ML20077M040

NRC Research Program on Plant Aging: Listing and Summaries of Reports Issued Through June 1991

ML20077M040
Person / Time
Issue date:	07/31/1991
From:	Eva Hill, Kondic N NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
References
NUREG-1377, NUREG-1377-R02, NUREG-1377-R2, NUDOCS 9108130174
Download: ML20077M040 (92)

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NUR13G-1377 Rev.2 NRC Researca Program on P: ant Aging: Lis~;ing anc Summaries of Reaorts Issuec T:arouga June :L991.

U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research N. N. Kondic, E. L llill

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O FARS 18AM 7 2 1377 R PDR

uns AVAILADlLITY NOTICE Availability of Reference Matorials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

1.

The NRC Pubhc Document Room 2120 L Stroot, NW, Lower Lovel, Washington, DC 20555 2.

The Superintendent of Documents, U.S. Government Printing Offico, P.O. Box 37082, Washington, DC 20013 7082 3.

The National TechnicalInformation Servico, Springfield VA 22101 Although the listing that follows represents the majority of documents citod in NRC publica-tions, it is not intended to be exhaustivo.

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NUREG-1377 Rev.2 NRC Research Program on Plant Aging: Listing and Summaries of Reports Issued Through June 1991 Manuscript Completed: May 1991 Date Published: July 1991 N. N. Kondic, !!. L liill' Division of Engineering Omce of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Wcshington, DC 20555 f a%,

(M),,

SProgram Management, Policy Development, and Analysis Staff

j AllSTRACT

'ihe U.S. Nuclear llegulatory Commission is conducting the Nucitar Plant Aging ltescarch (NPAll)Propam.'this is a comptchensive hatdwatc orsented engineering researth preparn focused on understanJing the aging mechanisms of corn;mnents and systems in nut! car }va er plantt 'the NPAR program also focuses on methoJs for t.imulating and monitoring the aging related degradation of these cornponents and systems, in addition,it provides secorn-rnendations for (flective maintenance to manage ag.tng and for the implementation of the research results in the regulatory process.

'Ihis document contains a listing end index of tcports rencrated in the NPAll Prorrain that were issued throurh June 1991 and summaries of those reivits.1:ath sununary desetil< s the elements of the rescatch coveted in the toport and outlines the significant results. l'or the convenience of the user, the reports at e indexed by personal author, corporate author, and subject.

ni NURl:0-1377

Contents Page gjj AD$traCl..............................................,.......,,,,,,,,,,,,,,,,,,,,,,,,,,,

y;j Preface ix Ack nowl edgem e nt........................................................................

xi Act onyms and Abbreviations................................................................

]

i n t rod u Cl io n...........................................................................

3 Main Citations and Summaries 55 Permnal Author Index....................................................................

65 Comorat e Au t h or I n d ex....................................................................

69 Subject index............................................................................

77 Ch r onological IJ sting.......................................................,,,,,,,,,,,,,,

l l

. NURilG-1377 v

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4 4~ _ _..., _,___,, _

PREFACE

'lhe Office of Nuclear Regulatory llesearch of the U.S.

ating system. level aging effects based on operating Nuclear Regulatory Commission (NRC)is conducting experience and nsk evaluation of the aging phenomena a hardware-oriented engineering research program has been largely completed. Significant accomplish.

dealing with the aging of nuclear power plant compo-ments have included identifying major technical r.afety nents and systems. ~1his program is der.cribed in issues and defining the risk significance of major light NUREG-1144, Rev,1. " Nuclear Plant Aging Re.

water reactor components and structures for applica.

search (NPAR) Program Plan," pubhshed in Septem-tion to plant life extension and license renewal.

ber 1987.

'Ihindocumentcontains summariesof NRC sponsored reports that were generated in the NPAR Program.

Sigmficant progress has been made in defining aging Each summary descnbes thc objectives of the research, degradation mechanisms and in developing effective identifies the contractor and the authors involved, and monitoring and surveillance methods for many of the outlines significant research results, if the readers of components and systems identified in NUREO-1144, this document need additional information on a par-Rev.1,These components and systems include motor-ticular report and the findings discussed therein, they operated valves, check valves, solenoid-operated are encouraged to contact the authors of that teport valves, electric motors, emergency diesel generators, directly.

chargers and inverters, circuit breakers and relays, bat-teries, auxiliary fcedwater pumps, and reactor protec-This report is updated annually to incorporate surnma-tion systems. Progress has also been made in develop-rics of new NPAR reports. Comments are welcorne ing models and approaches to evaluate the relative and will be considered in developing subsequent revi-impact of aging on risk.The Phase I research for evalu-

. ions of this document.

WA h ilton Vhs, Chief Electncal & Mechanical Engineering Ilranch Division of Engineering Office of Nuclear Regulatory Research vii NURihl377

ACKNOWLEDGEMENT The following valuable contributions to the preparation of this document are appreciated:

Members of IWS 1?lectrical & Mechanical linginecting tiranth,in particular, M. Vagins f or suggestions and advice and J. Vora with the other Nuclear Plant Aging 1(esearch program managers for their timely inputs.

Ina Schwartz of the lilt:ctronic Composition Services Section for expeditious and high-qual-ity proccasing of the text and the IlliS/l)l1 f.ccretarial sttdf, in particular, I!. I taw kins C. Itinn, and J. Williams for their effort on the manuscript.

NUlt!!G-1377 gg

i ACitONYMS AND AllilitEVIATIONS ISI inservice inspection AFW auxiliary feedwater Al.l!AP Aging and life lixtension Assessrnent KWU Kraftwerk Union Aktiengesellschaft, a German company Program ANSI American National Standards Institute 11!!(

l.icensec !! vent Report AShill American Society of hiet haniud lingi-LOCA loss-of-coolant accident D'"5 1.W R light water reactor AUXIP auxiliary feedwater pump hiCC motor cantrol center llNL llrookhaven Nationallaboratory h1CSA motor current signature analysis llOP balance of plant hilj!

maximum hkchhood estimate ll&W llabcock & Wilcox Cc..,

hiON, motor-operated valve llWR boiling water reactor me r opnald vake analym. amj teu CCW component cooling water rystem Cli Combusiton 11ngineering hist.ll main stear. line break CRD control rod drive NOK Ma Operations Analysis Center CVN Charpy V notch NPAR Nuclear Plant Aging itescarch Dill!

design basis event

1. P""

liCCAD electrical circuit characteritation and NPRDS Nuclear Plant Itcliability Data hystem diagnostic systern N1(C U.S. Nuclear llegulatoiy Commission 11 CS emergency wre cooling system NSAC Nuclear Safety Analysis Center 110 0 cmergency dietel generator NSSS nuelcar steam supply system I!PA clectrical penetration assembly POltV power-operated actief valve

!!PRI lilectric Powtr itescarch Institute PltA probabilistic risk assessment liSFAS engineered safety feature actuation system PWR pressurized water reactor FSAR l'inal Safety Analysis Report RCP reactor coolant pump Gl'RS generie flaw response spectra illlR residual heat removal 01 generic issue ICID resistance temperature dectector iIDR Ileissdampfreaktor, a decommissioned ICIS reactor trip system German reactor reactor watn cleanup llPCI high pressure coolant injection SCR tilicon controlled rectifier llPIS

! igh-pressure injection systern SOV solenoid-operated valve 11!!!!!

Institut e of lilectrical and I!!cetronics SQUG Seismic Gualification Utilities Group lingineers SSli safe shutdown earthquake INiil.

Idaho National linginecrmg I aboratory INPO Institute of Nucicar Power Operations TDR time-domain reficetometry Tl temporary instruction IPRDS In Piant Reliabdity Data Systems ISA Instrument Society of America TIRG Al.liX Technical integration Review Group for Aging and 1.ife !!xtension ischi inspection, suricillance, and condition momtoring UT ultrasonic testing l

N U RI.0 -1377 xi

l l

I.

INTRODUCTION tihable, reconunendations are made for improve-

'this document is a itsting and index of reports re-la;cd to the Nuclear Plant Aging itescarch (NPAlt) rnents.

Program issued through June 1991.'the first listing is

'the information contained in the reporis should be in alphanutaene order by r eport number and includes a of interest to those assessmg the aging and icliabihty of summary of each report'ltree indexes are provided to nuclear power plant components, including research-aid the user in tetrieving a specific report: Personal ers and designers as well as maintenance and opera-Author Index, Corporate Author index, anti Subject tions personnel.

Index. Finally, thes e is a listing in chronological order by date of publication.

Most of the documents ettedin this eport are avail-able from one of the following sources:

Most of the reports contain a descriptica of the 1.

'1he NitC Public thrument floom, 21201.

components or systems being examined and identify Street NW., lower !xvel Washington, DC.

the prir.cipal stressors leadtng to aging. '! hey fre-2.

'ihe Superintendent of Ihicuments, U.S. Gov-quently contain an analysis and statistical assessment ernment Printing Office, Post (Mfice llox of failure data obtained from Licensee I vent Iteports 37082, Washington, DC 20013-7082.

and other tources of component failure data for oper.

'the National Technical Information Senice, 3.

ating nuclear power plants. Current surveillance and Springfield, VA 221(31.

monitoring practices are also reviewed and, when iden-NUlt11G-1377 1

MAIN CITATIONS AND SUMh1 ARIES 1he reports listed in this compilation are aannged alphanumerically by report number, with unnum-bered repons preceding the numbered reports.'the bibliographic information is followed by a summary of each report.

ries of 14 repons are presented in this publication.

UNNUMilEltrD IEPOllTS

'thus the tesuits of these studies are made rnore readily 14tter Report, M. Subadhi, " Review of Aging Scismic available for rapid survey, directing attention to spe-Cctrelation Studies on Nuclear plant liquipment,"

cific r epons of interest and facilitating the utilliation of Ilrookhavcn National laboratory, January 1985-research results in the reg 6 tory process.

During the last decaJc, the tssue relatir$g to agir'g-

"the 14 reports are grouped into three catsgories:

seismic correlation of nuclear grade equipment and (1) early scoping and backgrotmd studies, including a their components has received special attention by suc/cy of aged power plant facdities, operating experi-both the NRC and the utility industr) with the aim of ence reviews of IJcensee Event Reports (IJIRs) to I.reventing catastrophic failures of aged nuclear pown identify aging trends, workshops to obtain experts' plant components during a seismic event. *1his report opinions, and aginga.sk considerations; (2) reports on summarizes the work performed by the Seismic Quali-developing a methodology fo; aning analysis and on fication Utilities Group (SQUG) based on real carth-evaluation and use of a signature analysis technique quake data, by NinT Cil for Sandia Nationallabora-(MOVATS); and (3) Phase I results of aging research tories, and by EPRI at Wyle, liased on the above, an on nine components, including electric motors, be.ttery outline of the work to be carried out at llNL under the chargers / inverters, electrical cables, pressure trans-NPAR scope relating to identifying the aged compo-mitters, diesel generators, motor operated valves, nents sensitive to seismic loadings is provided.

check valves, auxtliary feedwater pumps, and snubbers.

11ach summary has four sections: llackground, Sum-EQH, Inc., sponsored by the Seismic Qualification mary, Results/ Findings, and Utilization of Research Utilities Group has gathered a comparative data base Results in the Reg alatory Process.

on the performance of equipment in five fossil. fueled This report is considered a "living" document. *lhat plants consisting of 24 units and a high-voltage DC-lo-is, research results and summaries of additional se-AC c<mverter station. 'these plants have experienced four damaging California carthquakesof Richter Mag-tected reports may be added periodically, nitudes 5.1 to 6.6. Peak horizontal ground accelera-adon Rnlew Gr for Aging and IJfe Tnhnical lntg(111tGA1.EX),, Plan or Intef, ration of tions(PG A)of these carthquakestangedbetween0.2 g Extension Ag ng and life.lixtension Activities," (.S. Nuclear and 0.5 g..lhe actual carthquake-induced effects on equipment were compared with equipment qualifica.

Regulatory Commission, May 1987, tion data from three nuclear plants.

.!he TechnicalIntegration Review Group for Aging The objective of the pilot program was to deter-and life lixtension (TIRG A1.EX) was established to mine the feasibility of establishing criteria for assessin8 facilitate the plaaning and integration of NRC actwi-the seismic adequacy of equipment in nuclear power ties related to reactor aging and life extension.'lhe in-plants based on evaluation and application of data to be itial objectives of TIRG Al.EX w cre to identify techni-acquired on the characteristics and seismic perform-cal safety and regulatory policy issues related to reactor ance of equipment in nonnuclear power facihties that aging and life extension and to develop a plan to inte-have been subjected to strong motion earthquakes.

grate NRC and external activities to resolve the issues.

Application of the criteria would provide a valid basis This report contains the plan developed by TIR-for assessing the need for subsequent quahfication ef*

GA11X, which consists of the following main ele-forts in the nuclear industry and for defming the extent ments:

of the effort' L A summary and discussion of the major techni-cal saf ety and i egulatory policy issues associat ed Letter llepor1,l N. Rity, *h,ummaries of Research Re-with reactor aging and life extension.

ports Submitted in Connection with the Nuclear

2. An overview of ongoing programs and activities Plant Acing Research (NpAR) Program," Enginect-related to reactor aging and life extension,in.

ing and liconomics Research, Inc. O!ER). Reston, cluding both NRC and external programs and V A, September,1986, activities,

'the resvits of Phase I efforts in the NRC NPAR

3. Recommendations for future NRC actions to pr ogram for selected electrical and mechanical compo-address reactor aging and hfe extension in a nents since 19S4 have been published.To help main-tain cognizance of this wealth of information, summa-timely, efficient, and well. integrated manner.

NURiiG-1377 3

1 Main Citations and Summaries K.II. Iloopingarner and F.R. Zaloudek.

• Safety implica-NUMllEllED llEPORTS tions of Diesel Generator Agirg " Pacific Northwest I aboratory, Nudcar Safety, 31:484-189, October.

December 1990.

IINI, Technical Repod A-32Sil-26-M. II. Miller, ScopingTest on Contairtmert Pur e and Vent The emergency diesel generators in a nuclear

(",fifec"e t f 8't I

' " llrookhaven ational labora-powet plant have an important safety function in that they supply electric power for emergency core cooling Degradation of shaft scal material used in contain-and related emergency needs in the event of a loss of ment purge and vent butterfly valves may initiate valve i

offsite power. Typically, a plant has two redundant die-seal leakage thus breaching containment. A scopmg se! generators of 3,000 to 8.000 kW (5,000 to 10,000 test was pedormed to gather information on the be-hp)'

havior of the seal material (ethylene propylene)when exposed to severe accident conditions (l.c., stearn at Diesel generators have been identified as compo-350'F/120 psig and 400*F/232 psig). 'three separate nents wim significant safety importance, and their op-test sequences were perfortned with the test assembly crating history has shown perforrnance degradation monitored for leakage.1hc results of these tests re-and loss of reliabdity as a result of wear and aging. Con.

Vcated no seal leakage; however, shaft. seal degrada-sequently, em H gency diesel generators are included in

i. n was evident.

i the NRC NPAR program, the overall objectives of For two test sequences, the prescribed procedure which are to find the ciuses of aging related degrada-was revised to include malified temperature profiles tion and to recommend methods for managing it. Ily and scal testing sequences, ameliorating wear and performance related degrada.

tion, the relbbility of the emergency diesel generators Removal and inspection of thr: valve seat following can be improved and the risks associated with loss-of-some test sequences revealed minor remolding of the offsite-power events reduced.

seat material at the disc / body interface with no de-formities noted. Approximately one week later, cracks

'lhe NPAll Diesel Generator Study consisted of developed in the seat.1hc cracks were in an area that ww e compressed by the retaining ring and in noin-two phases. Phase I used plant operating experience, data, expert opinion, and statistical methods to pro-stance affected the scaling integrity of the valve.

duce a data base related to aging failures, their causes.

'the results of the scoping test revealed no shaft.

and corrective actions. Phase 11 included the develop.

seal leakage. 'ihe seal degradation and cracking was ment of a more appropriate testing and aging manage.

visuauy evident in the compressed retaining portion of ment program that could enhance the amilability and the seat, llowever, the result should not be construed reliability of these dicsci generators.

as representing the enttre ethylene propylene family (clastomers prepared from ethylene and propylene the purpose of this article is to discuss the principal monomers). Varying the relationship of these mono-causesof wear andaging.relateddegradationof emer.

mers affects the characteristics of the clastomer and its gency diesel generators and the effects on their rell, ability to withstand environmental conditions. It ability and availability and to describe rnethods by should also be noted that all mechanisms by which rub-w hich such degradation can be avoided or detected be, ber deteriorates with time are attributable to environ.

fore it becomes a nuclear safety concern. Operational mental conditions. The Parker Seal Company states inforination assembled on component and system fail-that it is enviromnent, not age, that is significant to seal urcs and their causes was reviewed to identify the im.

life, both in storage and in actual service.

portant aging and degradation factors for diesel gen-g g

g cratprs. One important factor contributing to wear and M. Subudhi, J. Ili ins, J, Curreri, M. Reich, degradation has been the fast starting and loading test F. Cifuentes, and Nehring. "Scismic Undurance procedure called for by Regulatory Guide 1.108 and Tests of Naturally Aged Small Electric Motors,"

the NRC Standard Plant Technical Specificationsc A-13rookhaven National laboratory, November 1985.

new regulatory approach was recommended to develop a more balanced aging management program that in-Two naturally aged 10-IIP electric motors were ob-tained from an older nuclear power plant that is ready cludes (1) slow-start t esting during which im portant oP' for decommissioning.The motors were utilized Io drive crating parameters are monitored, (2) analysis of data fan cooler units in an outside environment for 12 years. -

trends, (3) training, and (4) maintenance. 'Ihis ap-These motors were first tested for their dynamic char.

proach should improve safety by enabling the timely aeteristics.'they both were subjected to seismic excita-identification of aging related degradation that could tion with generic floor response spectra (GIWf) that lead to diesel generator failures so that maintenance encompass Safe Shutdown Earthquake (SSE) accel-could be performed in time to prevent actual failure.

crat6ns applicable to most nuclear plants in the NUREG-1377 4

i Main Citations and Summaries i

United States. He tests showed that the first funda-Such research will provide the rnaximum amount of ac-mental frequency is well above thb rigid range of an tual experience data to address the aging-seismic rela-earthquake frequency. Seismic testing was performed tionship in a practical manner. Ixssons learned from a with a motor both unk.aded and loaded by an attached review of these data can be used as input to develop hydraulic pump that served as a dynamometer. Signifi practical maintenance and operating procedures to en-cant operatmp parameters such as current, voltage, hance safety and improve plant reliability, and 'emperatur e were monitored before, dming, and after seismic loading, and no noticeable differences UNL rechnical Report A-3370- 346. A. C. Sugarman, were observed. Existing deficiencies in one of the mo.

M. W. Sheets, und M. Subudhi"l'esting Program for the Monitoring of Degradation in a Contmuous tot bearings and m the stator winding were not affected Duty 460 Volt Class "11",10-HP Electri': Motor,"

or magnified by the seismic excitations.

Hrookhaven National Laboratory, March 1986.

His report describes the test plan, includes details H s report presents an evaluation of potential of the procedure, and presents findings of the seismic mamt-nance techniques for monitoring age related ttsts and operating / static tests on both motors.

Jegradation in a continuous-duty 460-volt, Class B, His testing was part ni the NRC NPAR program, 10-HP electric motor. The program follows up the and its results are an integral part of the Brookhasea analyses and recommendations outlined in the draft Natonal laboratory's overall aging assessment of mo-of NUREG/CR-4156, " Operating thperience and tors, which was published as NUREG/CR-4156.

Aging-Seismic Assessment of Electric Motors," by M. Subudhi et al. In this study, the followig stressors UNL Technleal Report A-3270-12-B5, M. M. Silver, on dielectrics are evaluated: temperature, frequent R. Vasudevan, and M. Subudhi, " Pilot Assessment:

starts, overload, and high voltage gradient.

hapact of Aging on the beismic Perfonnance of Se-in general, the motor tests are conducted by con.

lected Equipment Types," llrookhaven National tinuously reversing motor direction for five hours, fo!-

Laboratory, December 1983, lewed by a half hour with the motor running under no he NRC han initiated a number of specific te-h1ad in a single direction and a half hour with the mot,tr search programs in support of the NPAR program, to turned off and stationary. During the half hour of run-better understand the impact of equipment aging on rting under no 13ad, measurements of bearing vibration plant safety and to secommend realistic operating and and movement of stator end turns (measured with ac-maintenance procedures to improse plant anilability celemmeters epoxicd to the end turns) were made.

and enhance safety.This pilot study was perf ormed to Also, a number of insulation tests were conducted.10 investNate the feasibility of using plant expenwee accelerate the degradation of the tes motor (including

~

data to assess the relationship between equipment ag-insul tion, bearings, and lubrication), a plug reverse mg and seismic performance capacity.

test..as performed.

After a brief review of availaMe Wattnadan on g

gg plant experience at many California Ons for conten!

vealed which insulation and bearing tests can best be and quality, data related t o pci formance maintemmcei used in utilities' procedures for preventive mainte

.nd failure history were collected for a sampie set o, nance, corrective maintenance, and surveillance for equipment types. This pdat study selected the equip-safety-related motors.

ment types for ir.vestigation 'from the highest pnonty group specified in a previous NPAR study.The equip This testing is meant for motors rated for contin, mem types studied were electric motors, motor-uous use. A separate test plan will be required for operated valves, relays, circuit breakers, and motor intermittent duty motors (e.g., valve actuator motors);

control centers.

such a plan should include typical valve actuator tests such as the open/close cycling test and the insulation The acquired equipment data consisted of installa-tests discussed in Section 4.0 of the presently reported tion date, chronological listing of preventive and cor-E*E"*'

rective maintenance activities, failed state and cause of failure, earthquake data (i.e., free. field acceleration, UNL rechnical Report A-3270-1246, IL Fullwood, Richter magnitude, date), and equipment status bef ore J. C. Iliggins, M. Subudhi, and J.1I. Taylor, " Aging and after the earthquake-and Life I!xtension Assessment Program (ALE P)

Systems 1 evel Plan," Brookhaven National labora.

Tne pilot study was successful in deinonstratirg tory, December 1986, that experience data can be extracted and utilized to address the relationship hetween seis.nic performance This system level program plan for ALEAP pre-capacity and aging of plant equipment. It is strongly sents and explains the BNL structured approach to recommended tnat future research be conducted to ac-assessing the effects of the aging of nuclear power I

quire experience data for other important equipment plant components and systems on safe operation and types and to investigate other California power plants.

the extension of plant operation beyond the originally 5

NUREG-1377

Main Citations and Summaries planned plant Itfe. It should be noted that this plan is cas of a system. Recommendations are made for im-prepared in a generic fashion and could be used by any-provements in pertinent regulatory guides, industry one for a system assessment.

standards, etc.This pogram plan delineates the goals De plan discusses the criteria for prioritizing plant, and major tasks to be completed in each phase. De system, and component selection for analysis to deter-current versi n f the program planisconsidered tobe mine the effects of aging.The use of failure modes and a draft and will be revised and updated as the first few effects analysis in conjunction with the results of natu-system assessments are completed using this method-ral and accelerated agirg tests are discussed as means I gy. This will produce a final proven methodology for undustanding and predicting the phenomena. The that can be applied to the remaining systems.

effects of aging on the failure rates of componet.ts are IINL Technical Report A-3270R-2-90, A. Fresco and being determined principally from plant data with M. Subudhi, " Aging Effects of important Balance of physical and phenomenological models used for inter.

Plant Systems in Nuclear Power Plants," Umok-potation in areas not included in the data base. These haven National 12iboratory, February 1990.

results will be integrated with a plant risk model to be used m in recent years, balance of plant (HOP) systems enouga.,addresstng the question "how old ts oldhave become major causes of plant trans:ents, e.g., the June 9,1985 loss-of all feedwater event at the Davis.

De NRC NPAR program has completed several Besse Nuclear l'ower Plant, and have received in-component-level aging assessments that include the creased attention from the nuclear industry and the identification of dominant component failure modes Nuclear Ren,ulatory Commission (NRC). This interim based on pl. nt operating experience.The studies pro-report describes the activities to date in a study of DOP vide recommendations for monitoring as well as miti-systems by B reckhaven Nationallaboratory in support gating age-related component degradations.

of the NRC Nuclear Plant Aging Research (NPAR)

Utilizing results from the component-level studies pr gr m. The initial phase of the study provides pre-and work performed by other NRC contractors for sys-liminary indications of those BOP systems that may tem-data assessment and system-level risk analysis, warrant a detailed study of agmg effects. An approach this program evaluates the impact of component fail-for accomplishing the overall objective of identifytng utes on plant system performance. The study performs the effects of agmg m these DOP systems on nuclear in-depth system-level failure data reviews, evaluates plant safety is sugge sted.

current irtlustry practices for system maintenance, His study on BOP systems covers all non-safety-re-testing, and operation and probabilistie risk assessment lated systeras neept for those associated with the nu.

8 PRA) techniques to understand and to predict the im-clear steam supply system (NSSS). Some non-safety-

. act of aging on system availability. Recommendations related systems.in the NSSS are being studied in other for improving the system performance by means of parts of the NPAR program.

degradation monitormg and timely preventive and cor-From the results of the study,it was concluded that rective maintenance are addressed. This project the frequency of unplanned reactor trips has often integrates its products with the BNL programs for op-been cited as an indicator of safety performance and crational safety reliability research and performance that the most frequent contributors to unplanned reac-mdicators.

tor trips caused by BOP systems are the power conver-The first phase of this research effort concentrates sion systems, i.e., the feedwater, main turbine, main on understanding various system designs from plant eiectric generator, main steam (usually the steam by-safety analysis reports, evaluating failure data from pass to the mam condenser), and condensate systems, plant operating experience data bases, applying PRA Other DOP systems contributing to unplanned reactor analy=es, assessing industry-wide surveillance and trips are support systems such as the electric distribu.

maintenance practices, and identifying system func-tion t,ystem and, less frequently, the circulating water, tionalindicators that are used to monitor the rate of service / instrument air, fire protection, and the heating, system degradation resulting from aging and service ventilation, and air conditioning systems. The electric wear.The program separates failures on demand from dbtribution system includes 120-V AC power distribu-time-dependent failures. It categorizes age related tion systems, the switchyard, large plant load users, the failures separately from random and design-type fail.

DC power system, and control centers. At the compo-utes. It produces results useful for the resolution of nent level, the feedwater regulating valves, the tur-pertinent unresolved safety issues and for review and bine-driven feedwater pumps, and the main turbine inspection of operating NPPs. The second phase, if electro-hydraulic control subsystem are frequent con-authorized and performed, will provide recommenda.

tributors. Failures in the main electric generators are tions for improving system performance through en.

also important as potential causes of reactor trips.

hanced maintenance practices and reliability monitor.

These results are substantially in agreement with ing, which will be focused on the most risk-sensitive ar-the results of an alternative approach in which impor-NUP EG-1377 6

Main Citations and Summaries tant HOP systems were categoriicd based on insights The in-depth performance-based inspections have from probabilistic risk assessments.

explored such areas as overall plant performance re-Preliminary recommendations are:

lated to maimenance, management support of mainte-nance, and the implementation of the maintenance L The frequency of unplanned reactor trips regram. Incorporated in the inspection criteria estab-should be considered the most important indica-

"shed for these general areas are specific attributes tor of current or potential near-term safety that are relevant to the understanding and managing of

problems, aging. The related activities include root cause evalu-

2. IlOP systems thrt significantly contribute to un-ation of equipment failure, trending of failure data,im-planned reactor trips should be included in the plementation of equipment qualification programs, NPAR program, control of spare parts, evaluation of test results (in-cluding postmaintenance testing), and implementation The next phase of the program will focus on several of condition monitoring techniques.

of the oldest nuclear plants. IJcensee Event Reports

'lhe maintenance team inspection reports were re-involving unplanned reactor trips from the beginninE viewed with the following objectives in mind:

of commercial operation to the present will be re-1.

Assess the evaluations of those portions of the viewed to determine if there is an mercasmg frequency maintenance program determined to be im-of unplanned trips caused by the identified IlOP sys-portant for understandmg and managing ag-tems.This group of plants willinclude Monticello and ing.

Yankee Rowe, the pilot plants for the life extension 2.

Es aluate the weaknesses noted in utility main-study.

tenance programs that could affect the ability Next, some plants of intermediate age and then of the plant to manage aging.

some of the youngest plants will be examined in the 3.

Determine the types of preventive mainte-same manner. Ihese three age groups will be com-nance activities and condition-monitoring pared and analyzed.The ultimate goal is to determine techniques that address plant aging.

whether agmg of individual BOP components is a sig-nificant factor affecting nuclear plant safety.

At this time,47 inspections have been completed, and 24 more are scheduled for the remainder of 199(L BNL Technical Report TR4270 90, W. Gunther, NRC has compiled the findings from these inspections

" Maintenance Team Ins; action Results: Insights in a computerized data base that assisted in identifyiag Related to Plant Aging,' Brookhaven National plants where the NRC inspection teams had concen,s laboratory, June 1990.

about how well the maintenance program accounted NRC is performing maintenance team inspections for the effects of aging. Ten reports contained specific in accordance with the NRC temporary instruction (TI) findings that the utility maintenance programs do not 2515/97 entitled " Maintenance Inspection" to deter-address aging. These findings are tabulated and sum-mine the effectiveness of the totalintegrated mainte-matized in this report.

nance process in nuclear power plants. As specified in it should be noted that nine of the above ten reports the TI,"the goal of the inspection effort is to empha-concluded that the overall mamtenance programs w ere size the use of plant experience, test and surveillance adequate, satisfactory, or good. The guidance and cri-data, recent component failures, [and] Probabilistic teria ptovided to the NRC maintenance inspection Risk Assessment (PRA) insights.. " to identify team allow a maintenance program to be judged good if strengths and weaknesses. Two volumes of inspection it effectively manages current problems even though it guidance supplement the T1 and direct the inspectors may not effectively manage the long-term aging effects to evaluate the maintenance /agmg relation. I or exam-on structures, systems, and components.

plc, the inspector is directed to determine the extent to llNL Technical Report TR-3270-9-90, E. Grove and which ma..agement is aware of plant aging.The inspec-W. Gunther,"An Operational Assessment of the tot is also required to evaluate the mvolvernent of cor-Habcock & Wilcox and Combustion Engineering porate management in maintenance activities that ad-Control Rod Drives," llrookhaven National labora-dress plant aging.

tory, September 1990.

More important than the cited guidelines is the ex.

Control rods and the associated drive and control pected assessment of maintenance program activities systerns, which ensure safe and reliable operation, are that reflect on the abihty of the plant technical staff to essential components of nuclear reactors. This report manage aging. Predictive and preventive maintenance describes an aging assessment of the Babcock & Wilcox programs rnust include condition monitoring. trendmg.

(H&W)and Combustion Engineering (CE) control rod and recordkeepinF in order to competently manage the drive (CRD) systems performed as a part of the NRC cffects of aging on a timely basis. Inadequacies found in NPAR program. Emphasis was placed on the specific these areas are indicative of the mabdity of the plant corr.piments of the systems that may be susceptible to staff to properly address and treat aging, aging-related degradation.This study along with a fail-7 NUREG-1377

m....

Main Citations and Summaries ure modes and effects analysis and a detailed review of The report gives an cumpic of an analysis of real utility maintenance practices and procedures will corn-data. In this cumple, the methods applied are unable plete Phase i of the aging assessment of these CRD sys-to discriminate among an exponential hazard function,

tems, a linear hard function, and a Weibull hazard func.

Of the fifteen plants that use CE CRD systems, tion. The MLE for the two parameters appears to have thirteen use a magnetic jack control element drive approximately a bivariate normal distribution under mechanism, and two use a rack and pinsn drive. In all the exponcatial or Weibull hazald model but not under eight B&W plants, the control rods are driven by a the linear hazard model, if the analysis using approxi.

roller nut /leadscrew drive mechanism. All CE reactors mate normality is carned out in any case, the results ap-had basically the same CE logic, control, and rod posi-pear similari r all three m dels,if some modelis pre-tion systems; all B&W reactors had basically the same ferred for theoretical or other reasons, this report indi-B&W logic, control, and rod position systems, des a w y to use it.

Commercially available operating experience data EGG-SSRE-9017. C.L. Atwood " User's Guide to bases were reviewed to identify failed components and PHAZE, a Computer Pro ard Function Estimation,, gram for Parametric Haz.

the resultant effects on plant operation for the Idaho National Engineer.

1980-1989 time period. Age-related failures that re.

ing laboratory, July 1990, sulted in significant plant events, including dropped rods, power reduction, and shutdown for B&W and CE The program PHAZE (for Parametric Hazard control rod drive systems were identified. Susceptibil' Function Estimation) performs statistical inference on sty of the system to such externalinfluences as mainte' a hazard function (also called a failure rate,or intensity nance errors and the operating environment was als function) based c1 reported failure times of compo-shown.

nents that are repaired and restored to service. Three parametric models are allowed: the exponential, lin-ear, and Weibull hazard models. He inference in.

EGG-SSRE-8972, C.L. Atwood, " Estimating Hazard cludes estimation (maximum likelihood estimators and Functions for Repairable Components," Idaho Na-tional Engineering Laboratory, May 1990-confidence regions) of the parameters and of the haz.

ard function itself, testing of such hypotheses as in.

This tutorial report, applying known formulas and creasing failure rate, and checking of the model as-tools in a way suitable for risk assessment, presents a sumptions under a choice of parametn,c models, parametric framework for performing statistical infer-Since the approach's concern is the failure behavior ence on a hazard function based upon such data of re-of components, these failures are assumed to be gov-pairable components as might be obtained from field erned by a Poisson process, with the time typically experience rather than from laboratory tests. His measured from the component's installation. It is fur-framework encompasses many possible forms of the ther assumed that, when a component fails, either it is hazard function, three of which are considered in some immediately repaired and placed back in service or it is detail. The theory is neatest and the asymptotic ap-replaced bya newcomponent. Failures of distinct com-proximations are most successful when the hazard ponents are assumed to be independent.Thus the data function-for a set of identical components-has the to be analyzed consist of sequences of failure times of 1

parametric form of a density in the exponential func-similar independent components.

tion family.The parameters are estimated based on se' quences of failure times when the components are re-This user's guide sketches only enough of the the-stored to service immediately after each failure. In cer-ory to permit PHAZE to be used; a full presentation of tain circumstances, the distribution of the failure the theory is giver, in a companion report. A typical counts does not depend on the parameter that deter-PHAZE application is described. This consists of an in.

mints the shape of the hazard function; this suggests tial exploratory phase, in whi-h the various model as-natural tools for diagnostic checks involvmg the mdt-sumptions are checked, and a final estimation phase, in vidual parameters. ne results presented include for-which the maximum likelihood estimator and a confi-mulas for maximum likelihood estimates (MLEs), tests dence interval are found for the hazard function at and confidence regions, and asymptotic distributions.

times of interest. The format of a data file is given with The confidence regions for the parameters are then examples. PHAZE is an interactive command-based

- translated into a confidence band for the hazard func-program; all the PHAZE commands are therefore 3

- tion. For the three examples considered in detail, a ta-listed and explained.

ble displays all the building blocks needed to program The program has been verified and validated, and the formulas on a computer; this table includes asymp-thisworU ummarized. Finally,someof thetechnical totic approximations when they are necessary to main-details 01 ot.: rest to statisticians and programm6rs are tain numerical accuracy. Diagnostic checks on the given. The appendix shows an erttire PHAZE session, model assumptions are outlined.

including both the user's commands and the program's NUREG-1377 8

Main Citations and Sumrnaries responses. Virtually all of the PllAZE commands are ing was stressed it was suggested that the effects of illustrated, and the resulting output is presented.

time-related degradation on the safety of the complete NUREG-1144, B. M. Morris and J. P. Vora, " Nuclear reactor system should be evaluated in terms of the risk Plant Aging Research (NPAR) Program Plan," U.S.

to the public. One should consider rnultiple causes that Nuclear Regulatory Commission, July 1985.

have typically been associated with abnormal occur-itaces. Since individual component failures create NUREG-1144, J. P. Vora, " Nuclear Plant Agi Research (NPAR) Program Plan," Rev.1, hS.

ptablems, time-related degradation will ultimately Nuclear Regulatory Commission, September 1987, have to be addressed in terms of maintenance, mont-tonng, surveillance, etc., of components.

The Nuclear Plant Aging Research (NPAR) pro-gram described in this plan is intended to resolve tech.

A brge number of phenomena that can cause fail-nical safety issues related to the aging degradation of urcs were discussed; a detailed list of parts / materials, electrical and mechanical components, safety and sup.

including lubricants and other additives that must be pon systems, and civil engineering structures used in consideret.1 was given; seemingly minor changes in the commercial nuclear power plants. ne aging period of chemical ccostituents of a material or in the manufac-interest includes the period covered by the original op-turing proccu can cause significant effects and changes erating license as well as the period of extended plant in the system / luring operation (e.g., water chemistry life that may be requested in utility applications for li-effects),

cense renewals.

Replacement parts were noted as a potential Emphasis has been placed on identifying and char-source of problemoThe effects of storage on parts and acterizing the mechanisms of material and component the possibility that new parts may be different from the degradation during service and utilizing the research original ones were mentioned, results in the regulatory process.The research includes evaluating methods of inspection, surveillance, condi-Here were extensive discussions on the limitations tion monitoring, and maintenance as means of manag.

of accelerated aging tee.s. The use of naturally aged ing and mitigating aging effects that may affect safe equipment for test purpeses was suggested. Sacrificial plant operation. Specifically, the goals of the program replacement of equipmeM was identified as a source are to:

for naturally aged plant equipment.

1. Identify and characterize aging mechanisms and Maintenance and surveihance in plants and their effects that could cause degradation of compo-relationship to time-related cagradation were exten-nents, systems, and civil engineering structures sively discussed.

and,if unchecked, impair plant safety.

NUREG/CP-0100, A. F. Beranek," Proceedings of the

2. Evaluate residual life of components, systems, International Nuclear Power Plart Aging bympo and civil structures and identify methods of in-sium," U.S. Nuclear Regulatory Commission, March spection, surveillance, and monitoring that will 1989.

ensure timely detection of aging effects before his report presents the proceedvigs of the Inter-loss of safety functions.

national Nuclear Power Plant Aging rymposium that

3. Evaluate the effectiveness of storage, mainte, was held at the Hyatt Regency Hotel in Bethesda, nance, repair, and replacement practices in Maryland, on August 30-31 and September 1,1988.

mitigating the rate and extent of degradation

,The Symposium was presented in cooperauon with the caused by aging.

American Nuclear Society, the American Society of Civil Engineers, the American Society of Machanical NUREG!CP-0036, (Compilation by) H. E. Bader and Engineers, and the Institute of Electrical and 13ectron-L A. Hanchey," Proceedings of the Workshop on its Engineers. There were approximately 550 partici-Nuclear Plant Ag64C, November 1982.ing," Sandra National I;tborato-pants from 16 countnes at the Symposium, ries, S AND82-2.

A total of 48 papers were presented in 7 technical The objective of the workshop, held August 4-5, sessms:

1982, in Hethesda Maryland, was to facilitate an ex-change of thoughts between the NRC and industry on

1. Aging Research Programs, time-related degradation and its influence on reactor

2. Aging of Structures and Mechanical Equip-safety. The spectfic goals were to define the problem,

ment, to discuss the state of knowledge on aging phenomena-

3. Aging of Electrical Equipment, and to antify future activities necessary to understand

4. Aging of Systems and Components, the problem'

5. Reliability, The need for a comprehensive program to identify

6. Role of Maintenance in Aging Management, l

the potential safety problems associated with plant ag-

7. Aging of Vessels and Steam Generators.

9 NUREG-1377

Main Citations and Summaries NUREG/CP-0105, Proceedings of the Seventeenth

3. Accounting for aging in design and operational Water Reactor Safety information Meeting, Vol. 3.

guidelines depends on several conservatisms to U.S. Nuclear Regulatory Commission. Paper by J.A.

provide prudent assurance that failures will not Christensen,"NPAR Approach to Controlling Ag-ing in Nuclear Power Plants," Pacific Northwest occur.The management of aging, however, re-laboratory,1 NL-SA-17487, March 1990, quires understanding and control of the time-For about 8 years, the NRC NPAR program has dependent degradation that actually occurs to been developing a technical understanding of and guid' implement and evaluate maintenance pro-grams.

ance for mitigating the effects of the time-dependent processes responsible for the aging.rclated detenora-

4. The design process considers aging-related deg-tion of structures, systems, and components that can radation of single components in estimating reduce safety margms m a nuclear power plant.

design lifetimes, but does not take into account the implications of common-cause failures due Controlof the effects of agingis at the center of the NPAR cfforts;it consists of three key elements:(1)se-to aging in redundant components or the ampli-lection of the structures, systems, and components in fied effect of aging in failure sequences involy-which aging-related degradation must be controlled, ing interactions of multiple components, ne (2) understanding the mechanisms and rates of the evaluation of these effects requires that accu-degradation, and (3) managing the degradation rate failure rate vs. time data for each compo-through effective surveillance and maintenance. These nent be applied using probabilistic methods that realistically model systems on a plant specific elements are implemented through vanous onEomg basis.

NRC and industry programs and initiatives as well as by conventional regulatory mstruments. Also, the three llecause of these kinds of considerations, specific elements are being addressed m a compik ion of good concerns, independent of plant design and operational practices that willintegrate the mformation developed parameters, over how and why structures, systems, and under Ni,AR and other studies of aging into a systems-components degrade with age must be resolved. The oriented format that tracks directly with the Safety generic content and stnicture of programs for address-Analysis Reports.

ing degradation due to agmg are discussed in this pa-per.

The need to mitigate time-dependent deterioration of NPP components is not a new or recent concept.

NUREG/CR-26 tl, J. P. Drago, R. J. Horkowski, D.11.

Degradation with time is, or should be, a prime consid.

Pike, and F. F. Goldberg. "The In-Plant Reliability cration in any design effort. The material specifications Data Base for Nuclear Power Plant Components:

Data Collection and Methodolop1)TM-8271, July Report, Oak and mechanical designs that characterize the struc-Ridge National laboratory, OR,,

tures, systems, and components of nuclear power 193),

plants reflect conscious, detailed concern on the part of the designer for the cffects of anticipated semce en-ne development of a component reliability data vironments and stressors on functionality over time.

base for use in nuclear power plant probabilistic risk as-The major codes and standards upor. which nuclear sessments and reliability studies is presented. The data power plant designs and msemce mspections are sources are the in-plant maintenance work request re-premised (e.g., ASME Hoiler and Pressure Vessel cords from a sample of nuclear power plants.nis data Code, Sections IH and XI) are based in large measure base is called the In. Plant Reliability Data System on recognized needs to achieve acceptable perform-GPRDSL lts features are compared with other data ance over a reasonable time whether or not age-related sources huch as the Licensee Event Report (LER) sys-degradation is explicitly addressed by the wording in kN Plant ReliaWiity Data (NPRD) sys-the codes. Even though time; dependent degradation is tem, and IEEE Standard 500. Generic descriptions of fundamental to the codes and standards that govem nuclear power plant systems formulated for IPRDS are the design, construction, and operation of nuclear outlined in the text.

power plants, additional concern over aging is war.

He major objective of the prog am described is to tanted for the following reasons:

provide an improved multipurpose data base. Compo-

1. The lack of specificity on time rates of deteriora-nents of each type of NSSS are included in the data tion in the codes requires the exertise of consid-

base, etable judgment in design and other functions, in addition to providing information on past failure De availability of explicit, detailed information rates and component down times, the IPRDS may be on aging rates and cysequences can result in used for:

more consistent judgments.

1. revising component test intervals and allowable 21 Developing technology will require that the un.

down times; derstanding of aging-related degradation be

2. identifying generic problems and recurring fail-pushed beyond that inherent in current codes.

ures; NUREG-1377 10 A

Main Citations and Summaries 3, identifying the variables (c.g., environment.op Preliminary results obtained from the pilot data emting mode, system, maintenance policy, etc.)

base in this report indicate WASH-1400 statistics to be that control component failure rates; nonconservative in reliability estimates for some valve types in certain failure modes.

4 providing an extensive data base against which to compare existing data sources (e.g., l.ERs NUREG/CR-3543, G. A. Murphy, R.11. Gallaher, and NPRDs) to assess the degree to which these M. L Casada, and 11. C. boy, " Survey of Operating data sources accurately reflect the actual com.

11xperiences from LIIRs to identify Aging 'Irends,"

Oak Rid e NationalIahoratory, ORNirNSIC-216, ponent reliability; 5, correlating current incidents with previous fail-

""" /7 this report describes the preliminary results of an utes, allowing for extrapolation in the near fu-assessment of information pertment to identifying age-

ture, related failures available in operating experience re-

6. identifying trends and patterns in the fail' ports. This assessment, by the Nuclear Operations ute characteristics of particular components or Analysis Center (NOAC) at Oak Ridge Nationallabo-aggregations of components; and ratory, utilized the computerized files of 1.icensec 7, identifying failure mechanisms over time for use

!! vent Reports (1.11Rs) and their predecessors to exam.

in defining the aging requirements for compo-ine age-related dyradation of safety-related equip-ment.

nent qualification, Abstracts of operating experience reports from NUREG/CR-3154, R. J. Ilorkowski, W. R. Kahl, T. I-commercial power plants reported from 1969 to 1982 Hebble, J. R.1 ra ola, and J. W. Johnson, "'lhe In-were surveyed. Over 7000 cvents were reviewed. Data Plant Reliability I ata Base for Nuclear Plant Com-included the system, component, subpart, the age-ponents: Interim Report-The Valve Component,"

related failure mechantsm, the severity, and the Oak Ridge National Laboratory, ORN!/FM-8647, method of detection of the failure, Wear, corrosion, December 1983.

crud, and fatigue were the identified failure mecha-This document details the collection and prelimi-nisms in over one-third of the 3098 age-related events.

nary analyses of data related to valves in thc In-Plant About two thirds of the failure severities were judged Reliability Data System (IPRDS).The data base is de-as a degraded state, and one-third were judged as cata-veloped primarily from historic:d records of corrective strophic failures. Pump and valve problems made up maintenance actions obtained directly from nuclear almost 30% of the (fed components. Almost two-plant maintenance files. A comprehensive valve popu-thirds of the reported failures were detected by routine lation is also included.This report presents data from surveillance testing indicating that such practices are one PWR and one BWR power plant.

effective techniques for monitoring and detecting age The report demonstrates the degree of distinction degradation of discrete components and systems. A and refinement in the reliability statistics that is possi-substantial number of events resulted from setpoint ble with data from the IPRI')S and suggests a general drift.

format for disclosure of suitable reliabbty statistics to NURI:G/CR-3818, N.11. Clark and D. L Berry, "Re-satisfy needs within the nuclear data-gathering com.

Q,irt of Results of Nuclear Power Plant, Aging munity. The examples given in the various tables and h3) nfi

[ti na l2boratories, figures suggest a useful method of comparing valve g

data and are representative of the degree to which reli-

.The objective of the workshops was to identify ability statistics for any particular valve can be ascer-whether there is any evidence of component or struc-tained.

tural time-related degradation, i.e., aging prob: cms, m One objective of this report is to examine the im-a nuclear power plant and, if so, what problems are of provement possible using IPRDS in refining the statis-greatest importance. Fifteen representatives from na-tics to ultimately focus on the reliability of particular tional laboratories, architect / engineers, nuclear steam valve types and valve operators in specific working en-supply system vendors, research firms, and one univer-vironments. Another objective is to generate com-sity participated. Questionnaires and group discussions ments from members of the nuclear data community as screened over 112 comp (ments believed to be suscepti-to the efficacy of the suggested formats for document-ble to excessive aging; pressure and temperature sen-ing valve information and the various methods used for sors, sahe operators, and snubbers emerged by con-comparison in this report.

sensus as the most important. Potential aging problems The report gives breakdowns of failure rates by fail-related to off. normal common-mode effects or prob-ure modes and by failure causes showing calculated lems that were just developing at the time werc outside maintenance frequencies and repair times. IPRDS re-the scope of the workshops because little av no first-pair time distributions, although unavailable from hand experience was available for these off. normal or f

IliRs, are also presented and esaluated.

yet to-be-explored circumstances. Recommendations 11 NURiiG-1377

i Main Citations and Summaries are made for a systematic approach to rating compo, plant and the two major : ore plant upgrades (repre-nents in terms of overall safety and for a cooperative senting a total of three distinct age proups)as well as to effon between industry research groups and regulatory evaluate previously developed surveillance technol-research groups to resolve known agmg problems and ogy, ne electrical testing was performed using the to identify off-normat oryct to-develop aging issues. In Electrical Circuit Characterization and Diagnostic

- addition to some well known aging mechanisms (e.g.,

(ECCAD) system developed by EG&G for the U.S.

- neutron embrittlement of pressure vessels) or prob-Department of Energy to use at TMI-2. Testing in-lems that manifest themselves as equipment failures cluded rneasurements of voltage, effective series ca-(e.g., stearn generator tube degradation), there is con-pacitance, effective senes inductance, impedance, cern that other types of aging problems may be devel-effective series resistance, de resistance, insulation re-oping. Their effects increase as nuclear power plants sistance, and time-domain reflectometry (l'DR) pa-get older, and some aging processes could eventually rameters. nc circuits evaluated included pressuriier affect power p' ant availability or safety, heaters, control red position indicator cables, miscella-neous primary system resistance ternpendure detec-NUREG/CR-3819, J. A. Rose, R., Survey of Aged tors (RTDs), nuclear instrumentation cables, and Steele, Jr.. K. G.

DeWall, and B. C. Cornwell, Power Ilant Facilities, Idaho National Engmeermg mfdY idection Dstem motor-oPcrated valves. It 3

Laboratory, EGG-2317, June 1985.

should be noted that the operabihty of these circuits was te se m an a er p ant opmtmn was con-The survey concentrated on component failures in c

at %pnpa em was no nec mtam LWR safety-related systems as determined from oper-the circuits in working condition following plant shut-ating histories. Only failurce that were determined to n, so n ort was eWenpat prposeme be age related were included.

in situ measurements and analysis of the data con-he age-related failure information gathered from firmed the effectiveness of the ECCAD system for de-the plant histories was analyzed for reoccurring failure tecting degradation of circuit connections and splices patterns. Early program emphasis was on isolating spe-along the high resistance paths; most of the problems

- cific equipment with high failure rates that were not al-were caused by corrosion. Results indicate a correla-ready the concern of other research efforts.The result' tion between the chronological age of circuits and cir-ing (gathered) data could not support the identification cult degradation.

of specific equipment. It did, however, imply a direct relationship between the failure and the failure NUREG/CR-4144, T. Davis, A. Shafaghi, R. Kurth, and mechanism.Thus the emphasis of the program was re-F. Leverenz,Importance Ranking Hased on Aging directed toward exploring the relationship of the fail-Consideration of Comp"onents included in Probabil-ure to the failure mechanism, istic Risk Assessments, Pacific Northwest labora-The results of this preliminary investigation indi-tory PNL-5389, April 1985.

cated that about 70% of the significant failures re-The method outlined in the report ranks power ported for the fluid systems analyzed were due to only plant components by using a risk-due-to-aging sensitiv.

four failure mechanisms (causes): crosion, corrosion, ity measure that describes the change in risk due to vibration, and foreign materials.This was subsequently changes in component failure rate (without describing verified by detailed study of several more plant systems closely the aging phenomena and the resulting time-and corroborated by field data obtained from person-dependent component failure rate).

nel interviews. In addition, there appears to be a strong The output from this study can be combined with correlation between the cause of component failure that from other studies (data, analytical or experimen-

- and the system m which the component operates.

tal) that identify the components most susceptible to De survey points out, with evidence, that the iden-

aging, tification and elimination of the system-level causes of The applications use average component unavail.

component failures is a viab:e approach to preventing ability equations currently emphiyed in probabilistic-and mitigatmg the major reported agmg effects, risk assessment (PRA) to calculate the risk-due-to-NUREG/CR-3956, M. R. Dinsel, M. R. Donaldson, and aging sensitivity. A more exact calculation is possible by F. T. Soberano, "In Situ Testing of the Shippingport using unavailability equations that include the time-Atomic Power Station Electrical Circuits," Idaho -

dependent characteristics of component failure rates:

National Engmeermg Laboratory, EGG-2443, April however, these time-dependent characteristics are not 1987.

well known. He risk-due-to-aging sensitivity measure This report discusses the results of electrical in situ presented here is therefore segregated from these testing of selected circuits and components at the Ship-time-dependent effects and addresses only the jime-pingport Atomic Power Station in Shippingport, Penn-independent portion of aging phenomena.He results sylvania.The goal was to determine the extent of aging identify the components that show the highest poten-or degadation of various circuits from the original tial for risk-due-to aging phenomena.

(

h NUREG-1377 12

hiain Citations and Summaries l

'Ihree operating NSSS were analyzed, and it was trends prior to failure and developing guidance for ci-found that the most risk-significant components are in fective and safe maintenance.

the auxiliary feedwater systern, the reactor protection NUREG/CR-4D4,11. D. Ilaynes, " Aging and Service system, and the service water systems, e.g., pumps, Wear of Electric Motor-Operated Valves Used in check valves, motor-operated valves, circuit breakers, Engineered SafetpFeature Systems of Nuclear and actuating circuits.

Power Plants: Apng " Assessments and Monitormg Future research on the time-dependent portion of Method Evaluations, Vol. 2, Oak Ridge National aging phenomena for these components is needed.o laboratory, ORNL-6170/V2, August 1989, completely describe the impact on risk.

Motor-operated valves (MOVs) are located in al-most all plant fluid systems. Their failures have re-NUREG/CR-4156, M. Subudhi, E. L Iturns, and J. It sulted in significar:t plant maintena..cc efforts. More Taylor," Operating Experience and Aging-Scismic important, the operational readiness of nuclear plant Assessment of Electric Motors,"llrookhaven safety related systems has often been affected by National Laboratory, IINI-NUREG-51861, J une MOV tiegradation and failure. Thus, in recent years, 1985.

MOVs ha$ e received considerable attention by the Nu-Alimited number of electnc motor categories with clear Regulatory Commission and the nuclear power direct safety significance were identified, and failures industry and were identified as a component for study due to insulation degradation were surveyed.

by the NRC NPAR program. In support of the NPAR Age-sensitive components (with respect to materi-Program, a comprehensive Phase 11 aging assessment als and design features) were reviewed, potential elec-n MOVs was performed by the Oak Ridge National trical and mechanical hazards were considered, opera-1.aboratory (ORNL), and the results of this study are tional and accident stressors were determined, and presented in this report.

monitorable f unctional indicators were identified'. 'the An evaluation of commercially available MOV contribution of pertinent seismic effects was assessed.

monitoring methods was carried out, as well as an as-and failure modes, mechanisms, and causes were re.

sessment of other potentially useful techniques.These viewed from existing data tuses.

assessments led to the identification of an effective, nonintrusive, and remote technique, motor current NUREG/CR--8234, W, L Greenstreet, G. A. Murphy, signature analysis (MCSA).The capabilities of moni-and D. M. Eissenberg," Aging and Sersice Wear of Electric Motor-Operated Valves Used in Engi-toring methods (especially MCSA) for detecting neered Safety-Feature Systems of Nuclear Power changes in operating conditions and MOV degradation Plants " Vol.1. Oak Ridge National laimratory, were investigated in controlled laboratory tests at ORNL-6170/V1, June 1 85.

ORNL, in situ MOV tests at a neighboring nuclear This report deals with motor-operated valves, fo-power plant, and the gate valve flow interruption blow-cusing on monitoring defects and degradation of nu-down test in Huntsville, Alabama.

clear plant safety equipment.The contents include the The background information and the work leading evaluation and identification of practical and cost-to the selected monitoring method are summarized be-effective methods for detecting, monitoring, and as-low. A primary objective of this study was to identify ef-sessing the severity, failure modes, and causes (mamly fectise methods for monitoring the condition of aging and service wear) of time-dependent degrada-motor-operated valves used in safety-related systems tion m nuclear plants. Also bemg considered are of nuclear power plants. In response to a need for im-manufacturer-recommended maintenance and sur~

proved methods for monitoring MOV condition, sev-veillance prachces and the selection of measurable pa-eral systems that use a variety of sensing devices and rameters (including functional indicators) for use in as-signal-processing equipment and provide signatures s singoperationalreadiness,establishingdegradation that yield useful diagnostic information have been de-trends, and detecting incipient failures. The report's veloped in the last few years. As part of the Phase 11 results are based on mformation derived from operat-MOV study, one of the motor-operated valve analysis ing experience records, nuclear mdustry reports,

& test systems (MOVATS) was evaluated in depth.

manufacturer-supplied information, and input from This evaluation and a desenplion of four other com-architect-engineer firms and plant operators' mercially available systems are included in this report.

Failure modes are identified for both the valve and in addition, the type and potential value of diag-the motor-operator assembly. I-or each failure mode, nostic information from many measurable parameters failure causes are listed by subcomponent or sub-were determined by ORNL tests using MOVs assembly,and parameters potentially useful for detect-mounted on test stands. 'the selected parameters are ing degradation that could lead to fallure are identtfied.

(1) valve stem position,(2) valve stem veh> city,(3) valve The method emerging from this analysis of the data stem strain, (4) torque-and limit-switch actuation can provide capabilities for establishing degradation (times of occurrence), (5) internal and external motor NUREG-1377 13

Main Citations and Summaries -

temperatures, (6) vibration (several locations), (7)

NUREG/CR-4257, S. Ahmed, A. Carfagno, and G.1 torque-switch angular position, and (8) motor current.

Toman, " Inspection, Surveillance, and Monitoring of Electrical Equipment inside Containment of Nu.

The tests led to the conclusion that the single uost clear Power Plants-With Applications to Electrical informative measurable parameter was also the one Cables," Vol.1, Oak Ridge National laboratory, that was most easily acquired, i.e., the motor current.

ORNUSUll/83-28915/1, August 1985.

MCSA was found to provide detailed information re-he purpose of this report is to describe currently lated to the condition of the motor, motor operator, available methodology for detecting and determining and valve across a wide range of values of parameters the amount and rate of age related detcrioration of and their variations,he recording and the analysis can safety related equipment. He general concepts of be donc during valve operation to render information monitoring equipment condition for this purpose are that characterizes transient and periodic occurrences.

described. nc goal is to detect deterioration in the in-Several tests were carried out to investigate the ca-cipient stage, prior to inservice failure and prior to the pabilities of monitoring methods (especially MCSA) point at which equipment can no longer be expected to perform its function when exposed to design basis acci-for detecting changes in operating conditions and dent conditions.

MOV degradation. Results from selected laboratory tests presented in the report illustrate examples of (1)

The application of condition monitoring is dis.

valve stem taper, (2) stem nut wear, (3) degraded volt' cussed specifically for electric cables.The goalis to de-age, (4) degraded valve st em l ubrication, (5) worm. gear termine the degree of cable degradation and to predict tooth wear, (6) obstruction m valve seat area. C) motor the remaining useful life. In situ nondestructive testing pinion disengagement, (8) degraded worm and worm-and destructive laboratory testing are discussed as are their limitations.

Interim recommendations gent lubrication, (9) stem packing adjustments, and (10) torque-switch settings.

are given for the implementation of a condition-monitoring program for cables.

In situ signature analysis tests were carried out by ORNL on a total of 20 aged MOVs at a neighboring NUREG/CR-4257, G. J. Toman, " Ins $ection, Surveil-nuclear power plant. Five of these MOVs were later lance, and Monitoring of Electrica Equipment in Nuclear Power Plants. Vol. 2: Pressure Transmit-retested after they were refurbished. Selected results ters." Oak Ridge National Laboratory ORNU from these tests are presented in this report and show, SUll/83-28915/3/V2, August 1986 for example, differences in motorcurrent signatures of This report describes the types of pressure trans-similar MOVs that were indicative of control-switch mitters commonly used in nuclear power plants setting variations and differences in component wear.

according to their application. The stresses that affect ne influences of refurbishing and inactivity on MOV these transmitters include ambient temperature, hu-operations were clearly seen in motor current signa-midity, radiation, process (fluid) medium pressure, and

- tures as well.

temperature.The most common effects of the stresses ORNL participated in the gate valve flow interrup-on the transmitters are calibration shifts. The evalu-tion blowdown test prograrn carried out under the di' ation-of failure data contained in Licensee Event rection of the Idaho National Engineering 12boratory Reports indicates that total failure of pressure trans-at Wyle 12boratories in Huntsville, Alabama. This test mitters occurs relatively infrequently, was an excellent opportunity for MOV diagnostic stud-Comparison of as-found and as-left calibration data les and, more important, a means for determmmg the s described as a partial means of evaluating the level of influences of hign blowdown flow on the operation of deterioration of a transmitter. Care must be taken to boiling water reactor isolation valves. The reduction in ensure that variations in method or procedure do not operatmg margin" of a MOV due to the presence of produce erroneous data and wrong conclusions. The a

Jditional valve running loads was imposed by high precision of the comparative measurements must also be high.

flow. This was observed in motor current and torque-switch angular-position signatures, a illustrated in this De evaluation of calibration data alone will not en.

study. In addition, the effects of differential pressure, sure the capability of operating underdesign basis acci-fluid temperature, and line voltage on MOV operation dent conditions. If, with time, steam or moisture pene-were clearly seen.

trates the transmitter housing, the transmitter elec-t ronics will become inaccurate and may fail.Therefore, The report presents information that should be the integrity of the housing seal must also be evaluated useful in resohing MOV issues concerning the NRC periodically to be able to predict continued perform-and the nuclear industry. Impartant areas not covered ance capability.

by the Phase II work are identified, and recommenda-Evaluation of inservice failures is recommended to tions for additional work are included, allow further differentiation between sudden failures NUREG-1377 14

Main Citations and Summaries NUREG/CR-4302, W, I. Greenstreet, G. A. Murphy, I

(having no precursor) and failures that can be detected Aging and R.11. Gallaher, and D. M. Eissenberg, "En meere,d in the incipient state. Such es aluations would aid in the Service Wear of Check Valves Used m I

further development of monitoring techniques, lle-Safety-Feature Systems of Nuclear Power tants, cause some of the transmitter failures are of the sud-Vol.1, Oak Ridge National laboratory, den type, periodic operability checks are an important ORNIA193/VI, Decemo :r 1985.

means of detectingIai!ures very soon after their occur-The report addresses detecting defects and moni-rence so that a significant number of failed (inactive or toring the degradation of nuclear plant safety cauip-inaccurate) transmitters do not remain undetected.

rnent.1he program is concerned with identifying and Acombination of operability rnonitoring and condi-evaluating practical and cost-effective methods for de-tion monitoring may be used toimprove the probability tecting, monitoring, and assessing the seventy of time-dependent degradation (aging and senice wear) of of successfully weathering aging proecsses and acci.

check valves in nuclear plants.1hese methods will al-dent conditions.

low degradation trends to be detected prior to failure and allow guidance for effective maintenance to be de-NUREG/CR-4279, S.11. Ilush, P. G. licaster, and veloped.

R. E. Dodge, " Aging and Service Wear of Ilydraulic The topics considered are failure modes and causes and Mechanical Snubbers Used on Safety-Related resulting from aging and service wear, manufacturer-Piping and Components of Nuclear Power Plants,"

recommended maintenance and survedlance practices, Vol.1, Pacific Northwest 12boratory, PNL-5479, and measurable parametets (including functionalindi-February 1986.

cators) for use in assessing operational readiness, es-This report presents an overview of hydraulic and tablishing degradation trends, and detecting incipient mechanical snubbers used on nuclear piping systems failure. The results presented are based on informa-tion derived from operating expenence records, nu-and components. The functions and functional re-clear industry reports, manufacturer-supplied infor-quirements of snubbers are outlined. 'lhe real versus mation, and plant operators, perceived need for snubbers is reviewed based primar.

ily on studies conducted by a Pressure Vessel Research Failure modes for check valves are identified and Committee. Tests conducted to qualify snubbers, to ac-are examined by identifying methods for detecting fail-cept them on a case-by-case basis, and to establish their ures and dif ferentiating between their causes. For each fitness for continued operation are reviewed.

failure mode, failure causes are listed by component or subassembly, and parameters potentially usefut for de-

,this report had two primary purposes:(1) to assess tecting degradation that could lead to failure are tabu-the effects of various aging mechanisms on hydrauhe lated.

and mechanical snubber operation (e.g., leaking of The report also identifies parameters potentially seals, functional failures) and (2) to determine the effi.

aseful for enhancing the detection of degradation and cacy of existing tests in determining the effects of aging incipient failure; these parameters mclude dimensions, and degradation mechanisms. These tests include bolt torque, noise, appearance, roughness, and breakaway force, drag force, vekscity/ acceleration cadm, p range for activation in tension or com pression, release rates within specified tension / compression limits, and NUREG/CR-4302. M.111laynes, " Aging and Service restricted thermal movement. The snubber operating Wear of Check Valves Used in Engineered Safety-experience was reviewed using licensee event reports Feature Systems of Nuclear Power Plants: Agmg, and other historictd data for a period of more than 10 Assessments and Monitorine Method Evaluations,

~

years. Data were stattsucally analyzed using arbitrary Vol. 2 Oak Ridec National I2boratory, snubber populations. Value-impact was considered in ORNIc6193/V2, April 1991.

terms of exposure to a radioactive environment for exa-

'lhe failures of check vaives have resulted in signifi-mination/ testing and in terms of the influence of lost cant maintenance efforts and, on occasion, in water snubber function and subsequent testing prognm ex-

• "' " F#

"*D"* W'* '

pansion on the costs and operation of a nuclear power and damage to flow system components. Ihese failures plant.The implications of the observed trends were as-have largely been attnbuted to severe degradation of sessed; recommendations include modifying or im-internal parts (e.g., hinge pins, hinge arms, disks, and proving the examination and testing procedures to en-dtsk nut pms) resulting from instability (flutter) of hance snubber reliabihty. Opt mization of snubber check valve disks under normal plant operating condi-populatiom, by selective removal of unnecessary snub-tions. Present survedlance requiremems for nuclear power plant check vahes have been inadequate for bers was also considered.

NUlWG-1377 15

._m i

i l

. Main Citations and Summaries timely detection and trending of such degradation be-cause neither the flutter nor the resulting wear can be types of abnormalities in the components so that cor.

rective measures can be implemented prior to loss of detected prior to valve failure. This report describes safety function. A field test program was carried out to evaluations carried out in support of the NRC NPAR program of the following developmental or commer-evaluate valve signature analysis as a surveillance method to achieve these results as well as to detect in, cially available methods for diagnostic monitoring of correct adjustments in motor-operated vahes. Ihc check valves:

technique specified in the title (MOVATS)is the sub-1.

Acoustic emission monitoring, ject of this report, in situ signature traces were ob.

2.

Ultrasonic inspections

• tained in 36 motor-operated valves at four nuclear plant sites. Described are the test equipment package, 3.

Magnetic flux signature analysis, the method of obtaining the signatures, and determina.

4.

Radiography, tions made as a result of analyring them. Ibcd on the 5.

Pressure noise signature analysts.

results of the signature-analysis technique and those

'Ilese evaluations were focused on the capability of cbtained from the field-test program, the capabihties each method to provide diagnostic information useful and hmitations of MOVATS are discussed.

in determining the effects of aging and service wear NUREG/CR-4457, J. L...dson and J. E. liardin, (degradation) and detecting failures and unoe:Jrao!c Nucbr 1o e Pl nts operating modes, Commercial supnHer. of three check d oN i i ngi erinE valve monitoring systems rceently panicipated m a laboratory, EGO-2488, July 1987.

ecmprehensive series of tests designed to evaluate the

' Itis report presents the results of a study of aging capability on each monhoring technology to detect the effects on safety-related batteries in nuc! car power pod tion, motion, and w car of check valve mternals and plants. The purpose is to evaluate the aging effects vale seat leakage. 'fhis report desenbes these tests, caused by battery operation in a nuclear facility and to which were directed by the Nuclear Industry Check evaluate maintenance, testing, and monitoring prac-Valve Group and carried out at the Utah Water Re-tices with respect to the effectiveness of these practices scarch Laboratory' in detecting and mitigating the effects of aging. The Each monitoring method is described and com-study follows the NRC NPAR apptoads and investi.

pared with the others, and areasin need of further de-gates the materials used in battery construction. It also velopment are identified. Examples of test data ac-identifies stressors and aging mechanisms, presents op-quired under controlled laboratory conditions and erating and testing experience : elated to aging effects, some field test data acquired at operating nuclear analyzes battery-failure event reports in various data plants are presented-bases, and evaluates recommended maintenance prac-tices. Data bases that were analyzed included the Of the methods examined, acoustic emission moni.

NRC's Licensee Event Report system, the Institut e for toring, ultrasonic inspection, and magnetic flux signa.

Nuclear Power Operations' Nuclear Plant Reliability ture analysis provided the greatest !cvel of diagnostic information. These three methods were shown to be Data System, the Oak Ridge National 12boratory's In-Plant Reliability Data System, and the S. M. Stoller useful in determimng check valve condition (e.g., disk Corporation's Nuclear Power Experience data base.

position, disk motion, and seaileakage), although none of the methods was, by itself, successful in monitormg NUREG/CR-4564. W. E. Gunther, M. Subudhi, and all three condition mdicators. However, the combma-J. H. Taylor, " Operating Experience and Aging-tion of acoustic emission with either ultrasonic or mag-Seismic Assessment of Battery Chargers and Inver-ters," Brookhaven National l2boratory, netic flux monitoring yields a monitoring system with BNL-NUREG-51971, June 1986, sufficient sensitivity to detect all major check valve op-Battcry chargers and inverters are vital components erating conditions. All three methods are still under development and are expected toimprove as a result of of the nuclear power plant electrical safety system.The further testing, analysis, and evaluation.

objectives of this program are to (1) identify concerns related to the aging and service wear of equipment op-QUREG/CR-4380, J. L Crowley and D. M. Eissenberg, erating in nue: ear power plants, (2) assess their possi-

" Evaluation of the Motor-O erated Valve Analysis ble impact on plant safety, (3) identify effective m-and Test System (MOVATS$ to Detect Degrada.

SPection, surveillance, and monitoring methods, and tion, incorrect Adjustments,and Other Abnormali.

(4) recommend suitable maintenance practices to miu-ties in Motor-Operated Valves," Oak Ridge gate aging-related concerns and diminish the rate of National 12boratory, ORNL-6226, January 1986.

degradation due to aging and service wear.

An important aspect of the NPAR program strat-The designs of battery chargers (3 types) and inver-l egy is to demonstrate the utility of condition monitor-ters (4 types) and the materiale for their construction ing, signature analysis, and other surveillance methods are i eviewed to identify age-sensitive components. Op-for detecting, differentiating, and trending various crational and accidental stressors are determined, and NUREG-1377 16

Mam Citations and Summaries their effect on promoting aging degradation are as-matenals of construction. operating requirements, and sessed. Variations in plant electrical designs, as well as modes of operation Failure modes and causes due to

~

system and component impacts were studied, l'ailure aging and senice wear are idcMified and explained, modes, mechanisms, ar3 causes were reviewed from and measurablc parameters (including functional indi-operating experience and existing data banks. The cators)for potential use in assessing operational reado study also considerci scismic effects on age-degraded ness, establishing degradation trends, and detecting in-components of battery chargers and inverters.

cipient failures are outlined.

A series of measures to correct present deficiencies

~lhe performance indicators that can be monitored in surveillance, monitoring, and insenice testing prac-to assess component detenoration due to aging or

]"

other relevant stressors are identified. Conforming tices is discussed.The main body of the report is sup.

with the NPAR strategy as outlined in the program plemented by 9 ramber of relevant appendices;in par-ticular, t major appendix is included on engineering plan, the study also includes a review of current stan.

and design information useful to assess operational dards and guides, maintenance programs, anc research readiness.

activities pertaining to safety-related battery chargers and inverters for nuclear power plants.

NUREG/C11-tS97, IL M. Kitch, J. S. Schionski, P. J.

S "M4 und W. V. Cesarski," Aging and,Senice l

NUREG/CR-4590, K. R lloopingarner, J. W. Vause, clear Power Plants. Vol. yn Pumps br N R Nu-u Agmp A har f Amhay beedw D. A. Dingee, and J. F. Nesbat, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating M nitoring Method Evaluations, Oak Ridge and Expert Experience," Vols. I and 2, Pacific National laboratory, ORNI A282/V2, June 1968.

Northwest laboratory, PNL-5832, August 1987.

'lhe subjects specified in the title are described and Pacific Northwest Iahoratory evaluated opera.

discussed in four major sections:

tional and expert experience pertaining to the aging

1. Failure causes, degradation of diesel generators in nuclear plant serv.

2. Description of inspection, suncillance, and ice, The research identtfied and characterized the condition monitoring (ISCM) methods, contribution of aging to emergency diesel generator

3. Evaluata.c of ISCM methods, and failures.

4. Role of maintenance in rJleviating aging and Volume I reviews diesel-generator experience to senice wear, identify the systems and components most subject to aging degradation and isolates the major causes of fail-Failure causes attributable to aging and service ure that may affect future operational readmess-wear are given and ranked in terms of importance.

Evaluations show that, as plants age, the percentage of Cause identifications are made on the basis of experi-aging related f.alutes increases and failure modes ence, postservice examinations, and in situ assess-change, A compilation is presented of recommended ments.

corrective actions for the aging-related failures identi' Measurable parameters related to failure causes fied, and the trend of these failures is discussed.This are identified. ISCM methods are cpecified, and evalu-study also includes a review of current relevant indus-ations are made based on Westinghouse experience.

try programs, research, and standards. Volume I pre-On the same basis, recommendations are given on in-sents the results of the Phase I research that identifies spection, surveillance, and condition monitoring. The the components and systems most susceptible to aging ISCM methods are mtended to yield required capabdi-degradation and the major causes of such degradation.

ties for establishing operational readmess as well as for Volume 2 reports the results of a workshop held on detecting and tracking degradation and its trends-May 28 and 29,1986, with industry representatives to The role of maintenance in alleviating and mitigat-discuss the technicalissues associated with aging of nu-ing aging and senice wcar effects is discussed, and the cleu senice emergency diesel generators.The techni-relationsh.p of maintenance to ISCM methods is iden-cal issues discussed most extensively were man /

tified. Predictive, preventive, and corrective mainten-machine interfaces, component interfaces, thermal ance practices are discussed and evaluated.

gradients of startup and cooldown, and the need for an Appendices contain a detailed discussion c.a ISCM accurate industry data base for trend analysis of the methods, fadute data base information, auxiliary feed-diesel generator system.

water pump ( AUXFP) installation lists (location sur-vey), a discussion of low-flow testing, auxilian feed-NUREG/CR-4597, M. l.. Adams and E. Makay," Acing water system desenptions (with flow-diagrams and i

i and Senice Wear et Auxiliary Feedwater l umpi for PWR Nuclear Power Plants. Vol.1: Operating schemes), AUXFP mmimum-floorate criteria, and j

Experience and Failure identification,' Oak Ridge guidelmes proposed by Westinchouse for full flow National laboratory, ORN14282/V1, July 198ti.

testing. Note: The drafi of this

01. 2 (with the same in this report, typical aunhary feedwater pump fea-title) v.as issued by Westmghouse Electric Corpora-tures are described in terms of configuration details, tion, Generation Technology Sptems Divnion, in 17 NU REG-1377

hiain Citations and Summaries April 19S6, coauthored by D. ht. Kitch, M. Vuckovich, fected by aging or degradation of structural materials, W. V. Cesarski, ano P. J. Sowatskey, and (4) performing research in support of all these needs. It should be stressed that there is no widely ac-NUREG/CR-4652, D. J. Naus. " Concrete Component cepted standardi7cd methodology for quantifying the Aging and Its Significance Relative to Life Exten-sion of Nuclear Power Plants," Oak Ridge National condition and capacity of an ;ndnidual concrete struc-ture.

laboratory, ORNIITM-10059, September 1986, The objectives of this study are to (1) expand upon NUREG/CR-4692. G. A. Mu hy and J. W. Cletcher the work that was tmtiated m the first two Electne 11 " Operating 11xperience eview of Failures of Power Operated Relief Valves and lilock Valves in Power Research Institute studies relative to longevity Nuclear Power Plants," Oak RidgOctober 1987, e National and life extension considerations of safety related con-I aboratory, ORN!JNOAC-233, crete components in light-water reactor (1.WR) facili-

'this report contains a review of nuclear power ties and (2) provide background that will logically lead plant operating events involving failures of power-to subsequent development of a methodology for operated relief valves (PORVs) and associated block assessing and predicting the effects of aging on the per.

valves (llVs). Of the 230 events identified.101 involved formance of concrete-based materials and compo-PORV mechanical failure, 91 were attributable to nents.

PORY control failure,6 involved design or fabrication Applications of safety-related concrete compo.

of the PORVs, and 32 involved IIV failures. The report nents to 13VR technology are identified, and pertinent contains a compilation of the PORV and IIV failure structures (containment buildings, contaimnent base events, including failure cause and severity.The events mats, biological shield walls, main building, and auxij.

are identified as to plant and valve manufacturer. An

(

iary buildings)and the materials of which they are con.

assessment of the need to upgrade PORVs and ilVs to structed (concrete, mild steel reinforcemcnt, pre, safety grade status concludes that such action would stressing systems, embedments, and anchorages) are improve PORV and ilV reliability, The greatest im, described. liistorical performance of concrete compo.

provement in reliability would result from using newer, nents was established through information presented more reliable PORV designs and improving testing, di.

on concrete longevity and component behavior in both agnostics, and maintenance applied to PORVs and 13VR and high-temperature gas-cooled reactor appli.

IIVs, particularly to the llV motor operators. A sum-cations. Also, a review of problems involving concrete mary of interviews conducted with four PORV manu-components in both general civil engineering and nu.

facturers is also included in the report.

clear power applications is given. The majority of the NUREG/CR-4715. G. J. Toman, V. P. llacanskas, T. A.

problems identified in conjunction with nuclear pcwer Shook, and C. C, ledlow, "An Aging Assessment applicatior's were minor; they include concrete crack-of Relay and Circuit Ilreakers and System Interac-ing, concrete voids, or low concrete strengths at an tions," llrookhaven National I aboratory, Franklin early age. Five incidents involving 1.WR concrete con-Research Center, Philadelphia, PA, tainments that are considered significant are described 13NirN1"1EG-52017, June 1987, in detail.

As pa: if the NRC NPAR program,1 ranklin Re-Potential eavironmental stressors and aging factors search Center analyzed the effects of agmg on safety-to which LWR safety-related components could be related circuit breakers and relays under contract to subjected are identified and discussed in terms of dura-llrookhaven National laboratory. Circuit breakers and bility factors related to the materials used to fabricate relays in a PWR safety injection system were evaluated the components. The current technology for detectmg with respect to the aging caused by system operation, concrete agmg phenomena ts also presented in terms

'Ihe effect of circuit breaker and rclay deterioration on of methods applicable to the particular matenal system the ability of the system to perform its safety functions

- that could experience deteriorating effects. Remedial was also evaluated. The st udy included nrotective, con.

.neasures for the repair or replacement of degraded trol, and logic relays, as well as molded-case and metal-

. concrete components and their effectiveness are dis-clad switchgear circuit breakers. Analysis of nuclear cussct. Finally, considerations relative to developing a power plant failure data confirmed that normally ener-damage q.shodology for assessmg the durability fac-gized relays commonly used in safety systems suffer tor, detenomtion rates, and predictionof structural re-from more rapid deterioration than do deenergized re-liabihty are outhned, lays. The failures were attributable to coil deteriora-tion, changes in dimensions of critical organte com-Conclusions and recommendations of the report ponents, and changes in characteristics of timing note the need for (1) obtaining aging data from decom-diaphragms from thermal deterioration. Some of the missioned plants, (2) using inservice inspection pro-failure modes will prevent fail-safe operation. The grams to provide aging trends, (3) developing a meth-electrical control and mechanical portions of metal-odology to quantitatively and uniformly (i.e., using the clad switchgear were found to be more failure prone same procedures) assess structural reliability as af-th m the main contacts and arc extinguishing systems.

NUREG-1377 18 iimiin irri iili i-r i i ri-nr n-i i r

Main Citations and Summaries Analysis of failure data for circuit breakers and relays Plant Reliability Data System, and plant maintenance indicated a general trend of increasing failure rates in records.'the purpose of the review was to evaluate the the period of 6 to 11 years following the start of com-potential significance of aging, including cycling, trips, rnercialoperation of the plants.The aging interaction and testing, as a contributor to degradation of the RTS and ESFAS. Tables show the percentage of events for study evaluated the interaction of aged relays and cir-RTS and ESFAS classified by cause, coruponents, nnd cuit breakers in a safety injection system with regard to subcomponents for each of the nuclear steam supply five events requiring the system to start operation.

system vendors. A representative Babcock and Wilcox Failure of redundant trains from common-mode fail-ure of a particular type of circuit breaker or relay is not plant was selected for detailed study.The NRC NPAR guidelines were followed in performir;g the detailed expected. Ilowever, the number of different types of study that identified materials susceptible to aging, potential failures supports the need for a strong stressors, environmental factors, and failure modes for maintenance and surveillance program to prevent mul-the RTS and ESFAS and the relevant generic instru-tiple age-related failures from affecting redundant mentation and control systems. Functional indicators safety trains.

f degradation are listed, testing requirements evalu-NUREG/CR-4731, V. N. Shah and P. E. MacDonald, ated, and regulatory issues discussed.

" Residual life Assessment of Major Light Water NUREG/CR-.4747, B. M. Meale and D. G. Satterwhite, Reactor Components," Vol.1. Idaho National Engi-

"An Aging Failure Survey of Light Water Reactor neering l2boratory, EGO-2469, June 1987.

Safety Systems and Components," Vol.1. Idaho NUREG/CR-4731, V. N. Shah and P. E. MacDonald, National Engineermg bboratory, EGG-2473, July

" Residual Life Assessment of Major Light Water 1987*

Reactor Components-Overview, Vol. 2, Idaho NUREG/CR-4747, B. M. Meale and D. G. Satterwhite, National Engmeeting Laboratory, EGG-2469, "An Aging Failure Survey 01 Light Water Reactor Safety S stems and Com nents," Vol. 2, Idaho November 1989.

Th.is report presents an assessment of the aging Nationaf Engineer'ng b ratory, EGG-2473, July (time-dependent degradation) of selected major light

193g, water reactor components and structures. The stres-This report describes the methods, analyses, re.

sors, possible degradation sites and mechanisms, sults, and conclusions of two different aging studies.

potential failure modes and currently used non.

The first study was a survey of light water reactor com-destructive examinations, inservice inspection (ISI),

ponent failures associated with 15 selected safety and and life assessment methods are discussed for major support systems. Analysts used computeriicd sorting light water reactor components, Volume 1 covers techniques to classify component failures into generic PWR and BWR pressure vessels, PWR containment failure categories. The second study was a careful exa-structures, PWR reactor coolant piping, PWR steam mination of component failure records to identify and generators, BWR recirculation pipmg, and reactor categorize the reported causes of component failures.

pressure vesset supports. Volume 2 covers PWR reac.

The systems evaluated in the failure-cause analysis tor coolant pumps, PWR pressunzers, PWR pres' were the auxiliary feedwater, Class IE electric power sunzer surge and spraylines, PWR reactor coolant sys-distribution, high-pressure injection, and service tem charging and safety injection nonles, PWR feed' water, Tables and figures indicate the systems and the water 1mes, PWR control rod dnve n,echanisms and re-components within the systems that are most affected actor mternals, BWR containments, BWR feedwater by aging. Engineering insights drawn from the data are and main steam lines, BWR control rod drive mecha-provided. Volume 2 presents allof the Volume 1 data nisms and reactor internais, PWR and BWR electrical from FY-86 combined with the data gathered in cables and connections, and PWR and BWR emer-

py_g7, gency dicel generators. Unresolved technical NUREG/CR-4769, W. E. Vesely, " Risk I! valuations of issues related to understanding and managing the ag-Aging Phenomena:He Linear Aging Reliability ing of these major components, including require-hlodel and its Extensions," Idaho National ments for advanced ISI and life assessment methods, lingineering Laboratory, EGG-2746, April 1987, are also discussed.

A model for failure rates of light water reactor safety system components due to aging mechanisms NUREG/CR-4740, L C. Meyer," Nuclear Plant-Aging has been developed from basic phenomenological con-Research on Reactor Protection Systems," Idaho siderations. In the treatment, the occurrences of dete-National Engineering I aboratory, EGG-2467, rioration are modeled as following a Poiss<m probabil-January 1985.

This report presents the results of a review of oper-ity process. He seventy of damage is allowed to have ating experience for the reactor trip system (RTS) and any distribution; however, the damage is assumed to the engmected safety feature actuating system (ES-accumulate independently. Finally, the failure rate is J

l FAS) reported in Licensee Event Reports (LERs), the modeled as being proportional to the accumulated Nuclear Power Experience data base, the Nuclear damage. Using this treatment. the linear aging-failure-NUREG-1377 19

hiain Citations and Summanes rate moJel is obtained. 'the applicabihty of the hnear determination of achievable accuracies in RTD cali-agmg model to vanous mechanisms is discussed. The bration.

e model is also extended to cover nonlmear and depend' ent aging phenomena. lhe implementation of the lin' NUREG/CR-4939, bl. Subudhi, W. I!. Gunther, J. I1.

car aging modelis demonstrated by applying it to the Tmlor, R. I nfaro, K. hl. Skreiner, A. C. Sugarman, aging data collected in the NRC NPAR program.

and h1. W. Sheets, *!mproving hiotor Reliability in Nuclear Power Plants; Volume 1: Performance IWaluation and hiaintenance Practices; Volume 2:

NUREG/CR-4819. V. P. llacanskas, O. C. Roberts, and Functional Indicator Tests on a Small Electric G. J. Toman, " Aging and Service Wear of Solenoid-hiotor Subjected to Accelerated Aging; Volume 3:

Operated Valves Used in Safety Systems of Nuclear Failure Analysis and Diagnostic Tests on a Naturally Power Plants. Vol.1: Operating Experience and Aged Electric Niotor: llrookhaven National Labora-Failure identification," Oak Ridge National 1abora-tory, llNicNUREG- $2031, November 1987.

tory, ORNIJSUl%/83-2891Sl4/V1, h1 arch 1987.

l'olume 1: Performance baluation and Maintenance An assessment of the types and uses of solenoid-Practices operated valves (SOVs)in nuclear power plant safety-This report presents recommendations for devel-related service is provided in the report.Through a de-eping a cost-effective program for petformance evalu-scription of the operation of each SOV combined with ation and maintenance of electric motors in nuclear knowledge of nuclear power plant applications and op-power plants. 'lhese recommendations are based on crational occurrences, the significant stressors respm-current industry practices, available techniques for sible for degradation of SOV performance are identi-monitoring degradation in motor components, manu-fied. A review of actual operating experience (includ-facturers' recommendations, operating experience, ing failure data) leads to the identification of potential and results from two laboratory tests on aged motors nondestructive in situ testing which, if properly devel-The test results (on a small and a large motor) provide oped, could provide Ihe methodology for monitoring the basis for recommendmg the various functional indi-the degradation of SOVs. Recommendations are out-cators for maintenance programs.

lined for continuing the study into the test methodol-ogy development phase, The overall preve,tive program is separated into two broad areas of activity aimed at mitigating the po-Nt'R,EG/CR--1928, II. N1. Hasemian, K. h1. Petersen, tential effects of equipment aging: performance evalu-L W. Kerlin, R. L. Anderson and K. E. Ilolbert, ation and equipment maintenance.The latter involves

,Deg'radation of Nuclear Plant I emperature Sen-actually maintaining the condition of the equipment, sors, Analysis and hicasurement Sersices Corpora, while the former m.volves monitoring degradation due tion, Knoxville, TN, June 19S7.

to aging. The monitoring methods are further catego-A program was established and initial tests were rized as periodic testing, surveillance testing, continu-performed to evaluate long-term performance of resis ous monitoring, and inspections.

tance temperature detectors (RTDs) of the type used

.Ilu.s study focuses on relevant methods and procc-in U.S. nuclear power plants.This report addresses the dures with the goal of maintaining the motors in a nu-effect of aging on RTD calibration accuracy and re-clear facility operationally ready, 'Dus mcludes an sponse time. The Phase I effort (lasting about evaluation of various functional indicators to deter-6 months) included exposure of 13 safety-grade RTD mine their suitability for trending assessments when elements to simulated LWR temperature regimes.

monitoring the condition of motor components. The Full calibrations were performed initially and monthly, intrusiveness of test methods and the present state of sensors were monitored and cross-checked continu.

the art for u. sing the test equipment in a plant environ-ously during exposure, and resp (mse time tests were ment are dtscussed, performed before and after exposure. Short term cali.

Implementation of the information provided in this bration drifts of as much as 1.8 F(l *C)were observed.

report will improve motor rehability in nuclear power Another result was that small respmse times were es.

plants. The study indicates the kinds of tests to con-sentially unaffected by the testing performed.

duc, how and when to conduct them, and to which mo-This program has demonstrated that there is a tm he m shouM be applied.

sound reason for concern about the accuracy of tem.

3 !nme 2: Functional Indicator Tests on a Small peratute measurements in nuclear powe'r plants, Electric Motor Subjected to Accclcrated Aging These limited tests should be expanded in a Phase II A 10-horsepower electric motor was artificially program to involve more sensors and longer exposures aged by plug reverse cycling for test purposes.The mo-to simulated LWR conditions in order to obtain statisti-tor was manufactured in 1967 and was in service at a cally significant data. Such data are needed to establish commercial nuclear power plant for twelve years. Vari-the length of meaningful testing or replacement inter-ous tests were performed on the motor throughbut the vals for safety-grade RTDs. An important corollary aging process. The motor failed after 3.79 million re-benefit from this expanded program would be a better versals (3 seconds per reversal) over seven months of NUREG-1377 20

Main Citations and Summaries ing and service wear degradation, and a review of cut-testing. Each test parameter was trended to assess its rent inspection, surveillance, and monitoripp methods, i

suitability in monitonag aging and service wear degra.

including manufacturer recommended surveillance dation in motors. Results and conclusions are discussed and maintenance practices.The detailed study identi-relative to the applicability of tne tests performed to fics materials susceptible to aging, various stressors, motor maintenance programs of nuclear power plants.

and environmental factors. Performance parameters Volume 3: failure Analysis and Diag *mstic Tests on a or iunctionalindimtors potentially useful in detecting Naturally Aged Large Electric Motor degradation are also identified, and preliminary rec-Stator coils of a naturally failed 400-hp motor from ommendations are made regardinginspection, surveil-the Ilrookhaven Nationall2boratory test reactor facil-lance, and monitoring methods.

In addition to the above engineering evaluation, the ity were tested for their dielectric integrities.The mo-components that contributed to system unavailability tor was used to drive the primary reactor coolant pump were determined, and the contribution of aging to for the last 20 years. Maintenarce activities on this mo' llPIS unavailability was evaluated. The unavailability tot during its entire service life w ere minimal, with the assessmpnt utilized an existing probabilistic risk assess-exception of meggering it periodically.The stator con-sisted of ninety individual coils, which were separated ment, the linear aging model, and generic failure data.

for testing. Seven different dielectric tests were per-NUREG/CR-1977, R. Steele, Jr. and J. G. Arendts, formed on the coils. Each set of data from the tested

• SiI AG Test Series: Seismic Research on an Aged coils mdicated a spectrum of variation depending on United States Gate Valve and on a Piping System in their aging conditions and characteristics. Ily compar-the Decommissioned Ileissdampfreaktor ()IDR):

ing the test data to baseline data, the test methods were Summary," Vol.1, Idaho National Engineering Iahoratory, EGG-2505, August 1989.

assessed for application to motor mainte-nance pro-NUREG/CR-8977, R. Steele, Jr. and J. G. Arendts, grams in nuclear power plants. Also included in this "SH AG Test Series: Seismic Research on an Aged study are results of an investigation to determine the United States Gate Valve and on a Piping System in cause of this motor's failme. The aged condition of a the Decommissioned lleissdampfreaktoi (llbR):

second identical primary pump motor, which is of the Appendices," Vol. 2, Idaho National Engineering same age and is presently in operation, is discussed.

Iaboratory, EGG-2505, August 1989, Recommendations relating to the applicabihty of each This report descnhes the investigation, results, and of the dielectric test methods to motor maintenance conclusions of the INEl. effort to determine the cause programs are formulated.

of the reduced performance of a naturally aged Crane NUREG/CR-1967, L C. Meyer," Nuclear Plant Aging gate valve with a Limitorque motor operator. The Research on High Pressure injection Systems,"

motor-operated valve served 25 years in the Ship-Idaho National Engineering laboratory, pingport Atomic Power Station as a feedwater isolation EGG-2514, August 1989.

valve before being refurbished and installed in a piping This report presents the results of a review of light system in the Heissdampfreaktor (HDR), where valve water reactor high pressure injection system (llPIS) operability in typical pressure and temperature envi-operating experience reported in the Nuclear Power mnments and during simulated earthquakes was stud-Experience Data llase, Licensee Event Reports ied. During the test program it was discovered that, un-(LERs), the Nuclear Plant Reliability Data System, der some hydraulic loadings, the motor operator failed to reach torque !cvels high enough to open the closing and plant records.

torque switch. Failure of the torque switch to open Operating expenence of nuclear power plants was evaluated to determine the significance of aging-caused the motor to stall. In normal plant service, stall-related service wear on equipment and its possible im-ing an operator motor can cause motor burnout and render the valve inoperable for subsequent safety pact on safety.The HPIS and those portions of related functions.

systems needed for operation of the HPIS were se-An extensive investigation was conducted to try to lected for detailed study in order to evaluate the poten.

isolate the cause of the poor performance of the motor tial significance of aging as a contributor to the degra-dation of that system. Tables show the percentage of operator. 'lhis investigation included follow-on in situ significant events for HPIS classified by cause, compo-tests at ilDR, dynamometer testing of the motor op-nent, and subcomponent for PWRs and llWRs. A rep-erator at the Limitorque laboratory, testing of the resentative Habcock and Wilcox plant was selected for torque spring at INEL, dynamometer testing of the motor alone at the Peericss Motor laboratory, and a detailed study.

mathematical analysis of the HDR power circuit.The The NPAR guidehnes provided the framework through which the effect of aging on HPIS was studied, investigation identified three causes of the motor-operator's poor performance: torque spring aging, and these guidelines were followed throughout the re-heating of the motor windings, and resistance in the de port,which presents an identification of failure modes power cabling at !IDR. The investigation also demon-a preliminary identification of failure causes due to ag-NUREG-1377 21 I

m..

Main Citations and Summaries strated that normal plant testing of valves is not ade.

NUREG/CR-4992, G. C. Roberts, V. P. Bacanskas, and quate to ensure proper performance under flow and G. J. Toman, " Aging and Service Wear of Multi-pressure loads in combination.

stage Switches Used in Safety stems of Nuclear Power Plants," Vol.1 Oak Ri e National labora.

During the follow-on tests at IIDR, we found that, tory, ORNI/SUll/83-28915/5 1 September 1987, when the valve was subjected to now loads and pres-An assessment of the types and uses of multistage sure loads in combination, the valve either torqued out switches in nuclear power plant safety related senice in the partially open position, stalled in the partially open position.or stalled in the iully closed position, de-is prosided. Through a description of the operation of pending on the load and the torque switch setting.The each type of switch combined with knowledge of nu-valve torqued out in the fully closed position only when clear power plant applications and operational occur-rences, the significant stressors responsible for multi-pressure and flow loads were very low, stage switch deterioration are identified. A review of Undersized power supply cabling resulting in high operating experience (failure data) leads to identifi-resistance has surfaced as a problem in at least two de cation of potential and recommended morutoring motor operators in the field. nough the other factors techniques for early detection of incipient failures.

contributed to the anomalous performance of the valve Although the operating experience does not justify ex-at HDR. undersized cabling was the main cause. The tensive deterioration monitoring of multistage NRC has recently issued an information notice regard, switches, nondestructive testing methods that could be ing the issue.

used to evaluate the condition of switches are identi-None of the three problems discovered during the fied. The report presents a detailed description of the llDR tests and follow-on investigation would be de-components, materials of construction, and operation tected during the normal in plant testing where the of each of the multistage switches included in the as-valves are subjected to no load or to pressure loads sessment. Also, it provides an analysis of failure data from the LER system. An analysis of the various failure alone. The problems are detectable only at higher modes of multistage totary switches and their related loadings, that is, flow loads in combination with pres-causes is also given.The existing recommended and re-sure loads, where the load slows the motor down to the extent that momentum cannot carry the unit through quired maintenance and surveillance practices are complete closure and torqueout-listed. Several techniques with a potential for assessing the condition of switch components and possibly pre.

dicting age related failures are identified. It is recom-NUREG/CR-4985. M. Subudhi, J. H. Taylor, mended that insenice failures be analyzed to deter.

J. Clinton, C, J. Czajkowski, and J. Weeks, " Indian mine whether the failures are due to random defects or Point 2 Reactor Coolant Pump Seal Evaluations,"

are the result of generic deficiencies that would require Brookhaven National laboratory, corrective action.

BNL-NUREG-52095, August 1987.

NUREG/CR-5008, R. D. Meininger and T. J. Weir, This report summarizes the findings on Westing-

" Development of a Testing and Analysis Methodol.

house reactor coolant pump (RCP) seal perfonnance ogy to Determine the Functional Condition of Sole-at Indian Point 2. his study considered a significant noid Operated Valves," Pentek, Inc., Coraopolis, number of RCP seal failures. Consolidated Edison in.

PA, September 1987, itiated a research effort to determine the causes of The objective of this research was to develop a sim-these failures and to develop appropriate ameliorative plc, reliable, condition-monitoring system that will action to enhance seal reliability. nc BNL work is an provide surveillance mformation without requmng dis-outgrowth of the first-phase effort performed by Fail.

connection or disassembly of solenoid. operated valves ure Analysis Associates. The objectives of the BNL (SOVs) installed in operating nuclear power plants, program are to determine the root causes of seat fail.

He information provided must be sufficiently reliable ure and to provide recommendations for improving to allow plant operators to conclude that valve per-l seal reliability.This program made notable advances in formance has or has not degraded to the point where understanding the root causes of RCP seal failure. For corrective maintenance becomes necessary, the first time, actual failed seals were examined in de-He required information is assumed to be obtain-tail in BNI.!s hot cell, and laboratory tests were con-able through analysis of in-rush current to the coil of ducted to determine failure causes. This report sum.

the SOV. Vanous SOVs were tested in an experimen.

marizes findings and presents conclusions and recom.

tal air system set up in the laboratory. In-rush current mendations based on review of plant operating and data acquired on degraded and new SOVs were ana-maintenance data, consultation with Westinghouse lyzed to determine behavior signature models.

and utilities, review of prior RCP seal studies (includ-Laboratory conditions provided the opportunity to ing previous BNL work), and visual and in-depth ex-simulate perturbations caused by the valve function, aminations of the first batch of service-exposed seals which would differ from actuation to actuation. A vis-received from the plant.

ual examination of this time.'

5g waveform re.

NUREG-1377 22

hiain Citations and Summaries vealed distinct and repeatable variations for dtfferent NUREG/CR-5052, J. C. liiggins, R. Lofaro, hi. Subudhi, R. Fullwood, and J.11. Taylor, v0ve anomalies,

" Operating Experience and Aging Assessment of This technique could identify gross changes and Component Cooling Water Systems in Pressurized render characteristic signatures that could be used for Water Reactors," lirookhaven National laboratory, various comparisons and to trend valve degradation 13NL-NUR11G-52117, July 1988.

mechanisms and their consequences over time.

An aging assessment of component cooling water (CCW) systems in PWRs was performed as part of the Utilization of the laboratory technique in an oper-NPAR program.The objectives were to provide a tech-ating nuclear power plant would be somewhat imprac.

nical basts for the identtfication and evaluation of deg-tical since the installed valves are not equipped with radation caused by age.The mformation generated will synchronous switching capability. Analytical research be used to assess the impact of aging on plant safety was therefore conducted to develop a technique to and to develop effective mitigating actions for the analyze similar electrical data obtained under asyn.

CCW system. The effect of time on this system was chronous conditions typical of an operating plant. For characterized by using the " Aging and Life Extension such field application l the technique developed would Assessment Program ( AlliAP) Systems Level Plan",

use a clip-on current probe, thus enabling all measure -

developed by Brookhaven National laboratory. Fail-ments to be made from outside the reactor building ure data from various nationti data ba;es were sc-without disturbing any electrical connections. The in, viewed and analyzed to identify predominant failure rush current to the solenoid-operated valve is analyzed modes, causes, and mechanisms in CCW systems, in real time using a personal computer and fast Fourier

' lime-dependent failure rates for major components transform techniques.

were calculated to identify aging trends. Plant-specific data were obtained and evaluated to supplement data NUREG/CR-5051, W li. Gunther, R. I ewis, and base results.

hl. Subudhi, " Detecting and hiitigating Battery A computer program (PRAAGE) was developed Charger and Inverter Aging," Brookhaven National and implemented to model a typical CCW system de-laboratory, BNL-NURiiG-52108, August 1988.

sign and perform probabilistic risk assessment (PR A)

This report is the second on the two-step approach calculations. Time-dependent failure rates were input for assessing the safety and operational aspects of bat-to the program to evaluate the effects of aging on the tery charger and taverter agmg m nuclear power plants-importance of a component with respect to system un-Analyses include an assessment of the recent operating availability. Time-dependent changes in component experiences with battery chargers and inverters and a importance and system unavailability with age were ob-discussion of improvements in reliability that may be served and discussed.

achieved through modification of the equipment's con-NUREG/CR-5053, W. Shier and bl. Subudhi, "Operat-figuration and an increased inspection frequency.The ing Experience and Aging Assessment of hiotor results are evaluated from a survey of the current Control Centers," llrookhaven National laboratory, maintenance and test practices used in nuclear power IINL-NUREG-52118, July 1988, plants, along with the manufacturerY recommenda-As part of the NRC NPAR program, an assessment tions for maintaining equipment operability. Advanced was made of the characteristics of aging and service designs for uninterruptible power systems, subcomp.

wear of motor control centers (h1CCs). htCCs perform onent improvements, and current monitoring and pro-an important function in the operation and control of a tective equipment are described and related to their large number of safety-related motors; thus the oper.

potential applicability in nuc! car power punts.

ability and rehability of h1CCs can affect the overall A naturally aged inverter aad battery charger were safety of nuclear plants, tested at l}N L to evaluate the naturally aged condition, This report follows the NPAR strategy and investi-the effectiveness of condition monitoring techniques, gates the operational performance, the design and and the practicality of selected maintenance and mont-rnanufacturing methods, and the current maintenance, toring procedures. A portion of this research effort is urveillance, and monitoring techniques APP ed to li covered m RIL No.159, Nuclear ilant Aging Re-h1CCs. A signif cant result described in this report con-scarch: Safety Related Inverters," November 9,1988.

cerns the identification of important h1CC failure A maintenance program for battery chargers and modes, causes, and mechanisms from plant opera-invertcrs is recommenJed. As described in this report, tional experience. Frequencies of failures determined such a program incorporates inspection, monitoring, for the various subcomponents of N1CCs are also de-testing. and repair actwities that should be performed scnbud. In addition, recommendations are provided for to detect and mitigate aging effects and thereby ensure functional indicators to monitor the performance of the operational readinesc of this important equipment h1CCs. These functional indicators will be evaluated throughout the plant's operating life.

during Phase 2 of the program.

)

23 NUREG-1377

IMain Citations and Summaries NUREG/ Cit-5057, K.11. lioopingarner and F. R.

that the operatmg parameters listed in the report are Zaloudek, " Aging Mitigation and Improved Pub within their inaximmn and minimum limits as applica-grams for Nuclear Service Diesel Generators,"

Pacifi': Northwest Laboratory, PNlr6397, ble. Ilowever, it is not necessary to trend every pa-December 1989.

rameter for effective results. Whn a limiting (maxi-rnum or minimum) value is being appeached, the utili-The study of diesel generator aging for the NRC ties should trend the approach to avoid failures and NPAll program was performed in two phases. In schedule repair before limits are exceeded.

Phase 1, plant operating experience and data were used Several recommendations were developed regard-to produce a new data base related to aging, reliability, ing maintenance procedures and training. One impor-and operational readiness of nuclear service diesel tant recornmendition is that tea-down of the diesel en-generators. Phase 11 is chiefly concerned with measures for mitigating the effects of aging.

gines solely for the purpose of inspection should be avoided unless there is a definite indication that opera-His report proposes a detailed management, test.

tion is degraded or there is an impending component ing, and maintenance program for emergency diesel failure based on performance data trends. He current generators based on studies and research developed in Practical periodic intrusive maintenance and engine Phase 11 of this effort. The proposed program would overhauls has been found to be less favorable for en-lead to three expected results:(1) reduction of several suring safety than engine overhauls based on monitor-of the stressors identtfied in Phase I that have been ing and trending results or on a need to correct specific shown to accelerate aging of diesel generators, (2) an engine defects. Herefore, this report recommends improved reliability and state of operational readiness, that the periodic overhaul requirements be reeval-and (3) an increased confidence in the future availabil-usted. Further, an understanding of the governor, as ity and reliability of diesel generators. He proposed well as of the engine / generator, must be developed by new program would integrate testing, inspection, providing the maintenance staff with adequate training

- monitoring, trending, maintenance, and other ele.

and motivation. Finally, this report recommends that ments for a better approach to mitigating diesel gen-engine inspections and preventive maintenance be in-eratoraging.The more important elements of the new creased to mitigate the aging and wear results of the vi.

proposed program are summarized in the following bration stressor, focusing on the engine and instrumen-paragraphs.

tation mountcd on the engine. Vibration cannot be The current fast starting and loading requirement climinated, but its effects can be mitigated by keeping for testing diesel generators can produce substantial fasteners / fittings tight and by fre;quently recalibrating harm and significant aging effects through the produc-Instrumentation subject to thts vibration.

tion of large mechanical and thermal stresses, inade.

He mission profile for th diesel generator is based quate lubrication during initial acceleration, high rotat.

or; a large-break LOCA with loss of all offsite power.

ing and - sliding pressures, overspeeding, etc. An With over 1000 reactor-years of operation in U.S. regu-improved testinh program including slow starting and latory history without r. large break 1.OCA, it may be loading would induce fewer aging effects in the emer, appropriate to redefine the~ mission profile for the gency diesel generator by largely eliminating a unique diesel. generator with consequent benefits. For a loss of offsite power, with or wi hout a small-break LOCA aging stressor. In the course of the monthly testing pro.

t gram, adequate data should be collected for about 30 cvent, the needs for emergency electric power and the engine operating parametera discussed in this report diesel mission profile are much less stringent. In this that could indicate degrading performance or an im, case, the need for power can be delayed and the emer-pending component failure. For many important com.

gency power need, are reduced,but the need for emer-ponents, the implementation of such a program could gency power may remain for several (3 to 4) days. ne detect approaching - performance failure and prevention of station blackout appears to be the most allow orderly repair, Monitoring and trending will not realistic mission envelope.The technical requirements be able to detect all components-with degraded for the diesel generator are very high reliability with performance, but the deterioration that will be de.

the durability to produce power until the emergency tected by the recommended tests is significant to aging Passes and the reactor cooling requirements drop off.

and reliability concerns, Condition monitoring and Acceptance of this mission envelope for the diesel-trending can provide important indications of possible generator system would result in a reduction of the ag-long-term component or system degradation. His ac-ing degradation of many important engine components tivity should detect many potential component / system through less harmful test requirements. In summary, a failures before the system actually fails. Cost and safety more practical mission envelope for the diesel genera-benefits would accrue from avoiding both equipment tor system would include an increased start and load damage and unscheduled downtime by anticipating time (within 5 minutes), with the power level reduced these failures and providing timely repair /

below the calculated full load (core and containment maintenance.ne monthly test program should ensure sprays not needed). From an overall mission. stand-l l

NUREG-1377 24

Main Citations and Summaries point, it appears that safety concerns are better served aging. Seal deterioration in the Valcor SOVs caused

]

by testing the engines for reliability rather than for leakage following DilE irradiationflhe naturally aged maximum starting accelerations and very rapid loading, Valcor SOV performed satisfactorily during the first high-temperature portion of the MSLil/LOCA profile which do not seem necessary, but malfunctioned during most of the rest of the test.

This portion of the NPAR study was initiated to de.

Deteriorationof the clastomericpartsof the ASCO velop for NRC consideration information on potential SOVs did not appear to be sufficient to account for the safety problems related to the aging of diesel genera.

observed failures to transfer, which evidently were tors. General applications of the study results were ex-caused by coil deterioration. Elastomeric parts of Val-pected for (1) improvement of diesel reliability, (2) cor SOVs, both from the naturally aged SOV and from modification of plant technical specifications, (3)im-the one that had not been aged, experienced substan-provement in the application of resources by the NRC and the utilitics, and (4) development of specific re-tial deterioration.

scarch information needed to change some regulatory NUREG/CR-5159, M. S. Kalsi, C. L llorst, and J. K.

requirements. All of these end uses of the research Wang," Prediction of Check Valve Performance and have been accomplished or are under active considera-Degradation in Nuclear Power Plant Systems," Kalsi tion. Collectively, the safety implications of these Engineenng, Inc., Sugar land, TX, Kill No.1559, changes and research recommendatiors are important.

May 1988.

Degradation and failure of swing check valves and NUREG/CR-5141, V. P. Itacanskas. G. J. Toman, and resulting damage to plant equ'pment has led to a need S. P. Caifagno, " Aging and Quahficatwn Research to develop a method to predict performance and degra-on SolenoiJ Operated Valves,"12ranklin Research dation of these valves in nuclear power plant systems.

Center, Norristown, PA, August 1988, This Phase 1 investigat bn developed methods that can Tests were conducted on three-way direct-actina be used to predict the stability of the check valve disk solenoid. operated valves (SOVs). Some SOVs I, when there are flow disturbances such as elbows, been aged naturally through service m nuclear powcz reducers, and generalized turbulence sources within 10 plants, and others were subjected to 1ccelerated aging.

pipe diameters upstream of the valve. Major findings Thermal agmg was conducted with both air and mtro-include the flow velocity required to achieve a full-gen as the process gas. Operational aging was simu-open stable disk position, the magnitude of disk motion lated by putting the specimens through operational cy-developed with these upstream disturbances (with flow cles at certain intervals during the accelerated thermal vel cities below full-open conditions), and disk natural aging with the environmental temperature controlled frequency data that can be used to predict wear and la-at a level representative of service conditions.The pro-tigue damage. Reducers were found to cause litt'.c or gram also included simulation of a design basis event no perf rmance degradation. Effects of elbowslocated (DBE) that consisted of gamma irradiation within 5 diameters of the check valve must be consid-and a main-steam line-break loss of-coolant accident cred, while severe turbulence sources have a signift-(MSLB/ LOC A) simulation. After each major segment cant effect at distances up to 10 diameters upstream of of the test program (aging, uradiation, and MSLB/

We vahm I OC A simulation), some of the valve specimens were Clearway swing check designs were found to be par-subjected to operational testing and then disassembled ticularly sensitive to manufacturing tolerances and in-for inspection and measurement of physical properties.

stallation variables making them hkely candidates for Performance of the Automatic Switch Co. (ASCO) prem ture failure. Reducmg the disk full-opening an-SOVs was aflected in the early stages of the program by gle on these designs results m sigmficant performance an organic deposit of undetermined origin. Removal of tmpmment the deposit climinated the problem.

NUREG/CR-5181, L C. Meyer and J. L Edson, A naturally aged ASCO SOV with Duna N seals

" Nuclear Plant Agmg Research:The IE Power cnd a new ASCO SOV with EPDM seals were sub-a al Engineering Laboratory, jected to accelerated aging with mtrogen as the process

$t m 5 gas.These valves were the only ones to go through the This in-depth engineering study of the Class IE entire test program without a failure to transfer and Power System is conducted in accordance with the without any significant leakage.

NRC NPAR program and guidelines.The report pro-Valcor Engineering Co. SOVs suffered from stick-vides (1) an identification of failure modes, (2) a pre-ing of the shaf t seal 0-rings, which made it impossible liminary identification of failure causes due to aging to complete the accelerated thermal aging. Repeated and service wear degradation, and (3) a review of cut-tests and ct inges in test procedures failed to alter this rent inspection, surveillance, and monitoring methods, includmg manufacturer recommended surveillance situation.

it is possible that the stresses of accelerated aging and maintenance practices Also, performance pa.

produced effects that are not representative of service rameters potentially useful in detecting degradation NUREG-1377 25

s

Main Citations and Summaries are identified in this report, and preliminary recom-high on the list because of frequent preventive main.

mendations are made regarding inspection, surveil-lance, and rnonitoring methods.

tenance tasks such as adding water and testing.)

The review of codes and standards included general A description of a typical Class IE power system is design criteria, regulatory guides, and IEEE standan ds.

presented for a pressurized water reactor (PWR) with there were three recommendations for regulatory specific maintenance information from a cooperating guides: (1) Regulatory Guide 1.118 should include the utility. The Class 1E power systems provide electric tssues f testing and inspection for the lightning pro-power for the safety systems in' the plant, including an tection system and power ground system, (2) Regula-emergency power source (usually dicsci generators)

} ry Guide 1.32 should address the issues of cleanliness and three subsystems: the alternating current (ac) in switchgear area, and (3) Regulatory Guide 1.9 power systems, the direct current (de) power system, should be extended to include the problems of dicsci and the vital ac power system. Each of the major Class genemtor aging.

IE power components is described, and the results of Approximaiely 40 IEEE standards applicable to component aging studies are summarized where appil.

Class IE power systems and associated cornponents cab 1::. De ac power rystem used in typical nuclear were reviewed and tabulated. The IEEE reviews each power plants is a dual-train cascading bus system that standard approximat ely every 5 years. ne authors rec-includes circuit breakers, transformers, relays, load ommend that aging be included in this review. Stan-centers, and rnotor control center switch gear. The dards provide design and application guidance but gen-de system includes battery chargers, batteries, inver, emHy do not provide specific recommendations for ters, and associated control breakers. He vital rnamtenance, testmg, mservice mspection, and mom-120lVac loads include the engineered safety feature toring of age-related degradation, cabinets and the reactor protection systems.

Aging research can play a supporting role in solving De review of operating experience included data outstanding safety issues. For example, component from the following generic data bases: Licensee Event degradation due to aging is one factor to consider in the Reports (LERs), Nuclear Plant Reliability Data Sys-plant coping analysis required by the NRC rule on sta-tion blackout.

tem (NPR DS), Nuclear Power Experience (NPE), and plant maintenance data from one cooperating utility.

NUREG/CR-S192, W. E. Gunther, " Testing of a the LER records indicate that the Class 1E power sub' Naturally Aged Nuclear Power Plant inverter and system failures were distributed as follows: emergency Battery Charger," llrookhaven National laboratory, power genemtion, 31.7%; medium voltage subsys' HNI.-NUREG-52158 September 1988.

tems, 21.2%; low-voltage ac (less than 600 V),19.8%;

A naturally aged inverter and battery charger ob.

and de system,9.8E ne most frequent component tained from the Shippingport facility were tested as failures were ctrcuit breakers,66.3%; mverters,9.9%;

part of the NPAR program, The objectives of this test-and batteries, 9.5% The leading causes of circuit ing were to evaluate the naturally aged equipment state, determine the effectiveness of condition-breaker faults were mechamcal malfunction, 25%;

electrical malfunction, 22%; and sticking, 7% The monitoring recommendations, and obtain insight into three leading causes of relay faults were drift,46%;

the practicality of preventive maintenance and moni.

electrical malfunction,11%; and sticking,10E ne toring methods.

NPRDS data review listed the Class 1E power compo-Testing indicates that the equipment has retained nents m oract of frequercy of failure as follows: diesel ts ability to respond toload transients. With the excep-engines, inverters, and circuit breakers. The overall tion of silicon controlled rectifiers (SCRs), which were fraction of Class lE electrical component failures re-found to be operating with case temperatures (*F) lated to aging was 32.7% However, because of system 20% higher than those during the acceptance test, redundancy and fail-safe design, only 2.4% of Class 1E component temperatures and circuit characteristics electrical component failures caused total loss of sys-were similar to original acceptance test measurements.

tem fustmn.

Based on these observations,it is concluded that the in.

verter and battery charger have not aged substantially.

Approximately 8% of all events in the NPE data base for all systems were associated with-the safety The two primary monitoring techniques employed were temperature measurements and electrical electrical system.The NPE listed breakers, motor con-waveform observation. Internal panel temperature trol centers, and switchgear as having the most fre-and individual component temperatures were re-quent failures, at 36.1 % This was followed by inverters corded at regular intervals during steady. state and and chargers,15%; diesel generators,10%; transform-transient operations. Thermocouples imbedded within ers, 3.4%; and batteries, 3E De plant data also the transformer and inductor windings and attached to showed that breakers caused the most work requests in SCR and capacitor surfaces provided a nonobtrusive the maimenance data base, ' allowed by batteries, the battery charger, and the generater. (Batteries were means of monitoring component operation. Readings taken were compared to original acceptance test data.

1 NUREG-1377 26

Main Citations and Summaries Circuit waveforms were observed on an hourly basis

4. Evaluate the importance of the aging ofindivid-during steady-state operation and at the time load tran-ual components and component groups on plant sients were applied. The inverter output voltage and

risk, the SCR gate current waveforms remained relatively

5. Applythe"RiskSignificanceof Compment Ag-constant regardless of the applied loads.

ing" methodology (being developed by W. E.

Finally, this test report recommends that individual Vesely of SAIC under the NPAR program) to fusing of filter capacitors be considered in order to pre-the prioritization, clude a capacitor failure in the short circuit mode from

6. Use operational failure data, rendering the inverter inoperable. Also, equipment ac-

7. Use expert judgment through an interdiscipli-cep'ance testing should be modified to obtain the most nary panel, limiting design operating conditions for all major sub-

8. Stress the importance of aging research to the componems. Results indicated that aging had not sub-resolution of generic safety issues and to user stantially affected equipment operation. On the other needs identified by the Office of Nuclear Reac-hand, the monitoring techniques employed were sensi-tor Regulation to aid NRC decision-makers but tive to changes in measurable component and equip-not to formally prioritize the components.

ment parameters. Thus comparing the monitoring re-NUREG/CR-5268, R. Lofaro, M. Subudhi, W. E.

sults with the original acceptance test data is a viable method of detecting degradation prior to catastrophic

,u he "d J.

t

actg, d

failure.

Heat i emoval System. Brookhaven Natior,al l2boratory, BNI.rNURIIG-52177, June 1989.

NUREG/CR-5248,1. S. Lo, D. B. Jarrell, and As part of ongoing efforts to understand and man.

E. P. Collins, "Prioritiza ion of TIRG A131X.

Recommended Components for Further Ac,ing age the effects of aging in nuclear power plants, an ag-Research," Pacific Northwest laboratory, $cience ing assessment of a vital system, the residual heat re-Applications International Corp., PNIA701, moval(RilR) system in boiling water reactors (BWRs),

November 1988, was performed. This report presents the results and in April 1986, the NRC established the Technical discusses the impact of RilR system aging on plant Integration Review Group for Aging and Life hten-safety. The work was performed as part of the NRC sion activities. In May 1987,TIRG AlliX finalized its NPAR program. The RilR study was done according plan (TIRG ALEX 1987), which identified the safety-to the methodology developed by BNL as part of the related structures and components that should be pri-Aging and life Extension Assessment Program oritized for subsequent evaluation in the NRC NPAR (AlliAP) System Level Plan. The sch cted approach program.This report documents the results of an ex-uses two parallel work paths, one applying determin-pert panel workshop established to perform the istic techniques and the other probabilistic techniques, prioritization activity. Prioritization was based primar-to characterize aging, ily on criteria derived from a specially developed risk-The deterministic work performed for the RHR n

based methodology that incorporates the effect on system study involved a review of past operating data plant risk of component aging and the effectiveness of from variou's national data bases. The data covered all current industry aging management practices in miti-operating rnodes of the R11R. They showed that ap-gating that aging.

proximately 70% of the failurer reported were due to An additional set of criteria was the importance of aging. The dominant cause of failure was found to be aging research on structures and components to the normal service, while the dominant failure mecha-resolution of generic safety issues and to identified nisms were wear and calibration drift. The predomi-regulatory needs. The resitant categorization was nant failure mode was leakage followed by loss of func-used to provide additional information to decision tion and wrong signal.The data also indicated that ap-makers but was not used to calculate final rankings.

proximately 65% of the failures were detected by the current test and inspection practices. Ilowever,27% of The expert panel workshop was conducted within the failures were not detected until an operational ab-the following ground rules:

n rm hty occurred. This shows that currently em-

1. Obtain all relevant information on aging of cut-phiyed maintenance and monitoring practices are not rent plants (i.e., during their original license completely successful in detecting all agmg degrada-puiod)'

tion. In evaluating the effect of failure on RilR per-

2. Develop an understanding of aging and its ef-formance, it was found that over 50% resulted in fccts (i.e., define the contribution of aging to degraded system operation, while approximately 20%

plant risk),

resulted in a loss of redundancy. Other significant ef-

3. Assess the adequacy of current industry prac-fccts of RHR failures include loss of shutdown cooling tices for managing component aging within ac-capability, radiological releases, reactor scrams, and ceptable levels of risk, actuation of engineered safety features. Actual plant 27 NUREG-1377

hiain Citations and Summaries records for hii11 stone Unit I were obtained and re-be significant. His will bc <tddressed in future viewed. The results showed consistency with data base work.

h"di"8

Data Analysis De probabilistic work entailed the imple'

l. Resuits have confirmed that generic failure -

mentat,on of a personal computer basco pro-i gram (PRAAGE-1988) developed to perform time-rates may not accurately represent individual dependent probabilistic risk assessment (PRA) plants for all applications. %c uncertainty in calculations. The RIIR model used was based on the risk estimates may be reduced by updating cal-Peach Bottom design. Time-dependent failure rates culations with actual plant data.

for maj,or components were developed from the data

2. Mechanical components in the R11R systern base findmgs and were used in the program to calculate show a low to moderate increase (8% to 17%

system availability and component importances for perycar)in failure rate with age, while electrical vanous ages. The PRA results showed that, when the components such as switches and sensors show time-dependent aging factors are accounted for, two little or no increase (0 to 3% per year).

significant system effects are scen: (1) system unavail.

Design Considerations ability increases moderately with age and (2) the rela-

1. Plants with a common suction line supplying all tive importances of components may change with age.

loops of the RilR while in the shutdown cooling For low-pressure coolant injection operation, miscali, mode should consider placing increased atten-bration of instrumentation was the most important tion on motcr-operated valves (MOVs) in the contributor to system unavailability. Ilowever, during later years, aging can cause motor operated valves to suction line daing later years of plant life since become equally important. PRA calculations for shut-aging can increae the probability of MOV fail-ute and lead to a temporary loss of shutdown down cooling operation showed these valves to be the cooling capability. Piping and other components most important contributors to unavailability through-in nontedundant supply lines should also be out plant life, considered.

The following conclusions resulted from this

2. Plants using a common minimum flow line for assessment:

two RiiR pumps should closely monitor pump Aging Effects performance since aging can degrade perform-ance and lead to dead-headed pump operation

1. Aging has a moderate impact on RIIR compo, and possible failure, nent failure rates (0 to 17% per year increase)

The findings presented in this report form a sound and system unavailability (2-fold to 4-fold in-technical basis for understanding and managing the ef-crease in 50 years). His contribution of aging fects of aging in RHR systems.The results also provide effects may be attributed to two factors: (1) the framework for future Phase !! work. Although the RHR is a safety system and has relatively strin, time-dependent aging effects appear to be moderate gent testing and monitoring requirements that for the RllR system, additional work is necessary to identify aging degradation before performance complete the aging assessment. Since this is predomi-is adversely affected and (2) the R11R system is nantly a standby system, exposure to operating stresses typically maintained in standby, which mini, is limited, which could contribute to the mitigation of mizes exposure to wear-related degradation.

aging effects, llowever, as plants continue to age and

2. Preliminan comparisons of unavailability for operating time increases, the RIIR system could expe-standby and continuously operating. systems rience rapid increases in failure rates, as was found in l

have shown that standby systems are potentially previous work on a continuously operating system. nis less severely affected by aging. Using this result should be addressed in future work. In addition, the i

as a basis, the differences in operation and man-relatively stringent tests and inspections performed for 1

agement of these two types of systems will be the RiiR system may contribute to the aging effects.-

fu-ther evaluated with the ultimate goal of de.

NUREG/CR -5280, M. Subudhi, W. Shier, and E.-Mac-

- veloping methods that are effective in mitigat.

Dougall,"A c-Related Degradation of Westtn-

-ing aging e77ects' ghouse 480- olt Circuit Breakers; Vol.1: Agmg-

3. Examination of plant-specific failure data has Assessment and Recommendations for improvmg Breaker Reliability," Brookhaven National labora-confirmed that failure trends for certain compo-tory, BNL-NUREG-52178, July 1990.

nents in some plants can differ from industry An aging assessment of the Westinghouse DS-se-averages. Although aging was found to have a ries low-voltage air circuit breakers (especially DS-206 moderate impact on the RllR system based on and DS-416) was performed as part of the NRC Nu-average values, the impact on plants for which clear Plant Aging Research (NPAR) program. These the data differ from these average values could breakers are used for Class IE applications in nuclear NUREG-1377 28

Main Citations and Summaries infonnation notices and bulletins pertaining to prob-power plants. DS-416 breakers, in particular, are used lems encountered in Class liibreakers. A review of op-for reactor trip applications. The fmdmgs from this erating experience suggested that burned.out cods, study form a technical basis for understanding aging ef-jammed operating mechanisms, and deteriorated con.

fects in DS-series breakers.

tacts were the dominant causes of failures. Although This study was mitiated following the failure of a failuresof the pole shaft weld were not included as one center pole lever weld in a reactor trip breaker at the of the genene problems, the NRC Augmented Inspec-McGuire Nuclear Station and the issuance of NRC tion Team had suspected that these welds were of sub-llulletin SS-01 on that subject. 'Ihe objectives of the standard quality, which could lead to their premature study are to characterite age-related degradation in the crackirg.

breaker assembly and to identify maintenance prac-This program involved a commercial grade Westin-tices to mitigate degradation elfccts.

ghouse DS-416 low voltage air circuit breaker that is

'Ihe design and operation of DS-206 and DS-416 typical of breakers used m nuclear power plants for breakers were reviewed in detail. Failure data from dan & appkanons. the test breaker was mecham,-

various operational data bases were analyzed (1) to enlly cycled for more than 36,000 full gcles with no identify all failure modes, causes, and mechanisms,(2) cicctncal load, thus accelerating the aging process that to assess the effectiveness of the requirements formu-could be attnbuted to brejiker cycles to help identify lated in NRC Hulletin 88-01, and (3) to recommend ac-age-related degradations. l'he test was conducted m tivities that would effectively detect and mitigate age-accordance with ANS!/ll!EE Standard 37.50 (1981)for related problems in breakers. The data bases included the hfe testing of circmt breakers.Threc phfferent pole Licensee Event Reports (1.ERs), Nuclear Plant Reli-snafts with weld configurations of approximately 60 de-ability Data System (NPRDS), in-Plant Reliability grees,120 degrees, and ISO degrees in the center-pole Data System (IPRDS), and Nuclear Power Experience lever (#3) were used to characterize crackmg m the (NPE). Additional operating experience data were ob-

@c legr welds. In addition, three operating mecha-tained from one nuclear station and two mdustrial nism umts and several other parts were replaced as breaker-service companies to develop aging trends for they became inoperable.

various subcomp(ments.The responses of the utdities The mechanical cychng test resulted in the follow-to NRC Halletin 8S-01 were analyzed to assess the fi.

ing conclusmns on the manufacturing and aging of nal resolution of failures of welds during reactor trips.

esunghouw Ewries breakers:

The predominant failure modes in nudear power 1.

Pole shafts used in this test program were plants along with the causes and mechanisms of failure found to have substandard welds. This raises were determined from the operating experience data.

questions as to the effectiveness of the quahty Instruction manuals including schematics and manu.

assurance program that was followed during facturers' maintenance manuals were analyzed to un.

welding.

derstand the effect of matenal aging during'the service life of the breakers. This analysis was augmented by 2.

Fracture of the top shaft lever suggested that technical discussions with maintenance and service correct electroplating procedures may not have been followed.

personnel from the electrical suppv industry. h1ainte-nance recommendations by the manufadurer to miti-3.

The sharper bendsat the neck of the hooks on gate age-related degradation, suggestions for improv-newly purchased reset springs-compared to ing the monitoring of age-related degradation, and m' an oider design-led to early spring failures.

~

puts from NRC inspectors involved in assessing break-L The hardness of the oscillator surface on er problems in the nuclear indusuy were reviewed.

r:cwly procured units was 30% less than on Volume 2 of this report presents the results irom a older units.

test program to assess degradation in breaker parts 5.

Wear, fracture, distortion, and normal fatigue through mechnical cy cling ihat simulated the operating kminated the aging procco, with wear bemg life of nuclear plant breakers.

the largest contributor.

NUREG/CR-5280, bl. Subudhi, E. hlacDougall, S.

6.

Excessive wear was evident m the ratchet Kochis, W. Wilhelm, and ll.S. Lee, " Age-Related whed, holding pawls, oscillator, drive plate, i

Degradation of Westinghouse 480Nolt Circuit motor crank and handie, cam segments, main Breakers; Vol. : Niccharucal Cychne of a DS-416 roller, and stop roller.

Hrcaker. Test Results," Brookhaven ' National I abo-7.

Structural components and contact assembly ratory, HNicNUREG-52178, November 1990.

pam showed few effects of agmg due to me-After N NicGmre event in 1%7 mvolving failure charucal gehng of the de weld in a reactor top breaker, the A pole shaf t with a reduced site weld could fail A.

NR('

m investigatmn of the probable causes.

at as icw as 3000 gdes.

Dunng inc Lt decade, NRC has mued a number of NURFG-1377 29

L i

I

' Main Citations and Summaries ne testing yielded many useful results. The sessment of cast stainless steel components, the main burned-out closing coils were found to be the result of binding in the linkages that are connected to this de-concern is loss of fracture toughness and impact en-ergy, Data and engineering models have been devel-vice. Among the seven welds on the pole shaft, #1 and oped to help predict the degree of embrittlement as a

1. 3 were the ones that cracked first and caused misalign.

function of thermal exposure history. For reactor inter-ment of the pole levers, which,in turn, led to many nals, irradiation history may also be a concern.

- problems with the operating mechanism, including burned-out coils, excessive wear in certain parts, and De minimum CVN impact er ergy after long-term overstressed linkages Based on these findings, a main-aging has been found to be proportional to the square tenance program designed to alleviate the age-related of the fraction of ferrite, the mean ferrite spacing, and degradations caused by mechanically cycling this type a chemical composition parameter, This model should of breaker is suggested.

be developed further for application to the assessment of components. A time. temperature parameter can be NUREG/CR-5314, C.E. Jaske and V.N. Shah, " life used to define lower bound trends to the available im-pact cricrgy values for cast stainless steels as a function Assessment Procedures for Major LWR Compt,,

of chemical composition and thermal exposure time, nents; Vol. 3, Cast Stainless Steel Components Idaho National Engineering laboratory,

.l'he report proposes a model using that parameter to EGG-2562. October 1990.

predict the impact energy decrease for any particular Many critical pressure boundary components m.

lot of cast stainless steel. This predicted impact energy commercial light water cactors (LWRs) are made of value or the predicted minimum impact energy value is cast stainless steels. Life assessment procedures are then used to estimate itacture toughness from correla-needed for these components because cast stainless tions between impact energy and fracture toughness at steels are subject to therrnal embrittlement during both room temp ature and 290*C. This approach long-term semcc at LWR temperatures. He compo-should provide a conservative estimate of fracture nents of concern mclude pump bodies, reactor coolant toughness for use in assessing the structural integrity of piping and fittings, surge imes (m a few plants), pres-cast stainless steel components, surizer spray heads, check valves, control rod drive

, inservice inspection (ISI)is needed to define type, mechanism housings, and control rod assembly hous-size, and location of any defects m cast stainless steel ings. These are made of grade CF-8, CF-8A, or components so that their structural integrity can be CF-8M stainless steel in U.S. LWRs; grade CF-3 ev lu ted. Use of mdiography during ISIis less practt-stainless steel is used in some foreign LWRs. The pur-cal than during fabrication. Conventional ultrasome pose of this project was to review the available data on testing (UT) methods for detecting flaws are not relt-therrnal embrittlement of cast stainless steels and to able in cast stainless steel components because its develop updated procedures for life assessment by key c arse gram structures result in a low signal-to-noise LWR cast stainless steel components.

ratio. Advanced UT methods being developed have Cast stainless steels have a two-phase microstruc-shown an improved capability t.o detect flaws in cast ture consisting of ferrite islands in an austenite matrix, stainless steel components and have been used in sev-

- With long-term egosure to LWR temperatures, other eral PWR plants. Becanse of Ihe difficulties with radi-phases form in the ferrite phase that cause it to become ography and UT methods in detecting and sizing flaws, hard and brittle, while the austenite remains ductile. If the application of the acoustic emission technique to the amount of ferrite is small and if it is distributed detecting crack growth in cast stainless steel needs to be evaluated, evenlyand finely throughout the austenite, the ptoper-ties of the casting are not significantly affected by the The report outlines a procedure developed for esti-thermal embrittlement of the ferrite. However, as the mating the current condition and residuallife of key amount of ferrite, its coarseness, and its uneven distn-LWR cast stainless steel components. The procedure bution increase, the increased thermal embrittlement s implemented in nine major steps. %c first three of the ferrite adversely affects the properties of the steps involve the collection, examination, and storage C^8I'"8' of records for fabrication and construction, inservice inspection, and operating history. The fourth step in-

' he properties most affected by thermal embrittle-volves a conservative fatigue and fracture mechanics ment are Charpy V-notch (CVN) impact energy and evaluation to determine the worst-case flaw size and fracture toughness (b). Both of these properties de-the minimum required fracture toughness at the end of crease as the degree of thermal embrittlement in-the next operating period. In the fif th step, the current creases. lf these values become too low, the structural condition of the matenal is assessed using a proposed integrity of a cast stainless steel componerit could be analytical model, microstructural data, or measured seriously impaired. Presently, more _ fatigue-crack-properties (or some combination of the three). In the growth data are needed for CF-8 and for all cast stain-sixth step, the results of the fourth and fifth steps are less steels in the high-cycle regime. Thus, for life as-combined to evaluate the structural integrity of the NUREG-1377 30 k

c

Main Citations and Summaries of the seal materials.De exceptionalleak integrity of component.ne seventh step establishes what actions the three designs tested in this progmm should not be (none, repair, replace, or shut down) are to be taken, assumed to apply to all other designs in use for at least and the eighth step establishes the plan for the next ISI. In the ninth step, the component is reevaluated two reasons:

and the steps are repeated as needed.

1. There are a large number of diverse designs in use. In particular, assemblies manufactured NUREG/CR-5334, D. B. Clauss, " Severe Accident prior to 1971 were not subject to national stan.

'l esting of Electrical Penetrat,on Assemblies',

dards and were of ten manufactured in the field, i

Sandia National Laboratories, SANDS 94)327 whereas the three tested in this program were November 1989 subject to rigorous quality assurance and were Since the Three Mile Island. incident, the risk and designed to meet the standards of IEEE consequences of severe accidents have been a major 317-1976 and IEEE 323-1974.

focus of reactor safety research.The performance of 1 He leak potential is highly dependent on the the containment building has a significant effect on ac-temperatures to which the assembly is sub-cident consequence; thus considerable effort has been jected. As research continues and more analyses directed toward understanding and predicting the of severe accident sequences are conducted, the functional failure of containments. He containment

" worst-case" loads may change. Therefore, the pressure boundaty typically mcludes numerous me-leakage potential must be reevaluated as the chamcal and electrical penetrations, each of which rep-understanding of severe accident loads is im-resents a potential leakage path, proved. Ileat transfer effects must be consid-Several studies completed in the early 1980s indi-cred to determine the temperature of the out-cated that electrical penetration assemblics could be board containment seals, which end up control-an important leak path that merited further study. A I ng the potential for leakage, report by the Oak Ridge National 1.aboratory on se-he results of these tests should not be construed as vere accident sequence analysis for BWR Mark I con-suggesting that all designs will not leak under severe tainments concluded that the temperatures m the accident conditions; the performance of all compo-drywell were high enough to possibly cause failure of nents of the containment pressure boundary must be the seals that could result in leakage. NUREG-0772 evaluated on a case-by-case basis with all loads consid-identified electrical penetration assemblies as having cred. ne performance is also affected by thermal and "one of the largest uncertainties associated with pre-radiation aging. Given good information on the con-dicting the amount of radionuclides released. These tainment loads, a heat transfer analysis to determine studies provided the major impetus for NRC to uutiate the approximate temperature profiles, knowledge of a research program on these assemblies. Sandia Na-the time temperature threshokis for the scalant mate-tional laboratories managed a program to conduct a rials used, and the proper exercise of engineering judg-background study and to recommend and perform tests ment, a reasonable evaluation of the leakage potential to generate data that could be used to assess the Icak of other designs can be made, potenttal when the assemblies are subjected to severe The electrical performance of the assemblics was accident cenditions. These tests are described in this monitored in these tests by measunng the insulation report.

resistance and electrical continuity of the conductors.

Electrical penetration assemblies are used to pr*

The resistance degraded rapidly during the severe acci-vide a leak-tight pass-through in nuclear power plant dent tests, although the rate depended more on the containment buildings for cicctrical cables with power, type of cable and loads than on the particular design control, and instrumentation applications. The design being tested. Under the specific severe accident condi-has evolved to a modular concept that consists of elec-tions that were simulated, the data suggest that all elec-trical conductors coatained within stamless steel tubes trical systems supplied in the Westinghouse assembly (modules) that are sealed, would have functioned for about 4 days; those supplied in the D. G. O'Brien would have functioned for about Three designs, D. G. O'Brien, Conax, and Westing-13 hours: and those supplied in the Conax may have house, were tested under sirnulated severe accident functioned for only about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Some cables would conditions for a PWR, a BWR Mark I drywell and a be cypected to function beyond these times. However, BWR Mark Ill wetwell, respectively, to generate engi-it must be noted that conclusions regarding the electri-neering data (leak rate, temperature. insulation resis-cal performance of systems inside the containment tance, and electrical continuity) for assessing their leak builJing based solely oninsulation resistance data must potential. None of the assembiics leaked during the se-be made with caution. The performance of the electri-vere accident 1:sts, which can be attributed to the use cal systems would depend on the voltage, currat, and of redundant seals and to the fact that the outboard impedance requirements for a specific conductor ap-

{

containment seah in all three designs were never ex-posed to temperatures that exceeded the service limits plication.

NUREG-1377 31

Main Citations and Summaries NUREG/CR-5379, D. B. Jarrell, A. B. Johnson, Jr.,

4. Perfonn a fault tree analysis of the senice P W Zimmerrnan, and M. L Gore, " Nuclear Plant Senice Water System Agmg Degradation Assess-water system of a typical plant to examine ment: Phase 1. Vol.1. Pacific Northwest Labora" faihire propagation and determine specific in-tory, PNL-6560, June 1989.

put requiremeMs of probabilistic risk analyses, The service water system represents the fint.1 heat

5. Develop an in depth questionnaire protocol for transfer loop between decay heat generated in the nu*

examining the information resources at a plant clear core and the safe dispersal of that heat energy where such resources are not available in the mto the environment.The objective of this assessment standard data bases. Subsequently, visit a nu-ts to demonstrate that aging phenomena m the senice clear power plant and solicit the required infor-mation..

water system can be identified and quantified so that aging degradation of system components can be de.

6. Analyre the information obtained frorn the in-tected and mitigated before the system availability is depth plant interrogation and draw contrasts reduced below an acceptable threshold. The following and conclusions in regard to the data base, goals of the assessment were directly derived from the NRC NPAR program plan:

7. Use the plant information to perform an interim assescment of degradation mechanisms and to

1. To identify the principal aging. degradation focus future investigations, mechanisms, to assess their impact on opera-The following is a summary of the conclusions of tional readiness, and to provide a methodology the assessment to date:

for mitigating the effects of senice water system agmg on nuclear plant safety,

1. Aging-related degadation of open senice water systeras, i.e., systems that have a direct

2. To examine the current surveillance specifica-interface to raw water without chemical control, tie s and evaluate their ability to provide accu-in nucicar plants is prevalent and constitutes a rate reliability information.

valid safety concern. Based on actual plant data,

3. To provide a means to evaluate the effective-E

"'I

1. 8" ness 'of maintenance on mitigatin8 a&n8-tk open pms is cope compoW h i

degradation phenomena, the accumulation of biologic and inorganic ma-terial. This conclusion directly contradicts the

4. To produce an inspection plan that optimizes results of a failure analysis performed using in-the effectiveness of inspections based on system formation obtained from the NPRD3 data base, risk reduction, which indicated that the torque switches of mo-

5. To use the information generated by this assess-tor operated valves were the prime cause of ment to resolve related genene tssues and pro-system failure, vide guidance for developing regulatory criteria

2. Based on multiple plant samplings, the current on aging and life extension, level of surveillance and posimaintenance test-The following approach was used duriog the initial ing performed on the system is not sufficient to i

phase of the assessment:-

accurately trend or detect sptem degradation due to aging phenomena.

1. Perform a literature search of government and private sector reports that are related to service

3. While postmaintenance surveillance doca give some mea:ure of the effectiveness of systent water, aging related degradation, and potential modification and repair efforts, sufficient infor-t L

methodologies for analysis.

mation on monitoring operational condition and

1. Assembic a data base that conmins a listing of Postmaintenance testing is not available to char-the configurations, characteristics, and water actenze more precisely the effectiveness of sources for the senice water systems in all com-mamtenance.

mercial nuclear power plants in the U.S.

4. To improve the accuracy of data to a point that

3.. Obtain and examine the available senice water would allow a high degree of confidence in the -

data from large generic data bases, i.e., the Nu-analysis of aging degradation, a root cause logic clear Plant Reliability Data System. Licensee scheme needs to be developed for use m defin-Event Reports, Nuclear Power Experience, in-mg the depth of knowiedge and the documen-spection reports, and other relevant plant refer, tation required to accurately characterize an ence data. Analyze the service water system of a agmg-related component failure.

specific power plant for aging-related degrada-

5. Clear resolution of relevant aging related safety i

tion phenomena from the avadable data ob-issues will require the specification of additional tained from this data base.

documentation of failure data and regulatory NUREG-1371 32

hiain Citations and Summaries i

test, noise analysis, and twer interrupt test.

requiremer.ts to ensure adequate safety mart n Two of the five methoos (noise analysis and under a ged or extended hfe conditions.

power inte trupt test)have the advantage of pro-NUiti:G/ Cit-531i3 11. M. liashemian. K. M. Petersen*

viding on line measurement capability at nor-

11. B. IMn, and J. J. Gingnch, 'T.ffect of Ahsing on

• "I OPCf"N"8 C""diIi"'

itesponse Time of Nuclear Plant hessure ensors."

Analysis and Measurement Sernes Corpotution,

2. 'the consequences of aging at simulated plant mnditions wer e calibrstion shifts and respom.e-Knoxville, TN, June 1989.

A research program was initiated to study thc ef.

time d gradation, the former being the more fccts of normal aging en the dynamic perforrnance of ptonoutaed problem, safety related pressure transmitters (i.e., sensors) in

3. 'the 1.l!R data base contains 1,325 cases of re-nuclear power nlants.1he project began with an ex-ported problems with pressure sensing systems ptrimental asser.sment of the conventional and new over a nine year period (19801988). potential testing metic3s for measuring the response time of age related cases account for 38% of the re-ported problems in this period. A notable num-pressur e transmittersflhis was followed by developing her of IJ!Rs r eported problems with bhickages, a laboratory setup and performing initial tests to study the aging characteristks of represmtative transmitters freeting, and void (bubble) formation in sensing of the type used in nuclear powe plants-lines.

'lhere is need to ensute tha the current testing

4. llegulatory Guide 1.118,11!!!!! Standard 338, tnethods, regulatory requirernen 4 and industry stan-and ISA f,tandard 67.06 can benefit f om minor dards and practi.es are adequatt to track age rel$ited recommended revisions to account for recent degradation.'the project examined the validity of the advances in performance testmg technologies 8vadsde methods for terponse time tetsting of pres-and from new information that has become sure transmitters and reviewed the historical data for availabic since these documents were initially evidence of performance degradation problems or renerated.

trends. Current intervals for response tirne testing and

.lhe sivmonth study of the dynamic pet formar.cc of calibrating pressure transmitters are based on refuel

• pressure transmitterta covered the foltowing alcas:

ing n nedules, appatently for two reasons:

gg

i. *lhere is nomethod available for on line calibra' An experimental assessment of the five meth-tion of pressure transmitters, and, until re-ods mentioned above involved laboratory test-cently, response time testing could oot be pu-ing of more than twenty pressure transmitters f ormed on line.

with all five methods. Resuhs showed thht the 4 1he available data base of degradation rates and methods are equally effective but vary widely in trends is not sufficiently f eliable to justify test-difficulty of implementation in nuclear power ing intervals longer than one refueling cycle-plants. Two of the five methods (noise analysis While testing based on refueling intervals may be and power interrupt test)can be performed re-adequate, th:re is concern that the rate of degradation rnotely on installed transmitters while the plant is at norrmil operating conditions.

of pressure transtaitter perfortnance may inercare as the current generation of plants becomes older, Fur-2, flgig Study. l2boratory research on aging was thermore, evline testing methods bued on new tech-initiated and preliminary results were obtained.

nologies are becoming available to permit more fre-1hc work involved response-time testing and quent testing of transmitters end to predict incipient mhbration checks of a number of transmitters Iallut es. These considerations have motivated research after exposure to heat, humidity, vibmtion, such as that covered in this report to ensure that practi-pressure, cycling, and overpressurization condi-cal test methods and adequatr. test schedules are used tions.1he effect of these conditions was an in-to verify proper and time'4 performance of safety

• crease in response time and calibration shifts, system pressure transmitters m nuclear power plants-the latter being the more pronounced problem.

the project included a search of the licensee event 3, nn.in, of Related Studies All published experi-report (1.ER) data base for pressure-sensing system mental work on aging of pressure transmitters problems and reviews of Regulatory Guide 1.118 and has concentrated on the effects of aging on of the industry candards on performance testing of static performance of the transmitters as op-pressure trans.nitters.1he following concludons have posed to the dynamic performance reported hereinJihe related studies concluded that aging been reached;

1. Five reasonably effective methods are available affects the performance of pressure transmit-for response. time testing of pressure transmit-ters and that temperature is the dominant strer-ten in nuclear power plantsflhese methods are sor. Most of the studies on performance of nu-referred to as step test, ramo test, frequency clear plant pressure transmitters were simn-l NURl!G477 33

~

blain Citations and Summaries sored ty the NRC. 'lhe only other saajor work

4. livaluation of accelerated methods for predict.

was perforrned by manufacturers for emiron.

ing seal hfe, rnental and seismic qualification of transmitters.

$ identification of seals most affected by aging.

llowever, the transmitter qualification data are tot suffietent to addrers normal aging.

'nw benefits to be derived from the research are prtne; pally safety related, including enhanced failurc The aging research awcred in this report was a fea.

predicuon ar.d schmic protection of safety rclated pip-sibility study;it used accelerated aging to axommodate ing anJ equipment,initigation of snubber aging effects, the short (6 months) duration of the project. Since ac.

reduction of staff radiation exposures,and reduction of celerated aging docs not necessarily simulate norrnal rad waste. Numerous technical benchts are also ex.

aging, the aging r caults in this report snust be viewed as pected, including the identification of aging trends,in.

tircliminary. Funhermore, this study was concerned lormation useful in developing guidelines for monitor.

with the perfonnance of the portion of the pressure-ing senice life, the technical bases for determining sensing system and electronics located in the harsh en.

Senice life, the effects of compression set in seals, and vironment of the plantt the power supply and other improvements in snubber design, materials, and comp (ments of the pressure sensing channel that are maintenance. Repclatory benefits anticipated include hicated in the control room, cable spreading room, or contnbutions to Standard Review Plans, Regulatory other mild environments were not studied.

Ouides, Plant Technical Specifications, and ash 1II/

ANSI Oht-4 Standards based on the broader, more NURl;G/CR-$386, D. P. Ilrown, O. R. Paltner !!. V, comprehensive data base that would be developed.

Werry, and D. IL Hlahnik *llasis for Snubber Aging

'the research proposed is designed to address the Research: Nuclear Plant Aging Research Program, foHowing questions about the aging of rnechanical and Pacific Northwest Iaboratory, lake !!ngmccrmg Company, Wyle I aboratories, PNie6911 January hydraulic snubberr'

-1990

1. Ilow do snubbers age and degrade?

'Ihis report proposes a research plan to address the 2.

Yh t are the failure characteristics of Inub-safety concerns of aging in snubbers used on piping and large equipment in commercial nuclear power plants.

3. What are the safety implications of snubber

'lhe proposed program will provide the structure for 8Ei"E7 the Phase !! Snubber Aging Studv for the NRC NPAR

4. What technical information is needed to im-program, to be perfotmed at nucIcar power plants and prove the performance and life expectancy of m test laboratories. 'Ihis research would be an exten.

snubbers?

sion of the work performed by the Pacific Northwest

'lhe results will contribute toward more reliable laboratory (PNL)in the Phase 1 Snubber Aging Study, and predictable snubbers in the nuclear pow er industry the primary objectives of which were to conduct an m-and thus will improve nuclear plant safety, implemen, itial aging assessment of snubbers and to evaluate the tation of the research plan will also provide a data base concept of reducing the number of snubbers in com-for use in addressing regulatory and snubber technol.

mercial nuclear power plants. Althout.h snubber re-ogy issues. 'the data base will be made available to nu-duction prqrsms may reduce their total population ty clear utilities, snubber rnanufacturers, snubber service 50 to 80% thiwill not in.tigate the concern for manag-companies, and the NRC. Planned interfaces will en-ing the aging of the remainirig snubbers, indeed, the sure technology transfer to utilitics and manufacturers, remaining snubners may become more important to plant safety than the original population. 'Ihe pr -

NUREG/CR-5404, D. A. Casada, "AuxiliaI National posed Phase 11 research work is based, in on a Feedwater System Aging Study," Vol.1, Oak Ridp study of snubbers m U.S. nuclear power plants by the lake lingineering Company conducted for PNL under laboratory, ORNIA566/VI htarch 1990.

the NPAR program. A survey of U.S. utilities con-

'lh s review of the auxiliary feedwater (AFW) sys, ducted for PNL by Wyle laboratones on the us of -

tem used at pressurind water reactor (PWR) plants

. snubbersin nuclear plants was also used + ' identify re-has been conducted under the auspices of the NRC scarch needs.

NPAR program. 'Ihc primary purposes of the review were to (1) determine the potential and historical

'the following are key elements of the proposed sources and modes of failure within the AFW system, snubber research:

(2) identify currently applied means of detecting 1 Review of existing senice data, known sources and rnodes of degradation and failure, 2.- Development of senice-life monitoring guide-and (3) evaluate the general effectiveness of current

lines, monitoring practices and identify specific areas where enhancements appear neded.

3. I! valuation of the effects of compression set in

'the seport reviews histo #al failure data available hydraulic seals, from the Nuclear Plant Reliabin'y Data System. Licen.

NUREO-1377 34

=

m____--_____mu_..m_

__.__..u._..

m___________.__

m_._____

_m____.

__,______,_________mm___,_m

Main Citations and Summaries see Event Report Sequence Coding and Search Sys-testing requirements and practices, other than at-tempting to focus on the extent to which aging (venus tem, and Nuclear Power Experience data bases. 'the design or operating practices, for example)is responsi-failure histories of AIM system components are con-ble for 1auure or degradation.

sidered from the perspectives of how the faUures were I

detected and the sigmficance of the failures. Results of An analysis of historical failut e data involving Alv a detaPed review of operating and morutoring practices systems was completed by 3 detailed Kriew of an exist-at a plant owned by a cooperating utility are presented.

ing AITV system and the associated monitoring prac-General system configuratisns and pertinent data are tices of a cooperating utilityJihe sing!c largest source provided for Westinghouse and liabcock and Wilcox of AITV system degradation, based upon the analysis of units.

historical failure data, is the turbine drive for AfW

'Ihc report includes an identification of the general pumps. !! should bc noted that the turbine proper has types of AIM system design eonfigurations, an analysis been a relatively reliable,and rugged piece of equip.

of historical failure data, and a detailed review of a co.

ment, llowever, the turbme auxiliaries, including the operating utility's AITV system design and their cur, governor control and the trip and throttic valve, have rent operating and monitoring practices, contributed substantially to the overall turbme llistorically, and partleularly since the Three Mile problems Island 2 accident, the Al'W system has been recog-nized as critical to successful mitigation of plant tran*

'Ihe sum of the failures of motor operators and air sients and accidents. In recent years, operating inci-operators for valves resulted in approximately the dents invol"ing failures of AITV system components same number of AITV system degradations as did fail-have been among the leading events identified in urcs of the turbine drives alone. Pump failures and i

NUkEG/CR-4674, Vols.1-8, " Precursors to Poten.

check valve failures were also significant cont.ibutors tial Severe Core Damage Accidents," in which the to system degradation.-

leading risk significant events are identified for several calendar years, in the years 1984 through 1986, seven For each type of component and for the various of the top ten events at i WRs, f rom a core damage risk g

gg gg gg standpoint, involved partial or total failure of the AIV detection were &signated and tabulated. The most e

system. Operational problems with these systems have notable feature of this aspect of the study was that been diverse in nature. The report lists six events re-fauures related to instrumentation and control domi-sulting in NRC llulletins and information Notices as nated the group of fauures that were detected during examples of the diverse types of failures involving the demand conditions (as opposed to failures detected as AIM systems. Numerous other operating experiences the result of periodic monitoring or routine observa-have resulted in feedback to the industry through both tions made by operators or other personnel). Many of the NRC and the institute of Nuclear Power Opera-the potential failure sources that were not detectable tions (INPO).

by the current monitoring practices were related to the in reviewing the role that aging plays in failures instrumentation and control portion of the system.

such as those of AIW systems, three important points must be conside;ed. First, a combination of factors, in*

It was also observed that a number of conditions cluding design, maintenance, operation, aging, and related to design basis demands are not being periodi.

Other considerations may be involved. These factors cally verified. lixamples of thesc include pump capaci-are not necessarily mdependent of one another-ties not being verified at design flow / pressure condi-tions, turbines not being verified to be capabic of deliv-Second, systems age only as the individual compo-ering required torque at low steam pressures, various nents age. Other studies performed under the NPAR control sequences not being checked, and automatic program address important components within the pump suction transfers not being tested.

AIM systern and discuss the aging stressors for these Individual components.

Anoth opdon mak ws that some compo-Third, a study performed by lNiil reviewed histori-nents or certam parts or asnects of components appear cal failure data from the Nuclear Plant Reliability Data to be tested in excess of what failure history mdicates Io System (NPRDS) and made judgments as to whether be appropriate. On the other hand, other aspects of or not individual failure episodes were related to aging.

certain parts of the AIM systems are either never llecause of the above three points, the ORNL ap-tested or receive less than thorough testing. It appears that improved testing requirements are needed in or-proach to the AlW system study has been to focus der to reduce excessive testing while at the same time attention on how and to what extent the various AIM ensuring that thorough performance serification is system components fail, how the failures have been and can be detected, and what is the value of existing conducted periodically.

NUREG-1377 J

35

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

Main Citations and Summaries M'RI:G/ Cit-5M6. K.G. DeWall and R. Steele, Jr.,

tests, the parametric study included varying toth the

• llWR Reactor Water Cleanup System 11cxible degree of inlet water subcoolin g and the pressure. 'Ihe Wedge Ne isolation Valve Qualtfication and liigh four Valve 11 tests were all performed at a normal linergy llow Conclusionrs,, Interruption Test: Vol.1, Analysis and llWR 10'F subcooling, and only the inlet pressure was Idaho National lingineering 1.nbora-vamd*

tory,liGG-2569 October 1989.

Test results show that, for both valve designs tested, Recent testing sponsored by the Nucicar Regula-the force required to open and close the valves at tem-tory Commission (NRC) showed that, for at least some peratures above 100*17 was significantly higher than gate valves installed m nuclear applications, the equa*

the force predicted by the valve manufacturers. Only tions used by industry to size the valve operators do not conservatively calculate the thrust needed to close the during the valve-opening tests at room temperature valves under design basis loadings. 'the tests als without flow did the typical industry vahe thrust equa-tion predict the valve response. Industry has assumed showed that the results of in situ valve testing at lower that the valve opening thrust requirements would be loadings cannot be extrapolated to design basis load-the highest wben the disk lifted off the seat. *lhis was mgs. 'lhis volume desenbes the testing conducted by determined not to be true for the valves tested.'the the Idaho National lingineering Laboratory (IN!!L) to highest opening loads (maximum thrust) with flow oc.

proviJe techmcal data for the NRC cffort regarding curred at different openings for both valves, but in both Generic issue 67 (Gi-87)*Tailure of IIPCI Steam Lme Without isolation." 'lhe test program also provides in.

cases, they were well off their respective seats. Valve.

fortnation applicable to Generic issue !!.C.6.1,"In Situ closing thrusts at full line-break flows were higher (up Testing of Valves," and a related document,113 Ilulle-to one third) than anticipated.

tin 8$--03, " Motor Operated Valve Common Mode The test results provide evidence for two concerns l' allures During Plant Transient Due to improper with MOVs in nuclear power plants, l'irst, proper sir.

Switch Ecttings."

ing of motor operators is complicated by the fact that the equation used for calculating the stem force Of the three boiling water reactor (llWR) proccss needed to close or open a gate valve does not have lines covered by 01-87, an unisolated break in the re-terms for the effects of temperature, degree of fluid actor water cleanup (RWCU) supply line was selected subcoolaig, internal valve clearances, and the differ-for the first phase of testing because such a break ences in the opening and closing forces that are not ac-

)

would have the greatest safety impact. All three GI-87 counted for by the stem rejection term. Second, effec-process lines have common features: all communicate tive in situ testing is very difficult because (1) the tests with the primary system, pass through containment, cannot be conducted at design basis conditions and (2) and have normally open isolation valve 5-even with the valve loadings properly quan.ified during To sneet the new valve operating criteria required the in situ tests, the results cannot be extrapolated to by IE Ilulletin 85-03 and Generic Letter No. 89-10,in -

design basis conditions because the final Ihrust varies dustry developed new diagnostic test equipment and depending on the extent to which disk friction rather methods for in situ motor operated valve (MOV) test.

than disk seating affects the torque switch, ing. lH llulletin 8$-03 succeeded in significantly im-The disk factor of 0.3 typically used in industry to proving the operability of the selected safety related calculate disk friction force is not conservative for valves because it caused many of the utilities to eitherof the valves tested. Adisk factorof 0.$ margin-reanalyze the design basis load for the applicable ally predicts the forces for one valve during hath open-MOVs and to reset the control switches accordingly*

ing and closing. 'Ihc response of the other valve is en-4 llowever, very little design basis testing of valves veloped by the 0.5 disk factor during opening but not has been conducted outside the plant to venfy the ana-during closing. Today's tools for analyzing valve re.

lylic assumptions used to determine valve switch set.

sponse to fluid k)adings are not sophisticated enough tings. Analytic assumptions are necessary because, in to detect small design differences that make large re-many cases, the utility cannot ter.t valves at design basis sponse differences. Temperature also affects the loadings in situ. 'the GI-87 testing provides some of thrust requirements of these gate valves, the;s valve operator sizing equations can be compared.

testing of prototypical valves at design basis loadings first measured valve responses with which indus' All the facts listed justify continued qualification-try In this initial test program, two representative and stress the need for industry to add new terrm to the RWCU isolation valves were subjected to the hydraulic equation or to increase the disk factor to a very conser.

qualification tests described in ANSI 1116.41, the quali-vative number to account for the missing terms in the i

fication standard for nuclear valves, and then to flow equation. Also, test results show that the stem factor is interruption tests at full RWCU pipe-break flow. In not a constant but changes with stem hiad, thus making all, fourteen flow interruption tests were performed, it very difficult to extrapolate normal in situ valve test-ten on Valve A and fc.ur on Valve 11. In the Valve A ing to design basis conditions.

NUR130-1377 36

'------'-----_______.m..

Main Citations and Summaries l

rnanufacturers listed m the survey is r eview ed and defi-When tests or improsed sitmg equations have de-termined the thrust needed to operate a valve at its de-ciencies in that methodohyy ate idenuhed.

sign basis loadmp, utilities can use one of several mod-Four botling water ieactor (llWR) systems the ern diagnostic systems to conservatively set the motor emergency coohng system, the high. pressure coolant operator control switches. Ilowever, this method may njection system, the reactor core isolation covhng n s-execed the allowable thrust on some valve designs, tem, and the reactor water (leanup nstem, were m-This job will be made easier and the r esult will be mor e cluded in the salve assernbly (haractettution. The conservative if both the torque and the thrust can be typical" containme nt isolation valve is a 3 to 10 in.,600 rneasured when the switches are set. If further research to 900 lb rate valve. The rnost common design is a cast proves lhnt there is a proportional telationship be-steel, fleuble wedge, picssurc+eal valve with a tween stem load and stem f actor, the degt ec of conser-1imitorque operator (AC inside and DC outside of vatisra can be reduced.

containment). "Ihe Anchor /Dailing Vahe Company manufactures approxtmately 40"e of the valves in the NUlti:G/ Cit-5406, K.G. DeWall and it. Steele, Jr..

four il%lR systerns.

"llWR lteac'or Water Cleanup System llexible

.the midgabon oi a high.cnergy pipe h eak is within Wedpe Gate Isolation Valve Quahfication and liigh the design basis for the above salve awembhes, with linergy 110w Interruption Test; Vol. 2, Data Re.

typical system design conditions of 1250 psi and 575'F.

r1, Idaho National lin neering 1aboratory, No flow testing has been perfermed under these condi-1G0-2569, OctoMr 198 nons to efy tk presmnpuons udynandactum

'lhis volume pr esents the 700 pages of actual meas-in the quahfication analysis calculations. Operator ured data from the gate valve test program describedin tmque md munp am MerminW uW caWa-Volume 1. 'lhey are provided for those readers who tions supphed by the valve vendor; torque settings m-wish to hok at the data and form their own opinion on adequate to dow We de muld result if the ongmal the performance of the test valves. l'or those readers cakulanons am not conmah-who wish to do their own analysis, the electronic data Most of the valve and operator manufacturers use are available from the Idaho dol 10ffice of Technol-the ame egintion with minor vanations in coe f beients ory Transfer, (208) $26-831h.

to site operators. In this equation, the required thrust 1 igure 1 of Volume 2 shows the test h>op in sche-to close the valve is equal to the sum of the disk drag matic f orm and identifies the instrument location and load due to dif ferential pressure, the stem end pres-numhets. Figuie 2 converts the differential pressures stu e load, and the pac king drag load.The se rvice condi-into flow rates kallons per minute).'lable 1 outlines tions used in the thrust equation are supplied by each the test sequence performed on each valve and corre-indisidual plant. l'our areas h;nc been identdied as htes the data as they are presented here. In the re-having the most influence on stem thrust icquire-maining figures, the header on each plot defines the ments:

valve ( A or 11), the test senes number, and the test step

1. Repeated tycling can has e a sign ficant eifeet on number. Table 2 lists the test parameters measured valve thrust requirements.

during blowdown tests, Table 3 displays the test step

2. The typical value of 0.3 for the dak friction matnx for quahheation and blowdown tests, and Table coefheient used by the industry is not conserva-4 lists the test steps and system pressure and tempera-tive for all cases.

ture for each of the tests performed,

3. 'lhe influeneeof massflow/momentumonvalve NUHl:GiCit-5406, K.G. DeWall and R. Steele. Jr.,

thrust icquirements may be sigmheant.

"IlWR Reactor Water Cleanup System llexible

4. Increased temperature causes a significant in-Wedge Gate Isolation Valve Quahfication and liig'h ercase in valve closure loads, linergy llow interruption Test; Vol. 3, Review of issues Associated with llWR Containment isolation

'lhe limited nurnber of tests performed to assess Valve Closure," Idaho National lingineering I abo the egaWity of the pate valve to interrupt the flow of ratory, EGG-2564, October 1959.

high. pressure steam has resulted m a relatively fre-This volume discusses research performed to de-quent inabihty to isolate portions of piping systems, The data now available surgest th?

iustry may be us.

velop techmcal insights for the NRC effort tepardmg mg nonconservative friction far < ; and pombly un-Genene issue 87. "l'ailure of 1IPCI Stearn 1.ine With-derestimating valve stem thru:i a quirements. Addi-out isolation." Volumes 1 and 2 describe the relevant tional work is needed to detcnmne whether oresent valve test propram.The research began w ith a su rvey to quahfication practices are adequate. Recommenda-characterue the population of normally open contain-tions for expandmg the quahfication of valve assem-ment isolation valves in those process lines that con-blies for high energy pipe break conditions ate pre-nect to the primary system and penetrate containment.

sented.

The quahheation methodology used by the vanous NURI&l377 37

Main Citations and Summaries NURrG/CR-5419, M. Villann, R. Fullwood and

4. Individualplant maintanance records forinstru.

M. S'ibucht, Aging Assessment of instrument Air ment and service air systems were found to be Systems in Nuclear Power Plants," ilrookhaven National Laboratory, llN1rNUREG-52212+

the most comprehensive source of data for per-January N forming aging analyses.

5. As a continuously operating system with mini-As par 1 of ongoing ciforts to understand and man-mal control room instrutnentation because ofits age the cifects of aging in nuclear power plants, an nonsafety classification, most problems in the aging assessment was performed for the instrument air air system are detected by kical monitodng and system, a system that recently luis been the subject of indication, walkdown typc inspection, and pre-much scrutiny. Despite its nonsafety classification, in-ventsve maintenance inspectton or survettlance.

strument air has been a factor in a number of poten.

6. Review of compressed air system designs and tially serious events.'lhis report presents the results of studies using a PRA based system model re-the assetsment and discusses the impact of aging of the vealed that the redundancy of key components instrument alt system on system availability and plant (compressors, dryers, instrument air / service air safety.1his work was performed as part of the NRC cross-connect valve) was an important factor in NPAR program. The ob,icctive of this study was to sptem availability.1he overall design configu-identify all the aging modes and their causes that ration affected the pervasiveness of air system should be mitigated to achieve reliable operation of all

problems, safety rdated air equipment. Also included is an in*

terim myiew of typical maintenance activities for air

7. Totalloss of air events are uncommon. The systemt in the nuclear power industry.

m' Mrity of events resulted in degraded opera-ti - ' low pressure, air quality out of limits). Nor-m.

vear of the system and contamination of the To perform the cornplex task of analyring an entire system, the Aging and Life Extension Assessment Pro-air dominate the problems of system failure, gram (ALEAP) System level Plan was developed by Procedures and testing for the responce of per-Drookhaven National 12toratory (13NL) and applied sonnel and equipment to these conditions successfully in previous studier,lhe work used two should be developed.

parallel work paths, one usmg deterministic techniques 8.11uman error was a significant cause of failures to assess theimpact of aging on in critical components Mn as compressors and perfonnance, and the other usm, compressed air system dryers, mail as at the system and intersystem g probabilistic meth-ods.The results f om both paths were usej to charac*

,;,el. Training should be nugmented in two key terire aging in the mstrument air system. The findings steas:(1) operation and maintenance of critical from this study, some of which have applications be*

ah 9' stem components and (2) understanding yond the instrum:nt and semce att systems, formed a the hcportance of instrument air to other plant technical basis for understanding the effects of aging in systems, particularly safety systems.

compressed att systems.1he major conclusions from

9. lhe outsidu 9 stems that were most often af-tlus work are:

fected by instrement air problems are contain4 ment isolation, Main feedwatcr/ main steam,

1. This study identified aging trends in component auxiliary feedwater, and the llWR scram syb-f ailure rates, the relative importance of compo-tem. The most commordy affected components -

r ents, and system unavailability. All these were ainoperated and solenotd-operated trends could have a deteriorating irnpact on sys.

V" IVC' tem availability and consequently on plant 10.The probabilistic work entaileo the develop-safety in iatet years.

ment of a computer program (PRAAGE IA)

2. Compressors, att system valves, and att dryers, usmg a PR A based instrument air systera inodel made up the majority of failures.The failures m

- to perform time-dependent PRA calculations.

passive components such as piping, after*

Time-dependent failure rates were developt41 coolers /moicture separators, and receivers in' from the data base and other inputs to the pro-creased with time, but these still constituted gram to calculate system unavailability and com-only a small percentage of overall failures.

ponent importances for various ages. The re-

3. The effectiveness and quantity of preventive

. sults showed that, when the time dependent ef.

fccts of aging for the worst case are accounted maintenance devoted to a component signifi-for, there are two significant system effects:(1) cantly reduced the number of failures experi-

. system unavailability increases moderately with enced. Ilowever, existing maintenance pro-age and (2) the relative importance of compo-grams within the industry lack uniformity, and nents changes with ag: During early operation, quality assurance is not rigorous because the leakage in both instrument air / service air piping system is classified as "nonsafety."

and support system piping was the most impor-NUREG-1377 38

Main C:tations and Summaries l

I

'ihe batteries in the test program had lead-calcium tant contnbutor to system unasadahdity;darmy plates and were manuf actured by C&D ilatteries. Dis.

the later years, agmp can eause compr ewirs and eussions with C&D personnel indicate that they are air drprs'fdters to become increasmgly typical of battenes presently bemg installed in nuclear important.

facthties. liac h cell had a rated 8. hour electrical capac-

'lhe imdmps presented in this report form a sound W im ampere hours;was7-5/8in.long,14-1/8in.

techrucal basic for understandmg and managing the ef-wd, and u-Ul6 in. high; and weighed about 240 fccts of apng m instrument air rystems.1:uture work pounds. lhey were obtained from a nuclear facility wdl include improvements to cuttent maintenance, where they were naturally aged to 13.5 years. Itecords monitoring, traming, survetHance, and off-normal re-provided by the nuclear f acihty indicate that the batter-sponse procedures to mitigate degradation due to ies were maintained and tested m accordance with practices that are consistent with those in 11!!!!! Std I

"E "E 450.

NUlti:GTit4448,3 A lidson. " Agmg livaluation of

.the batte ries w er e installed on a i. hake table using a Class Ili lla9c!!w Seisnue 'lesting," 2576, August new battery ratk purchased from the battery vendor Idaho Na.

tional IM.iocenng l aboratory, I!GO-and were tested to seismic spectra that are typical of lw0 those required for the safe shutdown carthquake (SS!!)

llatteries are the only installed source of electric in U.S. nuclear facthties. infor mation received frorn se-power to provide for namtoring plant conditions and lected nuclear plants and the lilectric Power Research wntrol of some systems et the nuclear reactor in the Institute (l!PRI) was used to specify the required re-oent of a station blackout Odloffsite power is lost and sponse spectrum (RRh) for the i,cismic tests.'ihe tests the diesel generators do not statt). Approximately 60 w ere conducted usmg four dif ferent seismic les els rep-mdividual 2-V ecils are connected together to form a resenting the best estimate for the RRS encornpassing typical 125-V de battery bank that has enough vohapc 504,85G,959,and 100% of the U.S. nuclear plants, and electrical capacity to proviJe the needed electne 1)uring the scismic tests, the batteries were dis-power for the petiod of time determined for each nu-charged at 2% of the.hhour rate while current and bat.

clear plant in accordance with NRC regulations.

tery voltages were monitored to detect the existence of Within the NRC Nuclear Plant Aging itesearch catastrophic f ailure.1)uring the prescismic, seismic, (NPAR) program, a Phase I study of battery at ng was and postseismic tests, alternative surveillance and i

performedand reportedin NURl G/CR-M57 " Aging monitormg methods were employed to determine of Class ll! Itatteries in Safety Systems of Nuclear whether other methods may be more sensitive to aging-Power Plants." 'Ihe study concluded that signincant related degradation of battenes tnan the standardvolt-ampere tests that determine their electrical capacity.

aging effects for old batteries are growth of positive

'lhe alternative monitonng methods employed were plates,loosemng of active material m plates that have (1) measurement of internal resistance, (2) measure-grown, loss of acthe material caused by passmg and rnent of capacitance, and (3) measurement of battery corrosion, and embrittlement of the lead gnds and polaritation (comparison of battery voltages measured straps.The results of these cifetts are decreased elec-while increasing d:scharge current with those obtained in al capacity and decreased seismic ruggedness that, with decreasmg discharpe current). 'lhese measute-during a seismie event, can lead to decreased electrical ments w ere suppested hs a result of investigations per-performance or complete failure-formed by the Westinghouse R&D Center for Sandia Since batte nes are susceptthte to aging degradation National laboratories in 1986.

that wuld cause ohl batteries to be vulnerable to se.

Results of the seismic tests indute that the capac-sete seismic events, a test program was conducted to ity of lead calcium batteries of this design did not de-deteimine if it is possiNe for the scismic ruggedness of crease as a result of shaking at scismic levels that in-aped battenes m nuclear plants to be inadequate, es en clude the most severe SSl; levels specified. In fact, the though the measur ed electncal capacity is satisfactory, average electrical capacity (ampere hours) of batteries in adJition, selected alternative surveillance rnethods tested at the 100% seismic level increased from a wete evaluated durmg the testing program to deter-preseismic capacity of 96% to a postseismic capacity of mine if any of them are hkely to be more sensitise to 48%. ~1he batteries did not show any degradation ex.

battery depradation than the surseillance and testing cept for some minor external damage.'the battery rack methods specthed in Ilitill Std 450-1957,"lFlill Rec-suffered sorne benJing of structuralcomponents, but it ommended Practice f ar Maintenance, Testing, and Re-performed its intended functions. Post test disassem-placement of Iarpe Storage flattenes for Generating bly of selected battenes showed that some corrosion of Stations and Substations" and Regulatory Guide the weld joint between the posithe plates and the bus /

1.1N. " Maintenance. Testing, and Replacemect of ternunal assembly had occurrcJ as a result of the natu-l 1arge I cad Storage liatteries for Nuclear Poacr ral ar mg process. I low eve r, this degradation did not m-Plant s,"

NURiiG-1377 39

hiain Citations and Summaries tctfere with the seistnic performance. hietallurgical cables, connections, and penetrations is dnided into examinations Rowed that a large grain structure ex.

two phases. Phase I, w hich is the subject of this report, isted at the wcld rea.nc larger grain structure of the consists of a review of applicable literature and evalu.

weld makes it su!.cqotible to corrosion and would ed ations of usage, operating experience, and current in.

plain the observed coosion, spection and surveillance rnethods. Phase 11, currently De results indicate that measurements of capacj.

planned only for cables, includes the development of tance and internal resistave can be obtained with re.

tmproved methods for inspection, surveillance, and I

peatability and may proside an indication of battery monitoring; application of monitoring methods to condition if the measuremerm were taken over the nat urally aged and in sit u cables; and r ecommendations hfetime o' the battery, Polarizaten and discharge cut, for utillring the research in the regulatory process.

tent interruption are two techniques that are capable his report includes a review of component usage of measuring internal resistance, wtJte discharge cur.

in nuclear power plants, a review of some commonly rent interruption is also capable of measuring battery used components and their matcrials of construction, a capacitance.nese measurements would be most use-review of the stressors that the components might be fulif they could be made while the batteries were new exposed to in both normal and accident environments, and then repeated at regular intervals to ottain a pat.

a compilation and evaluation of industry failure data, a tern of the change in battery characteristics with time.

discussion of comporient failure modes and causes, a

%c results of scismic tests on naturally agtd bat.

description of current industry testing and mainte.

teries nearly 14 years old showed that, when batteries nance practices, and a review of r.ome monitoring tech-are maintained and operated in accordance with IEGE niques that might be useful for ranitonng the condi.

Std 450 and Regulatory Guide 1.129, the following may tion of these components.

be expected of adequately designed and manufactured De conclusions of the study are:

lead calcium batteries:

7 1.

Cab!cs, connections, and EPAs ate highly rcli-1.

IJttle,if anf, electrical capacity will be lost hs a able devices under normal plant operating result of seismic shaking at levels that are conditions with no evidence of significant in-typical o! the most severc SSE levels specified creases in failure rate with aging. Conse.

for U.S. nuclear plants, quently, they receive little or no preventive 2.

Some internal damage to the plate separators rnaintenance. Under accident conditions, may be expected at the most severe seismic bowever, the reliability of these components is levels, liowever, this minor loss of seismic pretically unknown.

ruggedness is not expected to prevent the bat-

2. _ Agira effects that have the potential tolead to teries from providing at least 80% of rated common cause failuresdurmgaccident condi-capacity during and immediately following the tions have the highest significance.

most severe seismic event.

3.

hiany of the causes of failures for cables. con.

3.

Naturally aged batteries may sho,v evidence of nectiors and EPAs at accident conditions

)

corrosion at the joint between the positive would not cause ray detectable manifesta.

1 plates and the positive plate strap (bus). In a tions during normal operation because of the well made joint, this corrosion should not absence of high temperatures and humidities.

cause the seismic ruggedness to be inadequate The most important failure mode is expected for the most severe SSE evenis expected in the to be shorting (or reduced c!cctricalisolation).

U.S. Operation of batteries at elevated tem-Several different causes may result in this fail-perature or excessive charging could increase ure mode, the corrosion, which could then progress rap-4.

Plant operational experience is useful to the idly enough to result in inadequate scismic ruggedness.

extent that it may indicate some possible fast-acting degradation mechanisms for cables, NUREG/CR-5461, hfJ. Jacobus," Aging of Cables, connections, and EPAs that could lead to com.

Connections, and Electrical Penetration Assemblies mon cause failures under off normal environ.

Used in Nuclear Power Plants," Sandia National mental conditions. Ilowever, current LER laboratories, SANDS 9-2369, July 1990.

data provide a very limited data base for this his report covers the examination of the effects of purpose.

aging on cables, connections, and containment c!cetri.

5.

A significant number of manufacturers have cal penetration assemblies (EPAs)as part of the NRC produced cables, connections, and EPAs, re-NPAR program. Cables and connections are used in sulting in many different materials, degigns, nery electrical circuit in all nuclear power plants.

and connruction methods. Consequcntly, ge EPAs are included in every circuit that is inside con.

neric assescments of aging effects and vul-I tainment. His NRC sponsored ging assessment of nerabilities have become much more difficult, 1

NUREG-1377 40

_ -_ : L _

Main Citations and Summaries particularly where failure modes relate to in-ting up what the authors consider to be a reasonable internals monitoring program for U.S. utihties.

terfacing stresses, NUlti:G/CH-5491,11. P. Allen and A.11. Johnson, Jr.,

An experimental assessment of cables is currently

• Shippingport Station Ar.ing Evaluation," Pacific under way and will be documented in a future report.

Northw est laboratory, liNI A191, January 1990.

"Ihis report desenbes a research plan to address NUllEG/CR-5479,11. Damiano and R. C. Kryter, safety concerns on agmg of snubbers used on pipmg

" Current Applications of Vibration Monitoring and and equipment in commercial nuclear power plants.

Neutron Noise Analysis: Detection and Analysis of

'lhe work is to be performed under Phase 11 of the Structural Degradation of fleactor Vessel InterTials Snubber Aging Study of the NRC NPAll program with Oak Itidge National frorn Operational Aging,"ll398, February 1990.

the Pacific Northwest 1;iboratory (PNI.) as the prime laboratory, OltNIIlh!-

contractor, ltesearch conducted by PNL under Phase I "the detection of degradation in PWit internals due pr vided an initial assessment of snubber operation to operational aging is becoming more and more im, based pnmardy on a review of licensee event reports.

portant to U.S. utihties as the median age of U.S. nu.

.ne wou proposed is an extension of I hase I activities clear powe r plants incr eases. Monitoring and de tection and covers research at nuclear power plantt and in test of agmg ef f ects should aid in justifying plant life exten-laboratones. the report includes techmeal background sion and tesult in safer and more efficient operation on the design and use of snubbers in commercial nu-during the present and extended life period. It has

&ar power applications and a discussion of the pri-been demonstrated that monitoring programs based on rnary i i ute modes of loth hydraulic and mechanical neutron noise and vibration measurements utilizing snubbers. 'lhe anticipated safety, techmcal, and regu-signature analym an effectively detect, and in some latory benefits of the work, along with concerns of the cases diagnose, degradation of reactor vessel internals.

NitC and the utihties, are ah,o subjects of the report.

Such programs have the potential to reduce plant The Shippingport Atomic Power Station, presently downtime, make periodic maintenance more effective, (1989)in the fmal stages of decommissioning, has been and increase plant safety.

a major source of naturally aged equipment for the Monitonng of reactor internals can be con *idered a NPAll and other NltC programs. 'lhe evaluation of particular application of the generalconcept of predie-naturallyagedcomponentsisanelementof the NPAll tive maintenance, the techniques of which are aircady program strategy, llecause naturally aged components widely used in mdustry to monitor rotating machmery, and rnaterials experience the actual service related ex-Predictive maintenance will be further implemented as ternal stressors, corrosion and wear, testing procc-(1) its benefits become better documented, (2) f amis dures, and maintenance practices, the evaluation of anty with the techniques and their applications grows, such components is valuable. One is able to verify deg-and (3) better hardware and software become avail-radation models, to validate aging projections based on able. A similar statement could apply to the monitoring the extrapolation of accelerated test data, and to detect of reactorinternals. Although this monitoring hasbcen unexpected aging mechanisms (surprises) that could spotty in the U.S., the above mentioned techniques signtficantly affect the safety performances of compo-have been widely applied in Europe, particularly in nents or systems.

France and the Federalllepublic of Germany, where Despite their importance for plant studies, natu-they are currently (in 1989) 5 to 10 years ahead of those rally aged components of the desired type and vintage m this country. U.S. utilities could benefit from the ex-are not readily available.'Ihe best source of these com-perience in liurope, where, in many cases, intermds ponents is operational equipment from retired plants.

moratonng has been integrated into regular plant The decommissioning of the Shippingport Station, par-mamtenance programs. Ihus U.S utilities could im-ticularly because it was managed by the U.S. Depart-picment effective monitoring programs with a mini-ment of Energy, represents a valuable opportunity to mum of experimentation and wasted effott.

conduct in situ assessments at an aged reactor and to The r eport begins wnh a description of some promi-obtain a variety of naturally aged and degraded compo-nent rnechanisms through which degradation of reac-nents and samples for detaded aging evaluations by NitC contractors. As the first U.S. large scate, tot internals occurs; the cause of most cases of this deg-ccntral4tation nuclear plant, the Shippingport Station radation is flow-induced vibration. Other mechanisms are also reviewed. This is followed by a brief descop-parallels commercial pressuriicd water reactors in re-tion of vibration monitoring und neutron noise analy-actor, steam, auxiliary, support, and safety systems.

sis, inludmg a comparison ar.d evaluation of these two The 25-year service hfe (1957 to 1982) covers almost methods. Next, curtent pactices are summarized, and the entire period of currently operating reactors. Also, because of substantial modtfications during the examples of appheat. ions of these rnethods in both the miJ-19tius and 1970s, it offers unique examples of l

U.S. and liutope (mainly West Germany and France) identical or similar equipment used side by side with are given.The report concludes with guidehnes for set-SUllEG-1377 41

Main Citations and Summaries the origmal equipment but representing different vin.

pects of their actmticsflhe tre p ction Protram recog-l tages and degrees of aging. As patt of the Shippingpirt ni/cs that beens.ces may satisfy NRC requuements in i

Station aging evaluation work, more than 200 items, ways that ddfer among the beensees, and mtpection ranging in we from sma!! instruments and material guidance is therefore c< pressed m the form of per-l samples to main coolant pumps, have been removed formance objectives and evaluation entena. l'or the and shipped to designated laboratories.'these items in.

resident and regionalinspectors. pnicedures covering clude battery chargers, inverters, relays, breakers, ruth subjects as operations, maintenance, and surved-switches, power and control cables, electrical penetra.

lance h:ne been wntten. Some of these procedures tions, check valves, solenoid valves, and motor, contain guidance on derradation due to armg.

operated valves. Samples of piping from various plant systems also have been ncquired for radiological char-Associated with each NPAR study is the need to de.

acterization studies, and sampics from the pnmary sys-termine the tote of inspet tion, m untenance, and inoni-tem components wdl be used for material degradation tormg in counteracting the cifects of at ng and service i

wear.'lhe role of ma' tenance in managing aping isan m

studies.

important area where NRC emphasis has been ap.

Data and records relevant to the procurement, op.

plied. A review by the NRC of maintenance performed cration, and maintenance of these materials and com.

at semal plants concluded that most utihties do not ponents have been obtained to support the detailed ag-pedmm condition momtonng because of inadequate ing evaluations. In situ assessments of Shippingport knowledpc of degradation rnechanisms and measur-Station cornponents also have been conducted,includ-able condition. indication parameten. Ihe output from ing preremoval sisual and physical examinations of NPAR in this area could provide mformation needed components, tests of electncal circuits, and special to assist the mspectors to recogni/c are related con.

measurements to assist in the selection of specific com-ponents for further evaluation. Although detailed The types of information pencrated by NPAR that evaluations of the naturally aged compments and ma-were found to be relevant to mspection needs mclude:

terial from the Shippingport Station have not been completed, the results from preliminary studies indi-1.

TuruWmaHmluatorrNPAR reports identtfy cate the value of the aging information that may ulti-paranders that can be mcnitored or meas-mately be obtained ured to Atm adng &gtaanon. We inspeo tors can apply these results to enhance visual inspections (w alkdowns) and to evaluate licen-NUREG/CR-5507, W. Gunther and J. Taylor, "Results see programs for ensurmg the operability of from the Nuclear Piant Aging Research Program:

Their Use in Inspection Actmtics,"linvokhaven equipment and systems.

National I ahoratory, llNIcNUREG-52222, Sep-2.

Tuduremodes.cumes cffetis-Operatingexpe-tember 1940.

rience data evaluated in NPAR studies can

.lhe NRL, NPAR program has determined the sus' alert the inspectors to the prevalent failure ceptibility of nuclear power plant components and sys' mechanisms of systems and equipment. The tems to aging and the potential for aging to affect plant potential for changes in failure rate mth in-safety and availability. The pyram has also identified creasing age is usefulin evaluatmg preventive J

maintenance.

metho s for detectmg and mitigating the effects of ag-ing in cornpanents. A review of the NRC Inspection 3.

Stres3cs thar c uuse degradation-I nspectors e;m Program and discussions with NRC inspection person-benefit from knowing the environmental and nel revealed several areas where the NPAR results operational stresses t hat ea use or affcci degra-would be valuable to the inspectors. This report de-dation due to armg.

scribes the NPAR information that can enhance in-spection activitics and provides recommendations for To obtain a complete delineation of the NRC in-communicating this information to NRC inspectors.

spectors' needs, presentatiens summarizing the results

'ihese recommendations are based un' a detailed as, of the NPAR program were made to the resident in-sessment of the NRC Inspection Program and on feed.

spectors at tbree regions. 'their comments, supple-back from resident and regional inspectors.

mented by a wntten questionnaire, indicated that k

NPAR resultscan be of use to theinspectors when pro-The emphasis of the NRC Inspection t'rogram ison vided in a format directed to their activities. Itxamples evaluating the performance of licensees by focusing on of NPAR report summanes and inspection guides for requirements and standards associated with the admin-aging-related degradation of components and systems istrative, managerial, engineerng, and operational as-are included in the eport.

e NURiiG-1377 42

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

- - - - - - ~ - - - ~ '

hiain Citations and Sunimaries When the increase in core damage frequency is NUREG!CR-5510, W.11. Vesely, R.!!. Kurth, and S.ht.

Scalro," Evaluations of Core hicit' Frequency lif-targe foi a given surveillance and maintenance pro-Icets Due to Component Aging and hiamtenance,"

gram, examination of the detailed aging contnbutors Science Applications international Corporation, shows that relatively few comgments contribute. 'this SAIL-89/1744, June 1990.

implies that a " graded" maintenance program or,

'lhis report presents the results of a project to de-equivalently, a *prioritized" maintenance program can velop a methodology using probabilistic risk analysis cf f ectively controi the core damage irequency increase (PR A) and component aging models to quantify risk ef-due to aging. In such a maintenance progsam, most fects due to component and structural aging.'Ihc ap-components can have a lower level of maintenance if proach allows any present PRA and any aging model cornponents important to core damage frequency have f or the components and structurcs to be used. An im*

a higher level of maintenance.

portant part of the evaluations is that the effects of

.lhe dominant agingcontributors for the PWR were maintenance and survedlance programs in controlling found to be diesel generators, specific check valves and aging can be quantified.'Itese programs can be explic-motor operated valves in the emergency core cooling illy evaluated to determine their effectiveness in con-system, and motor-driven pumps and turbine-driven trolling aging impacts on system unavailability, core pumps in the auxiliary feedwater system. For the damage frequency, and pubhc nsk. Iloth point evalu-IlWR, the dominant aging contributors were the dic-ations and uncertainty evaluations can be carned out, sels, the motor driven pumps in the service water sys-and detailed contributors to the aging effects can be tem, and the turbine driven pumps in the reactor core identified and prioritized. PRA models are separated isolation system. The aging contribution from every from the aging models, allowmg available PRAs to be component in the PRA is provided and prioritized.

efficiently used m evaluating risk effects of aging.

'lhese detailed contributors include specific systems.

To demonstrate the rnethodology, two PRAs, one components, and failure modes and provide a compre-for a PWR and one for a BWR, were used to calculate hensive means of focusing aging analyses and aging the increase in core damage frequency caused by agir:g control efforts.

for given aging data and assumed surveillance nnd in addition to the point calculations, uncertainty maintenance programt.. The increase in core damage evaluations were carried out. For these evaluations, frequency due to aging was averaged over tune. Ihts ranges were assigned to each component aging rate, average increase m core damage frequency character-each effective overhaul interval, and each effective tied the effectiveness of the maintenance and surved-gg.c llance interval. These ranges described uncer-lance program in controlling aging effects. The average tainties and variations in the data. Ing. uniform distri-merease m core damage frequency can be added to the butions, which are flat distributions on a logarithmic baseline PRA core damage frequency to obtam the to-scale, were used for the uncertainty propagation. All tal projected core damage frequency under a given the variableswere treated asbeingindependent of one mamtenance and surveillance program with the actmg another for the evaluations.

aging process.

NUREG/CR-5519, J.C. hioycis, " Aging of Control and The agingof active components was modeled using clear Power i,ompressors and D crs Used in Nu-tants,

$cnice Air C the linear failure rate aging model in which the compo-ncnt failure rate h.nearly increases with age according tory, ORNL-6607/V1, July 1990.

to a characteristic aging rate. To demonstrate the

.lhis report discusses work performed as part of the methodology, four aging rate data bases were used:

NRC NPAR program on practical and cost effective

,l'IRG Al.F.X, h10DI, htOD2, and hlOD3. These data methods for detecting, monitoring, and assessing the bases demonstrated the effects of different aging rates wverity of time dependent degradation (aging and on the core damage frcquency for a given maintenance

. ice wear) of compressors and dryers used in the and survedlance program.

control and senice air systems of nuclear power plants.

Results obtained for dtfferent surveillance and The objective is to provide capabilities for establishing maintenance programs clearly show the sensitivity of degradation tiends prior to faihue and for developing the increase in core damage frequency to the type of guidance on effective maintenan;e programs.

maintenance program and the aging rates. The results The topies covered are failure modes and causes re-are significant from a technical standpoint because sidting from aging and senice wea r, manufacturer-rec-they explicitly quantify the imnacts that aging and ommended maintenance and sutveillance practices, maintenance can have. These evaluations are the first and measurable parameters (including functional indi-m) kr uw in assessing operational readiness and quantifications of aging and maintenance impacts us-g.

ing full-scale up to-date PRAs.

equipment condition (often rela'ed to degradation trends) and in detecting incipient failure. 'the results

' y,.

7e wm.. we e..-wm.""

are based on information derived f rom operating expc.

rience records, manufacturer. supplied information, 43 NURI!G-1377

&m Main Citations and Summaries and inputs from plant operators. For each failure clude moisture sensing within the deuccant column mode, failure causes are listed by subcomponent, and near the exit and perkwlic rnonitonng of the axial tem-potentially useful parameters for detecting degrada-perature profile within the column. Use of these meas-tion that could lead to failure arc identified.

urable parameters in the sun'eillance and monitoring A brief review of typical compressors and dryers in progtvn might reduce the level and duration of time-nuclear power plants showed that the nonlubricated directed out of semce mspection and maintenance, reciprocating compressors and the regenerative desic-thereby increasing availabuity and improving merall cant dryer are used in more plante than any other types system reHaMy.

for both service and control air systems, and the assess.

Nuclear plant control and senice air compressors ment was therefore focused on them. A general de.

and dryers are not usually considered as safety related scription of the equipment that includes illustrations, because the air systems are nct needed to bring the defined equipment boundaries, functional require-plant to a safe shutdown condition. An effective sur-ments, nd materials of construction is provided. Op.

veillance and monitoring program with preventive and erational stressors are categorized and listed in detail, corrective maintenance can provide reliable service Data bases and nuclear industry reports containing from nucle,ac plant compressors and dryers. Instances nuclear power plant operating experience were exam-s of mr supph due to compmsw or tu faUum ined. Dese data bases included the Licensee !! vent am ram becam q tp dndancy m most sWemt F r these reasons,it is recommended that no further R eport (LER) file as cataloged in t he Sequence Coding C

n qp n be meluded in the and Search System maintained by ORNL's Nucicar NI Er E Opentions Analysis Center, the Nuclear Power Expe-rience compilation maintained and published by the S.M. Stollers Corporation, the in Plant Rehability NUREG/CR-5546, S. P. Nowlen, "An Investigation of the Effects of Herrnal Ag' Cables," Sandia National ing on the I ire Data System containing maintenance records for one Damageability of Electric plant, and maintenance records obtair.ed from a coop-I aboratories, SAND 90-0696, May 1991.

crating utility for a second plant. Du ring the 1978-1988 This repott describes the results of a series of tests decade covered by the LER data, which represents ap-proximately 812 reactor years,22 compressor related performed to assess the effects of thermal aging on the and 16 dryer related events that resulted in loss of con' vulnerability of cables to fire-induced thermal damage.

trol air supply were reported. Equipment failure He tests were part of an effort in support of the NRC causes were diverse, with no single type of failure NPAR progtum to identify and investigate fire safety dominating.ne records available for the two commer.

i, sues for w hich plant aging might lead to an increased level of risk.

ctal plants indicated a sigmficant preventive and cor-rective maintenance effort to take care of senice wear From the standpoint of fire safety, cables represent and provide reliable equipment operation.

the single most important class of electrical equipment m a n uclear power pla nt. First, virt ually every plant sys-hiaintenance recommendations included in operat' tem includu power, control, and instrumentation ca-ing and maintenance manuals provided by equipment bles. Second, cable ** pinch" poims (that is, locations manufacturers were reviewed and compared to the where redundant train separation is reduced by the preventive maintenance practices at one plant. De merging of cable routings)often represent dominant user applied practices generally were m confoimance with or exceeded the manufacturers' recommenda-contributors to plant fire risk as determined by proba-bilistic risk assessment (PRA) analyses. %itd, cables tions. One troublesome aspect is ensuring the opera-represent the major combustible fuel loading for raost tional readiness of auxiliary compressors that are nor-plant areas.

mally idle for long periods but must provide backup senice for critical needs if the main control air supply The tests described here examined the thermal deteriorates. hianufacturer recommendc<1 mothball, damageability of two commonly used types of low-ing procedures do not appear practicalIor this applica-Dame-spread electric cables qualified to 1EEE-383:

tion; such friture causes as drive belt set, corrosion of 1.

A Neoprene-jacketed, cross-linked-polyethyl-internal parts, and small intemal water leaks may pre-ene-insulated (XPE), three-conductor, 12 sent a problem when the compressor is needed.

AWG,600V light power or control cable pro-hicasurable parameters that have a potential for duced by the Rockbestos Corporation and enhancing the capabilities for detecting incipient fail-marketed under the trade name brewall III.

urcs and examining degradation trends in compressors 2.

An et hylen e-propy' eae-rubbe r-insulated and dryers were identified. For compressors, they in.

(EPR). chlorosulfonated-polyethylene jack-clude periodic delivery capacity tests, trending of stage eted (CSPE or ilypalon), t#>o-conductor.16 temperaturet and pressures, and motor current signa-AWG, plus shield and drain,600V instrumen-ture analysis. hicasurable parameters for dryers in-tation or signal cable produced by BlW Cable l

NUREG-1377 44

Main Citations and Summaries l

l Systems incorporated and marketed under alvvc the damage threshold. 'lhe aged llockbestos the traJe name llostrad 71!

samples consistently displayed longer times to failure at a given temperature than did the unaged samples, 1or each of the two cable types, both unaged (i.e.,

ndicatu Oss vulnerabihty to thermal damage for the new from the cable reel) and thermally aged umples aged sartDies. The tirne to failure for the aged and were tested. No radiation aging was employed in these unaged lilW samples was not significantly dif ferent for 10515-exposure temperatures at which failure was observed

'ihe exposure conditions simulated during testing in teth aged and unaged samples.

were considered typical of those expected during an lt wmiso not ed that,in virtually every case, failure enclosure hre w hen the subject cables are not involved of the cables through conductor to-conductor shorting in the f te itself. The most signihcant difference be-resulted in the imtiation of intense, sustained, open tween the test evposures and anticipated actual expo-flaming in the cable rarnples. As the cables shorted, sures was that the tests invohed exposure at an ele-S par ks ignited the pases evolved from the cables. In no vated steady. state temperature whereas, m actual ex-cae was spontaneous ignition of the cables observed posur es, equipment would experience a transient time /

prior to electncal failure. These results indicate that temperature exposure.

the fail 6 e of energiicd cables is a rnechanism for fire in these cable exposure tests, the walls of the cham-spread.

ber and the air were preheated to the desired uniform The thermal damage threshold changes observed m steady-state exposure temperature. Two energi/ed ca-the tests on two of the most comrnon nuclear quahfied ble sampics w er e then quickly inserted through a small cables in current use in the U.S. nuclear industry are door to proside a near step change in environment not considered of suf ficient magnitude to significantly temperature for the cable samples.

alter risk estimates for scenarios involving cable ther-

'lhe cable samples were energized by a three phase mal damage.

204011 pow er source. liach of the threc conductors of 11 should be noted that these tests hase not ex-the llockbestos cables was connected to one phase of plored the impact of other fire environment effects the power source. In the case of the lilW cable, the two such as suppressant application and high humidity on conductors and the drain conductor were each con-cable survivalJihe failure threshotA gwen above per-nected to one phase of the power source. Irakage cur-tain to gross elect neal f ailure, b. Inost cases, significant rents between power phases were monitored continu-levels of current leakape were noted prior to gross fail-oudy. The time to ultimate cable failure, as deter-ure, and specthe applications must be examined to de-mined by the failure of a two-ampere fuse in any one of termine w hether such leakage could constitute the fail-the three phase circuits, was also recorded.Two meas-ure of a circuit to perform its design function. Also,be-ures of thermal darnageability can be made based on cause rnixed results were obtained for the two cable these tests-types tcsted, no direct conclusion regarding the impact One measure of fire damageability is the thermal of thermal aging on the fire vulnerability of any other damage threshold defmed,in the context of these tests, cable type can be drawn based solely on the results of these tests.

as a temperature range its upper limit is the lowest temperature at which electrical failure was observed NUltEG/ Cit-5555, W. Gunther and K. Sullivan," A in.

following exposures of up to 80 minutes, the lower Assessment of the Wcstinghouse pWit Control bo limit is the b'phest temperature for which no electrical

1) rive System," lirooknaven National I aboratory, failures were noted following exposur es of no less than liNI A > Ult!!G-52232 April 1991.

80 min"tes.

A study of the effects of aging on the Westinghouse l'or the llockbestos cable, the faliure threshold of control rod drive (CilD) system was performed as part the unaged cable was determined to be 325430*C, of the NitC N1'All program, its objective was to pro-whereas the the: mal damage threshold for the aged 3 ide a t echnical basis for identifying and evaluating the ramples was 350-365'C. l'or the lilW cable, the ther-degradation due to aping.

mal damage threshold of the unaged cable was esti-lhe Westinchouse CitD system consists of control mated at 365-370'C, whereas that of the aged samples rods and the rhechanical and electrical components was estimated at 345450rC. 'lhus the armg process tM wntrol the rod motion. The study exarnined the resulted in the opposite effect on the thermal damage design. construction, maintenance, and operation of thieshold for the two cable products. l'or the llockbes-the system to assess its potential for degradation as the los cable, aging increased the damare threshold by ap-nuclear plant ages and esaluated the extent to which j

pioximately 2545"C while, for the IllW cab!c. it de' agmg could affect the safety objectives of the system.

I creastd the threshold in approximately 20 C.

Studies ate also being conducted for the Combustion l

A second measure cf thermal damapeabihty is the lirgineenng. llabcock and Wilcox, and General filec-i relatise time to failure for exposure temperatures trie CitD 3pttms 45 NUltliG-1377

l Main Citations and Summaries

'Ihe operating experience for CRD systems as tem. Only a few plants are using circuit monitonng documented in the lacensee !! vent Reports (1.l!Rs),

techniques or nondestructive testing to monitor the Nuclear Plant Reliability Data System (NPRDS), and long term operational characteristics of the CRD sys.

Nuclear Power 11xperience (NPE) data bases was re-tem (a substantial portion of that system is considered viewed.1hese sources prmided an average of 30 not related to safety).1he reslonses from plants fur-unique failure eventsperyear over the last 10 years, of ther indicate that seme plants have modified the sys-which approximately 359'c ucre directly attributable to tem, replaced corr.ponents, or expanded preventive aging-related degradation.1hc review resulted in the maintenance. Several of these activities have effec-following observations:

tively addressed the aging issue. However, mainte.

1.

1hc majority of the n ported failures occurred nance practices appear to vary from one plant to an-in the electrical area,i.e., the power and logic other, possibly reflecting inadequactes at some plants.

cabinets, and the rod-position indication sub-Techniques to detect and mitigate the effects of ng-

system, ing, including advanced approaches by Westinghouse, 2.

Approximatcly 409o' of the reported failures the Japanese, and the French are described. Tresious resulted in a rod drop. which usually chal' research related ta the Westinghouse CRD system was lenges the reactor protection system and initi-discussed, e.g., NUR110-0641 on wear of control rod guide tubes, study 11613 on localized wear frorn the ates a reactor inp.

NRC Officeof Analysisandlivaluationof Operational 3.

Several failure modes such as rod position Data, and lipRI sponsored reports on plant life exten.

dnft and overheating of power cabinets are sion and control rod lifetime determination. Research common to many plants, which could indicate on the extension of plant life fora Westinghouse pWR, the need for generic resolutions, for example, identifies the latch, drive rod, and coil The normal operating and environmental stresses stack asserriblies as limited life components.

experienced by the system components were assessed

.lhe indings and recommendations of this aging stud / ma7 e summarized as follows:

b to determme their effect on the long term perform-ance of the system. For example, the regular stepping 1.

Aging related degradation of the Westing-action associated with control rod motion results in house CRD system can compromise the in-wear of the latch and drive rod components and in elec.

tended function of the system. 'Iherefore, trical surges on the control rod drive mechanism coils, means to detect and mitigate this degradation

'1he amount oflatch wear measured in another study,

in the safety and non safety related portions presented in this report along with results of other rc.

of the system should be pursued.

search efforts related to aging of the CRD system.

2.

The test requirements on the system (e.g.,

Other examples of stressors associated with aging.re-rod drop timmg) are important in determining lated degradation are high temperature, flow induced the operational readiness of the system, al-vibration, and particle debris carried by the coolant, though they cause some incremental wear on A failure modes and effects analysis of the Westing, the mechanical part of the system.

house controt rod drive system was aiso conducted, and 3.

The preventive maintenance, including in-components with a high safety significance y cre identi-spection and testing of the in containment ca.

i fied along with the likelihood of their failure.1his as-bles,. connectors, and coils should be in-l sessment was based on operating experience data and creased as these components age.The use of an evaluation of the susceptibility of the components to such predictive maintenance concepts as non-age related degradation. Several components that intrusive on-line monitoring techniques should receive attention as a plant ages were identified:

should be considered.1he method used to cables, coils, and connectors (in containment); latch as-perform the rod drop timing iest should be scrnbly; guide tube; and selected electronics within modtfied, or the degradation that can result power and logic cabinets, including the rod position in-from the present procedure of pulling the dicating system.

fuses should be accounted for.

- An evaluation of inspection, surveillance, monitor.

4.

The logic associated with the speed and mo-ing, and maintenance was accomplished with informa-tion control of the CRD system is complex <

tion from fifteen plants representing ten utilities. Re-Maintenance errors in this area have resulted sponses from most plants agreed that two of the re.

in unnecesssary reactor trips and additional quired tecimical specification tests (rod-drop timing stress to the CRD system. Repair and replace-and rod excercising) are beneficial in venfying the op-ment procedures for this portion of the system erational readiness of the system. Preventive mainte-should be evaluated for completeness and ac-nance activities for electrical components within con-curacy, and personnel training should be em-tainment dominate the overall maintenance of this sys-phasized.

NURiiG-1377 46

,---_-a--

--aa

~ ' - - ' --

hiam Citations and Summaries i

NUltt:G/ Cit-5558,11. Steele, Jr., K.G. DeWall, and flow interruption tests required more torque and sub-J.C. Watkins " Generic issue 87, I'lexible Wedre sequently more stem force to close than would be pre.

Gate Valve Test Program: Phase 11 liesults and dicted using the standard industry motor-operator siz-Analysis," Idaho National lingineering lateratory, ing equation for disk load calculations with a common 1100-2600, January 1991.

coeffic ent of friction.'the highest kiads recorded were the result of internal valve damage caused by the high-Qualifimtion and flow isolation tests were con-ducted to analyze the ability of selected boiling water differential.pt essure loads across the valve disk as it at.

reactor (llWil) process valves to perform their con-tempted to stop the flow.

tainment isolation f unctions at high-energy pipe break

'the high loads encountered during the test series conditions and other more normal flow conditions. Nu-raise the concem that some valves installed in nuclear merous parameters were measured to assess industry power plants may not have large enough motor opera-practices for predicting valve and motor operator re-tors to e nsure closure in the event of a design basis acci-quirements. ~lhe valves tested were representative of dent.

those used in llWit reactor water cleanup systems and

.lhe study into the phenomena affecting the stem high pressure coolant mjection (llPCI) steam lines.

loads in a motor-cperated gate valve continues. Ilow.

Among the objectives of this r ucarch program are i ever, the results to date indicate that the phenomena determine what factors affect the performance of mo-tak!ng place in,ide the gate valve are more complex tor-operated gate valves and to deternune how wellin-than previously thought.'lhe actual disk f actor is much dustry s analytic tools predict that performance.

higher than previously beliesed, but this factor can be

'this program supports the NitC's effort on a ge-moderated for some valve applications once the self-neric issue, GI-87,

• Failure of 1IPCI Steam 1.ine With-closing force balance on the valve disk is understood.

out isolation." GI-87 covers three tviling water reac-Physical inspection indicated that these valves were tor process lines: the llPCI turbine steam supplyline, very near their physical fragility limits at design basis the reactor isok. tion cooling (ItCIC) turbine steam conditions. The excessive bearing pressure between supply line, and the reactor water cleanup (ItWCU) the disk and the body guide materi:ds resulted in yield-ptocessline. All three of these processlinescommuni-ing, spalling, and gouging of the surf aces. In some of cate with the primary system, pass through contatn-the designs, the guide clearances were large enough to ment, and have normally open isolauon valves. Phe allow the disk to tilt during closure, which resulted in concern with the isolation valves is whether they will s gnificant damage to the sealing serfaces.

close m the event of a pipe break outside of the con-tainment. A release of high-energy steam or hot water NUlti:G/ Cit-5560, ll.ht. liashemian, D.D. lleverly, in the auxiliary building could result in common-cause D.W. hiitchell, and K.hi. Petersen, " Aging of Nu-failure of other components necessary to mitigate the clear Plant itesistance Temperature Detectors,"

Analysis and hicasurement Services Corporation, accident.

June 1990.

One of the major parts of the research program in-A mWWye research and development pro-cluded two full scale qualification and flow interrup-ject on aging of narrow-range resistance temperature tion test programs on flexible-wedge gate valves' h

TD) used in the primary coolant system of I hases I and 11, lhe Phase !! program was performed pressurized water reactors was carried out as part of m 1989 at the Kraftwcrk Umon (KW U) facilities near the NitC NPAll programJihe goal was to establish the 1 rankfurt, Germany. Among the valves tested, thre long term performance limits of these llTDs in order were 10-m. valves typical of those used in the i11 CI ap-to venfy that objective and adequate measures are im.

pbcations. One of the b-m. vahes was also tested at plemented to ensure safety, itCIC test conditions. In all, seventeen flow interrup.

'Ihe project was conducted in two phases. Phase I, a tion tests were perfonned, seven at design basis condi.

six month feasibility study, was completed in Junc

tions, 1987. The results, published in NUltliG/ Cit 4928,

.I.wo itWCU valves were tested during the earlier Degradation of Nuclear Plant Temperature Sensors,"

Phase 1 Test Program. As a result of that work, it was demonstrated the need for additional work in Phase 11, expected that the vahes would require more stem

.Ihis report presents the results of Phase 11, which was force to close than mdusry normally would have pre-conducted over a 30 month period beginning in Octo-dicted Therefore for the Phase 11 Program, the mo-bet 1987. 'lhe work involved laboratory testing of 72 tor-operator control switches were set at higher-than.

nuclear grade itTD clements representing several normal torque values to ensure valve closure, and the from each of four U.S. manufacturers.The limit for the strengths and weaknesses of a given valve design wete nitial accuracy of these IITDs was established, and a determined from the recorded data' procedure for performing precise calibration was de-

'lhe test results clearly showed that, for the GI-87 veloped. lixperimental aging of 30 of these itTDs at concerns, all valves that w cre subjected to design basis simulated reactor conditions resulted in five failures NUltliG-1377 47

Main Citations and Summaries and six major calibration shifts. Two failures occurred NUREG/CR-55ft3, M.S. Kalsi, C.I Ilorst, l.K. Wang.

in thermal aging, one in vibration aging, one in hurnid-and V. Sharma.

• Prediction of Check Valve Per-ity aging. and one in thermal geling.%c remaining 19 formance and Degradation in Nuclear Power Plant ifIDs performed well during the aging tests, maintain.

Systems-Wear and Impact Tests. Final 11eport, ing a drtit band ofi0.2'C.

September 1988 April 1990," Kalsi Engineering, I

Re shelf life drift of RTDs was also quantified.

Inc., Klil No.1656, August 1990.

%is involved testing 45 RTDs for storage effects: 24 that had been in normal storage at various nuc! car Check valve failures at nuclear power plants in re-power pla nts for periods of one to five years and 21 that cent years have led to serious safetI concerns and have were aged in the project. %c test resuhs for these 45 caused extenske damage to other plant components R' ids showed a shelf life drift band of i0.1'C. Most that had a significant negative impact on plant avail-of the storage drifts, the failures, and the normal aging ability. Swing check valve internals may experience drifts were found to occur in the first few months of ag, prernature deterioration if the disk is not firmly held ing. A potential remedy is to burr '.t the RTDs before oren against its stop. At the present time, no guide-they are calibrated and installed in the plant, lines exist for the prediction of degradation trends and he performance of nuclear plant RTDs is evalu.

the determination of suitable inspection intervals. A ated by response-time test!ng in addition to calibration, research program aimed at developing a reliable model These two procedures are independent and are there.

for quantitative predictions of wear and fatigue for fore done separately.The nuclear industry has about swing check valves was established as part of the NRC ten years of experience with R'ID response time result +

NPAR program to improve the safety and reliability of ing from periodic in situ measurements made in about their operation. His report covers Phase 11 of the re.

60 PWRs at least once in every fuel cycle. Representa-search on swing check valves. The work in Phase i was tive results of these measurements were reviewed to published in NUREO/CI'-5159.

identify the range of achievable response times and the response time degradation modes.

He goal of Phase II was to develop predictive mod-i Several coramercial grade RTDs were also aged els that could be used to quantify the degradation of and tested for comparison with nuclear grade RTDs.

swing check valves with flow disturbances close up.

De results showed that the average response time and stream of the valve at flow velocities that do not result calibration stability of nuclear grade RTDs is about in full disk opening. Two major causes of swing check twiec as good as that of the commercial grade RTDs.

valve failure are premature degradation due to wear in ne project addressed the following additional top.

the hinge pin and faugue in the disk stud connection to ics: sources of errors in R1D calibration, factors affect-the hinge arm. Accelerated wear tests were performed ing RTD accuracy and response time, failures of RTDs using aluminum hinge pins and bushings in 3 inch and as reported in the IliR and NPRDS data bases, and the j

International Temperature Scale of 1990 and its im.

6-inch valves to quantify wear experienced in the hinge pin area. A special disk instrumented with strain gages -

l-pact on ternperature measurements m nuclear power plants. The results of research performed in Phase 11

- was used in the 6 inch valve to measure the impact i

did not reveal any unanticipated or rnajor systematic forces and their rate of occurrence to quantify the fat aging problem in the performance of the RTDs tested.

tigue damage caused by the disk tapping against the Ihe nuclear industry s practice for verifying adequate stop. He wear and fatigue prediction models devel.

R1D accuracy and response time is to perform on line oPed in this PtoSram show Eood correlation with labo-cross cahbration and loop current step response tests at r tory test results as well as with a limited number of least once every fuel cycle,in light of the data obtained check valve failures at plants that had been sufficiently.

throughout this study, this approach is reasonable for documented.

I managing the aging of RTDs that do not have any ma.

jor design, fabrication, or installation deficiencies.

The results of this research allow inspection and RTDs that consistently maintain a suitable calibration maintenance activities to be focused on those check-and reponse time as determined by periodic testing can valves that are more likely to suffer p cmature degra-be used in the plant for their qualified life as specified dation.He methodology for quantitat{ive predict i

by the manufacturer. The manufacturers'specifica, wear and fatigue can be used to develop a sound and tions for the qualified life of nuclear grade R1Ds typi, effective preventive maintenance program.The results cally range from 10 to 40 years depending on the manu.

also indicate certain modifications in the valve design facturer and the conditions at which the RTDs are that may improve check valve performance and reli-used.

ability.

NUREG-1377 48 j

hiam Citations and Summanes NUltt:G/ Cit-M12, PX Samanta, W li. Vestly, F, lisu, component condition and signal for the correction of and ht. Subudhi. " Degradation Modeling with Ap-degradations before they hase significant impacts on phcation to Aging and Maintenance Effectiveness rehability and rtsk. In addition, the methodology was livaluation." llrookhaven National I aboratory, used to develop initial estimates of the ef fectiveness of IINL-NUllEG-52252, March 1991.

maintenance in preventing degradations from becom-An important element of the assessment of risk as-ing failures.

sociated with aging in nuclear power plants is the un-Specific applications of the theoretical approach dcrstandmg of the aging phenomena associated with resulted in quantitative models of comp nent degrada-components of safety systems.1his report describes a tion rates and component f ailure rates derived from r,tudy of aging pMnomena at the component levelin plant. specific data. As part of the data analysis, statisti-support of the NitC NPAR program to develop an ag-cal techniques that identtfy aging trends in failure and ing reliability model representing the agmg proecss ex-degradation data wer e developed. 'the aging trends can perienced by components m nuclear power plants un.

be of any kind and can exist in any segment of the data.

der presently existmg test and maintenance practices.

Specifically, an analysis of residual heat removal A new model was developed to process information on (R1IRhystem pump data shows a " bathtub" curve for wmpment degradation in order to analyze the degra-the de;radation rate where a Gstinct increasing trend dation procci.s and its implications.1he focus was en is observed at the later ages. Interestingly, the pump modchng the degradation rate,i.e., the rate at which f ailure rate does not shown any increasing trend for the degradations occur, with the specific objective of devel-same perioJ, which demonstrates the need to identify oping explicit relationships between degradation chat.

aging trends through analyses of component degrada-attenstics and the compment failure rate.

tions lhe research program goes beyond an analysis of

'these results are important first steps in showmg times of degradation and failure. l'irst, theoretical that degradations can be modeled to identify aging ef-models that relate the degradation rate of the comFF fcctL 1he theoretical methdology that was descloped nent to its failure rate are developed. With the rela-represents an advancement demonstrating that degra-tionships derived, information on component degrada-dation characteristics are eglicitly related to failure tion can be used to predict the component failure rate rates and hence ultimately to risk. The next step would and its significance. Specifically, this methodology can be to use the methodology and statistical techniques to use aging trends in the component degradation rate to develop and validate practical pn>cedures for predict-predict future aging trends in the component failure ing fatture rates due to aging from degradation data.

rate.

'this ability would provide powerful tools for analyzing

'lhe capabihty of making such a prediction is impor.

aging effects in terms of degradation data and for pre-tant because mformation on commment failure rates dicting their implications for reliability and risk, due to agmg is rcquired to quantify the effects of aging NUltEG/CR-%19, S.P. Nowlen, "1he impact of 1her-on core damage frequency and nsk.1his information is mal Aging on the I'lammabihty of lilectric Cables,'

also needed to quantify the effectiveness of a given Sandia National l aboratories, S AND904121, maintenance program in controlling the effects of ag-March 1991, ing on the core damage frequency and risk. Ilowever,

'lhis report describes tests on the fire vulnerabihty failure Cata are often sparse. On the other hand, degra-of aged electrical components performed for the NRC dation data are more abundaA because degradations NPAR program. The objective was to identify and in-occur at a higher rate than do failures.1hus the ineth-vestigate issues of plant aging that might result in an odology deTeloped in this report allows component increased fire risk at commercial nuclear power plants.

failure rates due to aging to be estimated from compu lhe particular issue investigated in these tests is the nent degradation rates. This has the potential of impact of thermal aging on the flammability of electri-greatly mcreasing the quantity and accuracy of compo-cal cables.

nent failure rates due to aging for use in risk evalu-The cable insulation represents the dominant ations of aging cifects, source of combustible materials in most nuclear power It is important that,in addition to the identification plant areas. Current USNRC standards require the use of aging trends in degradat on and failure data, the of low flame spread cables, as certified by the methodology allows maintenance indicators to be se.

IEl!!!-383 qualification standard, in all new installa-tions. Ilowever, should these cables lose their fire-lected in such a way that component degradations are related to impacts on reliabihty and nsk. When the retardant properties as a result of material aging, an in-crease in fire risk could result based on the role cable degradation indicators show significant impacts of deg-radation on the comgment failure rate and the result-installations have played in past fire risk assessments.

ing nsk. maintenance should be performed to correct To assess t his issue, four large-scale cable flammability the degradations. Thus the degradation indicators can testswere performed,Two commonly used types of nu-provide a practical and effectise means of monitoring clear grade electrical cables were tested in both the 49 NUREG-1377

hlaln Citations and Summaries new (unaged) and a thermally aged (through acceler-i ated aging)conditiom have not been tested,it is expected that similar gesults would be obtained. No f urther im estigation of this is.

l.

llockbestos l'IR11 Wall 111, 3-conductor,12 sue is recommended.

AWG, Neoprenc jacketed, cross linked poly.

ethylene (XPli) insulated light power or con

• NUltrG/CR-5655, hi.J. Jacobus and O.F. Fuehrer, trol cable, and

• Submergence and Illgh Temperature Steam Test-ing of Class 111 lilectrical Cables," Sandia National i

2.

11oston Insulated Wire (!!!W) llostrad 711, laboratories, S AND90-2629, biay 1991.

2 conductor with shield and drain,16 AWG, blany types of cainic are used throughout nuclear Ilypalon jacketed, ethylenc propylene rubber power plants in a wide variety of applications. Cab!c (llPR) insulated instrumentatitm cable, qualification typically includes thermal and raJiation Iloth of these cables are certified nuclear grade ca-oging intended to bring the cable to a defined "end of.

bles, meluding certification as low Dame. spread ca-Itic" condition before exposure to a simulated design-bles they are among the most commonly used cable basis accident. In some instances, cables must be quali.

types in U.S. commercial reactors, fied for submergence conditions. liigh. temperature Since these cables were certified us low flame steam testing of cables (beyond the design basis)is not spread, they were expmed to a fire that was more sc*

currently required for qualification.

vere than the standard exposure test. His exposure This report describes the results of high tcrnpera-was based on work performed by Factory htutual Re-ture steam testing and submergenec testing of 12 dif-search Corporation (Fh1RC). Fh1RC found that a dif-ferent cabic products. De cable products tested are ferent gas burner (fire source) configuration widt two typical of cables used inside containments of U.S. light cable trays placed face to face with insulating backet wat er s cattors and include prirnary insulations of cross-boards (as compared to a single open ladder tray used linked polyolefin (X110), ethylene propylene nibber in the standard test) would produce enhanced fire (liPR), silicone rubber (SR), polyimide, and chlorosul-pro agation. De tests described here used a similar fonated polyethylene (CSpli).

con iguration to induce flame spread in the sample ca-cablu were part of a larger test program in b es did not burn during testing, which four sets of cables u ere subjected to simultane.

tI uld le m ed ous thermal and radiation aging for 0 (unaged),3,6, During each of the four fire tests, it was obsened and 9 months. Following the aging, each set of cables that essentially all of the availab!c combustible materi.

was exposed to a simulated loss-of coolant accident als (the cable insulation andjacket materials) were con-(LOCA).

sumed by the fires Flame propagation to the full ne submergence test was performed on the cables height of the 16-foot vertical cabic trays was observed that had been aged for 6 months and then exposed to in all cases. Ilowever, up(m crammation of the test the simulated LOCA and the high-temperature steam data, it was found that, for both cable types, the aged test was performed on the cables that had been aged cable samples displayed a reduced nammability as for 3 months and also exposed to the I OCA. Iloth of compared to the, unaged cable samples. %is was re-these tests were added to the scope of the test program flected in reductions in both the rate of rise and the because the aged cables had completed all planned peak value of the measured fire heat release rates for testing and many of the cables had not yet failed. Mc the aged cable samples as compared to those for the unaged cables and the cables aged for 9 months were unaged cable samples.

not involved in either the submergence testing or the

%csc results indicate that, at least for the two cable high-temperature steam testmg and are therefore not types tested, thermal aging resulted in a decrease of discussed in this report.

material flammability, llence, for these two cable The submergence test used a solution close to that types, the issue of material agmg and cable flammabil-specified by ll! Ell 383-1974 for chemical spray during ity is not of concern. The use of material flammability I OCA simulations. He solution was maintained at parameters obtamed from tests of unaged cabic sam' about 95'C during the exposure, which lasted a total of plcs w,ll therefore provide conservative assessments of 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />. The high temperature steam test involved i

material llammability in a thermally aged condition.

crposure to steam at temperatures as high as 400'C (750'F). Cable insulation resistances were monitored These results are consistent with results of previous throughout the high temperature stearn test and at dis-cable aging studies. It has been observed that the proc-crcte times during the submergence test. Dielectric ess of thermal aging tends to drive off certain of the vithstand testing was performed before the submer-more volatile constituents of the cable insulation mate-gence and high temperature steam tests and at the end rials. Dis will leave less of these compounds available during a fire to support combustion; hence flammabil-of the submergence test. He cables that passed the post submergence dielectric test were subsequently ity is reduced somewhat. Although other cable types wrapped around a mandrel with a diameter 40 times NUREO-1377

$0

}

Main Citations and Summaries that of the cah!e and esposed to a fmal dielect te with-ognized that operation of a pump under low flow con-dttions can result in hydraulically unstable conditions stand test.

that can damage the pump, even though the rate of lhe conclusions from this study are:

flow is adequate for heat removal, 1.

'the results of the high temperature steam Nuclear Regulatory Commit.sion (NRC) llulletin te,t indicate the approtirnate thermal failure thresholds for each cable type. liPR cables 88-04 required utihties to examine (1)the potentialfor dead-headmg of pumps due to parallel pump competi-generally survived slightly higher tempera.

tion and (2) the adequacy of the minimum flow rate tures (370-400*C) than XI.PO cables (299-388'C) during the high-temperature provided for each safety telated pump. Utilities have reviewed the curreritty recommended minimum flow steam exposure. The XLPO insulated con.

rates with pump vendors and have examined existing ductors had noinsulation lef t at the end of the high temperature steam test. Silicone rubber system design provisions, operating controls, and his-failed in the range of 396' to 400'C, Ketite torical maintenance experience.

IV at 372

• to 382'C, and polyimide at 399'C.

Under the auspicesof the NRC'sNuclear Plant Ag-2, The results of the submergence test indicate ing Research Program, Oak Ridge National labora-tory has reviewed utility responses to llulletin 88-04.

that a number of cable types can withstand An assessment of the industry response and resultant submergence at elevated temperature, even conclusions and recommendations are presented.

after exposure to a loss-of coolant accident simulation. XI.PO cables generally performed PNic5722, D.11. lilahnik and R. l Goodman, better than EPR cables in the submergence "Op ling lixperience and Api,ng Assessment of test and in the post submergence dielectric C

rl$.

testing. lly the end of the final dielectric test at r (af ter the mandrel bend), only 1 of 11 (9.1%)

1his report provides a preliminary aging assess.

XLPO. insulated conductors,17 of 20 (85%)

ment of safety related room coolers for the emergency EPR insulated conductors, and 6 of 8 other core cooling system (ECCS) pump rooms in nuclear cables (silicone, Kerite IR, and Rockbestos power plants. The assessment conforms to the NRC coaxial) had failed, AR pmgram strateg and is based on limited 3.

A number of cables that performed well dur-mformation obtained through public and private data ing the submergence test failed post submer-bases, equipment vendors. utihty contacts, literature gence dielectric withstand testing (either be-se rehes, and expert optmon.

forc or after the mandrel bend). 'this indicates Description of the ECCS pump room cooler sys-that the IEE!! 383 dielectric withstand tests tems were based on FS ARs and vendor-supplied infor-and mandrel bends can induce failure of oth.

mation. Data from lliRs, review of maintenance te.

crwise functional cables. Note that this con.

clusion does not imply a criticism of the IEEE quests at a reactor plant, and discussions with person-nel that do utdity repair and rnaintenance work were 383 requirements, which are intended to pro, used to determine the operating experience of pump vide a level of conservatism in the testing, room alm. Fahre modes, causn, frequency rates, 4.

'lhe IEEE 383 dielec'cic withstand tests are and methods of detection are summanied from the op-very severe even if a mandrel bend test is not edng rec rds. Maintenance actions and modifica-perfortaed. This is evidenced by the failure of tions needed as a tesult of the operator expenence are nine conductors and the near failure of three addressed. Operational stressors are summarized, more in the post. submergence dielectric with-manuf cturer recommendations for maintenance and stand test, only two of which were showing a surveilance are listed, and aging and senice wear a

strong indication of degradation during the mon toring are briefly evaluated.

submergence test.

PNI 6287 K. R. lloopingarner, IL J. Kirkwood, and NUREG!CR-5706, D.A. Casada, "NRC !!ulletin 88-00 P. J. leniecky, " Study Group Review of Nuclear Potential Safety-Related Pump loss-An Assess.

Senice Diesel Generator Testing and Aging Mitiga-ment of Industiv Data." Oak Ridge National I abo.

tion " Pacific Northwest laboratory, March 1988.

ratory, ORN!4671 June 1991.

Nuclear utility plants are required to periodically As part of the NPAR program, the Pacific North-test safety-related pumps to demonstrate proper fune-w est Iaboratory is performing a diesel generator aging tioningof the pump. Ilistorically,a substantialnumber assessment study. In the on going NPAR Phase !! of l

of these pumps have been routinely tested at the flow the aging study, efforts have been focused on aging rate available through the pump's minimum flow recir-mitigation and other success strategies for improving

'I culation llow path, which in many cases was sized to nuclear plant diesel generator operation and mainte-

)

avoid overheating only. It has become more widely rec-nance and also increasing its rehability.

NUREG-1377 5l

Main Citations and Summaries A study group of diest.1 experts, the nothors of this 6.

Address fuel oil storage management to per.

report, met on April 29 nnd 30,1987, to resolve issues mit flexibdity and the use of a large fraction of on mitigating diesel generator aging and improving op-stored fuel before replact, ment, crations, tening, and maintenance. 'the focus of the study gcoup was to (1) adJten the diesel generator ag-litiminate many unnecessary and partially re-7.

dundant tests and engine starta in the ing streswrs resulting from the present peralic testm8 18 month test perks!(includmg those due to practices of the nuclear mdustry and (2) propose po*

tential mitigating measures. A new recommended test.

fah.c signsis),

ing program was developed and is documented in this 8.

litiminate, where possible, short engine run report.1he report lays out the condusions and recom-times and excessive idle times.

mendations of the study group.1hc caperts agreed

'Ihe technical bases for such changes to the specifi-that, if these recommendations are put into practice, cations are obtained from the NRC NPAR program many of the engine aging stressors (e.g., those due to and from research sponsored by the !!lectric Poact Re-fast start)could be reduced or climinated; another con-scarch Institute (liPRI)and the Nuclear Safety Analy-st:quence could be a reduction of fattures and an im-sis Center (NSAC) operated by !!PRI. Fast starts, fast provement in operability and rehability.

loading, and the large number of test runs are cited as acting to increase diesel generator stress and wear.

IWI-7516, K.H. Iloopingarner, *limergeng Diesel

'the results from this study confirm these stressors and Generator Technical Specifications Study Results/'

add enessive testing loads as another important stres.

Pacific Northwest Iaboratory, March 1991.

sor, 1his teport covers a study in support of the NRC pst,SA-lM07, A.ll. Johnson, Jr., D.ll, Jarrell, U.P.

NPAR program on the effects of aging on emergency dicsci generators (l!DG). 'the research was perforrned Sinha, and V.N. Shah, Understanding"and Manag-ing Corrosion in Nuclear Power Plants, Pacific in two phases. Phase 1 used plant operating experi-Northwest laboratory, August 1990.

ence, data, expert opinion, and Statistical methods to produce a new data base related to aging, reliability,

'the concept of understanding and mariaging corro-and operational readiness of nuclear service diesel sion in nuclear power plants is not new-in various generators. Phase 11 was chieflyconcerned with rneas-forms, this main theme of the report has been applied urcs for mitigating the effects of aging.

throughout the development and maturing of nuclear insights from a number of sources indicate that technology. Too often, however, understanding corro-there are many opportunities for improving the man-en has been based on reacting to it rather than on an-agement of liDG systems. I!xisting technical specifica-ticipating its occurrence. Regulatory and utility mitia.

tions, for example, could be modtfied to yield signifi-tives are creating a climate and framew;ork for more ef.

g cant safety benefits by reducing direct effects of aging fcctive application of the concept.1his report charac-

)

and increasing system reliability, 'lhus technical speci-terizes the framework and provides some illustrations fications related to the management, testing, and reli-f how We concept is being applied in support of the ability of emergene,' diesel generators were reviewed.

NRC NPAR, lleneficial specifications were identified as were those Although corrosion has not caused a major accident that could adversely influence aping and reliability, in a nuclear power plant, it has been a continuing cause I,otential improvements m. techn.ical specifications of an overall untimelydegradation of the plants and,in and engme and system management aimed at reducing particular, of nuclear safety-related systems and com-egmg effects and increasing reliability would:

ponents, often resulting in eactor shutdowns and ex-tendedoutages. !!xamplesof majorcorrosion induced 1.

Significantly reduce the number of total en-degradation include intergranular stress corrosion gine starts, craciing (IGSCC) of piping in boiling water reactor 2-Reduce the k)ad application rates for testing u adon bypass systems and denting, pitting, inter-r purposes by gradually adding load, E'""d"' "*C# ^)' "

• * **

• E C " " '"

tubes in pressuriicd water reactors. 'These types of 3.

Reduce the liDG testing loads 10 90% of the degradation involve significant phenomena that have continuous hiad rating or to the plant emer-been wid:ly recognised and investigat ed. Major invest-gency unit load, whichever is less, ments have been made to detect and mitigate the ef-t.

Increase the maximum 11D0 start time to 25 fects of these corrosion mechanisms. Numerous other to 30 seconds, mechanisms and phenomena have been observed in nuclear systems and components; some are obvious, 5.

Make recessary changes to support the reli.

some subtle.

ability emphasis of Regulatory Guide 1.9, Re-

'the failure of a valve to open or close on demand, vision 3, and delete statistical emphases, for example, may be a consequence of corrosion. 'lhis NUREG-1377 52

hiain Citations and Sutnmr ries ternperature dose tate data to supcipose when shif ted and similar considerations increase in importance as to the reference tempetature. 'the resulting super-plants age but take on an additional dimension with posed curve at the teference temperature extends to consideration of extendmg the licenses of the plants snuth low er dose rates that are experimentally inac(es-beyond the current 40 par period. Hoth the nudcar sible because of the long time petiods that would be r e-utilities e.nd the NRC are considering in detail the deg.

quired to simulate aging.'this procedure therefore al-radation inechanisms that may have spectal signift-lows meanmgful predictions to be made for long term, cance in hcense renewal.

hedose rate radiation agingconditions. Usmghistori-Mitigation of corrosion impacts in nuclear facihties cal data from Sandia's radiation-aging prograrn on nu-must involve more than technical considerations. Itclear pmer plant cable materiah, the authors have must involve attitudes of alertness and commitment on successfully applied the time temperature dose rate the part of reputators, plant management, and the superposition approach to four different materuit plant work force. It also requires timely and ampic ab hypalon, neoprene, pdyethylene, and PVC ;acket ma-heation of resourect tenal For two of these inaterials, extrapolattd predic-1here is a trend in the nuclear industry to advance bons based on the superimposed data weie f ound to be the managernentof agingphenomena,includingcorre I"C"'

"W' 'C

• C "Wd' D'" I" d* '"* ""'

sion. The fact that new corrosion phenomena ermtinue dear power plant wh to emerge, however, provides evidence that principics WYiJ,60103-X J,1; Gleason, R. A. Del'our, J. M of corrosion control still need to be aggresshely ap.

llammond, and P. A.1,0beski,'"1est Plan for the plied. I cssons learned in current reactor operation Comprehenshe Armg Assessmcnt of Circuit lircak-1 4 A ng Huanh need to be systematically and ef fectively applied to ed en and yays for Nudcar 1,, ar,)le ahoratones, (Ni AR)I rogram, Phase 11, %

tended operation and advanced reactor designs.

~

fluntsvine, Al, July 19A9

'lhis report defines the concept of understanding

'lhis entry refers to seven individually bound re-and rnanaging corrosion, references relemnt regula-ports, each presenting the test plan for a specine type tory and industry initiatives, and focuses on an ovet-view of how the concept is being applied, drawing on of circuit breaket or rclay:

results from the NPAR_ program. The ovetwiew in-60103-1 Molded case circuit bicakers cluder, a brief survey of corrosion impacts on major 60103 4 MnaHad eucmt breakers struct ur es, systems, and components, including service 60103-3 Auuliary relays water, steam generators, piping, and contamment.

N)l03-1 Control relay s Mitigation methods are briefly eviewed. The overview NH03 $

Protective relays is referenced to a majc.r data base that is being devel-60103 4 Timing relays oped to assist both utthties and reguhitors in the impor-60103-7 lilectronic relavs tant and respcmsible task of understandmg and manag-The purpose of these reports is to provide details of ing corrosion and other degradation mechanisms in nu-the tests planned for the types of cucuit breakers and clear plants. An oficctivenpplication of understanding relays under investigation m Phase il of the Compre-and managing these mechanisms is crucial not only to hensive Aging Assessment of Circuit lireakers and saie and economic operation of the nuclear plants, but RelaysJihis work is being performed by Wyle I abora-also to public perception of a sounoly designec, man.

tories for the NRC NP AR program, which is intended aged, and operated technology, to resolve technical safety issues related to the aging SAND 88-0754 UC-7H, K. T. Gillen and R. L Clough, degradation of electrical and mechanical components,

" Time Temperature Dose Rate Superposition: A safety systems, support systerns, and civil structures Methodology for Predicting Cable Depradation Un*

used in commercial nuclear power plants. The aging der A,mbient Nuclear Power Plant, Agmg Condi-period of interest includes the period of normal H-Luins, handia Nationallaboratones August 1988.

censed plant operation as well as the period of ex-tended plant life that may be requested m utility appli-Time temperature superposition is au empirical cations for heense renewals.

approach that has been used in polymers for more than The PhaseI report, NURl:G/CR-4715," An Aging 30 years to make thermal aging predictions dunng ev Assessment of Relays and Circuit lireakers and System perimentally inaccessible times. Given the historical Interactions," showed that relays and citetut breakers success of time temperature superposition, the are important nudear plant components that are sus-authors have expanded this approach for combined ra-ceptible to degradation with time. Thus Phase ll, a diation-thermal environments, yielding an empirical comprehensive aging assessment of relays and circuit time temperature-dose-rate shilting procedure. The breakers, was implemented to provide (l) a review and procedure derives an isothermal curve for a given venfication of improved inspection, surveillance, amount of material damage versus dose rate at a se-monitoring, and mamtenance mcthods; Q)in situ en lected reference temperature. 'Ihis is done by fmding aminations and data pathenny for operating equip-the Anhemus activation energy that causes higher-NURIG1377 53

i

~

Hain ('itations and Surntnaries ment; (3) pmtsenice cuminations of naturally areJ Methods are available to detett and initirate at ng i

components or wmgxments with sitnulated derrada-derradaoon and thereby to rmnirnite its unpact. ~lhe llori; (4) arl evaluJtion of the role of (naltltenance in r(ports Llcserabe the baLkrround of the rebearth mitigatmg the ellects of armg;(5) evaluations of ineth.

stratery, list and elaborate on the objectives of the re-ods for predicting residuals and service hie; and scanh. and delme the testing to be performed on natu-(6) recommendations for using rescanh results in the rally aped and derraded equipment m order to deter-regulatory procen Specific roals of the program are:

tuine the inethods most (Ifectise for detecting are der-

1. To identify and(haracterite aging cffects that,if radahon. [inphasis has ban placed on identifymp and unc hecked. could cause derradation of eomtw t haracterinng the met-hanisms of tnaterial and eornpo-nents and subsystems of circuit breakers anJ nent degradation during service and using researc h re.

relap and thereby impair plar.t safety, sults in the regulatory puccu.

2..I

.Jenufy methods ofinspection, surveillance,

'ihe testing consists of perforrning and evaluating oi and monitoring and to evaluate the residual hfe various methods of inspection, surveillance, wndition of components and subsystems of circuit break-rnonttoring, and maintenance, including simulated ers and relays that will ensure timely detection depdahon, to aid in determining the usefulnew of of signiheant agmg effects before loss of their these methods for managing toc effects of agingon safe safety function.

plant oper ation. 'lhe devictu hoten f or testmg ar e rep-resentatrve of circuit breakers and relap that have

3. To evaluate the effectivenen of storare, main-been in use in nut! car power plants New, used, and tenance, repair, and replacement practiets in aged specuncris up to 40 years old have been hwated initigating the rate and extent of deptadation from a variety of sources, including Wyle stoc k and nu.

caused by aging, elcar plants.

NUlti!G-1377 54 m

l'EllSONAL AUTilOlt INDEX

'thisindes lists,in alphabetical order, all panicipating authors of each report listed in the main citation listing. liach name is followed by the numbet and the tttk of the repens prepared by the author. If further information is needed, refer to the main citation by the report number.

NUld!G/ Cit 499YAgingand Service Wear of hiul-Adams, M.L.

tistage Switches Used in Safety Systems of Nuc! car NUltEG/ Cit-4597," Aging and Serme Wear of Aux-Power Plants,, \\ ol.1.

iliary l'ecdwater Pumps for PWit Nuclear Power NU111!G/ Cit-5141, " Aging and Qualification lle-Plants, Vol.1: Operating iixperience and Failure Iden-search on Solenoid Operated Valves."

ttfication."

llader, ll.E.

Ahmed, S.

NUltliG/CP-0036, *Prorcedings of the Workshop on NUltilG/CIN4257,

• Inspection, Surveillance, and Nuclear Plant Aging" hionitoring of lilectrical liquipment inside Contain-ment of Nucleai Power Plants-With Applications to Ilcranck, A.F.

lileettical Cables."

NUltEG/CP-0100, *1'roceedings of the International Nuclear Power Plant Aging Symposium "

Allen, R.P.

NUllEG/ Cit-5491, *Shippingport Station Aging Ilcrry,D.1,.

!! valuation."

NUllEG/ Cit-3818, "lleport of flesults of Nuclear Anderson, it.L.

llescrly, D.D.

NUlti!G/ Cit-4928,

• Degradation of Nuclear Plant NUltEGICR-5560, " Aging of Nuclear Plant itesis.

Temperature Sensors."

tance Temperature Detectors."

Arendis,.l.G.

ahnW, D.F.

NUltl!G/ Cit-4977, "Sil AG Test Series: Seismic Ite.

l'NI -5722," Operating thperience and Aging Assess-search on an Aged United States Gate Valve and on a ment of !!CCS Pump iloom Coolers."

Piping System in the 1)ccommissioned licissdampf.

reaktor (III)lt): Summaty," Vol.1.

llorkowski, it.J.

NUltliG/ Cit-4977. "S1I AG Test Series: Seismic lle-NURl!G/ Cit-2641, the in. Plant lleliability Data search on an Aged United States Gate Valve and on a llase for Nuclear Power Plant Components: Data Col-Piping System in the Decommissioned lle,issdampf-lection and Methodology lleport."

eaktor ()IDit): Appendices, Vol. 2.

NUlt!!G/ Cit-3154, "The in. Plant lleliability Data llase for Nuclear Plant Components: Interim Atwood. L,.L.

Repon-./lhe Valve component."

EGO-SSHl!-8972. " Estimating liarard Functions for Itepairable Components."

llrown, D.P.

NUREG/ Cit-5386, allasis for Snubber Aging Re-1100 -SSill!-9017," User's Guide to Pil AZli, a Com.

search: Nuclear Plant Aging Research Program."

puter Program for Parametric If arard I unction listi.

mation."

Iturns, E.L.

NUld!G/CR4156, " Operating thperience and Ag-llacanskas, V.P.

ing. Seismic Assessment of litectric Motors."

NUREG/ Cit-4715. "An Aging Assessment of Relays and Ciretut itreakers and System Interactions."

Ilu @,S ll NUlll!G/ Cit-4279, " Aging and Senice Wear of fly-NUREGICR-4s19. " Aging and Serviec Wear of Sole-draulie and Mechanical Snubbers Used on Safety-Re-noid. Operated Valves Used in Safety Systems of Nu-lated Piping and Components of Nuclear Power clear Power Plants. Vol.1: Operating i aperience and Plants,* Vol.1.

I:ailute identification "

NURiiG-1377 55

,.. ~

Personal Authorindex Carfagno, A.

Clough, R.L.

NUREO/CR-4257, " Inspection, Surveillance, and SAND 88-0754 UC-78, " Time-Temperature. Dose hionitoring of Electrical Equipment inside Contain-Rate Superposition: A Methodology for Predicting Ca.

rnent of Nucicar Power Plants-With Applications to ble Degradation Under Ambient Nuclear Power Plant Electrical Cables."

Aging Conditions."

NUREO/CR-5141, " Aging and Qualification Re-search on Solenoid Operated Valves."

Collins, E.P.

NUREO/CR-5248, *Prioritization of '11RGALEX-Casada, D.A.

Recornmended Components for Further Aging Research."

NUREO/CR-5404, " Auxiliary Feedwater System Ag-ing Study " Vol.1.

Cornwell, B.C.

NUREO/CR-5706, "NRC llulletin 88-04: Potential NUREO/CR-3819, " Survey of Aged Power Plant Safety Related Pump less-An Assessment of Indus.

Facilities."

try Data."

Crowley, J.L.

Casada, M.L.

NUREO/CR-4380, " Evaluation of the Motor.

NUREO/CR-3543," Survey of Operating Experiences Operated Valve Analysis and Test System (MOVATS)

Detect Degradat,on, Incorrect Adjustments, and from LERs to Identify Aging Trends."

t i

Other Abnormalitics in Motor Operated Valves "

Cesarski, W.V.

Curreri, J.

NUREO/CR-4597," Aging and Service Wear of Aux-iliary Feedwater Pumps for PWR Nuclear Power BNL Technical Report A-3270-11-85, " Seismic En-Plants, Vol. 2: Aging Assessments and Monitoring durance Tests of Naturally Aged Small Electric Motors."

Method Evaluations."

Czajkowski, C.J.

Christensen, J.A.

NUREO/CR-4985, " Indian Point 2 Reactor Coolant NUREO/CP-0105. Paper by J. A. Christensen, Pump Seal Evaluations "

"NPAR Approach to Controlling Aging in Nuc! car Power Plants."

Datulano, B.

Cifuentes, F.

NUREO/CR-5479,

• Current Applications of Vibra-tion Monitoring and Neutron Noise Analysis: Detec-DNL Technical Report A-3270-11-85, "Scismic tion and Analysis of Structural Degradation of Reactor Endurance Tests of Naturally Aged Small Electric VesselInternals from Operational A ing."

C Motors."

Davis, T.

Clark, N.H, NUREO/CR-4144, *lroponance Ranking Based on NUREG/CR-3818 " Report of Results of Nuclear Aging Consideration of Components Included in Power Plant Aging Workshop.

Probabilistic Risk Assessments.

auss, D.B.

DeFour, ILA.

NUREO/CR-5334." Severe AccidentTesting of Elec-WYLE 60103-X," Test Plan for the Comprehensive trical Penetration Assemblics."

Aging Assessment of Circuit Dreakers and Relays for Nuclear Plant Aging Research (NPAR) Program, Phase 11."

Cletcher, J.W.

NUREO/CR-4692, " Operating Experience Review of DeWall, K.G.

Failures of Power Operated Relief Valves and Block NUREO/CR-3819, " Survey of Aged Power Plant Valves in Nuclear Power Plams."

Facilities."

Clinton, J.

NUREG/CR-5406, "BWR Reactor Water Cleanup Syste Flexible Wedge Gate Isolation Valve Qualifi-NUREO/CR-4985," Indian Point 2 Reactor Coolant cation and High Energy Flow Int e rruption Test; Vol.1, Pump Seal Evaluations."

ulysis and Conclusion."

NUREO-1377 56

)

Personal Author Index NUREG/CR-4302, " Aging and Senice Wear of NURl!G/CR-5406, "HWR Reactor Water Cleanup Check Valves Used in lingineered Safety !:cature Sys-System l'lexible Wedge Gate isolation Valve Quahfi-tems of Nuclear Power Plants," Vol.1.

cation and liigh Energy I' low I nterruption Te st: Vol. 2, Data Repon."

N11RI!GICR-4380, "livaluation of the hiotor-OperatedValve AnalysisandTest System (htGVATS)

NUKEG/CR-5406, "IlWR Reactor Water Cleanup to Detect Degradation, incorrect Adjustments, and System llexible Wedge Gate isolation Valve Oualifi.

Other Abnormalities in biotor-Operated Valves "

cation and liigh Energy 110w Interre ption Test; Vol. 3, Review of issues Associated with IlWR Containment Fain, ll.E.

isolation Valve Closure "

NUREG/CR-5383 "liffect of Aging on Response NURl!GlCR-5558.

• Generic issue 87: Flexible Time of Nuclear Plant Pressure Sensors."

Wedge Gate Valve Test Program: Phase !! Results and Analysis."

Fragola, J.R.

NUREGICR-3iS4, 9hc In Plant Rehability Data 11ase for Nuclear Plant Comp (ments: Interim Re.

Dingee, D.A.

port-The Valve Comp (ment."

NUREGICR-4590," Aging of Nuclear Station Diesel Generators:livaluation of Operating and Expert Ex-Fresco, A.

perience," Vols. I and 2.

IlNI. Technical Repon A-327011-2-90, " Aging lif.

f ects of important ilalance of Plant Systems in Nuclear Dinsel, M.R.

Power Plants."

NUREG/CR-3956, *In Situ Testing of the Ship-pingport Atomic Power Station lilectrical Circuits."

Ftichrer, G.F.

NUREGICR-5655, " Submergence and liigh Tem.

perature Sicam Testing of Class 111 Electrical Cables."

Dodge, R.E.

NUREG/CR-4279," Aging and Senice Wear of fly-Fullel, R.

draulic and hiechanical Snubbers Used on Safety.

llNI, Technical Report A-3270-12-86, " Aging and Related Piping and Components of Nuclear Power 1.ife listension Assessment Program (AI.!!AP) Sys-Plants," Vol.1.

tems i evel Plan."

NUREG/CR-5052," Operating Experience and Aging Assess ent of Component Cooling Water Systems in Donaldson, M.R.

NUREG/CR-3956, "In Situ Testing of the Ship.

Pressunzed Water Reactors."

pingport Atomic Power Station Electncal Circuits.

NUREG/CR-5268, " Aging Study of Boiling Water j

Reactor Residual lleat Removal System."

i i

Drago, J.P.

NUREG/CR-5419," Aging Assessment ofInstrument NUREG/CR-2641, "the in Plant Reliability Data Air Systems in Nuclear Power Plants."

liase for Nuclear Power Plant Components: Data Col.

lection and hiethodology Report."

Gallaher, R.ll.

NURIIG/CR-3543, ' Survey of Operating Experiences Edson,Jl..

from 1 liRs to identify Aging Trends.

NUREGICR-4457, " Aging of Class III Itatteries in NURiiG/CR-4302, " Aging and Senice Wear of Safety Systerns of Nuclear Power Plants."

Check Valves Used in Engineered Safety Feature Sys-tems of Nuclear Power Plants," Vol.1.

NUREGICR-5181, " Nuclear Plant Aging Research;

'the lli Power System."

Gillen, K.T.

NURiiGICR-5448, " Aging I! valuation of Class 111 SAND 88-0754 UC-78, " Time Temperature-Dose llaneries: Setsmie I,esting.

Rate Superposition: A hiethodology for Predicting Ca-l ble Degradation Under Ambient Nuclear Power Plant Eissenberg, D.M.

Aging Conditions "

NUREG/CR-4234," Aging nnd Service Wear of Elec-G.ingrich, J.J.

tric hiotor Operated Valves Used in Engineered NUREG/CR-5383, "Effect of Aping on Response Safety-Feature Systems of Nuclear Power Plants."

Time of Nuclear Plant Pressure Sensors."

Vol.1.

NURl!G-1377 57

Personal Author Index Gleason, J.F.

NU Rl:G/CH-5051, *Detectmg and hiitigating liattery WYlli bO103-X, " Test Plan for the Comprehensive

' E " "" I"# C' "E

Agmg Assessment of Ctreuit Urcakers and Relays for NURl!GICR-5192," Testing of a Naturally Aged Nu-Nuclear Plant Ap.ing Research (NPAR) Program, clear Power Plant Inverter and llattery Charger."

NUltEG/CR-5268, "Armg Study of Iloiling Water Goldberg, F F.

Itcactor Residual lleat Removal System."

NURiiG/CR-2641. "'lhe In Plant Reliability Data NUREG/CR-S507,"Results from the Nuclear Plant

^E "E Nesearch Program: their Use in Inspection Ae-liase for Nuclear Power Plant Components: Data Col.

I lection and hiethoJoiogy Report."

tivities.,

NUREG/CR-5555, " Aging Assessment of the West-Goodmr.n, ILL.

inghouse PWR Control Rod Drive System."

PNI-5722,

• Operating Experience and Aging Assess-ment of ECCS Pump Room Coolers "

llammond, J M.

WYll! 60103-X, fest Plan for the Comprehensive gore, y,L*

Aging Assessment of Circuit liteakers and Relays for Nuclear Plant Aging Research (NPAR) Program, NURiiG/CR-5379," Nuclear Plant Service Water Sys' Phase 11."

tem Aging Degradation Assessment: Phase I," Vol.1.

Greenstreet, W.L.

llanclicy, L.A.

NUREG/CR--4234," Aging an<l Senice Wear of Elec-NUREG/CP-0036,

• Proceedings of the Wortshop on Nucler.t Plant Aging."

tric hiotor Operated Valves Used in Engineered Safety Feature Systems of Nuclear Power Plants,"

Vol.1.

liardin, J.E.

NUREG/CR--4457, " Aging of Class IE ilatteries in NUREGICR-4302, " Aging and Senice Wear of Safety Systems of Nuclear Power Plants."

Check Valves Used in Engineered Saiety-Feature Sys-tems of Nuclear Power Plants," Vol.1.

liasitemian, ll.M.

Grove, E.

NUREG/CR-4928, "Degradat;on of Nucleat Plant Temperature Sensors."

llNL Technical Report 'IR-3270-9-90, "An Opem tional Assessment of the liabcock & Wilcox and tom.' NUREG/CR-5383, "Effect of Ag.ing on Response bustion Engineering Control Rod Drives.

Time of Nuclear Plant Pressure Sensors "

NURiiG/CR-5560, " Aging of Nuclear Plant Resis-Guntlier, W.E.

tance Temperature Detectors."

llNLTechnical Report TR-3270-6-90, "htaintenance Team inspection Results: Insights Related to Plant llaynes, ll.D.

Aging."

NUREG/CR-4234, " Aging and Service Wear of Elec-DNL Technical Report TR-3270-9-90, "An Opera-tric h10 tor. Operated Valves Used in Engineered tional Assessrnent of the Babcock & Wilcox and Com.

safety. Feature Systems of Nuclear Power Plants: A -

E bustion IIngineering Control Rod Drives."

ing Assessments and hionitoring hiethod Enlu-ations," Vol. 2.

NUREG/CR-4564, " Operating Experience and A -

E NUREG/CR4302, " Aging and Service Wear of ing-Seismic Assessment of Hattery Chargers and in-Check Valves Used in Engineered Safety-Feature verters."

Systems of Nuclear Power Plants; Vol. 2, Aging Assess-NUREG/CR-4939, " improving hiotor Reliability in Nuclear Power Plants;" Volume 1: Performance Evaluation and hiaintenance Practices; Volume 2:

lleasler, P.G.

Functional Indicator Tests on a Small Electric hiotor NUREG/CR-4279, " Aging and Service Wear of liy-Subjected to Accelerated Aging; Volume 3: Failure draulic and hicchanical Snubbers Used i n Safety-Analysis and Diagnostic Tests on a Naturally Aged Related Piping and Comjunents of dumar Power Electric hiotor.

Plants," Vol.1.

Personal Author Indes liebble, T.L.

jneobus, M.J.

NURl!G/CR-3154, "Itc In Plant Reliability Data NUlWG/CR-5461, " Aging of Cables, Connections, liase for Nuclear Plant Components: Interim Re.

and Electrical Penetration Assemblies Used in Nu.

~

clear Power Plants."

port-lhe Valve Comp nent.a NUlWG/CR-5655, " Submergence and liigh Tem.

1Iiggins, J.C.

petature StcamTestingof Class 111IilectricalCables."

llNL Technical Report A-3270-Il-85, "Scismic lin.

Jarrell, D.II.

durance Te',ts of Naturally Aged Smail lilectric h19 NUlWG/CR-5248, "Prioritization of TIRGAtliX-torsy Recommended Compments for Further Aging IINL Technical Report A-3270-12-86, " Aging nnd llescarch."

Life lixtension Assessment Program (All!AP) Sys-NURIIG/CR-5379,* Nuclear Plant Scryice Water Sys-tems Level I lan' tem Aging Degradation Assessment: Phase 1,"Vol.1.

NUlWG!CR-5052,"Operatinglixperience and Agmg PNirSA-18407, " Understanding and hianaging Cor-Assessment of Component Cooling Water Systems in rosion in Nuclear Dawer Plants."

Prer,surized Water Reactors?

jnske, C.E.

Ilolbert, K.E.

NUlWG/CR-5314,"1.ife Assessment Procedures for NURl!G/CR-4928 " Degradation of Nuclear Plant hiajor LWR Components; Vol.3, Cast Stainless Steel Temperature Sensors.a Components."

Johnson, A.ll.

Iloopingarner, K.R.

NURl!G/CR-5379," Nuclear Plant Service Water Sys-

" Safety Implications of Diesel Generator Aging," Nu.

tem Aging Degradation Assessment: Phase 1," Vol.1.

clear Safcty, December 1990.

NURIIG/CR-5491, "Shippingport Station Aging NURl!G/CR-4590," Aging of Nuclear Station Diesel livaluation."

Generators: Evaluation o1 Operating and Expert Ex-PNL-SA-18407, " Understanding and hianaging Cor-perience, Vols. I and 2.

rosion in Nuclear Power Plants."

NUREGICR-5057, " Aging hiitigation and improved 30hDSOH,3 D,.

Programs for Nuclear Service Diesel Generators?

NURl!G/CR-3154, "the in. Plant Reliability Data PNic6287," Study Group Review of Nuclear Senice liase for Nuclear Plant Components: Interim T

Diesel Generator Testing and Agin;; hiitigation."

Report-The Valve Component."

PN1-7516, " Emergency Diesel Generator Technical NHN O' Specifications Study Results."

NUlWG/CR-3154, "the in Plant Reliability Data llase for Nuclear Plant Components: Interim Re-llorst, C.L.

port-The Valve Component."

NURiiG/CR-5159,

• Prediction of Check Valve Per-formance and Degradation in Nuclear Power Plant Kalsi, M.S.

Systems."

NUlWG/CR-5159," Prediction of Check Valve Per-formance and Degrada, ion in Nuclear Power Plant NURIIG/CR-5583, " Prediction of Check Vahe Per-S **S-f forinance and Degradation in Nuclear Power Plant Systems-Wear and Impact Tests "

NUREG/CR-5583, " Prediction of Check Valve Per-formance and Degradation in Nuclear Power Plant Systems-Wear and Impact Tests."

lloy,ll.C.

N URliG/CR-3543,

• Survey of Operating Experiences Kerlin, T.%'.

from 1 l!Rs to identify Aging Trends."

NUREG/CR4928, " Degradation of Nuclear Plant Temperature Sensors."

NUlWG/CR-5612," Degradation hiodeling with Ap.

Kirkwood, ll.J.

plications to Aging and hiaintenance Effectiveness PNL-6287," Study Group Review of Nuclear Senite Diesel Generator Testing and Aging hiitigation."

livaluation."

59 NU Rl!G-1377

.....J

Perscnal Author Index Kitch, D,hl.

Subjected to Accclerated Aging: Volume 3: Failure NURliG/ Cit-4597, " Aging and Servke Wear of Aux-Analysis and Diagnostic Tests on a Naturally Aged iliary Feedwater Pumps for PWR Nuclear Power Ucetric hiotor.

Plants, Vol. 2: Aging Aucssments and hionitonng NUREG/CR-5052," Operating Experbnce and Armg hiethod Evaluations."

Assessment of Component Cooling Water Systems in Pressurhed Water Reactors."

Kochis, S.

NUREG/CR-5268, " Aging Study of Iloihng Water NUREG/CR-5280, " Age-Related Degradation of Reactor Residual llent Removal System."

Westinghouse 480-Volt Circuit !!reakers; Vol 2, hie-chanical Cyding of a DS-416 litcaker. Test Results."

Lonrecky, l'.J.

PNI A287," Study Group Review of Nnclear Senice Kryter, R.C.

Diesel Generator Testing and Aging hiitigation."

NUREG/CH-5479, " Current Applications of Vibra.

Lubeski, P.A.

tion hionitoring and Neutron Noise Analysis: Detec.

tion and Analysis of Structural Degradation of Reactor M LE 60103-X," Test Plan for Comprehensive Aging VesselInternals from Operational Armg.

Assessment of Circun litcakers and R elays for Nuclear Plant Aging Research (NPAR) Prograu, Phase !!."

Kurth, R.

N1acDonald, P.E.

NURilG/CR-4144, "Lnportance Ranking liased on NURl!G/CR-4731 "Residualljic Assessment of hia-Aging Consideration of Components included in jar 1.ight Water Reactor Components," Voi.1.

Probabilistic Risk Assessments."

NUREG/CR-4731,"Residualljic Assessment of hia-NUREG/CR-5510, " Evaluations of Core hielt Fre-jor I.ight Water Reactor Comptments-Oveniew "

quency liffects Due to Component Aging and hiainte-Vol. 2.

51acDougall, E.

Lee, [3,g, NUREG/CR-5280, " Age-Related Degradation of NUREG/CR-5280, " Age-Related Degradation of Westinghouse 480-Vo!t Circuit lireakers; Vol.1, Ag.

Recommendadons for Improdng Westinghouse 480-Volt Circuit lircakers; Vol 2, hic-

"E Sy*g{"

chanictd Cycling of a DS-416 Ilreaker. Test Results."

NURl!G/CR-5280, " Age-Related Degradation of Leverenz, F.

Westinghouse 4SO-Volt Circuit fireakers; Vol 2 hie-NUREG/CR-4144, "Importance Ranking llased on Aging Consideration of Components included in h1akny, E.

Probabilistic Risk Assessments."

NUREG/CR-4597," Aging and Service Wear of Aux-Levy,1.5,.

iliary Feedwater Pumps for PWR Nuclear Power Plants, Vol.1: Operating Experience and Failure Iden-NUREG/CR-5248, "Prioritiration of TIRGAll!X-tification."

Recommended Components for Funber Aging Re-search."

h1c3Ic,II.hl.

NUREG/CR-4747,"An Aging Failure SurveyofIJght Lewis, R.

Water Reactor Safety Systems and componerts,"

NUR EG/CH-5051, " Detecting an l hiitigating liattery Vol.1.

Charger and Inverter Aging."

NUREG/CR-4747,"An Aging Failure Survey of Light Water Reactor Safety Systems and Components,"

Lodlow, C.C.

Vol. 2.

NUREG/CR-4715 "An Aging Assessment of Relays hleiningcr, R.D.

and Circuit Ilreakers and System Interactions "

NUREG/CR-5008, " Development of a Testing and aro,R.

Analysis hiethodology to Determine the Functional Condition of Solenoid Operated Valves."

NUREG/CR-4939, "Improvmg hic. tor Reliability in Nuclear Power Plants;" Volume 1: Performance hieyer, L.C.

Evaluation and hiaintenance Practices: Volume 2:

NUREG/CR-4740, " Nuclear Plant-Aging Research Functional Indicator Tetts on a Small lilectric hiotor on Reactor Protection Systems "

NUREG-1377 60

Personal Author index NUREG/CR-4967, " Nuclear Plant Aging Research Nowlen, S.P.

on liigh Pressure Injection Systems."

NUREG/CR-5546,*AnInvestigationof theEffectsof Thermal Atmg on the Fire Damagea.bility of Electric NUREG/CR-5181, "Nucicar Plant Aging Research:

rables "

The IE Power System."

NUREG/CR-5619,"Ihe Ir.pr 4hermal Agingen Miller, B.

the Flammability of Electric Cams "

IINL Technical R : port A-3270-11-26-84, " Scoping I3almer, G.R.

Test on Containment Purge and Vent Valve Seal NUREG/CR-53S6, allasis for Snubber Aging Re.

Material."

search: Nuclear Plant Aging Research Program."

Mitchell, D.W.

Petersen, K.M.

NUREG/CR-5560, " Aging of Nuclear Plant Resis NUREG/CR-4928, " Degradation of Nuclear Plant tance Temperature Detectors.,

Temperpture Sensors."

NUREG/CR-5383, "Effect of Aging on Response Morr.s, H.M.

Time of Nuclear Plant Pressure Sensors."

i NUREG-il44, " Nuclear Plant Aging Research (NPAR) Program Plan."

Pike, D.II.

NUREG/CR-2641, "The In-Plant Reliability Data Moyers, J.C.

Base for Nuclear Power Plant Components: Data Col-NUREG/CR-5519, Vol.1, " Aging of Control and lection and Methodology Report."

Service Air Compressors and Dryers Used in Nuclear Reich, M.

Power Plants."

BNL Technical Report A-3270-ll-PS, " Seismic En-durance Tests of Naturally Aged Srnall Electric Mo-Murgty, G.A.

tors.

NUREG /CR-3543, " Survey of Operating Experiences from LERs to Identify /...g Trends "

Rib, L.N.

NUREG/CR-423: " Aging a. senice Wear of Elec.

Letter Report, L N. Rib, " Summaries of Research tric Motor-Operated Valves Used in Engince td Reports Submitted in Connect on with the Nuclear i

Safety-Feature Systems of Nuclear Power Plants,"

Plant Agin;;Research(hPAR) Program," Engineering and Economics Research, Inc. (EER).

Vol.1.

NUREG/CR-4302, " Aging and Senice Wear of Roberts, G.C.

Check Valves Used in Engir.2ered Safety-Feature Sys-NUREG/CR-4819," Aging.md Senice Wear of Sole.

tems of Nuclear Power Plants," Vol.1.

noid Operated Valves UseJ in Safety Systems of Nu.

NUREG/CR 4692," Operating Experience Reviewof clear Power Plants, Vo'.1: Operating Experience and Failure Identification.

1. ailures of Power Operated Relief Valves and 111ock Valves in Nuclear Power Plants "

N'UREG/CR-4992, " Armg and Service Wear of Mul-tistage Switches Used in Safety Systems of Nuclear Naus, D.J.

Power Plants Vol. E NUREG/CR-4652,"ConcreteComponent Agingand Rose, J.A.

hs Significance Relative to Life Enension of Nuclear NUREG/CR-3819,"St..cy of Aged Power Plant Fa-Power Plants "

ciliti-s" Nehring, L Samanta, P.K.

BNL Techr.ical Repon A-3270-Il-85, "Scismic En' NURiiG/CR-5612," Degradation Modeling with Ap-durance fests of Naturally Aged Small Electne Mo-plaions to Aging and Maintenance Effectiveness tors.

livaluation.'

j i

Nesbitt,1F.

Satteru hite, D.G.

i NURl!GICR-1540 " Aging o Nuclet.r Statton Dies. !

NUREG/CR-4747." An Aging Failure Survey of Light r

Generaters: Evn:uation of Operating and Expert E -

Water Reactor Safety Ssstems and Comperents" Vol. 1.

perience " Vols I and 2.

I l

NUREG-1377 6i

6 Personal Author inacx NUREG/CR-4747."An Aging Failure Survey of Light NUREG/CR-5268, " Aging Study of Hoiling Water Water Reactor Safety Systems and Components" Reactor Residual lleat Removal System."

I' '

NUREG/CR-5280, " Age-P.clated Degradation of Westinghouse 480-Volt Circuit Dreakers; Vol.1. Ag-Scalzo, S.M.

ing Assessment and Recommendations for improving NUREG/CR-5510, " Evaluations of Core Melt Fre-Breaker Reliability "

^

quency Effects Due to Component Aging and Mainte-nanee "

Shook, T.A.

NUREG/CR-4715. "An Agmg Assessment of Relays y

Schlonski. J.S.

and circuit Breakers and System interections."

NUREG/CR-4992, "Agw 9'

• 7 evice WW. a Mul-(

tistage Switches Used in isafety Systems of Nuclear Silver, M.M.

Power Plants," Vol.1.

DNLTechnical Report A-3270-12-85," Pilot Assess-ment: Impact of Aging on the Seismic Performance of Sharaghi, A.

Selected Equipment Types."

NUREG/CR-4144, "Importance Ranking Based on Sinha, U.P.

Aging Consideration of Components included in PNL-SA-18407, "Understandmg and Managing Cor-Probabilistic Risk. Assessments."

rosion in Nuclear Power Plants."

Shah, V.N.

Skreiner, K.M.

NUREG/CR-4731," Residual Life Assessment of Ma-NUREG/CR-4939, " Improving Motor Reliability in jor Light Water Reactor Components," Vol.1.

Nuchar Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Volume 2:

NUREG/CR-4731,"Residuallife Assessment of Ma Functional Indicator re.ts on a Small Electric Motor jor Light Water Reactor Components-Ovemew,,,

Subjected to Accelerated Aginy Volume 3: Failure Vol. 2.

Analysis and Diagnostic Tests on a Naturally Aged NUREG/CR-5314," Life Assessment Procedures for Electric Motor.

Major LWR Components; Vol.3, Cast Stainless Steel Components."

Eoberano, F.T.

u e ng ftpe y PNL.-SA-It.407, " Understanding and Managing Cor-m rosion in Nuclear Pcwcr Plants."

p ngp rt Atonuc Power Stanon Ecct&al Ctremts.

Sowatskey, P.J.

Sharma, V.

NUREG/CR-4597," Aging and Service Wear of Aux-NUREG/CR-5583, " Prediction of Check Valve Per-iliary Feedwater Pumps for PWR Nuc! car Power formance and Degradation in Nuclear Power Plant Plants, Vol 2: Aging Asserments and Monitorirg Systems-Wear and Impact Tests."

Method Evaluations "

Sheets, M.W.

Steele, R.

BNL Technical Report A-3270-3-86, " Testing Pro.

NUREG/CR-3819," Survey of Aged Power Plant Fa-gram for the Munitoring of Degradation in a Continu.

cilities.

[

ous Duty 460 Volt Class "B",10-HP bicctric Motor."

NUREG/CR-4977, " SHAG Test Series: Seismic Re-

}

NUREG/CR-4939, " Improving Motor Reliabditv in search on an Aged United StLtes Gate Valve and on a Nuclear Power Plants;" Volume 1: Performance Evai-Piping System in the Decommissioned Heissdampf-untion and Maintenance Practices; Volume 2: Func-freaktor (HDR): Summary," Vol.1.

tional Indicator Tests on a Small Electric Motor NUREG/CR-4977, " SHAG Test Series: Seismic Re-I Subjected to Accelerated Agir.g: Volume 3: Failure sestch on an Aged United States Gate Valve and on a Analysis and Diagnostic Tests on a Naturally Aged Piping System in the Decommissioned Heissdampf-Electric Motor.

Creaktor (rIDR): Appendices," Vol 2.

Sh,er, W,.

NUREG/CR-5406, "BWR Reactor Water Cleanup t

System Bexible Wedge Gate Isolation Valve Qua'L#

l NUREG/CR-5053," Operating Experience and Aging cation and High Energy Flow Interruption T st; Vol.1, Asseosment of Motor Control Centers."

Analysis and Conclusion."

NUREG-1377 62 l

i

Personal Author Index NUREG/CR-5052," Operating Experience and Aging NUREG/CR-5406, "BWR Reactor Water Cleanup Assessment of Component Cooling Water Systems ;n System Flexible Wedge Gate Isolation Valve Qualifi-Pressurized Water Reactors."

cation and High Energy Flow Interruption Test; Vol. 2, Data Repen."

NUREG/CR-5053," Operating Experience and Aging Assessment of Motor Control Centers."

NUREG/CR-5406, "BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation Valve Qualifi-NUREG/CR-5768, " Aging Study of Boiling Water cation and liigh Energy Flow InterruptionTest: Vol. 3, Reactor Residual lleat Removal System."

Review of Issues Associated with BWR Containment NUREG/CR-5280, " Age-Related Degradation of Isolation Vahe Closure Westinghouse 480-Volt Circuit Breakers; Vol.1, Ag.

a NUREG/CR-5558, " Generic Issue 87: Flexible ing Assessment and Recommendations for improving Wedge Gate Valve Test Program: Phase 11 Results and Breaker Reliability "

NUREG/CR-5280, " Age-Related Degradation of

^""

Westinghouse 480-Volt Circuit Breakers; Vol 2, Me-chanical Cycling of a DS-416 Breaker. Test Results."

Subudhi, M.

- LetterReport,M.Subudhi,"Reviewof Aging-Scismic NUREG/CR-5419," Aging Assessment of Instrument Correlation Studies on Nuclear Plant Equipment,"

Air Systems in Nuclear Power Plants."

Brookhaven National Laboratory,~ January 1985.

NUREG/CR-5612," Degradation Modeling with Ap-UNL Technical Report A-3270-11-85, "Scismic En-plicatior.s to Aging and Maintenance Effectiveness durance Tests of Naturally Aged Small Electric Mo-Evaluation."

tors."

"E"*""'

BNLTechnical Report A-3270-12-85," Pilot Assess-UNL Technical Report A-3270-3-86, " resting Pro-ment: Impact of Aging on the Seismic Performance of gram for the Monitoring of Degradation in a Continu-Selected Equipment Types."

ous Duty 460 Volt Class "B",10-HP Electric Motor "

BNL Technical Report A-3270-3-86, " Testing Pro-NUREG/CR-4939, " improving Motor Reliability in gram for the Monitoring of Degradation in a Continu-Nude.ar ower Mants;" Volume 1: Performance ous Duty 460 Volt Class "B",10-HP Electric Motor."

Evaluation and Mainten, ace Practices; Volume 2:

Functional Indicator Tests on a Small Electric Motor BNL Technical Report A-3270-12-86, " Aging and Subjected to Accelerated Aging; Volume 3: Failure Life Extension Assessment Program (ALEAP) F-s.

Analysis and Diagnostic Tests on a Naturally Aged tems level Plan."

Electric Motor.

BNL Technical Report A-3270R-2-90, " Aging Ef-fccts of important Balance of Plant Systemsin Nuclear Sullivan, K.

NUREG/CR-5555, " Aging Assessmen af the West-Power Phnts "

inghouse PWR Control Rod Drive System."

NUREG/CR-4156, " Operating Experience and Aging-Scismic Assessment of Electric Motors "

NUREG/CR-4564, " Operating Experience and BNL Technical Report A-3270-11-85, " Seismic En-Aging-Scismic Assessment of Battery Chargers and durance Tests of Naturally Aged Small Electric Mo-tors."

Inverters."

BNL Technical Report A-3270-12-86, " Aging and NUREG/CR-4939, " Improving Motor Reliability in Life lixtension Assessment Program (ALEAP) Sys-Nuclear Power Plants;" Volume 1: Performance tems Level Plan."

Evaluation and Maintenance Practices; Volume 2:

Functional Indicator Tests on a Small Electric Motor NUREG/CR-4156, " Operating Experience and Subjected to Accelerated Aging: Volume 3: Failure Aging. Seismic Assessment of Electric Motors."

Analysis and Diagnostic rests on a Naturally Aged NUREG/CR-4564, " Operating Experience and Electric Motor.

Aging-Scismic Assessment of Battery Chargers and NUREGICR-4985," Indian Point 2 Reactor Coolant Inverters "

Pump Seal Evaluations?

NUREG/CR-4939, " Improving Motor Reliability in NU REG /CR-5051, " Detecting and Mitigating Hattery Nuclear Power Plants;" Volume 1: Performance i

Evaluation and Maintenance Practices; Volume 2:

Charger and Inverter Agmg?

NUREG-1377 63

m I

Pen.onal Author Index Functional Indicator Tests on a Small Electric hiotor Villaran, M.

Subjected to Accelerated Aging: Volume 3: Failure Analysis and Diagnostic Tests on a Naturally Aged NUREG/CR-5419 " Aging Assessment ofInstrument Electric hiotor.

Air Systems in Nuclear Power Plants."

NUREGICR-4985," Indian Paint 2 Reactor Coolant Yora, J.P.

Pump Seal Evaluations."

NUREG-1144,

• Nuclear Plant Aging Research NUREGICR-5052," Operating Experience and Aging (NPAR) Program Plen."

Assessment of Component Cooling Water Systems in Pressurized Water Reactors.*

NUREG-ll44, " Nuclear Plant Aging Research (NPAR) Prograrn Plan," Rev.1.

NUREG/CR-5268, " Aging Study of Boiling Water Reactor Residual Heat Removal System."

\\ Yang, J.K.

NUREG/CR-5507,"Results from the Nuclear Plant Aging Research Program: nctr Use in hispection Ac-NUREG/CR-5159, " Prediction of Check Valve Per.

tivities."

formance. and Degradation in Nuclear Power Plant Systems."

Toman, G.J.

NUREG/CR-5583, " Prediction of Check Valve Per.

NUREG/CR-4257, " Inspection, Surveillance, and fotmance and Degradation in Nuclear Power Plant Monitoring of Electrical Equipment inside Contain.

Systems-Wear and Impact Tests."

ment of Nuclear Power Plants-With Applications to Electrical Cables," Vol.1.

Watkins, J.C.

NUREGICR-4257,

• Inspection, Surveillance, and NUREG/CR-5558, " Generic Issue 87: Flexible Monitoring of Electrical Equipment in Nuclear Power Wedge Gate Valve Test Program: Phase II Results and Plants, Vol. 2: Pressure Transmitters."

Analysis."

NUREGICR-4715,*An Aging Assessment of Relays and Circuit Breakers and System Interactienr."

Weeks, J.

NUREG/CR-4819, " Aging and Service Waar of Sole-NUREGICR-4985," Indian Point 2 Reactor Coolant noid Operated Valves Used in Safety Syste.ms of Nu-Pump Seal Esaluadons.

clear Power Plants, Vol.1: Operating Experience and Failure Identification."

We.ir, T.J.

NUREGICR-5141, " Aging and Qualification Re.

NUREGICR-5008, " Development of a Testing and search on Solenoid Operated Valves."

Analysts Methodology to Determme the Functional Condition of Solenoid Operated Valves "

Vasudevan, R.

BNLTechnical Report A-3270-12-85," Pilot Assess.

Werry, E.V.

ment: Impact of Aging on the Seismic Performance of NUREG/CR-5386, " Basis for Snubber Aging Re-Selected Equipment bpes."

search: Nuclear Plant Aging Research P-ogram."

Vause,J.W.

Wilhelm, W.

NUREG/CR-4590, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating and Expert Ex-NUREG/CR-5280, " Age-Related Degradation of perience," Vols. I and 2.

Westinghouse 480-Volt Circuit Breakers; Vol 2. Mc-chanical Cycling of a DS-416 Breaker. Test Results."

Vesely, W.E.

NUREG/CR-4769, " Risk Evaluations of Aging Phe-Zaloudek, F.R.

nomena: The Linear Aging Reliability Model and Its

" Safety Implications of Diesel Generator Aging,"Nu-Extensions."

clear Safery, December 1990.

NOREGICR-5510. " Evaluations of Core Melt Fre-NUREG/CR-5057, " Aging Mitigation and improved quency Effects Due to Component Aging and Mainte-Programs for Nuclear Service Diesel Generators."

nance "

j NUREG!CR-5612, " Degradation Modeling with Ap.

Zimmerman, P.W.

plications to Aging and Maintenance Effectiveness NUR EG/CR-5379, " Nuclear Plant Service Water Sys-Evaluation."

tem Aging Degradation Assessment: Phase I," Vol.1.

]

NUREG-1377 t

64

i CORPORATE AUTHOR INDEX This index lists, in alphabetical order, the organizations that prepared the reports listed in this compila-tion. Usted below each organization are the numbers and titles of its reports. if further information is needed, refer to the main citation by the report number.

NUREG/CR-4715,"An Aging Assessment of Relays Analysis and Measurement Services Corp.

and Circuit 11reakers and System Interactions."

NUREG/CR-4928, " Degradation of Nuclear Plant NUREG/CR-4939. " Improving hiotor Reliability in Temperature Sensors "

Nuclear Power Plants;" Volume 1: Performance Eval-NUREG/CR-5383, "Effect of Aging on Response uation and hiaintenance Practices; Volume 2: Func-Time of Nuclear Plant Pressure Sensors."

tional Indicator Tests on a Small Electric Motor Subjected to Accelerated Aging: Volume 3: i'ailure NUREG/CR-5560, " Aging of Nuclear Plant Resis-Analysis and Diagnostic Tests on a Naturally Aged tance Temperature Detectors.

Electric Motor.

NUREG/CR-4985," Indian Point 2 Reactor Coolant Brookhaven National Laboratory (IINL)

Pump Seal Evaluations.,

Letter Report, M. Subudhi," Review of Aging Scismic NUREG/CR-5051," Detecting and Mitigating Hattery Correlation Studies on Nuclear Plant Equipment,"

Charger and Inverter Aging."

Erookhaven National Laboratory, January 1985.

NUREG/CR-5052,"OpciatingExperience and Aging HNL Technical Report A-3270-ll-26-84, " Scoping Assessment q Component g,dng Water Systems in Test on Containment Purge and Vent Valve Seal I ressurized Mter Reactors.

g g g,,

NUREG/CR -5M3,* Operating Experience and Aging HNL Technical Report A-3270-11-85, "Scismic En-Assessment of Motor Control Centers."

durance Tests of Naturally Aged Small Electric NUREG/CH-5192," resting of a Naturally Aged Nu-Motors."

clear Power Plant inserter and Battery Charger."

UNL Technical Report A-3270-12-85. " Pilot Assess-NUREG/CR-5268, " Aging Study of Boiling Water ment: Impact of Aging on the Seismic I erformance of Reactor Residual Heat Removal System."

Selected Equipment Iypes.,

NUREG/CR-520, " Age-Related Degradation of HNL Technical Report A-3270-3-86, " Testing Pro-Westinghouse 480-Volt Circuit Breakers; Vol.1, Ag-gram for the Monitoring of Degradation in a Continu-ing Assessment and Recommendations for Improving ous Duty 460 Volt Class "11",10-HP Electric Motor."

Drtaker Reliability."

HNL Technical Report A-3270-12-86, " Aging and

"' EG/CR-5280, " Age-Related Degradation of Life Extension Assessment Program (ALEAP) Sys-Westinghouse 480-Volt Circuit Hrcakers; Vol 2, Mc-chanical Cycling of a DS-416 Breaker. Test Results "

tems 1 evel Plan."

NUREG/CR-5419," Aging AssessmentofInstrument UNL Technical Report A-3270R-2-90, " Aging Ef-Air Systems in Nuclear Power Plants."

fects of Important Halance of Plant Systems in Nuclear Power Piants "

NUREG/CR-5507, "Results from tbc Nuclear Plant Aging Research Program:Their Use in Inspection Ac-BNLTechr.ical Report TR-3270-6 00," Maintenance tivities.

Team Inspection Results: Insights ilelated to Plant NUREG/CR-5555, " Aging Assessment of the West-Aging "

IINL Technical Report TR-3270-9-90, " An Opera-NUREG/CR-5612. " Degradation Modeling with Ap-tional Assessment of the Habcock & Wilcox r nd Com-bustion Engineenng Control Rod Drives."

plications to Aging and Maintenance Effectiveness Evaluation."

NUREG/CR-4156, " Operating Experience and Aging-Scismic Assessment of Electric Motors "

Engineering and Economics Research Inc, (EER)

N UREG 'CR-4564. " Operating Experience and Anmg-Seismic Assessment of Battery Ch srgers and Letter Report, L N. Rib, " Summaries of Research 1

Reports Submitted in Connection with the Nuclear Inverters.'

NUREG-1377 65

Corporate AuthorIndex Engineering and Economics Research, Inc.

Piping System in the Decommissioned lirissdampf-(EER) (Cont.)

reaktor (HDR): Appendices," Vol. 2.

Plant Aging Research (NPAR) Program," Engineering NUREG/CR-5181, " Nuclear Plant Aging Research:

and Economics Research, Inc. (EER).

'Ihe IE Power System."

Franklin Research Center NUREG/CR-5248, "Prioritization of TIRGALEX-Recommended Components for Further Aging Re-NUREG/CR-4715,"An Aging Assessment of Relays search."

and Circuit Breakers and System Interactions."

NUREG/CR-5314. " Life Assessment Procedures for NUREG/CR-5141, " Aging and Qualification Re.

Major LWR Components: Vol.3, Cast Stainless Stect search on Solenoid Operated Valves."

Components."

Idaho National Engineering Laboratory NUREGICR-5406, "BWR Reactor Water Cleanup (INEL)

System Flexible Wedge Gate Isolation Valve Oualifi-cation and High Energy Flow Interruption Test; Vol.1 EGG-SSRE-8972, " Estimating Hazard Functions fo:.

Analysis and Conclusion."

Repairable Components."

NUREG/CR-5406, "BWR Reactor Water Cleanup EGG-SSRE-9017, " User's Guide to PHAZE, a Com-System Flexible Wedge Gate Isolation Valve Qualifi-puter Program for Parametric Ilazard Function Esti-cation and Iligh Energy Flow Interruption Test; Vol. 2, mation."

Data Report.

NUREG/CR-3819," Survey of Aged Power Plant Fa-NUREG/CR-5406, "IlWR Reactor Water Cleanup cilities'"

System Flexible % cdge Gate Isolation Valve Qualifi.

cation and High Energy Flow Interruption Test; Vol.3, NUREG/CR-3956, "In Situ Testing of the Ship.

Review of Issues Associated with IlWR Containment pingport Atomic Power Station Electrical Circuits,"

Isolation Valve Closure "

NUREG/CR-4457, " Aging of Class IE Batteries in NUREG/CR-5448, " Aging Evaluation of Class IE Safety Systems of Nuclear Power Plants."

Batteries: Scismic Testing."

NUREG/CR-4731," Residual Ufc Assessment of Ma, NUREG/CR-5558. " Generic Issue 87: Flexible jor Ught Water Reactor Components," Vol.1.

Wedge Gate Valve's est Program: Phase 11 Results and NUREG/CR-4731," Residual Life Assessmeat of Ma-Analysis."

jor Ught Water Reactor Components-Oversiew,"

Kalsi Engineering. Inc.

Vol. 2.

NUREG/CR-5159, " Prediction of Check Valve Per-NUREG/CR-4740, " Nuclear Plant Aging Research formance and Degradation in Nuclear Power Plant on Reactor Protection Systems "

Systems "

NUREG/CR-4747,"An Aging Failure Survey of Light NUREG/CR-5583, " Prediction of Check Valve Per-Water Reactor Safety Systems and Components,"

formance and Degradation in Nuclear Power Plant Vol.1.

Systems-Wear and Impact Tests."

NUREGICR-4747,"An Aging Failure Survey of Light Lake Engineering Company Water Reactor Safety Systems and Components "

NUREG/CR-5386, " Basis for Snubber Aging Re-

- NUREG/CR-4769, " Risk Evaluations of Aging Phe.

scarch: Nuclear Plant Aging Research Program."

nomena: The Linear Aging Reliability Model and Its Nuclear Regulatory Commission (NRC)

Extensions."

Technical Integration Review Group for Aging and NUREG/CR-4967, " Nuclear Plant A on High Pressure Injection Systems." ging Researci life Extension (TIRG ALEX), " Plan for Integration of 3

Aging and Ufe-Extension Actwities."

NUREG/CR-4977, " SHAG Test Series: Seismic Re-yUREG-il44, " Nuclear Plant Aging Research search on an Aged United States Gate Valve and on a (NPAR) Program Plan.

Piping System in the Decommissioned Heissdampf-NUREG-Il44, " Nuclear Plant Aging Research reaktor (HDR): Summary," Vol.1.

(NPAR) Program Plan," Rev.1.

NUREG/CR-4977, " SHAG Test Series: Seismic Re-NUREG/CP-0!DO, "Proceedmgs of the International search on an Aged United States Gate Valve and on a Nuclear Power Plant Aging Symposium."

I NUREG-1377 66

m Corporate Authorindex Oak Ridge National Laboratory (ORNL)

NUREG/CR-4692,

• Operating Experience Review of Fakes of Power Operated Relief Valves and Block NUREG/CR-2641, "The In Plant Reliability Data Valves in Nuclear Power Plants.

Base for Nuclear Powcr Pitnt Components: Data Col-NUREG/CR-4819. " Aging and Senice Wear of Sole-lection and Methodo!qy Report."

noid-Operated Valves Used in Safety Systems of Nu-NUREG/CR-3154, "The in-Plant Reliability Data clear Power Plants, Vol.1: Operating Experience and Base for Nucicar Plant Components: Interim Failure Identification."

Report-The Valve Component."

NUREG/CR-4992," Aging and Senice Wear of Mul.

NUREG/CR-3543," Survey of Operating Experiences tistage Switches Used in Safety Systems of Nuclear Power Plants," Vol,1.

from LERs toidentify AgingTrends."

NUREGICR-5404," Auxiliary Feedwater System Ag-NUREG/CR-4234," Aging and Service Wear of Elec-ing Study, Vol.1.

tric Motor-Operated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants,"

NUREG/CR-5479, " Current Applications of Vibra-tion Monitoring and Neutron Noise Analysis: Detec-Vol,1.

tion and Analysis of Structural Degradation of Reactor NUREG/CR-4234," Aging and Senice Wear of Elec-VesselInternals from Operational Aging" tric Motor-Operated Valves Used in Engineered NUREG/CR-5519, Vol.1, " Aging of Control and Safety-Feature Systems of Nuclear Power Plants: Ag-Senice Air Compressors and Dryers Used in Nuclear ing Assessments and Monitoring Method Evalu.

Power Plants "

etions," Vol. 2.

NUREG.CR-5706, "NRC Bulletin 88-04: Potential NUREGICR-4257, " Inspection, Surveillance, and Safety-Related Pumpless-An Assessment ofIndus-hionitoring of Electrical Equipment Inside Contain.

try Data.

ment of Nuclear Power Plants-With Applications to Electrical Cables."

Pacine Northwest Laboratory (PNL)

NUREG/CR-4257, " Inspection, Surveillance, and

" Safety implications of Diesel Generator Aging " Nu-Monitoring of Electrical Equipment in Nuclear Power clear Safety, December 1990.

Plants, Vol. 2: Pressure Transmitters."

NUREG/CP-0105. Paper by J. A. Christensen, NUREG/CR-4302, " Aging and Senice Wear of "NPAR Approach to Controlling Aging in Nuclear Power Plants."

Check Valves Used in Engineered Safety-Feature Sys-tems of Nuclear Power Plants," Vol.1.

NUREGICR-4144, *1mportance Ranking Based on Aging Consideration of Components Included in NUREGICR-4302, " Aging and Senice Wear of Probabinstic Risk Assessments.

Check Valves Used In Engineered Safety-Feature Systemsof Nuclear Power Plants;Vol.2, Aging Assess-NUREG/CR-4279," Aging and Service Wear of Hy-nents and Monitoring Method Evaluations."

draulic and Mechanical Snubbers Used on Safety-Related' Piping and Components of Nuclear Power NUREG/CR-4380,

• Evaluation of the Motor-Plants," Vol.1.

OperatedValve AnalysisandTest System (MOVATS)

NUREG/CR-4590," Aging of Nuclear Station Diesel to Detect Degradation, Incorrect Adjustments, and Generators: Evaluation of Operating and Expert Ex-Other Abnormalities in Motor-Operated Valves."

perience," Vols. I and 2.

NUREGICR-4597," Aging and Senice Wear of Aux-NUREG/CR-5057," Aging Mitigation and improved iliary Feedwater Pumps for IWR Nucicar I ower Programs for Nuclear Service Diesel Generators."

Plants, Vol.1: Operating Experience and Failure Iden-NUREG/CR-5379,"N uclear Plant Senice Water Sys-tification."

tem Aging Degradation Assessment: Phase 1," Vol.1.

NUREG/CR-4597, " Aging and Senice Wear of Aux-NUREG/CR-5386, " Basis for Srubber Aging Re-iliary Feedwater Pumps for PWR Nuclear Power search: Nuclear Plant Aging Research Program."

Plants, Vol. 2: Aging Assessments and Monitoring Method Evaluations "

NUREG/CR-5491. "Shippingport Station Aging U"

NUREG/CR-4652," Concrete Component Aging and Its Significance Relative to Life Extension of Nuclear PNL-5722," Operating Experience and Aging Assess-

)

ment of ECCS Pump Room Coolers."

l Power Plants "

67 NUREG-1377

Corporate Author Index Pacific Northwest Laboratory (PNL)(Cont.)

NUREG/CR-5619,'The Impact of Thermal Aging on PNL-6287. " Study Group Review of Nuclear Service the riammability of Electric Cables."

3 Diesel Generator Testing and Aging Mitigation."

NUREG/CR-5653, " Submergence and High Tem.

PNL-7516, " Emergency Diesel Generator Technical perature Steam Testing of Class 1E Electrical Cables."

Specifications Study Results."

SAND 88-0754 UC-78, " Time-Temperature-Dose PNL-SA-18407, " Understanding and Managing Cor-rosion in Nuclear Power Plants, Rate Superposition: A Methodology.for Predicting Ca-ble Degradation Under Ambient Nuclear Power Plant Aging Condidons."

NUREG/CR-5008, " Development of a Testing and.

Analysis Methodology to Determine the Functional Scitnce Applications International Corp.

Condition of Solenoid Operated Valves."

NUREG/CR-5248, "Prioritization of TIRGALEX-Sandia National Laboratories (SAND) fe$h P"

^ 8 NUREG/CP-0036, " Proceedings of the Workshop on Nuclear Plant Aging."

NUREG/CK-5510, " Evaluations of Core Melt Fre-NUREG/CR-3818. " Report of Results of Nuclear

[ [-

E Power Plant Aging Workshop "

NUREG/CR-5334," Severe Accident Testing of Elec-trical Penetration Assemblics."

Wyle Laboratories NUREG/CR-5461, " Aging of Cables, Conrections, NUREG/CR-5386, " Basis for Snubber Aging Re-and Electrical Penetration Assemblics Used in Nu, search: Nuclear Plant Aging Research Program."

clear Power Plants "

WYLE 60103-X, " Test Plan for the Comprehensive NUREG/CR-5546 "An Investigation of the Effects of Aging Assessment of Cixuit Brcakers and Relays for nermal Aging en the Fire Damageability of Electric Nuclear Plant Aging Research (NPAR) Program, Cables."

Phase II."

NUREG-1377 68

~-

SUBJECT INDEX in this index, the reports are listed under one or more of the following subjects:

Aging, including Plans, Surveys, Analyses, Methods, and Models.

1.

Diesel Generators and Related Systems.

2.

Electric Power Systems, including Cables, Trays, Connectors, Circuit Bteakers, Switches, Penetra-3.

tions, and Related Components.

Electrical Equipment, including Transformers, Motors, Batteries, Chargers, and Inverters.

4.

Instrumentation, Mensurement, and Control Systems.

5.

Maintenance.

6.

Major Components: Teactor Vessels, Reactor Coolant Pumps, Steam Generators, Pressurizer 7.

and Structures (Inc.lu fing Containment).

8.

Monitoring.

Operating Experienc :, Field Results, and Related Data.

9.

10. Piping, including Va! <es, Seals, Supports, Snubbers, and Related Lamponents.

11. Probabilistic Risk As essment (FRA).

12. Safety and Protection Systems (Including injection Systems) and Their Components.

13. Seismic Effects and aging.

14. Service Water, Auxiliary Feedwater, Instrument Air, and Other Fluid Systems, including Pumps, Heat Exchangers, at 6 Related Components; Balance of Plant Systems and Cocoonents.

Rese subjects are not intended to include every subject covered in all the reports listed Nor do the represent a " standard" c r" official" list of subjects.They were selected to be most helpful to in able personnel seeking )ublished information on the various aspects of nuclear plant aging.

NUREG/CP--0036, " Proceedings of tL: Workshop 1, Aging, Including Plant, Surveys, Analyses, on Nuclear Plant Aging."

Methods, and Models Letter Report, L N. Rib, Summaries of Research NUREG/CP-0100, " Proceedings of the Imctna.

Reports Submitted in Cotilection with the Nuclear tional Nuclear Power Plant Aging Symposium."

Plant Aging Research (NP. tR) Program," Engineer.

NUREG/CR-3818," Report of Results of Nuclear ing and Economics Research, Inc. (EER).

E Technical Integration Review Group for Aging and NUREG/CR-4144,"importance Ranking Based on life Extension (TIRG ALEX), " Plan for Integration Aging Consideration of Components included in of Aging and Ilfe-Extension Activities."

Probabilinic Risk Assessments."

BNL Technical Report A-3270-12-86, " Aging and NUREG/CR-4652, " Concrete Component Aging Life Extension Assessment Program (ALEAP) Sys-and its Significance Relative to Life Extension of tems Level Ph.n."

Nuclear Power Plants."

EGG-SSRE-8972, " Estimating Hazard Functions NUREG/CR--4731, " Residual Life Assessment of for Repairable Components."

Major Light Water Reactor Components," Vol.1.

EGG-SSRE-9017, " User's Guide to PHAZE, a NUREG/CR-4731, " Residual Life Assessment of Computer Program for Parametric Hazard Function Major Light Water Reactor Components-Over-Estimation."

view," Vol. 2.

NUREG/CP-0105. Paper by L A. Christensen, "NPAR Approach to Controlling Aging in Nuclear NUREG/CR-4769,"RiskEvaluationsof AgingPhe-f nomena:The Linear Aging Reliability Model and its Power Mants, Extensions."

NUREG-1144, " Nuclear Plant Aging Research (NPAR) Program Plan."

NUREG/CR-5008," Development of a Testing and Analysis Methodology to Determine the Functional l

NUREG-1144, " Nuclear Plant Aging Research Condition of Solenoid Operated Valves."

l (NPAR) Program Pian." Rev.1.

NUREG-1377 69

~

Subject Index

1. Aging, Including Plans, Surveys, Analyses, PNL-7516, "Smergency Diesel Generator Technical

- Methods, and Models (Cont.)

Specifications Study Results."

NUREG/CR-5248, "Prioritization of TIRGALEX-

"$ *7 h ded Components for Further Aging

3. Electric Power Systems, including Cables'

. Frays, Connectors, Circuit Breakers, e

Switches, Penetrations, and Related Com.

NUREG/CR-5314, " Life Assessroent Procedures ponents for Major LWR Components: Vol.3, Cast Stainless Steel Components, BNL Technical Report A-3270-12-85, " Pilot As-sessment: Impact of Aping en the Seismic Perform-NUREG/CR-5386, " Basis for Snubber Aging Re.

ance of Selected Equenent Types."

search: Nuclear Plant Aging Research Program."

BNL Technical Report A-3270R-2-90, " Aging Ef-NUREG/CR-5491, "Shippingport Station Aging fects of Important Balance of Plant Systems in Nu-Evaluation."

clear Power Plants."

NUREG/CR-5507,"Results from the Nuclear Plant NUREG/CR-3956, "In Situ Testing of the Ship Aging Research Program: ncir Use in Inspection pingport Atomic Power Station Electrical Circuits.

Activities."

NUREG/CR-4257, ' Inspection, Surveillance, and NUREG/CR-5510, "Fvaluations of Core Melt Fre-M nitoring of Electrical Equipment Inside Contain-wer Plants-With Applications to quency Effects Due to Component Aging and Main-tenance.

,ct C

NUREG/CR-5583," Prediction of Check Valve Per-NUREG/CR-4715"An Aging Assessment of Relays formance and Degradation in Nuclear Power Plant and Circuit Breakers and System Interactiers."

Systems--Wear and Impact Teste "

NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Components-Over.

NUREG/CR-5612, " Degradation Modeling with view,a y og,2, Applications to Aging and Maintenance Effective-ness Evaluation."

NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems and Compo.

PNL-SA-18407, " Understanding and Managing nents,"

Corrosion in Nuclear Power Plants."

Vol.1.

SAND 88-0754 UC-78, Time Temperature-Dose NUREG/CR-4992, " Aging and Service Wear of Rate Superposition: A Methodology for Predicting Multistage Switches Used in Safety Syste.ns of Nu-

- Cable Degradation Under Ambient Nuclear Power clear Power Plants," Vol.1.

Plant Aging Conditions."

NUREG/CR-5181," Nuclear Ptant Aging Research:

WYLE 60103-X, " Test Plan for the Comprehensive The IE Power System."

Aging Assessment of Circuit Breakers and Relays for Nuclear Plant Aging Research (NPAR) Program, NUREG/CR-5280, " Age-Related Degradation of Phase II."

Westinghouse 480-Volt Circuit Breakers; Vol.1, Aging Assessment and Recommendations for Im-

2. Diesel Generators and Related Systems proving Breaker Reliability."

" Safety Implications of Diesel Generator Aging" NUREG/CR-5280, " Age-Related Degradation of Nuclear Safety, December 1990.

Westinghouse 480-Volt Circuit Breakers; Vol 2, Mc-chanical Cycling of a DS-416 Breaker. Test Results."

NUREG/CR-4590," Aging of Nuclear Station Die-sel Generators: Evaluation of Operating and Expert NUREG/CR-5461, " Aging of Cables. Connections, Experience," Vols. I and 2.

and Electrical Penetration Assemblies Used in Nu-clear Power Plants "

NUREG/CR-4731, " Residual Life Assessment of NUREG/CR-5546,"An Investigation of the Effects Major Light Water Reactor Components-Over-of Thermal Aging on the Fire Damageability of Elec-view," Vol. 2.

tric Cables."

NUREG/CR-5057, " Aging Mitigation and Im-NUREG/CR-5619, "The Impact of Thermal Aging proved Programs for Nuclear Service Diesel Genera-on the Flammability of Electric Cables."

tors "

NUREG/CR-5655, " Submergence and High Tem-PNL-6287," Study Group Review of Nuclear Scryice perature Steam Testing of Class lE Electrical Ca-Diesel Generator Testing and Aging Mitigation."

bles."

I NUREG-1377 70 1

Subject Index

3. Electric Power Systems, including Cables,

5. Instrumentation, Measurement, and Con-Trays, Connectors, Circuit Breakers, trol Systems Switches, Penetrations, and Related Com-BNLTechnical Report TR-3270-9-90, "An Opera-tional Assessment of the Baix;ock & Wilcox and ponents (Cont.)

Combustion Engineering Control Rod Drives."

SANDSS-0754 UC-78, " Time-Temperature-Dose Rate Superposition: A Methodology for Predictina NUREG/CR-4257, " Inspection, Surveillance, and Cable Degradation Under Ambient Nuclear Power Monitoring of Electrical Equipment in Nuclear Power Plants, Vol. 2: Pressure Transmitters."

Plant Aging Conditions."

NUREG/CR-4928," Degradation of Nuclear Plant WYLE 60103-X, " rest Plan for the Comprehensive Aging Assessment of Circuit Breakers and Relays for Temperature Sensors."

c car }llant Aging Research (NPAR) Program, NUREGICR-5383, "Effect of Aging on Response Time of Nuclear Plant Pressure Sensors."

NyREG/CR-5546," An 1nvestigation of the Effects of rhermal Aging on the h, re Damageability of Elec-

4. Electrical Equi ment, Including Trans-P formers, Motors, Battuiu, Chargers, and tric cables "

NUREG/CR-5555," Aging Assessment of tl Wes-Inverters BNLTechnical Report A-3270-11-85,"SeismicEn.

tinghouse 1 WR Control Red Drive System.}e durance Tests of Naturally Aged Sman Electric Mo-NUREG/CR-5560," Aging of Nuclear Plant Resis-tors."

tance Temperature Detectors "

BNL Technical Report A-3270-12-85, " Pilot As-NUREG/CR-5655, " Submergence and High Tem-sessment: Impact of Aging on the Seismic Perform.

perature Steam Testing of Class IE Electrical Ca-ance of Selected Equipment Types."

bles.,

BNLTechnical Report A-3270-3-86," Testing Pro-gram for the Monitoring of Degradation in a Con-

6. Maintenance tinuous Duty 460 Volt Class "B",10-HP Electric

" Safety Imphcations of Diesel Generator Aging,"

M'"or-Nuclear Safety, December 1990.

NUREG/CR-4156," Operating Experience and Ag-BNL Technical Report TR-3270-6-90, "Mainte-ing-Seismic Assessment of Electric Motors "

nanceTeam Inspection Results: Insights Related to Plant Aging."

NUREG/CR-4457 " Aging of Class 1E hatteries in NUREG/CR-4234, " Aging and Service Wear of Safety Systems of Nuclear Power Plants "

Electric Motor-Operated Valves Used in Engi-NUREG/CR-4564," Operating Experience and Ag-nected Safety-Feature Systems of Nuclear Power ing-Seismic Assessment of Battery Chargers and In-Plants," Vol. 'L verters "

NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Engineered Safety-Feature NUREG/CR-4939, " improving Motor Reliability in Systems of Nuclear Power Plants; Vol. 2, Aging As-Nuclear Power Plants;" Volume 1: Performance sessments and Monitoring Method Evaluations."

Evaluation and Maintenance Practices; Volume 2:

Functional Indicator Tests on a Small Electric Motor NUREGICR-4457," Aging of Class 1E Batteries in Subjected to Accelcrated Aging: Volume 3: Failure Safety Systems of Nuclear Power Plants "

Analysis and Diagnostic, rests on a Naturally Aged NUREG/CR-4564," Operating Experience and Ag-Electric Motor.

mg-Seismic Assessment of Battery Chargers and NUREG/CR-5051, "Detceting and Mitigating Bat-Inverters."

tery Charger and Inverter Aging" NUREG/CR-4597, " Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power NUREGICR-5053," Operating Experience and Ag.

Plants, Vol.1: Operating Experience and Failurc ing Assessment of Motor Control Centers."

Identification."

NUREGICR -5192, "l'esting of a Naturally Aged NUREG/CR-4597, " Aging and Service Wear of Nudear Power Plant Inverter and Battery Charger "

Auxiliary Feedwater Pumps for PWR Nuclear Power NUREG/CR-5448, " Aging Evaluation of Class IE Plants. Vol. 2: Aging Assessments and Monitoring Method Evaluations."

flatteries: Seismic Testing "

NUREG -1377 71

l Subject Index

6. Maintenance (Cont.)

PNL-SA-18407," Understanding and Managing Cor-NUREG/CR-4939, " Improving Motor Reliability in r sion in Nuclear Power Plants "

Nuclear Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Volume 2:

8, Mon, tor,mg i

Functional Indicator Tests on a Small Electric Motor BNLTechnical Report A-3270--3-86," Testing Pro-Subjected to Accelerated Aging Volume 3: Failure gram for the Monitoring of Degradation in a Con-Analysis and Diagnostic Tests on a Naturally Aged tinuous Duty 460 Volt Class *B",10-IIP Electric Electric Motor.

Motor."

NUREG/CR-5051, " Detecting and Mitigating Bat.

BNL Technical Report A-3270-12-86, " Aging and tery Charger and Inverter Aging."

Life Extension Assessment Program (ALEAP) Sys-tems Level Plan."

NUREG/CR-5057, " Aging Mitigation and Im-proved Programs for Nuclear Senice Diesel Genera-NUREG-1144, " Nuclear Plant Aging Research tors.-

(NPAR) Program Plan," Rev.1.

NUREG/CR-5181," Nuclear Plant Aging Research:

NUREG/CR-3543, " Survey of Operatind,,Experi-The IE Power System."

ences fr m I URs to identify Agmg Frends.

NUREG/CR-5280, " Age-Related Degradation of NUREGICR4234, " Aging and Senice Wear of Westinghouse 480-Volt Circuit Breakers; Vol.1 Electric Motor-Operated Valves Used in Engi-Aging Assessment and Recommendations for Im-nected Safety-Feature Systems of Nuclear Power Plants," Vol.1.

proving Breaker Reliability."

NUREG/CR-4234, " Aging and Senice Wear of NUREG/CR-5280, " Age-Related Degradation of Electric Motor-Operated Valves Used in Engi.

Westinghouse 480-Volt Circuit Breakers; Vol 2, Me-neered Safety-Feature Systems of Nuclear Power chanical Cycling of a DS-416 Breaker. Test Results."

Plants: Aging Assessments and Monitoring Method NUREG/CR-5519, Vol.1. " Aging of Control and Evaluations," Vol. 2.

Service Air Compressors and Dryers Used in Nuclear NUREG/CR-4257, " Inspection, Surveillance, and Power Plants."

Monitoring of Electrical Equipment Inside Contain-NUREG/CR-5612, " Degradation Modeling with ment @udearPowerPlants-With Applicationsto Electrical Cables.,,

Applications to Aging and Maintenance Effective-ness Evaluation."

NUREG/CR-4257, " Inspection, Surveillance, and PNL-5722, " Operating Experience and Aging As-Monitoring of Electrical Equipment in Nuclear Power Plants, Vol. 2: Pressue Transmitters."

sessment of ECCS Pump Room Coolers."

NUREG/CR-4302, " Aging and Senice Wear of

7. Major Components: Reactor Vessels, Re-Check Valves Used in Engineered Safety-Feature actor Coolant Pumps, Steam Generators, Systems of Nuclear Power Plants, Vol.1.

Pressurizers, and Structures (Including NUREG/CR-4457," Aging of Class IE Batteries in Containment)

Safety Systems of Nuclear Power Plants."

NUREG/CR-4652, " Concrete Component Aging NUREG/CR-4564," Operating Experience and Ag-and Its Significance Relative to Life Extension of ing-Seismic Assessment of Battery Chargers and Nuclear Power Plants."

Inverters."

NUREG/CR-4731, " Residual Ufe Assessment of NUREG/CR-4597, " Aging and Senice Wear of Major Light Water Reactor Components," Vol.1.

Auxiliary Feedwater Pumps for PWR Nuclear Power Plants, Vol.1: Operating Experience and Failure NUREG/CR-4731, " Residual Efe Assessment of Identification."

Major Light Water Reactor Components-Over-view," Vol. 2.

NUREG/CR-4597, "' Aging and Senice Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power

- NUREG/CR-5334, " Severe Accident Testing of Plants, Vol. 2: Aging Assessments and Monitoring Electrical Penetration Assemblics."

Method Evaluations."

NUREG/CR-5479," Current Applications of Vibra-NUREG/CR-4819," Aging and Service Wear of So-tion Monitoring and Neutron Noise Analysis: Detec-lenoid. Operated Valves Used in Safety Systems of tion and Analysis of Structural Degradation of Reac-Nuclear Power Plants, Vol.1: Operating Experience tor VesselInternals from Operational Aging."

and Failure Identification."

NUREG-1377 72

^-

Subject index NUREG/CR-2641, "The In Plant Reliability Data

8. Monitoring (Cont.)

llase for Nuclear Power Plant Conpnents: Data NUREG/CR-4939," Improving Motor Reliability in C llection and Methodology Report.

Nuclear Power Plants;' Volume 1: Performance NUREG/CR-3154, "The in Plant Reliability Data Evaluation and hiaintenance Practices; Volume 2:

Base for Nuclear Plant Components: Interim Functional Indicator Tests on a Small Electric Motor Report-The Valve Component."

Subjected to Accelerated Aging: Volume 3: Failure Analysis and Diagnostic Tests on a Naturally Aged NUREG/CR-3543, " Survey of Operating Experi-Electric Motor.

ences from 1.ERs to Identify Aging Trends."

NUREG/CR-4967," Nuclear Plant Aging Research NUREG/CR-3819,

• Survey of Aged Power Plant on liigh Pressure injection Systems."

Facilities "

NUREGICR-5008," Development of a Testing and NUREGICR-4156," Operating Experience and Ag-Analysis Methodelogy to Determine the Functional ing-Scismic Assessment of Electric Motors."

Condition of Solenoid Operated Valves."

NUREG/CR-4234, " Aging and Senice Wear of Electric Motor-Operated Valves Used in Engi-NUREG/CR-5051," Detecting and Mitigating flat.

nected Safety-Feature Systems of Nuclear Power tery Charger and Inverter Aging "

Plants," Vol.1.

NUREG/CR-5053," Operating Experience and Ag-NUREG/CR-4302, " Aging and Senice Wear of ing Assessment of Motor Control Centers.

Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1.

NUREG/CR-5057, " Aging Mitigation and Im-proved Programs for Nuclear Service Diesel Genera-NUREG/CR-4457," Aging of Class 1E Batteries in Safety Systems of Nuclear Power Plants."

10fS-NUREG/CR-5181," Nuclear Plant Aging Research:

NUREG/CR-4564," Operating Experience and Ag-ing-Seismic Assessment of Battery Chargers and ne IE Power System."

NUREG/CR-5192, " resting of a Naturally Aged NUREG/CR-4590," Aging of Nuclear Station Die-Nuclear Power Plant Inverter and Battery Charger."

sel Generators: Evaluation of Operating and Expert NUREG/CR-5461," Aging of Cables, Conne-tions, Experience," Vols. I and 2.

^

NUREG/CR-4597, " Aging and Senice Wear of t s,"

Auxiliary Feedwater Pumps for PWR Nuc! car Power cle Plants, Vol.1: Operating Experience and Failure NUREG/CR-5479 " Current Applicationsof Vibra-Identification."

tion Monitoring and Neutron Noise Analysis: Detec-tion and Analysis of Structural Degradation of Reac' NUREG/CR-4692, " Operating Experience Review tor Vessel Internals from Operational Aging.,

of Failures of Power Operated Relief Valves and-B1 ek Valves in Nuclear Power Plants "

NUREG/CR-5519, Vol.1, " Aging of Control and NUREG/CR-4715, "An Aging Assessment of Re-Senice Air Compressors and Dryers Used in Nuclear lays and Circuit Breakers and System Interactions."

Power Plants."

NUREG/CR-4740," Nuclear Plant-Aging Research NUREG/CR-5612 " Degradation Modeling with on Reactor Protection Systems "

Applications to Aging and Maintenance Effective-ness Evaluation."

NUREG/CR-4747, "An Aging Failure Survey of lj ht Water Reactor Safety Systems and Compo-p PNie6287," Study Group Review of Nuclear Service nents,_

Diesci Generator Testing and Aging Mitigation.-

Vol.1.

NUREG/CR-4747, "An Aging Failure Survey of

9. Operating Experience, Field Results, and 1ight } Vater Reactor Safety Systems and Compo-Related Data

nents,

" Safety Implications of Diesel Generator Aging" Vol. 2.

Nuclear Safety. December 1990.

NUREG/CR-4819," Aging and Service Wear of So-lenoid-Operated Valves Used in Safety Systems of BNI. Technical Report TR-3270-9-90, " An Opera-Nuclear Power Plants, Vel.1: Operating Experience tional Assessment of the Babcock & Wilcox and and Failure Identification."

Combustion Enginecting Control Rod Diives."

NUREG-1377 73 F"

?

Subject index

9. Operating Experience, Field Results, and PNL-7516

• Emergency DiesclGeneratorTechnical Related Data (Cont.)

Spec;fications Study Results "

NUREG/CR-4967,

• Nuclear Plant Aging Research on liigh Pressure Injection Systems.-

10. Piping, including Valves, Seals, Supports, NUREG/CR-4992, " Aging and Senice Wear of Snubbers, and Related Components hiultistage Switches Used in Safety Systems of Nu-llNIfl,echm. cal Report A-3270-II-26-84," Scoping clear Power Plants " Vol.1,

,iest on Containment Purge and Vent Valve Seal hiatenal.

NUREGICR-5052," Operating Experience and Ag-mg Assessment of Component Cooling Water Sys-HNL Technical Report A-3270-12-85, " Pilot As-sessment: IrnPact of A ing on the Seismic Perform.

tems in Pressun, zed Water Reactors.

8 ance of Selected Equipment Types.,,

NUREG/CR-5181,"Nucleac Plant Aging Research:

The lE Power System.

NUREG/CR-3154, "'Ihe In Plant Reliability Data B

h N N Pi m C w m W W NUREG/CR-5268, " Aping Study of Boiling Water Report-The Valve Component."

Reactor Residual IIcat Removal System."

NUREG/CR-4234, " Aging and Service Wear of NUREG/CR-5280, " Age-Related Degradation of Electric Motor-Operated Valves Used in Engi-Westinghouse 480-Volt Circuit Breakers: Vol. 1, nected Safety Feature Systems of Nuclear Power Aging Assessment and Recommendations for Im-Plants," Vol.1.

proving Breaker Reliability "

NUREG/CR-4234, " Aging and cenice Wear of NUREG/CR-5280, " Age-Related Degradation of Electric Motor-Operated Valves Used in Engi-Westinghouse 480-Volt Circuit Breakers; Vol 2. hte-neered Safety Feature Systems of Nuclear Power chanicai Cycling of a DS-416 Breaker. Test Results "

Plants: Aging Assessment and Monitoring Method NUREG/CR-5379, " Nuclear Plant Senice Water Evaluations " Vol. 2.

System Aging Degradation Assessment: Phase 1,"

NUREG/CR-4279," Aging and Senice Wear of IIy-Vol.1.

draulic and Mechanical Snubbers Used on Safety-l NUREG/CR-5383, *Effect of Aging on Response C! ied ping and Components of Nuclear Power Time of Nuclear Plant Pressure Sensors."

I t'

NUREG/CR 5404, " Auxiliary Feedwater System NUREG/CR-4302, " Aging and Senice Wear of Aging Study,,, Vol.1.

Check Valves Used m Engineered Safety-Feature Systems of Nuclear Power Plants" Vol.1.

NUREG/CR-5419, " Aging Assessment of Instru-l ment Air Systems in Nuclear Power Plants.,

NUREb/CR-4302, " Aging and Senice Wear of Check Valves Used in Engineered Safcty-Feature NUREG/CR-5461, " Aging of Cables, Connections, Systems of Nuclear Power Plants; Vol. 2, Aging As-and Electrical Penetration Assemblics Used in Nu.

sessments and Monitoring Method Evaluations "

clear Power Plants."

NUREG/CR-4380, " Evaluation of the Motor-NUREG/CR-5507,"Results from the Nuclear Plant Operated Valve Analysis and Test System Aging Research Program:'Ihcir Use in Inspection (MOVATS) to Detect Degradation Incorrect Ad-Activities *

.justments, and Other Abnormalities in Motor-Oper.

ated Vakes?

NUREG/CR-5519, Vol.1, " Aging of Control and Senice Air Compressors and Dryers Used in Nuclear NUREG/CR-4819, " Aging and Senice Wear of So-Power Plants."

lenoid. Operated Valves Used in Safety Systems of NUREGICR-5555," Aging Assepsment of the Wes-Nuclear Power Plants, Vol.1: Operating Experience tinghouse PWR Control Rod Dnve System."

and Failure Identification."

NUREG/CR-5560, " Aging of Nuclear Plant Resis-NUR EG/CR-4977, "SH AG Test Series: Scismic Re-tance Temperature Detectors."

8c rchonan AgedUnitedStatesGate Valveandona 1,iping System in the Decommtssioned Heissdampf-NUREG/CR-5706,"NRC Bulletin 88-04: Potential reaktor (HDR): Summary " Vol. L Safety-Related Pump Loss - An Assessment of In-ustry Data.

NUH EG'CR-4977 "SH AGTest Scries: Seismic Re-search on an Aged United States Gate Valve and on a PNL-5722 "Opert/ing Experience and Aging As-Piping System in the Decommissioned Heissdampf-sessment of ECCS Pump Room Coolers."

reaktor (HDR): Appendices," Vol. 2.

NUREG-1377 74

Subject Index

10. Piping, including Valves, Seals, Supports,

12. Safety and Protection Systems (Including Snubbers, and Related Components Injection Systems) and Their Components (Cont.)

NUREG/CR-4985,"Indisn Point 2 Reactor Coolant NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Enginected Safety-Feature Pump Seal Evaluations."

Systems of Nuclear Power Plants; Vol. 2, Aging As-NUREG/CR-5008, " Development of a Testing and sessments and Monitoring hiethod Evaluations."

Analysis hiethodology to Determine the Functional NUREG/CR-3819, " Survey of Aged Power Plant Condition of Solenoid Operated Valves."

Facilities."

NUREGICR-5141, "AginE and Qualification Re-NUREG/CR-4302, " Aging and Service Wear of search on So enoid Operated Valves.

Check Valves Used in Engineered SafetpFeature NUREG/CR-5159," Prediction of Check Valve Per-Systems of Nuclear Power Plants," Vol.1.

formance and Degradation in Nuclear Power Plant NUREG/CR-4731, " Residual life Assessment of Systems."

hiajor Light Water Reactor Components-Over-view," Vol. 2.

NUREG/CR-5386, " Basis for Snubber Aging Re.

search: Nuclear Plant Aging Research Program."

NUREG/CR-4740," Nuclear Plant-Aging Research on Reactor Protection Systems."

NUREG/CR-5406,"BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation valve Qualifi-NUREG/CR-4747, "An Aging Failure Survey of cation and High Energy Flaw interruption Test; Vol.

light Water Reactor Safety Systems and Compo-nents,"

1, Analysis and Conclusion."

Vol.1.

NUREG/CR-5406, *HWR Reactor Water Cleanup NUREG/CR-4747, "An Aging Failure Survey of System Flexible Wedge Gate Isolation Valve Qualifi-Light Water Reactor Safety Systems and Compo-cation and High Energy Flow Interruption Test; Vol.

nents,,

2, Data Report."

Vol. 2.

NUREG/CR-5406, "BWR Reactor Water Cleanup NUREG/CR-4967," Nuclear Plant Aging Research System Flexible Wedge Gate Isolation Valve Qualifi.

n High Pressure injection Systems."

ation and High Energy Flow Interruption Test; Vol.

3, Review of Issues Associated with BWR Contain-NUREG/CR-4992, " Aging and Service Wear of mer't isolation Valve Closure."

hiultistage Switches Used in Safety Systems of Nu-

• "I NUREG/CR-5491 "Shippingport Station Aging NUREG/CR-5558, " Generic Issue 87: Flexible Evaluation."

Wedge Gate Valve Test Program: Phase 11 Results NUREG/CR-5558, " Generic Issue 87: Flexible and Analysis."

Wedge Gate Valve Test Program: Phase 11 Results and Analysis."

13. Seismic Effects and Aging Letter Report, hi. Subudhi," Review of Aging-Seis-NUREGICR-5583," Prediction of Check Valve Pet.

mic Correlation Studies on Nuclear Plant Equip-formance and Degradation in Nuclear Power Plant ment," Brookhaven National Laboratory, January a

Systems-Wear and Impact Tests 1985.

PN1 SA-18407, " Understanding and Managing BNLTechnicalReport A-3270-11-85," Seismic En-Corrosion in Nuclear Power Ilants.

durance Tests of Naturally Aged Small Electric hiotors."

11. Probabilistic Risk Assessment (PRA) e n epm As-Aging Consideration of Components included in E (esnue Me NUREGICR-4144, "Importance Ranking Based on e men mpa Ag g n

""C#

Probabilistic Risk Assessments."

NUREG/CR-4156, " Operating Experience and NUREG/CR-5268, " Aging Study of Boihng Water Aging.Scismic Assessment of Electric hiotors "

Reactor Residual Heat Removal System."

NUREG/CR-4279," Aging and Service Wear of Hy-NUREG/CR-5510, " Evaluations of Core hielt Fre-draulic and hiechanical Snubbers Used on Safety-ouency Effects Due to Component Aging and hiain-Related Piping and Components of Nuclear Power tenance."

Plants " Vol.1.

75 NUREG-1377

SubintIndex

\\

t

13. Seismic Effects and Aging (Cont.)

NUREG/CR-4731. " Residual Life Assessment of NUREG/CR-4977," SHAG Test Scries: Seismic Re-h1 jor Light Water Reactor Components-Over-search on an Aged United States Gate Valve and on a y ew, Vol..

Piping System in the Decommissioned Heissdampf-NURIiG/CR-4747, "An Aging Failure Survey of I

l reaktor (HDR): Summary," Vol.1.

Light Water Reactor Safety Systems and Compo-NUREG/CR-4977," SHAG Test Series: Scismic Re-hen s "

j*

g search on an Aged United States Gate Valve and on a Piping System in the Decommissioned Heissdampf-NUREG/CR-4985, rindian Point 2 Reactor Coolant reaktor (HDR): Appendices," Vol. 2.

Pump Seal Evaluations."

NUREG/CR-5448," Aging Evaluation of Class 1E NUREG/CR-5052," Operating Experience and Ag.

Batteries: Seismic Testing.

ing Assessment of Component Cooling Water Sys-tems in Pressurized Water Reactors."

14. Service Water, Auxiliary Feedwater, In-NUREG/CR-5268, " Aging Study of Boiling Water strument Air, and Other Fluid Systems, Reactor Residual IIcat Removal System."

Including Pumps, IIcat Exchangers, and Related Components; Balance of Plant NUREG/CR-5379, " Nuclear Plant Senice Water Systems and Components hg'; tem Aging Degradation Assessment:

Phase I, DNL Technical Report A-3270R-2-90, " Aging Ef-fects of Important Balance of P19nt Systems in Nu.

NUREG/CR-5404, " Auxiliary Feedwater System clear Power Plants."

Aging Study," Vol.1.

NUREG/CR-4597, " Aging and Senice Wear of NURilG/CR-5419, " Aging Assessment of Instru-Auxiliary Feedwater Pumps for PWR Nuclear Power ment Air Systems in Nuclear Power Plants.

Plants, Vol.1: Operating Experience and Failure NUREG/CR-5519, Vol.1, " Aging of Control and Identification."

Service Air Compressors and Dryers Used in Nuclear Power Plants."

NUREG/CR-4597, " Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power NUREG/CR-5706,"NRC Bulletin 88-04: Potential Plants, Vol. 2: Aging Assessments and Monitoring Safety-Related Pump less - An Assessment of In-Method Evaluations."

dustry Data."

NUREG/CR-4731 " Residual Life Assessment of PNL-SA-18407, " Understanding and Managing Major Ligh' Water Reactor Components," Vol.1.

Corrosion in Nuclear Power Plants "

l l

l

[

i I

NUREG-1377 76

i CIIRONOLOGICAL LIST 1NG (in arder of publication) 1.

'NU REG /CR-2641, J. P. Drago, R. J. Borkowski,

10. NUREG/CR-4156, M. Subudhi, E. L 13 urns, and D. H. Pike, and F. F. Goldberg,"The In-Plant Reli-J. H. Taylor, " Operating Expetience and Aging-ability, Data Base for Nuclear Power Plant Compo-Seismic Assessment of Electric Motors /' Brook-nents: Data Collection and Methodology Report,"

haven National labora tory, B NIA. U REG -51861, Oak Ridge National Laboratory, ORN11rM-8271, June 1985 July 1982.

11. NUREGICR-4234 W.L Greenstreet,G. A.Mur-2.

' NU REG /C P-0036, (Compilation by) B. E. Bader phy, and D. M. Eissenberg, " Aging and Service and I. A. Hanchey," Proceedings of the Workshop Wear of Electric Motor-Operated Valves Used in on Nuclear Plant Aging," Sandia Nationallabora.

Engineered Safety-Feature Synems of Nuclear tories, SAND 82-2264C, November 1982.

Power Plants," Vol.1, Oak Ridge Nationallabora-tory, ORNL-6170/V1, June 1985.

3.

NUREG/CR-3154, R. J. Borkowski, W. K. Kahl,

12. NUREG-ll44 B. M. Morris and J. P. Vora, T. L Hebble J. R. Fragola, and J. W. Johnson, "The In. Plant Reliability Data Base for Nuclear

" Nuclear Plant Aging Research (NPAR) Program Plant Components: Interim Report-The Valve Plan," U.S. Nuclear Regulatory Commission, July

1985, Component," Oak Ridge National laboratory, ORN1/PM-8647, December 1983.

13.

NUREG/CR -4257, S. Ahmed, A. Carfagno, and 4.

NUREG/CR-3543, G. A. Murphy, R. B. Gallaher, G. J. Toman, "Insi'ection, Surveillance, and Moni-M. L Casada, and H. C. Hoy, " Survey of Operating toring of Electrical Equipment inside Containment of Nuclear Power Plants-With Applications to Experiences from LERs toidentify AgingTrends,"

Electrical Cables," Oak Ridge National labora-Oak Ridge National laboratory, ORNL.

tory, ORNI./SUB/83-28915/1, August 1985.

NSIC-216, January 1984.

14.

BNL Technica' Report A-3270-11-85, J. H.

5.

NUREG/CR-3818, N. H. Clark and D. L Berry, Taylor, M. Subudhi, J. Higgins, J. Curreri, M.

" Report of Results of Nuclear Power Plant Aging Reich. F. Cifuentes, and T. Nehring, *Scismic En-Workshop," Sandia National Laboratories, durance Tests of Naturally Aged Small Electric SAND 84-0374, August 19S4.

Motors," Brookhaven National 12toratory, 6.

BNL Technical Report A-3270-11-26-84, November 1985.

~

B. Miller, " Scoping Test on Containment Purge 15.

BNL Techm. cal Report A-3270-12-85, M. M. Sil-and Vent Valve Seal Material," Brookhaven Na-ver, R. Vasudevan, and M. Subudhi," Pilot Assess-tional Laboratory, December 1984.

ment: Impact of Aging on the Seismic Performance of Selected Equipment Types," Brookhaven Na-7.

Letter Report, M. Subudhi, " Review of Aging-ti nal 12boratory, December 1985.

Seismic Correlation Studies on Nuclear Plant Equipment " Brookhaven National laboratory, 16.

NUREG/CR-.4302, W. L Greenstreet, G. A. Mur-phy, R. B. Gallaher, and D. M. Eissenberg, " Aging nd Sersice Wear of Check Valves Used in Engi-8.

NUREGICR-4144 T.

Davis, A.

Shafaghi, neered Safety-Feature Systems of Nuclear Power R. Kurth, and F. Leverenz, "Importance Ranking Plants," Vol.1, Oak Ridge National Laboratory, Based on Aging Consideration of Components In-ORNL-6193/V1, December 1985.

cluded in Probabilistic Risk Assessments," Pacific Northwest laboratory, PNL-5389, April 1985.

17.

NUREG!CR-4380, J. L Crowley and D. M. Eis-senberg," Evaluation of the Motor-OperatedValve 9.

N OREG/CR-3819, J. A. Rose, R. Steele, Jr., K. G.

Analysis and,f est System (MOVATS) to Detect DeWall, and B. C. Cornwell, " Survey of Aged Degradation. Incorrect Adjustments, and Other Power Plant Facilities," Idaho National Engineer.

Abnormalities in Motor-Operated Valves," Oak ing laboratory, EGG-2317, June 1985.

Ridge National Iaboratory, ORNL-6226, January 1986.

18.

NUREG/CR-4274, S. H. Bush, P. G. Heasier, and chNcChthn$d, f[$[

R. E. DoJpe." Aging and Service Wear of Hydraulic 77 NUREG-1377

Chronological Listing and Mechanical Snubbers Used on Safety-Related Experience and Failure Identification," Oak Ridge Piping and Components of Nuclear Power Plants,"

National l2boratory, O RN USU B/83-28915/4 /V 1, Vol.1. Pacific Northwest Laboratory, PNL-5479, March 1987.

February 1986.

28. NUREG/CR-3956, M.

R.

Dinsel, M.

R.

19. BNLTechnicalReport A-3270-3-86, A.C. Sugar.

Donaldson, and F. T. Soberano, "In Situ Testing of man, M. W, Sheets, and M. Subudhi, " resting Pro-the Shippingport Atomic Power Station Electrical gram for the Monitoring of Degradation in a Con.

Circuits," Idaho National Engineering 12boratory, -

tinuous Duty 460 Voh Class "B",10-lip Electric EGG-2443, April 1987.

Motor," Brookhaven National laboratory, March 1986.

29. NUREG/CR-4769, W. E. Vesely, " Risk Evalu.

ations of Aging Phenomena:The Linear Aging Re-

20. NUREG/CR-4564 W. E. Gunther, M. Subudhi, liability Model and its Extensions," Idaho National and J. H. Taylor, " Operating Experience and Ag.

Engineering Laboratory, EGG-2476, April 1987.

ing-Seismic Assessment of Battery Chargers and Inverters," Brookhaven National laboratory,

30. Technical Integration Review Group for Aging and BNL-NUREG-51971, June 1986.

Life Extension (TIRGALEX), " Plan for Integra-tion of Aging and Life-Extension Activities," U.S.

21. NUREG/CR-4597, M. L Adams and E. Makay, Nuclear Regulatory Commission, May 1987.

" Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants, Vol.1: Op-

31. NUREG/CR-4715, G. J. Toman, V. P. Bacanskas, crating -Experience and Failure Identification,"

T. A. Shook, and C. C.12dlow,"An Agmg Assess-Oak Ridge National Laboratory, ORNL-6282/V1, ment of Relay and Circuit Breakers and. System July 1986' Interactions," Brookhaven National Laboratory, Franklin Research Center Philadelphia, PA,

22. NUREG/CR-4257, G. J. Toman, " Inspection, Sur-BNL-NUREG-52017, June 1987.

veillance, and Monitoring of Electrical Equipment

32. NUREG/CR-4731, V. N. Shah and -P. E. Mac-m Nuclear Power Plants. Vol. 2: Pressure Frans-mitters, Oak Ridge National 12boratory, ORNU Donald," Residual Life Assessment of Major Light SUB/83-28915/3/V2, August 1986, Water Reactor Components," Vol.1, Idaho Na-tional Engineering laboratory, EGG-2469, June

23. Letter Report, L N. Rib, " Summaries of Research 1987.

Reports Submitted in Connection with the Nuclear

33. NUREG/CR-4928, H. M. Hashemian K. M.

Plant Agmg Research (NPAR) Program," Engi-neering and Economtes Research (EER)Inc., Res-Petersen, T. W. Kerlin, R. L Anderson, and K. E.

ton, VA, September 1986, Holbert, " Degradation of Nuclear Plant Tempera-ture Sensors," Analysis and Measurement Senices

24. NUREG/CR-4652, D.J. Naus, " Concrete Compo-

'E '" "' "

nent Aging and its Significance Relative to Life

34. NUREG/CR-4457, J. L Edson and J. E. Hardin, Extension of Nuc! car Power Plants," Oak Ridge National laboratory, ORN11rM-10059, Septem-

" Aging of Clau IE Batteries in Safety Systems of ber 1986, Nuclear Power Plants," Idaho National Engineer-ng laboratory, EGG-2488, July 1987.

25. PNI-5722, D. E. Blahnik and R. L Goodman,

35. NUREGICR-4747, B. M. Meale and D. G. Sat.

" Operating Experience and Aging Assessment of terwhite,"An Aging Failure Survey of Light Water ECCS Pump Room Coolers," Pacific Northwest Reactor Safety Systems and Components," Vol.1, 12boratory, October 1986.

Idaho Nationai Engineering laboratory, EGG-2473, July 1987.-

26. BNL Technical Feport A-3270-12-86, R.

i

-- Fullwood, J. C. Higgins, M. Subudhi, and J. H.

36. NUREG/CR-4590, K. R. Hoopingarner, J. W.

Taylor, " Aging and Life Extension Assessment Pro-Vause, D. A. Dingee, and J. F. Nesbitt, " Aging of gram (ALEAP) Systems level Plan," Brookhaven Nuclear Station Diesel Generators: Evaluation of National laboratory, December 1986.

Operating and Expert Experience," Vols. I and 2, Pacific Northwest laboratory, PNL-5832, August 27.

NUREG/CR-4819, V.

P. Bacanskas, G. C.

1987.

Roberts, and G. J. Toman, " Aging and Service Wear of Solenoid-Operated Valves Used in Safety 37.

NUREG/CR-4985, M. Subudhi, J. H. Taylor, J.

Systems of Nuclear Power Plants, Vol.1: Operating Clinton, C. J. Czajkowski, and J. Weeks, " Indian NUREG-1377 78 i

)

Chronological Listing Point 2 Reactor Coolant Pump Seal Evalua-

46. NUREG/CR-4597, D. M. Kitch, J. S. Schlonski,

-tions," Brookhaven National Laboratory, BNL-P. J. Sowatskey, and W. V. Cesarski, " Aging and NUREG-52095, August 1987.

Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants, Vol. 2: Aging Assess-

38. NUREG-1144, J. P. Vora, " Nuclear Plant Aging ments and Monitoring Method Evaluations," Oak Research (NPAR) Program Plan," Rev.1 U.S.

Ridge National laboratory, ORNir6282/V2, June 1988-Nuclear Regulatory Commission, September 1987.

47.

NUREG/CR-4747, B. M. Meale and D. G. Sat.

39. NUREG/CR-4992, G.

C.

Roberts, V,

P.

terwhite,"An Aging Failure Surveyof Light Water Bacanskas, and O. J. Toman, " Aging and Service Reactor Safety Systems and Components," Vol. 2, Wear of Multistage Switches Used in Safety Sys-Idaho National Engineering laboratory, tems of Nuclear Power Plants," Vol.1, Oak Ridge EGG-2473, July 1988.

National Laboratory, ORNIJSUB/83-28915/5/VI, September 1987.

48.

NUREG/CR-5052, J. C. Higgins, R. Iofaro, M. Subudhi, R. Fullwood, and J. H. Taylor, "Oper.

- 40.

NUREG/CR-5008, R. D. Meininger and T. J.

ating Experience and Aging Assessment of Compo-Weir, " Development of a Testing and Analysis nent Cooling Water Systems in Pressurized Water Methodology to Determine the Functional Condi-Reactors," Brookhaven National laboratory, tion of Solenoid Operated Valves" Pentek, Inc.,

BNieNUREG-52117, July 1988.

Coraopolis, PA, September 1987.

49.

NUREG/CR-5053, W. Shier and M. Subudhi, 41.

NUREG/CR-4692, G. A. Murphy and J. W.

" Operating Experience and Aging Assessment of Motor Control Centers," Brookhaven National Cletcher II," Operating Experience Review of Fail, ures of Power Operated Relief Valves and Block Laboratory, BNieNUREG-52118, July 1988.

Valves in Nuclear Power Plants," Oak Ridge Na-

50. NUREG/CR-5051, W. E. Gunther, R, Lewis, and t onal Laboratory, ORN11NOAC-233, October M. Subudhi, " Detecting and Mitigating Battery Charger and inverter Aging" Ilrookhaven Na-tional laboratory, BNicNUREG-52108, August 42.

NUREG/CR-4939, M. Subudhi, W. E. Gunther, 1988.

J. H. Taylor, R. lofaro, K. M. Skreiner, A. C.

Sugarman, and M. W. Sheets, " Improving Motor S t.

NUREG/CR-5141, V. P. Bacanskas, G. J. Toman, Reliability in Nuclear Power Plants;" Volume 1:

and S. P. Carfagno, " Aging and Qualification Re-Performance Evaluation and Maintenance Prac-search on Solenoid Operated Valves," Franklin tices; Volume 2: Functional Indicator Tests on a Research Center, Norristown, PA August 1988.

Small Electric Motor Subjected to Accelerated Aging; Volume 3: Failure Analysis and Diag-52.

' SAND 88-0754 UC-78, K. T. Gillen and R. L nostic Tests on a Naturally Aged Electric Clough, ' Time-Temperature-Dose Rate Superpo-Motor; Brookhaven National laboratory, UNL-sition: A Methodology for Predicting Cable Degra-NUREG-52031, November 1987.

dation Under Ambient Nuclear Power Plant Aging Conditions,"Sandia National Laboratories, August

43. NUREGICR-4740, L C. Meyer, " Nuclear Plant-1988-Aging Research on Reactor Protection Systems,

53. NUREG/CR-5192, W. E. Gunther, "TestinS of a Idaho National Engineering laboratory, Naturally Aged Nuc! car Power Plant inverter and EGG-2467, January 1988.

Battery Charger," Brookhaven National 12bora-tory, llNL-NUREG-52158, September 1988.

'44, PNL-6287, K. R. Hoopingarner, B. J. Kirkwood, f-and P. J. Lonzecky, " Study Group Review of Nu-

54. NUREG/CR-5248, I. S. Levy, D. B. Jarrell, and clear Service Diesel Generator Tc. sting and Aging E. P. Collins, "Prioritization of TIRG ALEX.

Mitigation," Pacific Northwest Laboratory, March Recommended Components for Further Aging Re-

)

1988-search," Pacific Northwest Idooratory, Science Ap-plications International Corp., PNL-6701, Novem-

- 45.

NUREG/CR-5159, M. S. Kalsi, C. L Horst, and ber 1988.

J. K. Wang, " Prediction of Check Valve Perform-ance and Degradation in Nuclear Power Plant Sys-tems." Kalsi Engineering. Inc., Sugar land, TX,

• D $ $ $"y y % C y M g6 l

KE! No.1559, May 1988-NPAIM.ogram 79 NUREG-1377 i

ChronologicalIJsting

55. NUREG/CP-0100, A. F. Beranek,

• Proceedings of 64.

NUREG/CR-5406, K, G. DeWall and R. Steele, the International Nuclear Power Platit Aging Sym-Jr., *BWR R eactor Water Cleanup System l'lexible posium," U.S. Nuclear Repclatory Commission, Wedge Gate Isolation Valve Qualification and March 1989, liigh Energy Flow Interruption Test; Vol.1, Analy-sis and Conclusion," Idaho National Engineering

56. NUREG/CR-5268, R. lefaro, bl. Subudh<. W. E.

Imboratory, EGG-2569, October 1989.

Gunther, W. Shier, R. Fullwood, and J. H. Wylor,

65. NUREG/CR-5406, K. G. DeWall and R. Steele,

" Aging 5tudy of Boiling Water Reactor Resdual lleat Removal System," BNis-NUREG-52177, Jr.,"BWR Reactor Water Cleanup System Flexible Brookhaven National laboratory, June 1989.

Wedge Gate Isolation Valve Qualification and High Energy Flow Interruption Test; Vol. 2, Data Report " Idaho National Engineering laboratory,

57. NUREG/CR-5379, D. B. Jarrell, A.11. Johnson,

, Octok M Jr., P. W. Zimmerman, and M. L Gore, " Nuclear Plant Service Water System Aging Degradation As-

66. NUREG/CR-5406, K. G. DeWall and R. Steele, sessment: Phase 1," Vol.1, Pacific N orthwest Labo' Jr.,"BWR Reactor Water Cleanup System Flexiole ratory, PNir6560, June 1989.

Wedge Gate Isolation Valve Qualification and High Energy Flow InterruptionTest; Vol.3, Review

58. NUREG/CR-5383, H. M. Hashemian, K. M.

of Issues Associated with BWR Containment Isola-Petersen, R. E. Fain, and J. J. Gingrich, "Effect of tion Valve Closure," Idaho National Engineering Aging on ResponseTime of Nuclear P; ant Pressure Irboratory, EGO-2569, October 1989,-

Sensors," Analysis and Measurement Services Cor-poration, Knoxville, TN, June 1989.

67. NUEEG/CR-4731, V. N. Shah and P. E. Mac-Donald " Residual Life Assessment of Major Light

59. WYLE 60103-X, J. F. Gleason, R. A. DeFour, J.

Water Recctor Components-Overview," Vol. 2 M. Hammond, and P. A. Lubeski, " Test Plan for (Draft), Idaho National Engineering laboratory, the Comprehensive Aging Assessment of Circuit EGG-2469, November 1989.

Breakers and Relays for Nuclear Plant Aging Re-search (NPAR) Program, Phase 11," Wyle Labora-NUREGICR-5334 D. B. Clauss, " Severe Acci; 68.

tories,13untsville, AL, July 1989*

dentTestingof Electrica! Penetration Assemblies, Sandia National laboratories, - SAND 89-0327, 60.

NUREG/CR-4234,~ H. D. Haynes, " Aging and Service Wear of Electric Motor-Operated Valves

69. NUREG/CR-5057, K. R. Hoopingsrner and F. R.

Used in Engineered Safety-Feature Systems of Nu.

Zaloudek, " Aging Mitigation at.d Ireproved Pro-clear Power Plants: Aging Assessments and Moni-grams for Nuclear Service Diesel Genetetors," Pa.

toring Method Evaluations," Vol. 2, Oak Ridge

- cific Northwest laboratory, PNL4397, Dc.cember National Laboratory, ORNL-6170/V2, August

1989, 1989.

70. NUREG/CR-5386, D. P. Brown, G. R. Palmer, E, 6L NUREG/CR-4967 L C. Meyer, " Nuclear Plant V. Werry, and D. E. Blahnik " Basis for Snubber Aging Research on High Pressure injectiora Sys-Aging Research: Nuclear Plant Aging Research tems," Idaho National Engineering Laboratory, Program, Pacific Northwest Laboratory, take En-EGG-2514, August 1989.

gineering Company, Wyle l2boratories, PNL-6911, January 1990.

. 62. NUREG/CR-4977, R. Steele, Jr. and J. G.

- 71. NUREG/CR-5419, ht Villaran, R Fullwood, and -

.Arendts, SHAG Test Series: Seismte Research on M. Subudhi," Aging Assessment of Instrument Air an Aged United States, Gate Valve and on a Piping Systems in Nuclear Power Plants," B rookhaven Na-System m the Decotymtsstoned Heissdampfreak tor tional Laboratory, BNirNUREG-52212, January i

(HDR): Summary, Vol.1. Idaho National Erigi-1990' neering Laboratory, EGC-2505, Augnst 1989.

72. NUREG/CR-5491, R. P. Allen and A. B. Johnson

63. NUREG/CR-4977 R. Steele, Jr. cA J. G.

Jr., "Shippingport Station Aging Evaluation." Pa-Arendts," SHAG Test Series: Seismic Resea@ on cific Northwest Laboratory, PNL-7191, January an Aged United States Gate Valve and on a Pipir,

1990.

e System in the Decommissioned Heissdampireaktor (HDR): Appendices," Vol. 2, Idaho National Engi-73.

HNL Technical Report A-3270R-2-90, A. Fresco neering 12boratory, EGG-2505, August 1989.

and M. Subudhi, " Aging Effects of important Bal-NUREG-1377 80

Chronological Listing ance of Plant Systems in Nucicar Power Plants,"

Assessment and Recommendations for improving ilrookhaven National laboratory, February 1990.

Breaker Reliability," Hrookhaven National Iabo-ratory, llNL-NUREG-52178, J uly 1990.

74.

NUREG/CR-5479, H. Damiano and R. C. Kryter,

" Current Applicationsof Vibration Monitoringand 84.

NUREG/CR-5461, M. J. Jacobus, " Aging of Ca.

Neutron Noise Analysis: Detection and Analysisof bles, Cor nections, and Electrical Penetration As-Structural Degradation of Reactor Vesselinternals semblics Used in Nuclear Power Plants," Sandia from Operational Aging

• Oak Ridge National Nationallaboratories, SANDS 9-2369, July 1990.

Iaboratory, ORNIll M-11398, February 1990.

85.

N UREG /CR-5519, Vol.1 J. C. Moyers, " Aging of 75.

NUREG/CP-0105, hoceedings of the Seven.

Control and Service Air Compressors and Dryers teenth Water Reactor Safety Information Meeting, Used in Nuclear Power Plants," Oak Ridge Na-Vol. 3, U. S. Nuclear Regulatory Commission. Pa-tioned laboratory, ORNL-6607/V1, July 1990.

per by J. A. Christensen,"NPAR Approach to Con-trolling Aging in Nuclear Power Plants," Pacific 86.

NUP.EG/CR-5448.J. L Edson," Aging Evaluation Northwest laboratory, PNL-SA-17487, March of Class 1E Hatteries: Seismic Testing " Idaho Na-1990.

tional Engineering laboratory, EGG-2576, August 1990.

76.

NUREG/CR-5404, D. A. Casada, " Auxiliary Feedwater System Aging Study " Vol.1, Oak Ridge 87.

NUREG/CR- $583, M. S. Kalsi, C. L Horst, J. K.

National laboratory, ORNL-6566/VI, h1 arch Wang and V. Sharma, " Prediction of Check Valve Performance and Degradation in Nuclear Power 1990.

Plant Systems-Wear and Impact Tests," Kalsi En-77.

EGG-SSRE-8972, C. L Atwood, " Estimating gineering,Inc., KEI No.1656, August 1990.

Ilazard Functions for Repairable Components,"

Idaho National Engineeringlaboratory, May 1990.

88.

PNL-SA-18407, A. H. Johnson, Jr., D. B. Jarrell, U. P. Sinha, and V. N. Shah, " Understanding and Managing Cortc' ion in Nuclear Power Plants," Pa-

'/8.

NUREG/CR-5181, L C. Meyer and J. L Edson, cific Northwest Laboratory, August 1990.

" Nuclear Plant Aging Research The lE Power Sys-tem," Idaho National Engineering Iaboratory, 89.

UNL Technical Report TR-3270-9 90, E. Grove EGG-2545, May 1990.

and W. Gunther,"An Operational Assessment of the Babcock & Wilcox and Combustion Engineer.

79.

HNL Technical Report TR-3270-6-90, W. Gun-ing Control Rod Drives," Brookhaven Natioaal ther, " Maintenance Team inspection Results: In-laboratory, September 1990, sights Related to Plant Aging," Brookhaven Na-tional Laboratory, June 1990.

90.

NUREG/CR-5507, W. Gunther and J. Taylor, "Results from the Nuclear Plant Aging Research 80.

N U REG /C R-5510, W. E. Vesely, R. E. Kurth, and Program: Their Use in Inspection Activities,"

S. M. Scal 70," Evaluations of Core Melt Frequency Hmokhaven National Laboratory, UNL-Effects Due to Component Aging and Mainte-NUREG-52222, Septemoer 1990, nance " Science Appiications laternational Corpo-ration, SAIC-S9/1744, June 1990.

91.

NUREG/CR-5314, C. E. Jaske and V. N. Shah,

" Life Assessment Procedures for Major LWR 81.

NUREG/CR-5560, H. M. Hashemian, D. D.

Components; Vol.3, Cast Stainless Steel Compo-Deverly, D. W. Mitchell, and K. M. Petersen, " Ag-nents," Idaho National Engineering I2boratory, ing of Nuclear Plant Resistance Temperature De-EGG-2562, October 1990, tectors," Analysis and Measurement Senices Cor-92.

NUREG /CR-5280, M. Subudhi, E. MacDougal, S.

poration, June 1990.

Kochis, W. Wilhelm, and B. S. Lee, " Age-Related (U.

EGG-SSRE-9017, C. L Atwood," User's Guide to Degradation of Westinghouse 480-Volt Circuit PH AZil, a Computer Program for Parametric Haz.

Breakers; Vol. 2, Mechanical Cycling of a DS-416 mJ Function Estimation," Idaho National Engi-Breaker. Test Results," Hmokhaven National neenq laboratory, July 1990.

laboratory, UNL-NUREG-52178, November 1990.

83. NUREG/Ch -5280. M. Subudhi, W. Shier, and E.

MacDougall, " Ar-Related Degradation of Wes-93.

K. R. Hoopingarner and F R. Zaloudek " Safety tinghouse 480Nolt Cbcuit Breakers:Vol.1. Aging implications of Diesel Generator Aging," Pacific 81 NUREG-1377

I Chronological Listing Northwest 1aboratory, Nuclear Safety, 31:484-489,

98. PNIA516, K. R. Ilotpingarner, " Emergency Die.

October-December 1990, sel Generator Technical Specifications Study Re.

suits," Pacific Northwest laboratory, March 1991.

94. NUREG/CR-5558, R. Steele, Jr., K.O. DeWall.

99. NUREG/CR-1302 H. D.11aynes, " Aging and and J. C, Watkins, " Generic issue 87: Flenb!c Service Wear of Check Valves Used in Enginected Wedge Gate ValveTest Program: Phase 11 Results Safety-Feature Systems of Nuclear Power Plants; and Analysis," Idaho National Engineeringl2bora.

Vol. 2, Aging Assessments and Monitoring Method tory, EGO-2600, January 1901.

Evaluations," Oak Ridge National Laboratory, April 1991.

95. NUREG/CR-5555, W. Gunther and K. Sullivan,

" Aging Assessment of the Westinghouse PWR 100. NUREG/CR-5546, S. P. Nowlen, "An lavestiga-Control Rod Drive System," Brookhaven National tion of the Effects of 1hermal Ag, g on the Iire m

laboratory, ilN!s-NUREG-52232, March 1991.

Damageabilityof Elec'.ricCables, Sandia National Laboratories, SAND 90-0696, May 1991.

96. NUREG/CR-5612, P. K. Samanta, W. E. Vesely, 101. NUREG/CR-5655, M. J. Jacobus and G. F. Fuch-F, lisu, and M. Subudhi, " Degradation Modeling rer," Submergence and High Temperature Steam with Applications to Aging and Main'enance Effec.

Testing of Class IE Electrical Cables," Sandia Na-tiveness Evaluation." Brookhaven National Labo.

tional laboratories, S AND90-2629, May 1991.

ratory, UNIs-NUREG-52252, March 1991, 102. NUREG/CR-5706, D. A. Casada, "NRC Bulletin 88-04: Potential Safety-Related Pump loss - An

97. NUREG/CR-5619, S. P. Nowlen, "The Impact of Assessment of Industry Data," Oak Ridge National 1hermal Aging on the Flammability of Electric Ca-laboratory, ORNL-6671, June 1991.

bles," Sandi.v National Iaboratories, SAND 90-2121, March 1991, NUREG-1377 82

U 8, NUCLE AR REGULATORY COMM:SSON 1, 54EPORT NUMBER NRC PORM $36 Supp., Flev...RC. Add Vol.

(Assigned by N nd A00.ndum Nam-(249) t#iCM 1102 t"-

" **V 1 i

3200 3202 BIBLIOGRAPHIC DATA SHEET NUREG-1377 ts instrucuans on in. r.v.rs.

2. TifLf. AND SVBillLE 3 OAIE f(EPOHT PUUUS,tO NRC Research Program on Plant Aging: Listing and Summaries uourH l

vtAR of Reports issued hrough June 1991 3993

4. FN OR GRANT NUMDER 6 rYPL OF Itt PORT

6. AJTF,0 Rib)

Technical N. N. Kondic, E. L Hill

7. FTR'OD COVERED (inclusiv. D.i.s) 1981-1991

8. fi.kFOH.MiNQ OHGANIZ.ATION - NAME m.ND.no m.8tE 68 (rf NRC, previo. Divi. ion. Offic. or R.gson, U. S. Nucl..r Regul. tory Com A

ADD m mna ae.ss; it contr cior, previo. n mne.oe..sa Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555

9. SPONSOHING ORGANI2 ATION - NAME.nd m.mng.ESS (if NHC. typ. "6.m..s. boy."; le contr.ctor, provia. NRC Division, Ottic. or R.g AND ADD'4 U. S. Nucl..r R. gut. tory Committen, de.ss )

Same as above.

10. tit #'PLEMENT AHY NOTES

11. ABSTRACT (200 words or I.ss)

The U.S. Nuclear Regulatory Commission is conducting the Nuclear Plant Aging Research (NPAR) Program. This is a comprehensive hardware-oriented engineering research program focused on understanding the aging mecha-nism:: of components and systems in nuclear power plants. The NPAR program also focuses on methods for simu-lating and monitoring the aging-related degradation of these components and systems. In addition, it provides ree-ommendations for effective maintenance to manage aging and for the implementation of the research results in the regulatory process.

This document contains a listing and index of reports generated in the NPAR program that were issued through June 1991 and summaries of those reports. Each summary describes the elements of the research covered in the report and outlines the significant results. For the convenience of the user, the reports are indexed by personal author, corporate author, and subject, is Av^ueurY srATEMeNT tr. xEv woRosiOEsewTORs (u.i worai or pnr.... in.1 wm...i.i r....ren.r. in ioc.ung in. r. port. i Unlimited Nuclear Plant Aging Research (NPAR) aging mechanisms Unclassified aging mitigation (This keport) compilation Unclassified life extension is. ruMota Oe eAc.es plant aging

16. PRtCE NRC FORM 336 (2-69)

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