ML20246F588

NRC Research Program on Plant Aging: Listing and Abstracts of Reports Issued Through February 1, 1989

ML20246F588
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Issue date:	08/31/1989
From:	Kondic N NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
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NUREG-1377, NUDOCS 8908310069
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Availability of Reference Materials Cited in NRC Publications-

.l 'I - Most documents cited in NRC publications will be available from ono.of the following - Sourcest 1,. The NRC Public Document Room, 2120 L Street, NW, Lower Level, Washington, DC l , 20555 3: 2/ The Superintendent

• of Documents, U.S. Government Printing Office, P.O.' Box 37082, Washington, DC 20013-7082 3,

The Nationa! 'echnical Information Service, Springfield. VA 22161 ) Although the listing that follows reprer.ents the majority of documents cited in NRC publica-tbns, it is not intended to be exhaustive. j RcSronced doctrnents available for inspection and copying for a fee from the NRC Public Document Roorn 4ncludo NRC correspondence ;and internal NRC memoranda; NRC Office of l inspection and Enforcement bulletins, circulars, information notices, inspection and investi-pation notices; Licensee Event. Reports; vendor reports and correspondence'; Commission papers; and applicant and licensee documents and correspondence. -I The following documents in the NUREG series are available for purchase from the GPO Sales ~j Propigm; formal NRC staff and contractor reports, NRC-sponsored conference proceedi -l 'ings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regula-j tions in the Code of Federal Regulatisos. and Nuclear Regulatory Commission lssuances. Documents available from the Nationa' Technica, information Service include NUREG series 'l repc/ts and technical reports prepared by other fecoral agencies and reports prepared by .l tho' Atomic Energy Comniission, foreruntier egency to the Nuclear Regulatory Commission. ( Documerns availab!e from pub'ic and special technical libraries include all open literature items, such as books, jourrel and periodical articMs, and transactions. Federal Register r$odcas, federal and state legislation, and congfestbnal reports can usually be obtairied irom *hese liuraries, j j l Documents such es theses, desertations, foreign reports and translations, and non-NRC j conference proceedirtgs are available for purchase from the organization sponsoring the ] publication cited. I Singio copies of NRC draft reports nre availabte free., to the extent of supply, upon written request to the O$ce of information Resources Management Distribution Section, U.S. l Nuclear Regulatory Carnmiscion, Wathington DC 20555. l l Copies of industry codes tend standards used in a substantive mannoi in the NRC segulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there fo: reference use Dy the public. Codes and standards are usually copy-righted and niay be purchased from the originating organization or, if they are American National Standards. from the Americ6n National Sta.3dards institute,1430 Broadway, New York, NY 10018. l 1 i

-{ " g+ m- .g %g;, NUREG-1377. l 4 pa. i l. 1 i l l NRC:Research Program on l Plant Aging: Listing and Abstracts.of Reports Issued j Tarough Februm y 1.1989 i ' Manuscript Completed: July 1989 Date Published August 1989

N. N. Konaic Division of Engineering Office of Nuclear kegulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 f.. =%y i

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ABSTRACT The U.S. Nuclear Regulatory Commission is conducting the Nuclear Plant Aging Research (NPAR) Program. This is a comprehensive hardware-oriented program focused on understanding the aging mechanisms of components and systems in nuclear plants. The NPAR program also foceses on methods for simulating and monitoring the a@;ng-related degradation of these components cn systems. This document contains a listing and index of reports generated in the NPAR Program that w0re issued through February 1,1989, and abstracts of those reports. Each abstra:t 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-1 1 iii 4 L___________ i

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-\\r i i g< TABLE 0F' CONTENTS 4 Page .4PSTRACT c................ iii PREFAC.T........... .m.............................................. vii ACKNOWLEDGEMENT.....e............................................... ix

INTRODUCTION........,..................<..........................

1 . MAIN CITATIONS AFD ABSTLACTS....................................... 3' t PERSON #L AUTH0R INDEX.............................................. 33 CORPOR ATE A UTHO R I NDEX.........................................,,.. 47 SUBJECT INDEX..................................................... 51 l 1-CF20N0 LOG I CAL L I STI NG............................................. 61 I i 4 1 i 1 V 1 l l i

l 1 l 1 PREFACE The Office of Nuclear Regulatory Research of the U. S. Nuclear Regulatory Commission (NRC) is conducting a hardware-oriented engineering research program dealing with the aging of nu: lear power plant components and systems. This program it c'escribed in hilREG-1144, Rev.1, " Nuclear Plant Aging Research (NPAR) Program Plan," published in September 1987. Significant progress has been made in defining aging degradation mechanisms and sites, and effective monitoring and surveillance methods were developed for many of the components and systems identified in NUREG-1144, Rev.1. Some of these components and systems are motor-operated valves, check valves, solenoid-operated valves, electric motors, emergency diesel generators, chargers and inverters, circuit breakers and relays, t,atteries, auxiliary feedwater pumps, and reactor protection systems. Progress has also been made in developing models and approaches to evaluate the relative impacts of aging on risk. The Fhase I reseerch for evsluating system-level aging effects based on operating I experience and risk evaluation of the aging phenomena has been largely complet-ed. Significant accomplishments have been achieved in identifying major technical safety issues ond defining the risk significance of major light water reactor components and structures for application to plant life extensioc and license renewal issues. This document contains suurnaries of NRC-sponsored reports that were generated in the NPAR Prcgram. Each summary describes the objectives of the research, identifies the contractor and the authors involved, and outlines significant research results. If the readers of this document need additional information on a particular report and the findings discussed therein, they 'are encouraged to contact either the appropriate NRC project manager or contractor. This report is intenced to be a living document. It will be updated periodi-cally to incorporate the sununaries of fiFAR-generated reports as they are completed. Comments on this document are welcome end will be considered in developing subsequent revisions. '~] l 44 " ! ;1 tong @, Chief Electrical & Mechanical Engineering Branch Division of Engineerir.g Office of Nuclear Regulatory Research vii

x i ACKNOWLEDGEMENT The following valuable contributions ~to the-preparation of'this-document-are appreciated: Members of RES Electrical & Mechanical Engineering Branch, in particular, h M. Vagins for suggestions and advice, J. Vora and the other Nuclear Plant Aging Research ' program managers for. their inputs, and S. Harty for her work on the text. From the RES Program Managenent Policy Development & Analysis Staff, E. Hill, Technical Editor, for his involvement and extended efforts to improve the L format, style, and structure of the report. l The Electronic Composition Services Unit for expeditious and high-quality l processing of the manuscript. ix

V INTRODUCTION l-This document is a listing and index of reports related to the Nuclear L Plant Aging Research (NPAR) Program issued through February 1, 1989.. The first listing is in alphanumeric order by-report number and includes an abstract of L each report. Three indexes are provided to aid the user in-retrieving a specific report: Personal Author Index, Corporate Author Index, and Subject Index. Finally, there is a listing in chronological order by date of-publication. Most of the reports contain a description of the components or systems being examir:ed and identify the principal stressors leading to aging, They frequently'contain an analysis and statistical assessment of failure data obtained from Licensee Event Reports and other sources of component failure data for operating nuclear power plants. Current surveillance and conitoring practices are also reviewed and, when identifiable, recommendations are made for improvements. The information contained in the reports should be of interest to those assessing the aging and reliability of nuclear power plant components, including researchers and designers as well as maintenance and operations personnel. Most of the documents cited in this report are available from one of the following sources: 1. The NRC Public Document Room, 2120 L Street NW., Lower Level, Washington, DC. 2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082, Washington, DC 20013-7082. 3. The National Technical Information Service, Springfield, VA 22161. l l I l l I _A

l I MAIN CITATIONS AND ABSTRACTS l. The reports listed in this compilation are arranged alphanumerically by report i number, with unnumbered reports preceding the numbered reports. The biblio- ~ graphic information is followed by a brief abstract of each report. UNNUMBERED REPORTS I Letter Report, M. Subudbi, " Review of Aging-Seismic Correlation Studies on Nuclear Plant Equipment," Brookhaven National Laboratory, January 1985. During the last decade, the issue relating to aging seismic correlation of nuclear grade equipmcnt and their components has received special attention by ) both the NRC and the utility industry with the aim of preventing catastrophic failures of aged nuclear power plant components during a seismic event. This report summarizes the work performed by the Seismic Qualification Utilities Group (SQUG) based on real earthquake data, by NUTECH for Sandia National Laboratories, and by EPRI at Wyle. Based on the above, an outline of the work to be carried out at BNL under the NPAR scope relating to identifying the aged components sensitive to seismic loadings is provided. EQE, Inc., sponsored by the Seismic Qualification Utilities Group has gathered a comparative data base on the performance of equipment in five fossil-fueled plants consisting of 24 units and a high-voltage DC-to-AC converter station. These plants have experienced four damaging California earthquakes of Richter Magnitudes 5.1 to 6.6. Peak horizontal ground acceleration (PGA) of these earthquakes ranged between 0.2 g and 0.5 g. The actual earthquake-induced effects on equipment were compared with equipment qualification data from three nuclear plants. The objective of the p' lot program was to determine the feasibility of establishing criteria for assessing the seismic adequacy of equipment in nuclear power plants based on evaluation and application of data to be acquired on the characteristics and seismic performance of equipment in nonnuclear power facil-l ities that have been subjected to strong-motion earthquakes. Application of the criteria would provide a valid basis for assessing the need for subsequent qualification efforts in the nuclear industry and for defining the extent of the effort. Letter Report, L. N. Rib, " Summaries of Research Reports Submitted in Connection with the Nuclear Plant Aging Research (NPAR) Program," Engineering and Economics j Research (EER) Inc., Reston, VA, September, 1986. i The results of Phase I efforts in the NRC NPAR program for selected electri-j cal and mechanical components since 1984 have been published. To help maintain cognizance of this wealth of info mation, summaries of 14 reports are presented in this publication. Thus the results of these studies are made more readily I l

MAIN CITATIONS AND ABSTRACTS l l l available for rapid survey, directing attention to specific reports of interest l and facilitating the utilization of research results in the regulatory process. The 14 rep 9rts are grouped into three categories: (1) early scoping and background studies, including a survey of aged power plant facilities, operating experience reviews of Licensee Event Reports (LERs) to identify aging trends, workshops to obtain experts' opinions, and aging / risk considerations; (2) reports on developing a methodology for aging analysis and on evaluation and use of a signature analysid technique (MOVATS); and (3) Phase I results of aging research on nine components, including electric motors, battery chargers / inverters, elec-trical cables, pressure transmitters, diesel generators, motor operated valves,. l check valves, auxiliary feedwater pumps, and snubbers. Each summary has four sections: Background, Summary, Results/ Findings, and Utilization of Research Results in the Regulatory Process. This report is considered a "living" document. That is, research results and summaries of additional selected reports may be added periodically, j Technical Integration Review Group for Aging and 1.ife Extension (TIRGALEX), " Plan for Integration of Aging and Life-Extension Activities," U.S. Nuclear Regula-tory Commission, May 1987. The Technical Integration Review Group for Aging ar;d Life Extension (TIRGALEX) was established to facilitate the planning and integration of NRC j activities related te reactor aging

d life extensicn.

The initial ot>jectives i of TIRGALEX were to identify technical safety and regulatory policy issues related to reactor aging and life extension and to develop a plan to integrate i NRC and external activities to resuive the issues. This report contains the j plan developed by TIRGALEX, which consists of the following main elements: l A summary and discussion of the major technical safety ni regulatory policy issues associated with reactor aging and life exter.sion. e An overview of ongoing programs and activities related to reactor aging and life extension, including both fvRC and external progJams and activities. i Recommendations for future NRC acticas to address reactor Lging and e life extension in a timely, efficient, and well-integrated manner. NUMBERED REPORTS SNL Technical Report A-3270-11-26-84, B. Miller, " Scoping Test on Containment i Purge and Vent Valve Seal Material," Brookhaven National Laboratory, December 1984. 1 Degradation of shaft seal material used in containment purge and vent butterfly valves may initiate valve seal leakage thus breaching containtner.t A scoping test was performed to gather information on the behavior of the seal l material (ethylene propylene) When exposed to severe accident conditiens (i.e.. l steam at 350 F/120 psig and 400 F/232 psig). Three separate test sequences were performed with the test assembly monitored for leakage. The results of these i tests revealed no seal leakage; however, shaf t-seal degradation was cvident. i i 4 1

_ _ _ _ _ - _ = _ _ i 1 l MAIN CITATIONS AND ABSTRACTS For two test sequences, the prescribed procedure was revised to include modified temperature profiles and seal-testing sequences. Removal and inspection of the valve seat following some test sequences revealed minor remolding of the seat material at the disc / body interface with no 4 ' deformities noted. Approximately one week later, cracks developed in the seat. The cracks were in an area that would be compressed by the retaining ring and in no instance affected the sealing integrity of the valve. j The results of the scoping test revealed no shaft-seal leakage. The seal degradation and cracking was visually evident in the compressed retaining por-g tion of the seat. However, the ro ult should not be construed as representing i the entire ethylene propylene family (elastomers prepared from ethylene and pro-pylene monomers). Varying the relationship of these monomers affects the char-atteristics of the elastometer and its ability to withstand environmental condi-tions. It should also be noted that all mechanisms by which rubber deteriorates with time tre attributable to environmental conditions. The Parker Seal Company states that it is environment, not age, that is significant to seal life, both in storage and in actual service. l BNL Technical Report A-3270-11-85, J. H. Taylor, M. Subudhi, J. Higgins, J. Curreri, M. Reich, F. Cifuentes, and T. Nehring, " Seismic Endurance Tests of Naturally Aged Small Electric Motors," Brookhaven National Laboratory, November 1985. Two naturally aged 10-HP electric motors were obtained from an older nuclear power plant that is ready for decommissioning. The motors were utilized to drive fan cooler units in an outside environment for 12 years. These motors were first tested for their dynamic characteristic:s. They both were subjected to seismic excitation with generic floor response spectra (GFRS) that encompass Safe Shut-down Earthquake (SSE) accelerations applicable to most nuclear plants in the United States. The tests showed that the first fundamental frequency is well above the rigid range of an earthquake frequency. Seismic testing was performed with a motor both unloaded and loaded by an attached hydraulic pump that served as a dynamometer. Significant operating parameters such as current, voltage, and temperature were monitored before, during, and after seismic loading, and i no noticeable differences were observed. Existing deficiencies in one of the motor bearings and in the stator winding were not affected or magnified by the seismic excitations. This report describes the test plan, includes details of the procedure, and presents findings of the seismic tests and operating / static tests on both motors. This testing was part of the NRC NPAR program, and its results are an inte-gral part of the Brookhaven National Laboratory's overall aging assessment of motors, which was published as NUREG/CR-4156. 5 E_____________

-= l ) l l I MAIN CITATIONS AND ABSTRACTS BNL Technical Report A-3270-12-85, M. M. Silver, R. Vasudevan, and M. Subudhi, " Pilot Assessment: Impact of Aging on the Seismic Performance of Selected Equipment Types," Brookhaven National Laboratory, December 1985. i The NRC has initiated a number of specific research programs in support of j the NPAR program, to better understand the impact of equipment aging on plant safety and to recommend realistic operating and maintenance procedures to improve plant availability and enhance safety. This pilot study was performed to investigate the feasibility'of using plant experience data to assess the relationship between equipment aging and seismic performance capacity. After a brief review of available information on plant experience at many l California sites for content and quality, data related to performance, mainte-l nance, and failure history were collected for a sample set of equipment types. 1 This pilot study selected the equipment types for investigation from the highest i priority group specified in a previous NPAR study. The equipment types studied j were electric motors, motor-operated valves, relays, circuit breakers, and ] motor control centers. I i The acquired equipment data consisted of installation date, chronological l listing of preventive and corrcctive maintenance activities, failed state and cause of failure,' earthquake data (i.e., free-field acceleration, Richter magni-tude, date), and equipment status before and after the earthquake. The pilot study was successful in demonstrating that experience data can be extracted and utilized to address the relationship between seismic perform-ance capacity and aging of' plant equipment. It is strongly recommended that future research be conducted to acquire experience data for other important equipment types and to investigate other California power plants. Such research will provide the maxinum amount of actual experience data to address the aging-seismic relationship 'n a practical manner. Lessons learned from a review of these data can oe used as input to develop practical maintenance and operating procedures to enhance safety and improve plant reliability. BNL Technical Report A-3270-3-86, A. C. Sugarman, M. W. Sheets, and M. Subudhi " Testing Program for the Monitoring of Degradation in a Continuous Dun 460 Volt Class "B",10-HP Electric Motor," Brookhaven National Laboratory, March 1986. J \\ This report presents an evaluation of potential maintenance techniques for monitoring age-related degradation in a continuous-duty 460-volt, Class B, 10-HP electric motor. The program follows up the analyses and recommendations outlined in the draft of NUREG/CR-4156, " Operating Experience and Aging-Seismic Assessment j of Electric Motors," by M. Subduhi et al. In this study, the following stresses i on dielectrics are evaluated: temperature, frequent starts, overload, and high voltage gradient. l l 1 In general, the motor tests are conducted by continuously reversing motor l direction for five hours, followed by a half hour with the motor running under no load in a single direction and a half hour with the motor turned off and sta-l tionary. During the half hour of running under no load, measurements of bearing j vibration and movement of stator end turns (measured with accelerometers epoxied i i i 6 l

!A j I, MAIN CITATIONS'AND ABSTRACTS j 1 to the end' turns) were made. Also, a number of insulation tests were conducted. To accelerate the degradation'of the test motor (including insulation, bearings, and lubrication), a plug reverse test was perf ormed. 'The results of the exploratory testing program revealed which insulation and bearing tests can best be used in utilities' procedures for preventive main-i tenance, corrective maintenance,'and surveillance for safety-related motors. -) This testing.is meant for motors rated for continuous use. A separate test plan will be required for intermittent-duty motors (e.g., valve actuator mc. tors);. such a plan should include typical valve actuator tests such as the open/close cycling test and the insulation tests discussed in Section 4.0 of the presently reported program. 1 BNL Technical Report A-3270-12-86, R. Fullwood, J. C. Higgins, M. Subudhi, and J. H. Taylor, " Aging and Life Extension Assessment Program (ALEAP) Systems Level Plan," Brookhaven National Laboratory, December 1986. l I This system level program plan for ALEAP presents and explains the BNL structured approach to assessing the effects of the aging of nuclear power plant components and systems on safe operation and the extension of plant operation beyond the originally planned plant life. It should be noted that this plan j is prepared in a generic f ashion and could be used by anyone for a system i assessment.' The plan discusses the criteria for prioritizing plant, system, and l component selection for analysis to determine the effects of aging. The use of j failure modes and effects analysis in conjunction with the results of natural and accelerated aging tests are discussed as means for understanding and pre-dicting the phenomena. The effects of aging on the failure rates of components are being determined principally from plant data with physical and phenomenolog-ical models used for interpolation of areas not included in the data base. These results will be integrated with a plant risk model to be used in addressing the question "how old is old enough." The NRC NPAR program has completed several component-level aging assessments that include the identification of dominant component f ailure modes based on plant operating experience. The studies provide recommendations for monitoring as well as mitigating age related component degradations. Utilizing results from the component-level studies and work performed by other NRC contractors for system-data assessment and system-level risk analysis, this program evaluates the impact of component f ailures on plant system perfor-mance. The study performs in-depth system-level failure-data reviews, evaluates current industry practices for system maintenance, testing, and operation and probabilistic risk assessment (PRA) techniques to understand and to predict the impact of aging on system availability. Recommendations for improving the system performance by means of degradation monitoring and timely preventive and corrective maintenance are addressed. This project integrates its products with the BNL programs for operational saf ety reliability research and performance indicators. 7

MAIN CITATIONS AND ABSTRACTS The first phase of this research effort concentrates on understanding various system designs from plant safety analysis reports, evaluating failure data from plant operating experience data bases, applying PRA analyses, assessing industry-wide surveillance and maintenance practices, and identifying system functional indicators that are used to monitor the rate of system degradation resulting from aging and service wear. The program separates failures on demand from time-dependent failures. It categorizes age-related failures separately from random and design-type failures. It produces results useful for the resolution of pertinent unresolved safety issues and for review and inspection of operating NPPs. The second phase, if authorized and performed, will provide recommendations for improving system performance through enhanced maintenance practices and reliability monitoring, which will be focused on the most risk-sensitive areas of a system, Recommendations are made for improve-ments in pertinent regulatory guides, industry standards, etc. This program plan delineates the goals and major tasks to be completed in each phase. The current version of the program plan is considered to be a draft and will be revised and updated as the first few system assessments are completed using this methodology. This will produce a final proven methodology that can be applied to the remaining systems. NUREG-1144, B. M. Morris and J. P. Vora, " Nuclear Plant Aging Research (NPAR) Program Plan," U.S.11uclear Regulatory Commission, July 1985. NUREG-1144, J. P. Vora, " Nuclear Plant Aging Research (NPAR) Program Plan," Rev.1, U.S. Nuclear Regulatory Commission, September 1987. The Nuclear Plant Aging Research (NPAR) program described in this plan is intended to resolve technical safety issues related to the aging degradation of electrical and mechanical components, safety and support systems, and civil engi-neering structures used in commercial nuclear power plants. The aging period of interest includes the period covered by the original operating license as well as the period of extended plant life that may be requested in utility applica-tions for license renewals. Emphasis has been placed on identifying and characterizing the mechanisms of material and component degradation during service and utilizing the research results in the regulatory process. The research includes evaluating methods of inspection, surveillance, condition monitoring, and maintenance as means of managing and mitigating aging effects that may affect safe plant operation. Specifically, the goals of the program are to: Identify and characterize aging mechanisms and ef f ects that could cause degradation of components, systems, and civil engineering structures and, if unchecked, impair plant safety. l Evaluate residual life of components, systems, and civil structures and identify methods of inspection, surveillance, and monitoring that will ensure timely detection of aging effects before loss of safety functions. Evaluate the effectiveness of storage, maintenance, repair, and replace-ment practices in mitigating the rate and extent of degradation caused by aging. I 8

l MAIN CITATIONS AND ABSTRACTS j NUREG/CP-0036, (Compilation by) B. E. Bader and L. A. Hanchey, " Proceedings of the .. Workshop on Nuclear Plant Aging," Sandia National Laboratories, SAND 82-2264C, November 1982. The objective of the workshop, hela August 4-5, 1982, in Bethesda, Maryland, was to facilitate an exchange of thoughts between the NRC and industry on time-l related degradation and its influence on reactor safety. The specific goals were l to define the problem, to discuss the state of knowledge on aging phenomena, and I to identify future activities necessary to understand the problem. l The need for a comprehensive program to identify the potential safety I problems associated with plant aging was stressed. It was suggested that the effects of time-related degradation on the safety of the complete reactor system 1 should be evaluated in terms of the risk to the public. One should consider multiple causes that have typically been associated with abnormal occurrences. l Since individual component failures create problems, time-related degradation will ultimately have to be addressed in terms of maintenance, monitoring, surveillance, etc., of components. A large number of phenomena that can cause failures were discussed; a f detailed list of parts / materials, including lubricants and other additives that i must be considered, was given; seemingly minor changes in the chemical constitu-ents of a material or in the manufacturing process can cause significant effects and changes in the system during operation (e.g., water chemistry effects). i Replacement parts were noted as a potential source of problems. The effects of storage on parts and the possibility that new parts may be different from the original ones were mentioned.

l f

There were extensive discussions on the limitations of accelerated-aging i tests. The use of naturally aged equipment for test purposes was suggested. Sacrificial replacement of equipment was identified as a source for naturally aged plant equipment. Maintenance and surveillance in plants and their relationship to time-related degradation were extensively discussed. NUREG/CR-2641, J. P. Drago, R. J. Borkowski, D. H. Pike, and F. F. Goldberg, "The In-Plant Reliability Data Base for Nuclear Power Plant Components: Data Collec-tion and Methodology Report," Dak Ridge National Laboratory, ORNL/TM-8271, July 1982. The development of a component reliability data base for use in nuclear power plant probabilistic risk assessments and reliability studies is presented. The data sources are the in plant maintenance work request records from a sample of nuclear power plants. This data base is called the In-Plant Reliability Data System (IPRDS). Its features are compared with other data sources such as the Licensee Event Report (LER) system, the Nuclear Plant Reliability Data (NPRD) system, and IEEE Standard 500. Generic descriptions of nuclear power plant systems formulated for IPRDS are outlined in the text. 9 ~

{ MAIN CITATIONS AND ABSTRACTS The major objective of the program described is to provide an improved multipurpose data base. Components of each type of NSSS are included in the data base. In addition to providing information on past failure rates and component down times, the IPRDS may be used for: revising component test intervals and allowable down times; identifying generic problems and recurrin9 failures; identifying the variables (e.g., environment, operating mode, system, maintenance policy, etc.) that control component failure rates; providing an extensive data base against which to compare existing data sources (e.g., LERs and NPRDs) to assess the degree to which these data sources accurately reflect the actual component reliability; correlating current incidents wit', previous failures, allowing for extrapolation in the near future; identifying trends and patterns in the failure characteristics of e particular components or aggregations of components; and identifying failure mechanisms over time for use in defining the e aging requirements for component qualification. NUREG/CR-3154, R. J. Dorkowski, W. K. Kahl, T. L. Hebble, J. R. Fragola, end J. W. Johnson, "The In-Plant Reliability Data Base for Nuclear Plant Components: Interim Report--The Valve Component," Oak Ridge National Laboratory, DRNL/TM-8647 December 1983. This document details the collection and preliminary analyses of data related to valves in the In-Plant Reliability Data System (IPRDS). The date base is developed primarily from historical records of corrective maintenance actions obtained directly from nuclear plant maintenance files. A comprehensive valve populatior, is also included. This report presents data from one PWR and one BWR power plant. The report demonstrates the degree of distinction and refinement in the reliability statistics that is possible with data from the IPRDS and suggests a general format for disclosure of suitable reliability statistics to satisfy needs within the nuclear data gathering community. The examples given in the various tables and figures suggest a useful method of comparing valve data and are repre-sentative of the degree to which reliability statistics for any particular valve can be ascertained. One objective of this report is to examine the improvement possible using IPRDS in refining the statistics to ultimately focus on the reliability of par-ticular valve types and valve operators in specific working environments. Another objective is to generate comments from members of the nuclear data 3D

l MAIN CITATIONS AND ABSTRACTS community as to the efficacy'of the suggested formats for documenting valve information and the various methods used for comparison in this report. The report gives breakdowns of failure rates by failure modes and by failure causes showing calculated maintenance frequencies and repair times. IPRDS repair time distributions, although unavailable f rom LERs, are also presented and evaluated. Preliminary results orWined from the pilot data pase in this report . indicate WASH-1400 statiszia to be nonconservative in reliability estimates for some valve types in cert.in failure modes. NUREG/CR-3543, G. A. Murphy, R. B. Gallaher, M. L. Casada, and H. C. Hoy, " Survey of Operating Experienc s f roa LERs to Identify Aging Trends," Dak Ridge National Laboratory, ORNL-NSIC-216, January 1984. This report describes the pre 19r.inary results of an assessment of informa-tion pertinent to identifying age-related failures available in operating exper-ience reports. This assessment, by the Nuclear Operations Analysis Center (NOAC) at Oak Ridge National Laboratory, utilized the computerized files of Licensee Event Reports (LERs) and their predecessors to examine age-related degradation of safety-related equipment. Abstracts of operating experience reports from commercial power plants reported from 1969 to 1982 were surveyed. Over 7000 events were reviewed. Data included the system, component, subpart, the age-related failure mechanism, the severity, and the method of detection of the failure. Wear, corrosion, crud, and fatigue were the identified f ailure mechanisms in over one-third of the 3098 age-related events. About two-thirds of the failure severities were judged as a degraded state, and one-third were. judged as catastrophic failures. temp and valve problems made up almost 30% of the Tailed components. Almost two-thirds of the reported failurcs were detected by routine surveillance testing indicating that such practices are effective techniques for monitoring and detecting age degradation of discrete components and systems. A substantial number of events resulted from setpoint drift. NUREG/CR-3818, N. H. Clark and D. L. Berry, " Report of Results of Nuclear Power Plant Aging Workshop," Sandia National Laboratories, SAND 84-0374, August 1984. The objective of the workshops was to identify whether there is any evidence of component or structural time related degradation, i.e., aging problems, in a nuclear power plant and, if so, what problems are of greatest importance. Fif teen representatives frorn national laboratories, architect / i engineers, nuclear steam supply system vendors, research firms, and one un ver-sity participated. Questionnaires and group discussions screened over 112 components believed to be susceptible to excessive aging; pressure cnd temperature sensors, valve operators, and snubbers emerged by consensus as the most important. Potential aging problems relate to off-normal common-mode effects or problems that were just developing at the time were outside the scope of the workshops because little or no first-hand experience was available for these off-normal or yet-to-be-explored circumstances. Recommendations are made for a systematic approach to rating components in terms nf overall saf ety and 11

MAIN CITATIONS AND ABSTRACTS j 1 i for a cooperative effort between industry research groups and regulatory i' research groups to resolve known aging problems and to identify off-normal'or yet-to-develop aging issues. In addition to some well-known aging mechanisms (e.g., neutron embrittlement of pressure vessels) or problems that manifest themselves as equipment failures (e.g., steam generator tube degradation), there is concern that other types of bging problems may be developing. Their I effects increase as nuclear power plants get older, and some aging processes j could eventually affect power plant availability or safety. J 8 NUREG/CR-3819, J. A. Rose, R. Steele, Jr., K. G. DeWall, and B. C. Cornwell, " Survey of Aged Power Plant Faci,ities," Idaho National Engineering Laboratory, EGG-2317, June 1985. i, 1 The survey concentrated on component failures in LWR safety related systems l as determined from operating histories. Only failuren that we u determined to i be age related were included. l The age-related failure information gathered from the plant histories was aneilyzed for reoccurring failure patterns. Early program emphasis was on isolating specific equipment with high failure rates that were not already the 4 concern of_other research efforts. The resulting (gathered) data could not support the identification of specific equipment. It did, however, imply a direct relationship between the failure and the failure mechanism. Thus the emphasis of the program was redirected toward es d oring the relationship of the failure to the failure mechanism. The results of this preliminary investigation indicated that about 70% of the significant failures reported for the fluid systems analyzed were due to only four failure mechanisms (causes): erosion, corrosion, vibration, and for-eign materials. This was subsequently verified by detailed study of several more plant systems and corroborated by f k.ld data obtained frem perscnnel interviews. In addition, there appears to be a strong correlation between the cause of com-ponent failure and the system in which the cor.ponent operates. The survey poirts out, with evidence, that tr.e identification and elimina-tion of the sytte6-level ca &rs cf component failuren is a viable approach to preventing and mitigatino the majer reported aging effects. WEG/Dh3956, M. R. Dinsel, M. R. Donaldson, ana F. T. Soberano, "In Situ Testing i of the Shippingport Atomic Power Station Electrical Circuits," Idaho National Engineering Laboratory, EGG-2443, April 1987. This report discusses the results of electrical in situ testing of selected circuits and components at the Shippingport Atomic Power Station in Shippingport, Pennsylvania. The goal was to determine the extent of aging or degradation of l-various circuits from the original plant and the two major w re plant upgrades (representing a total of three distinct age groups) as wel? 6s to evaluate pre-viously developed surveillance technology. The electrical testing was performed l using the Electrical Circuit Cha acterization and Diagnostic (ECCAD) system developed by EG&G for the U.S. Department of Energy to use at THI-2. Testing included measurements of voltage, effective seriet capacitance, effective series 12 l

MAIN CITATIONS AND ABSTRACTS inductance, impedance, effective s6cies resistance, dc resistance, insulation resistance, and time-domain reflectometry ( MR) parameters. The circuits eval-uated included pressurizer heaters, c00 trol rod position indicator cables, mis-cellaneous primary system resistance temperature detectors (RTDs), nuclear instrumentation cables, and safety injection system motor-operated vahes. It should be noted that the operability of these circuits was tested seve al years after plant operation was concluded at Shippingport. 7here was no neet to retain the circuits in working condition f6110 wing nlant shutdown, so no effort was expended for that purpose. The in situ measurements and analysis of the data confirmed the effectiveness of the ECCAD syste for detecting degradation of circuit connections and splices a'long the high-resistarm paths; ruost of the problems were caused by corrosioa. Results indicate a correlation between the chronological age of circuits and circuit degradation. NUREG/CR-4144, T Davis, A. Shafaghi, R. Kurth, and F. Leverenz, "Importance Ranking Based on Aging Consideration of Camponents Included in Probabilistic Risk Assessments," Pacific Northwest Laboratories, PNL-S389, April 1985. The method outlined in the report ranks power plant components by using a risk-due-to-aging sensitivity measure that describes the change in risk due to changes in component failure rate (without describing closely the aging pher.omena and the resulting time-dependent component failure rate). The output from this study can be combined with that from other studies (data, analytical or experimental) that identify the components most susceptible to aging. The applications use average component unavailability equations currently employed in probabilistic risk assessment (PRA) to calculate the risk-due-to-aging sensitivity. A more exact calculation is possible by using unava'11bility equations that include the time-dependent characteristics of component f ailure rates; however, these ti:ne-dependent characteristics are not well known. The risk-due-to-6ging sensitivity measure presented here is therefore Mgregated from these time-dependent ef fects and addresses unly the time-independent per-tion of aging phenomena. The results identify the components that show the highest potential for risk-due-to-aging phenomena. l Three operating SSSS wre analyzed, and it was found that the mst risk-significant components are in the auxiliary feednter system, the reactor i protection system, and the service *mter systems, e.g., pomps, check valves, motor-operated valves, circuit breakers, and actuating circuits. Future research on the time-dependent portion of aging phenomena for these components is needed to completely describe'the impact on risk. NUREG/CR-4156, M. Subudhi, E. L. Burns, end J. H. Taylor, " Operating Experience and Aging-Seismic Assessment of Electric Motors," Brookhaven National Labora-tory, BNL-NUREC-51861, June 1985. A limited number of electric motor categories with direct safety signifi-cance were identified, and failures due to insulation degradation, were surveyed. 1 13 i I

MAIN CITATIONS AND ABSTRACTS Age-sensitive components (with respect to materials and design features) were reviewed, potential electrical and mechanical hazards were considered, operational and accident stressors were determined, and monitorable functional indicators were identif.ied. The contribution of pertinent seismic effects was assessed, and failure modes, mechanisms, and causes were reviewed from existing data bases. NUREG/CR-4234, 'W. L Greenstreet, G. A. Murphy, and D. M. Eissenberg, " Aging and Service Wear of Electric Motor-Operated Valvec Used in Engineered Safety-Feature Systems of Naclear Power Plants," Vol.1, Oak Ridge National Laboratory, ORNL-6170/VI, June 1985. This report deals with motor-operated valves, focrsing on monitoring defects and degradation of nuclear plant safety equipment. The contents include the evaluation and identification of practical and cost effective methods for detecting, monitoriag,- and assessing the severity, failure modes, and causes (mainly aging and service wear) of time-dependent degradation in nuclear plants. .Also being considered are manufacturer-recommended maintenance and surveillance practices and the selection of meas'urable parameters (including functional indi-cators) for use in assessin 3 operational readiness, establishing degradation trends, and detecting incipient failures. The report's results are based on information derived from operating experience records, nuclear industry. reports, manufacturer-supplied i' formation, and input from architect-engineer firms and n plant operators. Failure modes are identified for both the valve and the motor-operator assembly.. For each failure mode, failure causes are listed by subcomponent or subassembly, and parameters potentially useful for detecting degradation that could lead to failure are identified. The method emerging from this analysis of the d3ta can provide capabilities for establishing degradation trends prior to failure and developing guidance for effective and safe maintenance. NUREG/CR-4257, 5. Ahmed, A. Carfagno, and G. J. Toman, " Inspection, Surveillance, ~ and Monitoring of Electrical Equipment Inside Containment of Nuclear Power l l Plants With Applications to Electrical Cables," Dak Ridge National Laboratory, ORNL/508/83-28915/1, August 198$. L The purpose of this report is to describe currently available methodology for detecting and determining the amount and rate of age-related deterioration of safety-related equipment. The general concepts of monitoring equipmerit con-dition for this purpose are described. The goal is to detect deterioration in the incipient stage, prior to inservice failure and prior to the point at which equipment can no longer be expected to perform its function when exposed to design basis accident conditions. The application of condition monitoring is discussed specifically for electric cables. The goal is to determine the degree of cable degradation-and to predict the remaining useful life. In situ nondestructive testing and destruc-tive laboratory testing are discussed as are their limitation.i. Interim recom-mendations are given for the implementation of a condition-monitoring program for cables. 14 i

L l HAIN CITATIONS AND ABSTRACTS NUREG/CR"4257 G. J. Toman, ":nspection, Surveillance, and Monitoring of Electrical S Equipment in Nuclear Power Plants. Vol. 2: Pressure Transmitters," Oak Ridge National Laborat.ory, ORNL/SUB/83-28915/3/V2, August 1986. This report describes the types of pressure transmitters commonly used in nuclear power plants according to their application. The stresses that affect these transmitters include ambient temperature, humidity, radiation. process (fluid) medium pressure, and temper.ature. The most common effects of the stresses on the transmitters are calibr3 tion shifts. The evaluation of failure data contained in Licensee Event Reports indicates that total failure of pres-sure transmitters occurs relatively infrequently. Comparison of avfound and as-lef t calibration data is described as a partial means of evaluating the level of deterioration of a transmitter. Care i must be taken to ensure that variations in method or procedure do not produce erroneous data and wrong conclusions. The precision of the comparative measure-ments must also be high. The evaluation of calibration data alone will not ensure the capability of operating under design basis accident conditi0ns. If, with time, steam or mois-ture penetrates the transmitter housing, the transmitter electronics will become inaccurate and may fail. Therefore, the integrity of the housinn seal must also be evaluated periodically to be able to predict ccatinued perfore Mce capability. Evaluation of inservice failures is recommended to allow further differen-tiation between sudden failures (having no precursor) and failures that can be detected in the incipient state. Such evaluations would aid in the further development of monitoring techr.iques. Because some of the transmitter failures are of the sudden type, periodic operability checks are an important means of detecting failures very soon after their occurrence so that a significant number of failed (inactive or inaccurate) transmitters do not remain undetected. A combination of operability monitoring and condition monitoring may be used to improve the probability of successfully weathering aging processes and accident conditions. NUREG/CR-4279, S. H. Bush, P. G. Heasier, and R. E. Dodge, " Aging and Service Wear of Hydraulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nucitar Power Plants " Vol.1, Pacific Northwest Laboratories, s PNL-5C9, February 1986. This report presents an overview of hydraulic ed mechanical snubbers used on nuclear pip'ng systems and components. The functions and functional require ments of snubbers are outlined. The real versus perceived need for snubbers is reviewed based primarily on studies conducted by a Pressure Vessel Research Committee. Tests conducted to qualify snubbers, to accept them on a case-by-case basis, and to establish their fitness for continued operation are reviewed. This report had two primary purposes: (1) to assess the effects of various aging mechanisms on hydraulic and mechanical snubber operation (e.g., leaking of seals, functional failures) and (2) to determine the efficacy of existing tests in determining the effects of aging and degradation mechanisms. These tests 15

o MAIN CITATIONS AND ABSTRACTS m. include breakaway force, drag force, velocity / acceleration range for activation I in tension or compression, release rates within specified tension / compression l limits, and restricted thermal movement. The snubber operating experience was reviewed using licensee event reports and other historical data for a period of more than 10 years. Data were statistically analyzed using arbitrary snubber l populations. Value-impact was considered in terms of exposure to a radioactive environment for examination / testing and in terms of the influence of lost snubber function and subsequent testing program expansion on the costs and operation of a nuclear power plant. The implications of the observed trends were atsessed; recommendations include modifying or improving the examination and testing proce-dures to enhance snubber reliability. Optimization of snubber populations by selective removal of unnecessary snubbers was also considered. NUREG/CR-4302, W. L. Greenstreet, G. A. Murphy, R. B. Gallaher, and D. M. Eissenberg, " Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol. 1, Oak Ridge National Laboratory, ORNL-6193/V1, December 1985. The report addresses detecting defects and monitoring the degradation of nuclear plant safety equipment. The program is concerned with identifying and evaluating practical and cost-effective methods for detecting,-monitoring, and assessing the severity of time-dependent degradation (aging and service wear) of check valves in nuclear plants. These methods will allow degradation trends to be detected prior to failure and allow guidance for effective maintenance to be developed. The topics considered are failure modes and causes resulting from aging and service wear, manufacturer-recommended maintenance and surveillance practices, and measurable parameters (including functional indicators) for use in assessing operational readiness, establishing degradation trends, and detecting incipient failure. The results presented are based on information derived from operating experience records, nuclear industry reports, uanuf acturer-supplied information, and plant operators. Failure modes for check valve. are identified and are examined by iden-tifying methods for detecting failures and differentiating between their causes. For each failure mode, failure causes are listed by component or subassembly, and parameters potentially useful for detecting degradation that could lead to failure are tabulated. The report also identifies parameters potentially useful for enhancing the detaction of degradatio.n and incipient f ailure; these parameters include dimen-sions, bolt torque, noise, appearanch roughness, and cracking. NUREG/CR-43M, J. L. Crowley and D. M. Eissenberg, " Evaluation of the Motor-Operated Valve Analysis and Test System (MOVA15) to Detect Degradation, Incorrect Adjustments, and Othe Abnormalities in Motor-Operated '!alves," Gak Ridge National Laboratory, ORNL-6226, January 1986. An important aspect of the NPAR program strategy is to demonstrate the utility of condition monitoring, signature analysis, and other surveillance 16

MAIN CITATIONS AND ABSTRACTS methods for detecting, differentiating, and trending various types of abnorma-lities in the components so that corrective measures can be implemented prior to loss of safety function. A field test program was carried out to evaluate valve signature analysis as a surveillance method to achieve these results as well as to detect incorrect adjustments in motor operated valves. The technique specified in the title (MOVATS) is the subject of this report. In situ signa-ture traces were obtained in 36 motor-operated valves at four nuclear plant i sites. Described are the test equipment package, the method of obtaining the signatures, and determinations made as a result of analyzing them. Based on the results of the signature-analysis technique and those obtained from the 4 field-test program, the capabilities and limitations of MOVATS are discussed. NUREG/CR-4457, J. L..Edson and J. E. Hardin, " Aging of Class 1E Batteries in j Safety Systems of Nuclear Power Plants," Idaho National Engineering Laboratory. j EGG-2488, July 1987. l 1 This report presents the results of a study of aging effects on safety-f related batteries in nuclear power plants. The purpose is to evaluate the aging i effects caused by battery operation in a nuclear facility and to evaluate main-tenarce, testing, cnd monP " ring practices with respect to the effectiveness of these practices in detectii.g and mitigating the effects of aging. The study follows the NRC NPAR approach and investigates the materials used in battery con-struction. It also identifies stressors and aging mechanisms, presents operating and testing experience related to aging effects, analyzes battery-failure event reports in var.ous data bases, and evaluates reccomended maintenance practices. Data bases that were analyzed included the NRC's Licens-ee Event Report system, the Institute for Nuclear Power Operations' Nucleer Plant fieliebility Data System, the Oak Ridge Natio.nal Laboratory's In-Plant Reliability Data System, and the S. M. Stoller Corporation's Nuclear Power Experience data base. NUREG/CR-4564, W. E. Gunther, M. Subudhi, and J. H. Taylor, " Operating Experience l and Aging-Seismic Assessment of Battery Chargers and Inverters," Brookhaten National Laboratory, BNL-NUREG-51971, June 19G6. Battery chargers and inverters are vital components of the nuclear power plant electrical safety system. The objectives of this program are to (1) iden-l L tify concerns related to the aging und service wear of equipment operating in l nuclear power plants, (2) assess their possible imoact on plant safety, (3) identify effective inspection, surveillance, and monitoring methods, and (4) recommend suitchie maintenance practices to mitigate aging-related concerns and diminish the rate of degradatina due to aging and service wear. The designs of battery chargers (3 types) and inverters (4 types) and the materials for their construction are reviewed to ideni.ify age-sensitive compon-ents. Operational and accidental stressors are determined, and their effect on promoting aging degradation are assessed. Variations in plant electrical designs, as well as system and component impacts v:ere studied. Failure modes, mechanisms, and causes were reviewed frem operating experience and existing data banks. The study also considered seismic effects on age-degraded components of battery chargers and inverters. The performance indicators that can be monitored te assess component dete-rioration due to aging or other relevant stressors are identified. Conforming 17

> i MAIN CITATIONS AND ABSTRACTS with the NPAR strategy as outlined in the program plan, the study also includes a review of current standards and guides, maintenance programs, and research activities pertaining to safety-related battery chargers and inverters for nuclear power plants. NUREG/CR-4590, K. R. Hoopingarner, J. W. Vauso, D. A. Dingee, and J. F. Nesbitt, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating and Expert-Experience," Vols. I and 2, Pacific. Northwest Laboratories, PNL-5832, August 1987. Pacific Northwest Laboratory eveluated operational and expert experience pertaining to the aging degradation of diesel generators in nuclear plant service. The research identified and characterized the contribution of aging to emergency diesel generator failures. Volume 1 reviews diesel generator experience to identify the systems and components most subject to aging degradation and isolates the major causes of failure that may affect future operational readiness. Evaluations show that, as plants age, the percentage of aging-related failures increases and failure modes . change. A compilation is presented of. recommended corrective actions for the aging related failures identified, and the trend of these failures is discussed. This study also includes a review of current reievant industry programs, research, and stanidards, Volume 1 presents the results of the Phase I research that identifies the components and systems most susceptible to aging degradation and the major causes of such degradation. Volume 2 reports the iesults of a workshop held on May 28 and 29,1986, with industry representatives to discuss the technical issues associated with aging of nuclear service emergency diesel generators. The technical issues dis-cussed most extensively were man / machine interfices, component interfaces, ther-mal gradients of startup and cooldown, and the need for an accurate industry data base for trend analysis of the diesel generator system. NOREG/CR-4597, M. L. Adams and E. Makay, " Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants. Vol. 1: Operating Experience and Failure Identification," Dak Ridge National L'aboratory, ORNL-6282/V1, July 1986. In this report, typical auxiliary feedwater pump features are described in terms of configuration details, materials of construction, cperating require-ments, and modes of operati,on. Failure modes and causes due to aging and set-vice wear are identified and explained, and measurable parameters (f acluding functional indicators) for potential use in assessing operationa? readiness, establishing degradation trends, and detecting incipient failures are outlined. A series of measures to correct present deficiencies in surveillance, monitoring, and inservice testing practices is discussed. The main body of the report is tuppiemented by a number of relevant appendices; in particular, a major appendix is included on engineering and design information useful t0 assess operational readiness. 18 __._._m _mm

l .:4 j [ MAIN CITATIONS AND ABSTRACTS i I' l

NUREG/CR-4597, D. M. Kitch, J. S. Schlonski, P. J. Sowatskey, and W. V. Cesarski, j

" Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants. Vol. 2: Aging Assessments and Monitoring Method Evaluations," Dak l Ridge Aational Laboratory, ORNL-6282/V2, June 1988. The subjects specified in the title are described and discussed in four I majorsections: l 1. Failure causes, 2. Description of inspection, surveil?ance, and condition monitoring l (ISCM) methods, 3. Evaluation of ISCM methods, and 4. Role of maintenance in alleviating aging and service wear. Failure causes attributable to aging and service wear are given and ranked in terms of importance. Cause identifications are made on the basis of exper-ience, postservice examinations, and in situ assessments. Measurable parameters related to failure causes are identified. ISCM methods are specified, and evaluations are made based on Westinghouse experience. On the same basis, recommendations are given on inspection, surveillance, and condition monitcring. The ISCM methods are intended to yield required capabil-ities for establishing operational readiness as well as for detecting and track-ing degradation and its trends. The role of maintenance in alleviating and mitigating aging and service wear effects is discussed, and the relationship of maintenance to ISCM methods is identified. Predictive, preventive, and corrective maintenance practices are discussed and evaluated. Appendices contain a detailed discussion on ISCM methods, failure data base information, auxiliary feedwater pump (AUXFP) installation lists (location sur-vey), a discussion of low-flow testing, auxiliary feedwater system descriptions (with flow-diagrams and schemss), AUXFP minimum-flow-rate criteria, and guide-lines proposed by Westinghouse for full-fiow testing. Note: The draft of this Vol. 2 (with the same title) was issued by WestinghouseTTectric Corporation, Generation Technology Systems Division, in April 1986, coauthored by D. M. Kitch, M. Vuckovich, W. V. Cesarski, and P. J. Sowatskey. NUREG/CR-4652, D. J. Naus, " Concrete Component Aging and Its Significance Relative to Life Extension of Nuclear Power Plants," Dak Ridge National Laboratory, ORNL/TM-10059, September 1986. The objectives of this study are to (1) expand upon the work that was initiated in the first two Electric Power Research Institute studies relative to longevity and life extension considerations of safety-related concrete com-ponents in light-water reactor (LWR) facilities and (2) provide background that will logically lead to subsequent development of a methodology for assessing and predicting the effects of aging on the performance of concrete-based mate-rials and components. Applications of safety related concrete components to LWR technology are identified, and pertinent structures (containment buildings, containment base 19

H l MAIN CITATIONS AND ABSTRACTS 1 mats, biological shield walls, main building, and saxiliary buildings) and the materials of which they are constructed (concrete, mild steel reinforcement, pre-stressing systems, embedments, and anchorages) are oescribed. - Historical per-formance of concrete compor.ents was established through information presented on concrete longevity and component behavior in both LWR and high-temperature gas-cooled reactor applications. Also, a review of problems involving concrete components in both general civil engineering and nuclear power applications is given. The majority of the problema identified in conjunction with nuclear power applications were minor; they include concrete cracking, concrete voids, or low concrete strengths at an early age. Five incidents involving LWR con-crete containments that are considered significant are described in detail. Potential environmental stressors and aging factors to which LWR safety-related components could be subjected are identified and discussed in terms of durability factors related to the materials used to fabricate the components. The current technology for detecting concrete aning phenomena it; also presented in terms of methods applicable to the particular material system that could experience deteriorating effects. Remedial measures for the repair or replacement of degrade 9 concrete components and their effectiveness are discussed. Finally, considerations relative to developing a damage methodology for assessing the durability factor, de a rioration rates, and prediction of structural reliability are outlined. Conclusions and recommendations of the report note the need for (1) obtain- 'ig aging data from decommissioned plants, (2) using inservice inspection pro-grams to provide aging trends, (3) developing a methodology to quantitatively and uniformly (i.e., using the same procedures) assess structurai reliability as affected by aging or degradation of structural materials, and (4) performing l research in support of all these needs. It should be stressed that there is no l widely accepted standardized methodology for gaantifying the condition and l capacity of an individual concrete structure. NUREG/CR-4692, G. A. Murphy and J. W. Cletcher II, " Operating Experience Review of Failures of Power Operated Relief Valves and Block Valves in Nuclear Power i Plants," Dak Ridge National Laboratory, ORNL/N0AC-233, October 1987. j This report contains a review sf nuclear power plant operating events involvin; failures of power-operated relief vsives (PORVs) and associated block valves (BVs). Of the 230 events identified, 101 involved PORV mechanical failure, 91 were attributable to PORV control failure, 6 involved design or fabrication of the PORVs, and 32 invoh ed BV failures. The report contains a compilation of the PORV and BV failure events, includirg failure cause and i severity. The events are identified as to plant and valve manufteturer. An assessment of the need to upgrade PORVs at:d BVs to safety grade status con-l cludes that such action would improve PORV and BV reliao'ility. lhe greatest improvement in reliability would resdft from using newer, more reliable PORV designs and improving testing, diagnostics, and maintenance applied to PORVs and SVs, particularly to the BV moter operators. A sumrury of interviews con-ducted with four PORV manufacturers is also included in the report. 20 b____________________._____

1 MAIN CITATIONS AND ABSTRACTS NUREG/CR-4715, G. J. Toman, V. P. Bacanskas, T. A. Shook, and C. C. Lodlow, "An.Aqiig Assessment of Relay and Circuit Breakers and System Interactions," BrooLaann National Laboratory, Franklin Research Center, Philadelphia, PA, 3 BNL-NUREG-52017, June 1987. i As part of the NRC NPAR program, Franklin Research Center analyzed the effects of aginn on safety related circuit breakers and relays un br contract to Brookhaven Fac ional Laboratory. Circuit breakers and relays in a PWR safety injection system were evaluated with respect to the aging caused 5y system opera-tion. The effect of circuit breaker and relay deterioration on the ability of the system to perform its safety functions was also evaluated. -The study included protective, cbntrol, and logic relays, as well as molded-case and metal-clad switchgear circuit breakers. Analysis of nuclear power plant failure data confirmed that normally energized relays commonly used in safety systems suffer from more rapid deterioration than do deenergized relays. The failures were attributable to coil deterioration, changes in dimensions of critical organic components, and changes in characteristics of timing diaphragms from thermal deterioration. Some of the failure modes will prevent fail-safe Opera-tien. The electrical control and mechanical portions of metal-clad switchgear were found to be more failure prone than the main coriacts and arc extinguish-ing systems. Analysis cf failure data for circuit breakers and relays indicated a general trend of increasing failure rates in the period of 6 to 11 years following the start of commercial operation of the plants. The aging interaction study evaluated the interaction of aged relays and circuit breakers in a safety injection system with r7 gard to five events requiring the system to start operation. Failure of redundant trains f rort common-mode failure of a particular type of circuit breaker or relay is not expected. However, the number of different types of potential failures supports the need for a strong maintenance and surveillance program to prevent multiple age-related failures from affecting redundant safety trains. NUREG/CR-4731, V. N. Shah and P. E. MacDonald, " Residual Life Assessment of Major Light Water Reactor Components," Vol.1, Idaho National Engineering Laboratory, EGG-2469, June 1987. NUREG/CR-4731, V. N. Shah and P.E. MacDonald, " Residual Life Assessment of Major Light Water Reactor Components--0verview," Vol. 2 (Draft), Idaho National Engineering Laboratory, EGG-2469. March 1988. This report presents an assessment of the aging (time-dependent degradation) of selected major light water reactor coaponents and structures-The stressors, possible degradation sites and mechanisms, potential failure modes and currently usad nondestructive examinations, inservice inspection (ISI), and life assessment methods are discussed for major light water reactor components. Volume 1 covers PWR and BWR pressure vessels, FWR containment structures, PWR reactor coolant piping, PWR deam generators, BWR recirculation piping, and reactor pressure vessel stpports. Volume 2 covers PWR reactor coolant pumps, PWR pressurizer, PWR pressurizer surge and spray lines, PWR reactor coolant system charging and safety injection nozzles, PWR feedwater lines, PWR control rod drive mechanisms aad reactor internals, BWR containments, BWR feedwater and main steam lines, BWR control rod drive mechanisms and reactor internals, PWR and BWR electrical cables 21

MAIN CITATIONS AND ABSTRACTS and connections, and PWR and BWR emergency diesel generators. Unresolved techni-csl issues related to understanding and managing the aging of these major compon-ents, including requirements for advanced ISI and life assessment methods, are also discussed. NUREG/CR-4740, L. C. Mey r, " Nuclear Plant-Aging Research on Reactor Protection Systems," Idaho National Engineering Laboratory, EGG-2467, January 1988. This report presents the results of a review of operating experience for the reactor eip system (RTS) and the engineered safety feature actuating system (ESFAS) reported in Licensee Event Reports (LERs), the Nuclear Power Experience data base, the Nuclear Plant Reliability Data System, and plant maintenance records. The purpose of the review was to evaluate the potential significance of aging, including cycling, trips, and testing, as a contributor to degradation of the RTS and ESFAS. Tables show the percentage of events for RTS and ESFAS classified by cause, componaats, and subcompanents for each of the nuclear steam supply system vendors. A representative Babcock and Wilcox plant was selected for detailed study. The NRC NPAR guidelines were followed in performing the detailed study that identified materials susceptible to aging, stressors, environmental factors, and failure snodes for the RTS and ESFAS and the relevant generic instrumentation and control systems. Functional inJicators of degradation are listed, testing requirements evaluated, and regulatory issues discussed. NUREG/CR-4747, B. M.'Mesle and D. G. Satterwhite, "An Aging Failure Survey of { Light Water Reactor Safety Systems and Components," Vol. 1, Idaho National Engineering Laboratory, EGG-2473, July 1987. NUREG/CR-4747, B. M. Meale and D. G. Satterwhite, "An Aging failure Survey of { Light Water Reactor Safety Systems and Components," Vol. 2, Idaho National Engineering Laboratory, EEG-2473, July 1988. This report describes the methods, analyses, results, and conclusions of two different aging studies. The first study was a survey of light water reac-tor component failures associated with 15 selected safety and support systems. Analysts used computerized sorting techniques to classify component failures l into generic failure categories. The.second study was a careful examination of component failure records to identify and categorize the reported causes of com-ponent fa; lures. The systems evalucted in the failure-cause analysis were the auxiliary feedwater, Class 1E electric power distribution, high pressure injec-tion, and service water. Tables and figures indicate the systems and the compon-ents within the systems that are most affected by aging. Engineering insights drawn from the data are provided. Volune 2 presents all of the Volume I data from.FY-86 combined with the data gathered in FY-87. i NUREG/CR-4769, W. E. Vesely, " Risk Evaluations of Aging Phenomena: The Linear Aging Reliability Model and Its Extensions," Idaho National Engir.eering Laboratory, EGG-2746, April 1987. A model for failure rates of light w&ter reactor safety system components due to aging mechanisms has been developed from basic phen-omenelogical consid-erations. In the treatment, the occurrences of deterioration are modeled as following a Poisson probability process. The severity of damage is allowed to j 22 1

MAIN CITATIONS AND ABSTRACTS have any distribution; however, the damage is assumed to accumulate indcpen-dently. Finally, the failure rate is modeled as being proportional to the accumulated damage. Using this treatment, the linear aging-fail rre-rate model is obtained. The applicability of the linear aging model to various mechanisms is discussed. The taodel is also extended to cover nonlinear and dependent aging phenomena. The implementation of the linear aging model is deinonstrated by. applying it to the aging data collected in the NRC NPAR program. HUREG/CR-4G19, V. P. Bacanskas, G. C. Rc6berts, and G. J. Toman, " Aging and Service Wear of Solenoid-Operated Valves Used in Safety Systems of Nuclear Power Plants. Vol. 1: Operating Experience and Failure Identification," Dak Ridge Natior.ai j Laboratory, ORNL/ SUS /83-28915/4/V1, March 1987. 1 An assessment of the types and uses of solenoid-operated vaives (SOVs) in l nuclear power plant safety-related service is provided in the report. Through a description of the operation of each 50V combined with knowledge of nuclear l power plant applications and operational occurrences, the significant stressors responsible for degradation of S0V performance are identified. A review of actual operating experience (including failure data) leads to the identification i i of potential nondestructive in situ testing which, if properly developed, cnuld l provide the methodology for monitoring the degradation of SOVs. Recommendations are cutlined for continuing the study into the test methodology development phase. NUREG/CR-4928, H. M.11asemian, K. M. Petersen, T. W. 'Kerlin, R. L. Anderson and K. E. Holbert, " Degradation of Nuclear Plant Temperature Sensors " Analysis s and Measurement Services Corporation, Knoxville, TN, June 1987. A program was established and initial testr were performed to evaluate long-term performance of resistance temperature detectors (RTDs) of the type used in U.S. nuclear power plants. This report addresses the effect of aging on RTD calibration accuracy and response time. The Phase I effort (lasting j about 6 months) included exposure of 13 safety grade RTD elements to simulated LWR temperature regimes. Full calibrations were performed initially and monthly, sensors were monitored and cross-checked continuously during exposure, and response time tests were performed befcre and after exposure. Short-term calibration drifts of as much as 1.8 F (1 C) were observed. Another result was that small response times were essentially unaffected by the testing performed. l This program has demonstrated that there is a sound reason for concern l about the accuracy of temperature measurements in nuclear power plants. These limited tests S N uld be expanded in a Phase 11 program to involve more sensors and longer exposures to simulated LWR conditions in order to obtain statist-ically significant data. Such data are needed to establish the leagth of mean-ingful testing or replacement intervals for safety grade RTDs. i$n important corollary benefit from this expanded program would be a better determination of achievable accuracies in RTD calibration. l 23

l [ MAIN CITATIONS AND ABSTRACTS b NUREG/CR-4939, M. Subudhi, W. E. Gunther, J. H. Taylor, R. Lofaro, K. M. Skreiner, l A. C. Sugarman, and M. W. Sheets, " Improving Motor Reliability in Naclear Power l Plants;" Voleme 1-Performance Evaluation and Maintenance Practices; Volume 2: Functional Indicator Tests on a Small Electric Notor Subjected to Accelerated Aging: Volume 3: Failufe Analysis and Diagnostic Tests on a Naturally Aged Electric Motor; Brookhaven National Laboratory. BNL-NUSEG 52031, November 1987. Volume 1: Performaste Evaluation and Maintenance Practices This report presents recommendations for developing a cost-effective program for performance evaluation and maintenance of electric motors in nuclear power plants. These recommendations are based on current industry practices, svailable techniques for monitoring degradation in motor components, manufac-turers' recommendations, operating experience, and results from two laboratory tests on aged motors. The test results (on a small and a large motor) provide the basis for recommending the various functional indicators for maintenance programs. The overall preventive program is separated into two broad areas of activity aimed at mitigating the potential ef fects of equipment aging: performance evalu-ation and equipment maintenance, The latter involves actually maintaining the condition of the.equipmeat, while the former involves inonitoring degradatico due to aging. The monitoring methods are further categorized as periodic testing, surveillance testing, centinuous monitoring, and inspections. This study focuses on relevant methods and procedures with the goal of maintainir:g the motors in a nuclear facility operationally ready. This includes an evaluation of various functional indicators to determine their suitability for trending assessments when monitoring the condition of motor components. The intrusiveness of test methods and the present state of the art for using the test equipment in a plant environment are discussed. Implementation of the information provided in this report will improve motor reliability in nuclear power plants The study indicates the kir;ds of tests to conc'uct, how and when to conduct them, and to which rtotors the tests should be applied. Volume 2: Functional Indicator Tests 09 a Small Electric Motor Subjected to Accelerated Aging A 10-horsepower electric motor was artificially aged by plug reverse cycling for test purposes. The motor was manufactured in 1967 and was in service at a commercial nuclear power plant for twelve years. Various tests were performed on the motor throughout the aging process. The motor failed after 3.79 million reversals (3 seconds per reversal) over seven months of test-ing. Each test parameter was trended to assess its suitability in monitoring aging and service wear degradation in motors. Results and conclusions are dis-cussed relative to the applicability of the tests performed to motor mainte-nance programs of nuclear power plants. I f 1 i 24

MAIN CITATICNS AND ABSTRACTS-Volume-3: Failure Analysis and Diagnostic Tests on a Naturally Aged Lerge Electric Motor Stator coils of a naturally failed 400-hp motor from the Brookhaven National Laboratory test re6ctor facility were tested for their dielectric integrities, i The motor was used to drive the primary reactor coolant pum' for the last 20 -years. Mainter.ance ectivities oh'this motor during its entire service life were minimal, with the exception of meggering it periodically. The stator consisted of ninety-individual coils, which were separated for testing. -Seven different dielectric tests were performed on the coils. Each set of data from the tested coils indicated a spectrum of variation depending on their aging conditions and characteristics. By comparing the test data to baseline data,.the test methods were assessed for application to motor maintenance programs in nuclear power plants. Also included in this study are results of an investigation to deter-mine the cause of this motor's failure. The aged condition of a second_ identical primary pump motor, which is of the same age and is presently in operation, is discussed. Recommendations relating to the applicability of each of the dielectric test methods to motor maintenance programs are formulated. NUREG/CR-4967, L. C. Meyer, "Nuclesr Plant Aging Research on High Pressure Injection Systems," Idahc National Engineering Labcratory, EGG-2514, November 1987. This report presents the results of a review of light water reactor high-pressure injection system (HPIS) operating experience, imported in the Nuclear Power Experience Data Base, Licensee Event Reports (LERs.), the Nuclear Plant Reliability Data System, and plant records. Operating experience of occlear power plants was evaluated to determine the significance of aging-related service year on equipment and its possible impact on safety. The HPIS and those portions of related systems aceded for operation of the HPIS were selected for detailed study in order to evaluate the potential signifi:ance of aging as a contributor to the degradation of that system. Tables show the percentage of significant events for HPIS classified by cause, component, and subcompo;.snts for PWRs and BWRs. A representative Babcock and Wilcox plant was selected for detailed study. The NPAR guidelines provided the framework. through which the effect of aging on HPIS was studied, and these guidelines were follcwed threughout the report, which presents an identification of f4ilure modes, a preliminary identification of f ailure casses due to aging and service wear degradation, and j a review of current inspection, surveillance, and monitoring methods, including manufacturer-recommended surveillance aRd maintenance practices. The detailed study identifies materials susceptible to aging, various stressors, and environ-mental factors. Performance parameters or functional indicaters potentially l useful in detecting degradation are also identified, and preliminary recommenda-tions are made regarding inspection, surveillance, and monitoring methods. i { \\ 25 L

MAIN CITATIONS AND ABSTRACTS NUREG/CR-4985, M. Subudhi, J. H. Taylor, J. Clinton, C.'J. Czajkowski, and J. Weeks, " Indian Pcint 2 Reactor Coolant Pump Seal Evaluations," Brookhaven National Laboratory, BNL-NUREG-52095, August 1987. This report summarizes the findings on Westinghouse reactor coolant pump- .(RCP) seal performance at Indian Point'2. This study considered a significant number of RCP seal failures. Consolidated Edison initiated a research effort to i deter 6aine the caus's of these failures and to develop appropri;.te ameliorative e action to enhance seal reliability, The,0NL work is an outgrowth of the first-phase eff9rt performed by Failure Analysis Associates. The objectives of the BHL prcgram are to determine the root causes of seal failure and to provide recommendations for improving seal reliability. This prograin made notable j advances in understanding the root causes of RCP' seal failure. For the first time, actual failed seals were examined in detail in BNL's hot cell, and labora-tory tests were conducted to determine failure causes. This report suomarizes i findings and presents conclusions'ar.d recommendations based on review of plant f operating and maintenance data, consultation with Westinghouse and utilities, review of prior RCP seal studies (including previous BNL work), and visual and in-depth examinations of the first batch of service-exposed seals received from the plant. NUREG/CR-4992, G. C. Roberts, V. P. Bacanskas, and G. J. Toman, " Aging and Service Wear of Multistage Switches Used in Safety Systems of Nuclear Power Plants," Vol.1, Oak Ridge National Laboratory, ORNL/SUB/83-28915/5/V1, September 1987. An assessment of the types and uses of multistage switches in nuclear power plant safety-rolated service is provided. T h ough a description of the opera-tion of each type of switch combined with knowledge of nuclear power plant applications and operational occurrences, the significant stressors respons-ible for multistage switch deterioration are identPied. A review of operating experience (failure data) leads to identification of potential and recomended monitoring techniques for early detection of incipient failures. Although the operating experience does not justify extensive deterioration monitoring of multistage switches, nondestructive testing meth6ds tnat could be used to ~ svaluate the condition of switches are identified. The report presents a xietailnd description of the components, materials of construction, and operation i of each of the multistag? switches included in the assessmenL Als6, it prnvides j an analysis of failure data from the LER sys'Jm. An analysis of the various failure modes of multistage rotary switches and their related causes is also ghv The existing recommended and required maintenance and surveillance practiu s are listed. Several techaiques with a potential for assessing the condition of switch components and possibly predict:ng age-related failures are identified. It is recommended that inservice failures be analyzed to determine whether the failuret are due to random defects or are the result of generic deficiencies that would require corrective action. NURPG/CR-5008, R. D. Meininger and T. J. Weir, " Development of a Testing and Analysis Methodology to Determine the Functional Conditinn of Solenoid Operated Valves," Pentek, Inc., Coraopolis, PA, September 1987. ihe objective of this research was +n develop a simple, relia' ole, condition-monitoring system that will provide survei lance information without requiring l 26

MAIN CITATIONS AND ABSTRACTS disconnection or disassembly of solenoid-operated valves (SOVs) installed in nperating nuclear' power plants. The information provided must be sufficiently reliable to a' low plant operators to conclude that valve performance has or has not degraded to the point where corrective maintenance becomes necessary. The requirea information is assumed to be obtainable through analysis of in-rmh current to the coil of the 50V. Various SOVs were tested ir. an experi-mental air system set up in tne laboratory. In rush current data acquired on degraded cnd new SOVs were analyzed to determine behavior signature models. Laboratory conditions provided the opportunity to simulate perturbations caused by the valve function, which would differ from. actuation to actuation. A visual examination of this time varying waveform revealed distinct and repcatab'e variations for different valve anomalies. This technique could identify gross changes and render characteristic rignatures that could be used for various comparisons and to trend volve degradation mechanisms and their consequences over time. Utilization of the laboratory technique in an operating nuclear power plant would be somewhat impractical since the installed valves are not equipped witn synchronous switching capability. Analytical research was therefore conducted to develop a technique to analyze similar e7ectrical data obtained under asynch-l ronous conditions typical of an operating plant. For such field application, the technique developed would use a clip-on current proba, thus enabling all measure-ments to be made from outside the reactor building uitbout disturbing any elec-trical connections. The in-rush current to the solenoid-operated valve is analyzed in real time using a personal computer and fast Fourier transform j techniques. l i NUREG/CR-5051, W. E. Gunther, R. Lewis, end M. Subudhi, " Detecting and Mitigating i Battery Charger and Inverter Aging," Brookhaven National Laboratory, l BNL-NUREG-52108, August 1988. 1 This report is the tecnnd on the two-Mep approach for assessing the safety l and operational aspects of t?.ttery charger and inverter aging in nuclear powet plants. Analyses include an assessment of the recent operating experiences with i battery chargers and inverters and a discussico 9f improvements in reliability i that may be achieved through modification of tae equipment's configuration and an increased inspection frequency. The results are evaluated from a survey of the current maintenance and test practices used in 6uclear power plants, along with the manufacturer's recommendations for maintaining equipment operability. l Advanced designs for uninterruptible pcwer systems, subcomponent improvements, I l and current monitoring and protective equipment are described and related to j l their potential applicability in nuclear power plants. i A naturally aged inverter and battery charger were tested at BNL to evaluate I the naturally aged condition, the effectiveness of condition monitoring techni-l ques, and the practicality of selected maintenance and monitoring procedures. A i portion of this research effort is covered in RIL No. 159, " Nuclear Plant Aging i Research: Safety-Related Inverters," November 9, 1988. l i 27 b __

MAIN CITATIONS AND ABSTRACTS j i A maintenance progren for battery chargers and inverters is recommended. As described in this report, such a progra") incorporat&s inspection, monitoring, testing, and repair activities that should be performed to detect and mitigate l aging effects and thereby ensure the operational readiness of this important equiplaent throughout the pihnt'h operating life. ' NUREG/CR-5052, J. C. Higgins, R. Lofaro, M. Subucihi, R. Fullwood, and J. H. Taylor, l i " Operating Experience and Aging Assessment of Component Cooling Water Systems in Pressurized Water Reactors," Brookhaven National Laboratory, BNL-NUREG-52117, July 1988. i An aging assessment of conrponent cooling water (CCW) systems in PWPs was performed as part of the NPAR program. The objectives were to provide a techni-l cal basis for the identification and evaluation Of degradation caused by age. Tne information generated will be used to assess the impact of aging on plant safety and to develop effective mitigating actions for the CCW system. The offect of time on this system was characterized by using the " Aging and Life Extension Assessment Program (ALEAP) Systems Level Plan", developed by Brook-j haven Nctional Laboratory. Failure data from various national data bases were reviewed and analyzed to identify predominant failure modes, causes, and mech-l anisms in CCW systems. Time-dependent f ailure rates for major components were calculated to identify aging trends. Plant-specific data were obtained and i evaluated 10. supplement data base results. A computer program {PRAAGE) was developed and implemented to model a typical CCW system desich and perform probabilistic risk assessment (PRA) calculations. Time-dependent failure rates were input to the program to evaluate the effects of aging on the irportance of a component with respect to system unavailability. Time-dependent changes in component importance and system unavailability with. Ege were observed and discussed. NUREG/CR-5053, W.. r and M. Subudhi, " Operating Experience and Aging Assessment of Motor Cor, trol Centers," Brookhaten National Laboratory, BNL-NUREG-52118, JuP/ 1988. As part of the NRC NPAR program, an assessment was made of the character-istics of aging and service wear of motor control tenters (MCCs). MCCs perform an important function in the operation and control of a large number of safety-related motors; thus the operability and reliability of PCCs can affect the overall safety of nuclear plants. This-report follows the NPAR strategy and investigates the operational performance, the design and manufacturing methods, and the current maintenance, surveillance, and monitoring techniques applied to MCCs. A significant result described in this report concerns the identification of important MCC f ailure modes, causes, ant! mechanisms from plant operational experience. Frequencies of failures determined for the various subcomponents of MCCs are also described. In addition, recommendations are provided for functional indicators to monitor the performance of MCCs. These functional indicators will be evaluated during Phase 2 of the program. 28

MAIN CITATIONS AND ABSTRACTS NUREG/CR-5141, V. P. Bacanskas, G. J. Toman, and S. P. Carfagno, " Aging and Qualification Research on Solenoid Operated Valves," Franklin Research Center, Norristown, PA, August 1988. Tests were conducted on three-way direct acting solenoid-operated valves (50Vs). Some SOVs had been aged naturally through service in nuclear power plants, and others were subjected to accelerated aging. Thermal aging was cun-ducted with both air and nitrogen as the process gas. Operational aging was simulated by putting the specimens through operational cycles at certain inter-vals during the accelerated thermal aging with the environmental temperature controlled at a level representative of service conditions. The program also included simulation of a design basis event (DBE) that consisted of gamma irra-diation and a main-steam-line-break loss-of-coolant accident (MSLB/LOCA) simula-tion. After each major segment of the test program (aging, irradiation, and MSLB/LOCA simulation), some of the valve specimens.were subjected to operational testing and then disassembled for inspection and measurement of physical proper ties. Performance of th,e Automatic Switch Co. (ASCO) SOVs was affected in the early stages of the program by an organic deposit of undetermined origin. Removal of the deposit eliminated the problem. A naturally aged ASCO S0V with Buna N seals and a new ASCO SOV with EPDM seals were subjected to accelerated aging with nitrogen as the process gas. These valves were the only ones to go through the entire test program withcut a failure to transfer and without any significant leakage. Valcor Engineering Co. SUVs suffered from sticking of the shaft se61 0 rings, which made it impossible to complete the accelerated thermal aging. Repeated tests and changes in test procedures failed to alter this situation, lt is possible that the stresses of accelerated aging produced effects that are not representative of service aging. Saal deterioration in the Vakor 50Vs caused leakage following DBE irradiation. The naturally aged Valccr 50V performed satisfactorily during the first high-temperature portion of the MStB/ LOCA prnfile bat malfunctioned during most of the rest of the test. Deteriorate;n of the elastomeric parts of the ASCO SOVs did not appear to l De suf ficient to account for the observed f ailures to transfer, which evidently were caused by coil deterioration. Elastomeric parts of Valcor SOVs, both from the naturally aged 50V and from the one that had not been aged, experienced suisuntial deterioration. NUREG/CR-5159, M. S. Kalsi, C. L. Horst, and J. K. Wang, " Prediction of Check Valve Performance and Degradation in Nuclear Power Plant Systems," Kalsi Engineering, Inc., Sugar Land, TX, KEI No. 1559, May 1988. Degradation and failure of swing check valves and resulting damage to plant equipment hss led to a need to develop a method to predict performance and degradation of these valves in nuclear power plant systems. This Phase I investigation developed methods that can be used to predict the stability of the check valve disk when there are flow disturbances such as elbows, reducers, 29

i I MAIN CITATIONS AND ABSTRACTS and generalized turbulence sources within 30 pipe diameters upstream of the valve. Major findings include the flow velocity required to achieve a full-open stable disk position, the magnitude of disk motion developed with these upstream disturbances (with flow velocities below full open conditions), and disk natural frequency data that can be used to predict wear and fatigue damage. Reducers were found to cause little or no performance degradation. Effects of elbows located within 5 diameters of the check valve must be considered, while severe turbulence sources have a significant effect at distances up to 10 diameters upstream of the valve. Clearway swing check designs were found to be particularly sensitive to inanufacturing tolerances and installation variables making them likely candidates for premature failure. Reducing the disk. full-opening angle on these designs results in significant performance improvement. NUREG/CR-5192, W. E. Gunther, " Testing of a Naturally Aged Nuclear Power Plant i Inverter and Battery Charger," Brookhaven National Laboratory, BNL-NUREG-52158, September 1938. A naturally aged inverter and battery charger obtained from the Shippingport facility were tested as part of the NPAR program. The objectives of this testing were to evaluate the naturally aged equipment state, determine the effectiveness of condition-monitoring recommendations, and obtain insight into the practicality of preventive maintenance and monitoring methods. Testing indicates that the equipment has retained its ability to respond to load transients. With the exception of silicon controlled rectifiers (SCRs), which were found to be operating with case temperatures ( F) 20% higher than tbse during the acceptance test, component temperatures and circuit characteristics were simi W to original acceptance test measurements. Based on these observations, it is concluded trat the inverter and battery charger have not aged substantially. The two primary monitoring techniqu~es employed were tempi.rature measurwents and electrical waveform observaticrc. Internd panel temperature and indivLwl comolent temperatures were recorded at regular intervals during steady-state and transient operations. Thermocouple imbedded within the transformer and inductor windirgs and attached to SCR and capacitor surfaces provided a nonobtrusive means of monitoring component operation. Readings taken were compared to original acceptance test data. Circuit waveforms were observed or, an hourly basis during steady-state operation and at the time load transients were applied. The inverter output voltage and the SCR gate current waveforms remained relatively constant regardless of the applied loads. Finally, this test report recommends that individual fusing of filter capacitors be considered in order to preclude a capacitor failure in the short circuit mode from rendering the inverter inoperable. Also, equipment acceptance testing should be modified to obtain the most limiting design operating condi-tions for all major subcomponents. Results indicated that aging had not sub-stantially affected equipment operation. On the other hand, the monitoring 30

MAIN CITATIONS AND ABSTRACTS 1 techniques employed were sensitive to changes in measurable componcnt and equipment parameters. Thus comparing the monitoring results with the original acceptance test data is a viable method of detecting degradation prior to catastrophic failure. PNL-5722,' D. E1 Blahnik and R. L. Goodman, " Operating Experience and Aging Assessment of ECCS Pump Room Coolers," Pacific Northwest Laboratories, October 1986. This' report provides a preliminary aging assessment of safety-related room coolers for the emergency core cooling system (ECCS) pump rooms in nuclear power planto. The assessment conforms to the NRC NPAR progras strategy and is based on limited information obtained through public and private data bases, equip-ment vendors, utility contacts, literature searches, and expert opinicn. Description of the ECCS pump room cooler systems were based on FSARs gnd vendor-supplied information. Data from LERs, review of maintenance requests at a reactor plant, and discussions with personnel that do utility repair and main-tenance work were used to deterraine the operating experience of pump room coolerr. Failure modes, causes, frequency rates, and methods of detection are summarized from the operating records. Maintenance actions and mortifications naeded ai, n result of the operator experience are addressed. Operational stressors are sum-marized, manufacturer recommendations for maintenance and surveillance are listed, and aging and service-wear monitoring are briefly avaluated. PNL-6287, K. R. Hoopingarner, B. J. Kirkwood, 6nd P. J. Lon2ecky, " Study Group Review of Nuclear Service Diesel Generator Testing and Aging Mitigction," Pacific Northwest Laboratories, March 1988. As part of the WPAR program, the Pacific 5'uthwest isbaratory is perform-ing a ditsel generatcr aging assessment study. In the on gaing NPAR Phase 11 of the aging study, efforts have been focused on aging mitigation and other success strategies for iTproving nuclear plant diesel generator operation and maintenance and also increning its reliability. A study group of diesel experts, th( mtbors of this report, set on April 29 and 30,1937, to resolve issues an mitigating di~esel generator aging and improving operations, testing, and maintenance. The focus of the study group was to (1) address the diesel generator aging stressors resulting froa the present periodic testing practices of the nuclear industry and (2) propose potential mitigating measures. A new recommended testing program was developed and is documented in this report. The report lays out the conclusions and recommendations of the study group. The 2xperts agreed that, if these recom-Mendations are put into practice, many of the engine aging stressors (e.g., those due to fast start) could be reduced or eliminated; another consequence could be a reduction of f ailures and an improvement in operability and reliability. 31

MAIN CITATIONS AND ABSTRACTS SAND 88-0754 UC-78, K. T. Gillen and R. L Clough, " lime-Temperature-Dose Rate Superposition: .A Methodology for Predicting Cable Degradation Under Ambient Nuclear Power Plant Aging Conditions," Sandia National Laboratories, August 1988. Time-temperature superposition is an empirical appr.cach that has been used in polymers for more than 39 years to make thermal aging predictions during experimentally inaccessible times. Given the historical success of time-temperature superposition, the authors have expanded this appro$ch for combined radiation-thermal environments, yieldino an empirical time-temperature-dose-rate shifting procedure. The procedure derives an isothermal curvc for a given hinount of material damage versus dose rate at a selected refereni.e temperature. This it done by finding the Arrhenius activation er,ergy that causes higher-temperature dose-rate data to superpose when shifted to the reference temperature, The resulting superposed curve at the reference temperature exterids to much lower dose rates that are experimentally inaccessible because of the long time periods that would be required to simulate aging. This procedure therefore allows meaningful predictions to be made fo long-term, low-dose-rate radiation aging conditions. Using historical data from Scndia's radiation-aging progri.m on nuclear power plant cable materials, the authors have successfully applied the time-temperature-dose-rate superposition approach to four different materials: hypalon, neoprene, polyethylene, and PVC jacket material. For two of these materials, extrapolated predictions based on the superimposed data were found to be in excellent agreement with 12 year, low-dose-rate nuclear power plant results. 32 \\

PERSONAL AUTH0R INDEX This index lists in alphabetical oroer all participating authors of each report listed in the main citation listir.g. Each name is followed by the number and the title of the reports prepared by the author. If further information is needed, refer to the min citation by the report numoer. ADAMS, M.L. NUREG/CR-4$97, " Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants. Yn1. 1: Operating Experience and Failure Identification." AHMED, S. NUREG/CR-4257, " Inspection, Surveillance, and Monitoring of Electrical Equipment Inside Containment of Nuclear Power Plants--With Applications to Electrical Cables." ANDERSON, R.L. I NUREG/CR-4928, " Degradation of Nuclear Plant Temperature Sensors." BACANSKAS, V.P. NUREG/CR-4715, "An Aging Assessment of Relay ar.d Circuit breakers and Sysum '{ Interactions." NUREGICR-4 SIS,

• Aging and Service Hear of Scienotd-Operated Valves Used in Safety Systeins of Nuclear Powr Plants. Vol.1: Oper,Hng Experience and Failure Identification."

NOREG/CR-4992, " Aging and Service Wear of Multistate Switches Used in Safety Systems of Nuclear Pcw?r Plants," Vol. J. NURtG/CR-5141, " Aging and Qualification Research on Solenoid Operated Valves." BADER, B.E. NUREG/CP-0036, " Proceedings of the Workshop on Nuclear Plant Aging." l BERRY, D.L. NUREG/CR-3818, " Report of Results of Nuclear Power Plant Aging Workshop." BLAHNIK, D.E. PNL-5722, " Operating Experience and Aging Assessment of ECCS Pump Room Coolers." i 1 33 I l \\

p. PERSONALNUTH0R.INDEX BORK0WSKI, R.J. NUREG/CR-2641, "The In-Plant Reliability Data Base for Nuclear Power Plant ' Components: -Data Collection and Methodology Report."- NUREG/CR-3154, "The In-Plant Reliability Data Base for Nuclear Plant Components: ; Interim Report--The Valve Component." BURNS,~E.L. NUREG/CR-4156, " Operating Experience and Aging-Seismic Assessment'of Electric Motors." BUSH, S.H. NUREG/CR-4279, " Aging and Service Wear of Hydraulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Plants," Vol. 1. CARFAGNO, A. NUREG/CR-4257, " Inspection, Surveillance, and Monitoring of Electrical Equipment-Inside Containment of Nuclear Power Plants--With Applications to Electrical Cables."' NUREG/CR-5141, " Aging and Qualification Research'on Solenoid Operated I Valves." CASADA, M L. hCRE3/CR-?S43, " Survey td Operating Experiences from LERs to hientify Aging l Trer.ds." CESARSK1, W.Y, NU4EG/CR-4597, " Aging and Service Vear of Auxiliary Feedwater Pumps for PHR Huclear Power Plants. Vol. 2: Aging Assessments and Monitoring Method . Evaluations." CIFUENTES, F. BNL Technical Report A-3270-11-85, " Seismic Endurance Tests of Naturally Aged Small Electric Motors." CLARK, N.H. NUREG/CR-3818, " Report of Results of Nuclear Power Plant Aging Workshop." CLETCHER, J.W. NUP.EG/CR-4692, " Operating Experience Review of Failures of Power Operated Relief Valves and Block Valves in Nuclear Power Plants." l 34 ..____.=__m__._____. ._.m_J

FERSONAL AUTHOR INDEX CLIN'ON, J. NUREG/CR-4985, " Indian Point 2 Reactor Coolant Pump Seal Evaluations." CLOUGH, R.L. SAND 88-0754 UC-78, " Time-Temperature-Dose Rate Superposition: A Methodology for Predicting Cable Degradation Under Ambient Nuclear Power Plant Aging Conditions." CORNWELL, B.C. NUREG/CR-3819, " Survey of Aged Power Plant facilities." CROWLEY, J.L. NUREG/CR-4380, " Evaluation of the Motor-0perated Valve Analysis and Test System (MOVATS) to Detect Degradation, Incorrect Adjustments, and Other Abnormalities in Motor-Operated Valves." CURRERI J. BNL Technical Report A-3270-11-85, " Seismic Endurance Tests of Naturally Aged Small Electric Motors." CZAJKGkSKI, C.J. j HUPEC/CR-4985, " Indian Point 2 Reactor Coolant Pump Seal Evchstions." I DAY!S, T., 1 f Hl' REG /CR-4144, "Importance Ranking Based on Aging Consideration af l Components Included in Pros 6bilistic Risk Assessments." DEWALL, K.G. NUREG/CR-3819, " Survey of Aged Power Plant Facilities." DINGEE, D.A. NUREG/CR-4590, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating and Expert Experience," Vols. I and 2. DINSEL, M.R. NUREG/CR-3956, "In Situ Testing of the Shippingport Atomic Power Station Electrical Circuits." D0DGE, R.E. NUREG/CR-4279, " Aging and Service Wear of Hydraulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Plants," Vol.1. 35

PERSONAL AUTH0R INDEX DONALDSDN M.R. NUREG/CR-3956, "In Situ Testing of the Shippingport Atomic Power Station Electrical Circuits." DRAGO, J.P. N'5ES/CR-2641, "The In-Plant Reliability Data Base for Nt. clear Power Plant Components: Data Collection and Methodology Report." EDSON, J.L. NUREG/CR-4457, " Aging of Class IE Batteries in Safety Systems of Nuclear Power Plants." EISSENBERG, D.M. NUREG/CR-4234, " Aging and Service Wear of Electric Motor-Operated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1. NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1. NilREG/CR a38U.

• Evaluation of the Motor-Operated Valve Analysis and Test System (MCLAYS) to Dstect Degradation, Incorrect Adjustments, and Other e

4 Abrarmalities in Motor. Operated Valves." FRAGOLA, J.R. j NURE6/CR-315t "The In-Plant Reliability Data Base for Nucicer Plant Components: Interim Repcrt--The Vdve Co conent.' FULLWOOD, R. BNL Technical Report A-3270-12-86, "/,ging and Life Extension Assessment l Program (ALEAP) Systems Level Plan." NUREG/CR-5052, " Operating Experience and Aging Assessment of Component i Cooling Water Systems in Pressurized Water Reactors." GALLAHER, R.D. NUREG/CR-3543, " Survey of Operating Experiences from LERs to Identify Aging Trends." NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol. I. GILLEN, K.T. SAND 88-0754 UC-78, " lime-Temperature-Dose Rate Superposition: A Methodology for Predicting Cable Degradation Under Ambient Nuclear Power Plant Aging Conditions." 36

PERSONAL AUTH0R INDEX GOLDBERG, F.F. NUREG/CR-2641, "The In-Plant Reliability Data Base for Nuclear Power Plant Components: Data Collection and Methodology Report." GOODMAN, R.L. PNL-5722,

• Operating Experience and Aging Assessment of ECCS Pump Room Coolers."

GREENS ~iREET, W.L.- NUREG/CR-4234, " Aging and Service Wear of Electric Motor-Operated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1. NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1. GUNTHER, W.E. NUREG/CR-4564, " Operating Experience and Aging-Seismic Assessment of Battery Chargers and Inverters." NUREG/CR-4939 "Inraving Motor ReMebfif ty in Nuclear Power P)ents;* Volume 1: Pertornance Eve.luation end Maintenece Practices; Volut? h Functional Indicatnr hsts on a Emell Electric Ntcr Subjected to Accelerated Aging; Vuinma 3: Failure Analysis and Diagnostic Tests on a Nctura51y Aged Electric Motor. MUREG/CR-5051, " Detecting end Mitigating Battery Charger end inverter Aging." NUREG/CR-5192, " Testing of a Naturally Aged Nuclear Power Plant Inverter and Battery Charger." HANCHEY, L.A. NUREG/CP-0036, " Proceedings of the Workshop on Nuclear Plant Aging." HARDIM, J.E. NUREG/CR-4457, " Aging of Class 1E Batteries in Safety Systems of Nuclear Power Plants." l HASHEMIAN, H.M l NUREG/CR-4928, " Degradation of Nuclear Plant Temperature Sensors." HEASLER, P.G. NUREG/CR-4279, " Aging and Service Wear of Hydraulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Plants," Vol. 1. l 37

p+ ' - ~ ~ - -

---

~r--

.;t. PERSONAL' AUTHOR-INDEX 'HEBBLE, T.L. NUREG/CR-2154, "The'In-Plant' Reliability. Data' Base for Nuclear' Plant. Componentsi Interim Report--The Valve Component." HIGGINS, J.C. BNL Technical Report A-3270-11-85, " Seismic Endurance Tests of.Ncturally' -Aged Small Electric Motors." BNL' Technical Report A-3270-12-86, " Aging and Life Extension Assessment Program (ALEAP)SystemsLevel. Plan."- ( NUREG/CR-5052, " Operating Experience and Aging Assessment of Component Cooling. Water Systems in Pressurized Water Jeactors." l HOLBERT, K.E; NUREG/CR-4928, "De@adation of Nuclear Plant Temperature Sensors."- H00PINGARNER, K.R p NUREG/CR 4590,:" Aging of Nucle w Station Diesel Generators' Evaluation of

4 '

. Operating and Txpert Experience," Vols. 1.and 2. PNL-6267, "$tedy Croup Revirew of Wuclear Service Diesel Generator Testing s.nd Aging Kitigation." b HORST -C.L. 1 NUREG/CR-5159, " Prediction of Check Valve Performance and Degradation in Nuclear Power Plant Systems." H0Y, H.C. NUREG/CR-3543, " Survey of Operating Experiences from LERs to identify Aging -Trends." - JOHNSON, J.W.'

NUREG/CR-3154, "The In-Plant Reliability Data Base for Nuclear Plant Components:: Interim Report--The Valve Component."

KAHL, W.K. NUREG/CR-3154, "The In.> Plant Reliability Data Base for Nuclear Plant Components: Interim Report--The Valve Component." KALSI, M.S. NUREG/CR-5159, " Prediction of Check Valve Performance and Degradation in Nuclear Power Plant Systems." 38 e_

i i PERSONAL AUTH0R INDEX I KERLIN, T.W. NUREG/CR-4928, "Degradat hn of Nuclear Plant temperature Sensors." KIRKWOOD, B.J. PNL-6287, " Study Group Review of Nuclear Service Diesel Generator Testing i and Aging Mitigation." KITCH, D.M. NUREG/CR-4597, " Aging and Service Wear of Auxiliery feedwater Pumps for PWR Nuclear Power Plants. Vol. 2: Aging Assessments and Monitoring Method Evaluations." KURTH, R. NUREG/CR-4144, "Importance Ranking Sned on Aging Consideration of Components included in Probabilistic Risk Assessments." j LEVEREN1, f. HUREG/CR-4144, "importance Ranking cased an Acing Considetetier of Components Included in Probabilistic Risk Assessments." LEWIS, R. NUkEG/CR-5051, " Detecting and Mitigating Battery Charger and Inverter l Aging." LODLOW, C.C. NUREG/CR-4715, "An Aging Assessment of Relay and Circuit Breakers anci System l Interactions." i LOFARD, R. HUREG/CR-4939, " Improving Motor Reliability in Nuclear Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Volume 2: functional Indicator Tests on a Small Electric Motor Subjected to Accelerated Aging; Volume 3: Failure Analysis and Diagno:; tic Tests on a Naturally Aged Electric Motor. NUREG/CR-5052, " Operating Experience and Aging Assessment of Component i l Cooling Water Systems in Pressurized Water Reactors." LONZECKY, P.J. PNL-6287, " Study Group Review of I;uclear Service Diesel Generator Testing and Aging Mitigation." MACDONALD, P.E. NUREG/CR-4731, " Residual Life Assessment of Majer Light Water Reactt;r Components," Vol. 1. 39 l

PERSONAL AUTHOR.INDEX i 7 MACDONALD, P.E. (Cont.) NtPEG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Components--Overview," Vol. 2 (Draft), MAKAY, F. AUREG/CR-4597, " Aging and Service Wear of Auxiliary Feedwater Purips for PWR Nuclear Power Plants. Vol. 1: Operating Experience and Failure Identification." MEALE, B.M. NUREG/CR-4747, "An Aging f ailure Survey of Light Water Reactor Safety Systems and Components," Vol. 1. NUREG/CR-4747, *An Aging Failure Survey of Light Water Reactor Safety Systems and Components," Vol. 2. MEININGER, R.D. NUREG/CR-5006, " Development of a Testing and Analysis Methodology to Determine tne Functional Condition of Solenoid Operated Valves." MEYER, t.C. NUREG/CR-4740, "huclear Plant-Aging Research on Reactor Protection Systems." NUREG/CR-4967, " Nuclear Plant Aging Research on High Pressure Injection Syst ems. " MILLER, B. BNL Technical Report A-3270-11-26-84, " Scoping Test on Containment Purge and Vent Yalve Seal Material." MORRIS B.M. YJREG-1144, B. M. Morris and J. P. Vo a, "Nucler. Plant Aging Pesearch r (NPAR) Program Plan." MURPHY, G.A. NUREG/CR-3543, " Survey of Operating Experiences f rom LERs to Identify Aging Trends." D NUREG/CR-4234, " Aging and Service Wear of Electric Motor-Operated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1. NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1. NUREG/CR-4692, " Operating Experience Review of Failures of Power Operated Relief Valus and Block Valves in Nuclear Power Plants." 40 u- -

PERSONAL AUTHOR JNDEX V: l l NAUS, D.J. NUREG/CR-4652, " Concrete Component Aging and Its Significance Relative to' Life Extension of Nuclear Power Flants." NEHRING, T. BNL Technical Report A-3270-11-85, " Seismic' Endurance Tests of Naturally Aged Small Electric Motors." NESBITT, J.F. NUREG/CR-4590, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating and Expert Experience," Vols. 1 and 2. PETERSEN, K.N. NilREG/CR-4928~ " Degradation of Nuclear Plant Temperature Sensors." PIKE, D.H. NUREG/CR-2641, "The In-Plant Reliability Data Base for Nuclear Power Plant Components: Data Collection and Methodology Report." i REICH, M. BNL Technical Report A-3270-11-85, " Seismic Endurance Tests of Naturally Aged Small Electric Motors." RIB, L.N. Letter Report, L. N. Rib, " Summaries of Research Reports Submitted in Connection with the Nuclear Plant Aging)Research (NPAR) Program," Engineering and Economics Research (EER Inc. j i ROBERTS, G.C. NUREG/CR-4819, " Aging and Service Wear of Solenoid-0perated Valves Used in Safety Systems of Nuclear Power Plants. Vol.1: Operating Experience and failure Identification." i NUREG/CR-4992, Aging and Service Wear of Multistage Switches Used in Safety Systems of Nuclear Power Plants," Vol. 1. ROSE, J.A. NUREG/CR-3819, " Survey of Aged Power Plant Facilities." !.ETTERWHITE, D.G. NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems and Components," Vol.1. 1 I 41 j

PERSONAL AUTHOR INDEX SATTERWHITE, D.d. (Cont.) NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems 3 and Components," Vol. ?. SCHLONSKI, J.S. NUREG/CR-4992, " Aging and Service We.ar of Multistage Switches Used in Safety f Systems of Nuclear Power Plants," Vol. 1. SHAFAGHI, A. NUREG/CR-4144, "Importance Ranking Based on Aging Consideration of Components Included in Probabilistic F.ish Assessments." i SHAH, Y.N. NUREG/CR-4731, " Residual Life Assessneat of Major Light Water Reactor Components," Vol. 1. I NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Coc.iponents--Overview," Vol. 2 (Draft). SHEETS, M.U. SNL Technical Report A-3270-3-86, " Testing Program for the Monitoring of Degradation in a Continuous Duty 460 Volt Class "B",10-HP Electric Motor." NtiREG/CR-4939, " Improving Motor Reliability in Nuclear Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Volume 2: Functional Indicator Tests on a Small Electric Motor Subjected to Accelerated Agir.g; Volume 3: Failure Analysis and Diagnostic Tests on a Naturally Aged Electric

Motor, i

SHIER, W. NUREG/CR-5053, " Operating Experience and AginD Assessment of Motor Control Centers." SHOOK, T.A. NUREG/CR-4715, "An Aging Assessment of Relay and Circuit Breakers and System Interactions." SILVER, M.M. BNL Technical Report A-3P70-12-G5, " Pilot Assessment: Impact of Aging on the Seismic Performance of Selected Equipment Types." SKREINER, K.M. NUREG/CR-4939, " Improving Motor Reliability in Nuclear Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Volume 2: Functional 42

PERSONAL' AUTHOR INDEX SKREINER,K.M.'(Cont.) Indicator. Tests on a Small Electric Motor Subjected to Accelerated' Aging; Volume-3: Failure Analysis and Diagnostic Tests on a Naturally Aged Electric Motor. SOBERANO, F.T. NUREG/CR-3956, "In Situ' Testing of the Shippingport Atomic Power Station Flectrical Circuits."- 4 S0WATSKEY, P.J. WOREG/CR-4597, " Aging and Service Wear of Auxiliary Feedwater Pumps for PWR' Nuclear Power. Plants. Vol. 2-Aging Assessments and Ponitoring Method Evaluations." STEELE, R. NUREG/.CR-3819,

• Survey of Aged Power Plant Facilities."

SUBUDHI, M. Letter Report, M._Subudhi, " Review of Aging-Seismic Correlation Studies en Nuclear Plant Equipment " Brookhaven National Laboratory, January 1985. BNL Technical Rep 6rt A-3270-11-85, " Seismic Endurance Tests of Naturally - Aged Small Electric Motors." BNL Technical Report A-32'/0-12-85, " Pilot Assessitent: Impact of Aging on the Seismic Performance of Selecteo Equipment Types." 'BNL Technic 61 Report A-3270-3-86, " Testing Program for the Monitoring of Degradation in a Contir<uous Duty 460' Volt Class "B",10-HP Electric Motor." BNL Technical Report A-3270-12-86, " Aging and Life Extension Assessment Program (ALEAP) Systems level Plan." NUREG/CR-4156, " Operating Experience and Aging-Seismic Assessir.ent of Electric Motors." l-NUREG/CR-4564, " Operating Experience and -Aging-Seismic Assessment of Battery i Chargers and Inverters." NUREG/CP.-4939, " Improving Motor Reliability in Nuclear Power Plants;" Volume 1: Perforinance Evaluation and Maintenance Practices; Volume 2: Functional Indicator Tests on a Small Electric Metur Subjected to Accelerated Aging; { Yolume 3 Failure Analysis and Diagnostic Tests on a Naturally Aged Electric Motor. f NUREG/CR 4985, " Indian Point 2 Reactor Coolant Pump Seal Evaluations." NUREG/CR-5051, " Detecting and Mitigating ' Battery Charger and Inverter Aging." 1

l. .[ 1, 1 i PERSONAL AUTHOR INDEX SUBUDHI,M.'('ont.) C NUREG/CR-5052, " Operating Experience and Aging Assessment of Component Cooling Water Systems in Pressurized Water Peactors.* l NUREG/CR-5053, "0perating Experience and Aging Assessment of Motor Control Centers." u -SUGARMAN,'A.C. BNL Technical Report A-3270-3-86, " Testing Program for the Monitoring of- 'l Degradt. tion in a Continuous Duty 460 Volt Class "B",10-HP Electric Motor." .NUREG/CR-4939, " Improving Motor Reliability in Nuclear Power Plants;" Volume 1:. Performance Evaluation and Maintenance Practices; Volume 2: Functional-Indicator Tests on a Small Electric Motor Subjected to Accelerated Aging; Volume 3: Failure Analysis and Diagnostic Tests on a Naturally Aged Electric Motor. TAYLOR, J.H. BNL Technical Report A-3270-11-85, " Seismic Endurance Tests of Naturally Aged Small Electric Motors." 5NL Technical Report A-3270-12-86, " Aging and Life Extension Assessment l Program (ALEAP) Systems Level Plan." j NUREG/CR-4156, " Operating Experience and. Aging-Seismic Assessment of Electric Motors." NUREG/CR-4564, " Operating Experience and Aging-Seismic Assessment of Battery Chcrgers and Inverters." .NUREG/CR-4939 " Improving Motor Reliability in Nucle r Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Voluine 2: Functional ~ Indicator Tests on 6 Small Electric Motor Subjected to Accelerated Aging; Volume 3: Failure Analysis and Diagnostic Tests on a Naturally Aged Electric Motor. NUREG/CR-4985, "ind %n Point 2 Reactor Coolar.t Pump Seal Evaluations." 4 NUREG/CR-5052, " Operating Experience and Aging Assessment of Component i Cooling Water Systems in Pressurized Water Reactors." j TDMAN, G.J. NUREG/CR-4257, ' Inspection, Surveillance, and Monitoring of Electrical i Equipment Inside Containment of Nuclear Power Plants--With Applications to Electrical Cables." L l NUREG/CR-4257, " Inspection, Surveillance, and Monitoring of Electrical ) Eqfipment in Nuclear Power Plants. Vol. 2: Pressure Transmitters." l f 44 _ ~ _ - _ _ _ _ _ _

_ _ _ _ _ _ _ _ _ _ _ _ _ _ ~ _ PERSONAL AUTH0R INDEX TOMAN, G.J. (Cont.) N0rtEG/CR-4715, "An Aging Assessment of Relay and Circuit Breakers and System Interactions." NUREG/CR-4819,

• Aging and Service Wear of Solenoid-Operated Valves Used in Safety Systems of I..iclear Power Plants. Vol.1: Operating Experience and Failure Identificat on."

NUREG/CR-5141, " Aging and Qualification Research on Solenoid Operated Valves." VASUDEVAN, R. BNL Technical Report A-3270-12-85, " Pilot Assessment: Impact cf Aging on the Seismic Performance of Seiected Equipment Types." VAUSE J.W. NUREG/CR-4590, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating ard Expert Experience," Vols. I and 2. VESELY, W.E. NUREG/CR-4769, " Risk Evaluations of Aging Phenomena: The Linear Aging Reliability Model and Its Extensions." VORA, J.P. NUREG-1144, B. M. Morris and J. P. Vora, " Nuclear Plant Aging Research (NPAR)ProgramPlan." NUREG-1144, J. P. Vora, " Nuclear Plant Aging Research (NPAR) Program Plan," Rev. 1. WANG, J.K. NUREG/CR-5159, " Prediction of Check Valve Performance and Degradation in Nuclear Power Plant Systems." WEEKS, J. NUREG/CR-4985, " Indian Point 2 Reactor Coolant Pump Seal Evaluations." WEIR, T.J. NUREG/CR-5008, " Development of a Testing and Analysis Methodology to Determine the Functional Condition of Solenoid Operated Valves." 45

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CORPORATE AUTH0R INDEX l i This index lists, in alphabetical order, the organizations that prepared the reports listed in this compilation. Listed 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. ANALYSIS AND MEASUREMENT SERVICES CORP. NURE3/CR-4928, " Degradation of Nuclear Plant Temperature Sensors.' BROOKHAVEN NATIONAL LABORATORY (BNL) Letter Report, M. Subudhi, " Review of Aging-Seismic Correlation Studies on Nuclear Plant Equipment," Brookhaven National Laborato y, January 1985. BNL Technical Report A-3270-11-26-84, "Sceping Test on Containment Purge and Vent Valve Seal Material." BNL Technical Report A-3270-11-85, " Seismic Endurance Tests of Naturally Aged Small Electric Motors." BNL Technical Report A-3270-12-85, " Pilot Assessment: Impact of Aging on the Seismic Performance of Selected Equipment Types." BNL Technical Report A-3270-3-86, " Testing Program for the Monitoring of Degradation in a Continuous Duty 460 Volt Class "B",10-HP Electric Motor." BNL Technical Report A-3270-12-86, " Aging and Life Extension Assessment Pro-gram ( ALEAP) Syste:as tevel Plan." NUREG/CR-4156, " Operating Experience and Aging-Seismic Assessment of Electric Motors." NUREG/CR-4564, " Operating Experience and Aging-Seismic Assessment of Battery Chargers and Inverters." NUREG/CR-4715, "An Aging Assessment of Relay and Circuit Bt eakers and System j Interactions." NUREG/CR-4939, " Improving Notor Reliability ir Nuclear Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Volume 2: Functional Indicator Tests on a Small Electric Motor Subjected to Accelerated Aging; Volume 3: Failure Analysis and Diagnostic Tests on a Naturally Aged Electric Motor. NUREG/CR-4985, " Indian Point 2 Reactor Coolant Pump Seal Evaluations." 47

10 1 'l b CORP 0h!.TE AUTi4R INDEX BROOKHAVEN NATIONAL LABORATORY (BNL) (Coiit.). '.NUREG/CR-5051,

• Detecting and Mitigating ' Battery Charger and Inverter Aging."

l l

NUREG/CR-5052,'"Opbrating Experience an'd Aging-Assessment of Component' Cool-ing Water Systems in Pressurized Water Reactors."

NUREG/CR-5053, " Operating Experience and Aging Assessment of Motor Control Centers." NUREG/CR-5192, " Testing of a Naturally Aged Nuclear Power Plant Inverter and - i Battery Charger." l ENGINEERINGANDECONOMICSRESEARCH,INC.(EER) Letter Report, L. N. Rib, " Summaries of Research Reports Submitted in Connec-tion with the Nuclear Plant Aging'Research (NPAR) Program," Engineering and j Economics Research (EER) Inc.- ) FRANKLIN RESEARCH CENTER NUREG/CR-4715,."An Aging Assessment of Relay and Circuit Breakers and System ' Interactions." NUREG/CR-5141, " Aging and Qualifiestion Research on Solenoid Operated Valves." IDAH0-NATIONALENGINEERINGLABORATORY(INEL) NUREG/CR-3819, " Survey of Aged Power Plant Facilities." NUREG/CR-3956, "In Situ Testing of the Shippingport Atomic Power Station Electrical Circuits." NUREG/CR-4457, " Aging of Class IE Batteries in Safety Systems of Nuclear Power Plants." l NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Compo-nents," Vol. 1. NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Components--0verview," Vol. 2 (Draft). HUREG/CR-4740, " Nuclear Plant-Aging Research on Reactor Protection Systems." NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems and Components," Vol. 1. NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems and Components," Vol. 2. NUREG/CR-4769, " Risk Evaluations of Aging Phenomena: The Linear Aging Reli-ability Model and Its Extensions." 48

CORPORATE-AUTHOR INDEX IDAHO NATIONAL ENGINEERING LABORATORY (INEL) (Cont.) NUREG/CR-4967, " Nuclear Plant Aging Research on High Pressure Injection Systems." KALSI ENGINEEMNG, li'C. NUREG/CP.-5159, " Prediction of Check Yalve Performance and Degradation in Nuckar Power Plant Systemsf NUCLEAR REGULATORY CCMMISSION (l'RC) TechnicalIntegrationReviewGroupforAgingcndLifeExtension(TIRGALEX), " Plan for Integration of' Aging and Life-Extension Activities." NUREG-1144, B. M. Morris and J. P. Vora, "flutlear Plant Aging Research (NPAR) Program Plan." NUREG-li44, J. P. Vera, " Nuclear Plant Aging Research (NPAR) Program Plan," Rev. 1, GAK RIDGE NATIONAL LABORATORY (ORNL) NUREG/CR-2641, "The In-Plant Reliability Data Base for Nuclear Power Plant Components: Data Collection and Methodology Report." NUREG/CR-3154, "The In.. Plant Reliability Data Base for Nuclear Plant Compo-nents: Interim Weport--The Valve Ccmponert." NUREG/CR-3543, " Survey of Operating Experiences from LERs to Identify Aging Trends." NUREG/CR-4234, " Aging and. Service Wear of Electric Motor-0perated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1. NUREG/CR-4257, " Inspection, Surveillance, and Monitoring of Electrical Equip. ment Inside Containment of Nuclear Power Plants--With Applications to Elec-trical Cables." NUREG/CR4257, " Inspection, Surveillance, and Monitoring of Electrical Equip-ment in Nuclear Pcwer Plants. Vol. 2: Pressure Transmitters." NUREG/CR-4302, " Aging and Service Hear of Check Valves Used in Engineered Safety-Feature Systems cf Nuclear Power Plants," Vol.1. NUREG/CR-4380, " Evaluation of the Motor-0perated Valve Analysis and Test System (MOVATS) to Detect Degradation, Incorrect Adjustments, and Other Ab-normalities in Motor-0perated Valves." 1 NUREG/CR-4597, " Aging and Service Wear of Auxiliary feedwater Pumps for PWR Nuclear Power Plants. Vol. 1: Operating Experience and Failure Identification." 49

I CORPORATE AUTHOR INDEX CAKRIDGENATIONALLABORATORY(ORNL)(Cont.) .NUREG/CR-4597, " Aging and Servite Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants. Vol. 2: Aging Assessments and Monitoring Method Evalations." NUREG/CR-4652, " Concrete Component Agfng and Its Significance Relative to Life Extension of Nuclear Power Plants." j l NUREG/CR-4692, " Operating Experience Review of Failures of Power Operated Relief Yalves and Block Valves in Nuclear Power Plants." I i NUREG/CR-4819, "tiging and Service Wear of Solenoid-0perated Valves Used in Safety Systems of Nuclear Power Plaats. Vol.1: Operating Experience and -Failure Identification." 1 NUREG/CR-4992, " Aging anc' Service Wear of Multistage Switches Used in Safety 1 Systems of Nuclear ~ Power Plants," Vol. 1. PENTEK, INC. NUREG/CR-5C/08, " Development of a Testing and Analysis Methodology to I Determine the Functional Condition of Solenoid Operated Valves." PACIFIC NORTHWEST LABORATORIES (PNL) NUREG/CR-4144, *!mportance Ranking Based on Aging Consideration of Components Included in Probabilistic Risk Assessments." NUREG/CR-4279, " Aging and Service Wear of Hydraulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Plants," Vol.1. i NUREG/CR-4590, " Aging of Nuclear Station Diesel Genebators: Evaluation of Operating and Expert Experience," Yols. I and 2. j PNL-0722, " Operating Experience and Aging Assessment Of ECCS Pump Room Coolers." PKL-6287, " Study Group Review of Nucleae Service Diesel Generator Testing and i Aging Hitigation." SANDIA NATIONAL LABORATORIES (SAND) i NUREG/CP-0036, " Proceedings of the Workshop on Nuclear Plant Aging." NUREG/CR-3818, " Report of Results of Nuclear Power Plant Aging Workshop." i i SAND 88-0754 UC-78, " Time-Temperature-Dose Rate Superposition: A Methodology ] for Predicting Cable Degradation Under Ebient Nuclear Power Plant Aging 1 Conditions." I l 50 L 1 l

SUBJECT INDEX In this inder, the reports are listed under one or more of the following subjects: 1. Aging, including plans, surveys, anhl tes, methods, and models; f 2. Diese? generators and related systems; 3. Electric power systems, including cables, trays, connectors, circuit breakers, switches, and related components; 4. Electrinal equipment, including motors, batteries, chargers, and inverters; 5. Instrumentation and measurement methods; 6. Maintenance; 7. Major components: reactor vessels, steam generators, pressurizer, and structures (including containment); 6. M6nitoring; 9. Operating experience, field results, and related data;

10. Piping, including valves, seals, supprts, snubbers, and related components;

11. Probabilistic risk assessment (PRA);

12.. Safety and protection systems (including injection systems) and their components; 13.

Seismic effects on aging; 14 Service veter, auxiliary feedwater, and other fluid systems, including pumps, heat exchargers, and related components. These subjects are not intended to include every subject covered in all the reports listed. Nor do they represent a " standard" or " official" list of subjects. They were selected to be most helpful to knowledgeable personnel seeking published information on the various aspects of nuclear plant aging. AGING, INCLUDING PLANS, Sun EYS, ANALYSES, METHODS, AND MODELS Letter Report, L. N. Rib, " Summaries of Research Reports Submitted in Connec-tion itith the Nuclear Plant Aging Research (NPAR) Program," Engineering and Economics Research (EER) Inc. Technical Integration Review Group for Aging and Life Extension (TIRGALEX), " Plan for Integration of Aging and Life-Extension Activities." 1 BNL Technical Report A-3270-12-86, " Aging and Life Extension Assessment Pro-gram (ALEAP) Systems Level Plan." i NUREG-1144, B. M. Morris and J. P. Vora, " Nuclear Plant Aging Research (NPAR) l Program Plan." NCREG-1144, J. P. Vora, " Nuclear Plant Aging Research (NPAR) Program Plan," l Rev. 1. I 51

7 m-l! L p SUBJEi:T INDEX I 1 ACING, INCLUDING PLANS, SURVEYS, ANALTSES, NETHODS, AND MODELS (Cent.) I NUREG/CP-0036, " Proceedings of the Workshop on Nuclear Plant Aging." i NUREG/CR-3818. " Report of Reruits of iuclear_ Power P' ant Aging Workshop." NUREG/CR-4144, "Importance Ranking Btrea on Aging Consideration of Components 1 Included in Probebilistic Risk Assessments." NUREC/CR-4652, " Concrete Componant Aging and Its Significance Relative.to Life Extension of Nuclear Power Plants." NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Compo-nents," Voi. 1. NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Compo- -nents --Overview," Vol. 2 (Draft). NUREG/CR-4769, " Risk Evaluations of Aging Phenomenn The Linear Aging Reli-ability Podel and Its Extensions." NUREG/CR-5008, " Development cf a Testing and Analysis Methodology to Deter-mine the Functional Condition cf Solenoid Operated Valves." SAND 88-0754 0C-78, " Time-Temperature-Dose Rate Superposition: A Methodology for Predicting Cable Degradation Under Ambient Nuclear Power Plant Aging Conditions." DIESEL GENERATORS AND RELATc0 SYSTEFS NUREG/CR-4690, " Aging of Nuclear Station Diesel Generators: Evelvation of Operating and Expert Experience," Vols. I and 2. NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Compo-nents --0verview," Vol. 2 (Draft). PNL-6287, " Study Group Review of Nuclear Service Diesel Generator Testing and . Aging Mitigation." ELECTRIC PCKER SYSTEMS, INCLUDING CABLES, TRAYS, CONNECTORS, CIRCUIT BREAKERS, SWITCHES, AND RELATED COMP 0NENTS BNL Technical Report A-3270-12-85, " Pilot Assessment: Inipact of Aging on the Seismic Performance of Selected Equipment Types." l L NUREG/CR-3956, "In Situ Testing of the Shippingport Atomic Power Station L Electrical Circuits." NUREG/CR-4257, " Inspection, Surveillance, and Monitoring of Electrical Equip-ment Inside Centairrut of Nuclear Power Plants--With Applications to Elec-trical Cattles." NUREG/CR-4715, "An Aging Assessment of Relay and Circuit Breakers and System Interactions." 52 t-u--

'n SUBJECT INDEX ELECTRIC POWER SYSTEMS, INCLUDING CABLES SWITCHES, AND RELATED COMP 00ENTS (Cont.), TRAYS, CONNEC10RS, CIRCUI L NUREG/CR-4731, " Residual !.ife Assessment of Major Light Water Reactor Compo-nents --Overview," Vol. 2 (Dr.sf t). NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems ar,d Components," Vol. 1. FAEG/CR-4992, " Aging and Service Wear of Multistsge Switches Used in Safety-tystems of tluclear Power Plants," Vol. 1. SAND 88-0754 UC-78, ' Time-Temperature-Dose Rate Superposition: A Methodology for Predicting Cable Degradation Unahr Ambient duclear Power Plant Aging Conditions." ELECTRICAL EQUIPMENT, iWCLUDING MOTORS, BATTERIES, CI!ARGERS, AND IFVERTERS BNL Technical Report A-3270-11-85, " Seismic Endurance tests of Naturally Aged Small Electric Motors." BNL Technical Report A-3?70-12-85, "Pi)at Assessment: Impact of Aging on the Sehmic Performance of Selected Equipment Types." BNL Technical Repo-t A-3270-3-86, " Testing Program for the Monitoring of DegradLticu in a Continuous Duty 460 Volt Clars "B",10-HP Electric tbtor." NUREG/CR-4156, " Operating Experience and Aging-Seismic Assessment of Electric Motors." NUREG/CR-4457, " Aging of Class IE Batteries in Safety Systems of Nucleu Power Plants." NU5G/CR-4564, " Operating Experience and Aging-Seitmic Assessment of Battery Chargers and Inverters." U!JREG/CR-4939, " Improving Motor Reliability in Wuclear Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Yolume 2: Functional Indicator Tests on a Small Elertric Motor Subjected to Accelerated Agir.g; Volume 3: failure Analysis and Diagnostic Tests on a Naturally Aged Electric

Motor, l

NbisEG/CR-5051, " Detecting and Mitigating Bittery Charger and Inverter Aging." NUREG/CR-5053, " Operating Experience and Aging Assessment of Motor Ctr. trol 4 Centers." HURG/CR-5192, " Testing of 6 Naturally Aged Nuclear Power Plant inverter and Battery Charger." INSTRUMENT / TION AND MEASUREMENT METHODS NUREG/CR-4257, " Inspection, Surveillance, and Monitoring of Electrical Egn y-ment in Nuclear Power Plants. Vol. 2: Pressure Transmitters." 53

i '[ g 4, ' SUBJECT INDEf u. LINSTRUMENTATIONANDMEASUREMENTHETHODS(Cont.) NOREG/CR-4928 " Degradation of Nuclear Plant Temperature Sensors." l II 1MAlh!ENANCE NUREG/CR-4234, " Aging an0 Service Wear of Electric Notor-0perated Valves Used ) l- 'in Engineered Safety-FeElvre Systems of Nuclear Powe* Platts." Vol.1. 1 l. 1; . NUREG/CR-445/, " Aging of Class 1E Batteries in Safety Systems of Nuclear ) l Power Plants." l NUREG/CR-4564,'" Operating Experience.and Aging-S;ismic Assessment of Battery Chargers and Inverters." j NUREG/CR-4597, " Aging ana Service Wear of' Auxiliary Feedweter Pumps for PWR Nuclear Fower Plants. Vol. 1:.0perating Experience and failure Identification." NUREG/CR-4597, " Aging and Service Fear of Asiliary Fcedwater Puinps for PWR ' Nuclear Power Plants. Y91. 2: Aging Assessments and Monitorin's Method 5 valuations." ~ l NUREG/CR-493E, "Imprnving Motor Reliability in Nuclear Power Plants;" Yolume lL h Performan g Evaluetion and Maintenance Practices; Volume 2: Functinrtal Indicator Tests on a Small Electric Motor Subjected to Accelerated. Aging; Volume 3: failure Analysis and Diagriostic fests on a Naturally Aged Electrf c -Motor." NUREG/CR-5051, *Dotecting and Mitigating Batter; Charger and Inverter Ag'ng." PNL-5722, " Operating Experience and Aging Assessment of ECCS Pump Room Coolers." l MAJOR COMPONENTS: REACTOR YESSELS, STEAf! GENERATORS, PRESSURIZER, AND STRUC-TURES (INCLUDING CONTAINMENT) isVRE(i/CR-4652, " Conc ~ ete Component Aging and Its Significance relative to r Life Extension of Nuclear Power Plants." NUREG/CR-6731. " Residual Life Assessment of Ma,icr Light Water Reactor Coalpo-l nents," Vol. 1. ~ l NUREG/CR-4731, " Residual Life Assessment of Major Light Yater Reccter Compo-necta --Overview," Vol. 2 (Draft). MONIT0 KING BNL Technical Report A-3270-3-86, " Testing Program for the Monitoring of Deo~radation in a Continuous Duty 460 Volt Class "B",10-HP Electric Motor." BRL Technical Report A-3270-12-85, "Apfng and Life Extension Assessment Pro-gram (ALEAP) Systems Level Plan." E4 ~_.

_ = _ _ - _. } 4 4 ~ -SUBlECT INDEX P 1 . It0tlITORING (Cr.nt.) g. h0 REG-1144, J. P. Vora, " Nuclear Plant Agin.q.Research'(flPAR) Program Plan," . Rev. - 1 ~.. 'NUREG/CR-3543,;" Survey of Operating Experiences from LERs to Identify Aging Trends."'- NUREG/CR-62b7, " Inspection, Surveillance, and Monitoring of Electrical Equip-ment Inside Containment' of Nuclear: Power Plants--With Applications to Elec-trical. Cables." NUREG/CR-4257, " Inspection, Surveillance, and Monitoring of Electrical Equip-ment in Nuclear Power Plants. Vol. 2: Pressure Transmitters." NUREG/CR-4302, " Aging and Service Wear of Check Yalves used in Engineered Safety-Feature Systems of Neclear Fower Plants," \\'ol.1. NUREG/CR-4457. " Aging of Class IE Batteries in Safety Systems of hucle.ar Power Plants." .NUREG/CR 4!564, "Operatug Experience and Aging-Seismic Assestment of Battery Chargers and Inverters." NUREG/CR-4597, " Aging and Service ~ Wear of Auxiliary Feedwater Pumps for PWR Nuclear i'ower Plants. Vol. 1: Operating Experience and Failure Identification."' NUREG/CR-4597, " Aging and SePvice Wear of Auxiliary feedwater Pomps for PWR Nuclear Power Plantsv Vol ?: Aging Assessments and Monitoring Method Evaluations." EUDEG/CR-481S,." Aging and Service Uear of Solencid-Operated Valus Used in Safety Systens of Nuclear Powar Plants. Vol.1: Operating Experience and Failure Identification." NUREG/CR-4939, "Improvin] Motor Re' liability in Nuclear Power Plantt,;" Volume 1: Performance Evaluation and Maintenance Practices; Volume 2: Functional Indicator Tests on a Small Electric Motor Su'ojected to Accelerated Aging; l Volume 3: Failure Analysis end Diapf.ostic Tests on a Naturally Aged Electric b Motor. NUREG/CR-4967, " Nuclear Plant Aging i'esearch on High Pressure Injection Systems." HUREG/CR-5008, " Development of a Testir:9 and Analysis Methodology to Deter-mine the Functional Condition of Solenoid Operated Valves." i. NUREG/CR-5051, " Detecting and Mitigating Battery Charger arid Inverter Aging." NUREG/CR-5053, " Operating Experience and Aging Assessment of Motor Control Centers." 55 w

r I SUBJECT INDEX l MCHITORING (Cont.) NUREG/CR-5192, " Testing of a Naturally Aged Nuclear P9wer Plant Inverter and Battery Charger." l PNL-6287, " Study Group Review of Nuclear Service Diesel Generator Testing u,d Agirg Mitigation." OPERATING EXFERIENCE, FIELD RESULTS, AND RELATED DATA N!! REG /CR-2641, "ha In-Plent Reliability Data Base for Nuclear Power flant Components: Data Collection and Methodology Report." NUREG/CR-3"M, "The In-Plant Reliability Data Base for Nuc! car Plant Compo-nents: Interb Report -The Valve Component." 1 NUREG/CR-3543, "Survry of Operating Experiences from LERs to Ioentify Aging Trends." NUREG/CR-3819, "3urvey of Aged Power Plant Facilities." NUREG/CR-4156, " Operating Experience and Aging-Seismic As;essment of Electric Hotors." NUREG/CR-42$4, " Aging and Service Wear of Elec?-ic Notor-0perated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Yol.1. NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems af Nuclear Power Plants," Vol.1. MJREG/CR-4457, " Aging of Class 1.E Batteries in Safety Systems of Nuclear l Power Plants." NUREG/CR-4564, " Operating Experience and Aging-Seismic Assessment of P2ttery Chargers and Inverters." NUREG/CR-4590, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating and Expert Experience," Yols. 1 and 2. NUREG/CR-4597, " Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants. Vol. 1: Operating Experience and failure Identification." NUREG/CR-4692, " Operating Experience Review ef failures of Power Operated Relief Valves and Block Valves in Nuclear Power Plants. NUREG/CR-4715, "An Aging Assessment of Relay and Circuit Breakers and System Interactions." NUREG/CR-4740, " Nuclear Plant-Aging Research on Recctor Protection Systems." NUREG/CR-4747, "An Aging failure Survey of 1,ight Water Reactor Safety Systems 3 and Components " Vol. 1. 56 c-d

SUfklECT INDEX OPERATING EXPERIENCE, FIELD RESULTS, AND REIATED DATA (Cont.) NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems and Components," Vol. 2. NUREG/CR-4819, " Aging and Service Wear of Solenoid-Operated Valves Used in Safety Systems of Nuclear Power Plants. Vol.1: Operating Experience and Failure Identification." NUREG/CR-4967, " Nuclear Plant Aging Research on High Pressure Injection Systems. " NUREG/CR-4992, " Aging and. Service Wear of Multistage Switches Used in Safety Systems of Nuclear Power Plants," Vol.1. NUREG/CR-5052, " Operating Experience and Aging Assessment of Component Cool-ing Water Systems in Pressurized Water Reactors." PNL-5722, " Operating Experience and Aging Assessment of ECCS Pump Room Coolers " PIPING, INCLUDING VALVES, SEALS. SUPPORTS, SNUBBERS, At'D RELATED COMPONENTS BNL Technical Report A-3270-11-26-84, " Scoping Test cn Containment Purge end Vent Valve Seal Material." BNL Technical Report A-3270-12-85, " Pilot Assessment: Inipact of Aging on the Seismic Performance of Selected Equipment Types." NUREG/CR-3154, "The In-Plant Reliability Data Base for Nuclear Plant Compo-nents: Interin Report--The Valve Component." NUREG/CR-4234, " Aging and Service Wear of Electric Motor-0perated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1. NUREG/CR-4279, " Aging and Service Wear of Hydraulic ar.d Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Plants," Vcl. 1. NUREG/CR-4302, " Aging and service Wear of Check Vaives Used in Engineered Safety-Feature Systens 0f Nuclear Power Plants," Vol.1. NUREG/CR-4380,)" Evaluation of the tiotor-Operated Valve Analysis and Test System (POVATS to Detect Degradation, Incorrect Adjustments, and Other Ab-normalities in Motor-0perated Valves." NUREG/CR-4819, " Aging and Service Wear of Solenoid-0perated Valves Used in Safety Systems of NJclear Power Plants. Vol.1: Operating Experience and Failure Identification." NUREG/CR-4985, " Indian Point 2 Reactor Coolant Pump Seal Evaluations." NUREG/CR-5000, " Development of a Testing and Analysis Methodology to Deter-mine the Functional Condition of Solenoid Operated Valves." 1 I ] l 57 l __s

j. SUBJECT INDEX l l-1 PIPINC, INCLUDING VALVES, SEALS, SUPPORTS, SNUBBERS, AND RELATED COMP 0NENTS j NUREG/CR-5141, " Aging and Qualification Research on Solenoid Operated Valves." NUREG/CR-5159,

• Prediction of Check Valve Perfomance and Degradation in Nuclear Power Plant Systems."

l PROBABILISTICRISKASSESSMENT(PRA) l

1. UREG/CR-4144, "Importance Ranking Based on Aging Consideration of Components Included in Probettistic Risk Assessments."

SAFETY AND PROTECTION SYSTEMS (INCLUDING INJECTIO.N SYSTEMS) AND THEIR COMPONENTS NUREG/CR-3819, " Survey of Aged Power Plant Facilities." NUREG/CR-4302, " Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems-of Nuclear Power Plants," Vol.1. i NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reactor Compo-nents --Overview," Vol. 2 (Draft). NUREG/CR-4740, " Nuclear Plant-Aging Research on Reactor Protection Systems." NUREG/CR-4747, "An Aging Failure Surycy of Licht Water Reactor Safety Systems and Components," Vol. 1. l NIJREG/CR-4747, "An Aging failure Survey of Light Water Reactor Safety Systems and Components," Vol. 2. NUREG/CR-4967, " Nuclear Plant Agina Research on High Prcssure Injection Systems." ~ NUREG/CR-4992, " Aging anti Service Wear of Multistage Switches Used in Safety Systems of Nuclear Power Plants," Vol. 1. SEISMIC EFFECTS ON ACING Letter Report, M. Subudhi, " Review of Aging-Seismic Correlation Studies on Nuclear Plant Equipment," Brookhaven National Laboratory, January 1985. BNL Technical Report A-3270-11-P5, " Seismic Endurance Tests of Naturally Aged Small Electric Motors." BNL Technicel Report A-3270-12-85, " Pilot Assessment: Impact of Aging on the Seismic Performance of Selected Equipment Types." NUREG/CR-4156, " Operating Experience and Aging-Seismic Assessment of Electric Motors." NUREG/CR-4?79, " Aging and Service Wear of Hydraulic and Mechanical Snubbers Used on Scfety-Related F'iping and Compcnents of Nuclear Power Plants," Vol.1. 58

- - _ - _ - = _ _ _ -. SUBJECT INDEX SERVICE WATER, A.UXILIARY FEEDWATER, AND OTHER FLUID SYSTEMS, INCLUDING PUMPS, HEAT EXCHANGERS,'AND RELATED COMPONENTS NUREG/CR-4597, " Aging and Service Wear of Auxiliery F*edwater Pumps for PWR Nuclear Power Plants. Vol. 1: Operating Experience and failure Identification." NUREG/CR-4597, " Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants, Vol. 2: Aging Assessments and Monitoring Method Evaluations." NUREG/CR-4731., " Residual Life Assessment of Major Light Water Reactor Compo-nents," Vol. 1. NUREG/CR-4731, " Residual Life Assessment of Major Light Water Reacter Compo-nents --Overview," Vol. 2 (Draft). NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems and Components," Vol. 1. HUREG/CR-4985, " Indian Point 2 Reactor Coolant Pump Seal Evaluations." NUREG/CR-5052, " Operating Experience ano Aging Assessment cf Component Cool-ing Water Systems in Pressurized Water Reactors." 59

b; m g r ) 1 CHRON0 LOGICAL LISTING l (Inorderofpublication) 1.* NUREG/CR-2641, 'J. P. Drago, P. J. Borkows'ki, D.11. Pike, and F. F. Goldberg,

"The' In-Plant Reliability' Data Base for Nuc? car Power Plant Components:

Data Collection and Methodology. Report," Oak Ridge National Laboratory, ORNL/TM-8271, July 1982. 2.* WUREG/CP-0036, (Co:npilation by) B. E. Bader and L. A. Hanchey, " Proceedings of the Workshop on Nuclear Plant Aging," Sandia National Laboratories, SAND 82-2264C, November 1982. 3. NUREG/CR-3154, R. J. Borkowski, W. K. Kdl, T. L. Hebble, J. R. Fragola, and. J. W. Johnsoa, "The In-Plant Reliability Data Base for Nuclear Plant ' Components: Interim Report--The Vaive Component," Oak 8"ge National Laboratory, ORNL/TM-8647, December 1983. 4. NUREG/CR-3543, G. A. Murphy, R. B. Gallaher, M. L. Casada, and H. C. Hoy, " Survey of Operating Experiences from LERs to Identify Aging Trends," Cak Ridge National Laboratory, ORNL-hSIC-216, January 1984. 5. NUREG/CR-3818, N. H. Clark and D. L. Berry, " Report of Results of Nuclear-Power Plant Aging Workshop," Sacdia National Laboratories, SAND 84-0374, August'1984. 6. BNL Technical Report A-3270-11-26-84, B. Miller, " Scoping Test on Containment Purge and Vent Valve Seal Material," Brookhaven National Laboratory, December 1984. 7. Letter Report, M. Subudhi, " Review of Aging-Seirmic Correlation Studies on Nuclear Plant Equipment," Brookhaven National Lahoretory, Jarmary 1985. 8. NUREG/CR-4144, T. Davis, A. Shafsshi, R. Kurth, and F. Leverenz, "Importance . Ranking Based on Aging Consideration of Components Included in Probabilistic Risk Assessments," Pacific Northwest Laboratories, PNL-5389, April 1985. 9. NUREG/CR-4156, M. Subudhi, E. L. Berns, and J. H. Taylor, " Operating Experience and Aging-Seismi: Assessment of Electric Motors," Brookhaven National Laboratory, BNL-NUREG-51861, June 1985.

10.. NUREG/CR-4234, W, L. Greenstreet, G. A. Murphy, and D. M. Eissenberg, " Aging ar.d Service Wear of Electric Motor-OptratEd Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1, Oak Ridge National Laboratory, ORNL-6170/VI. June 1985.

• Not produced as a part of the NRC NPAR Program, but used in developing the early liPAR praram plan.

61

) l CHRON0!.0GICAL L1 STING 1

11. NUREG/CR-3819, 'J. ' A. Rose, R. Steele,' Jr., K. G. DeWall, and B. C. Cornwell, 3

" Survey of Aged Power Plant Facilities," Idaho National Engineering - Laboratory, EGG-2317, ' June 1985. - q .32?.NUREG-1144,.B. M. Morris and J. P. Vora, " Nuclear Plant Aging Research '(NFAR) Program Plan," U.S. Nuclear Regulatory Commission, July.1985. 13.' NUREG/CR 4257, S.'Ahmed, A. Carfagno, and G. J. Toman, " Inspection, Surveillance, and Monitoring of Electrical Equipment Inside Containment of Nuclear Power Plants--With Applications to Electrical-Cables," Oak Ridge National Laboratory, ORNL/SUB/83-28915/1, August 1985.

14. BNL Technical Report A-3270-11-85, J. H. Taylor, M. Subudhi, J. Higgins, J.

Curreri,.M. Reich, F. Cifuentes, and T. Nehring, " Seismic Endurance Tests of Naturally. Aged Small Electric Motors," Brookhaven National Laboratory, November 1985.

15. BNL Technical Report A-3270-12-85, M. M. Silver, R. Vasudevan, and M, Subudhi, " Pilot Assessment:

Impact of Aging on the Seismic = Performance of Selected Equipment Types," Brookhaven National Laboratory, December 1985.

16. NUREG/CR-4302, W. L. Greenstreet, G. A. Murphy, R. B. Gallaher, and D. M.

Eissenberg,:" Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants," Vol.1, Oak Ridge National Laboratory, ORNL-6193/V1, December 1985.

17. NUREG/CR-4380, J. L. Crowley and D. M. Eissenberg, " Evaluation of the Motor-Operated Valve Analysis and Test System (M0 VATS) to Detect Degradation,. Incorrect Adjustments, and Other Abnormalities in Motor-

.0perated Valves," Oak Ridge National Laboratory, ORNL-6226, January 1986. 18.- NUREG/CR-4279, S. H. Bush, P. G. Heasier, and R. E. Dodge, " Aging and Service Wear of Hydraulic and MecLanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Plants," Vol. 1 Pacific Nurthwest Laboratories, FNL-5479, February 1986.

19. BNL Technical Report A-3270-3-86, A. C. Sugarman, M. W. Sheets, and M.

Subudhi, " Testing Program for the Monitoring of Degradation in a Continuous Duty 460 Volt Class "B", 10-HP Electric Motor," Brookhaven National ' acoratory, March 1986.

20. NUREG/CR-4564, W. E. Gunther, M. Subudhi, and J. H. Taylor, " Operating Experience and Aging-Seismic Assessment of Battery Chargers and Inverters,"

Brookhaven National Laboratory, BNL-NUREG-51971, June 1986.

21. NUREG/CR-4597, M. L. Adams and E. Makay, " Aging and Service Wear of Auxiliary Feedwater Pumps for PWR Nuclear Power Plants. Vol.1: Operating Experience and FM lure Identification," Oak Ridge National Laboratory, ORNL-6282/V1, i 1986.

Toman, " Inspection, Surveillance, and Monitoring of

22. NUREG/CR-4257,u.

o. Electrical Equipment in Nuclear Power Plants. Vol. 2: Pressure Transmitters," Dak Ridge National Laboratory, ORNL/SUB/83-28915/3/V2, August 1986. 62

CHRON0 LOG 1 CAL LISTING

23. NUREG/CR-4652, D. J. Naus, " Concrete Component Aging and Its Significance Relative to Life Extension of Nuclear Power Plants," Dak Ridge National Laboratory,. ORNL/TM-10059, September 1986.

24. Letter Report, L. N. Rib, " Summaries of Research Reports Submitted in i

Connection with the Nuclear Plant Aging Research (NPAR) Program," Engineering and Economics Research (EER) Inc., Reston, VA, September 1986.

25. PNL-5722, D. E. Blahnik and R. L. Goodman, " Operating Experience and Aging Assessment of ECCS Pump Room Coolers," Pacific Northwest Laboratories, October 1986.

26. BNL Technical Report A-3270-12-86, R. Fullwood, J. C. Higgins, M. Subudhi, and J. H. Taylor, " Aging and Life Extension Assessment Program'(ALEAP)

Systems Level Plan," Brookhaven National Laboratory, December 1986.

27. NUREG/CR-4819, V. P. Bacanskas, G. C. Roberts, and G. J. Toman, " Aging and Service Wear of Solenoid-0perated Valves Used in Safety Systems of Nuclear Power Plants. Vol. 1: Operating Experience and Failure Identification," Oak Ridge National Laboratory, ORNL/SUB/83-28915/4/V1, March 1987.

28. NUREG/CR-3956, M. R. Dinsel, M. R. Donaldson, and F. T. Soberano, "In Situ Testing of the Shippingport Atomic Power Station Electrical Circuits," Idaho National. Engineering Laboratory, EGG-2443, April 1987.

20 NUREG/CR-4769, W. E. Vesely, " Risk Evaluations of Aging Phenomena: The Linear Aging Reliability Model and Its Extensions," Idaho National Engineering Laboratory, EGG-2476, April 1987. j

30. Technical Integration Review Group for Aging and Life Extension (TIRGALEX), " Plan for Integration of Aging and Life-Extension Activities,"

U.S. Nuclear Regulatory Commission, May 1987. i

31. NUREG/CR-4731, V. N. Shah and P. E. MacDonald, " Residual Life Assessment of Major Ligiit Water Peactor Components," Vol.1, Idaho National Engineering Laboratory, EGG-2469, June 1987.

32. HUREG/CR-471 G. J. Toman, V. P. Bacanskas, T. A. Shook, and C. C. Lodlow, l "An Aging Asse nment of Relay and Circuit Breakers and System Interactions," Brookhaven National Laboratory, Franklin Research Center, Philadelphia, PA, BNL-NUREG-52017, June 1987.

33. NUREG/CR-4928, H. M. Hashemian, K. M. Petersen, T. W. Kerlin, R.

L. Anderson, and K. E. Holbert, " Degradation of Nuclear Plant Temperature Sensors," Analysis and Measurement Services Corporation, Knoxville, TN, June 1987. 34. NUREG/CR-4747, B. M. Meale and D. G. Satterwhite, "An Aging Failure Survey of Light Water Reactor Safety Systems and Components," Voi,1, Idaho National Engineering Laboratory, EGG-2473, July 1987.

35. NUREG/CR-4457, J. L. Edson and J. E. Hardin, " Aging of Class IE Batteries in Safety Systems of Nuclear Power Plants," Idaho National Engineering Laboratory, EGG-2488, July 1987.

1 63

p CHRON0 LOGICAL L1STit!G

36. ' NUREG/CR-4985, M. Subudhi, J. H. Taylor, J. Clinton, C. J, Czajkowski, and J. Weeks,'" Indian Point 2 Reactor Coolant Pump Seal Evaluations," Brookhaven National Laboratory, BNL-NUREG-52095, August 1987.

37. NUREG/CR-4590, K. R.' Hoopingarner, J; W. Vause, D. A. Dingee, and J. f.

Nesbitt, " Aging of Nuclear Station Diesel Generators: Eveluation of Operating and Expert Experience," Vols. I and 2, Pacific. Northwest Laboratories, PNL-5832, August 1987.

38. NUREG-1144, J. P. Vora, " Nuclear Plant Aging Research (EPAR) Program Plan,"

Rev.1, U.S. Nuclear Regulatory Commission, Septembt.r 1987.

39. NUREG/CR-4992, G. C. Roberts, V. P. Bacanskas, and G. J. Toman, " Aging and

' Service Wear of Multistage Switches Used in Safety Systems of Nuclear Power Plants," Vol.11, Oak Ridge National Laboratory, ORNL/SUB/83-28915/5/V1, September 1987.

40. NUREG/CR-5008,.R. D. Meininger and T. J. Weir, " Development of a Testing and Analysis Methodology to Determine the Functional Condition of Solenoid Operated Valves," Pentek, Inc., Coraopolis, PA, September 1987.

41.. NUREG/CR-4692, G. A. Murphy and J. W. Cletcher 11. " Operating Experience-Review of Failures of Power Operated Relief Valves and Block Valves in Nuclear Power Plants," 0ak Ridge National Laboratory, ORNL/N0AC-233, October 1987.

42. NUREG/CR-4967, L. C. Meyer, " Nuclear Plant Aging Research on High Pressure Injection Systems," Idaho National fagineering Laboratory, EGG-2514, November 1987.

43. NUREG/CR-4939, M. Subudhi, W. E. Gunther,.J. H. Taylor, R. Lofaro, K. M.

Skreiner, A. C. Sugarman, and M. W. Sheets, " Improving Motor Reliability in Nuclear Power Plants;" Volume 1: Performance Evaluation and Maintenance Practices; Volume 2: Functional Indicator Tests on a Small Electric Motor Subjected to Accelerated Aging; Volume 3: Failure Analysis and Diagnostic Tests on a Naturally Aged Electric Motor; Brookhaven National Laboratory, BNL/NUREG-52031, November 1987. .44. NUREG/CR-4740, L. C. Meyer, " Nuclear Plant-Aging Research on Reactor Protection Systems," Idaho National Engineering Laboratory, EGG-2467, January 1988.

45. NUREG/CR-4731, V. N. Shah and P. E. MacDonald, " Residual Life Assessment of Major Light Water Reactor Components--0verview," Vol. 2 (Draft), Idaho National Engineering Laboratory, EGG-2469, March 1988.

46. PNL-6287, K. R. Hoopingarner, B. J. Kirkwood, and P. J. Lonzecky, " Study Group Review of Nuclear Service Diesel Generator Testing and Aging Mitigation," Pacific Northwest Laboratories, March 1988.

47. NUREG/CR-5159, M. S. Kalsi, C. L. Horst, and J. K, Wang, " Prediction of Check Valve Performance and Degradation in Nuclear Power Plant Systems," Kalsi Engineering, Inc., Sugar Land, TX, KEI No.1559, May 1988. 64

E l-CHRON0 LOG 1 CAL LISTING- . NUREG/CR-4597, D. M.' Kitch, J. S. Schlonski, P. J. Sowatskey, and W. Y.

48. '

Cesarski.." Aging.and Service Wear of Auxiliary.Feedwater Pumps for PWR lluclear Power Plants. Vol. 2: Aging Assessments and Monitoring fiethod Evaluations," Dak Ridge National Laboratory, ORNL-6282/V2, June 1988.

49. HUREG/CR-4747, B. M. tieale and D. G. Satterwhite, "An Aging Failure Survey ofL Light Water Reactor Safety Systenis and Components," Vol. 2,-Idaho-National Engineering Laboratory, EGG-2473, July 1988.

!n

50. MUREG/CR-5052,' J. C. Higgins, R. Lofaro, M. Subudhi,' R. ' Fullwood, and J. H.

Taylor, " Operating Experience and _ Aging Assessment' of Component Cooling. Water Systems in Pressurized Water Reactors," Brookhaven National Laboratory, BNL-NUREG-52117, July 1988. l

51. NUREG/CR-5053, W.-Shier and M. Subudhi, " Operating. Experience and Aging i

Assessment of Motor Control-Centers, Brookhaven National Laboratory, BNL-NUREG-52118, July 1988.

52. NUREG/CR-5051, W. E. Gunther, R. Lewis, and M. Subudhi,." Detecting and Mitigating Battery Charger and Inverter Aging," Brookhaven National Laboratory, BNL-NUREG-52108, Aug.1988.

53.. NUREG/CR-5141, V. ' P. Bacanskas, G. J. Tcman, and S. P. Carf agno, " Aging and Qualification Research on Solenoid Operated Valves," Franklin Research Center, Norristown, PA, August 1988.

54.* SAND 88-07E4 UC-78, K. T. Gillen and R. L. Clough, " Time-Temperature-Dose Rate Superposition: A Methodology for Predicting Cable Degradation Under Ambient Nuclear Power Plant Aging Conditions," Sandia National Laboratories, August 1988.

55. NUREG/CR-5192, W. E. Gunther, " Testing of a Naturally Aged Nuclear Power Plant Inverter and Battery Charger," Brookhaven National Laboratory, BNL-NUREG-52158, September 1988.

• Sponsored by the Department of Energy (DOE); although concerned with nuclear power plant aging, it is not a part of the NRC NPAR program.

1 I J ) u 65 4 )

NJ.C FOIM 3:,6 UL NUCLE AM RE GULATO:1Y COMMISSION 1 RE PORT NUMBL R 5*c%c2' %C.2,2,'L"' ihE"I'" **

m. nor BIBLIOGRAPHIC DATA SHEET

. (See metrucroons on the reserse) NUREG-1377' 2, Tn tE AND susuitE NRC Research Program on Plant Aging: Listing and Abstracts 3 DATE REPORT PUBLISHED of Reports Issued Through February 1, 1989 l em v i a *. August 1989

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9. 5,PONSO, RING ORGANIZATION - N AME AND ADDR E SS rst mac. repe seme e eao.e". sr cearrerror. oro,pe sec oar..n orr.ne o, nepen. u s sucher meeusererr Comma e,a meu. ooareni Same as above

10. SUPPLEMENT ARY NOTES

11. ABSTRACT framwee er7us The U.S. Nuclear Regulatory Commission is conducting the Nuclear Plant A.ging Research i

(NPAR) Program. This is a comprehensive hardware-oriented program focused on under-standing the aging mechanisms of components and systems in nuclear plants. The NPAR Program also focuses on methods for simulating and monitoring the aging-related degra-dation of these components and systems. This document contains a listing and index of reports generated in the NPAR Program that were issued through February 1,1989, and abstracts of those reports. Each ab-stract 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 per-sonal author, corporate author, and subject.

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