were damaged, CCW heat exchanger clogged.
Vogtle 1 March 20,1990 Imss of all safety ac power in cold shutdown. Offsite power was lost due to a truck accident. The emergency diesel-generator i
i ~
tripped upon start.
i 5
- ~-.
--. -. ~
Table 2.4 Recent Events Resulting in the l>>ss of SW Function
)
i Plant LER Number Description
)
Millstone 1 LER 245/90-16 Storm-induced high winds and seas caused an excessive amount of seaweed to accumulate on the traveling screens for the service i
water intake resulting in degraded ESW system. The operator
{
tripped the reactor from 45 percent of full power.
Fitzpatrick LER 333/90-23 Excessive fouhng on the trascling screen caused the screen to bow -
inward due to high differential pressure across the screen. The l
operator scrammed the reactor from 45 percent of full power.
}
i ANO-1, Unit 2 LER 368/91-12 Debris from the lake bypassed the screens and clogged the senice j
water pump strainers of both loops. The licenscc declared both loops of the senice water inoperable with the reactor in startup conditions.
t Nine MUe Point 1 LER 220/92-05 Complete loss of the ESW system occurred when the licensee in advertently closed all gates to the senice water intake bay whi!c I
the reactor was shut down.
i I
i 2.2 Review of I'lant-Specific the ESW system to plant internal events CDF is about 20 j
Probabilistic Risk Assessments P ""! I r th* "***' P' "ts nd 46 percent for the older i
plants.
t 2.2.1 Contribution of ESW system to Risk of m.2 PWR Plants Core Damage l
For the GI-153 study, seven PWR plant-specific PR As.
For the GI-153 study,11 plant-specific PRAs were re-werc reviewed. Table 2.6 tabulates the contribution of the i
viewed to evaluate the contribution of the ESW sysicm to ESW system to CDF for these seven plants. Given in the the risk of core (lamage. The results of the review plus table are four additional.PWR plant specific PRAs re-data from the six PRAs revicwed under the EPRI-spon-viewed under the EPRI-sponsored study (Ref. 8).
sored study (Ref. 8) indicate that this coniribution is sig-nific mt for some plants. The contribution of the ESW As can be seen in Table 2.6, the contribution to the total system to the risk of core damage from these 17 plan 6spe-internal events CDF varies from 2.4E-07 to 1.6E-05 per ctfic PRAs can be divided into four groups by separatinE reactor-year for newer PWR plants and from 1.5E-08 to i
BWRs from PWRs and by separating older plants from 6.7E.05 per reactor-year for older PWR plants, or ranges newer plants, using 1976 as the dividing year.
from < 1 to 19 percent for the newer plants and also for
[
older plants.The average contribution of the ESW system
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2.2.1.1 BWR Plants to plant in crnal events CDF is about 7 percent for the newer plants and 12 percent for the older plants.
l For the GI-153 study, four BWR plant-specific PRAs were reviewed. Table 2.5 tabulates the contribution of the 4
2.2.2 Discussion of Plant-Specific ESW system to CDF for these four plants. Given in the Probabilistic Risk Assessments table are the results of two additional BWR plant-specific f
PR As reviewed under the EPRI-sponsored study (Ref. 8).
Tables 2 5 and 2.6 show a broad range of the contribution I
to CDF from plant to plant.This variation may be due to the degree that a plant is dependent on senice water, the As can be seen in Table 2.5, the contnbution to the total reliability of the USW system itself, and to some extent, f
internal events CDF varies from 5.6E-07 to 3.0E-05 per the differences in modeling assumptions and scope of reactor-year for newer BWR plants and from L4E-06 to each NRC-sponsored PRA program. 'Diese conclusions 2.7E-04 per r eactor-year for older BWR plants. or ranges are consistent with the observations in NSAC-148 (Ref.
from 14 to 26 percent for the newer plants and from 30 to 8). Other observations are discussed in the following sec-l 65 percent for older plants. The average contribution of tions.
. _ ~
l Table 2.5 ESW System Contribution to Cose Damage Frequency,llWRs Onternal Events)
Total CDF ESW System ESW System j
(Internal CDF l
l Plant Type Events)
Contribution Contribution l
t i
Plant 11 1 Older 9.9E-05 3.011-05 30 i
l l
Plant B.2 Older 2.91i-04 1.9E-04 65 -
Plant 11-3 Older 4.5E-06 1.4 E-06 32 Plant B-4 Newer 4.11!-06 5.6E-07 14 l
f Plant C*
Older 4.7E-04 2.7E-04 57 Plant P Newer 3.0E-04 3.0E-05 26 BWR Average 2.0E-04 8]E-05 37 i
t L
- Plants were sJentifwd in NSAC-148 (Ref. 8). PIM for Plant C included internal and ex1ernal events.
{
I L
i Table 2.6 ESW System Contribution to Core Damage Frequency, PWHs Onternal Esents)
+
1 l
Total CDF ESW System ESW System l
Onternal CDF G~
Plant Type Events)
Contribution Contribution i
Plant P-1 Older 1.3E-04 1.4 E-05 11 I
j Plant P-2 Older 1.4 E-04 2.6E-05 19 l
Plant P-3 Older 7.1E-05 3.4 E-06 5
Plant P-4 Newer 1.4 E-05 1.8E-06 19 i
Plant P-5 Older 8.81i-05 1.lE-05 12 l
i l
Plant P.6 Older 4.0E-05 1.5E-08
<1 6
l t
Plant P-7 Newer 5.7E-05 2.4 E-07
<1 1
Plant A*
Older 6.7E-05 1.2E-06 2
j Plant 11' Newer 1.7E-04 1.6E-05 9
Plant D' Older 3.7E-04 6.7E-05 18 i
Plant E' Newer 2.0E-04 9.7E-06 5
l 1
PWR Average 8.2E-05 1.4E-05 10 l
- Plants were identified in NSAC-148 (Ref. E). PRAs for these plants included internal and external events.
I 2.2.2.1 Dominant Sequences For PWRs, medium and small lossef-coolant-accident Dominant sequences with regard to total CDF due to all sequences tend to dominate contribution to total plant l
causes are distinctly different in PWR and HWR plants.
internal events CDF (r:mging from 30 to 86 percent).
j 7
~-~.
When considermg the contribution of the ESW system to failure. 'Ihe second failure mode can occur if the dis-the risk of core damage to PWRs, the ESW system tends charge check valve fails to reclose in one pump train of a to also dominate m these sequences v.here the ESW multiple pump syst em where the pumps are all eross-tied.
system fa ls to provide cooling to the high and low pres-The failure of the check valve to reclose occurs when one i
sure cooling systems cither in the injection phasc or in the of the operating pumps is shut down.This allows flow recirculation phase nf operation. *Ihat is, the ESW system from the other pump (s) to recirculate back through the fails to provide cooling to the injection system pumps or idle pump resulting in functional failure of the system.
I pump room coolers, thereby causing loss of injection; or the ESW system fails to provide cooling to the residual i
2m Mects of Water Quality on ESW System heat removal (RHR) heat exchanger, thereby causing
. Die NRC-sponsored study on service water system aging _
f failure of low pressure recirculation.
showed that poor water quality which results in sitting, However, for BWRs the contribution of the ESW system corrosion and fouling can cause ESW system component to the risk of core damage is predommantly m sequences failures (Ref. 24, 25). The EPRI-sponsored study (Ref. 8) where the ESW system fails to provide cooling to the also indicated that water quality has a major effect on the RilR system in the suppression pool cooling mode (due reliability of the ESW system. However, because water to loss of an ac bus as an initiator or loss of offsite power).
quahty varies from plant to plant and because generic
[
data instead of plant-specific data were used in the PRAs Station blackout sequences were considered in the Ge-reviewed for the GI-153 study to quantify the system I
neric Issue 23 as the prime accidents that would result m models, the effect of water quality would not be expected f
j failure of the reactor pump scals of PWRs. However, note to be a major finding in the GI-153 study. Therefore, that when the GI-153 study reviewed the dominant acci-water quality problems (e.g., above-average maintenance j
dent sequences for the contribution of the ESW system to cutage times for the service water system and higher r
1 the risk of core damage, station blackout sequences were component failures rates) typically would not be ac-not considered because safety systems which depend on counted for.
)
the ESW system will not be functional following a station Nackout regard! css of the availability of the ESW system.
Nevertheless. r GI-153 sensitivity analysis was performed Hus may partly cxplam why the GI-153 study did not for the pilot plant ESW system components partly to j
reveal the seal failure as one of the dommant sequences determine the importance of water quality problems.The as discussed above. Other aspects of the differences be-following component failure modes were selected as be-tween the G1-23 study and the GI-153 study m terms of ing susceptibic to water quality problems:
the risk assessments of the ESW system are discussed in Section 3.4.1.
~
check valve back leakage e
check. valve failure to open 2.2.2.2 ESW S,5 stem Failure Modes a
air-operated valve failure to open i
e air-operated valve unavailability because of
(
The ESW system dominant failure modes found from the e
i review of the 11 NRC-sponsored PRAs tend to have some maintenance I
common aspcets in different plants, even though the sys-plugged manual valve j
q tem configuration Er cach plant is unique. These failure motor-driven pump failure to run o
modes are listed in Table 23, together with their relative i
- t r-drive pump unavailability because of contributions to internal events CDF at those sites wherc mamtenance they werc among the dominant contributors. Note that no single common failure mode for all the 11 PRAs was found-The analysis was performed by adjusting the probabilities of all of the selected failure modes by a multiplier and
[
The two most common ESW system faults were the de-recalculating the CDF for the existing accident sequence l
pendency of the ESW system on motor-operated isolation cut sets, increasing the prouabilities of these failure valvcs to open on demand to supply cooling water to modes by a factor of 3 resulted in a 40 percent increase in safety-related loads and failure of the standby service the total CDF from internal events. Increasing the prob-i water pumps to start. Two other subtle failure modes are abilitics by a factor of ten increased the total internal (1) failure to isolate nonessential cooling water loads and events CDF by 150 percent. 'Dius, the sensitivityanalysis (2) failare of cross-tied pumps as a result of back flow showed that water quality could have a significant effect through a pump discharge check valve. The first failure on the contribution of the ESW systern to CDF, as was mode c;m result m madequate couting of the essential also concluded in the EPRI. sponsored report (Ref.8).
l hiads because of the diversion of water away from the The insights pained from this analysis support the need to essential loads and the potennal for pump runout and use plant-specific data as indicated in Section 4.3.
t r
[
e
Table 2.7 1:SW System l'ailure Modes Dominant at More Than One Site Percent Contribution to Internal Events FSW S3 stem Failure Mode Plant Type CDF s.
Failure of ESW system motor-opemted or air-operated Plant B-2 BWR 18 isolaties valve to open on demand to supply cooling water to Plant B-3 BWR 4
safety-related loads Plant 11-4 MWR 7
Plant P-1 PWR 6
Plant P-5 PWR 11 Failure of standby service water pump to stan Plant B-1 BWR 4
Plant B-3 BWR 2
Plant B-4 BWR 3
Common-mode f:ulures of pumps Plant B-1 BWR 5
Plant B-4 BWR 4
Plant P-4 PWR 13 Plant P-5 PWR 1
Plant P-2 PWR 7
Plant P-3 PWR 4
Failure of isolation v:dve to isolate nonessential cooling Plant B-2 BWR 34 water huds Plant P-3 PWR 2
Plugged manual valve Plant P-1 PWR 1
Plant P-2 PWR 2
Unavailability of ESW loop because of maintenance Plant B-2 BWR 12 Plant B-1 BWR 19 2.3 Evaluation of Selected age are possible even for a plant with a relatively low risk pertaining to the ESW rystem.
Plant-Specific PRAS In this section, the staff discusses the evaluation of the 2.3.1 Plant 11-2 contribution of the ESW system to the risk of core dam-age for Plant B-2 and Plant B-3.These two BWR plants The ESW system of Plant B-2 includes the reactor build-were selected because Plant B-2 represents the upper ing closed cooling water (RBCCW) system and the reac-end in terms of this contribution and Plant B-3 represents tor building service water (RBSW) system. The vul-the lower end as shown in Table 2.5. Plant B-2 was evalu-nerabilities of the ESW system were investigated under ated under the Unresolved Safety Issue (USI) A-45 pro-the USI A-45 propram (Ref. 25), and the dominant ESW gram (Ref. 25). The results of the evaluation are summa-component failures and unavailabilities were identified as rized here to illustrate the safety significance of the ESW the following:
system. A detailed value/ impact analysis of Plant B-3, selected as the GI-153 pilot plant, was performed to dem-(1) failure of the RBCCW noneritical header isolation onstrate that cost-effective modifications to reduce the valve (motor-operated valve) to isolate nonsafety contribution of the ESW system to the risk of core dam-loads (31%)
(2) failure of the RBCCW isolation valve (two motor-flood were found significant in regard to plant CDF and I
operated valves) to safety loads to open (189c')
were evaluated for potential modifications.The proposed l
modifications related to the ESW system include (1)
)
- 13) failure of the RBSW noncritical header isolation strengthening of the RBCCW heat exchanger supports, valve (motor-operated valve) to isolate nonsafety (2) addition of a 1-hour fire barrier around RBSW power l
loads (3%)
cables, and (3) development of procedures Ior safe shut-down during very high flood crests. Implementation of l
(4) unavailability of one of 1he two RBSW loops be-modifications to the ESW system for external events and
}
cause of maintenance (5%)
for internal events could reduce the plant notal CDF from 4.4E-4 to 1.8E-4 per reactor-year, or a reduction of i
(5) unavailabihty of one of the two RBCCW loops be-2.6E-4 per reactor year. These modifications u.: Smd cause of maintenance (7%).
to be cost-effective.
2.3.1.1 Contribution of ESW system to Internal 2.3.2 Plant 11-3 Iaents CDF De contribution of each potential vulnerability to the 2.3.2.1 ESW System Vulnerabilities l
plant internal events CDF is as follows:
?
i For the GI-153 scoping study (Ref. 9), a detailed value/
l j
Vulnerability ACDF per impact analysis of a BWR-4/ Mark I plant (Plant B-3)was reactor year performed. Although similar in general design, the ESW system of Plant B-3 is substantially different from that of 1
9.0E-5 Plant B-2. De ESW system of Plant B-3 for safety-re-
[
2 5.2E-5 lated equipment consists of the emergency senice water 3
0.9E-5 system, the emergency heat sink, the high pressure serv-4 1.5E-5 ice water system and the reactor building cooling water 5
2.0E-5 system. Ilowever, the dominant basic events which con-tribute to the plant CDF involve the emergency senice Total 1.9E_4 water (EMSW) system and the emergency heat sink l
. (Ells) only, i
The study showed that the modifications to reduce vul' The EMSW system of Plant B-3 is designed to provide -
I nerabilities 1, 2 and 3 above would be cost-effective.
7 cooling wat er for the diesel-engine coolers and the emer-
[
nese modifications include the following:
pency core cooling system equipment room coolers during i
q a loss of offsite power.The system consists of two full-ca-Addition of a bypass line around the normally closed pacity pumps installed in parallel and is common to both
~
1 e
3 motor-operated isolation valves m the RBCCW Units 2 and 3.nc normal suction source for the pumps is pump discharge a pond.He pump discharge piping consists of two head-f ers with service h) ops to supply the diesel-engine coolers Addition of a second isolation valve to the RBCCW and selected equipment coolers. A common discharge noncritical supp!y header header routes the system effluent back to the pond.The configuration of the EMSW system is showed in detail in Addition of automatic closure logic to the second the scoping study report (Ref. 9).
i e
3 isolation valve for the noncritical header of the RBSW system Both EMSW pumps start automatically whenever Implementation of Ihese modtfications would reduce the standby diesel-generators are started. One of the EMSW t
plant's internal events CDF by 1.5E-4 per reactor-year, pumps is m nually shut off if both pumps are runnmg.
u hich represents about 52 percent of the plant totalinter-Should the EMSW pumps fail, the EMSW may be oper-nal events CDF.
ated m conjunction with the Ells.
i 2.3.1.2 Contribution of ESW system to External Thc Ells functions as a backup to t he EMSW system.The Events CDF Ells consists of an mduced-draft coohng towerand a full l
capacity Emergency Cooling Water (ECW) pump. The The exte nal events identified under the USI A-45 pro-cooling tower will be used to cool the return water from gram included seismic events, fire, internal flood, exter-the ESW system loads, and the ECW pump deliveis the l
nal flood. extreme wind, and lightning. liowever. only the cooling water back to the ESW system through the i
vulnerabilities rclated to seismic events, fire, and external EMSW piping and valving.
NUREG-1461 10 i
Ilased <m the design and the operational modes of the maintenance to senfy operability. Therefore, the failure EMSW system described above vulnerabildies and asw to restore EMSW components is a direct result of the ciated potennal modificruions were identdied as follows:
operator f ailing to follow plant procedures.
l'ulncrubihty 1: Operator 1 ails to Operate Ibe FCW Madification 3: Provide Addnional Operator Training Pump and/or Revise Procedures l
Following a loss of offsne power, the ECW pump auto-
~llus is the same modification as proposed for Vulnerabil-J matically starts after a 22-second time delay following an i!F -
)
1 emergency diesel generator auto start. If the discharge pressure for the EMSW pumps ap;, cars normal the op-t{ulnnyn 3: Discharge Check Valve Failures Fail erator will shutdown the ECW pump. If, later in the ross-Fied EMSW Pumps accident, the operating EMSW pump trips and the
.lhis failure of the nmning EMSW pump is caused by standby EMSW pump fails to start or run, the operator failure (hack leakage) of the standby EMSW pump dis-must manually restart the ECW pump. This vulnerability charge check valve. That is, when the standby pump is addrenes the operator failure to restart the ECW pump secured following auto-actuation with the chosen pump following a delayed failure of the EMSW pumps.
running, the flow from the operating pump recirculates i
back through the failed check valve and the standby pump Modification 1: Addition of alhird EMSW Pump and results in functional failure of the operating pump.
I
'lhe addition of a third EMSW pump that would auto start Modification 4: Addition of a Second Pump Discharge on diesci auto start or low EMSW system pressure would Check Valve j
i increase the reliability of the EMSW system. Such an j
addition would reduce dependance on operator actions to This modification would prorde a second pump discharge
)
initiate the cmcrgency heat sink and add fienbility in check valve m senes with the existing pump discharge response to a loss of offsite power accident.
check valve to reduce the probability of this occurrence.
i a
Modification 2: Atkhtion of Standby Auto Actuation Modification 5: Inct case System Functional Testmg Fre-I logic for the liCW Pump quency for EMSW Pump Di< charge Check Valves i
l 1he addition of standby auto-actuation logic would de-This modification would increase the EMSW system test
[
mand the liCW pump to auto start and the pump dis-frequency from quarterly (current frequency) to monthly.
I J
charge valve to open on low EMSW system pressure aft er By doing so, the probability of a check valve failing to l
)
the emergency diesel generator auto start s:gnal has been reclose would decrease by a factor of three.
i received.1his modification in of feet woulJ have the ECW j
pump respond as if it were a third EMSW pump and l'ulnuaN/hy 4:
EMSW Pump Hardware Faults would climinate the dependency of the ECW pump on the opcmtor following a low of offsite power transient.
The EMSW consists of two redundant, cross-tied pump j
trains. The success criteria established is one of two i
Mothfication 3: Provide Additional Opemtor Training, EMSW pumps operating delivenng flow or the ECW j
]
Revise Procedures Add Additional Alarms in the Con.
pump delivering flow to the ESW system. Dominant iail-j trol Room ure modes pencrally consist of one EDG failing, which i
fails one EMSW pump, and the resulting available pump This modification wou'd provide additional operator failing to start /run, and the operator fails to initiate the l
training and revision of procedures to enhance the opera.
ECW pump.
(
tors response to conditions requ ring the starting of the Modification 1: Addition of a 'lhird EMSW Pump j
ECW pump. Add:tional alarms and indication would be 1 tis s the same modification as proposed forVulnerabil-l provided to aid the operator.
"Y l'ulncichility 2: Failure to Restore EMSW Components l
After Maintenance
%dificmion 2:
Addition of Standby Auto Actuation logic for the ECW Pump Fa!!ure to restore an EMSW motor dn.ven pump or EMSW diesel generator cooling compcments defeats lhis is the same mothfication as proposed for Vulnerabil-one-half of the EMSW system. Restoration of these com-ity 1.
ponents after mainten:mcc is performed using wntten
~
procedures with independent venlication. Functional l'ulncrabiluy 5:
Failure of EMSW to Cool the EDGs testing of these components is performed following Due to AOV Failures 11 NUREG-1461
-- I
The EMSW provides cooling water to the emergency For the analysis, ESW system failures that are a direct dicsci pencrators.1he EMSW outlet header from each consequence of the external event and random failures emergency dicsci pencrator contains an air operated iso-that cause a component to be unavailable at the time of lation valve that is signaled open when its respective die-the external event were considered. The analysis showed set pencrator starts. Failure of this vah e to open on diesel that the overall fire induced CDF is 2.0E-05 per reactor-start defeats EMSW coohng to that diesel which results in year. of which 65 percent or 1.5E.05 is the contnbution of failure of the respective diesel generator. Failure of the the ESW system. For scismic-induced CDF, using the diesel generators following a loss of offsite power results frequency assumed by Lawrence Livermore National in a station blackout.
Laboratory (11NL), the total scismic-induced CDF is 7.7E-05 per reactor-year and the contribution of the ESW Modification 6: Addition of a Check Valve in Series to system to that amount is 2.4E-05 per reactor-year. If the the Diesel Generator AOVs seismic frequency assumed by EPRI is used, the total seismic. induced CDF is 3.1 E-06 per reactor-year and the
'Ihis modification would remove Ihe demand on the contribution of the ESW system to that amount is 1.1E-06 AOVs to open on diesel start by making them normally per reactor-year.
open valves and installing a check valve in series with the Table 2.10 shows the contributions of the ESW system to AOV.
internal and external events CDF. These values are based on the 11NL seismic hazard estimates. As c:m be Modification 7:
Addition of a Swing, Self. Cooled Die-seen from Table 2.10, the ESW system contribution to the sel Generator risk of core damage increases from 1.4E-06 per reactor yar for internal events only to 3.8E-05 per reactor year Tius modification consists of the addition of a swing, self-c nsidering both internal and external events. As a result, cooled diesel pencrator that would auto start in the event the valuebmpact ratio for Modihcatmn 2 was reduced of loss of normal sources of power 10 the onsite power fmm a value of $3R0 per pmmem for internal events syst em ~
only to 5380 per person-rem if both internal and external events were considered.
lhe vulnerabilities and the proposed moJifications are summarized in Table 2.8.
2.33 Loss of ESW System as Initiator 23.2.2 Value/1mpact Anal sis less of the ESW system as an initiator was not considered 3
in the USl A-45 study and the NUREG-1150 analysis A value/ impact analysis was performed for Plant 11-3, and (Ref. 25 an 26). The report on the USl A-45 study did not the results of the analysis are given in Table 2.9. The give a reason for excluding loss of the ESW system as an analysis is presented in detail in Reference 10. As ccm be nitiator. NUREG-1150 indicated that, on the basis of a seen from Table 2.9, Mochfication 2 scores the lowest general review, the possible catises for the ESW system valuchmpact ratio of $3SO per person-rem (internal and being an initiator were unlikely. No detailed analysis was external events) which represents a cost-effective meas' performed to substantiate this conclusion.
ure for reducing nsk.
How ever, on the basis of the rewew of ESW system oper-Note that the external events CDF for Plant 11-3 were ating experience (Ref. 7), the frequency of :omplete loss dominant for the valuchmpact analysis. The effect of of the ESW system is relatively high (1.8E-U per reactor-external events is discussed in the following section-year). Consequently. loss of the ESW system should be considered a credible initiator. Shoald the effect of the 2.3.23 Contribution of ESW to External Esents cdp ESW system as an initiator be considered, the proposed modifications would become even more favorable be-The scoping st udy evaluated the ext ernal events contribu-cause of the increase of the contribution of the ESW tions for Plant 11-3 using the analysis provided in system to the risk of core damage.
NUREG-1150 (Ref. 27). NUREG-1150 provides an analysis of external events including hazards duc to carth-23.4 Uncertainty Considerations quakes, fires, external and internal flooding, and extreme winds and tornadoes for Plant 11-3. The analysis showed Uncertainty arises from the selection of the database used that all external hazards except fire and seismic events are to determine parameter values, modeling assumptions, negligib!c contributors to the risk of core damage. For the and completeness of the analysis. Because the database scopmg study, the fire and scismic events assessments of and model from the previous NRC.sp(msored PRAs NUREG-1150 werc analyzed to determine the contribu-(NUREG-1150, US1-45. and IREP) were used for the tion of thc ESW fadures.The detailed analysis is provided GI-153 study, the uncertainties of these PRAs wdl be-in Reference 10.
come the inherent uncertainties of the GI.153 study.
NUREG-1461 12
Table 2.8 ESW System Yulnerabilities and Modifications Vulnerability Alternatise Modifications 1.
Failure of operator to 1.
Addition of a third emergency service water operate cmergency heat sink pump
- ESW pump hardware faults 2
Addition of standby auto-actuation logic
- failure to restore ESW components after maintenance for the emergency cooling water pump 3.
Additional operator training. revised procedures,and additional alarrns in the control room 2.
Failure of discharge check 4.
Addition of a second pump discharge valve causing failure of cross-tied ESW pumps check valve 5.
Increased system testing frequency 3.
Failure of ESW to cool the emergency diesel 6.
Addition of check valves in series to the air generators because of failure of air-operated valves operated valves 7.
Addition of a swing, self-cooled diesel generator Table 2.9 Value/ Impact Analysis for Indisidual Modification (Plant 11-3)
(Internal and External Esents CDF)
Value/ Impact
.$CDF Cost Ratio Modification Per RY S/P REM 1
9.4E-6 12,000K 37K 2
1.2 E-5 150K 0.38K 3
0.0E4 0 NA NA 4
3.9E-6 1,200K 9K 5
2.9E-6 35K 3K 6
2.0E-6 1,800K 27K 7
1.1E-5 21,000K.
59K Table 2.10 ESW System Contribution to Core Damage Frequency (Plant 113)
Plant CDF ESW System Percent of PER Contribution ESW System Reactor year to CDF Contribution Internal Events 4.5E-06 1.4E-06 31 External Events 9.7E-05 3.7E-05 38.
Total 1.0E-04 3.SE-05 38 l
13 NUREG-1461 l
.J
The value/ impact results for Plant 11-2 were taken directly discount rate and remaming plant life The results of the from the USI A 45 study (Ref.26).'lhe uncertaintyanaly-sensitivity analysis are presented in detail in Reference sis was presented in Section S of NUREGiCR-4767 (Ref.
- 10. For Modification 2 the resultsare givenin Table 2.11.
26). For Plant 15-3, the contribution of the ESW system to The central value shown in this table is the value calcu-the risk of core damage for each alternative modification lated by the scoping study, the low and high values are was analyzed by using the NURI:G/CR-4550 (Ref. 28) cMeulated by vaning those parameters and assumptions sequences as appropnate. 'lhe costs for engineering, in-indicated above usmp an assigned factor. For instance, a rtallation, operations and maintenance for implementing factor of 2 above and below the central values was as-the modtficanons were estimated by using the costs esti-signed for the risk reduction, which represents the uncer-mated for the USl A-45 study and then multiplied by the tamtics in the PRA. As can be seen from Tab!c 2.11, the consumer price index for the years 1985 to 1992 to ac-value/ impact ratio is very sensitive to the PR A unces tain.
count for inflation.'lhe onsite dose (40.000 person. rem) tics and is much less sensitive to other parameters (event for Plant 11-3 was assumed to the same as that used for the consequence and cost). It could vary from $30 per person-similar reactor of the USI-45 studyJlhe methodology for rem (high risk reduction) to $4,000 per person. rem (Iow the Gl.153 value/ impact analysis followed the same risk reduction), with a mean of 5380 per person. rem.'lhe methodo!on used for the USI A-45 study.
results of the sensitivity study indicate that the valuchm-pact analysis for the GI-153 scoping study did not gener-A sensitivity study (Ref.10) for Plant 11-3 was performed ate any significant additional source of uncertainty.
by varying (1) the estimate of initial cost and of recurring Therefore, decision making on determining whether cost (2) the parameters for replacement power, loss of Molfication 2 is favorabic will rely on the acceptance of investment, onsite cleanup. and onsite dosage from an the NUREGICR-4550 PR A and its associated uncertain-accident. G) the calculated nsk reduction. and (4) the ties (Ref. 28), which is one of the NUREG-1150 plants.
Tahir 2.11 Sensitivity of Modification 2 Cost /llenefit Ratio to Variations in Assumptions for Plant 113 (Internal and external esents CDF) (KS per person. rem)
Calculated Risk 1.ow Central liigh Reductions (Note 1)
(1.)
(C)
(II)
Event Consequence Estimates (Note 2) 1.
C 11 1.
C 11 1.
C 11 Cost Estimates (Note 3)
N L
3.6 33 2.e
.a4 3
.22
.027
.02
<.01 C
4.4 40 3.4
.42 3S 3
.036
.03
.013 II 6.0 5.6 4.9
.59
.55
.47
.053
.046
.03 Notes:
1.
Central value = valacs used for the value/ impact analysis for the scoping study; low =.5 x Central; liigh = 2.0 x Central. The rationale for these assigned multipliers was discussed in Reference 10.
2.
I ow =.2 x Central; 11igh = 5.0 x Central.
3, I ow
.8 x Central: High = 1.6 x Central.
N UREG-1461 14
3 ESW-REI ATED REGULATORY ACTIVITIES Ther c arc a number of on-going regulatory activities aim-confirm that the ESW system will perform its in-e ing to improve the ESW system reliability. These activi-tended function in accordance with the plant's li-ties are summarued in the following sections.
censing basis confirm the adequacy of maintenance pmetices, op-e 3.1 SWS System Operational crating and emergency pnadun s, and training Performance Inspection Program The reliability of the ESW system wal be improved w hen these actions are implemented.
The NRC has initiated a Service Water System Opera-tional Performance lnspection (SWSOPI) program to sat-3.3 Individual Plant Examination isfy the following objectives (Ref 22):
I,rogram verify that the SWS system is capable of meeting GL SS-20, Supplement 1 (Ref. 29) tequested each licen-a thermal and hydraulic design requirements see to perform an Individual Plant Examination (IPE) for severe accident vulnerabilities from internal events.
identify and evaluate SWS design vulnerabilities NUREG-1335 (Ref. 30) provides guidance on what each e
IPE should involve, Related io ESW concerns, licensees are to consider as part of their IPE the following:
e asses SWS operation, maintenance, surveillance, testing, and associated personnel training Interfaces and dependencies of front-line systems with the supports systems assess ihe unavailabdity of the SWS that is the result e
of planned maintenance, surveillance, and compo.
Generic and plant-specific initiating events nent fadures Equipment unavailability because of test and e
maintenance assess the licensce's planncd or completed actions in e
Common-cause failures response to GL 89-13 Iluman failures, both in maintenance and operation The NRL, staff has completed pilot ESW system mspec-and failures to recover and mitigate tions at St. Lucie, Gmna, Quad Cities, and South Texas. A common findmg from these inspections is that licensees Internal flooding have not been successfulin confirming the existing licens-As-built, as-operated system configurations ing basis for their ESW systems. Also, potential single failures that could prevent the ESW system from per-forming its safety function have been identified (R ef. 22).
In addition, GL 88-20, Supplement 4 (Ref. 31) requires ne NRC staff plans to conduct the SWSOPI program at each licensee to perform an Individual Plant Exammauon o xternaments@M)of theirplant.Risexaminat-most sites to ensure that issues pertaining to this system are adequately resolved.
i n should identify severe accident vulnerabihties result-ing from seismic events, internal fires, high winds and tornadoes, external floods and t ransportation and nearby I"'"I'Y "'dd'"
3.2 Implementation of Generic Letter 89-13 Requirements ne nSw vulnerabilities and associated risks are difficult to extract from a PRA if the specific ESW system insights Generic Letter 89-13 requires each licensee to take the are not taken into consideration.Therefore, the insights following actions-from the GI-153 study are prosided in Section 4.3 to serve as guidar.ce for any ESW risk assessment.
o implement surveillance and control techniques 3.4 ESW-Related Generic Issues e
conduct heat transfer testing of ESW cooled heat 3.4.1 Genen.c Issue 23, " Reactor Coolant exchangers Pump Seal Failures" implement a routine inspection and rnaintenance GI-23 identified loss of coo!ing water to the reactor cool.
o program ant pump (RCP) seals as one of the mechanisms causing 15 NURI G-1461
failur e of 1he scals and subsequent seal LOCA following a A review of icing operating experience was conducted and station blackout accident. The proposed GI-23 resolution only one LWR operating event was identified.1his event is currently applied to PWRs only. For IlWRs, the nsk (PWR) was caused by icing on traveling screens for the associated with seal LOCA was considered insigmficant senice water intake in 1976 befor e comrnercial operation at the time the proposed resolution was issued for pubhc (Table 2.3). No similar icing events were observed to comments; but this position has been under re-evaluation occur after commercial operation. Although this survey in response to the public comments.The cooling water to indicates that blockage of water supply caused byicingis a the RCP seals for PWRs is normally provided by the rare event, the consequence could be safety signific;mt.
Component Cooling Water (CCW) system which is typi-Therefore, the effects of icing on the reliability of the cally supported by the ESW system for most plants.
ESW system is identified as one of the potential risk 1herefore, resolution of GI-23 may has e direct impact on contnbutor to ESW system vulnerabilities (see Section the assessment of the contribution of the ESW system to 4.3).
the risk of core damage for PWRs. It should be noted that the risk assessment from the GI-153 study reviewed only previous PRAs in the areas related to the ESW system 3D MUIIIICII3IICO NIIl0 and did not spectfically review seal LOCA.1hese i' ras may not have considered seal LOCA, or they may have con'sidered seal LOCA but used a differe'nt leakage lhe NRC is issuing for public comment a draft Regula-model from what was used by GI-23.
tory Guide DG-1020, " Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," to endorse an indastry guidance document (NUM ARC 93-01, Revision 3.4.2 Generie issue 51, " Improving tlie 2A dated July 9,1992), Industry Guideline for Monitor-lleliabil.ity of Open L,.ye:e S,erv. ice-Water ing the Effectiveness of Maintenance at Nuclear Power System,
Phmt, to implement the maintenance rule stated in 10 CFR Part 50.65.The maintenance rule requires commer-GI-51 deals with senice water system fa ling.1hc preb.
lems of biofouline are considered to be resowed by imple.
cial nuclear power plant licensees to monitor the effec-meming the bas'eline fouling program, as required by tiveness of maintenance activitics for safety-sigmficant Generic Letter W 13 (Ref. 5 k 1herefore, biofouling-re, plant equipment to minimize the likelihood of failures lated problems were not meluded in GI-153 scope of and events caused by the lack of effective maintenance, The NRC staff believes that NUMARC 93-01 provides work.
methods acceptable for complying with the provisions of
'h" * " **'"""'" " I"'
3.4.3 Generic Issue 130, " Essential Senice Water Pump Failures at Multi-Unit Sites" Effective maintenance is important to operability, avail-ability, and reliability of safety equipment. It is especially GI-130 is hmited to seven PWR sites of multi-unit con-important to the ESW system because the system involves figurations with two ESW pumps per plant. This generic a raw water supply. As indicated earlier, review of operat-issue is considered resolved by issuance of GL91-13 (Ref.
ing experience showed that deposition of sediment, 32).
biofouling, corrosion and crosion, and intrusion of foreign material and debris has caused failure and degradation of 3.4.4 Generic Issue Ib32. " Ice Effects on the system. Effective maintenance is expected to mitigate Safety-Itelated Water Supplies" these causes and improve the ESW system reliability.
Although GL69-13 and the SWSOPI program have pte-This issuc deals with the concern of the potential effects viously emphasized the need for effective maintenance, of extreme cold weather resulting in ice buildup on vari.
implementation of the maintenance rule by following the ous water supplies. This issue was evaluated and deter-N UM ARC 93-01 guidance would further strengthen this mmed to be subsumed in GI-153.
important area.
NUREG-1461 16
4 INSIGilTS 1
The insights gained from the woping smJy d:wused in 4,2 ])onijnant Sequences and Ollier Section 2 are summarued in the lonowing sections:
gractors Influencing tiie ESW System Reliability Dominant sequences with regard to total CDF due 4.} (,Ont rib ut.lOn of. ille ESW Systent to all causes are distinctly different in PWR and to tlie Risk or core namage nwR piants. wnen considering the contribution of t he ESW syst em to the risk of core damage at PWRs, e
The contribution of the ESW system to the risk of wqamn inmhing small and medium i OCA tend cor e damage is sienificant for so}ne BWRs as we!! as to be dominant. Ilowever, for if% Rs, the contribu.
for some PWRs (preater than about 10E-5 per rca tion of the ESW system to the nsk of core damage is tor ) car) predominant in sequences invohing loss of an at bus as an initiator or loss of of fsite power.
e ESW system dominant failure modes found e
Re ESW system appears to be more important for HWRs than for PWRs in terms of the contnbution from the review of the NRCsponsored PR As tend of the ESW sTstem to inter nal events CDF(1.41M%
!".9 mme mmmon asps in Mmnt plann to 2.7E-04 p'er reactor-year for llWR plants com-gee Iab!c 2.7). even though the system configura-pared uith 1.$E-08 to 6.7E-05 per reactor-y ear for Nn for cach p ni s umque. I he two myommon ES system faults were the dependency of the ESW PWR phmisk system on motor-operated isolation vahes to open on demand to supply coohng water to safety-related e
The coninhution of the ESW system to the risk of loads and failure of the standby senice water pumps core damage is higher for older p! ants (operating to start. Two other subtle failure modes are (1) fail-date before 19M) than that f or newer plants (aver.
m to isolate nonessential cooling water loads and arcs as 38 percent for older HWRs,1S percent for (2) fade of noss-tied pumps as a result of back newer HWRs and 12 percent for older PWRs. 7 flow from a failed pump discharpe check valve.
percent for newer PWRs).
hv could have a sicmficant effect on the contribution of the ESW system to CDF, as was also e
The contribution of the ESW system to the nsk of concluded in the EPRI-sponsored report (Ref.8).
core damage appears to be plant spee:fic.
Hus, whee watn quahty is pmt, plant-specific equipment failure data should be used in bcu of generic data.
External events could ufect substantially the total e
contribution of the ESW system to the risk of core 4.3 ESW-Related Risk Data froni PRA damage. For the pilot plant, external es ents in-creased this contribution more than an or der of mar.
Reports nitude (from 1.4E-06 to 3EE-05 per reactor-year).
De insights from extracting the ESW-related risk data 110 wever, the impact of external events CDF could from previous NRC-sponsored PRAs and reviewing the vary substantially from plant to plant-IPE submittals result in the following list which provides the areas important to determining ESW c3 stem reliabil-ity:
Inss of the ESW system was not considered as an e
initiating event in the NRC-sponsored PRAs (Ref.
interfaces and dependencies of front-line systems on e
9). Should loss of this system be included in the the ESW system and the risk contribution resultmg PRAs, the estimated contribution of the ESW sys-from these system interfaces and dependencies tem to the risk of core damage is expected to in-cr ease.
e initiating events (internal and external) including appropnate ESW system failures, such as icing or blockare of an intake screen, and the risk contnhu-The value/ impact analysis for the pilot plant demon-tion re$ultinn from these imtating events e
strat es that cost-effectise measures are possible.
even for a plant with a relatively low internal events dominant sequences that uwlve failures or depra-e CDR dation of the ESW system 17 N UREG-1461 i
a fault tree model for the ESW system. 'lhe model ESW train / subsystem unavailability because of e
maintenance while at power should address vulnerabilities of the ESW system to single failures and should consider subtle failures, such as bypass flow through an idle pump and failed the ef fect of water quality on the liSW system com-check valves in cross-tied pump discharge headers.
e ponents and plant CDF plant-specific or generic data used for ESW compo-e e
operator errors, required manual operations, mamtenance, and recovery actions involving the the capability of the methodology to identify vul-e liSW system. In particular, the need to isolate non-nerahdities associated with loss of the ESW system essential cooling loads during accident conditions should be considered.
uncertainties and sensitivities of the PR A NUREG-1461 18 l
l
i 1-1 i
5 RECOMMENDATION i
The staff recommends that the msights from GI-153 re.
With due regard for the insights from GI-153, the ESW garding the contnbution of the liSW system to the risk of system operability, availabihty and reliabihty can be im-l
(
cor e damage serve as a complement to the SWSOPI pro-proved significantly by implementing the SWSOPI pm-j gram to assist in such areas as determining inspection gram, Generic Letter 89-13 requiremems, the IPE pro-i priorities. Other GI-153 information such as dominant gram, the GI-23 rule-making, the maintenance rule, and j
sequences, ESW failure modes, and the effects of water the industry sponsored EPRI research program results.
quahty could also assist the program in identifymg and herefore, the staff concludes that ESW system reliabil-cvaluating ESW vulnerabilities. In addition we note that ity is being adequately addressed by the on-going regula-i the plant-specific PRA developed under the Individual tory and industry initiatives. As a result, the objective of l
Plant Examination (IPli) program is also a meaninpful GI-153 is achieved and the issue should be considered complement to the SWSOP1 program to achieve one of "RESOINED." The need for future action (s) on ESW its objective, i.e., to identify and evaluate the liSW system system reliability is expect ed to be determined from these vulnerahihties.
on-going programs.
j i
f t
i i
i I
l l
l
)
i l
I I
i i
i i
19 NUREG-1461
--.~.--.- - -..--.-
i i
6 REFERENCE 1.
U.S. Nuclear Regulatory Commission, NUREG/
Policy Statement. Federal Register, Vol. 51, p.
CR-5526," Analysis of Risk Reduction Measures 26044, August 4,1986.
l Applied to Shared Essential Senice Water Systems at Multi-Unit Sites," Brookhaven National labo-13.
U. S. Nuclear Regulatory Commission, Memortm-ratory, June 1991.
dum from J. Heltemes to R. Bernero et. al.,*Regu-i latory Analysis Guidelines," July 10,1992.
j 2.
U.S. Nuclear Regulatory Commission, NUREG/
CR-2797,"Evalu-ion of Events Involving Senice 14.
U. S. Nuclear Regulatory Cnmmission, NUREG/
Water System in ~.aclear Power Plant," Oak Ridge CR-2497, ORN1JNS1C-182. Volume 1, "Precur-National laboratory, November 1982.
sors to Potential Severe Core Damage Accidents:
l 1969-1979, A Status Report," Oak Ridge National 3.
IE Bulletin 80-24, U.S. Nuclear Regulatory Com-12tboratory, June 1982.
mission November 21,1980.
15.
U. S. Nuclear Regulatory Commission, NUREG/
4, IE Hulletin 81-03, U.S. Nuclear Regulatory Com-CR-3591, " Precursors to Potential Severe Core mission, April 10,1981.
Damage Accidents: 19S0-1981, A Status Report,"
l Oak Ridge National I;iboratory, Volumes 1 and 2, -
5.
Generic lxtter 89-13. "Senice Water System July 1984.
Problems Affecting Safety Related Equipment,
U.S. Nuclear Regulatory Commission, July 18, 16.
U. S. Nuclear Regulatory Commission, NUREG/
1989.
CR-4674, " Precursors to Potential Severe Core Damage Accidents: 19S4, A Status Report," Oak s
6.
NUREG-0933, "A Prioritization of Generic Safety Ridge National laboratory, Volumes 3 and 4. May 1ssues," U.S.
Nuclear Regulatory Commission, 1987.
June 1989.
17.
U. S. Nuclear Regulatory Commission, NUREG/
7.
U.S.
Nuclear Regulatory Commission, CR-4674, " Precursors to Potential Severe Core NUREG-1275,"Operatmg Experience Feedback Damage Accidents: 1985, A Status Report,' Oak Repon-Senice Water System Failure and Degra-Ridge National Laboratory, Volumes 1 and 2, De-i dations," Vol. 3, November 1988.
. cember 1956.
j i
3 8.
Electric Power Research Institute and Nuclear 18.
U. S. Nuclear Regulatory Commission, NUREG/
Safety Analysis Center, NSAC-148, " Service CR-4674, " Precursors to Potential Severe Core I
Water Systems and Nuclear Plant Safet,," pre-Damage Accidents: 1986, A Status Report." Oak l
1 pared by Pickard, Iowe and Garrick, Inc., May Ridge National I aboratory, Volumes 5 and 6, May i
1990.
19S8.
9.
U. S. Nuclear Regulatory Commission, NUREG/
19.
U. S. Nuclear Regulatory Commission, NUREGI CR-5910. "Inss of Essential Service Water in CR-4674, " Precursors to Potential Severe Core LWRs (GI-153) Scoping Study," Sandia National Damage Accidents: 1987, A Status Report," Oak i
I;tboratory, August 1992.
Ridge National laboratory, Volumes 7 and 8, July 1989.
10.
Science and Engineering Associates, Inc., SEASF-
~
i LR-92-013, Rev.1, " Supplemental Study of Ge-20.
U. S. Nuclear Regulatory Commission, NUREG/
l neric issue 153, l. ass of Essential Senice Water in CR-4674, " Precursors to Potential Severe Core LWRs," August 1992.
Damage Accidents: 1988, A Status Report," Oak i
Ridge National 12iboratory, Volumes 9 and 10, i
11.
U. S. Nuclear Regulatory Commission, Memoran-February 1990.
I dum from E. l_ Jordan to E. S. Beckjord, "Imple-mentation of 1he Safety Goals," September 6,1990.
21.
U. S. Nuclear Regulatory Commission NUREG/
CR-4674. " Precursors to Potential Severe Core Damage Accidents: 1989, A Status Report," Oak j
' 12.
U. S. Nuclear llegulatory Commission, " Safety Ridge National laboratory, Volumes 11 and 12, Goals for t he 0;cration of Nuclear Power Plants "
August 1990.
1 21 NUREG-1461
i l
l
[
- 22. SECY-92-355," Implementing Senice Water Sys-28.
U. S. Nuc! car Regulatory Commission. NUREGI l
tem Operational Performance Inspection CR-4550, " Analysis of Core Damage Frequency:
1 (SWSOPIs), October 20.1992.
Peach llottom. Unit 2, Internal Events," Sandia National I aboratory, Vol. 4. Rev.1, August 1989.
23.
U. S. Nuclear Regulatory Commission, Informa-tion Notice 92-49: R ecent less or Severe Degrada-29.
Generic Letter 88-20, Supplement 1 " Individual tion of Senice Water Systems, July 2,1992.
Plant Examination for Severe Accident Vul-nerabilities-10 Cl R 50.54(f)" U.S. Nuclear 24.
U. S. Nuclear Regulatory Commission, NUREG/
Regulatory Commission, August 29,1989.
CR-5643," Insights Gained from Aging Research,-
Ilrookhaven National laboratory, March 1991.
30.
U.S.
Nuctcar Regulatory.
Commission, 25.
U. S. Nuclear Regulatory Commission, NUREG/
NUREG-1335, " Individual Plant Examination:
CR-5379, Vol. 2 " Nuclear Plant Sen ice Water Sys-Submittal Guidance," July 19S9.
tem Aging Degradatior. Assessment," Pacific
~
Northwest Laboratory, October 1992.
31.
Generic Letter 88-20, Supplement 4 " Individual Plant Examination of External Events (IPEEE)for
.6.
U. S. Nuclear Regulatory Commission, NUREG/
Severe Accident Vulnerabilities-10 CFR
( R4767, Shutdown Decay Heat Removal Analy-50.54(f)," U.S. Nuclear Regulatory Commission, sis of a General Electne 15% R4/ Mark I, Sandia
""' *' I99 I' National laboratory, J uly 1987.
27.
U.S.
Nuclear Regulatory Commission, 32.
Generic 1 etter 91-13. " Request for Information NUREG-1150, "Scvere Accident Risks: An As.
Related to the Resolution of Generic issue 130,10 sessment of Five U.S. Nuclear Power Plants,"Vol-CFR 50.54(f)," U.S. Nuc! car Regulatory Commis-umes 1 and 2, December 1990.
sion, September 19,1991.
NUREG-1461 22
NGC rORM 3%
U r,. NUCLEAR REGULATORY COMM;S$10N
- 1. REPORT NUMBER G-89)
( Assigned t?y NRC, /sdd Vol..
NRCM t 102.
Supp., Rev.. and Addendsm Num-
%21. N'1' BlBlJOGRAPHIC DATA SHEET t* 5 a"v - )
isee insuuctons on tne reverse)
- 2. TH ti AND Lui.,TM LL
- 3. DATE Rt' PORT INULISHED Regulatory Analysis for the Resolution of Generic Issue 153:
yoy7g l
ycyn less of Essential Senice Water in LWRs i
August 1993
- b. A u ; e sun i:n
- b. T YPE OF RE PORT T..M. Su
- 7. etR:OD covtRED tinciusive Dates)
- 8. PLhF OHMNG OhGANsZ AllON - NAME AND ADDhESS ut NhC. gov 6 Divmon. Othee or Reg >on. u.S fJucicar AcQJ;alOry COramisson, ard matling addtess; (f Contractor, POvide narne and mahing add +ess.)
Division of Safety Issue Resolution Office of Nuclea'r Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
- 9. SPON50HNG UhGAN;Z ATiON - NAME AND ADDHE SS (It NRC. type "E ame as asove' ; if ccatractor, provroe NRC Diviseon, Omce or Region, U. S. Nacistar Reptatory Corrmssion, and rnailarig modress. I Same as above a sum tu nT Am Nem
- 11. AE STRACT GD3 wates or less) in this report, the staff of the U.S. Nut! car Regulatory Commission provides a repulatory analysis for the proposed reso-lution of Generic Issue 153 (GI-153). " Loss of Essential Sen ice Water in LWRs " GI-153 deals with the concerns per-taining to the reliability of essential senice water (liSW) system and related problems for all light water reactors except the seven multi-unit sites addressed by GI-130. " Essential Sen-ice Water Pump failures at Multi-Unit Sites " On the basis of the technical findings of a scopmg study for GI-153, the staff recommends that the insights pained from the study sen c as a complement to the on-going ESW performance inspection program. The staff also concludes that ESW system reliability is being addressed by various on-poing regulatory programs. Therefore, the staff recommends that GI-153 should be considered " RESOLVED." ne need for future action (s) on ESW reliability is expected to be determined froni these on-poing programs.
- 12. FJY WORDS/DrScPIPTORS (List woras or phrases that will assist researche s 6n locaterig the reprt.)
- 13. AVAILADILITY STAT EMENT Unlimited
- 14. SECURITY CLAS$1FICATION Regulatory Analysis for G,eneric Issue 153 LWRs Essential service water (ESW)
Unclassified Ghis Report)
Unclassified
- 15. NUh&LR OF PAGES
- 16. F4tfCE NROFORM335 (2-49)
m Printed on recycled paper eceral Recycling Program r
153:
'AUGUdl 1993 l
LOSS OF ESSENTIAL SERVICE WATER IN LWRS I
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