ML20062N289
| ML20062N289 | |
| Person / Time | |
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| Issue date: | 12/31/1993 |
| From: | Vora J NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
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| NUREG-1377, NUREG-1377-R04, NUREG-1377-R4, NUDOCS 9401120296 | |
| Download: ML20062N289 (118) | |
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NUREG-1377 Rev.4 NRC Research Program on Plant Aging: Listing anc 1
Summaries of Reports Issuec Throug:a Se7: ember 1993 U.S. Nuclear Regulatory Commission i
i Office of Nuclear Regulatory Research J. P. Vora i
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1 AVAILABILITY NOTICE i
Availabihty of Reference Materials Cited in NRC Publications Most documents cited in NRC pubhcations will be available from one of the following sources:
1.
The NRC Public Document Room. 2120 L Street, NW., Lower Level, Washington, DC 20555-0001 2.
The Superintendent of Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, DC 20402-9328 3,
The National Technical Information Service, Spongfield, VA 22161 Although the hsting that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.
Referenced documents available for inspection and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC bulletins, circulars, information notices, inspection and investigation notices; licensee event reports; i
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NUREG-1377 Rev.4 NRC Research Program on Plant Aging: Listing and Summaries of Reports Issued Through September 1993 Manuscript Completed: October 1993 Date Published: December 1993 J. P. Vora Division of Engineering Office of Nuclear Regulatory Research U.S. Nu' clear Regulatory Commission Washington, DC 20555-0001
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ABSTRACT
'Ihe U.S. Nuclear Regulatory Commission is conducting the Nuclear Plant Aging Research (NPAR) Program. This is a comprehensive hardware-oriented engineering research program focused on understanding the aging mechanisms of components and systems in nuclear power plants. The NPAR program also focuses on methods for simulating and monitoring the aging-related degradation of these components and systems. In addition,it provides recom-mendations for effective maintenance to manage agmg and for implementation of the re-
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search results in the regulatory process.
This document contains a listing and index of reports generated in the NPAR Program that -
were issued through September 1993 and summaries of those reports. Each summary de-scribes 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.
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iii NUREG-1377
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CONTENTS Page Abstract...............
iii lYeface.............
vii Acronyms and Abbreviations ix Introduction.....
1 Main Citations and Summaries............
3 Personal Author Index.......................................
71 Corporate Author Index..
85 Subject Index.........
91 101 Chronological Listing.............
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PREFACE The Office of Nuclear Regulatory Research of the U.S.
Significant accomplishments have included identifying Nuclear Regulatory Commission (NRC)is conducting major technical safety issues and defining the risk sig-a hardware-oriented engineering research program nificance of major light water reactor components and dealing with the aging of nuclear power plant compo-structures.The Nuclear Plant Aging Research program nents and systems. This program is described in continues to provide the technical bases and regulatory NUREG-1144, Rev. 2, " Nuclear Plant Aging Research guidelines for the license renewal rulemaking, which is (NPAR) Program Plan-Status and Accomplish-considered a top priority for the NRC.
i mentrs" published in June 1991.
Significant progress has been made in defining aging This document contains summarics of NRC-sponsored degradation mechanisms and in developing effective reports that were generated in the NPAR Program, monitoring and surveillance methods for many of 1he Each summary describes the objectives of the research, components and systems identified in NUREG-1144, identifies the contractor and the authors involved, and Rev. 2. These components and systems include motor-outlines significant research results. If the readers of operated valves, check valves, solenoid-operated this document need additional information on a par-valves, electric motors, emergency diesel generators, ticular report and the findings discussed therein, they chargers and inverters, circuit breakers and relays, bat-are encouraged to contact the authors of that report teries, auxiliary feedwater pumps, and reactor protec-directly.
tion systems. Progress has also been made in develop-ing models and approaches to evaluate the relative impact of aging on risk.The phase I research for evalu-This report is updated annually to incorporate summa-ating system-level aging effects based on operating ex-rics of new NPAR reports. Comments are welcome and perience and risk evaluation of the aging phenomena will be considered in developing subsequent revisions has been completed.
of this dottument.
Milton Vagins, Chief Electrical & Mechanical Engineering Ilranch Division of Engineering Office of Nuclear Regulatory Research -
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l ACRONYMS AND ABBREVIATIONS APN auxiliary feedwater ISA Instrument Society of Amenca ALARA as low as reasonably achievable (radia-ISCM inspection, surveillance, and condition tion level) monitoring ALEAP Aging and ljfe Extension Assessment ISI inservice inspection Program KWU Kraftwerk Union Aktiengesellschaft, a ANSI American National Standards Institute German company ASME American Society of Mechanical Engi-LER Licensee Event Report neers LOCA loss-of-coolant accident AUXFP auxiliary feedwater pump LWR light-water-reactor BNL Brookhaven National Laboratory MCC motor control center BOP halance of plant MCSA motor current signature analysis B&W Babcock & Wilcox Co.
MLE maximum likelihood estimate BWR boiling water reactor MOV motor-operated valve CCW component cooling water MOVATS motor-operated valve analysis and test CE Combustion Engineering system CFR Code of Federal Regulations MSLB main steam line break CRD control rod drive NOAC Nuc! car Operations Analysis Center CVN Charpy V-notch NPE Nuclear Ibwer Experience DBE design basis event NPRDS Nuclear Plant Reliability Data System ECCAD electrical circuit characterization and NRC U.S. Nuclear Regulatory Commission diagnostic system NPAR Nucicar Plant Aging Research ECCS emergency core cooling system NSAC Nuclear Safety Analysis Center EDG emergency diesel generator NSSS nuclear steam supply system EPA clectrical penetration assembly PORV power-operated relief valve EPRI Electric Power Research Institute PRA probablistic risk assessment ESFAS engineered safety feature actuation PWR pressurized water reactor system RCP reactor coolant pump FSAR Final Safety Analysis Report RHR residual heat removal GFRS generic flaw response spectra RTD resistance temperature detector GI generic issue RTS reactor trip system GSI generic safety issue RWCU reactor water cleanup IIDR licissdampfreaktor, a decommissioned SCR silicon controlled rectifier German reactor SOV solenoid operated valve IIPCI high-pressure coolant injection SOUG Seismic Qualification Utilities Group HPIS high-pressure injection system SSE safe shutdown earthquake IEEE Institute of Electrical and Electronics TDR time-domain reflectometry Engineers TIRGALEX 'Ibchnical Integration Review Group for INEL.
Idaho National Engineering 1.aboratory Aging and Life Extension INPO Institute of Nuclear Power Operations Tl temporary instruction IFRDS In-Plant Reliability Data Systems Ur ultrasonic testing ix NUREG-1377
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INTRODUCTION 1
'Ihis document is a listing and index of reports related
'Ihe information contained in the reports should be of to the Nuclear Plant Aging Research (NPAR) Program interest to those assessing the aging and reliability of issued through September 1993. The first listing is in nuclear power plant components, including research-alphanumeric order by report number and includes a ers and designers as well as maintenance and opera-1 summary of each report.'fhree indexes are provided to tions personnel.
aid the user in retrieving a specific report: Personal Author index, Corporate Author Index, and Subject Most of the documents eited in this report arc available Index. Finally, there is a listing in chronological order from one of the following sources:
by date of publication.
1.
'Ihe Superintendent of Documents, U.S. Govern-men t Printing Office,1bst Office Box 37082, Wash-Most of the reports contain a description of the compo-ington, DC 2fK)l3-7082.
nents or systems being examined and identify the prin-cipal stressors leading to aging. 'Ihey frequently 2.
The National Technical Information Service, c<mtain an analysis and statistical assessment of failure Springfield, VA 22161.
data obtained from Licensee Event Reports and other i
sources of comp (ment failure data for operating 3.
'The NRC Public Document Room, 2120 L Street nuclear power plants. Current surveillance and moni-NW., Lower Level, Washington, DC. Mailing Ad-toring practices are also reviewed and, when identifi-dress: NRC Public Document Room, Washington, able, recommendations are made for improvements.
DC 20555.
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MAIN CITATIONS AND SUMMARIES He reports listed in this compilation are arranged alphanumerically by report number, with unnum-bered reports preceding the numbered reports. The bibliographic information is followed by a summary of each report.
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UNNUMBERED REPORTS 14 reports are presented in this publication. Thus the results of these studiesare made more readilyavailable Letter Report, M. Subudhi, " Review of Aging-Seismic for rapid survey, directing attention to specific reports Correlation Studies on Nuclear Plant Equipment,"
ofinterest and facilitating the u tilization of research re-Ilrookhaven National laboratory, January 1985.
sults in the regulatory process.
During the last decade, the issue relating to aging-The 14 reports are grouped into three categories:
seismic correlation of nuclear-grade equipment and (1) carly scoping and background studies, including a their components has received special attention by survey of aged power plant facilities, operating experi-both the NRC and the utility industry with the aim of ence reviews of Licensee Event Reports (LERs) to preventing catastrophic failures of aged nucicar power identify aging trends, workshops to obtain experts' plant components during a seismic event. His report opinions, and aging / risk considerations; (2) reports on I
summarizes the work performed by the Seismic Quali-developing a methodology for aging analysis and on fication Utili ies Group (SOUG) based on real carth-evaluation and use of a signature analysis technique t
I quake data, by NUTECII for Sandia National Labora-(MGVATS); and (3) Phase I results of aging research on tories, and by EPRI at Wyle. Based on the above, an nine components, including electric motors, battery outline of the work to be carried out at flNL under the chargers / inverters, electrical cables, pressure transmit-l NPAR scope relating to identifying the aged compo-ters, diesel generators, motor-operated valves, check l
nents sensitive to seismic loadings is provided.
valves, auxiliary feedwater pumps, and snubbers. Each EQE, Inc., sponsored by the Seismic Qualification summary has four sections: Hackground, Summary, Re-l Utilities Group has gathered a comparative data base suits / Findings, and Utilization of Research Results m on the performance of equipment in five fossil-fueled the Regulatory Process.
plants consisting of 24 units and a high-voltage DC-to-This report is considered a *1iving" document. That -
AC converter station. Dese plants have experienced is, research results and summaries of additional se-four damaging California carthquakes of Richter Mag-lected reports may be added periodically.
nitudes 5.1 to 6.6. Peak horizontal ground accelerations I
(PGA) of these carth uakes ranged between 0.2 g and Techm, cal Integration Review Group for Aging and Life 4
0.5 g. The actual earthquake-m. duced effects on equip-Extension (TIRGALEX),, Plan for Integration of Aging and Life-Extension Activitics," U.S. Nuclear ment were compared with equipment qualification data from three nuclear plants.
Regulatory Commission, May 1987.
The 'Ibchnical Integration Review Group for Aging He objective of Ihe pilot program was to determine I
the feasibility of establishing criteria for assessing the
~ and Life Extension (TIRGALEX) was established to facilitate the planning and integration of NRC activi-l-
seismic adequacy of equipment in nuclear powerplants ties related to reactor aging and life extension. He in-i based on evaluation and application of data to be ac-itial objectives of TIRG ALEX were to identify techei-l quired on the characteristics and seismic performance cal safety and regulatory policy issues related to reactor of equipment in nonnuclear power facilities that have aging and life extension and to develop a plan to inte-
- been subjected to strong-motion earthquakes. Applica-grate NRC and external activities to resolve the issues.
tion of the criteria would provide a valid basis for as-This report contains the plan developed by TIR-sessing the need for subsequent qualification efforts in GALEX, which crmsists of the following main ele-the nuclear industry and for defining the extent of the ments:
effort.
- 1. A summary and discussion of the major techni-Letter Repost L N. Rib," Summaries of Research cal safety and regulatory policy issues associated Reports Submitted in Connection with the Nuclear w th reactor aging and life extension.
Plant Aging Research (NPAR) Program."
- 2. An overview of ongoing programs and activities Engineering and Eccmomics Research, Inc. (EER),
related to reactor aging and life extension, in-Reston, VA, September,1986, ciuding both NRC and external programs and The results of Phase I efforts in the NRC NPAR activities.
program for selected electrical and mechanical compo-
- 3. Recommendations for future NRC actions to nents since 1984 have been published.Tb help maintain address reactor aging and life extension in a cognizance of this wealth of information, summaries of timely, efficient, and well-integrated manner, e
Main Citations and Summaries K.IL lhmpingarner and Elt. Zaloudek, " Safety NUMBERED REPORTS 1mplications of Diesci Generator Aging," Pacific Northwest Laboratory, Nuc/ car Safety. 31:484-489, IINL Techm, cal Report A-3270-11-26-84, II. Miller, October. December 1990-Scopmg Test on Contamment Purge and Vent Valve Scal Material," llrookhaven National Labora-The emergency diesel generators in a nuclear tory, December 1984.
power plant have an important safety function m that they supply electric power for emergency core cooling Degradation of shaft-scal material used in contain-and related emergency needs in the event of a loss of ment purge and vent butter 0y valves may initiate valve offsite power.'lypically, a plant has two redundant dic, seal leakage thus breaching containment. A scoping sel gen erators of 3,(XX) to 8,(XX) kW (5,(XX) to l0,(XX) hp).
test was performed to gatherinformation on the behav-ior of the seat material (ethylene propylene) when ex-Diesel generators have been identified as compo-posed to severe accident conditions (i.e., steam at nents with significant safety importance, and their op-350'F/120 psig and 400*F/232 psig). Three separate crating history has shown performance degradation test sequences were performed with the test assembly and loss of reliability as a result of wear and aging. Con-monitored for leakage. The results of these tests re-sequently, emergency dicscl generators are included in Vcaled no seal leakage; however, shaft scal degrada-tion was evident.
the NRC NPAR program, the overall objectives of which are to find the causes of aging-related degrada-For two test sequences, the prescribed procedure tion and to recommend methods for managing it. Ily was revised to include modified temperature profiles ameliorating wear-and performance-related degrada-and scal-testing sequences.
tion, the reliability of the emergency dicsci generators Removal and inspection of ',nc valve seat following c:m he improved and the risks associated with loss-of-some test sequences revealed minor remolding of the offsite power events reduced.
seat material at the disc /txidy interface with no de.
formities noted. Approximately one week later, cracks The NPAR Diesel Generator Study c4msisted of devchiped in the scat. The cracks were in an area that two phases. Phase I used plant operating experience, would be compressed by the retaining ring and in no in-data, exPcrt oP nion, and statistical methods to produce stance affected the scaling integrity of the valve.
i a data base related to aging failures, their causes, and corrective actions. Phase II included the development The results of the scoping test revealed no shaft-of a more appropriate testing and aging management seal leakage. The seal degradation and cracking was program that could enhance the availability and reli.
visually evident in the compressed retaining portion of the scat. Ilowever, the result should not be construed ability of these dicsci generators.
as representing the entire ethylene propylene family The purpose of this article is to discuss the principal (clastomers prepared from ethylene and propylene causes of wear-and aging-related degradation of emer-monomers). Varying the relationship of these mono-mers affects the characteristics of the clastomer and its gency dicsci generators and th e effects on their reliabil.
ity and availability and to describe methods by which ability to withstand environmental conditions. It should also be noted that all mechanisms by which rubber de-such degradation can be avoided or detected before it teriorates with time are attnbutable to environmentgd becomes a nuclear safety concern. Operational infor, conditions. The Parker Scal Company states that it is mation assembled on component and system failures envinmment, not age, that is significant to seal life, and their causes was wiewed to identify the important both m storage and m actual semcc.
agmg and degrads. ion factors for diesel generators.
One important factor contributing to wear and degra-IINL Technical Repo, t A-3270-11-85, J. I1. 'Ih lor, r
dation has been 1he fast starting and loading test procc-M. Subudhi, J. Iligms, J. C urren, M. Reic,
dure called for by Regulatory Guide 1.108 and the NRL, F. Cifuentes, and I. Nehring, "Scismic Endurance Standard Plant'11 chnical Specifications. A new regula-Tests of Naturally Aged Small Electric Motors,"
tory approach was recommended to develop a more Ilrookhaven National Laboratory, November 1985.
balanced aging management program that includes (1)
Two naturally aged 10-1IP clectric motors were ob-slow-start testing during which im portant operating pa-tained from an older nuclear power plant that is ready rameters are monitored,(2) analysis of data trends,(3) for decommissioning.The motors were utilized to drive training, and (4) maintenance. This approach should fan cooler units in an outside environment for 12 years.
improve safety by enabling the timely identification of These motors were first tested for their dynamic char-aging-related degradation that could lead to diesel gen-acteristics.They both were subjected to seismic cxcita-crator failures so that maintenance could be performed tion with generic floor response spectia (GFRS) that in time to prevent actual failure.
encompass Safe Sh utdown Earthquakc (SSE) accelera-NUREG-1377 4
hiain Citations and Summaries tions applicable to most nuclear plants in the United and to investigate other California power plants. Such States. The tests showed that the first fundamental fre-research will provide the maximum amount of actual quency is well above the rigid range of an earthquake experience data to address the aging-scismic relation-frequency. Seismic testing was performed with a motor ship in a practical manner. Lessons learned from a both unloaded and loaded by an attached hydraulic review of these data can be used as input to develop pump that served as a dynamometer. Significant oper-practical maintenance and operating procedures to en-ating parameters such as current, voltage, and tem-hance safety and improve plant reliability, perature were monitored before, during, and after seis-mic loading, and no noticeable differences were ob-IINL Technical Repor1 A-3270-3-86, A., C. Sugarman, served. Existing deficiencies in one of the motor bear.
hl. W. Sheets, and M. Subudhi,"'Ibstmg Program for ings and in the stator winding were not affected or mag-the Monitoring of Degradation in a C.ontmuous nified by the seismic excitations.
1)uty 460 Volt Class "11",10-IIP Electric hiotor,"
ilrookhaven National Laboratory, hiarch 1986.
- lhis report describes the test plan, includes details This report presents an evaluation of potential of the procedure, and presents findings of the seismic maintenance techniques for monitoring age-related tests and operating / static tests on both motors.
degradation in a continuous-duty 460-volt, Class II, j
This testing was part of the NRC NPAR program, 10-IIP clectric motor. The program follows up the I
and its results are an integral part of the llrookhaven analyses and recommendations outlined in the draft National Laboratory's overall aging assessment of mo-of NUREG/CR-4156, " Operating Experience and I
tors, which was published as NUREG/CR-4156.
Aging-Scismic Assessment of Electric hiotors," by h1. Subudhi et al. In this study, the following stressors llNL Technical Report A-3270-12-85, M. hl. Silver, on dielectrics are evaluated: temperature, frequent R. Vasudevan, and hi. Subudhi, " Pilot Assessment:
starts, overload, and high voltage gradient.
Impact of Aging on the Scismic Performance of Sciccted Equipment 'lypes,"llrookhaven National In general, the motor tests are conducted by con-Laboratory, )cccmber 1985.
tinuously reversing motor direction for five hours, fol-lowed by a half hour with the motor running under no
'Ihe NRC has initiated a number of specific re-load in a single direction and a half hour with the motor scarch programs in support of the NPAR program, to turned off and stationary. During the half hour of run-better understand the impact of equipment aging on ning under no load, measurements of bearing vibration plant safety and to recommend realistic operating and and movement of stator end turns (measured with ac-maintenance procedures to improve plant availability celerometers epoxied to the end turns) were made.
and enhance safety. This pilot study was performed to Also, a number of insulation tests were conducted. 'lo investigate the feasibility of using plant experience data accelerate the degradation of the test motor (including l.
to assess the relationship between equipment aging insulation, bearings, and lubrication), a plug reverse
.l and seismic performance capacity.
test was performed.
After a brief review of available information on
'the results of the exploratory testing program re-plant experience at many California sites for content vealed which insulation and bearing tests can best be and quality, data rclated to performance. maintenance, used in utilitics' procedures for preventive mainte-and failure history were collected for a sample set of nance, corrective maintenance, and surveillance for equipment types. This pilot study selected the equip-safety-related motors.
ment types for investigation from the highest priority
'Ihis testing is meant for motors rated for contin-group specified in a previous NPAR study. The equip-uous use. A separate test plan will be required for ment types studied were electric motors, motor-intermittent-duty motors (e.g., valve actuator motors);
operated valves, relays, circuit breakers, and motor such a plan should include typical valve actuator tests control centers.
such as the open/close cycling test and the insulation The acquired equipment data consisted of installa-tests discussed in Section 4.0 of the presently reported tion date, chronological listing of preventive and cor-program.
rective maintenance activitics, failed state and cause of failure, earthquake data (i.e., free-field acceleration, IINL Technical Report A-3270-12-86, R. Fullwood, Richter magnitude, date), and equipment status before J. C. liiggins, M. Subudhi, and J. It 'lhylor, " Aging and after the earthquake.
and Life t.xtension Assessment Program (ALEAP)
Systems Level Plan." Ilrookhaven Nationa! Labora-The pilot study was successful in demonstrating that tory, December 1986.
cxperience data can be extracted and u tilized to address This system level program plan for ALEAP the relationship between seismic performance capacity presents and explains the llNL structured approach to and aging of plant equipment. It is strongly recom-assessing the effects of thc aging of nuclear power plant mended that future research be conducted to acquire components and systems on safe operation and the experience data for other important equipment types extension of plant operation beyond the originally 1
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planned plant life. It should be noted that this plan is areas of a system. Recommendations are made forim-prepared in a generic fashion and could be used by any-provements in pertinent regulatory guides, industry one for a system assessment.
standards, etc. This program plan delineates the goals The ptan discusses the criteria for prioritizing plant, and major tasks to be completed in each phase. The system, and component selection for analysis to deter-current version of the program plan is considered to be mine the effects of aging.The use of failure modes and a draft and will be revised and updated as the first few -
effects analysis in conjunction with the results of natu-system assessments are completed using this method-ral anel accelerated aging tests are discussed as means I gy. This will produce a final proven methodology J
that car died to the remaining systems.
for understanding and predicting the phenomena.The effects of aging on the failure rates of components are BNL 'P ' '
...eport A 3270R-2-M A. Fresco and being determmed principally from plant data with 14 ',uoudhi, " Aging Effects of Implar ts," Brook-i: ant Halance of physical and phenomenological models used for inter-Plant Systems in Nuclear Power I polation in areas not included in the data base.These haven National Laboratory, February 1990.
results will be integrated with a plant risk model to be In recent years, balance of plant (LOP) systems used m addressing the question "how old is old have become major causes of plant transie:Js, e.g., the enough.
June 9,1985 loss-of-all-feedwater event at tia Davis.
The NRC NPAR program has completed several Hesse Nuclear Power Plant, and have reccied in-component-level aging assessments that include the creased attention from the nuclear industry anf the identification of dominant component failure modes Nuclear Regulatory Commission (NRC). 'Ihis intesn based on plant operating experience.1he studies pro-report describes the activities to date in a study of HOP vide recommendations for monitoring as well as miti.
systems by B rookhaven National Laboratory in support gating age related component degradations.
of the NRC Nuclear Plant Aging Research (NPAR)
Utilizing results from the component-level studies P' 8#"*'.Ihc initial phase of the study provides pre-limm ry mdications of those HOP systems that may and work performed by other NRC contractors for sys-warrant a detailed study of agmg effects. An approach tem-data assessment and system-level risk analysis, this f r accomplishmg the overall objective of identifymg program evaluates the impact of comp (ment failures on the effects of aging in these HOP systems on nuclear plant system performance. The study performs in-plant safety is suggested, depth system-level failure-data reviews, evaluates cur-rent industry practices for system maintenance, testing, This study on BOP systems covers all non-safety-l and operation and probabilistic risk assessment (PRA) related systems except for those associated with the techniques to understand and to predict the impact of nuclear steam supply system (NSSS). Some non-safety-aging on system availability. Recommendations for im.
related systems in the NSSS are being studied in other proving the system performance by means of degrada.
parts of the NPAR program.
tion monitoring and timely preventive and corrective From the results of the study,it was concluded that maintenance are addressed.
This project the frequency of unplanned reactor. trips has often integrates its products with the HNL programs for op-been cited as an indicator of safety performance and crational safety reliability research and performance that the most frequent contributors to unplanned reac-indicators.
tor trips caused by HOP systems are the power conver-The first phase of this research effort concentrates sion systems, i.e., the feedwater, main turbine, main on understanding various system designs from plant electric generator, main steam (usually the steam by-safety analysis reports, evaluating failure data from pass to the mam condenser), and condensate systems.
plant operating experience data bases, applying PRA Other HOP systems contributing to unplanned reactor analyses, assessing industry-wide surveillance and trips are support systems such as the electne distribu-maintenance practices, and identifying system func.
ti n system and, less frequently, the circulating water, tional indicators that are used to monitor the rate of service / instrument air, fire protection, and the heating, system degradation resulting from aging and service ventilation, and air conditioning systems.1he electric wear.The program separates failures on demand from distribution system includes 120-V AC power distribu-time-dependent failures. It categorizes age-related tion systems, the switchyard,large plant load users, the failures separately from random and design-type fail.
DC power system, and control centers. At the compo-nent level, the feedwater regulating valves, the ures. It produces results useful for the resolution of pertinent unresolved safety issues and for review and turbine-driven feedwater pumps, and the main turbine inspection of operating NPPs. The second phase, if electro-hydraulic control subsystem are frequent con-authorized and performed, will provide recommenda-tributors. Failures in the main electric generators are tions for improving system performance through en.
also important as potential causes of reactor trips.
hanced maintenance practices and reliability monitor-These results are substantially in agreement with ing, which will be focused on the most risk-sensitive the results of an alternative approach in which NUREG-1377 6
Main Citations and Summaries
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important HOP systems were categorized based on in-The in-depth performance-based inspections have sights from probabilistic risk assessments.
explored such areas as overall plant performance re-Preliminary recommendations are:
lated to maintenance, management support of mainte-nance, and the implementation of the mamtenance
- 1. The frequency of unplanned reactor trips program. Incorporated in the inspection criteria estab-l should be considered the most important indica-Ished for these general areas are specific attributes l
tor of current or potential near-term safety that are relevant to the understanding and managing of l
problems.
aging. The related activities include root cause evalu.
(
- 2. IlOP systems that significantly contribute to un-ation of equipment failure, trending of fah:e data, im-planned reactor trips should be included in the plementation of equipment qualification programs, NPAR program.
control of spare parts, evaluation of test results (includ-ing postmaintenance testing), and implementation of The next phase of the program will focus on several condition-monitoring techniques.,
of the oldest nuclear plants. Licensee Event Reports the maintenance team inspection reports were re-involving unplanned reactor trips from the beginning of viewed with the following objectives in mind:
commercial operation to the present will be reviewed 1.
Assesstheevaluationsof thoseportionsof the l
to determine if there is an increasing frequency of un.
maintenance program determined to be im-l' planned trips caused by the identified BOP systems.
portant for understanding and managing ag-This group of plants will include Monticello and Yan.
mg.
kee Rowe, the pilot plans for the life extension study.
2.
Evaluate the weaknesses noted in utility main-Next, some plants al intermediate age and then tenance programs that could affect the ability some of the youngest rilants will be examined in the of the plant to manage aging.
same manner. 'lhese three age groups will be com.
3.
Determine the types of preventive mainte.
pared and analyzed. The ultimate goal is to determine nance activities and condition-monitoring whether aging of individual BOP components is a sig.
techniques that address plant aging.
nificant factor affecting nuc! car plant safety.
At this time,47 inspections have been completed, and 24 more are scheduled for the remainder of 1990.
HNL Technical Report TR-3270-6-90, W. Gunther, NRC has compiled the findings from these inspections -
" Maintenance Team,Ingection Results: Insights n a computerized data base that assisted in identifying Related to Plant A g, Brookhaven National mtmy, June 1 plants where the NRC inspection teams had concerns-about how well the maintenance program accounted NRC is performing maintenance team inspections for the effects of aging. 'Ibn reports contained specific in accordance with the NRC temporary instruction (TI) findings that the utility maintenance programs do not 2515/97 entitled " Maintenance Inspection" to deter-
. address aging. 'Ihese findings are tabulated and sum-mine the effectiveness of the totalintegrated mainte-marized in this report.
nance process in nuclear power plants. As specified in It should be noted that nine of the above ten reports theTI,"the goal of the inspection effort is to emphastze concluded that the overa!! maintenance programs were the use of plant experience, test and surveillance data, adequate, satisfactory, or good.The guidance and crite-recent component failures, [and) Probabilistic Risk As-ria provided to the NRC maintenance inspection team sessment (PRA) msights-." to identify strengths and allow a maintenance program to be judged good if it ef-weaknesses. Two volumes of inspection guidance sup-fcctively manages current prob! cms even though it may plement th e TI and direct the mspectors to evaluate the not effectively manage the long-term aging effects on mamtenance/agmg relation. For example, the inspec-structures, systems, and components.
tor is directed to determine the extent to which man-agement is aware of plant aging. The inspector is also llNL Technical Report TR-3270-9-90, E. Grove and W.
Gunther,"An Operatmnal Assessme cock & Wilcox and Combustion Engm,nt of.the Hab-required to evaluate the involvement of corporate eermg Con-management in maintenance activitics that address trol Rod Drives," Brookhaven National Laboratory, plant agmg.
September 1990.
More important than the cited guidelines is the ex-Control rods and the associated drive and control pected assessment of maintenance program activities systems, which ensure safe and reliable operation, are that reflect on the ability of the plant technical st< to essential components of nuclear reactors. This report manage aging. Predictive and preventive ma:menance describes an aging assessment of the Babcock & Wilcox programs must include condition monitoring, trending, (B&W) and Combustion Engineering (CE) control rod
<md recordkeeping in order to competently manage the drive (CRD) systems performed as a part of the NRC cffects of aging on a timely basis. Inadequacies found in NPAR program. Emphasis was placed on the specific these areas are indicative of the inability of the plant components of the systems that may be susceptible to staff to properly address and treet aging.
aging-related degradation. This study along with a 7
Main Citations and Summaries failure modes and cf fccts analysis and a detailed review the first 3 years, increases faster in the later years, of utility maintenance practices and procedures will reaching approximately the same value as the degrada-complete Phase I of the aging assessment of these CRD tion rate at the end of 10 years of operation.'Ihisbehav-ior indicates the ineffectiveness of maintenance in pre-systems.
Of the fifteen plants that use CE CRD systems, venting degradations from transforming into failures as thirteen use a magnetic jack c(mtrol element drive the air compressors age, mechanism, and two use a rack and pinion drive. In all Another important application of degradation mod-eight il&W plants, the ctmtrol rods are driven by a cling is to predict the age-related failure rate from the roller nut /lcadscrew drive mechanism. All CE reactors degradation rate. Experience shows that time-had basically the same CE logic, c(mtrol, and rod posi-dependent failures generally pass through a degrada-tion systems; all 13&W reactors had basically the same tion state first. An observed increased aging contribu-Il&W logic, control, and rod position systems.
tion in the degradation rate can signal an expected Commercially available operating experience data increase in aging-related failure rates. A simple linear bases were reviewed to identify failed components and relationship between these two parameters is studied, the resultant effects on plant operation for the considering any delayed effect that degradations may 1980-1989 time period. Age-related failures that re.
have on failures. An example of an application of the sulted in significant plant events, including dropped data on RiiR pumps shows a time lag of 2 years for rods, power reduction, and shutdown for 13&W and CE degradations to affect the occurrence of failures.
control rod drive systems were identified. Susceptibility For additional applications, extensions of degrada-of the system to such external influences as mainte-tion modeling are presented.The extended models the nance errors and the operating environment was also authors are developing will explicitly show the reliabil-shown.
ity effects of different maintenance and test intervals, emnt mMnknance aM W chendes, anMm llNL Technical Report A-3270 6-21-91, E lisu, W. E.
cnt repair times. Thus, the extended model will allow Vescly, E. Grove. M. Subudhi, and P. K. Samanta, the user to evaluate the reliability effects of different
" Degradation Modeling: Extensions and Applica-tions," lirookhaven National Laboratory, June maintenance programs.
1991. Available from the NRC Public Document EGG-SSRIL-8972, C,L Atwood,"Estimatinfdaho llazard Functions for Repairable Components,"
R06*'
Component degradation modeling includes model-National Engincenng Laboratory, May 1990.
ing of occurrences of component degradations and This tutorial report, applying known formulas and analyses of these ocurrences to understand the degra-tools in a way suitable for risk assessment, presents a dation process and its implications, one of them being parametric framework for performing statistical infer-that degradation rate senes as a precursor of the fail-ence on a hazard function based upon such data of re-ure rate.'lhe degradation modeling discussed in the re-pairable comp (ments as might be obtained from field port focuses on the analysis of times of degradation and experience rather than from laboratory tests. This failure occurrences in order to understand the aging framework encompasses many possible forms of the degradation of components. This report reflects the hazard function, three of w hich are considered in some previous llNL work related to the same basic concepts detail. The theory is neatest and the asymptotic ap-and the mathematical development of a simple degra-proximations are most successful when the hazard dation model. Using the degradation modeling meth-function-for a set of identical components-has the odology, failure data on residual heat removal (RIIR) parametric form of a density in the exponential func-pumps and service water (SW) pumps werc analyzed to tion family. The parameters are estimated based on detect indications of aging and to infer the effective-sequences of failure times when the comp (ments are ness of maintenance in preventing age-related degra-restored to service immediately after each failure. In dations from transforming to failures. In this report, certain circumstances, the distribution of the failure further applications and extensions of degradation counts does not depend on tne parameter that deter-modeling are discussed-mines the shape of the hazard function; this suggests Additional applications of degradation modeling natural tools for diagnostic checks involving the indi-are carried out for air compressors, which are continu-vidual parameters. The results presented include for-ously operating components. It was demonstrated that mulas for maximum likelihood estimates (MLEs), tests the analysis of degradation occurrences is useful in and confidence regions, and asymptotic distributions.
understanding the aging ~ process and the role of The confidence regions for the parameters are then maintenance in that process. For the air compressors, translated into a ccmfidence band for the hazard func-the failure rate and degradation rate show an early de-tion. For the three examples ccmsidered in detail, a creasing trend followed by a significtmt increasing table displays all the building blocks needed to program trend that indicates effects of aging.The failurc rate, the formulas on a computer; this tab!c includes asymp-which is significantly lower than the degradation rate in totic approximations when they are.necessary to NUREG-1377 8
3 Main Citations and Summaries j
l maintain numerical accuracy. Diagnostic checks on the ineluding txnh the user's commands and the program's model assumptions are outlined.
responses. Virtually all of the PilAZE commands are The report gives an example of an analysis of real illustrated, and the resulting output is presented.
data. In this example, the methtxis applied are unable EGG-SSRE-9777, J. C. Watkins, R. Steele, Jr., and K.
to discriminate among an exponential hazard function, G. DeWall,
- Isolation Valve Assessment (IVA) a linear hazard function, and a Weibull hazard func.
Software Version 3.10, User's Manual," Idaho Na-tion. The MLE for the two parameters appears to have tional Engineering Laboratory, June 199L Available approximately a bivariate noimal distribution under from the NRC Public Document Room.
the expmential or Weibull hazard modelbut not under The Isolation Valve Assessment (IVA) software is a the linear hazard model. If the analysis using approxi.
PC-based computer program developed to assess the l
mate normality is carried out in any case, t he results ap.
performance capabilities of a 5-degree flexwedge pear similar for all three models. If some model is pre.
motor-operated gate valve with a Limitorque operator ferred for theoretical or other reasons, this report indi-in the closing direction.The software provides easy-to-cates a way to use it.
use data entry screens to input system design basis con-ditions and selected valve and motor-operator parame-EGG-SSRE-9017, C.L Atwo<x1, " User's Guide to ters. Context-sensitive help information is available to PHAZE.,a Computer Program for l'arametric clarify the required input and to provide typical input llazard I unction Estimation, Idaho National values in the event actual values for the valve-motor-Engineering Laboratory, July 1990.
operator unit are not available.
%c program PilAZE (for Parametric IIAZard The software enables the calculation and display of I, unction Estimation) performs statistical inference on the stem thrust, operator torque, and motor torque re-a hazard function (also called a failure rate or intensity quired to operate the valve using both the standard in-function) based on reported failure times of compo' dustry equation and the correlation developed by the nents that are repaired and restored to service. Three Idaho National Engineering Laboratory (INEL).nese parametric models are allowed: the exponential, lin-calculations (of stem thrust, operator torque, and mo-car, and Weibull hazard models. The mference m-tor torque) pertain to unit capability of delivery at ciudes estimation (maximum likelihood estimators and nominal, minimum, and maximum voltage conditions.
c(mfidence regions) of the parameters and of the haz-ne results obtained are then displayed to aid the user ard function itself, testmg of such hypotheses as m-in determining whether the motor-operated valve creasing failure rate, and checking of the model as-(MOV)is capable of functioning at design basis condi-sumptions under a choice of parametric models.
tions.
Since the approach's concern is the failure behavior A series of four graphs can also be displayed to aid of components, these failures are assumed to be gov-the userin evaluating the functionality of an MOV.nc crned by a Poisson process, with the time typically first graph compares the required versus available stem measured from the compment's installation. It is fur-thrust as a function of stem factor for tx)th nominal and ther assumed that, when a compment fails, either it is degraded voltage conditions. The second graph pro-immediately repaired and placed back in service or it is vides the same information as a function of stem to replaced by a new component. Failures of distinct com-stem nut coefficient of friction. The third graph dis-ponents are assumed to be independent.Thus the data plays the conversion of operator torque to stem thrust to be analyzed consist of sequences of failure times of for a series of stem to stem nut coefficients of friction-similar independent compments.
He fourth graph relates the stem thrust to torque This user's guide sketches only enough of the the-switch settings for a number of stem to stem nut coeffi.
ory to permit Pil AZE to be used; a full presentation of cients of friction.
the theory is given in a companion report. A typical
%c software cim also be used to assess low flow / low Pil AZE application is described. nis consists of an in-differential pressure diagnostic test results to deter-itial exploratory phase, in which the various model as-mine whether the response of a valve is typical of the sumptions are checked, and a final estimation phase, in valves tested by the INE L If typicality is confirmed, the which the maximum likeliho(x1 estimator and a confi-INEL correlation can then be used to assess the re-dence inteival are found for the hazard function at sponse of the valve at design basis conditions. A single times ofinterest.The format of a data file is given with graph is needed to assist in this assessment. This graph, examples. PIIAZE is an interactive command-based available on the software, relates the evaluation of the program; all the PIIAZE commands are therefore low flow / low differential pressure diagnostic test data I
listed and explained.
to the results of a similar evaluation of the valves tested ne program has been verified and validated, and by INEL this work is summarized. Finally, some of the technical The IVA software was developed by the INEL un-details of interest to statisticians and programmers are der contract to the Office of Nuclear Regulatory Re-l given. The appendix shows an entire PilAZE session, scarch of the NRC.%c software has been verified and L
~
l
Main Citations and Summaries validated in accordance with the requirements of the valve manufacturer was tested; however, among the six American Society of Mechanical Engineers quality as-valves with seven internal configurations, most of the surance requirements for nuclear facility applications.
design features (except the double disc design) of most
'Ihe NRC has made the software available to the public of the wedge gate valves used to formulate Generic Is-and it may be obtained through INEL.
sue 87, " Failure of IIPCI Steam Line Without Isola-tion"(GI-87, September 1985), were tested.
EGG-SSRE-9926, R. Steele, Jr., J. C. Watkins, and K.
G. DeWall, " Evaluation of EPRI Draft Report
.Ihc typicality of the hardware used m.,
the INEL NP-7065-Review of NRC/INEL Gate Valve Test testing and the actual test conditions provided repre-Program," Idaho National Engineering 1.aboratory, sentative GI-87 operating conditions for evaluating November 1991. Available from the NRC Public representative GI-87 valves. Although the EPRI Document Room.
report questioned whether the test conditions were ap-(
This report documents an INEL evaluation of an plicable to GI-87 valves, the research tests performed EPRI draft report that was critical of the NRC/INEL sufficiently simulated the worst case conditions and gate valve test program and was being used by some provided a correlation that can be used once the design utilities to discount the applicability of the INEL gate basis conditions for a valve are known and the valve's valve test dataflhe authors' purpose was to resolve the predictable behavior established.
comments of the review team sp(msored by EPRI and For a predictable valve, that portion of the closure provide strong technical bases for areas of agreement stroke that occurs just after flow isolation, when the p
and disagreement between the INEL and EPRI posi-disc is riding fully on the seats, is very well defined on a L
tions. Areas of disagreement between EPRI and the stem force versus time trace.The sliding friction ordisc
(
NRC/INEL include predictable versus nonpredictable factor used to size a valve must be based on a correct i
valve performance, motor-operator sizing equatiors, calculation of this point. Determination of this point typicality of the test hardware, applicability of the and the corresponding final stem force will help utili-Phase I and the Phase II (work schedules) test condi-ties evaluate performance margins or establish that the tions, assessment of valve response as indicated in the valve is representative of the valves tested by the INEL measured stem force, and the validity of NRC docu-for predictable extrapolat;on purposes.
ments issued relating to valve testing. 'lhe INEL posi-In conclusion, the NRC documents reviewed in the tions as described in the report are summarized below.
EPRI report have some omissions in the description of The most important questions raised by the INEL circumstances, but after two years of additional analy-testing was how to determine whether or not a valve ses, the conclusions presented in the documents are will perform predictably. The EPRI report suggested still valid.
that valves with different disc designs might perform EGG-SSRE-10039, T.H. Hunt and M.E. Nitzel, "An differently from those tested m the NRC/INEL pro-Evaluation of the Effects of Valve Body Erosion on gram. The results from INEL, EPRI-Marshall facility, Motor-Operated Valve Operability," Idaho National and European testing have all shown that a large num-Enginecrmg Laboratory, May 1993.
ber of the valves tested performed unpredictably. Test Engineers at the Idaho National Engineering: abo-expenence mdicates that mternal damage occurs as the ratory evaluated the effects of crosion-induced vanc body-guide to disc-guide clearances increase. The wall thinning on motor-operated valve operability.The threshold level where current valve designs begin to authors reviewed reports that identified the extent and perform unpredictably needs to be deterrr.ined and ac~
k3 cation of crosion damage in nuclear plant valves and counted for.
chose a globe valve with severe erosion damage to as-The INEL analyses of the Phase I and Phase 11 test sess the potential for loss of operability. They devel-results were not complete when the Ct'R1 review was oped a finite element morJel of the selected valve and performed. At that time, INEL was trying to analyze performed a structural analysis with valve closing the test results using standard industry equations. Only forces to analyze the effects of the erosion on structural after INEL relinquished its efforts to fit test data into integrity. The results indicate that sufficient margin to the standard industry equation could the apparent ran-yield stress remained. Therefore, crosion-related wall domness in the results be correlated with fluid subcool-thinning is not likely to create an operability probicm ing and pressure effects. At that point, the conclusion for motor-operated valves.
was drawn that, because the disc factor includes other 3.ISTIR 4485, E D. Martzloff and A. G. Perr terms along with sliding friction, the mdustry's tradt-
.' Annotated llibliography: Diagnostic Me ods and tional stem force equation used to size motor operators Measurement Approaches'Ib Detcet Incipient is incornplete, regardless of the disc area term or the Defects Due to Aging of Cables." National Institute disc factor used.
of Standards and 'lechnology, July 1991.
INEL selected 5-degree flerwedge gate valves for This annotated bibliography has been prepared to testing based on surveys of equipment installed in document the literature search conducted as an initial nuclear service. Not every variation produced by every task for the project of assessing existing test methods NUREG-1377 10
Main Citations and Summaries for detection of incipient faults in nuclear power plant applied to low-voltage cables. Previous investigations cables. The combined listing presented in this report generally involved cables rated SkV or higher, while can help future work that may be performed by other the objective of this program focused on the lowervolt-
. organizations involved in similar studies. As a further ages associated with the safety systems of nuclear aid to researchers, the references provided in each of power plants, the papers included in this bibliography have been con-The system implemented for this demonstration solidated in a citations list. A hard copy of each paper was based on commercially available signal analysis included in this bibliography has been retained at NIST.
hardware and software packages, customized for the specific purposes of the project.The test specimensin-This report is organized in three sections:
cluded several cables of the type found in nuclear
- Alphabeticallisting of all the reviewed docments; power plants, with artificial defects introduced at vari-ous points of the cable.The final demonstration of the
- Review of authored papers; system included a cable with several defects inflicted at
- Consolidated citations.
locations kept undisclosed to the test operator. Using a The initial search covered the 1970-1986 period, combination of the standard signal analysis built into-combining several data bases and incidental referrals to the system and experience-based detailed analysis of reports with limited circulation.The motivation of this selected portions of the data base, the operator was search, at the outset of the project, was to identify any able to identify the existence and k) cation of all these and all test methods reported in the literature that undisclosed defects.
might be applied, refined, or developed into a test These results indicate that, indeed, partial dis-method for in situ assessment of the cable condition.
charge analysis is capable of detecting incipient defects An additional set of papers was reviewed, covering the in low-voltage cables. There are, however, technical 1986-1990 period. Each review is shown on a separate and nontechnical limitations that need further sheet giving retrieval information, author's abstract, exploration before this method can be accepted in the table of contents of a reviewer's summary,identifica-industry.
tion of the technology involved, brief remarks on the contents and applicability to the subject of in situ test
,lSTIR 4787, E I. Mopsik, "The Use of Time-Domain Dielectric Spectroscopy 7b Evaluate the Lifetime of.
methods, and identification,if any, of the criteria and Nuclear Power Station Cables," National Institute test methods used to characterize aging or residual life of Standards and Technology, April 1992.
of the msulation.
'this report describes the work that has been under-Since these remarks are made from the point of taken at NIST to see if the method of Time-Domain view of applicability to the limited subject of assessing Diclectric Spectroscopy can evaluate the aging in reac-l in situ testing, they should not be construed as an over-tor cables with the goal to ultimately estimate lifetimes.
I all, definitive review of the papers. For example, some I
publications might contain a relevant description of a lhe Time-Domain Dielectric Spectrometer is used method applied in assessing the aging of the insulation in a method for measuring dielectric properties over and thus provide a lead toward a good candidate the frequency range from 104 to 10A lh by obtaining method for other purposes. Also, papers are still listed the sample response to a step application of voltage.
that were identified in the search as potentially rele-The data are then converted to the frequency domain vant but found upcm review not to be relevant.
by a numerical Laplace transform. This method ren-This inclusion of irrelevant papers can save the ders high precision and sensitivity and allows acquisi-l readers the unnecessary effort of acquisition and re-tion of the entire frequency range, typically at least view of seemingly relevant papers that might eventu-seven decades,in a time that is less than that required ally prove not applicab!c to their problem. Conversely, for one cycle of the lowest frequency of interest.
I significant papers may have been missed in the search, One of the current problems in the operation of ag-ano their shsence from this compilation should not be ing nuclear power plants is the estimation of the life-construed as a egative judgment of their significance.
time of the electricalinsulation used inside the reactor NISTIR 4487. E D. Martzloff, E. Simmon, J. P. Steiner, confinement. This insulation, having been manufac-and R. J. Van lirt.nt, " Detection of Inc' ient tured for mstallation m reactors, has been subject to Defects in Cablet by Partial Discharge 'gnal lifetime testing prior to certification for use. This test-Analysis," National Institute of Standards and ing program has been quite successful, as there have Technology, July 1992.
been too few failures to make any changes in the re-As one of the objective of a program aimed at as-quirements.
sessing existing test methods 1 Tr in situ detection of in-As the plants age, however, the question has arisen cipient defects caused by ch, in cables, a laboratory whether the life remaining in the cable insulation can test system was implemented to demonstrate that the be estimated so that the cables can be use6 beyond their partial discharge analysis method can be successfully original certified life. If this were possible, costly and 11 NUREG-1377
l-Main Citations and Summaries difficult overhauls of the reactor wiring could be NUREG-1144, B. M. Morris and J. P. Vora, " Nuclear avoided. Given the conservative nature of the certifica-Plant Aging Research (NPAR) Program Plan," U.S.
tion process, this possibility of extended lifetime cer.
Nuclear Regulatory Commtssion, July 1985, tainly exists.
NUREG-1144. J. P. Vora, " Nuclear Plant Aging The original tests on the cables were acceptance Research (NPAR) Program, Plan," Rev.1 U.S.
tests in which the cables were subjected to an environ-Nuclear Regulatory Commission, September 1987.
ment more severe than found in service, and the cables NUREG-1144, " Nuclear Plant Aging Research (NPAR) had to maintain their integrity. These tests have re.
Program Plan, Status and Accomplishments," Revi-suited in a situation in which cables are possibly used sig'2, U.S. Nuclear Regulatory Commission, June for only a small part of their potentiallifetime. For esti-mation of a lifetime, such tests provide little informa-The Nuclear Plant Aging Research (NPAR) pro-tion other than a possible minimum allowable expo-gram described in this plan is intended to resolve tech-sure. These acceptance tests do not follow changes in nical safety issues related to the aging degradation of the cables as they ageflhe effects of accelerated aging electrical and mechamcal components, safety and sup-used in these tests are not studied. Finally, there is no port systems, and civil engineering structures used m indication at all of the ultimate failure mechanisms of commercial nuclear powv plants. The aging period of the cables and when and how the failures might occur.
interest includes the perioo overed by the original op-erating license as well as the pc.xxl of extended plant
'Ihe degradation of electricalinsulation under com-life that may be requested in utility applications for li-bined thermal and radiation stresses is a well-known cense renewals.
phenomenon. Unfortunately, the rate at which this Emphasis has been placed on identifying and char-degradation ogcurs is highly material-dependent, and a acterizing the mechanisms of material and component small change m additives can make a large difference.
degradation during service and utilizing the research In addition, the mechanisms are very complex, with results in the regulatory process.The research includes many possible deterioration scenarios. All these factors evaluating methods of inspection, surveillance, c(mdi-make theoretical prediction of lifetimes very difficult.
tion monitoring, and maintenance as means of manag-They also create the possibility of relatively large vari-ing and mitigating aging effects that may affect safe t ns in the useful lifetime m a given material (compo-plant operation. Specifically, the goals of the program are to:
This investigation was to determine whether some
.l.
Identify and characterize aging mechanisms and measure of cable aging (e.g., deterioration) could be effects that could cause degradation of compo-found. Ideally, undesired changes would vary smoothly nents, systems, and civil er:gineering structures with aging and would display measurable changes that and, if unchecked, impair pant safety.
arc above any possible sample-to-sample variations. Fi-
- 2. Evaluate residual life of compments, systems, nally, there might be a possibility foradding instrumen-and civil structures and identifv methods of in-tation (preferably nonintrusive) without major modifi-spection, surveillance, and monitoring that will cations to any reactor, ensure timely detection of aging effects before Lifetime estimation is complicated by the require-loss of safety functions.
ment that any electrical cable must remain functional
- 3. Evaluate the effectiveness of storage, mainte-during a loss-of-coolant accident (LOCA). Since such nance, repair, and replacement practices in miti-an event can put severe thermal and radiation stresses gating the rate and extent of degradation caused on the cables inside a reactor, the cables must not be by aging.
close to failure prior to such an event. It should be NUREG/CP-0036, (Compilation by) B. E.11ader and stressed that although a cable at the end of its useful L. A. Hanchey," Proceedings of the Workshop on life could appear to be quite normal,its mechanical and Nuclear Plant Aging," Sandia National electrical properties may deteriorate quickly when the Laboratories, SAND 82-2264C, November 1982.
insulation is near failure.
'Ihe objective of the workshop, held August 4-5, The estimation of lifetime for reactor cables cur-1982, in liethesda, Maryland, was to facilitate an ex-rently in use isvery difficult because of the combination change of thoughts between the NRC and industry on of the above considerations. Not only is a convenient time-related degradation and its influence on reactor.
measure of age difficult to establish, but all measure.
safety.The specific goals were to define the problem, to ments must be conducted in a relatively harsh environ.
discuss the state of knowledge on aging phenomena, ment. Furthermore, any estimation must attempt to and to identify future activities necessary to understand predict the future, including the possibility of an event the problem.
more severe (with respect to lifetime) than the total en.
The need for a comprehensive program to identify vironmental aging allowable to a real (accident-free) the potential sadety problems associated with plant ag-endpoint.
ing was stressed. It was suggested that the effects of NUREG-1377 12
Main Citations and Summaries time-related degradation on the safety of the complete NUREG/CP-4)l05, Proceedings of the Seventeenth reactor system should be evaluated in terms of the risk Water Reactor Safety Information Meeting, Vol. 3, to the public, One should consider multiple causes that U.S. Nuclear Regulatory Commission. Paper by J.A.
have typically been associated with abnormal occur.
Chnstensen,"NPAR A proach to Controllmg Ag-ing in Nuclear Power P ants, Pacific Northwest rences. Sm.cc mdit,idual component failures create I aboratory, PNL-SA-17487, March 1990.
problems, time-related degradation will ultimately have to be addressed in terms of maintenance, moni-For about 8 years, the NRC NPAR program has toring, surveillance, etc., of components.
been devehying a techm. cal understandingof and gu d.
ance for mitigatmg the effects of the time-dependent A large number of phenomena that can cause fail-processes responsible for the aging-related deteriora-urcs were discussed; a detailed list of parts / materials, tion of structures, systems, and componems that can including lubricants and other additives that must be reduce safety margins in a nuc! car power plant, considered, was given; seemingly minor changes in the Controlof the effectsof agingisat the centerof the chemical constituents of a material or in the manufac" NPAR cfforts;it consists of three key clements:(1)se-turing process can cause significant effects and changes lection of the structures, systems, and comp (ments in in the system during operation (e.g., water chemistry which aging-related degradation must be controlled, effects).
(2) understanding the mechanisms and rates of the deg-Replacement parts were noted as a potential source radation, and (3) managing the degradation through ef-fcctive surseillance and maintenance. These elements of problems. The effects of storage on parts and the l
possibility that new parts may be different from the are implemented through various ongoing NRC and in-original ones were mentioned.
dustry programs and initiatives as well as by conven-tionalregulatmyinstruments. Also,the threeelements There were exter.sive discussions on the limitations are being addressed in a compilation of good practices of accelerated-aging tests. The use of naturally aged that will integrate the information developed under equipment for test purposes was suggested. Sacrificial NPAR and other studies of aging into a systems-ori-replacement of equipment was identified as a source ented format that tracks directly with the Safety Analy-for naturally aged plant equipment.
sis Reports.
l Maintenance and surveillance in plants and their The need to mitigate time-dependent deterioration relationship to time-related degradation were exten_
of NPP components is not a new or recent concept.
sisely discussed.
1)cgradation with time is, or should be, a prime consid-eration in any design effort.The material specifications and mechanical designs that characterize the struc-NUREG/CP-0100, A. E ilcranck," Proceedings of the tures, systems, and components of nuclear power International Nuclear Power Plant Aginf' plants reflect conscious, detailed concern on the part of Sm si m, U.S. Nuclear Regulatory ( ommission, the designer for the effects of anticipated service envi-ronments and stressors on functionality over time.The This report presents the proceedings of the Interna-major codes and standards upon which nuclear power tional Nuc! car Power Plant Aging Symposium that was plant designs and inservice inspections are premised held at the ilyatt Regency Ilotel in liethesda, Mary-(e.g., ASME iloiler and Pressure Vessel Code, Sections land, on August 30-31 and September 1,1988. The III and XI) are based in large measure on recogniecd Symposium was presented in cooperation mth the needs to achieve acceptable performance over a rea.
Americm Nuc! car Society, the American Society of sonable time whether or not age-related degradation is Civil Engineers, the American Society of Mechanical explicitly addressed by the wording in the codes. Even Engineers, and the Institute of Electrical and Electron-though time-dependent degradation is fundamental to ics Engineers, There were approximately 550 partici.
the codes and standards that govern the design, con-pants from 16 countries at the Symposium.
struction, and operation of nuclear power plants, addi-tional concern over aging is war ranted for the following A total of 48 papers were presented in 7 technical reasons; sessions:
- 1. 'Ihc lack of specificity on time rates of deteriora-
- 1. Aging Research Programs, tion in the codes requires the exercise of consid-
- 2. Aging of Structures and Mechanical Equip-erable judgment in design and other functions.
- ment, The availability of explicit, detailed information
- 3. Aging of Electric:d Equipment, on aging rates and consequences c;m result in
- 4. Aging of Systems and Components, more consistent judgments.
- 5. Reliability,
- 2. 1)eveloping technology will require that the un-
- 6. Role of Maintenance in Aging Management, derstanding of aging-related degradation be
- 7. Aging of Vessels and Steam Generators.
pushed beyond that inherent in current codes.
l 13 NUREG-1377
Main Citations and Summaries
- 3. Accounting for aging in design and operational
- 2. identifying generic problems and recurring fail-guidelines depends on several conservatisms to ures; provide prudent assurance that failures will not
- 3. identifying the variables (e.g., environment, op-occur.1he management of agmg, however, re-erating mode, system, maintenance policy, etc.)
quires understandmg and control of the time-that control component failure rates; dependent degradation that actually occurs to L pmviding an extensive data base against which -
implement and evaluate maintenance pro-to comparc existing data sources (e.g., LERs and grams.
NPRDs) to assess the degree to which these data -
- 4. The design process considers aging-related deg-sources accurately reflect the actual component -
radation of single components in estimating reliability; design lifetimes, but does not take into account
- 5. correlating current incidents with previous fail-the implications of common-cause failures due ures, allowing for extrapolation in the near fu-to aging in redundant components or the ampli-ture;
~
fied effect of aging in failure sequences involv-ing interactions of multiple components. The
- 6. identifying trends and patterns m. the fail-evaluation of these effects requires that accu-ure charactenstics of particular components or rate failure rate vs. time data for each compo-aggregations of components; and nent be applied using probabilistic methods that
- 7. identifyingfailure mechanismsovertime foruse redistically model systems on a plant-specific in defining the aging requirements for compo-basis.
nent qualification.
Because of those kinds of considerations, specific NUREG/CR-3154. R. J. Ilorkowski, W. K. Kahl, T L concerns, independent of plant design and operational liebble, J. R. Fragola, and J. W. Johnson, "Ihe In-parameters, over how and why structures, systems, and Plant Reliability Data Base for Nuclear Plant Com-components degrade with age must be resolved. The puents: Interim Report-The Valve Component,"
generic content and structure of programs for address.
Oak Ridge National Laboratory, ORN11FM-8647, December 1983.
ing degradation d ue to aging are discussed in this paper.
This document details the collection and prelimi-nary analyses of data related to valves in the in-Plant NUREG/CR-2641, J. P. Drago, R. J. Horkowski, D. II.
Reliability Data System (IPRDS). The data base is de-Pike, and E E Goldberg,"The In-Plant Reliability veloped pnmanly from histoncal records of corrective -
Data Base for Nuclear Power Plant Components:
maintenance actions obtained directly from nuclear-Data Collection and Methodology Report," Oak plant maintenance files. A comprehensive valve popu.
Ridge National Laboratory, ORNI11'M-8271, July lation is also included. This report presents data from 1982.
one PWR and one BWR power plant.
The development of a component reliability data The report demonstrates the degree of distinction base for use in nuclear power plant probabilistic risk and refinement in the reliability statistics that is possi-assessments and reliability studies is presented. The ble with data from the IPRDS and suggests a general data sources are the in-plant maintenance work format for disclosure of suitable reliability statistics to request records from a sample of nuclear power plants.
satisfy needs within the nuclear data-gathering com-This data base is called the In-Plant Reliability Data munity. The examples given in the various tables and '
System (IPRDS). Its features are compared with other figures suggest a useful method of comparing valve data sources such as the Licensee Event Report (LER) data and are representative of the degree to which reli-system. the Nuclear Plan t Reliability Data (NPR D) sys-ability statistics for any particular valve can be ascer-tem, and IEEE Standard 500. Generic descriptions of tained.
nuclear power plant systems formulated for IPRDS are One objective of this report is to examine the im-outlined in the text.
provement possible using IPRDS in refining the statis-The major objective of the program desenbed is to tics to ultimately focus on the reliability of particular provide an improved multipurpose data base. Compo.
valve types and valve operators in specific working envi-nents of each type of NSSS are included in the data ronments. Another objective is to generate comments base.
fmm members of the nuclear data community as to the efficacy of the suggested formats for documenting in addition to providing information on past failure valve information and the various methods used for rates and component down times, the IPRDS may be comparison in this report.
used for:
1hc report gives breakdowns of failure rates by fail-
- 1. revising component test intervals and allowable ure modes and by failure causes showing calculated down times; maintenance frequencies and repair times. IPRDS re-NUREG-1377 14-
_q u
l e
Main Citations and Summaries pair time distributions, although unavailable from outside the scope of the workshops because little or no
[
LERs, are also presented and evaluated.
first-hand experience was available for these off-Preliminary results obtained from the pilot data normal or yet-to-be-explored circumstances. Recom-base in this report indicate WASH-1400 statistics to be mendations are made for a systematic approach to rat-nonconservative in reliability estimates for some valve ing components in terms of overall safety and for a co-3 types in certain failure modes.
operative effort between industry research groups and regulatory research groups to resolve known aging NUREG/CR-3543, G. A. Murphy, R. H. Gallaher, problems and to identify off-normal or yet-to-develop M. L. Casada, and H. C. Ifoy, " Survey,of Operating aging issues. In addition to some well-known aging Experiences from LERs to identify Agmg1 rends,"
mechanisms (e.g., neutron embrittlement of pressure Oak Rid e National Latmratory, ORNL-NSIC-216, vessels)or problems that manifest themselves as equip.
January 9%
ment failures (e.g., steam generator tube degradation),
1his report describes the preliminary results of an there is concern that other types of aging problems may assessment ofinformation pertinent to identifying age-be developing. Their effects increase as nuclear power related failures available in operating experience plants get older, and some aging processes could even-I reports. This assessment, by the Nuclear Operations tually affect power plant availability or safety.
Analysis Center (NOAC) at Oak Ridge National Laboratory, utilized the computerized files of Licensee NUREG/CR-3819, J. A. Rose, R. Steele, Jr., K. G.
Lvent Reports (LERs) and their predecessors to DeWall, and 11. C. Cornwell, Survey of Aged Ibwer Plant Facilities." Idaho National Engineering i
examine age-related degradation of safety-related Laboratory, EGG-2317. June 1985.
equipment.
The survey concentrated on component failures in Abstracts of operating experience reports from LWR safety-related systems as determined from oper-commercial power plants reported from 1%9 to 1982 ating histories. Only failures that were determined to were surveyed. Over 7(X)0 events were reviewed; Data be age related were included.
included the system, component, subpart, the age-The age-related failure information gathered from related failure mechanism, the severity, and the the plant histories was analyzed for reoccurring failure method of detection of the failure. Wear, corrosion, patterns. Early program emphasis was on isolating spe-crud, and fatigue were the identified failure mecha-cific equipment with high failure rates that were not al-nisms in over one-third of the 3098 age-related events-ready the concern of other research efforts. The result.
Almut two-thirds of the failure severities were judged ing (gathered) data could not support the identification j
as a degraded state, and one-third were judged as cata-of specific equipment. !! did, however, imply a direct re-st rophic failures. Pump and valve problems made up al-lationship between the failure and the failure mecha-most 30% of the failed components. Almost two-thirds nism.Thus the emphasis of the program was redirected of the reported failures were detected by routine sur" toward exploring the relationship of the failure to the veillance testing indicating that such practices are ef-failure mechanism.
fcctive techniques for monitoring and detecting age The results of this preliminary investigation indi-degradation of discrete components and systems. A cated that about 70% of the significant failures re-substantial number of events resulted from setpoint ported for the fluid systems analyzed were due to only drift.
four failure mechanisms (causes): erosion, corrosion, NUREG/CR-3818. N. IL Clark and D. L lierry, vibration, and foreign materials. This was subsequently
" Report of Results of Nuclear Power Plant Aging verified by detailed study of several more plant systems Workshop " Sandia National Laboratories, and corroborated by field data obtained from pers(m nel SANDS 4-0374, August 1984, interviews, in addition, there appears to be a strong The objective of the workshops was to identify correlation between the cause of comp < ment failure whether there is any evidence of component or struc_
and the system in which the component operates.
tural time-related degradation, i.e., aging problems, in The survey points out, with evidence, that the iden-
- a nuclear power plant and, if so, what problems are of tification and climination of the system-level causes of greatest importance. Fifteen representatives from component failures is a viable approach to preventmg national laimratories, architect / engineers, nuclear and mitigating the major reported aging effects.
steam supply system vendors, research firms, and one NUREG/CR-3956, M. R. Dinsel, M. R. Donaldson, and
. university patticipated. Questionnaires and group dis-E T. Soberano, "In Situ Testing of the Shi Atomic Power Station Electrical Circuits,ppingport cussions screened over 112 components believed to be Idaho susceptible to excessive aging; pressure and tempera.
National Engineering Laboratory, EGG-2443, April ture mm valve operators, and snubbers emerged by 1987.
con'ensus as the most important. Ibtential aging prob-This report discusses the results of electrical in situ len.s related to off-normal common-mode effects or testing of selected circuits and components at the Ship-publems that were just developing at the time were pingport Atomic Power Station in Shippingport, Penn-15 NUREG-1377
l i
Main Citations and Summaries sylvania.'Ihe goal was to determine the extent of aging identify the components that show the highest poten-or degradation of various circuits from the original tial for risk-due-to-aging phenomena, plant and the two major core plant upgrades (repre-Three operating NSSS were analyzed, and it was senting a total of three distinct age groups)as well as to found that the most risk-significant components are in evaluate previously developed surveillance technology.
the auxiliary feedwater system, the reactor protection The electrical testing was performed usmg the Electri' system, and the service water systems, e.g., pumps, cal Circuit Characterization and Diagnostic (ECCAD) check valves, motor-operated valves, circuit breakers, system developed by EG&G for the U.S. Department and actuating circuits-of Energy to use at TMI-2. Testing included measure-Future research on the time-dependent portion of ments of voltage, effective series capacitance, effective series inductance, impedance, effective series resis-agmg phenomena for these components is needed to tance, de resistance, insulation resistance, and time-completely describe the impact on risk.
domain reflectometry (FDR) parameters. The circuits NUREC/CR-4156. M. Subudhi, E. L Burns, and J. H.
ev'duated included pressurizer heaters, control rod p '
'Ihylor, " Operating Experience and Aging-Seismic sition indicator cables, miscellaneous primary system Assessment of Electric Motors," Brookhaven resistance temperature detectors (RI Ds), nuclear in-National Laboratory, ilNI NUREG-51861, June strumentation cables, and safety injection system 1985.
motor-operated valves. It should be noted that the op-A limited number of electric motor categories with erability of these circuits was tested several years after direct safety significance were identified, and failures plant operation was concluded at Shippingport. There due to insulation degradation were surveyed.
was no need to retain the circuits m working condition following plant shutdown, so no effort was expended Age-sensitive components (with respect to materi-for that purpose.The in situ measurements and analy.
als and design features) were reviewed, potential elec-sis of the data confirmed the effectiveness of the trical and mechanical hazards were considered, opera-tional and accident stressors were determined, and ECCAD system for detecting degradation of circuit monitorable functional indicators were identified.The connections and splices along the high-resistance paths: most of the problems were caused by corrosion.
contribution of pertinent seismic effects was assessed, Results indicate a correlation between the chronologi.
and failure modes, mechanisms, and causes were re-cal age of circuits and circuit degradation.
viewed from existing data bases.
NUREG/CR 4234. W. L Greenstreet, G. A. Murphy, NUREGICR-4144, T. Davis, A. Shafaghi, R. Kurth, and and D. M. Eissenberg, " Aging and Service Wear E Leverenz,"Importance Ranking flased on Aging of Electric Motor-Operated Valves Used in Considention of Components included in Engineered Safety-Feature Systems of Nuclear Probabilistic Risk Assessments," Pacific Northwest Power Plants," Vol.1. Oak Ridge National Laboratory, PNIe5389, April 1985.
Laboratory, ORNL-6170/V1, June 1985.
The method outlined in the report ranks power This report deals with motor-operated valves, fo-plant components by using a risk-due-to-aging sensitiv.
cusing on monitoring defects and degradation of nu-
. ity measure that describes the change in risk due to clear plant safety equipment. The contents include the changes in component failure rate (without describing evaluation and identification of practical and cost-closely the aging phenomena and the resulting time.
effective methods for detecting, rnonitoring, and as-dependent component failure rate).
sessing the severity, failure modes, and causes (mainly aging and service wear) of time-dependent degrada-The outpet from this study can be combined with tion in nuclear plants. Also being considered are manu-that from other studies (data, analytical or experimen-facturer-recommended maintenance and surveillance tal) that identify the components most susceptible to practices and the selection of measurable parameters aging.
(including functional indicators) for use in assessing op-erational readiness, establishing degradation trends, The applications use average component unavail-ability equations currently empk>yed in probabilistic and detecting incipient failures. The report's results are based on information derived from operating expe-risk assessment (PRA) to calculate the risk-due-to-rience records, nuclear industry reports, manufac-aging sensitivity. A more exact calculation is possible by turer-supplied information, and input from architect-using unavailability equations that include the time-engmeer firms and plant operators, dependent characteristics of comp (ment failure rates; however, these time-dependent characteristics are not Failure modes are identified for both the valve and well known.The risk.due-to-aging sensitivity measure the motor-operator assembly For each failure mode, presented here is therefore segregated from these failure causes are listed by subcomponent or sub-time-dependent effects and addresses only the time-assembly, and parameters potentially useful for detect-independent portion of aging phenomena.The results ing degradation that could lead to failu re are identified.
NUREG-1377 16
i Main Citations and Summaries
%c method emerging from this analysis of the data strain, (4) torque-and limit-switch actuation (times of can provide capabilities for establishing degradation occurrence), (5) internal and external motor tempera-trends prior to failure and developing guidance for ef-tures, (6) vibration (several h) cations), (7) torque-fective and safe maintenance.
switch angular position, and (8) motor current.
. 'lhe tests led to the conclusion that the single most NUREG/CR-4234, II. D. Ila nes, " Aging and Service Wear of Electric Motor perated Valves Used in mformative measurable parameter was also the one Engineered Safety-Feature Sstems of Nuclear that was most easily acquired, t.c., the motor current.
l'ower Plants: Apng Assesments and Monitoring MCSA was found to provide detailed information re-
}
Method Evaluations " Vol. 2, Oak Ridge National lated to the condition of the motor, motor operator, Laboratory, ORNL-6170/V2, August 1989.
and valve across a wide range of values of parameters Motor-operated valves (MOVs) are k)cated in al-and their variations.He recording and the aralysis can most all plant fluid systems. Their failures have re-be donc during valve operation to render m, formation sulted in significant plant maintenance efforts. More that characterizes transient and periodic occurrences.
important, the opermioaal readiness of nuclear plant Several tests were carried out to investigate the ca-safety-related systems has often been affected l'y MOV pabilities of monitoring methods (especially MCSA) degradation and failure. Thus. in recent years, MOVs for detecting changes in operating conditions and have received considerable attention by the Nuclear MOV degradation. Results from selected laboratory Regulatory Commission and the nuclear power indus-tests presented in the report illustrate examples of (1) try and were identified as a component for study by the valve stem taper, (2) stem nut wear,(3) degraded volt-NRC NPAR program. In support of the NPAR Pro-age, (4) degraded valve stem iubrication, (5) worm-gear gram, a comprehensive Phase 11 aging assessment on tooth wear, (6) obstruction in valve seat area,(7) motor MOVs was performed by the Oak Ridge National pinion disengagement, (8) degraded worm and worm-Laboratory (ORNL), and the results of this study are gear lubrication, (9) stem packing adjustments, and (10) presented in this report.
torque-switch settings.
An evaluation of commercially available MOV in situ signature analysis tests were carried out by monitoring methods was carried out, as well as an as-ORNL on a total of 20 aged MOVs at a neighboring nu-sessment of other potentially useful techniques.These clear power plant. Five of these MOVs were later assessments led to the identification of an effective, retested after they were refurbished. Selected results nonintrusive, an,d remote technique, motor current from these tests are presented in this report and show, signature analysis (MCSA). The capabilities of moni-for example, differences in motor current signatures of toring methods (especially MCSA) for detecting similar MOVs that were indicative of control-switch I
changes in operating conditions and MOV degradation setting variations and differences in component wear.
were mvestigated m controlled laboratory tests at The influences of refurbishing and inactivity on MOV ORNL, m situ MOV tests at a neighboring nuclear operations were clearly seen in motor current signa-j power plant, and the gate valve flow mterruption blow-tures as well.
l down test in lluntsville, Alabama.
The background information and the work leading ORNL participated in the gate valve flow interrup-to the selected monitoring method are summarized be-tion blowdown test program carried out under the di-I low. A primary objective of this study was to identify ef-rection of the Idaho National Engineering Laboratory fcctive methods for monitoring the condition of at Wyle Laboratories in Iluntsville, Alabama.This test i
motor-operated valves used in safety-related systems w s n excellent opportunity for MOV diagnostic stud-of nuclear power plants. In respcmse to a need for im.
ies and, more important, a means for determming the I
proved methods for monitoring MOV condition, sev.
mfluences of high blowdown flow on the operation of eral systems that use a variety of sensing devices and boiling water reactor isolation valves.The reduction m signal-processing equipment and provide signatures operating " margin" of a MOV due to the presence of that yicId useful diagnostic information have been de.
additional valve running k> ads was imposed by high veloped in the last few years. As part of the Phase !!
flow.his was observed in motor current and torque.
MOV study, one of the motor-operated valve analysis switch angular-position signatures, as illustrated in this
& test systems (MOVKrS)was evaluated in depth.This study. In addition, the effects of differential pressure, evaluation and a description of four other commer, fluid temperature, and ime voltage on MOV operation cially available systems are included in this report.
were clearly seen.
In addition, the type and potential value of diag-The report presents information that should be use-nostic information from many measurable parameters fut in resolving MOV issues concerning the NRC and were determined by ORNL tests using MOVs mount ed the nuclear industry. Important areas not covered by on test stands. The selected parameters are (1) valve the Phase 11 work are identified, and recommendations stem position, (2- ) valve stem velocity, (3) valve stem for additional work are included.
17 NUREG-1377
a Main Citations and Summaries NUREG/CR-4257, S. Ahmed, A. Carfagno, and G. J.
Evaluation of inservice failures is recommended to Tbman,
- Inspection, Surveillance, and Monitoring of allow funher differentiation between sudden failures Electrical Equipment Inside Containment of (having no precursor) and failures that can be detected Nuclear Power Plants-With A plications to in the incipient state. Such evaluations would aid in the Electrical Cables, Vol.1, Oak idge National Laboratory, ORNL/SUll/83-28915/1, August 1985.
further development of monitoring techniques. Be -
cause some of the transmitter failures are of the sudden The purpose of this report is to describe currently type, periodic operability checks are an important available methodology for detecting and determining means of detecting failures very soon after their occur-the amount and rate of age-related deterioration of rence so that a significant number of failed (inactive or safety-related equipment. The general concepts of inaccurate) transmitters do not remain undetected.
monitoring equipment condition for this purpose are described. 'The goal is to detect deterioration in the in.
A combination of operability monitoring and condi-cipient stage, prior to inservice failure and prior to the tion monitoring may be used to improve the probability point at which equipment can no longer be expected to of successfully weathering aging processes and accident perform its function when exposed to design basis acci-conditions.
dent conditions.
The application of condition monitoring is dis-NUREG/CR-4279, S.11. Bush, P. G. Heasler, and cussed specifically for electric cables. The goal is to de-R. E. Dodge, " Aging and Service Wear of Hydraulic termine the degree of cable degradation and to predict and Mechanical Snubbers Used on Safety-Related the remaining usefullife. In situ nondestructive testing Piping and Components of Nuclear Power Plants,"
and destructive latmratory testing are discussed as are Vol.1, Pacific Northwest Laboratory, PNIe5479, their limitations. Interim recommendations are given February 1986.
for the implementation of a condition-monitoring This report presents an overview of hydraulic and progra.n for cables.
mechanical snubbers used on nuclear piping systems and components.'"he functions and functional require-NUREG/CR-4257, G. J. Toman, " Inspection, ments of snubbers are outlined. The real versus per-Surveillance, and Monitormg of Electrical ceived need for s ubbers is reviewed based primarily on Equipment in Nuclear Power Plants, Vol.
e National studies conduded by a Pressure Vessel Research Com.
" Pressure Transmitters." Oak Ridg/3/V2, mittee. Tests conducted to qualify snubbers, to accept Laboratory, ORNL/SUB/83-28915 August 1986.
them on a case-by-case basis, and to establish their fit-ness for continued operation are reviewed.
This report describes the types of pressure transmit-ters commonly used in nuclear power plants according This report had two primary purposes:(1) to assess -
to their application. The stresses that affect these the effects of various aging mechanisms on hydraulic transmitters include ambient temperature, humidity, and mechanical snubber operation (e.g., leaking of radiation, process (fluid) medium pressure, and tem-seals, functional failures) and (2) to determine the effi-perature. The most common effects of the stresses on cacy of existing tests in determining the effects of aging the transmitters are calibration shifts. The evaluation and degradation mechanisms. These tests include of failure data contained m Licensee Event Reports in-breakaway force, drag force, vekicity/ acceleration dicates that total failure of pressure transmitters occurs range for activation in tension or compression, release relatively infrequently
- rates within specified tension / compression limits, and Comparison of as-found and as-left calibration data restricted thermal movement. The snubber operating is described as a partial means of evaluating the level of experience was reviewed using licensee event reports deterioration of a transmitter. Care must be taken to and other historical data for a period of more than 10 ensure that variations in method or procedure do not years. I)ata were statistically analyzed using arbitrary produce erroneous data and wrong conclusions. The snubber populations. Value-impact was considered in -
precision of the comparative measurements must also terms of exposure to a radioactive environment for be high.
examination / testing and in terms of the influence of The evaluation of calibration data alone will not lost snubber function and subsequent testing program ensure the capability of operating under design basis expansion on the costs and operation of a nuclear accident conditions. If, with time, steam or moisture power plant. The implications of the observed trends penetrates the transmitter housing, the transmitter were assessed; recommendations include modifying or electronics will become inaccurate and may fail.There-improving the examination and testing procedures to fore, the integrity of the housing seal must also be enhance snubber reliability. Optimization of snubber evaluated periodically to be able to predict continued populations by selective removal of unnecessary snub-performance capability, bers was also considered.
NUREG-1377 18
E Main Citations and Summaries NUREG/CR-4302, W. L Greenstreet, G. A. Murphy, evaluations carried out in support of the NRC NPAR R.11. Gallaher, and D. M. Eissenberg, " Aging and program of the following developmental or commer-Service Wear of Check Valves Used m Engmeered Safety-Feature Systems of Nuclear Power Plants,"
cially available methods for diagnostic monitoring of check valves'-
Vol.1 Oak Ridge National Laboratory, ORNL-6193/V1, December 1985.
1.
Acoustic emission monitoring, The report addresses detecting defects and moni.
2.
Ultrasonic inspections, toring the degradation of nuclear plant safety equip-3.
Magnetic flux signature analysis, ment. The program is concerned with identifying and 4.
Radiography, evaluating practical and cost-effective methods for de-5.
Pressure noise signature analysis.
tectmg, momtormg, and assessing the severity of time-dependent degradation (aging and service wear) of
'lhese evaluations were focused on the capability of check valves in nuclear plants. Ihese methods will al-each method to provide diagnostic information useful low degradation trends to be detected prior to failure m determining the effects of aging and service wear and allow guidance for effective maintenance to be de, (degradation) and detecting failures and undesirable veloped.
operating modes. Commercial suppliers of three check The topics considered are failure modes and causes valve monitoring systems recently participated in a resulting from aging and service wear, manufacturer-comprehensive series of tests designed to evaluate the recommended maint enance and surveillance practices, capability of each monitoring technology to detect the and measurable parameters (including functional ind
position, motion, and wear of check valve internals and cators) for use in assessing operational readiness, es-valve seat leakage. 'Ihis report describes these tests, tablishing degradation trends, and detectmg meipient which were directed by the Nuclear Industry Check failure.The results presented are based on mformation Valve Group and carried out at the Utah Water Re-denved from operating experience records, nuclearm, -
search Laboratory.
dustry reports, manufacturer-supplied information, Each monitoring method is described and com-and plant operators.
pared with the others, and areas in need of further de-Failure modes for check valves are identified and velepment are identified. Examples of test data ac-are examined by identifying methods for detecting fail' quired under controlled laboratory conditions and urcs and differentiating between their causes. For each iicid test data acquired at operating nuclear plants are presented.
failure mode, failure causes are listed by component or subassembly, and parameters potentially useful for de-Of the methods examined, acoustic emission moni-tect ng degradation that could lead to failure are tabu-toring, ultrasonic inspection, and magnetic flux signa-ture analysis provided the greatest level of diagnostic information. These three methods were shown to be The report also identifies parameters potentially useful for enhancing the detection of degradation and useful in determining check. valve condition (e.g., disk incipient failure; these parameters include dimensions, position, disk motion, and seat leakage), although none of the methods was, by itself, successful in monitorina bolt torque, noise, appearance, roughness, and all three condition indicators. However, the combina-
- cracking, tion of acoustic emission with either ultrasonic or mag-NUREG/CR-4302, M.D. Haynes, " Aging and Service netic flux monitoring yields a monitoring system with Wear of Check Valves Used in Engineered Safety, sufficient sensitivity to detect all major check valve op-1 Feature Systems of Nuclear Power Plants: Aging crating ccmditions. All three methods are still under f
Assessments and Monitoring Method Evaluatmns,"
development and are expected to improve as a result of Vol. 2, Oak Ridge National Laboratory, further testing, analysis, and evaluation.
ORNL-6193/V2, April 1991.
The failures of check valves have resulted m.sigmfi-NUREG/CR--4380, J. L Crowley and D. M. Eissenberg, cant maintenance efforts and, on occasion, in water
" Evaluation of the Motor-Operated Valve Analysis and 'Ibt System (MOVATS) to Detect Degradation, hammer, overpressurization of low-pressure systems, Incorrect Adjustments, and Other Abnormalities in and damage to flow system components.These failures Motor-Operated Valves," Dak Ridge have largely been attributed to severe degradation of National Laboratory, ORNL-6226, January 1986.
internal parts (e.g., hinge pins, hinge arms, disks, and An important aspect of the NPAR program strategy disk nut pins) resulting from instability (flutter) of is to demonstrate the utility of condition monitoring, check valve disks under normal plant operating condi-signature analysis, and other surveillance methods for tions. Present surveillance requirements for nuclear detecting, differen tiating. and trending various types of power plant check. valves have been inadequate for abnormalities in the components so that corrective timely detection and trending of such degradation be-measures can be implemented prior to loss of safety cause neither the flutter nor the resuhing wear can be function. A field test program was carried out to evalu-detected prior to valve failure. This report describes ate valve signature analysis as a surveillance method to 19 NUREG-1377
p Main Citations and Summanes achieve these results as well as to detect incorrect ad-modes, mechanisms, and causes were reviewed from justments in motor-operated valves. The technique operating experience and existing data banks. The specified in the titie (MOVATS)is the subject of this re-study also considered scismic effects on age-degraded port. In situ signature traces were obtained in 36 motor-components of battery chargers and inverters.
operated valves at four nuclear plant sites. Described The performance indicators that can be monitored are the test equipment package, the method of obtain-to assess component deterioration due to aging or ing the signatures, and determinations made as a result other relevant stressors are identified. Conforming of analyzing them. Based on the results of the signa-with the NPAR strategy as outlined in the program ture-analysis technique and those obtained from the plan, the study also includes a review of current stan-field-test program, the capabilities and limitations of dards and guides, maintenance programs, and research MOVAB are discussed.
activities pertaining to safety-related battery chargers and inverters for nuclear power plants.
NUREG/CR-4457, J. L Edson and J. E. Hardin,
" Aging of Class lE Batteries in Safety Systems of NUREG/CR-4590, K. R. Hoopingarner, J. W. Vause, Nuclear Power Plants," Idaho National tingineering D. A. Dingee, and J. E Nesbitt, " Aging of Nuclear Laboratory, EGG-2488, July 1987.
Station Diesel Generators: Evaluation of Operating and Expert Experience " Vols.1 and 2, Pacific This report presents the results of a study of aging Northwest laboratory, PNIr5832, August 1987.
effects on safety-related battenes m nuclear power Mific Nodwest Laboratory evaluated opera-plants. De purpose is to evaluate the agmg effects tional and expert experience pertaining to the aging caused by battery operation in a nuclear facility and t degradation of diesel generators in nuclear plant serv-evaluate maintenance, testmg, and momtoring prac-ice. The research identified and characterized the tices with respect to the effectiveness of these practices contribution of aging to emergency diesel generator m detecting and mitigating the effects of aging.The f,;;ures.
study follows the NRC NPAR approach and investi-Volume 1 reviews diesel-generator experience to gates the materials used in battery construction. It also identify the systems and components most subject to identifies stressors and aging mechanisms, presents op erating and testing experience related to aging effects, aging degradation and isolates the maj,or causes of fail-ure that may affect future operational readiness.
analyzes battery-failure event reports in various data Evaluations show that, as plants age, the percentage of bases. and evaluates recommended maintenance prac-tices. Data bases that were analyzed included the aging-related failures mcreases and failure rnodes NRC's Licensee Event Report system, the Institute for change. A compilation is presented of recommended corrective actions for the agmg-related failures identi-Nuclear Power Operations
- Nuclear Plant Reliability fied, and the trend of these failures is discussed. This Data System, the Oak Ridge National Laboratory's In-Plant Reliability Data System. and the S. M. Stoller study also includes a review of eurrent relevant industry Corporation's Nuclear Power Experience data base.
programs, research, and standards. Volume 1 presents the results of the Phase I research that identifies the NUREG/CR-4564, W. E. Gunther, M. Subudhi, and components and systems most susceptible to aging deg-J. H. Thylor, " Operating Experience and Aging-radation and the major causes of such degradation.-
Seismic Assessment of Hattery Chargers and Volume 2 reports the results of a workshop held on Inverters," Brookhaven National Laboratory, May 28 and 29,1986, with industry representatives to HNL-NUREG-51971, June 1986.
discuss the technical issues associated with aging of nu-Battery chargers and inverters are vital components clear service emergency diesel generators.The techni-of the nuclear power plant electrical safety system.The cal issues discussed most extensively were man /
objectives of this program are to (1) identify concerns machine interfaces, component interfaces, thermal related to the aging and service wear of equipment op-gradients of startup and cooldown, and the need for an crating in nuclent power plants, (2) assess their possi-accurate industry data base for trend analysis of the die-ble impact on plant safety, (3) identify effective inspec-sel generator system.
tion, rurveillance, and monitoring methods, and (4) recommend suitable maintenance practices to mitigate NUREG/CR-4597, M. L Ada,ms and E. Makay, " Aging rafi aging-related concerns and diminish the rate of degra-Nuclea Po er t,
ol 1 O dation due to aging and service wear.
Experience and Failure Identification " bak Rifge De designs of battery chargers (3 types) and inver-National Laboratory, ORN1-6282/V1, July 1986.
ters (4 types) and the materials for their construction in this report, typical auxiliary feedwater pump fea-are reviewed to identify age-sensitive components. Op-tures are described in terms of configuration details, crational and accidental stressors are determined, and materials of construction, operating requirements, and their effect on promoting aging degradation are as-modes of operation. Failure modes and causes due to sessed. Variations in plant electrical designs, as well as aging and service wear are identified and explained, system and component impacts were studied. Failure and measurable parameters (including functional indi-NUREG-1377 '
20 lo
~ --
i Main Citations and Summaries i
cators) for potential use in assessing operational readi-NUREGICR-4652, D. J. Naus, " Concrete Component ness, establishing degradation trends, and detecting in-Aging and its Significance Rblative to Life cipient failures are outlined.
Extenin of Nucicar Powcr Plants," Oak Ridge National L aboratory. ORNL/rM-10059, September A series of measures to correct present deficiencies 1986.
in surveillance, monitoring, and inservice testing practices is discussed. He main body of the report is
'Ihe objectives o{ this study are to (1) expand upon i
the work that was mitiated in the first two Electne supplemented by a number of relevant appendices; in Power Research Institute studies relative to longevity particular, a major appendix is included on engineering and life extension considerations of safety-related con-and design information useful to assess operational crete components m light. water reactor (LWR) facih-readiness.
ties and (2) provide background that will logically lead NUREG/CR-4597, D. M. Kitch, J. S. Schlonski, P. J.
to subsequent development of a methodology for Sowatskey, and W. V. Cesarski, " Aging and Service assessmg and predicting the effects of aging on the per-Wear of Auxiliary Feedwater Pumps for PWR formance of concrete-based materials and compo-Nuclear Power Plants," Vol. 2, " Aging Assessments nents.
I and, Monitoring Method Evaluations," Oak Ridge Applications of safety-related concrete comp National Laboratory, ORNL-6282/V2, June 1988.
i nents to LWR technology are identified, and pertment The subjects specified in the title are described and structures (containment buildings, containment base discussed in four major sections:
mats, biological shield walls, main building, and auxil-
- 1. Failure causes, iary buildings) and the materials of which they are con-i structed (concrete, mild steel reinforcement, pre-
^
- 2. Description ofinspection, surveillance, and con-stmssing systems, embedments, and anchorages) are ditior monitoring (ISCM) methods, described. Histoncal performance of concrete compo-
- 3. Evaluation of ISCM methods, and nents was established through information presented
- 4. Role of maintenance in alleviating aging and on concrete longevity and component behavior in both service wear.
LWR and high-temperature gas-cooled reactor appli-i pqems n ng mnmte ca w
Failure causes attributable to aging and service qa 8"""#" #
- "E "#
"E'""*
l wear are given and ranked in terms of imponance.
- E ## "I
- "* E
""I Cause identifications are made on the basis of experi-problems identified in conjunction with nuclear power ence, postservice examinations, and m. situ assess ~
applications _wcre minor; they include concrete crack-ments.
ing, concrete voids, or low concrete strengths at an Measurable parameters related to failure causes early age. Five incidents invohing LWR ccmcrete con-are identified. ISCM methods are specified, and evalu-tainments that are considered significant are described ations are made based on Westinghouse experience.
in detail.
On the same basis, recommendations are given on in-Potential environmental stressors and aging factors spection, surveillance, and condition momtoring. The to which LWR safety-related components could be sub-ISCM methods are mtended toyleid required capabih-jected are identified and discussed in terms of durabil.
ties for establishing operational readiness as well as for ity factors related to the materials used to fabricate the detecting and tracking degradation and its trends.
components.The current technology for detecting con.
'Ihe role of maintenance in alleviating and mitigat-crete aging phenomena is also presented in terms of j
ing aging and service wear effects is discussed, and the methods applicable to the particular material system relationship of maintenance to ISCM methods is iden-that could experience deteriorating effects. Remedial tified. Predictive, preventive, and corrective mainten-measures for the repair or replacement of degraded ance practices are discussed and evaluated.
concrete components and their effectiveness are dis-Appendices contain a detailed discussion on ISCM cussed. Finally, considerations relative to developing a methods, failure data base information, auxiliary feed.
damage methodology for assessing the durability fac-water pump (AUXFP) installation lists (location sur-tor, deterioration rates, and prediction of structural re-vey), a discussion of low. flow testing, auxiliary feed, liability are outlined.
water system descriptions (with flow-diagrams and Conclusions and recommendations of the report schemes), AUXFP minimum. flow-rate criteria, and note the need for (1) obtaining aging data from decom-guidelines proposed by Westinghouse for full-flow test-missioned plants, (2) using inservice inspection pro-
-ing. Note:'Ihe draft of this Vol. 2 (with the same titic) grams to provide aging trends, (3) developing a meth-was issued-by Westinghouse Electric Corporation, odology to quantitatively and uniformly (i.e., using the Generation Technology Systems Division, in April same procedures) assess structura! reliability as af-1986, coauthored by D. M. Kitch, M. Vuckovich, W. V.
fected by aging or degradation of structural materials.
Cesarski, and P. J. Sowatskey, and (4) performing research in support of all these 21 NUREG-1377
Main Citations and Summaries needs. It should be stressed that there is no widely ac-indicated a general trend of increasing failure rates in cepted standardized methodology for quantifying the the period of 6 to 11 years following the start of com-condijon and capacity of an individual concrete struc-mercial operation of the plants.He aging interaction ture.
study evaluated the interaction of aged relays and cir-cuit breakers in a safety injection system with regard to NUREG/CR-4692, G. A. Murphy and J. W. Cletcher II, five events requiring the system to start operation. Fail-
" Operating Experience Review of Failures of Power ure of redundant trains from common-mode failure of Operated Relief Valves and Block Valves m Nuclear Power Plants," Oak Ridge National Laboratory, a particular type of circuit breaker or relay is not expected. Howcver, the number of different types of ORNUNOAC-233, October 1987.
This report contains a review of nuclear power plant potential failures supports the need for a strong operating events involving failures of power-operated maintenance and surveillance program to prevent mul-relief valves (PORVs) and associated block valves tiple age-related failures from affectmg redundant (BVs). Of the 230 events identified,101 involved PORV safety trains.
mechanical failure,91 were attributable to PORV con-NUREG/CR-4731, V. N. Shah and P. E. MacDonald, trol failure, 6 involved design or fabrication of the
" Residual Life Assessment of Major Light Water PORVs, and 32 involved BV failures. The report con-Reactor Components," Vol.1. Idaho National Engi-tains a compilation of the PORV and BV failure events, neering Lateratory, EGO-2469. June 1987.
including failure cause and severity. The events are NUREG/CR-4731, V. N. Shah and P. E. MacDonald, identified as to plant and valve manufacturer. An as-
" Residual Life Assessment of Ma{or Light Water sessment of the need to upgrade FORVs and BVs to Reactor Components-Overview, Vol. 2. Idaho safety-grade status concludes that such action would National Engmeering Laboratory, EGG-2469, improve PORV and BV reliability. The greatest im.
November 1989.
provement in reliability would result from using newer, This report presents an assessment of 11e aging more reliable PORV designs and improving testing, di-(time-dependent degradation) of selected major light agnostics, and maintenance applied to PORVs and water reactor components and structures. The stres-BVs, particularly to the BV motor operators. A sum-sors, possible degradation sites and mechanisms, mary of interviews conducted with four PORV manu-potential failure modes and currently used non-facturers is also included in the report.
destructive examinations, inservice inspection (ISI),
and life assessment methods are discussed for major NUREG/CR-4715, G. J. 'Ihman, V. P. Bacanskas, T. A.
light water reactor components. Volume I covers PWR Shook, and C. C. Lodlow, "An Agm Assessment of Relays and Circuit Breakers and ystem Interac-and BWR pressure vessels, PWR containment struc-tions," Brookhaven National Laboratory, Franklin tures, PWR reactor coolant piping, PWR steam gen-Research Center. Philadelphia, PA, erators, BWR recirculation piping, and reactor pres-BNL-NUREG-52017, June 1987.
sure vessel supports. Volume 2 covers PWR reactor As part of the NRC NPAR pragram, Franklin Re-coolant pumps, PWR pressurizers PWR pressurizer scarch Center analyzed the effects of aging on safety-surge and spray lines, PWR reactor coolant system charg, g and safety mjection nozzles, PWR feedwater m
related circuit breakers and relays under contract to B rookhaven National 12boratory. Circuit breakers and lines, PWR control rod drive mechanisms and reactor relays in a PWR safety injection system were evaluated internals, BWR contamments, BWR feedwater and mam steam lines, BWR control rod drive mechanisms with respect to the aging caused by system operation.
The effect of circuit breaker and relay deterioration on and reactor internals, PWR and BWR electrical cables the ability of the system to perform its safety functions and connections, and PWR and BWR cmergencydiesel was also evaluated.The study inc!uded protective, con-generators. Unresolved technical issues related to un-trol, and logic relays, as well as molded-case and metal-derstanding and managing the aging of these major
-clad switchgear circuit breakers. Analysis of nuclear components, including requirements for advanced ISI and life assessment methods, are also discussed.
power plant failure data confirmed that normally ener-gized relays commonly used in safety systems suffer NUREG/CR-4740, L C. Meyer, " Nuclear Plant-Aging from more rapid deterioration than do deenergized Research on Reactor Protection Systems," Idaho relays. The failures were attributable to coil deteriora-National Engineering Laboratory, EGG-2467, tion, changes in dimensions of critical organic com-January 1988.
ponents, and changes in characteristics of timing This report presents the results of a review of oper-diaphragms from thermal deterioration. Some of the ating experience for the reactor trip system (RTS) and failure modes will prevent fail-safe operation. The the engineered safety feature actuating system (ES-electrical control and mechanical portions of metal-FAS) reported in Licensee Event Reports (LERs), the clad switchgear were found to be more failure prone Nuclear Power Experience data base, the Nuclear than the main contacts and arc extinguishing systems.
Plant Reliability Data System, and plant maintenance Analysis of failure data for circuit breakers and relays records.The purpose of the review was to evaluate the NUREG-1377 22
,n.
=- _--
i Main Citations and Summaries i
l i
potential significance of aging, including cycling, trips, is also extended to cover nonlinear and dependent ag-and testing, as a contributor to degradation of the RTS ing phenomena. The implementation of the linear ag-
{
and ESFAS. Thbles show the percentage of events for ing model is demonstrated by applying it to the aging RTS and ESFAS classified by cause, components, and data collected in the NRC NPAR program.
subcomponents for each of the nuclear steam supply NUREG/CR-4819, V. P. Bacanskas, G. C. Roberts, and synem vendors. A representative llabcock and Wilcox G. J. Toman, " Aging and Service Wear of Solenoid-I
- p. ant was selected for detailed study. The NRC NPAR Operated Valves Used in Safety Systems of Nuclear J
guidelines were followed in performing the detailed Power Plants," Vol.1, "Operatmg Experience and study that identified materials susceptible to aging.
Failure Identification," Oak Ridge National Labora-i stressors, environmental factors, and failure modes for tory, ORNL/SUB/83-28915/4/V1, March 1987.
the RTS and ESFAS and the relevant generic instru-An assessment of the types and uses of solenoid-mentation and control systems. Functional indicators operated valves (SOVs) in nuclear power plant safety-I of degradation are listed, testing requirements evala-related service is provided in the report. Through a de-ated, and regulatory issues discussed.
scription of the operation of each SOV combined with NUREG/CR-4747, H. M. Meale and D. G. Satterwhite, knowledge of nuclear power plant applications and op-
"An Aging Failure Survey of Light Water Reactor crational occurrences, the sigmficant stressors respon-Safety Systems and Components," Vol.1, Idaho sible for degradation of SOV performance are identi-National Engineering Laboratory, EGG-2473, July fied. A reviewof actualoperating experience (including 1987.
failure data) leads to the identification of potential n ndestructive in situ testing which, if properly devel-NUREG/CR-4747 IL M. Meale and D. G. Satterwhite, "An Aging Failure Survey of Light Water Reactor oped, could provide the methodology for momtonng Safety Systems and Components," Vol. 2, Idaho the degradation of SOVs. Recommendations are out-National Engineering Laboratory, EGG-2473, July lined for contmuing the study into the test methodol-1988.
ogy development phase.
His report describes the methods, analyses, re-NUREG/CR-4819, R. C. Kryter, " Aging and Service I
sults, and conclusions of two different agmg studies-Wear of Solenoid-Operated Valves Used in Safety The first study was a survey of light water reactor com-Systems of Nuclear Power Plants," Vol. 2,"Evalu-i ponent failures associated with 15 selected safety and ation of Monitoring Methods," Oak Ridge National support systems. Analysts used computerized sorting Laboratory, ORNL/IM-12038, July 1992.
techniques to classify component failures into generic Solenoid-operated valves (SOV) were studied at failure categories. The second study was a careful exa-Oak Ridge National Laboratory as part of the USNRC mination of component failure records to identify and Nuclear Plant Aging Research (NPAR) Program.The categorize the reported causes of component failures.
primary objective of the study was to identify, evaluate, De systems evaluated in the failure-cause analysis and recommend methods for inspection, surveillance, were the auxiliary feedwater, Class IE electric power monitoring, and maintenance of SOVs that can help distribution, high-pressure injection, and service water.
ensure their operational readiness, that is, their ability Thbles and figures indicate the systems and the compo-to perform required safety functions under all antici-nents within the systems that are most affected by ag-pated operating conditions.The failure of one of these ing. Engineering insights drawn from the data are pro-small and relatively inexpensive devices could have se-vided. Volume 2 presents all of the Volume 1 data from rious consequences under certain circumstances.
FY-86 combined with the data gathered in FY-87.
An earlier (Phase I) N PAR program study described SOV failure modes and causes and identified measur-NUREGICR-4769, W. E. Vesely, " Risk Evaluations of Aging Phenomena:He Linear Aging Reliability able parameters thought to be linked to the progression
' Model and its Extensions," Idaho National of ever-present degradation mechanisms that may ulti-Engineering Laboratory, EGG-2746, April 1987.
mately result in functional failure of the valve. Using A model for failure rates of light water reactor this earlier work as a guide, the present (Phase II) study safety system components due to aging mechanisms has focused on devising and then demonstrating the effec-been developed from basic phenomenological consid.
tiveness of techniques and equipment with which to crations. In the treatment, the occurrences of deterio.
measure performance parameters that show promise ration are modeled as following a Poisson probability for detecting the presence and trending the progress of process. ne severity of damage is allowed to have any such degradations before they reach a critical stage.
distribution; however, the damage is assumed to accu-Intrusive techniques requiring the addition of mag-mulate independently. Finally, the failure rate is mod-netic or acoustic sensors or the application of special cled as being proportional to the accumulated damage.
test signals were investigated briefly, but major empha-Using this treatment, the linear aging-failure rate sis was placed on the examination of nonintrusive, con-model is obtained. The applicability of the linear aging dition-indicating techniques that can be applied with
.model to various mechanisms is discussed. The model minimal cost and impact on plant operathn. These in-23 NUREG-1377 l
l
Main Citations and Summaries clude monitoring coil mean temperature remotely by iblume 1:ltrformance Evaluation and Afaintenance i
means of coil de resistance or ac impedance, verifying Practices unrestricted SOV plunger movement by measuring This report presents recommendations for develop-current and voltage at their critical bistable (pull-in and ing a cost-effective program for performance evalu-drop-out) values, and detecting the presence of shorted ation and maintenance of c!cetric motors in nuclear turns or insulation breakdown within the solenoid coil power plants. These recommendations are based on using interrupted-current test methods. The first of current industry practices, available techniques for these techniques, though perhaps the simpicst concep-monitoring degradation in motor components, manu-tually, will likely benefit the nuclear industry most be-facturers' recommendations, operating experience, i
cause SOVs have a history of failure in senice as a re-and results from two laboratory tests on aged motors.
sult of unwitting operation at excessive temperatures.
The test results (on a small and a large motor) provide Experimental results are presented that demon-the basis for recommendinp " ; various functionalindi-strate the technical feasibility and practicality of the c tors for maintenance programs.
monitoring techniques assessed in the study, and rec-The overall preventive program is separated into
)
ommendations for further work are provided.
two broad areas of activity aimed at mitigating the po-tential effects of equipment aging: performance evalu-NUREG/CR-4928, H. M. Ilashemian, K. M. Petersen, ation and equipment maintenance.The latter involves T. W. Kerlin, R. L. Anderson and K. E. Holbert, actually maintaining the condition of the equipment,
" Degradation of Nuclear Plant Temperature while the former involves monitoring degradation duc Sensors," Analysis and Measurement Services to aging. The monitoring methods are further catego-Corporation, Knoxville, TN, June 1987.
rized as periodic testing, surveillance testing, continu-I A program was established and initial tests were ous monitoring, and inspections.
performed to evaluate long-term performance of resis.
This study focuses on relevant methods and proce-tance temperature detectors (RTDs)of the type used in dures with the goal of maintaining the motors in a nu-.
t U.S. nuclear power plants. This report addresses the clear facility operationally ready. This includes an effect of aging on RfD calibration accuracy and re-evaluation of various functional indicators to deter-i sponse time. The Phase I effort (lasting about mine their suitabilie for trending assessments when l
6 months) included exposure of 13 safety-grade KrD monitoring the coi.
..on of motor components. He cicments to simulated LWR temperature regimes. Full intrusiveness of test methods and the present state of i
calibrations were performed initially and monthly, sen-the art for using the test equipment in a plant environ-sors were monitored and cross-checked continuously ment are discussed.
during exposure, and response time tests were per-Implementation of the information provided in this formed before and after exposure. Short-term calibra-report will improve motor reliability in nuclear power tion drifts of as much as 1.8*F (1 C) were observed.
plants.The study indicates the kinds of tests to conduct, Another result was that small response times were es-how and when to conduct them, and to which motors sentially unaffected by the testing performed.
the tests should be applied.
This program has demonstrated that there is a iblumc 2; Functionallndicator 7bsts on a Small i
sound reason for concern about the accuracy of tem-Electric Motor Subjected to Accelerated Aging perature mcasurements in nuclear power plants.These limited tests should be expanded in a Phase H program A 10-horsepower electric motor was artificially to mvolve more sensors and longer exposures to simu-aged by plug reverse cycling for test purposes.The mo-i lated LWR conditions m order to obtain statistically tor was manufactured in 1%7 and was in service at a sigmficant data Such data are needed to establish the commercial nuclear power plant for twelveyears Vari-length of meaningful testing or replacement intervals ous tests were performed on the motor throughout the for safety-grade RfDs. An important corollary benefit aging process. The motor failed after 3.79 million re-from thn expanded program would be a bet t er determi-versals (3 seconds per reversal) over seven months of i
nation of achievable accuracies m KrD calibrat,on.
testing. Each test parameter was trended to assess its '
I i
suitability in_ monitoring aging and service wear degra-NUREGICR-4939. M. Subudhi, W. E. Gunther, J. H.
dation in motors. Results and conclusions are discussed l
'laylor, R. Lofaro, K. M. Skreiner, A. C. Sugarman, relative to the applicability of the tests performed to-t and M W. Sheets, " Improving Motor Reliability in motor maintenance programs of nuclear power plants.
Nuclear Power Plants": Volume 1, " Performance Evaluation and Maintenance Practices": Volume 2, Volume 3: Failure Analysis and Diagnostic Tests on a
" Functional Indicator Tests on a Small Electric Naturally Aged Large Electric Motor l
- Motor Subjected to Accelerated Aging"; Volume 3, Stator coils of a naturally failed 400.hp motor from l
" Failure Analysis and Diagnostk Tests on a Natu.
the Brookhaven National Laboratory test reactor facil-rally Aged Electric Motor,"llrookhaven National ity were tested for their diclectric integrities. The mo-Laboratory. HNL-NUREG-52031, November 1987.-
tor was used to drive the primary reactor coolant pump NUREG-1377 24
Main Citations and Summaries for the last 20 years. Maintenance activities on this mo-In addition to the above engineering evaluation, the tor during its entire service life were minimal, with the components that contributed to system unavailability exception of meggering it periodically.The stator con-were determined, and the contribution of aging to sisted of ninety individual coils, which were separated HPIS unavailability was evaluated. The unavailability for testing. Seven different dielectric tests were per-assessment utilized an existing probabilistic risk assess-formed on the coils. Each set of data from the tested ment, the linear aging model, and generic failure data.
coils indicated a spectrum of variation depending on their aging conditions and characteristics. By compar-NUREG/CR-4977, R. Steele, Jr. and J. G. Arendts'ed
" SHAG Test Series: Seismic Research on an Ag ing the test data to baseline data, the test methods were United States Gate Wlve and on a Piping System in assessed for application to motor mainte-nance pro-the Decommissioned Heissdampfreaktor (HDR):
grams in nuclear power plants. Also included in this Summary," Vol.1, Idaho National Engineering study are results of an investigation to determine the Laboratory, EGG-2505, August 1989.
cause of this motor's failure.The aged condition of a NUREG/CR 4977, R. Steele, Jr. and J. G. Arendts, seccmd identical primary pump motor, which is of the SIIAG rest Series: Seismic Research on an Aged 7
Y n operation, is discussed.
United States Gate Valve and on a Pipmg System m i
same age and is Presenti Recommendations relating to the applicability of each the Decommissioned Heissdampfreaktor (HDR):
of the dielectric test methods to motor mamtenance Appendices." Vol. 2. Idaho National Engineering programs are formulated.
Laboratory, EGO-2.505, August 1989.
His report describes the investigation, results, and NUREG/CR--4967, L C. Meyer," Nuclear Plant Aging conclusions of the INEL effort to determine the cause Research on High Pressure Injection Systems" of the reduced performance of a naturally aged Crane Idaho National Er'tineering Laboratory, EGG-2514, gate valve with a Limitorque motor operator. The August 1989.
motor-operated valve served 25 years in the Ship-pingport Atomic Power Station as a feedwaterisolation This report presents the results of a review oflight y lve before being refurbished and installed in a piping -
water reactor high-pressure injection system (HPIS) system in the Heissdampfreaktor (HDR), where valve operating experience reported in the Nuclear Power perability m typical pressure and temperature envi-Experience Data Base, Licensee Event Reports (LERs), the Nuclear Plant Reliability Data System, and
[onments and dunng simulated carthquakes was stud-ied. During the test program it was discovered that, plant records.
under some hydraulic loadings, the motor operator Operating experience of nuclear power plants was failed to reach torque levels high enough to open the evaluated to determine the significance of aging-closing torque switch. Failure of the torque switch to related service wear on equipment and its possible im-open caused the motor to stall. In normal plant service, pact on safety.The HPIS and those portions of related stalling an operator motor can cause motor burnout systems needed for operation of the HPIS were se-and render the valve inoperable for subsequent safety lected for detailed study in order to evaluate the poten-functions.
tial significance of aging as a contributor to the degra-An extensive investigation was conducted to try to dation of that system.'Ihbles show the percentage of isolate the cause of the poorperformance of the motor significant events for HPIS classified by cause, compo-operator. This investigation included follow-on in situ nent, and subcomponent for PWRs and BWRs. A rep-tests at HDR, dynamometer testing of the motor op-resentative Habcock and Wilcox plant was selected for crator a: the Limitorque laboratory, testing of the detailed study, torque spring at INEL, dynamometer testing of the motor alone at the Peerless Motor laboratory, and a The NPAR guidelines provided the framework -
through which the effect of aging on HPIS was studied,-
mathematical analysis of the HDR power circuit. 'The and these guidelines were followed throughout the re.
investigation identified three causes of the motor-port, which presents an identification of failure modes, operator's poor performance: torque spring aging, a preliminary identification of failure causes due to ag.
heating of the motor windings, and resistance in the de ing and service wear degradation, and a review of cur.
power cabling at HDR. The investigation also demon-rent inspection, surveillance, and monitoring methods, strated that normal plant testing of valves is not ade-including manufacturer-recommended surveillance quate to ensure proper performance under flow and and maintenance practices. De detailed study identi.
pressure loads in combination.
fies materials susceptible to aging, various stressors, During the follow-on tests at HDR, we found that, and emironmental factors. Performance parameters or when the valve was subjected to flow loads and pressure functional indicators potentially useful in detecting loads in combination, the valve either torqued out in degradation are also identified, and preliminary recom-the partially open position, stalled in the partially open mendations are made regarding inspection, surveil-position, or stalled in the fully closed position, depend-I lance, and monitoring methods, ing on the load and the torque switch setting.The valve
'l 25 NUREG-1377 -
1 j
- Main Citations and Summaries torqued out in the fully closed position only when pres-multistage switch deterioration are identified. A review sure and flow loads were very low.
of operating experience (failure data) leads to identifi-Undersized power supply cabling resulting in high cation of potential and recommended monitoring resistance has surfaced as a problem in at least two de techniques for early detection of mcipient failures.
motor operators in the field. Though the other factors Although the operating experience does not justify ex-tensive detenoration momtonng of multistage contributed to the anomalous performance of the valve switches, nondestructive testing methods that could be at HDR, undersized cabling was the main cause. The used to evaluate the condition of switches are identi-NRC has recently issued an information notice regard.
fied.He report presents a detailed description of the ing the issue.
c mp nen, ma e a s ns uq n,an operadon None of the three problems discovered during the of each of the multistage switches meluded in the as-HDR tests and follow-on investi ation would be de-sessment. Also, it provides an analysts of failure data E
tected during the normal in-plant testing where the from the LER system. An analysis of the various failure valves are subjected to no load or to pressure loads modes of multistage rotary switches and their related alone. Fhe problems are detectable only at higher kiad-causes is also given.The existing recommended and re-ings, that is, flow loads m combmation with pressure quired maintenance and surveillance practices are loads, where the load slows the motor down to the ex-listed. Several techniques with a potential for assessing tent that momentum cannot carry the unit through the condition of switch components and possibly pre-complete closure and torqueout.
dicting age-related failures are identified. It is recom-mended that inservice failures be analyzed to deter-NUREG/CR-4985, M. Subudhi, J. II. Thylor, mine whether the failures are due to random defects or J. Clinton, C. J. Czajkowski, and J. Weeks, " Indian Point 2 Reactor Coolant Pump Seal Evaluations."
are the resuit of generic deficiencies that would require Brookhaven National 12boratory, corrective action.
BNL-NUREG-52095, August 1987.
NUREG/CR-5008, R. D. Meininger and T. J. Weir, This report summarizes the findings on Westing-
" Development of a lesting and Analysis Methodol-house reactor coolant pump (RCP) seal performance at ogy to Determine the Functional Condition of Sole-Indian Point 2. This study considered a significant num-noid Operated Valves," Pentek. Inc., Coraopolis, ber of RCP seal failures. Consolidated Edison initiated PA, September 1987.
a research effort to determine the causes of these fail.
The objective of this research was to develop a sim-urcs and to develop appropriate ameliorative action to plc, reliable, condition-monitoring system that will pro-enhance seat reliability.The BNL work is an outgrowth vide surveillance information without requiring discon-of the first-phase effort performed by Failure Analysis nection or disassembly of solenoid-operated valves Associates.The objectives of the BNL program are to (SOVs) installed in operating nuclear power plants.
determine the root causes of seat failure and to provide The information provided must be sufficiently reliable recommendations for improving seal reliability. This to allow plant operators to conclude that valve per-program made notable advances in understanding the formance has or has not degraded to the point where root causes of RCP seal failure. For the first time, ac-
. corrective maintenance becomes necessary.-
tual failed seals were examined in detail in BNL's hot The required information is assumed to be obtain-cell, and laboratory tests were conducted to determine able through analysis of in-rush current to the coil of failure causes. This report summarizes findings and the SOV. Various SOVs were tested in an experimental presents conclusions and recommendations based on air system set up in the laboratory. In-rush current data review of plant operating and maintenance data, con-acquired on degraded and new SOVs were analyzed to sultation with Westinghouse and utilities, review of determine behavior signature models.
prior RCP seal studies (including previous BNL work),
Laboratory conditions provided the opportunity to and visual and in-depth examinations of the first batch s mulate perturbations caused by the valve function,-
of service-exposed seals received from the plant.
which would differ from actuation to actuation. A vis-ual examination of this time-varying waveform re-NUREGICR-4992, G. C. Roberts, V. P. Bacanskas, and vealed distinct and repeatable vanations for different G. J.1bman, " Aging and Service Wear of Multistage valve anomalies.
Switches Used in Safety Systems of Nuclear Ibwer Plants," Vol.1 Oak Ridge National Laboratory, This technique could identify gross changes and ORNL/SUB/83-28915/5/VI, September 1987.
render characteristic signatures that could be used for An assessment of the types and uses of multistage various comparisons and to trend valve degradation switches in nuclear power plant safety-related service is mechanisms and their consequences over time.
provided. Through a description of the operation of Utilization of the laboratory technique in an operat-each type of switch combined with knowledge of nu-ing nuclear power plant would be somewhat impracti-clear power plant applications and operational cal since the installed valves are not equipped with syn-occurrences, the significant stressors resp (msible for chronous switching capability. Analytical research was NUREG-1377 -
26
Main Citations and Summaries i
)
therefore conducted to develop a technique to analyze to develop effective mitigating actions for the CCW similar electrical data obtained under asynchronous system.The effect of time on this system was character-conditions typical of an operating plant. For such field ized by using the " Aging and Life Extension Assess-application, the technique developed would use a clip-ment Program (ALEAP) Systems Level Plan", devel-on current probe, thus enabling all measurements to be oped by Brookhaven National Laboratory. Failure data made from outside the reactor building without dis-from various national data bases were reviewed and turbing any electrical connections. The in-rush current analyzed to identify predominant failure modes, to the solenoid-operated valve is analyzed in real time causes, and mechanisms in CCW systems. Time-using a personal computer and fast Fourier transform dependent failure rates for major components were i
techniques.
calculated to identify aging trends. Plant-specific data were obtained and evaluated to supplement data base NUREG/CR-5051, W. E. Gunther, R. Lewis, and results.
M. Subudhi, " Detecting and Mitigating Battery A computer program (PRAAGE) was developed -
Charger and Inverter Agm' g," Brookhaven National Laboratory, UNL-NURLG-52108, August 1988.
and implemented to model a typical CCW system de-sign and perform probabilistic risk assessment (PRA) j This report is the second on the two-step approach ca culations. lime-dependent failure rates were input i
for assessing the safety and operational aspects of bat-to the program to evaluate the effects of aging on the tery charger and inverter aging in nuclear power plants.
a c nmonent wn mpectjo nstem un-mp na Analyses include an assessment of the recent operating avail bihty. 'Ume-dependent changes m component e
experiences with battery chargers and inverters and a imp rtance and system unavailability with age were ob-l discussion of improvements in reliability that may be achieved through modification of the equipment's ccm-figuration and an increased inspection frequency. The NUREG/CR-5053, W. Shier and M. Subudhi, "Operat-i results are evaluated from a survey of the current ing Experience and Aging Assessment of Motor maintenance and test practices used in nuclear power Control Centers," Brookhaven National Laboratory, BNL-NUREG-52118, July 1988.
plants, along with the manufacturer's recommenda-tions for maintaining equipment operability. Advanced As part of the NRC NPAR program, an assessment j
designs for uninterruptible power systems, subcomp.
was made of the characteristics of agmg and service onent improvements, and current monitoring and pro.
wear of motor control centers (M CCs). MCCs perform tective equipment are described and related to their an important function in the operation and control of a potential applicability in nuclear power plants.
large number of safety-related motors; thus the oper-ability and reliability of MCCs can affect the overall A naturally aged inverter and battery charger were safety of nuclear plants.
tested at UNL to evaluate the naturally aged condition,
.I.his report follows the NPAR strategy and invest.i-the effectiveness of condition monitoring techniques, and the practicality of selected maintenance and moni-g tes the operational peiformance, the design and manufacturing methods, and the current maintenance,
,j toring procedures. A portion of this research effort is surveillance, and monitoring techniques applied to covered in RIL No.159, " Nuclear Plant Aging Re-MCCs. A significant result described in this report con-scarch: Safety-Related Inverters," November 9,1988.
cerns the identification of important MCC failure A maintenance program for battery chargens and modes, causes. and mechanisms from plant operational inverters is recommended. As described in this report, experience. Frequencies of failures determined for the such a program incorporates inspection, monitoring, various subcomponents of MCCs are also described. In testing, and repair activities that should be performed addition, recomer.endations are provided for functional to detect and mitigate aging effects and thereby ensure indicators to monitor the performance of MCCs.These the operational readiness of this important equipment functional indicators will be evaluated during Phase 2 throughout the plant's operating life.
of the program.
NUREG/CR-5052, J. C. Higgins, R. Lofaro, NUREG/CR-5057, K. R. Hoopingarner and F. R.
M. Subudhi, R. Fullwood, and J. H. Taylor,
'Z.aloudek, " Aging Mitigation and Improved Pro-
" Operating Experience and Aging Assessment of grams for Nuclear Service Diesel Generators,"
Component Cooling Water Systems in Pressurized Pacific Northwest Laboratory, PNL-6397, Water Reactors," Brookhaven National Laboratory, December 1989.
BNL-NUREG-52117, July 1988.
The study of diesel generator aging for the NRC An aging assessment of component cooling water NPAR ' program was performed in two phases. In Phase
- (CCW) systems in PWRs was performed as part of the I, plant operating experience and' data were used to NPAR program.The objectives were to provide a tech-produce a new data base related to aging, reliability, nical basis for the identification and evaluation of deg-and operational readiness of nuclear service diesel gen-radation caused by age. 'Ihe information generated will crators. Phase 11 is chiefly concerned wit h measures for be used to assess the impact of aging on plant safety and mitigating the effects of aging.
27 NUREG-1377
,m l
l Main Citations and Summaries 1
This report proposes a detailed management, test.
failure based on performance data trends. 'lhe current ing, and maimenance program for emergency diesel practical periodic intrusive maintenance and engine generators bat ed on studies and research developed in overhauls has been found to be less favorable for ensur-phase II of this effort. The proposed program would ing safety than engine overhauls based on monitoring lead to three expected results: (1) reduction of several and trending results or on a need to correct specific en-of the stressors identified in phase I that have been gine defects. Therefore, this report recommends that shown to accelerate aging of diesel generators, (2) an the penodic overhaul requirements be reevaluated.
improved reliability and state of operational readiness, Further, an understanding of the governor, as well as of and (3) an increased confidence in the future availabil-the engine / generator, must be developed by providing ity and reliability of dicscl generators. 'Ihe proposed the maintenance staff with adequate training and moti-new program would integrate testing, inspection, moni-vation. Finally, this report recommends that engine in-toring, trending, maintenance, and other elements for spections and preventive maintenance be increased to a better approach to mitigating diesel generator aging.
mitigate the aging and wear results of the vibration The more important elements of the new proposed stressor, focusing on the engine and instrumentation program are summarized in the following paragraphs.
mounted on the engine. Vibration cannot be climi.
The current fast starting and loading requirement n ted, but its effects can be mitigated by keeping fas-for testing diesel generators can produce substantial teners/ fittings tight and by frequently recalibratmg m-harm and significant aging effects through the produc-strumentation subject to this vibration, tion of large mechanical and thermal stresses, inade.
The mission profile for the diesel generator is based quate lubrication during initial acceleration, high rotat-on a large-break LOCA with loss of all offsite power.
ing and sliding pressures, overspeeding, etc. An With over 1000 reactor-years of operation in U.S. regu-t improved testing program including slow starting and latory history without a large-break LOCA, it may be loading would induce fewer aging effects in the emer.
appropriate to redefine the mission profile for the gency diesel generator by largely climinating a unique diesel-generator with consequent benefits. Fora loss of aging stressor. In the coursc of the monthly testing pro-offsite power, with or without a small-break LOCA gram, adequate data should be collected for about 30 event, the needs for emergency electric power and the engine operating parameters discussed in this report diesel mission profile are much less stringent. In this i
that could indicate degrading performance or an im-case, the need for power can be delayed and the emer-pending component failure. For many important com-gency pow cr needs are reduced, but the need for emer-ponents, the implementation of such a program could gency power may remain for several (3 to 4) days. The detect approaching performance failure and allow prevention of station blackout appears to be the most orderly repair. Monitoring and trending will not be able realistic mission envelope.The technical requirements to detect all components with degraded performance, for the diesel generator are very high reliability with the but the deterioration that will be detected by the rec-durability to produce power until the emergency passes ommended tests is significant to aging and reliability and the reactor cooling requirements drop off, concerns. Condition monitoring and trending can pro-Acceptance of this mission envelope for the diesel-vide important indications of possible long-term com-generator system would result in a reduction of the ag-ponent or system degradation. This activity should de-ing degradation of many important engine components tect many potential component / system failures before through less harmful test requirements. In summary, a the system actually fails. Cost and safety benefits would more practical mission envelope for the diesel gencra-accrue from avoiding both equipment damage and un-tor system would include an increased start and load scheduled downtime by anticipating these failures and time (within 5 minutes), with the power level reduced
^
providing timely repair / maintenance.'Ite mor.thly test below the calculated full load (core and containment i
l program should ensure that the operating parameters sprays not needed). From an overall mission stand-listed in the report are within their maximum and mini-point, it appears that safety concerns are better served mum limits as applicable. Ilowever, it is not necessary by testing the engines for reliability rather than for to trend every parameter for effective results. When a maximum starting accelerations and very rapid loading,.
limiting (maximum or minimum) value is being ap-which do not seem necessary.
proached, the utilities should trend the approach to avoid failures and schedule repair before limits are ex-
'Ilis portion of the NPAR study was initiated to de-ceeded.
velop for NRC consideration information on potential safety problems related to the aging of diesel genera-Several recommendations were developed regard-tors. General applications of the study results were ex-ing maintenance procedures and training. One impor-
.pected for (1) improvement of dicscl reliability. (2) tant recommendation is that teardown of the dicsci modification of plant technical specifications, (3) im-engines solely for the purpose of inspection should be provement in the application of resources by the NRC avoided unless there is a definite indication that opera-and the utilitics, and (4) development of specific re-tion is degraded or there is an impending component search information needed to change some regulatory NUREG-1377 28
Main Citations and Summaries requirements. All of these end uses of the research NUREG/CR-5159 M. S. Kalsi, C. L Horst, and J. K.
have been accomplished or are under active considera-Wang," Prediction of Check Valve Performance and tion. Collectively, the safety implications of these Degradation in Nuclear Power Plant Systems," Kalsi changes and research recommendations are important.
Engineering, Inc., Sugar Land, TX, KEI No.1559, May 1988.
NUREG/CH-5141, V. P. Hacanskas, G. J.1bman, and S.
Degradation and failure of swing check valves and P. Carfagno, " Aging and Qualification Research on resulting damage to plant equipment has led to a need Solenoid Operated Valves." Franklin Research to develop a method to predict performance and degra-Center, Nornstown, PA, August 1988-dation of these valves in nuclear power plant systems.
Tests were conducted on three-way direct-acting so.
This Phase I investigation developed methods that can lenoid-operated valves (SOVs). Some SOVs had been be used to predict the stability of the check valve disk when there are flow disturbances such as elbows, aged natumily through senice in nuclear power plants, and others were subjected to accelerated aging. Ther-reducers, and generalized turbulence sources within 10 mal aging was conducted with both air and nitrogen as pipe diameters upstream of the valve. Major findings the process gas. Operational aging was simulated by include the flow vekicity required to achieve a full-putting the specimens through operational cycles at open stahic disk position, the magnitude of disk motion certain intervals during the accelerated thermal aging developed with these upstream disturbances (with flow i
with the environmental temperature controlled at a vehicitics below full-open conditions), and disk natural level representative of senice conditions. The program frequency data that can be used to predict wear and fa-also included simulation of a design basis event (DHE) tigue damage. Reducers were found to cause little or no performance degradation. Effects of elbows located that consisted of gamma irradiation and a main. steam-line-break loss-of-coolant accident within 5 diameters of the check valve must be consid-(MSLB /LOCA) simulation. After each majorsegment cred, while severe turbulence sources have a significant l
of the test program (aging, irradiation, and MSLH/
cffect at distances up to 10 diameters upstream of the valve.
LOCA simulation), some of the valve specimens were subjected to operational testing and then disassembled Clearway swing check designs were found to be par-for inspection and measurement of physical properties.
ticularly sens?ive to manuf acturing tolerances and in-stallation variabMs making them likely candidates for Performance of the Automatic Switch Co. (ASCO)
SOVs was affected in the early stages of the program by premature failure. Reducing the disk full-opening an-
~
an organic deposit of undetermined origin. Removal of gle on these designs results m significant performance the deposit climinated the problem.
mprovement.
A naturally aged ASCO SOV with Buna N seats and NUREG/CR-5181, L. C. Meyer and 1 L. Edson, a new ASCO SOV with EPDM seals were subjected to
" Nuclear Plant Aging Research: The IE Power System," Idaho National Engineering Laboratory, accelerated aging with nitrogen as the process gas.
GG-2545, May 1990.
These valves were the only ones to go through the en-tire test program without a failure to transfer and with.
This in-depth engineering study of the Class IE out any significant leakage.
Power System is conducted in accordance with the NRC NPAR program and guidelines.The report pro-Valcor Engineering Co. SOVs suffered from stick-n on of a mm s, (2) a pm-ing of the shaft seal O-rings, which made it impossible ga O) an limmary identification of failure causes due to aging to complete the accelerated thermal aging. Repeated and senice wear degradation, and (3) a review of cur-tests and changes in test procedures failed to alter this rent inspection, surveillance, and monitoring methods, i
meluding manufacturer-recommended surveillance it is possible that the strewes of accelerated aging and maintenance practices. Also, performance pa-produced effects that are not representative of senice rameters potentially useful in detecting degradation aging. Seal deterioration in the Valcor SOVs caused are identified in this report, and preliminary recom-J leakage following DHE irradiation.The naturally aged mendations are made regarding inspection, surveil-Wlcor SOV performed satisfactorily during the first lance, and monitoring methods.
high-temperature portion of the MSLH/L.OCA profile A description of a typical Class IE power system is -
but malfunctioned during most of the rest of the test.
presented for a pressurized water reactor (PWR) with Deterioration of theclastomericpartsof the ASCO
_ specific maintenance information from a cooperating-SOVs did not appear to be sufficient to account for the utility. The Class lE power systems provide electric observed failures to transfer, which evidently were power for the safety systems in the plant, including an caused by coil deterioration. Elastomeric parts of Val-emergency power source (usually diesel generators) cor SOVs, both from the naturally aged SOV and from and three subsystems: the alternating current (ac) the one that had not been aged, experienced substan-power systems, the direct current (de) power system, tial deterioration.
and the vital ac power system. Each of the major Class 29 NUREG-1377
. ~..
L
.1 l
1 Main Citations and Summaries i
4 1E power components is described, and the results of Approximately 40 IEEE standards applicable to component aging studies are s,. 'marized where appli-Class IE power systems and associated components cable. The ac power system used in typical nuclear were reviewed and tabulated. The IEEE reviews each power plants is a dual-train cascading bus system that standard approximately every 5 years.The authors rec-includes circuit breakers, transformers, relays, load ommend that aging be included in this review. Stan-centers, and motor control center switch gear. The dards provide design and application guidanct but gen-de system includes battery chargers, batteries, inver-erally do not provide specific recommendations for ters, and associated con trol breakers. The vital 120-Vac maintenance, testing, inservice inspection, and moni-loads. include the engineered safety feature cabinets toring of age-related degradation, and the reactor protection systems.
Aging research can play a supporting role in solving The review of operating experience included data outstanding safety issue,s. For example, component from the following generic data bases: Licensee Event egra non he to agmg is ondactono conWer in ik Reports (LERs), Nuclear Plant Reliability Data System yac "
(NPRDS), Nuclear Ibwer Experience (NPE), and plant maintenance data from one cooperating utility.
NUREG/CR-5192, W. E. Gunther, " Testing of a lhe LER records indicate that the Class 1E power sub-Naturally Aged Nuclear Power Plant Inverter and system failures were distributed as follows: emergency llattery Charger," Ilrookhaven National Laboratory, power generation,31.7%; medium-voltage subsystems, IlNLJUREG-52158, September 1988. _
21.2%; low-voltage ac (less than 600 V),19.8%; and de A naturally aged inverter and battery charger ob-system. 9.8% The most frequent component failures tained from the Shippingport facility were tested as weru circuit breakers,66.3%; inverters,9.9%; and bat-part of the NPAR program. The objectives of this test-teries, 9.5% The leading causes of circuit breaker ing were to evaluate the naturally aged equipment faults were mechanical malfunction,25%; electrical state, determine the effectiveness of condition-malfunction,22%; and sticking,7% The three leading monitoring recommendations, and obtain insight into 6
causes of relay faults were drift,46%; electrical mal.
the practicality of preventive maintenance and moni-function,11%; and sticking,10% The NPRDS data re.
toring methods.
view listed the Class IE power components in order of
'Ibsting indicates that the equipment has retained frequency of failure as follows: diesel engines, inver-its ability to respond to load transients. With the excep-ters, and circuit breakers.The overall fraction of Class tion of silicon controlled rectifiers (SCRs), which were IE electrical component failures related to aging was found to be operating with case temperatures (* F) 20%
32.7% However, because of system redundancy and higher than those during the acceptance test, compo-fail-safe design, only 2.4% of Class IE electrical com.
nent temperatures.and circuit characteristics were ponent failures caused total loss of system function.
similar to original acceptance test measurements.
Based on t hese observations, it is concluded that the in-Approximately 8% of all events in the NPE data verter and battery charger have not aged substantially.
base for all systems were associated with the safety electrical system. The NPE hsted breakers, motor con-The two primary monitoring techniques employed were temperature measurements and electrical trol centers, and switchgear as having the most fre-quent failures, at 36.1% Ihis was followed by inverters waveform observation. Internal panel temperat ure and and chargers,15%; diesel generators,10%; transform-individual component temperatures were recorded at -
ers, 3.4%; and batteries. 3% Ihe plant data als regular intervals during steady-state and transient op-showed that breakers caused the most work requests in erations. Thermocouples imbedded within the trans-the maintenance data base, followed by batteries, the former and inductor windings and attached to SCR and battery charger, and the generator. (llatteries were capacitor surfaces provided a nonobtrusive means of-high on the list because of frequent preventive mam-monitoring component operation. Readings taken tenance tasks such as adding water and testmg.)
were compared to original acceptance test data.
Circuit waveforms were observed on an hourly basis The review of codes and standards included general during steady-state operation and at the time load tran-design criteria, regulatory guides, and IEEE standards.
sients were applied. The inverter output voltage and There were three recommendations for regulatory the SCR gate current waveforms remained relatively.
guides:(1) Regulatory Guide 1.118 should include the constant regardless of the applied loads.
F issues of testing and inspection for the lightning protec-Finally, this test report recommends that individual tion system and power ground system (2) Regulatory fusing of filter capacitors be considered in order to pre-Guide 1.32 should address the issues of cleanliness in clude a capacitor failure in the short circuit mode from switchgear area, and (3) Regulatory Guide 1.9 should rendering the inverter inoperable. Also, equipment ac-be extended to include the problems of diesel genera-ceptance testing should be modified to obtain the most tor aging.
limiting design operating conditions for all major sub-
-i
' NUREG-1377 30
n.
l i
Main Citations and Summaries components. Results indicated that aging had not sub-
- 8. Stress the importance of aging research to the stantially affected equipment operation. On the other resolution of generic safety issues and to user hand, the monitoring techniques employed were sensi-needs identified by the Office of Nuclear Reac-tive to changes in measurable component and equip-tor Regulation to aid NRC decision-makers but ment parameters. Hus comparing the monitoring re-not to formally prioritize the components.
sults with the original acceptance test data is a viable NUREG/CR-5268, R. Lofaro, M. Subudhi, W. E.
method of detecting degradation prior to catastrophic Gunther, W. Shier, R. Fullwood, and J. ii.'lhylor, failure.
" Aging Study of Boiling Water Reactor Residual licat Removal S stem," Brookhaven National NUREG/CR-5248,1. S. Levy, D. H. Jarrell, and Laboratory, BN rNUREG-52177, June 1989.
E. P. Collins, "Prioritization of TIRGALEX-Recommended Components for Further Aging As part of ongoing efforts to understand and man.
Research " Pacific Northwest Laboratory, Science age the effects of aging in nuclear power plants, an ag.
Applications International Corp., PNL-6701, ing assessment of a vital system, the residual heat re-November 1988-moval (Ri f R) system in boiling water reactors (HWRs),
In April 1986, the NRC established the Technical was performed. This report presents the results and i
Integration Review Group for Aging and Life Exten-discusses the impact of RilR system aging on plant sion activities. In May 1987, TIRGALEX finalized its safety. The work was performed as part of the NRC plan (FIRGALEX 1987), which identified the safety.
NPAR program.The RiiR study was done according to related structures and components that should be pri-the methodology developed by UNL as part of the Ag-oritized for subsequent evaluation in the NRC NPAR ing and Life Extension Assessment Program (ALEAP) program. This report documents the results of an ex-System Level Plan. The selected approach uses two pert panel workshop established to perform the priorit-parallel work paths, one applying deterministic tech-ization activity, Prioritization was based primarily on niques and the other probabilistic techniques, to char-criteria derived from a specially developed risk-based acterize aging.
methodology that incorporates the effect on plant risk The deterministic work performed for the RilR sys- -
of component aging and the effectiveness of current in-tem study involved a review of past operating data from dustry aging management practices in mitigating that various national data bases.The data covered all oper-agmg.
ating modes of the RilR. They showed that approxi-4 An additional set of criteria was the importance of mately 70% of the failures reported were due to aging.
aging research on structures and components to the De dominant cause of failure was found to be normal resolution of generic safety issues and to identified service, while the dominant failure mechanisms were regulatory needs. The resultant categorization was wear and calibration drift. The predominant fail-used to provide additional information to decision mak-ure mode was leakage followed by loss of function and crs but was not used to calculate final rankings.
wrong signal. The data also indicated that approxi-The expert panel workshop was conducted within mately 65% of the failures were detected by the current test and mspection practices.110 wever,27% of the fail-the following ground rules:
ures were not detected until an operational abnormal-
- 1. Obtam all relevant information on aging of cur-ity occurred. This shows that currently employed rent plants (i.e., during their ongmal license maintenance and monitoring practices are not com-period),
pletcly successful in detecting all aging degradation. In
- 2. Develop an understanding of aging and its ef-evaluating the effect of failurc on RIIR performance,it fects (i.e., define the contribution of aging to was found that over 50% resulted in degraded system plant risk), -
operation, while approximately 20% resulted in a loss 4
- 3. Assess the adequacy of current industry prac.
of redundancy. Other significant effects of R11R fail-
' tices for managing component aging within ac-urcs include loss of shutdown cooling capability, radio-ceptable levels of risk, logical releases, reactor scrams, and actuation of engi-nected safety features. Actual plant records for Mill-
- 4. Evaluate theimportanceof theagingofindivid-stone Unit I were obtained and reviewed. The results ual components and component groups on plant showed consistency with data base findmgs.
- risk, 5 Apply the" Risk Significance of Component Ag-
. probabilistic work entailed the imple-C mentation of a personal-computer-based pro-ing" methodology (being developed by W. E.
gram @RAAGE-1988) developed to perform time-Vesely of SAIC under the NPAR program) to dependent probabilistic nsk assessment (PRA) the prioritization, calculations. De RHR model used was based on the
- 6. Use operational failure data +
Peach Bottom design. Time-dependent failure rates for
- 7. ' Use expert judgment through an interdiscipli-major components were developed from the data base nary panel,
- findings and were used in the program to calculate sys-31 NUREG-1377
=
Main Citations and Summaries I
tem availability and component importances for var.-
components such as switches and sensors show ous ages.The PRA results showed that, when the time-little or no increase (0 to 3% per year),
dependent aging factors are accounted for, two signifi-cant system effects are seen: (1) system unavailability Design Considerations increases moderately with age and (2) the relative im-portances of components may change with age. For
- 1. Plants with a common suction line supplying all loops of the RHR while in the shutdown cooling low-pressure coolant injection operation, miscali-mode should consider placing increased atten-bration of instrumentation was the most important tion on motor-operated valves (MOVs)in the contributor to system unavailability. However, during suction line during later years of plant life since later years, aging can cause motor-operated valves to become equally important. PRA calculations for shut-aging can increase the probability of MOV fail-down cooling operation showed these valves to be the ure and lead to a temporary loss of shutdown most important contributors to unavailability through-cooling capability. Piping and other components out plant life.
in nonredundant supply lines should also be considered.
The following conclusions resulted from this
- 2. Plants using a common minimum flow line for assessment:
two RHR pumps should closely monitor pump Aging Effects performance since aging can degrade perform-
- 1. Aging has a moderate impact on RHR comp -
ance and lead to dead-headed pump operation nent failure rates (0 to 17% per year increase) and possible failure.
t and system unavailability (2-fold to 4-fold in-The findings presented in th.is report form a sound crease in 50 years). This contribution of aging technical basis for understanding and managing the ef-effects may be attributed to two factors: (1) fects of agmg m RHR systems. The results also provide RHR is a safety system and has relatively strin, the framework for future Phase H work. Although the gent testing and monitoring requirements that time-dependent aging effects appear to be moderate identify aging degradation before performance for the RHR system, additional work is nccessary to is adversely affected and (2) the RHR system is complete the aging assessment. Since this is predome typically maintained in standby, which mini-nantly a standby system, exposure to operating stresses mizes exposure to wear-related degradation.
is limited, which could contribute to the mitigation of
- 2. Prelimina'Y comparisons of unavailability for
" "E "E" ""
standby and continuously operating systems operating time increases, the RHR system could expe-have shown that standby systems are potentially rience rapid increases in failure rates, as we found in less severely affected by agmg. Using this result previous work on a continuously operating sysicm.This as a basis, the differences in operation and man-should be addressed in future work. In additica, the agement of these two types of systems will be relatively stringent tests and inspections performed for further evaluated with the ultimate goal of de-the RHR system may contribute to the aging effects.
veloping methods that are effective in mitigating aging effects-NOREG/CR-5280, M. Subudhi, W. Shier, and E. MacDougall, " Age-Related Degradation of
- 3. Examination of plant-specific failure data has Westinghouse 480-Volt Circuit Breakers," Vol.1.
confirmed that failur e trends for certain compo_
" Aging Assessment and Recommendations for nents in some plants can differ from industry I*{mr vin {lreaker Reliability," Brookhaven averages. Although aging was found to have a
, iy ik tory BNieNUREG-52178, moderate impact on the RHR system based on average values, the impact on plants for which
,An aging assessment of the Westinghouse DS-the data differ from these averace values could series low-voltage. air circuit breakers (especially be significant. This will be addressed in future DS-206 and DS-416) was performed as part of the NRC work.
Nuclear Plant. Aging Research (NPAR) program.
These breakers are used for Class IE applications in Data Analysis.
nuclear power plants.' DS-416 breakers, in particular,.
- 1. Results have ccmfirmed that generic failure are used for reactor trip applications.The findings from rates may not accurately represent individual this study form a technical basis for understanding ag-plants for all applications. He uncertainty in ing effects in DS-series breakers, risk estimates may be reduced by updatingcalcu' lations with actual plant data.
This study was initiated following the failure of a center pole lever weld in a reactor trip breaker at the
- 2. Mechanical components in the RHR system McGuire Nuclear Station and the issuance of NRC show a low to moderate increase (8% to 17% per Hulletin 88-01 on that subject. The objectives of the year) in failure rate with age, while electrical study are to characterite age-related degradation in the l
NUREG-1377 32
Main Citations and Summaries breaker assembly and to identify maintenance practices standard quality, which could lead to their premature to mitigate degradation effects, cracking.
The design and operation of DS.206 and DS-416
'lhis program involved a commercial grade West-breakers were reviewed in detail. Failure data from inghouse DS-41610w-voltage air circuit breaker that is various operational data bases were analyzed (1) to typical of breakers used in nuclear power plants for identify all failure modes, causes, and mechanisms, (2) class IE applications. The test breaker was mechani-i to assess the effectiveness of the requirements formu-cally cycled for more than 36,000 full cycles with no lated in NRC B ulletin 88-01, and (3) to recommend ac-ciectrical load, thus accelerating the aging process that livities that would effectively detect and mitigate age-could be attributed to breaker cycles to help identify related problems in breakers. The data bases included age-related degradations.The test was conducted in ac-Licensee Event Reports (LERs), Nuclear Plant Reli-cordance with ANSl/IEEE Standard 37.50 (1981) for ability Data System (NPRDS), In-Plant Reliability the life testing of circuit breakers.Three different pole Data System (IPRDS), and Nuclear Power Experience shafts with weld configurations of approximately 60 de-(NPE). Additional operating experience data were ob-grees,120 degrecs, and 180 degrees in the center-pole tained from one nuclear station and two industrial lever (#3) werc used to characterize cracking in the pole breaker-service companics to develop aging trends for lever welds. In addition, three operating mechanism various subcomponents. The responses of the utilitics units and several other parts were replaced as they be-to NRC Hulletin 88-01 were analyzed to assess the final came inoperable, resolution of failures of welds during reactor trips.
The mechanictd cycling test resulted in the follow-The predominant failure modes in nuclear power ing conclusions on the manufacturing and aging of plants along with the causes and mechanisms of failure Westinghouse DS-series breakcrs:
were determined from the operating experience data.
1.
Pole shafts used in this test program werc Instruction manuals including schematics and manu-found to have substandard welds. This raises facturers' maintenance manuals were analyzed to questions as to the effectiveness of the quality understand the effect of material agingduring the serv-assurance program that was followed during ice life of the breakers. This analysis was augmented by welding.
technical discussions with maintenance and service 2.
Fracture of the trip shaft lever suggested that personnel from the electrical supply industry. Mainte-correct electroplating procedures may not nance recommendations by the manufacturer to have been followed.
mitigate age-related degradation, suggestions for 3.
The sharper bends at the neck of the hookson improving the monitoring of age,related degradation, newly purchased reset springs-compared to and inputs from NRC mspectors mvolved in assessmg an older design-led to early spring failures.
breaker problems m the nuclear mdustry were reviewed.
4.
The hardness of the oscillator surface on newly procure.mits'was 30% less than on Volume 2 of this report presents the results from a older units.
test program to assess degradation m breaker parts 5.
Wear, fracture, distortion, and normal fatigue through mechnical cycling that simulated the operating life of nuclear plant breakers.
dommated the aging process, with wear being the largest contributor.
NUREG/CR-5280, M. Subudhi, E. MacDougall, S.
6.
Excessive wear was evident in the ratchet Kochis W. Wilhelm, and H.S. Lee, " Age-Related wheel, holding pawls, oscillr. tor, drive plate, Degradation of Westinghouse 480 Volt Circuit motor crank and handle, cam segments, main Hrcakers," Vol. 2,"Mcchanical Cyctmg of a DS-416 roller, and stop roller.
Breaker. Test Results," Brookhaven National Labo-ratory, HNL -NUREd-52178, November 1990.
7.
Structural components and contact assembly
' After the McGuire event in 1987 invoMng failure
. pans showeQew dects of aging due to me-cham, cal c#ng.
of the center-pole weld in a reactor trip breaker, the 8.
A pole shaft with a reduced size weld could fail NRC initiated an investigation of the probable causes.
During the last decade, NRC has issued a number ofin-at as few as 3000 cycles.
formation notices and bulletins pertaining to problems The testing yielded many useful results. The encountered in Class 1E breakers. A review of operat-burned-out closing coils were found to be the result of ing experience suggested that _ burned-out coils, binding in the linkages that are connected to this de-jammed operating mechanismsiand deteriorated con-vice. Among the seven welds on the pole shaft, #1 and tacts were the dominant causes of failures. Although
- 3 were the ones that cracked first and caused misalign-failures of the pole shaft weld were not included as one -
ment of the pole !cvers, which, in turn, led to many of the generic problems, the NRC Augmented Inspec-problems with the operating mechanism, including tion Team had suspected that these welds were of sub-burned-out coils, excessive wear in certain parts, and 33 NUREG-1377
~
l i
1.
0 Main Citations and Summaries i
overstressed linkages. Based on these findings, a main-The minimum CVN impact energy after long-term tenance program designed to alleviate the age-related aging has been found to be proportional to the square degradations caused by mechanically cycling this type of the fraction of ferrite, the mean ferrite spacing, and a of breaker is suggested.
chemical-composition parameter. This model should be developed further for application to the assessment NUREG/CR-5314, C.E. Jaske and V.N. Shah, " Life of components. A time-temperature parameter can be Assessment Procedures for Major LWR Compo.
used to define lower-bound trends to the available im-nents: Vol. 3, Cast Stainless Steel Components,"
pact energy values for cast stainless steels as a function Idaho National Engineering Laboratory, EGG-2'562.
of chemical composition and thermal exposure time.
October 1990.
The report proposes a model using that parameter to Many critical pressure boundary components in predict the impact energy decrease for any particular lot f cast stainless steel. This predicted impact energy commercial light water reactors (LWRs) are made of cast stainless stecis. Life assessment procedures are value or the predicted minimum impact energy value is needed for these components because cast stainless then used to estimate fracture toughness from correla-steels are subject to thermal embrittlement during tions between impact energy and fracture toughness at long-term service at LWR temperatures. The compo.
both room temperature and 290 C. Dis approach nents of concern include pump bodies, reactor coolant should prov.de a conservative estimate of fracture piping and fittings, surge lines (in a few plants), pres.
toughness Qr use in assessing the structu ral integrity of surizer spray heads, check valves, control rod drive cast stair.iess steel components, mechanism housings, and control rod assembly hous-ings. These are made of grade CF-8, CF-8A, or Insevice inspection (ISI) is needed to define type, CF-8M stainless steel in U.S. LWRs; grade CF-3 stain-size, and ocation of any defects in cast stainless steel l
less steel is used in some foreign LWRs. The purpose of c mponents so that their structural integrity can be this project was to review the availabic data on thermal cv luated. Use of radiography during ISI is less practi-embrittlement of cast stainless stects and to develop cal than during fabrication. Conventional ultrasome updated procedures for life assessment by key LWR testing (UF) methods for detecting flaws are not reli-cast stainless steel components.
able in cast stainless steel components because its coarse grain structures result in a low signal-to-noise Cast stainless steels have a two-phase microstruc*
ratio. Advanced Ur methods being developed have ture consisting of ferrite islands in an austenite matrix.
shown an improved capability to detect flaws in cast -
With long-term exposure to LWR temperatures, other stainless steel components and have been used in sev-phases form in the ferrite phase that cause it to become cral PWR plants. Because of the difficulties with radi-hard and brittle, while the austenite remains ductile. If ography and Ur methods in detecting and sizing flaws, the amount of ferrite is small and if it is distributed evenly and finely throughout the austenite, the proper-the application of the acoustic emission technique te detecting crack growth in cast stainless steel needs to ties of the casting are not significantly affected by the
~
be evaluated
- thermal embrittlement of the ferrite. However, as the amount of ferrite,its coarseness and its uneven distri-He report outlines a procedure developed for esti-bution increase, the increased thermal embrittlement mating the current condition and residuallife of key of the ferrite adversely affects the properties of the LWR cast stainless steel components.The procedure is casting-implemented in nine major steps.ne first three steps The propedies most affected by thermal embrittle.
involve the collection, examination, and storage of
, ment are Charpy V-notch (CVN) impact energy and records for fabrication and construction, inservice in-fracture toughness (J c). Both of these properties de.
spection, and operating history. The fourth step in-i crease as the degree of thermal embrittlement in-volves a conservative fatigue and fracture mecham,es creases. If these values become toolow, the structural ev luation to determine the worst-case flaw size and integrity of a cast stainless steel component could be se.
the minimum required fracture toughness at the end of riously impaired. Presently, more fatigue-crack-growth the next operating period. In the fifth step, the current data are needed for CF-8 and for all cast stainless condition of the material is assessed using a proposed stecis in the high-cycle regime. Thus, for life assess.
analytical model, microstructural data, or measured ment of cast stainless steel components, the main con.
properties (or some combination of the three). In the cern is loss of fracture toughness and impact energy.
sixth step, the results of the fourth and fifth steps are Data and engineering models have been developed to combined to evaluate the structural integrity of the help predict the degree of embrittlement as a function component. The seventh step establishes what actions of thermal exposure history. For reactor internals, irra-(none, repair, replace, or shut down) are to be taken, diation history may also be a concern.
and the eighth step establishes the plan for the next ISI.
NUREG-1377 34
Main Citations and Summaries In the ninth step, the component is reevaluated and the sumed to apply to all other designs in use for at least steps are repeated as needed.
two reasons:
- 1. 'Ihere are a large number of diverse designs in NUREG/CR-5334, D. B. Clauss, " Severe Accident use. In particular, assemblies manufactured Testmg of Electncal Penetration Assemblics,"
prior to 1971 were not subject to national stan-
)
Sandia National Laboratories, SAND 89-0327, dards and were often manufactured in the field, November 1989 whereas the three tested in this program were Since the Three Mile Island incident, the risk and subject to rigorous quality assurance and were consequences of severe accidents have been a major fo-designed to meet the standards of IEEE cus of reactor safety research.The performance of the 317-1976 and IEEE 323-1974.
i containment building has a significant effect on acci-dent consequence; hus considerable effort has been
- 2. The leak potential is highly dependent on the directed toward understanding and predicting the func.
temperatures to which the assembly is sub-tional failure of containments. The ccmtainment pres.
jected. As research continues and more analyses j
sure boundary typically includes numerous mechanical of severe accident sequences are conducted, the and c!cctrical penetrations, each of which represents a
" worst-case" loads may change. 'Iherefore, the.
potential leakage path.
leakage potential must be reevaluated as the understanding of severe accident loads is Several studies completed in the early 1980s indi-impm Heat transfer effects must be consid-cated that electrical penetration assemblics could be an ered to determ, e the temperature of the m
important leak path that merited further study. A re-utboard containment seals, which end up con-port by the Oak Ridge National Laboratory on severe accident sequence analysis for BWR Mark I contain-twiling the potential for leakage, ments concluded that the temperatures in the drywell The results of these tests should not be construed as were high enough to possibly cause failure of the seals
. suggesting that all designs will not leak under severe ac-t that could result in leakage. NUREG-0772 identified cident conditions; the performance of all components electrical penetration assemblies as having "one of the of the containment pressure boundary must be evalu-largest uncertainties associated with predicting the ated on a case-by-case basis with all loads considered.
amount of radionuclides released. These studies pro-The performance is also affected by thermal and radia-vided the major impetus for NRC to imtiate a research program on these assemblies. Sandia National Labora-tion aging. Given good information on tpe contamment loads, a heat transfer analysis to determme ie :mproxi-tories managed a program to conduct a background ate temperature profiles, knowledge of the time-study and to recommend and perform tests to generate data that could be used to assess the leak potential temperature thresholds for the sealant materials med, when the assemblics are subjected to severe accident and the proper exercise of engineering j udgmea, a rea-conditions. The>c tests are described in this report.
s nable evaluationof theleakage pientialofotherde-signs can be made.
Electrical penetration assemblics are used to pro-
. vide a leak-tight pass-through in nuclear power plant The electrical performance of the assernblics was containment buildings for electrical cables with power, monitored in these tests by measuring the insulation j
control, and instrumentation applications. The design resistance and electrical continuity of the conductors.
has evolved to a modular concept that consists of elec-The resistance degraded rapidly during the severe acci-trical conductors contain, d within stainless steel tubes dent tests, although the rate depended more on the (modules) that are sealcu.
type of cable and loads than on the particular design be-ing tested. Under the specific severe accident condi-Three designs, D. G. O'Brien, Conax, and Westing-tions that were simulated, the data suggest that all elec-house, were tested under simulated severe accident trical systems supplied in the Westinghouse assembly-conditions for a PWR, a BWR Mark I drywell, and a w uld have functioned for about 4 days; those supplied BWR Mark III wetwell, respectively, to generate engi_
in the D. G. O'Brien would have functioned for about neering data (leak rate, temperature, insulation resis.
tance, and electrical continuity) for assessing their leak 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and those supplied in the Conax may have potential. None of the assemblics leaked during the se.
functioned for only about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. Some cables would vere accident tests, which can be attributed to the use of be expected to function beyond these times. However, redundant seals and to the fact that the outboard con-it must be noted that conclusions regarding the electri-tainment seats in all three designs were never exposed cal performance of systems inside the containment to temperatures that exceeded the service limits of the building based solely on insulation resistance data must seal materials. The exceptional leak integrity of the-be made with caution.The perfonnance of the electri-three designs tested in this program should not be as-cal systems would depend on the voltage, current, and 35 NUREG-1377 l
s Main Citations and Summaries l
impedance requirements for a specific cond uctor appli-
- 3. 1oprovide a means to evaluate the effectiveness cation.
of maintenance on mitigating aging-degradation phenomena.
NUREG/CR-5378, A.J. Wolford, C.L Atwood, and W.S. Roesener, " Aging Data Analysis and Risk
- 4. 1b producc an inspection plan that optimizes the Assessment-Development and Demonstration effectiveness of insPcetions based on System risk Study,"2567, August 1992. Idaho National Engineering Laboratory, reduction.
EGG-
- 5. Tb use the information generated by this assess-This work develops and demonstrates a probabilis.
ment to resolve related generic issues and pro-tic risk assessm ent (PRA) approach to assess the effects vide guidance for developing regulatory criteria of aging and degradation of active components on plant on agmg and life extension.
risk. The work (a) develops a way to identify and quan-The following approach was used during the initial tify age-dependent failure rates of active components phase of the assessment:
and then incorporate them into PRAs, (b) demon-
- 1. Perform a literature search of government and strates these tools by applying them to a fluid-mechani-private sector reports that are related to service cal system, using the key elements of a NUREG-1150 water, aging-related degradation, and potential PRA. and (c) presents them in a step-by-step approach, methodologies for analysis, to be used for evaluating risk significance of aging phe-
- 2. Assemble a data base that contains a listing of '
nomena in systems of interest.
the configurations, characteristics, and water Statistical tests are used for detecting increasing sources for the service water systems in all com-failure rates and for testing data-pooling assumptions mercial nuclear power plants in the U.S.
and model adequacy. The component failure rates are
- 3. Obtain and examine the available service water assumed to change over time, with several forms used data from large generic data bases, i.e., the Nu-i to model the age dependence-exponential, Weibull, clear Plant Reliability Data System, Licensee and linear. Confidence intervals for thc age-dependent Event Reports, Nuclear Power Experience,in-failure rates are found and used to develop inputs to a spection reports, and other relevant plant refer.
PRA modelin order to determine the plant core dam-ence data. Analyze the service water system of a age frequency. This approach was used with plant-specific power plant for aging-related degrada-specific data, obtained as maintenance work requests, tion phenomena from the available data ob-for the auxiliary feedwater system of an older pressur-tained from this data base.
ized water reactor. It can be used for extrapolating
- 4. Performafaull-trecanalysisof theservicewater present trends into the near future and for supporting system of a typical plant to examine failure risk-based aging-management decisions.
propagation and determine specific input re-NUREG/CR-5379, D. H. Jarrell, A. H. Johnson, Jr.,
quirements of probabilistic risk analyses.
P. W. Zimmerman, and M. L Gore, " Nuclear Plant
- 5. Develop an m depth questionnaire protocol for Service Water System Aging Depradation Assess.
examining the information resources at a plant ment: Phase 1," Vol.1 Pacific Northwest Labora-where such resources are not available in the tory, PNL-6560, June 1989, standard data bases. Subsequently, visit a nu-The service water system represents the final heat transfer loop between decay heat generated in the nu.
. clear power plant and solicit the required infor-mation.
clearcore and the safe dispersal of that heat energyinto
- 6. Analyze the information obtained from the in-the environment.The objective of this assessment is to depth plant interrogation and draw contrasts demonstrate that aging phenomena in the service and conclusions in regard to the data base.
water system can be identified and quantified so that
- 7. Use the plant information to perform an interim aging degradation of system components can be de-assessment of degradation mechanisms and to tected and mitigated before the system availability is re-focus future investigations.
duced below an acceptable threshold. The following The following is a summary of the conclusions of the goals of the assessment were directly derived from the assessment to date:
NRC NPAR program plan:
- 1. Aging-related degradation of open service
- 1. Tb identify the principal aging-degradation water systems, i.e., systems that have a direct mechanisms, to assess the,r impact on opera-interface to raw water without chemical control, i
tional readiness, and to provide a methodology in nuclear plants is prevalent and constitutes a for mitigating the effects of service water system valid safety concern. Based on actual plant data, aging on nuclear plant safety-the primary degradation mechanism found in
- 2. "10 examine the current surveillance specifica.
the open systems is corrosion. compounded by tions and evaluate their ability to provide accu-the accumulation of biologic and inorganic ma-rate reliability information.
terial. This conclusion directly contradicts the NUREG-1377 36
.- ~
Main Citations and Summaries 1
results of a failure analysis performed using in-to effectively control the water chemistry properties formation obtained from the NPRDS data base, when possible and to use biocidal agents where which indicated that the torque switches of necessary.
motor-operated valves were the prime cause of system failure.
A methodology for producing a complete root-cause analysis was developed as a result of needs identi-
- 2. Based on multiple plant samplings, the current fied in the Phase I assessment for a more formal proce-level of surveillance and postmaintenance test-dure that would lend itself to a generic, standardized ing performed on the system is not sufficient to approach, it is recommended that this, or a similar accurately trend or detect system degradation methodology, be required as a part of the documenta-due to aging phenomena.
tion for corrective maintenance performed on the
- 3. While postmaintenance surveillance does give safety-related portions of SWSs to provide an accurate -
some measure of the effectiveness of system f cus f r effective management of agmg, modification and repair efforts, sufficient infor-NUREG/CR-5383, H. M. Hashemian, K. M. Petersen, mation on momtormg operational condition and R. E. Fain, and J. J. Gingrich, "Effect of Aging on postmaintenance testingis not available to char-Response Time of Nuclear Plant Pressure Sensors,"
acterize more precisely the effectiveness of Analysis and Measurement Services Corporation, maintenance.
Knoxville, TN, June 1989.
- 4. 'Ib improve the accuracy of data to a point that A research program was initiated to study the ef-would allow a high degree of confidence in the fects of normal aging on the dynamic performance of analysis of aging degradation, a root cause logic safety-related pressure transmitters (i.e., sensors) in scheme needs to be developed for use in defin, nuelear power plants. The project began with an ex-ing the depth of knowledge and the documenta-penmental assessment of the conventional and new tion required to accurately characterize an testing methods for measuring the response time of pressure transmitters. Th,s was followed by developing aging-related component failure.
i a laboratory setup and performing initial tests to study
- 5. Clearresolution of relevant aging-related safety the aging characteristics of representative transmitters I
issues will require the specification of additional of the type used in nuc! car power plants.
documentation of failure data and regulatory requirements to ensure adequate safety margin There is need to ensure that the current testing under aged or extended life conditions.
methods, regulatory requirements, and industry stan-dards and practices are adequate to track age-related i
NUREG/CR-5379, D.B. Jarrell, L.L. Larson, degradation. The project examined the validity of the R.C. Stratton, S.J. Hohn, M.L. Gore, " Nuclear vailable methods for response-time testing of pres.
Plant Service Water System Aging Degradation sure transmitters and reviewed the historical data for Assessment," Volume 2, Pacific horthwest Labora-esidence of performance degradation problems or tory, PNL-7916. October 1992.
trends. Current intervals for response-time testing and The second phase of the aging assessment of nu.
calibrating pressu re transmi.ters are based on refueling clear plant service water systems (SWSs) was per-schedules. apparently for two reasons:
formed by the Pacific Northwest Laboratory to support
- 1. There is no method available for on.line calibra-thc NRC's Nuclear Plant Aging Research (NPAR) pro-tion of pressure transmitters, and, until re-gram.The SWS was selected for study because of its es.
cently, response-time testmg could not be per-sential role in the mitigation of and recovery from acci.
formed on line.
dent scenarios involving the potential for core melt,
- 2. The available data basc of degradation ratesand -
and because it is subject to a variety of aging mecha-trends is not sufficiently reliable to justify test-nisms. 'the objectives of the SWS task under the NPAR ing intervals longer than one refueling cycle <
program are to identify and characterize the principal While testing based on refueling intervals may be age related degradation mechanisms relevant to this adequate, there is concern that the rate of degradation system, to assess the impact of aging degradation on op-of pressure transmitter performance may increase as erational readiness, and to provide a methodology for the current generation of plants becomes older. Fur-the management of aging on the service water aspect of.
thermore, on-line testing methods based on new tech-nuclear plant safety.
nologies are becoming available to permit more fre-The primary degradation mechanism in the SWSs, quent testing of transmitters and to predict incipient as stated in the Phase I assessment and confirmed by
- failuresflhese considerations have motivated research '
the analysis in Phase 11, is corrosion compounded by such as that covered in this report to ensure that practi-biologic and inorganic accumulation. It then follows -
cal test methods and adequate test schedules are used j
that the most c!fective means for mitigating degrada-to verify proper and timely performance of safety-l
' tion in these systems is to pursue appropriate programs system pressure transmitters in nuclear power plants.
l 37 NUREG-1371
Main Citations and Summaries The project included a search of the licensee event
- 3. Review ofRelated Studies. All published experi-report (LER) data base for pressure-sensing system mental work on aging of pressure transmitters problems and reviews of Regulatory Guide 1.118 and of has concentrated on the effects of aging on static the industry standards on performance testing of pres-performance of the transmitters as opposed to sure transmitters.The following conclusions have been the dynamic performance reported herein.The reached:
related studies concluded that aging affects the
- 1. Five reasonably effective methods are available performance of pressure transmitters and that for response-time testing of pressure transmit.
temperature is the dominant stressor. Most of ters in nuclear power plants.These methods are the studies on performance of nuclear plant referred to as step test, ramp test, frequency Pressure transmitters were sponsored by the test, noise analysis, and power interrupt test.
NRC.The only other major work was performed
'fwo of the five methods (noise analysis and by manufacturers for environmental and seismic power interrupt test) have the advantage of pro.
qualification of transmitters. Ilowever, the viding on-line measurement capability at nor.
transmitter qualification data are not sufficient mal operating conditions.
to address normal agmg.
- 2. The consequerices of aging at simulated plant The aging research covered in this report was a fea-conditions were calibration shifts and response.
sibility study; it used accelerated aging to accommodate time degraiatian, the former being the more the short (6 months) duration of the project. Since ac-pronourceci problem.
celerated aging does not necessarily simulate normal S ng, the aging results in this report must be viewed as i
- 3. The LER data base contains 1,325 cases of re.
preliminary. Furthermore, this study was concerned ported problems with pressure-sensing systems with the performance of the portion of the pressure-over a nine-year period (1980-1988). Potential sensing system and electromes kicated in the harsh en-age-related cases account for 38% of the re-vir nment of the plant; the power supply and other ported problems in this period. A notable num-components of the pressute-sensmg channel that are ber of LERs reported problems with bhyckages, h)cated in the control room, cable spreading room, or freezing, and void (bubble) formation in sensing ther mild environments were not studied.
lines.
- 4. Regulatory Guide 1.118, IEEE Standard 338, NUREG/CR-5386, D. P. Ilrown, G. R. Palmer, E. V.
and ISA Standard 67.06 can benefit from minor Werry, and D. E. illahnik " Basis for Snubber Aging, recommended revisions to account for recent Research: Nuclear Plant Aging Research Program, advances in performance testing technologies Pacific Northwest Laboratory, Lake Engmeenng and from new information that has become mpany, Wyle Laboratones, PNL-6911, January.
available since these documents were initially generated.
This report proposes a research plan to address the safety concerns of aging in snubbers used on piping and The six-month study of the dynamic performance of larg equipment in commercial nuclear power plants.
pressure transmitters covered the following areas:
The proposed program will provide the structure for 1, Assessment ofResponscJime Testing Methods.
the Phase 11 Snubber Aging Study for the NRC NPAR An experimental assessment of the five meth-program, to be performed at nuclear power plants and ods mentioned above involved laboratory test-in test laboratories. This research would be an exten-ing of more than twenty pressure transmitters sion of the work performed by the Pacific Northwest with all five methods. Results showed that the Laboratory (PNL)in the Phase I Snubber Aging Study, methods are equally effective but vary widely in the primary objectives of which were to conduct an int difficulty of implementation in nucicar power itial aging assessment of snubbers and to evaluate the plants. 'lko of the five methods (noise analysis concept of reducing the number of snubbers in com-and power interrupt test) can be performed re-mercial nuclear power plants. Although snubber re-motely on installed transmitters while the plant duction programs may reduce their total population by is at normal operating conditions.
50 to 80%, this will not mitigate the concern for manag-
- 2. Aging Study Laboratory research on aging was ing the aging of the remaining snubbers. Indeed, the re-initiated and preliminary results were obtained.
maining snubbers may become more important to plant The work involved resp (msc-time testing and safety than the original population. The proposed calibration checks of a number of transmitters Phase 11 research work is based, in part, on a study of '
after exposure to heat, hu midity, vibration, pres-snubbers in U.S. nuclear power plants by the Lake En-sure, cycling, and overpressurization conditions, gineering Company conducted for PNL under the The effect of these conditions was an increase in NPAR program. A survey of U.S. utilities conducted response time and calibration shifts, the latter for PNLby Wyle 12boratories on the use of snubbersin being the more pronounced problem.
nuclear plants was also used to identify research needs.
1 l
38
~
=
u..
as.
a w
a.
-2 ua a.
. =, _
Main Citations and Summaries The following are key elements of the proposed sources and modes of failure within the AFW system, snubber research:
(2) identify currently applied means of detecting known
- 1. Review of existing service data, sources and modes of degradation and failure, and (3) evaluate the general effectiveness of current monitor-
- 2. Development of service-life monitoring guide-ing practices and identify specific areas where enhance ~
- lines, ments appear needed.'
- 3. Evaluation of the effects of compression set in The report reviews historical failure data available hydraulic seals' from the Nuclear Plant Reliability Data System, Licen-
- 4. Evaluation of accelerated methods for predict-see Event Report Sequence Coding and Search Sys-ing seat life, tem, and Nuclear Power Experience data bases.1he i
- 5. Identification of seals most affected by aging.
failure histories of AFW system components are con.
The benefits to be derived from the research are sidered from the perspectives of how the failures were principally safety related, including enhanced failure detected and the significance of the failures. Results of prediction and seismic protection of safety-related pip.
a detailed review of operating and monitoring practices ing and equipment, mitigation of snubber aging effects, at a plant owned by a c(x)perating utility are presented.
reduction of staff radiation exposures, and reduction of General system configurations and pertinent data are rad waste. Numerous technical benefits are also ex.
provided for Westinghouse and Babcock and Wilcox pected, including the identification of aging trends,in.
- units, formation useful in developing guidelines for monitor-
'lhe report includes an identification of the general ing service life, the technical bases for determining types of AFW system design configurations, an analysis service life, the effects of compression set in seals, and of historical failure data, and a detailed review of a co-improvements in snubber design, materials, and operating utility's AFW system design and their cu-maintenance. Regulatory benefits anticipated include rent operating and monitoring practices.
contributions to Standard Review Plans, Regulatory Historically, and particularly since the lhree Ms Guides, Plant Technical Specifications, and ASME/
Island 2 accident, the AFW system has been recognized ANSI OM-4 Standards based on the broader, more as critical to successful mitigation of plant transients comprehensive data base that would be developed-and accidents. In recent years, operating incidents in-The research proposed is designed to address the volving failures of AFW system components have been
~
following questions about the aging of mechanical and among the leading events identified in NUREG/ '
hydraulic snubbers:
CR-4674, Vols.1-8, " Precursors to Potential Severe
- 1. 1Iow do snubbers age and degrade?
Core Damage Accidents," in which the leading risk-
- 2. What are the failure characteristics of snubbers?
significant events are identified for several calendar years. In the years 1984 through 1986, seven of the top
- 3. What are the safety implications of eubber ten events at PWRs, from a core damage risk stand-aging?
point, involved partial or total failure of the AFW sys-
- 4. What technical information is needed to im-tem. Operational problems with these systems have prove the performance and life expectancy of been diverse in nature. The report lists six events re-snubbers?
sulting in NRC Bulletins and Information Notices as The results will contribute toward more reliable examples of the diverse types of failures involving the and predictable snubbers in the nuclear power industry ARV systems. Numerous other operating experiences
. and 'thus will improve nuclear plant safety. Implemen'.
have resulted in feedhack to the industry through both t
tation of the research plan will also provide a data base the NRC and the Institute of Nuclear Power Opera-for use in addressing regulatory and snubber technol.
tions (INPO).
ogy issues.1he data base will be made available to nu -
In reviewing the role that aging plays in failures
- clear utilities, snubber manufacturers. snubber service such as those of AFW systems, three important points companies, and the NRC. Planned interfaces will en-must be considered. First, a combination of factors, in-sure technology transfer to utilities and manu facturers.
'cluding design, maintenance, operation, aging, and other considerations may be involved. These factors a not necesMy in&gndent M m anomen NUREGICR-5404, D, A. Casada, " Auxiliary Feedwater Second, systems age ' nly as the individual compo-System Aging Study," Vol.1. Oak Ridei: National o
I'aboratory, ORNL-65tWVI, March IWO.
nents age, Other studies performed under the NPAR
~
This review of the auxiliary edwater (AFW) vys.
program address important components within the tem used at pressurized water reactor (PWR) plams AFW system and discuss the aging stressors for these has been conducted nnder the auspices of the NRC individual componentsc NPAR program. The primary purposes of the review Third, a study performed by IN EL reviewed histori.
were to (1) determine the potential and historical'
. cal failure data from the Nuclear P! ant Reliability Data i
39 NUREG-1377
Main C tations and Summaries System (NPRDS)and made judgments as to whether or tested or receive less than thorough testing. It appears not individual failure episodes were related to aging.
that improved testing requirements are needed in or-Hecause of the above three points, the ORNL ap.
der to reduce excessive testing while at the same time proach to the AFW system study has been to focus ensuring that thorough performance verification is attention on how and to what extent the various AFW conducted periodically.
system components fail, how the failures have been and NUREG/CR-5404, LD. Kucck, " Auxiliary Feedwater can be detected, and what is the value of existing testing System Aging Phase I Follow-on Study," Volume 2, requirements and practices, rather than attempting to Oak Ridge National Laboratory, ORbL-6566/V2, focus on the extent to which aging (versus design or July 1993.
operating practices, for example)is responsible for fail-The Phase I study found a number of significant ure or degradation.
Auxiliary Fecdwater System functions that were not tested and verified operable by periodic surveillance An analysis of historical failure data involving AFW testing. In addition, the Phase I st udy identified compo -
systcms was completed by a detailed review of an exist" nents actually degraded by the periodic surveillance ing AFW system and the associated monitoring prac-tests. Thus, it was decided that this follow-on study tices of a cooperating utility. The single largest source would not deal with aging assessments or in situ exami-of AFW system degradation, based up(m the analysis of nation but would instead focus on the testing omissions historical failure data, is the turbine drive for AFW and equipment degradation found in Phase 1.
pumps. It should be noted that the turbine proper has in this follow-on study, the deficiencies in current been a relatively reliable and rugged piece of equip-monitoring and oprrating practice are categorized and ment. Ilowever, the turbme auxihartes, meludmg the evaluated. Areas of component degradation caused by governor control and the trip and throttle valve, have current practices are discussed. Recommendations are contributed substanttally to the overall turbine made for improved diagnostic methods and test proce-pro ems.
dures that will ver fy operability'without degrading The sum of the failures of motor operators and air equipment, operators for valves resulted in approximately the same NUREG/CR-5406, K.G. DeWall and R. Steele, Jr.,
number of AFW system degradations as did failures of "HWR Reactor Water Cleanup Svstem Flexible
~
the turhine drives alone. Pump failures and check valve Wedge Gate Isolation Valve Qualification and High failures were also significant contributors to system Energy Flow Interruption Test," Vol.1, " Analysis degradation.
and Lonclusions," Idaho National Engineering Laboratory, EGG-2569, October 1989.
For cach type of component and for the various Recent testing sponsored by the Nuclear Regula-sources of component failurcs, the methods of failure tory Commission (NRC) showed that, for at least some detection were designated and tabulated. The most gate valves installed in nuclear applications, the equa-notable feature of this aspect of the study was that tions used by industry to size the valve operators do not 3
failures related to instrumentation and control domi-conservatively calculate the thrust needed to close the nated the group of failures that were detected during
. valves under design basis loadings..The tests also i
demand conditions (as opposed to failures detected as showed that the results of in situ valve testing at lower l-the result of periodic monitoring or routine observa-loadings cannot be extrapolated to design basis load-j.
tions made by operators or other personnel). Many of ings. This volume describes the testing conducted by the potential failure sources that were not detectable the Idaho National Engineering Laboratory (INEL) to i
I by the current monitoring practices were related to the provide technical data for the NRC effort regarding instrumentation and control portion of the system-Generic issuc 87 (GI-87)" Failure of IIPCI Steam Line It was also observed that a number of cimditions Without Isolation." The test program also provides in-fonn don applicable to Generic Issue II.E.6.1, in Situ related to desian basis demands are not being periodi-cally verified. 5xamples of these include pump capaci-R' sting of Valves, and a related document, IE Bulletm 85-03. Motor Operated Valve Common Mode Fail-A tics'not being verified at design flow / pressure condi-tions, turbines not being verified to be capable of deliv-ures Durmg Plant Iransient Due to Improper Switch
'8" "E'
ering required torque at low steam pressures, various control sequences not being checked, and automatic Of the three boiling water reactor (BWR) process pump suction transfers not being tested.
lines covered by GI-87, an unisolated break in the reaci tor water cleanup (RWCU) supply line was selected for Another observation made was that some compo the first phase of testing because such a break would nents or certain parts or aspects of components appear.
have the greatest safety impact. All three GI-87 proc; to be tested in excess of what failure history indicates to ess lines have common features: all communicate with be appropriate. On the other hand, other aspects of.
the primary system, pass through containment, and certain parts of the AFW systems arc either never have normally open isolation valves.
1 NUREG-1377 40
Main Citations and Summaries To meet the new valve operating criteria required by tests, the results cannot be extrapolated to design basis IE 11ulletin 85-03 and Generic Letter No. 89-10, indus-conditions because the final thrust varies depending on try developed new diagnostic test equipment and meth-the extent to whichdisk friction rather than disk seating ods for in situ motor-operated valve (MOV) testing. IE affects the torque switch.
Bulletin 85-03 succeeded in significantly improving the lhe disk factor of 0.3 typically used in industry to operability of the selected safety-related valves be-calculate disk friction force is not conservative far cause it caused many of the utilities to reanalyze the de-either of the valves tested. A disk factor of 0.5 margi+
sign basis load for the applicable MOVs and to reset the ally predicts the forces for one valve during both gen-control switches accordingly.
ing and closing. The response of the other vake is en-Ilowever, very little design basis testing of valves veloped by the 0.5 disk factor during opening but not j
has been conducted outside the plant to verify the ana.
during closing. lbday's tools for analyzing valve re-lytie assumptions used to determine valve switch set _
sponse to fluid loadings are not sophisticated enough to tings. Analytic assumptions are necessary because, in detect small design differences that make large re-many cases, the utility cannot test valves at design basis sponse differences. Temperature also affects the thrust loadings in situ.1hc GI-87 testing provides some nf t he requirements of these gate valves.
first measured valve responses with which industry's All the facts listed justify continued qualification valve operator sizing equations can be compared.
testing of prototypical valves at design basis loadings and stress the need for industry to add new terms to the In this initial test program, two representative equa n or to increase the disk factor to a very conscr-RWCU isolation valves were subjected to the hydraulic V ive numkr to account for the missing terms in the qualification tests described in ANSI H16.41, the quali-equ tion. Also, test results show that the stem factoris fication standard for nuclear valves, and then to flow not a constant but changes with stem load, thus making interruption tests at full RWCU pipe-break flow. In all, it v ry difficult to extrapolate normal in situ valve test-fourteen flow interruption tests were performed, ten ng to gn s conditions.
on Valve A and four on Valve H. In the Valve A tests, When tests or irnproved sizing equations have de-the parametnc study included varying both the degree ter med the thrust needed to operate a valve at its de-
)
of inlet water subcooling and the pressure. The four sign basis loading, utdities can use one of several mod-
)
Valve B tests were all performed at a normal HWR ern diagnostic systems to conservatively set the motor 10 F subcooling, and only the inlet pressure was operator control switches. However, this method may varied.
exceed the allowable thrust on some valve designs.This Test results show that, for both valve designs tested, job will be made casier and the result wdl be more con-the force required to open and close the valves at tem-servative if both the torque and the thrust can be meas-peratures above 100 F was significantly higher than the ured when the switches are set. If further research force predicted by the valve manufacturers. Only dur-proves that there is a proportional relationship be-ing the valve-opening tests at room temperature with-tween stem load and stem factor, the degree of conser-out flow did the typical industry valve thrust equation vatism can be reduced.
predict the valve response. Industry has assumed that NUREGICR-5406, K.G. DeWall and R. Steele, Jr.,
the valve-opening thrust requirements would be the "BWR Reactor Water Cleanup System Flexible highest when the disk hfted off the seat. This was deter-Wedge Gate Isdation Valve Qualification and fligh mined not to be true for the valves tested.The highest Energy Flow Interruption Test," Vol. 2, " Data opening loads (maximum thrust) with flow occurred at Report," Idaho National Engineering Laboratory, different openings for both vahes, but in both cases, EGG-2569, October 1989.
they were well off their respective seats. Valve-closing This volume presents the 700 pages of actual meas-thrusts at full line-break flows were higher (up to one ured data from the gate valve test program described in third) than anticipated-Volume 1. They are provided for those readers who The test results provide evidence for two concerns wish to hiok at the data and form their own opinion on with MOVs in n uclear power plants. First, proper sizing the performance of the test valves. For those readers who wish to do their own analysis, the electronic data of motor operators is complicated by the fact that the equation used for calculating the stem force nxded to -
are available from the Idaho DOE Office of1bchnol-close or open a gate valve does not have term! for the ogy Transfer, (208) 526-8318.
effects of temperature, degree of fluid subcooFng,in-Figure 1 of Volume 2 shows the test hiop in sche-ternal valve clearances, and the differences in the matic form and identifics the instrument location and opening and closing forces that are not accounted for by numbers. Figure 2 converts the differential pressures the stem rejection term. Second, effective in situ test-into flow rates (gallons per minute).1hble 1 outlines -
ing is very difficult because (1) the tests cannot be con--
the test sequence pcrformed on each valve and corre-ducted at design basis conditions and (2) even with the lates the data as they are presented here. In the re-valve loadings properly quantified during the in situ maining figurcs, the header on each plot defines the 41 NUREG-1377
~
)
l I
Main Citations and Summaries valve (A or B), the test series number, and the test step
- 1. Repeated cycling can have a significant effect on number. *Ihble 2 lists the test parameters measured valve thrust requirements.
during blowdown tests,'Ihble 3 displays the test step
- 2. The typical value of 0.3 for the disk friction coef-matrix for qualification and blowdown tests, and Table 4 ficient used by the industry is not conservative lists the test steps and system pressure and tempera-for all cases.
ture for each of the tests performed.
- 3. Theinfluenceof massflow/momentumonvalve NUREG/CR-5406, K.G. DeWall and R. Steele, Jr.,
thrust requirements may be significant.
"HWR Reactor Water Cleanup System Flexible
- 4. Increased temperature causes a significant in-Wedge Gate Isolation Valve Qualification and Ifigh crease in valve closure loads.
Energy Flow Interruption Test," Vol. 3, " Review of Issues Associat The limited number of tests performed to assess the Valve Closure.,ed with HWR Containment isolation capability of the gate valve to interrupt the flow of high-Idaho National Engineering Laboratory, EGG-2569, October 1989.
pressure steam has resulted in a relatively frequent m-ability to isolate portions of piping systems.,The data This volume discusses research performed to de-now available suggest that industry may be using non-velop technical insights for the NRC effort regarding conservative friction factors and possibly underestimat.
i Generic issue 87, " Failure of HPCI Steam Line With-ing valve stem thrust requirements. Additional work is out Isolation." Volumes 1 and 2 describe the relevant needed to determine whether present qualification valve test program.The research began with a survey to practices are adequate. Recommendations for expand-charactenze the population of normally open contain-ing the qualification of valve assemblics for high-ment isolation valves in those process lines that con-energy pipe break conditions are presented.
nect to the primary system and penetrate containment.
The qualification methodology used by the various NUREG/CR-5419, M. Villaran, R. Fullwood, and manufacturerslisted in the survey is reviewed and defi-M. Subudhi, " Aging Assessment of Instrument Air ciencies in that methodology are identified.
Systems in Nuc! car Power Plants," Hrookhaven National Laboratory, HNL-NUREG-52212, Four boiling water reactor (HWR) systems. the January 19%).
cmergency cooling system. the high-pressure coolant mjc: tion system, the reactor core isolatien cooling sp-As part of ongoing efforts to understand and man.
tem, and the reactor water cleanup system, were m-age the effects of aging in nucIcar power plants, an duded in ihe v:dve assembly characterization. The aging assessment was performed for the instrument air -
" typical"contamme nt isolation valve is a 3 to 10 in.,600 system, a system that recently has been the subject of v
to WO lb gate valve. The most common design is a cast much scrutiny. Despite its nomarcty classification, in-l steel. flexible wedge, pressure-scal valve with a strument air has been a factor in a number of poten-Iamitorque operator ( AC inside and DC outside of tially serious events.This report presents the results of containmenth The Anchor / Darling Valve Company the assessment and discusses the impact of aging of the manufactures approximately 40% of the valves m the instrument air system on system availability and plant four HWR mtems.
safety. This, work was performed as part of the NRC
(
The mitigation of a high-energy pipe break is within AR program. The objective of this study was to iden.
the desien basis for the'above vidve assemblies, wah tify all the aging modes and their causes that should bc gated to achieve reliable operation of all safety-m tvpical system deugn conditions of 1250 psi and 575"F.
$a flow testing has been performed under these condi-related air equipment. Also included is an interim re-tions to verify the presumptions used by manufacturers new of typical maintenance activitics for air systems in "I P""
"D in the uualification analysis calculations. Operator torque switch settings arc determmed using calcula.
tions supplied by the valve vendor: torque settings in-To perform the complex task of analyzing an entire adequate to close the valve could resuh H the original dram (, the Agmg and Life Extension Assessment wstem calculat om are not conservatwe, Al EAP) System Level Plan was developed by
~
W hven National laboratory (HNL) and applied Most of the vahe and operator manufacturers use successfully in previous studies. The work used two the same equation wit h minor variations in coefficients parallel work pat hs one using deterministic techniques u si/c eperators. In this equation. the required thrust to assess the impact of agmg on compressed air system to cime the vahe is equal to the sum of the disk drag performance, and the other using probabilistic meth-
' ud due to differential pressm e, t he stem end pressure odt Ihe results from both paths were used to charac-loadiand the packdg dmg hud;The savice conditions terite aging in the instrument air system.Tne findmgs med m the thrust equanon ;tre suppbed by each mdi-from this study, some of which have applications be-vidual plant. Four areas have been identined as having yond the instrument and service air systems, formed a the most influence on stem thrust requirements:
technical bas for under>tanding the effects of aging in
-42
Main Citations and Summaries L
compressed air systems. The major conclusions from
- 9. The outside systems that were most often af-this work are:
fected by instrument air problems are contain-i ment isolation, main feedwater/ main steam,
- l. This study identified aging trends in component auxiliary feedwater, and the BWR scram system.
failure rates, the relative importance of compo-
%c most commonly affected components were nents, and system unavailability. All these air-operated and solenoid-operated valves.
trends could have a deteriorating impact on sys-10.The probabilistic work entailed the develop-tem availability and consequently on plant safety ment of a computer program (PRAAGE-IA) m later years.
using a PRA-based instrument air system model
- 2. r mpressors, air system valves, and air dryers, to perform time-dependent PRA calculations.
o t
made up the majority of failures.The failures in Time-dependent failure rates were developed passive components such as pipmg, after-from the data base and other inputr to the coolers / moisture separators, and receivers in-program to calculate system unavailability and creased with time, but these still constituted component importances for various ages. The only a small percentage of overall faibires.
results showed that, when the time-dependent
- 3. The effectiveness and quantity of preventive effects of aging for the worst case are accounted 1
maintenance devoted to a component signifi.
for, there are two significant system effects:(1) cantly reduced the number of failures experi-system unavailability increases moderately with enced. However, existing maintenance pro-age and (2) the relative importance of compo-grams within the industry lack uniformity, and nents changes with age. During early operation, quality assurance is not rigorous because the leakage in both instrument air / service air piping j
system is classified as "nonsafety."
and support system piping was the most i
imp rtant contributor to system unavailability;
- 4. Individual plant maintenance records forinstru-dunng the later years, agmg can cause compres-ment and service air systems were found to be s rs and air dryers / filters to become increasingly the most comprehensive source of data for per-important.
forming aging analyses.
The findings presented in this report form a sound
- 5. As a contmuously operating system with mim-technical basis for understanding and managing the ef-4 mal controi room instrumentation because ofits fccts of aging in instrument air systems. Future work n(msafety classification, most problems m the w11 include improvements to current maintenance, air system are detected by local monitoring and monitoring, training surveillance, and off-normal re-mdication, walkdown-type mspection, and pre-sponse procedures to mitigate degradation due to ventive maintenance mspection or surveillance.
aging.
- 6. Review of compressed air system designs and NUREG/CR-5448, J.L. Edson, " Aging Evaluation of studies using a I RA-based system model re-vealed that the redundancy of key components National Engineering Laboratory, h," Idaho Class 1E Batteries: Seismic Testin G G-2576,
- (compressors, dryers, mstrument air / service air August 1990.
cross-connect valve) was an important factor in Batteries are the only installed source of electric system availability. The overall design configu-power to provide for monitoring plant conditions and ration affected the pervasiveness of air system control of some systems of the nuclear reactor in the problems.
event of a station blackout (all offsite power is lost and
- 7. 'Ibral-loss-of-aireventsare uncommon.The ma-the diesel generators do not start). Approximately 60 jority of events resulted in degraded operation individual 2-V cells are connected together to form a (Iow pressure, air quality out of limits). Normal typical 125-V de battery bank that has enough voltage wear of the system and contamination of the air and electrical capacity to provide the needed electric dominate the problems of system failure. Proce-power for the period of time determined for each nu-dures and testing for the response of personnel cicar plant in accordance with NRC regulations.
and equipment to these conditions should be Within the NRC Nuclear Plant Aging Research 1
^
developed.
(NPAR) program, a Phase I study of battery aging was
- 8. Iluman error was a significant cause of failures performed and reported in NUREG/CR-4457," Aging -
in critical components such as compressors and of Class lE Batteries in Safety Systems of Nuclear -
dryers, as well as at the system and intersystem Power Plants." The study concluded that significant ag-level. Training should be augmented in two key ing effects for old batteries are growth of positive areas: (1) operation and maintenance of critical plates, loosening of active materialin' plates that have air system components and (2) understanding grown, loss of active material caused by gassing and cor-the importance of instrument air to other plant n@n, and embrittlement of the lead grids and straps.
systems, particularly safety systems.
The results of these effects are decreased electrical ca-43 NUREG-1377.
1
['
Main Citations and Summaries pacity and decreased scismic ruggedness that, during a while increasing discharge current with those obtained scismic event, can lead to decreased electrical perform-with decreasing discharge current). These measure-ance or complete failure, ments were suggested as a result ofinvestigations per-Since batteries are susceptible to aging degradation f rmed by the Westinghouse R&D Center for Sandia that could cause old batteries to be vulnerable to severe National Laboratories in 1986.
scismic events, a test program was conducted to deter-Results of the seismic tests indicate that the capac-mine if it is possible for the scismic ruggedness of aged ity of lead-calcium batteries of this design did not de-batteries in nuclear plants to be inadequate, even crease is a result of shaking at seismic levels that in-though Ihe measured electrical capacity is satisfactory.
clude 11 e most severe SSE levels specified. In fact, the In addition, selected alternative surveillance methods averagt electrical capacity (ampere-hours) of batteries were evaluated during the testing program to deter-tmd at the 100% scismic level increased from a mine if any of them are likely to be more sensitive to prescismic capacity of 9% to a postseismic capacity of battery degradation than the surveillance and testing 98% The batteries did not show any degradation ex-methods specified in IEEE Std 450-1987, "lEEE Rec.
ccpt for some minor external damage.The battcry rack ommended Practice for Maintenance, Testing, and Re suffered some bending of structural components, but it placement of Large Storage Hatteries for Generating performed its intended functions. Post-test disassem-Stations and Substations," and Regulatory Guide 1.129, bly of selected batteries showed that some corrosion of
" Maintenance, Testing, and Replacement of Large the wcld joint between the positive plates and the bus /
Lead Storage Batteries for Nuclear Power Plants /'
terminal assembly had occurred as a result of the natu-ral aging process. However, this degradation did not
.Ihc batteries in the test program had lead-calcium interfere with the seismic performance. Metallurgical plates and were manufactured by C&D Hatteries. Dis-examinations showed that a large grain structure ex-cussions with C&D personnel indicate that they are typical of batteries presently being installed m nuclear isted at the weld area. The larger grain structure of the wcld makes it susceptible to corrosion and would ex-facilities. Each cell had a rated 8-hour electrical capac-plain the observed corrosion.
ity of 1350 ampere-hours: was 7-5/8 in. long,14-1/8 in.
wide, and 22-1/16 in. high; and weighed about 240
.Ihe results indicate that measurements of capaci.
pounds. They were obtained from a nuclear facility tance and mternal resistance ctm be obtained with re-where they were naturally ag~ed to 13.5 years. Records peatability and may provide an mdication of battery provided by the nuclear facility indicate that the batter-condition if the measurements were taken over the life-
[
ies were maintained and tested in accordance with time of the battery. Polarization and discharge current practices that are consistent with those in IEEE Std interruption are two techniques that are capable of, CO.
measuring internal resistance, while discharge current interruption is also capable of measuring battery ca-The battcries were installed on a shake table using a pacitance. These measurements would be most useful new battery rack purchased from the battery vendor if they could be rnade while the batteries were new and and were tested to scismic spectra that are typical of then repeated at regular intervals to obtain a pattern of l
those required for the safe shutdown earthquake (SSE) the change in battery characteristics with time.
~
in U.S. nuclear facilities. Information received from se.
1:
lected nuclear plants and the Electric Power Research The results of seismic tests on naturally aged batter-Institute (EPRI) was used to specify the required re-ies nearly 14 years old showed that, when batteries are sponse spectrum (RRS) for the scismic tests. The tests maintained and operated in accordance with IEEE Std t
were conducted using four different seismic levels rep-450 and Regulatory Guide 1.129, the following may be resenting the best estimate for the RRS encompassmg expected of adequately designed and manufactured 50% 85% 95%, and 100% of the U.S. nuclear plants.
1.
- Little, if any, electrical capacity will be lost as a During the seisrnic tests, the batteries were dis-result of seismic shaking at levels that are charged at 2% of the 3-hour rate while current and bat-typical of the most severe SSE levels specified
.I l.
tery voltages were monitored to detect the existence of for U.S. nuclear plants.
catastrophic failure. During the prescismic, seismic,
(
and postseismic tests, alternative surveillance and 2.
Some internal damage to the plate separators monitoring methods were employed to determine m y be expected at the most severe seismic whether other methods may be more sensitive to aging-levels. However, this minor loss of seismic related degradation of batteries than the standard volt-ruggedness is not expected to prevent the bat-ampere tests that determine their electrical capacity.
teries from providing at least 80% of rated The alternative monitoring methods employed were capacity during and immediately following the
- (1) measurement of internal resistance, (2) measure, most severe scismic event.
i ment of capacitance, and (3) measurement of battery 3.
Nat urally aged ba t teries may show evidence of
[mlarization (comparison of battery voltages measured '
corrosion at the joint between the positive I
I NUREG-1377 44
l i
Main Citations and Summaries plates and the positive plate strap (bus). In a tions during normal operation because of the well.made joint, this corrosion should not absence of high temperatures and humidities.
cause the seismic ruggedness to be inadequate The most important failure mode is expected for the most severe SSE cvents expected in the to be shorting (or reduced electrical isolation).
U.S. Operation of batteries at elevated tem-Several different causes may result in this fail.
perature or excessive charging could increase ure mode.
the corrosion, which could then progress rap-4.
Plant operational experience is useful to the idly enough to result m madequate scismic extent that it may indicate some possible fast-ruggedness.
acting degradation mechanisms for cables, c nnections, and EPAs that could lead to com-NUREGICR-5461, MJ. Jacobus. " Aging of Cables, on-cause failures under off-normal environ-Connections, and Electrical Penetration Assemblics Used in Nuclear Power Plants," Sandia National mental conditions. However, current LER 4
Laboratorics, SAND 89-2369, July 1990.
data provide a very limited data base for this This report covers the examination of the effects of purpose.
aging on cables, connections, and containment electri-5.
A significant number of manufacturers have cal penetration assemblics (EPAs) as part of the NRC produced cables, connections, and EPAs, re-NPAR program. Cables and connections are used in sulting in many different materials, designs, every electrical circuit in all nuclear power plants.
and construction methods. Consequently, ge-EPAs are included in every circuit that is inside contain-neric assessments of aging effects and vul-l ment.'Ihis NRC-sponsored aging assessment of cables, nerabilities have become much more difficult, connections, and penetrations is divided into two particularly where failure modes relate to in-phases. Phase I, which is the subject of this report, con-terfacing stresses.
sists of a review of applicable literature and evaluations An experimental assessment of cables is currently of usage, operating experience, and current inspection under way and will be documented in a future report.
and surveillance methods. Phase 11, currently planned NUREG/CR-5479,11. Damiano and R. C. Kryter.
only for cables, includes the development of improved Current Applications of Vibration Momtonng and methods for m.specuon, surveillance, and monitoring:
Neutron Noise Analysis: Detection and Analysis of application of monitoring methods to naturally aged Structural Degradation of Reactor Vessel Internals and in situ cables; and recommendations for utilizing from Operational Aging," Oak Ridge National the research in the rcgulatory process.
Laboratory, ORNL/Inll398, February 1990.
This report includes a review of component usage in The detection of degradation in PWR internals due nuclear power plants, a review of some commonly used to operational aging is becoming more and more impor-components and their materials of construction, a re-tant to U.S. utilities as the median age of U.S. nuclear view of the stressors that the compcments might be ex-power plants increases. Monitoring and detection of posed to in both normal and accident environments, a aging effects should aid in justifying plant life extension compilation and evaluation of industry failure data, a and result in safer and more efficient operation during discussion of component failure modes and causes, a the present and extended life period. It has been description of current industry testing and mainte.
demonstrated that monitoring programs based on neu-nance practices, and a review of some monitoring tech.
tron noise and vibration measurements utilizing signa-niques that might be useful for monitoring the condi.
ture analysis can effectively detect, and in some cases tion of these components.
diagnose, degradation of reactor vessel internals. Such progr ms have the potential to reduce plant downtime, The conclusions of the study are:
1.
Cab!cs, connections, and EPAs are highly reh.-
make periodic maintenance more effect ve, and.
increase plant safety.
able devices under normal plant operatmg Monitoring of reactor internals can be considered a -
conditions with no evidence of significant in-creases in failure rate with aging. Conse-particu{ar application of the general concept of predic-tqe m mtenance, the techniques of which are already.
quently, they receive little or no preventive widely used m mdustry to monitor rotating machinery.
maintenance..Under accident conditions, Predictive maintenance will be further implemented as -
however, the reliability ofIhese components is (1)its benefits become better documen ted, (2) famihar-practically unknown.
ity with the techniques and their applications grows, 2.
Aging effects that have the potential to lead to and (3)better hardware and software become available.
common-cause failuresduringaccident condi-
- A similar statement could apply to the monitoring of tions have the highest significance.
reactor internals. Although this monitoring has been 3
Many of the causes of failures for cables, con-spotty in the U.S., the above-mentioned techniques nections, and EPAs at accident conditions have been widely applied in Europe, particularly in would not cause any detectable manifesta-France and the Federal Republic of Germany, where-45 NUREG-1377
l Main Citations and Summaries they are currently (in 1989)5 to 10 years ahead of those NUREG/CR-5491, R. P. Allen and A. B. Johnson, Jr.,
l in this country. U.S. utilities could benefit from the ex-
"Shippingport Station Aging Evaluation," Pacific perience in Europe, where, in many cases, internals Northwest Laboratory, PNL-7191, January 1990.
r monitoring has been integrated into regular plant maintenance programs. Thus U.S. utilities could im-This report describes a research plan to address safety concerns on aging of sn ubbers used on piping and piement effective momtonng programs with a mini" mum of experimentation and wasted effort.
equipment in commercial nuclear power plants. *lhe i
work is to be performed under Phase 11 of the Snubber i
'the report begins with a description of some promi.
Aging Study of the NRC NPAR program with the Pa-
)
nent mechanisms through which degradation of reactor cific Northwest Laboratory (PNL) as the prime con-l internals occurs; the cause of most cases of 1his degra.
tractor. Research conduct ed by PNL under Phase I pro-dation is flow-induced vibration. Other mechanisms vided an initial assessment of snubber operation based are also reviewed.This is followed by a brief description primarily on a review of licensee event reports. The of vibration monitoring and neutron noise analysis,in, work proposed is an extension of Phase I activities and cluding a comparison and evaluation of these two covers research at nuclear power plants and in test methods. Next, current practices are summarized, and laboratories.The report includes technicalbackground examples of applications of thet:c methods in both the on the design and use of snubbers in commercial nu-U.S. and Europe (mainly West Germany and France) clearpowerapplicationsandadiscussionof theprimary are given. The report concludes with guidelines for failure modes of both hydraulic and mechanical snub-setting up what the authors consider to be a reasonable bers. The anticipated safety, technical, and regulatory internals-monitoring program for U.S. utilities, benefits of the work, along with concerns of the NRC and the utilities, are also subjects of the report.
NUREG/CR-5490, E. V. Werry, "Regulato Instru.
The Shippingport Atomic Power Station, presently ment Review: Management of Aging c LWR (1989)in the final stages of decommissioning, has been Major Safety-Related Components," Vol.1, Pacific a major source of naturally aged equipment for the Northwest Laboratory, PNL-7190, October 1990.
NPAR and other NRC programs. The evaluation of naturally aged components is an element of the NPAR r
This report is the first volume of a review of U.S.
nuclear plant regulatory instruments to determine the program strategy. Because naturally aged components and materials experience the actual ser ice-related ex-amount and kind ofinformation they contain on man-ternal stressors, corrosion and wear, testing proce-aging the aging of safety-related components in U.S.
dures, and maintenance practices, the evaluation of nuclear power plants. The review was conducted for such components is valuable. One is able to verify deg-the U.S. Nuclear Regulatory Commission (NRC) by radation models, to validate aging projections based on Ihe Ibcific Northwest Laboratory (PNL) under the the extrapolation of accelerated test data, and to detect NRC NucIcar Plant Aging Research (NPAR) program.
unexpected aging mechanisms (surprises) that could Eight selected regulatory instruments, including regu-significantly affect the safety performances of compo-lations, regulatory guides. technical specifications, nents or systems.
standards, Code of Federal Regulations and others, were reviewed for safety-related information on five Despite their importance for plant studies, natu-selected components: reactor pressure vessels, steam rally aged components of the destred type and vintage generators, pressurizers, primary piping, and emer.
re not readily available. The best source of these com-gency diesel generators. Volume 2 is tentatively sched-ponents is operational equipment from retired plants.
uled for FY 1994, and it will cover selected major
'Ihe decommissioning of the Shippingport Station, par-safety related components, e.g., pumps, valves and
. ticularly because it was managed by the U.S. Depart--
cables.
ment of Energy, represents a valuable opportunity to conduct in situ assessments at an aged reactor and to -
?
'Ihe focus of the review was on 26 NPAR-defined obtain a variety of naturally aged and degraded compo-safety-related aging issues, including examination, nents and samples for detailed aging evaluations by 1
inspection, maintenance and repair, excessive / harsh NRC contractors As ' the first U.S. large-scale, testing, and irradiation embrittlement. The major con-central-station nuclear plant, the Shippingport Station clusion of the review is that safety related regulatory parallels commercial pressurized water reactors in re-instruments do provide implicit guidance for aging' actor, steam, auxiliary, support, and safety systems.The '
l management, but include little explicit guidance. The 25-year service life (1957 to 1982) covers almost the en-major recommendation is that the instruments be tire period of currently operating reactors. Also, be-l revised or augmented to explicitly address the manage-cause of substantial modifications _ during the ment of aging.
mid-1960s and 1970s,it offers unique examples ofiden-NUREG-1377
-46
)
Main Citations and Summaries j
1 I
tical or similar equipment used side by side with the istrative, managerial, engineering, and operational as-original equipment but representing different vintages pects of their activities.The inspection Program recog-I and degrecs of aging. As part of the Shippingport Sta-nizes that licensees may satisfy NRC requirements in tion aging evaluation work, more than 200 items, rang-ways that differ among the licensecs, and inspection ing in size from small instruments and material samples guidance is therefore expressed in the form of perform-to main coolant pumps, have been removed and ance objectives and evaluation criteria, For the resident shipped to designated laboratories. These items in-and regional inspectors, procedures covering such sub-clude battery chargers, inverters, relays, breakers, jects as operations, maintenance, and surveillance have switches, power and control cables, electrical penetra.
been written. Some of these procedures contain guid-tions, check valves, solenoid valves, and motor.
ance on degradation due to aging, operated valves. Samples of piping from various plant Associated with each NI%R study is the need to de-
~
systems also have been acquired for radiological char-termine the role ofinspection, maintenance, and moni.
acterization studies, and samples from the pnmary sys-toring in counteracting the effects of aging and service tem components will be used for material degradation wear. The role of maintenance in managing aging is an studies.
important area where NRC emphasis has been applied.
Data and records relevant to the procurement, op.
A review by the NRC of maintenance performed at sev-eration, and maintenance of these materials and com.
eral plants concluded that most utilitics do not perform ponents have been obtained to support the detailed ag, condition monitoring because of inadequate knowl-ing evaluations. In situ assessments of Shippingport edge of degradation mechanisms and measurable condition-mdication parameters. The output from Station components also have been conducted, includ-NPAR in this area could provide information needed to ing preremoval visual and physical examinations of assist the inspectors to recognize age-related concerns, components, tests of electrical circuits, and special measurements to assist in the selection of specific com-The types of information generated by NPAR that ponents for further evaluation. Although detailed were found to be relevant to inspection needs include:
evaluations of the naturally aged components and ma-1.
Functionalindicators-NI%R reports identify terial from the Shippingport Station have not been parameters that can be monitored or meas-completed, the results from preliminary studies indi-ured to detect agmg degrelation.The inspec-cate the value of the aging information that may ulti-ps can apply these results to mhance visual s
mately be obtained.
mspections (walkdowns) and to evaluate licen-see programs for ensuring the operability of NUREG/CR-5507, W. Gunther and J. 'Iaylor, "Results equipment and systems.
from the Nuclear Plant Aging Research Program:
'Iheir Use in Inspection Activitics," Brookhaven 2.
Failure modes. causes, effects-Operating expe-National Laboratory, BNL-NUREG-52222, rience data evaluated in NPAR studies can September 1990, alert the inspectors to the prevalent failure mechanisms of systems and equipment. The
.The NRC NPAR program has determmed the sus-potential for changes in failure rate with in-ceptibility of nuclear power plant components and sys-creasing age is useful in evaluating preventive tems to aging and the potential for aging to affect plant maintenance.
safety and availability.The program has also identified methods for detecting and mitigating the effects of ag-3.
Stresses that cause degradation -Inspectors can ing in components. A review of the NRC Inspection benefit from knowing the environmental and Program and discussions with NRC inspection person-operational stresses that cause or affect degra-nel revealed several areas where the NPAR results dation due to aging.
- I would be valuable to the inspectors. This report de-scribes t he NIWR information that can enhance inspec.
To obtain a complete delineation of the NRC in-tion activities and provides recommendations for com.
spectors' needs, presentations summarizing the results municating this inforrnation to NRC inspectors. Thesc Of the N1%R program were made to the resident in-recommendations are based on a detailed assessment spectors at three regions. Their comments, supple-of the NRC Inspection Program and on feedback from mented by a written questionnaire, indicated that NPAR results can be of use to the inspectors when pro-resident and regional inspectors.
vided in a format directed to their activities. Examples i
The emphasis of the NRC Inspection Program is on of NPAR report summaries and inspection guides for evaluating the performance oflicensees by focusing on aging-related degradation of compcments and systems requirements and standards associated with the admin-are included in the report.
t 47 NUREG-1377
' Main Citations and Summaries NUREG/CR-5510, W.E. Vesely, R.E. Kurth, and S.M.
When the increase in core damage frequency is Sca!7o," Evaluations of Core Melt
- Frequency large for a given surveillance and maintenance pro-Effects Due to Component Aging and Mainte-gram, examination of the detailed aging contributors nance," Science Applications International Corpora-shows that relatively few comp (ments contribute. This tion, SAIC-89/1744, June 1990.
implies that a. graded,, maintenance program or,
' Itis report presents the results of a project to de-equivalently, a "prioritized" maintenance program can velop a methodology using probabihstic nsk analysis effectively control the core damage frequency increase (PRA) and component agmg models to quantify risk ef-due to aging. In such a maintenance program, most fccts due to compment and structural aging. The ap-components can have a lower level of maintenance if proach allows any presen t PRA and anyaging model for components important to core damage frequency have the components and structures to be used. An impor-a higher level of maintenance.
tant part of the evaluations is that the effects of mainte-
'Ihe domm. ant agingcontributors for the PWR were nance and surveillance programs in controlling agmg can be quantified. These programs can be explicitly found to be diesel generators, specific check valves and evaluated to determine their effectiveness in control, motor-operated valves m the cmcrgency core cooling ling aging impacts on system unavailabhity, core dam-system, and motor-driven pumps and turbine-driven age frequency, and public risk. Both point evaluations pumps in the auxiliary feedwater system. For the BWR, and uncertainty evaluations can be carried out, and de.
the dominant aging contributors were the diescis, the tailed contributors to the aging effects can be identified motor-driven pumps in the service water system, and and prioritized. PRA models are separated from the ag-the turbine-driven pumps in the reactor core isolation ing models, allowing available PRAs to be efficiently system. Ihe aging contribution from every component used in evaluating risk effects of aging.
in the PRA is provided and prioritized. These detailed c ntributms mclude specific systems. components,and
'Ib demonstrate the methodology, two PRAs, one um n pnm mmp en means of for a PWR and one for a BWR, were used to calculate focusing aging an lyses and agmg control efforts.
the increase in core damage frequency caused by aging for given aging data and assumed surveillance and In addition to the point calculations, uncertainty maintenance programs. The increase in core damage evaluations were carried out. For these evaluations, frequency duc to aging was averaged over time. This av.
ranges were assigned to each component aging rate, erage increase in core damage frequency charactcrized each effective overhaul interval, and each effective sur-the effectiveness of the maintenance and surveillance veillance interval. These ranges described uncertain-program in controlling aging effects. The average in.
tics and variations in the data. Log-uniform distribu-crease in core damage frequency can be added to the tions, which are llat distributions on a logarith mic scale, baseline PRA core damage frequency to obtain the to.
were used for the uncertainty propagation. All the vari-tal projected core damage frequency under a given ables were treat ed as being independent of one another maintenance and surveillance program with the acting for the evaluations.
aging process.
NUREG/CR-5515, H.H..Neely, N.M. Jeanmougin, and The aging of active components was modeled usinE J.J. Corugcdo, " Light Water Reactor Pressure Isola-the linear failure rate aging model m.which the compo-tion Valve Performance Testing," Energy 'Ibchnology acnt failure rate 1mearly increases with age accordmg Engineering Center, July 1990.
to a characteristic aging rate. 'Io demonstrate the meth-
, The Light Water Reactor Wlve Performance Test-odology, four aging rate data bases were used: TIR-GALEX, MOD 1, MOD 2, and MOD 3 These data mg Pngram was mitiated by the NRC to evaluate leak-bases demonstrated the effects of dilferent aging rates age as an mdication of valve condition, provide input to Section XI of the ASME Code,and evaluate emission on the core damage frequency for a given maintenance monitoring for condition and degradation and inservice and surveillance program.
inspection techniques. Six typictd check and gate valves Results obtained for different surveillance and
- # purchased for testing at typical plant conditions maintenance programs clearly show the sensitivity of (550.F at 2250 psig) for'an assumed number of cycles the m.erease m core damage frequency to the type of for a 40-year plant lifetime. Tests revealed that there maintenance program and the aging rates. The results were variances between the test results and the present -
arc significant frorn a technica! standpoint because they statement of the Code; however, the testing was not explicitly quantify the impacts that aging and mainte-conclusive. The lifecycle tests showed that high tech nance can have. these evaluations are the first quanti-acoustic emission can be utilized to trend smallleaks, fications of aging and maintenance impacts using full-that specific motor signature measurement on gate scade up-to-date i ras.
dves can tre nd and indicate potential failure, and that inservice inspection techniques for check valves were shown to be both feasible and an execlient picycntive
- The cun ent NRC terminology uses" core damage"(asin the accompa, oyinpummary) in<tead ol " core meli."
maintenance indicator. Lifecyc!c testing performed NUREG-1377
.48 1
1
.-~
i Main Citations and Summaries I
i l
}
1 here did not cause large valve leakage typical of some records available for the two commercial plants indi-
)
plant operation. Other testing is required to fully un-cated a significant preventive and corrective mainte-derstand the implication of these results and the re-nance effort to take carc of service wcar and provide re-qu red program to fully implement them.
liable equipment operation.
Maintenance recommendations included in operat-NUREG/CR-SS19, J.C. Moyers, " Aging of Control and ing and maintenance manuals provided by equipment r
Service Air Compressors and Dryers Used m 4
Nuclear Power Plants " Oak Ridge National 12bora-manufacturers were reviewed and compared to the 3
tory, ORNir(407/VI, July 1990.
preventive maintenance practices at one plant. 'lhe user-applied practices generally were in conformance l
,lhts report discusses work performed as part of the with or exceeded the manufacturers' recommenda-NRC NPAR program n practical and cost-effective tions. One troublesome aspect is ensuring the opera-methods for detectmg. momforing, and a5.sessing the tional readiness of auxiliary compressors that are nor-severity of time-dependent degradation (agmg and mally idle for long periods but must pmvide backup semcc wear) of compressors and dryers used in the service for criticd needs if the main control air supply control and semcc air systems of nuclear power plants.
deteriorates. Manufacturer-recommended mothball-Ihc objective is to provide capabilities ior establishmg ng procedures do not appear practical for this applica-degradation trends prior to failure and for developing tion; such failure causes as drive belt set, corrosion of guidance on effective mamtenance programs.
internal parts. and small internal water leaks may pre-l The topics covered are failure modes and causes re-sent a problem when the compressor is needed.
sulting f rom aging and service wear, manufacturer-Measurable parameters that have a potential for j
r ecommended rnain tenance and surveillance practices.
enhancing the capabdities for detecting incipient fail-and measurable parameters (including functional mdi-urcs and examining degradation trends in compressors i
cators) for use in assessing operational readiness and and dryers were identified. For compressors, they in-i equipment condition (often related to degradation clude periodic delivery capacity tests, trending of stage trends) and in detecting incipient failure. The results temperatures and pressures, and motor current signa.
are based on information derived from operating expe-ture analysis. Measurable parameters for drycrs in-rience records, manufacturer-supplied information, clude moisture sensing within the desiccant column and inputs from plant operators. For cach failure near the exit and periodic monitoring of the axial tem-l mode, failure causes are listed by subcomponent, and perature profile within the column. Use of these meas-4 potentially useful parameters for detecting degrada-urable parameters in the surveillance and monitoring tion that could lead to failure are identified.
program might reduce the level and duration of time-
- A brief review of typical compressors and dryers in directed out-of service inspection and maintenance, nuclear power plants showed that the nontubricated re.
thereby increasing availability and improving overall i
ciprocating compressors and the regenerative desic-system reliability.
j cant dryer are used in more plants than any other types Nuclear plant control and service air compressors for both service and cont rol air systems, and the assess-and dryers are not usually considered as safety related ment was therefore focused on them. A general de-because the air systems are not needed to bring the.
scription of the equipment that includes illustrations, plant to a safe shutdown condition. An effective sur-J defined equipment boundaries, functional require-veillance and monitoring program with preventive and mentsiand materials of construction is provided. Op-corrective maintenance c:m provide reliable service crational stressors are categorized and hsted in detail.
from nuclear plant compressors and dryers. Instances Data bases and nuclear industry reports containing of loss of air supply due to compressor or dryer failure nucicar power plant operating experience were exam.
are rare because of the redundancy in most systems.
ined.1hese data bases included the Licensec Event For these reasons, it is recommended that no further Report (LER) file as cataloged in the Sequence Coding ccmsideration of this equipment be included m the and Search System main tained by ORNils Nuclear Op.
NPAR program.
erations Analysis Center, the Nuc! car Power Experi-NUREGICR-5546. S. P. Nowlen, "An Investigation of ence compilation main tained and published by the S.M.
the Effccts of Thermal Aging on the Fire Sto!! cts Corporation, the In-Plant Reliability Data Sys-Damageability of Electric Cabics," Sandia National tem containing maintenance records for one plant, and Laboratories, SANDWO6% May 1991.
maintenance records obtained from a cooperating util.
1his report describes the results of a series of tests ity for a second plant. During the 1978-1988 decade performed to assess the effects of thermal aging on the covered by the LER data, which represents approxi-vulnerability of cab!cs to fire-induced thermal damage.
mately 812 reactor-years,22 compressor-related and 16 The tests were part of an effort in support of the NRC dryer-related events that resulted in loss of control air NPAR program to identify and investigate fire safety is-supply were reported. Equipment failure causes were sucs for which plant aging might lead to an increasci diverse, with no single type of failure dominating.1he level of risk.
j 4
49 NUREG-1377 i
Main Citations and Summaries A
i From the standpoint of fire safety, cables represent of thermal damageability can be made based on these the single most important class of electrical equipment tests.
in a nuclear power plant. First, virtually every plant sys-One measure of fire damageability is the thermal tem includes power, control, and instrumentation ca-darnage threshold defined, in the context of these tests,
.l bles. Second, cable " pinch points (that is, locations as a temperature range. Its upper limit is the lowest i
where redundant train separation is reduced by the temperature at which electrical failure was observed merging of cable routings) often represent dominant following exposures of up to 80 minutes. The lower i
contnbutors to plant fire risk as determined by proba-limit is the highest temperature for which no electrical bilistic risk assessment (PRA) analyses. Third, cables failures were noted following exposures of no less than
~j represent the major combustible fuel loading for most 80 minutes.
plant areas.
For the Rockbestos cable, the failure threshold of The tests described here examined the thermal the unaged cable was determined to be 325-330 *C, damagcability of two commonly used types of low-whereas the thermal damage threshold for the aged flame-spread electric cables qualified to IEEE-383:
samples was 350-365'C. For the BlW cable, the ther-
[
1.
A Neoprene-jacketed, cross-linked-polyethyl-mal damage threshold of the unaged cable was esti-ene-insulated (XPE), three-conductor, 12 mated at 365-370*C, whereas that of the aged samples AWG,600V light P wer or control cable pro-was estimated at 345-350*C. Th us the aging process re-t duced by the Rockbestos Corporation and sulted m the opposite effect on the thermal damage marketed under the trade name Firewall 111.
threshold for the two cable products. For the Rockbes-tos cable, aging increased the damage threshold by 2.
An ethylene-propylene-rubber-insulated approximately 25-35*C while, for the HIW cable, it -
(EPR), chlorosulfonated-polyethylene-jack-decreased the threshold by approximately 20*C.
eted (CSPE or Hypalon), two-conductor,16 A second measure of thermal damageability is the AWG, plus shield and drain,600V instrumen-relative time to failure for exposure temperatures tation or signal cable produced by BIW Cable above the damage threshold. The aged Rockbestos Systems incorporated and marketed under the samples consistently displayed longer times to failure trade name Bostrad 7E.
at a given temperature than did the unaged samples,in-For each of the two cable types, both unaged (i.e.,
dicating less vulnerability to thermal damage for the new from the cable reel) and thermally aged samples aged samples. The time to failure for the aged and were tested. No radiation aging was employed in these unaged BIW samples was not sigmficantly different for tests.
exposure temperatures at which failure wasobserved m both aged and unaged samples.
1 The exposure conditions simulated during testing were considered typical of those expected durmg an en-
. It was also noted that, in virtually every case, failure l
closure fire when the subject cables are not involved in of the cables through conductor-to-conductor shorting l
the fire itself. The most significant difference between resulted in the initiation of intense, sustained, open the test exposures and anticipated actual exposures was flaming in the cable samples. As the cables shorted, that the tests mvolved expmure at an elevated steady-sparks ignited the gases evolved from the cables. In no state temperature whereas, m actual exposures, equip-caw was spontaneous ignition of the cables observed -
prior to electrical failure. These results indicate that '
I ment would experience a transient time / temperature the failure of energized cables is a mechanism for fire exposure.
spread.
In these cable exposure tests, the walls of the cham-The thermal damage threshold changes observed in ber and the air were preheated to the desired uniform the tests on two of the most common nuclearqualified
~
I steady. state exposure temperature. Two energized ca-cables in current use in the U.S. nuclear industry are ble samples were then quickly inserted through a small not considered of sufficient magnitude to significantly door to provide a near step change in environment tem-alter risk estimates for scenarios involving cable ther-perature for the cable samples.
mal damage.
The cable samples were energized by a three-phase It should be noted that these tests have not explored 208-volt power source. Each of the three conductors of the impact of other fire environment effects such as I
the Rockbestos cables was connected to one phase of suppressant application and high humidityon cable sur-the power source. In the case of the BlW cable, the two vival. The failure thresholds given above pertain to--
conductors and the drain conductor were cach con-gross electrical failure. In most cases, significant IcVels nected to one phase of the power source. Leakage cur-of current leakage were noted prior to gross failure, l
rents between power phases were monitored continu-and specific applications must be examined to deter-ously.The time to ultimate cable failure, as determined -
mine whether such leakage could constitute the failure
-I by the failure of a two-ampere fuse in any one of the of a circuit to perform its design function. Also,because l
three phase circuits, was also recorded. Two measures mixed results were obtained for the two cable types NUREG-1377 -
50
i l
Main Citations and Summaries tested, no direct ccmclusion regarding the impact of induced vibration, and particle debris carried by the thermal aging on the fire vulnerability of any other ca-coolant.
ble type can be drawn based solely on the results of A failure modes and effects analysis of the Westing-4 these tests.
house control rod drive system was also conducted, and I
NUREG/CR-5555, W. Gunther and K. Sullivan, " Aging emnponents with a high safety significance were identi-Assessment of the Westinghouse PWR Control Rod fied along with the likelihood of their failure. This as-Drive System," linokhaven National Laboratory, sessment was based on operating experience data and HNL-NUREG-52232, March 199L an evaluation of the susceptibility of the components to A study of the effects of aging on the Westinghouse age-related degradation. Several components that c<mtrol rod drive (CRD) system was performed as part should receive attention as a plant ages were identified:
of Ihe NRC NPAR pmgram. Its objective was to pro _
cables, coils, and connectors (in containment); latch as-vide a technical basis for identifying and evaluating the sembly; guide tube: and selected electronics within degradation due to aging, power and logic cabinets, including the rod-position in-The Westinghouse CRD system consists of control dicating system.
i rods and the mechanical and electrical components An evaluation of inspection, surveillance, monitor-that c(mtrol the rod motion. The study exammed the ing and maintenance was accomplished with informa-design, construction, maintenance, and operation of tion from fifteen plants representing ten utilities.
the system to assess its potential for degradation as the Responses from most plants agreed that two of the
-i nuc! car plant ages and evaluated the ' extent to which required technical specification tests (rod-drop timing aginr could affect the safety objectives of the system.
and rod execrcising) are beneficial in verifying the op-Studies are also being conducted for the Combustion crational readiness of the system. Preventive mainte-Engmeermg, Habcock and Wilcox.and General Elec.
nance activitics for electrical components within con-tric CRD systems, tainment dominate the overall maintenance of this sys-tem. Only a few plants are using circuit-monitoring The operating experience for CR D systems as docu-mque or nondestructive testing to monitor the mented in the Licensee Event Reports (LERs). Nu-lonplerm operational characteristics of the CRD sys-cicar Plant Reliability Data System (NPRDS). and Nu-tem (a substanttal portion of that system is ccmsidered clear Power Experience (NPE) data bases was re-not rdaad to saW Re mpom fmm plana fun viewed. 'fhese sources provided an averace of 30 ther indicate that some plants have modified the sys-uniaue failure events per year ever the la<,t Riycars, of
' " W""#
which approximately 35% were directiv attributable to
- ' U"# "*P""E""hese activities have effec-P"C m n nnw ral of t aging-related degradation. The review resulted in the tivelv addrq %ed the aging issue. However. mainte-followinnThe majonty of the reported failures occurred observations:
nance praeves appear to vary from one plant to an-1.
- other, possibly reflecting inadequacies at some plants.
m the electrical area i.e., the power and logic Rchniques to detect and mitigate the effects of ag-cabinets, and the rod position mdication sub' ing including advanced approaches by Westinghouse, system.
the Japanese, and the French are desenbed. Previous 2.
Approximately 40's of the reported failures research re!ated to the Westinghouse CRD system was resulted in a nxi drop, which usually chal-discussed, e g., NUREG-Ell on wear of control md lenges the reactor pmtection system andiniti-guide tubes, study E613 on localized wear from the arcs a reactor trip.
NRC OUice of Analysis and Evaluation of Operational 3.
Several failure modes such as rod position drif t Data, anJ EPRI-sptmsored reports on plant life exten-and overheating of power cabincts are com.
sion and control rod lifetime determination. Research mon to many plants, which could indicate the on the extension of plant hfe fora Westinghouse PWR, need for generic resolutions, for' example, identifies the latch, drive rod, and coil stack asser9blies as hmited-hfe components.
The normal operating and environmental stresses experienced by the system components were assessed The nnMgs and recommendations of ths eng to determine 'their effect on the long-term perform-study may be summari7ed as follows:
ance of the+ystem. Far exampic, the regular stepping 1.
Arinprelated decradation of the Westing action associated with control rod motion results in home r'RD sutem can compromise the in-wear of the latch and drive rod components and m clec.
wnded function of the system. Therefore.
Incal surces on the control rod drive mechanism co A means in detect and mitigate this degradation tne afety-nd non safety-related portions
'lhe amo' int of latch wear measured in another study is n
t presented in this report along wi resuhs of otbet ofthec. tem, hould be pursued.
research efforts related to aging of the CRD synem.
2.
The test requiremems on the sptem(c.g.. rod-Olher cumpics of stressors asmciated with aging-Jrop timind arc important in determining the' tebted degradation are high t empe ratu re, fh gerader.ai rcaJiness of the system.ahhough
$1 NUREG-1377
Main Citations and Summaries they cause some incremental wear on the me-One of the major parts of the research progam in-chanical part of the system.
cluded two full-scale qualification and flow interrup-tion test programs on flexible-wedge gate valves, 3.
The preventive maintenance, including in-spection and testing of the in-containment ca.
Phases I and II. The Phase Il program was performed in bles, c<mnectors, and coils should be increased 1989 at the Kraftwerk Union (KWU) facilities near-as these components age.The use of such pre-Frankfurt, Germany. Among the valves tested, three dictive maintenance concepts as nonintrusive were 10-in. valves typical of those used in the HPCI ap-on-line monitoring techniques should be con-plications. One of the 6-in. valves was also tested at sidered.The method used to perform the rod-RCIC test conditions. In all, seventeen flow interrup-drop timing test should be modified, or the tion tests were performed, seven at design basis c(mdi-degradation that can result from the present tions.
procedure of pulling the fuses should be ac-counted for.
Two RWCU valves were tested during the earlier j
Phase 1 Test Program. As a result of that work, it was expected that the valves would require more stem force ti ntr h CRI e ic pl
. Maintenance errorsin this area have resulted to close than m, dusry normally would have predicted, in unnecesssary reactor trips and additional Therefore, for the Phase Il Program, the motor-opera-stress to the CR D system. Repair and replace-tor control switches were set at higher-than-normal ment procedures for this portion of the system torque values to ensure vahe closure, and the strengths j
should be evaluated for completeness and ac.
and weaknesses of a given valve design were deter-j curacy, and personnel training should be em-mined from the recorded data.
i f
phasized.
The test results clearly showed that, for the GI-87 concerns, all valves that were subjected to design basis j
NUREG/CR-5558. R. Steele, Jr., K.G. DeWall, and J.C. Watkins, " Generic issue 87, Flexible Wedge flow interrupti n tests required more torque and sub-Gate Valve Test Program: Phase il Results and sequently more stem force to close than would be pre-1 Analysis," Idaho National Engineering Laboratory, dicted using the standard industry motor-operator siz-i EGd-26(0, January 1991.
ing equation for disk load calculations with a common
-i Qualification and flow isolation tests were con-coefficient of frictionfrhe highest loads recorded werc ducted to analyze the ability of selected boiling water the result of internal valve damage caused by the high-reactor (HWR) process valves to perform their contain-differential-pressu re loads across the valve disk as it at-ment isolation functions at high-energy pipe break con-tempted to stop the flow.
i ditions and other more normal flow conditions. Nu-merous parameters were measured to assess industry The high loads encountered during the test series ;
practices for predicting valve.and motor operator re"-
raise the concern that some valves installed in nudear 4
quirements. The valves tested were representative of.
power plants may not have large enough motor opera.
those used in HWR reactor water cleanup systems and -
tors to ensure closure in the event of a design basis acci-high-pressure coolant injection (HPCI) steam lines.
dent.
Among the objectives of this research program are to determine what factors affect the performance of mo.
The. study into the phenomena affecting the stem I
l tor-operated gate valves and to determine how wellin-loads in a motor-operated gate valve continues. How-l l
dustry's analytic tools predict that performance.
cver, the results to date indicate that the phenomena i:
This program supports the NRC's effort on a ge-taung place inside the gate valve are more complex -
l _
neric issue, Gl-87, " Failure of HPCI Steam Line With-than previously thought.The actual disk factor is much l
out isolation." GI-87 covers three boiling water reac-higher than previously believed, but this factor can be l'
tor process lines: the 1IPCI turbine steam supply line, mJerated for some valve applicat' ions once the self.
the reactor isolation coolingiRCIC) turbine steam sup closing force balance on the valve disk is understood.
i ply line, and the reactor water cleanup (RWCU) pmc-
. a mypection mdicated that these valves were ess line. All three of these process lines communicate with the primary sptem, pass through containment, very ne r their physical fragdity limits at design basis
)
and have normally open isolation valves. The concern c nditions.The excessive bearmg pressure between the j-with the isolation VMves is whether they will close in the disk and the body guide matcrials resulted in yiciding, l
event of a pipe break outside of the contamment. Are.
spalliv,and gougingof the surfaces.In somt of thede-l.
lease of high-energy steam or hot waterin the auxiliary signs, the guide clearances were large enough to allow building could result in common-cause failure of other the disk to tilt during closure, which resulted in signifi.
comp (ments necessary to mitigam the accident.
cant damage to the scaling surfaces.
NUREG-1377 52
i Main Citations and Summaries NUREG/CR-5560, H.M. Hashemian, D.D. Heverly, The project addressed the idewing additional top-D.W. Mitchell, and K.M. Petersen, " Aging of Nu-ics: sources of errors in RfD calibration, factors affect-clear Plant Resistance Temperature Detectors,"
ing RfD accuracy and response time, failures of RfDs Analysis and Measurement Serv ces Corporation, as reported in the LER and NPRDS data bases, and the JuncI M International Temperature Scale of 1990 and its impact J
, A comprehensive research and development pro-on temperature measurements in nuclear power i
ject on aging of narrow-range resistance temperature plants. The results of research performed in Phase II detectors (RID) used in the primary coolant system of did not reveal any unanticipated or major systematic pressurized water reactors was carried out as part of the aging problem in the performance of the KrDs tested.
NRC NPAR program. The goal was to establish the The nuclear industry's practice for verifying adequate long-term performance limits of these KPDs m order RfD accuracy and response time is to perform on-line to verify that objective and adequate measures are im-cross calibration and loop current step response tests at l
plcmented to ensure safety.
least once every fuel cycle. In light of the data obtained The project was c(mducted in two phases. Phase 1, a throughout this study, this approach is reasonable for six-month feasibility study, was completed in June 1987.
managing t he aging of RfDs that do not have any major i
'The results, published in NUREG/CR--4928, "Degra-design, fabrication, or installation deficiencies. RIDS dation of Nuclear Plant Temperature Sensors." dem-that consistently maintain a suitable calibration and j
onstrated the need for additiond work in Phase II.This reponse time as determined by periodic testing can be report presents the results of Phase 11, which was con-used in the plant for their qualified life as specified by ducted over a 30-month period beginning in October the manufacturer. The manufacturers' specifications 1987. The work involved laboratory testing of 72 nu-for the qualified life of nuclear grade RI Ds typically j
cicar grade RfD clements representing several from range from 10 to 40 years depending on the manufac-each of four U.S. manufacturers. The limit for the turer and the conditions at which the RIDS are used.
l initial accuracy of these RfDs was established, and a procedure fo'r performing precisc calibration was NUREG/CR-5583, M.S. Kalsi, C.L Horst, J.K. Wang, j
and V. Sharma "Predictmn of Check Valve Per-i developed. E*Pcrimental aE ng of 30 of these RIDS at i
formance and Degradation in Nuclear Power Plant simulated reactor conditions resulted m. five failures Systems -Wear and Impact Tests. Final Report, and six major calibration shifts. Two failures occurred m September 1988-April 1990," Kalsi Engineering, thermal agmg, one m vibration aging, one in humidity Inc., KEI No.1656, August 1990, aging, and one in thermal cycling. The remaining 19 Check valve failures at nuclear power plants in re-KrDs performed well during the aging tests, maintam-cent years have led to serious safety concerns and have ing a drift band of 0.2 L.
caused extensive damage to other plant components l
The shelf-life drift of RfDs was also quantified-that had a significant negative impact on plant availabil-This involved testing 45 RfDs for storage effects: 24 ity. Swing check valve internals may experience prema-l that had been in normal stomge at various nuclear ture deterioration if the disk is not firmly held open power plants for periods of one to fiveycars and 21 that.
against its stop. At the present time, no guidelines exist were aged in the project. The test results for these 45 for the prediction of degradation trends and the deter.
RIDS showed a shelf hfe drift hand of iO.l *C. Most of mination of suitable inspection intervals. A rescarch -
the storage drifts, the failures, and the normal aging program aimed at developing a reliable model for drifts were found to occur in the first few months of ag-quantitative predictions of wear and fatigue for swing ing. A potential remedy is to burn in the KrDs before check valves was established as part of the NRC NPAR they are calibrated and installed in the plant.
program to improve the safety and reliability of their The performance of nuclear plant RIDS is evalu.
operation. This report covers Phase II of the research ated by respcmse. time testingin addition to calibration.
on swing check valves. The work in Phase I was pub-These two procedures are independent and are there-lished in NUREG/CR-5159.
fore done separately. The nuclear industry has about The goal of Phase 11 was to develop predictive mod.
ten years of experience with RfD response time result-cis that could be used to quantify the degradation of ing from periodic in situ measurements made in about swing check valves with flow disturbances close up.
60 PWRs at least once in every fuel cycle. Representa-stream of the valve at flow velocitics that do not result tive results of these measurements were reviewed to
- n fuG iisk opening. Two major causes of swing check identify the range of achievable response times and the valve fanure are premature degradation due to wear in response time degradation modes.
the hinge pin and fatigue in the disk stud connection to Several commercial grade RrDs were also aged and the hinge arm. Accelerated wear tests were performed tested for comparison with nuclear grade RIDS. 'Ihe using aluminum hinge pins and bushings in 3-inch and
(
results showed that the average response time and cali-6-inch valves to quantify wear experienced in the hinge bration stability of nuclear grade RIDS is about twice as pin area. A special disk instrumented with strain gages good as that of the commercial grade RIDS.
was used in the 6-inch valve to measure the impact i
53 NUREG-1377
Main Citations and Summaries forces and their rate of occurrence to quantify the fa-
'lhe research program goes beyond an analysis of tigue damage caused by the disk tapping against the times of degradation and failure. First, theoretical stop. The wear and fatigue prediction models devel-models that relate the degradation rate of the compo-oped in this program show good correlation with labo-nent to its failure rate are developed.With the relation-ratory test results as well as with a limited number of ships derived, information on component degradation check valve failures at plants that had been sufficiently can be used to predict the component failure rate and documented.
its significtmce. Specifically, this methodology can use The results of this research allow inspection and aging trends in the component degradation rate to pre-dict future aging trends in the component failure rate.
maintenance activities to be focused on those check val,es that are more likely to suffer premature degra-The capability of making such a prediction is impor-dation. The methodology for quantitative prediction of tant because information on component failure rates wear and fatigue can be used to develop a sound and due to aging is required to quantify the effects of aging effective preventive maintenance program. lhe results on core damage frequency and risk.This information is also indicate certain modifications in the valve design also needed to quantify the effectiveness of a given that may improve check valve performance and reli-maintenance program in controlling the effects of ag-ability.
ing on the core damage frequency and risk. However, failure data arc often sparse. On the other hand. degra-NUREG/CR-5587. W.E. Vesely, " Approaches for Age-dation data are more abundant because degradations 1]cpendent I robabilistic Safety Assessments with..
occur at a higher rate than do failures. Thus the meth.
Emphasis on I rioritization and Sensitwily Studies, Science Apphcations International Corporation, odology developed m. this report allows component SAIC-92/1137, August 1992.
failure rates due to aging to be estimated from compo-nent degradation rates.This has the potential of greatly This report desenbes approaches for incorporating MN*N" "Y
commment agmg rehabdity models into a probabdistic m ratMuc m agg kr m in M Muam of ag-safety assessment (PSA) or probabilistic risk "E '"C" assessment (PRA) of a nuclear power plant. These ap-proaches and procedures are described from a techni-It is important that. in addition to the identification cal standpoint and are not to be interpreted as having of aging trends in degradation and failure data, the any regulatory implications. Component aging failure methodology allows maintenance indicators to be
~
rate models and test and maintenance aging control selected in such a way t hat component degradations are models are presented. Different approaches for carry-related to impacts on reliability and risk. When the deg-ing out the aging evaluations are given. Demonstra-radation indicators show sigmficant impacts of degra-tions are cis en invohing prionnang aging contributors, dation on the component fadure rate and the resulting evaluating mainicnance effectiveuss, carrying out risk, maintenance should be performed to correct the time-dependent evaluations, and mying out uncer-degradat ions. Th us the degradation indicators can pro-tainty and sensnivity analyses of c.ging eticcts.
uJe a practical and effective means of monitoring component condition and signal for the correction of
. NUREGICR-%12, P.K. Samanta. W.E. Vesely, E lisu-denradations before they have significant impacts on and M. hubudhi. " Degradation Modehng vth Ap-miIabihty and risk. In addition. the methodology was obcation to Aging ano Mamtenance Ef fecuveness used to develop initial estimates of the eficctiveness of Evaluanons Brookhasen National Laboratory, maintenance in preventing degradations from becom.
HNL-N UR EG-52252. March 1491.
ing failures.
An important element of the assessment of risk as-sociated with aging in nuclear pmer plants is the un-Specific appbcations of the theoretical approach derstanding of ihe aging phenomena associated with resulted in quantitative models of comp (ment degrada-components of safety systems. This report describes a lion rates and component failure rates derived from study of aging phenomena at the component level m plant-spectfic data. As part of the data analysis support of the NRC NPAR program to Jevelop an ag.
statistical techniques that identify aging trends in fail-ing reliabibty model representing the aging process ex.
ute and degradanon data were developed. The aging perienced by components in nucitar power plants un-t rends can be of any kind and can exist in any segment of der pesently exnting test and maintenance pracaces.
the data. Specifically, an analysis of residual heat-re.
A ntw model was developed to process mformation on moul (RllR) system pump data shows a "hathtub" component degradauon m order to analyze the degra-curve for the degradation rate where a distinct increas-dation process and its imphcanont The focus was on ing trend is obsers ed at th e later ages. In terestingly, the madctmg the degradation rate. te.. tne rate at which pump fatture rate does not show n any increasmg trend degradations occur, with the *pecific onjectis e of devel-for the same pecod, which demonstrates the need to oping explicit relationships between degradation char identify aging tren61hrough amlyses of component actenstics and de component fadure rate.
degradations NUREG-1377 54 l
)
Main Citations and Summaries lhese results are important first steps in showing ferent gas burner (fire source) configuration with two that degradations can be modeled to identify aging ef-cable trays placed face to face with insulating backer fects. 'Ihe theoretical methdology that was developed boards (as compared to a single open ladder tray used in represents an advancement demonstrating that degra-the standard test) would produce enhanced fire propa-dation characteristics are explicitly related to failure gation.The tests described here used a similar configu-rates and hence ultimately to risk.1he next step would ration to induce flame spread in the sampic cable be-
-l be to use the methodology and statistical techniques to cause, if the cables did not burn during testing, little
]
develop and validate practical procedures for predict-would be learned.
ing failure rates due to aging from degradation data.
During each of the four fire tests, it was observed Ih:s ability would provide powerful tools for analyzing M
i of the available combustible materi-l aging effects m terms of degradation data and for pre-als (the cable irisulation and jacket materials) were cc n-dicting their implications for reliability and risk.
g; p
- g g;;
NUREG/CR-5619, S.P. Nowlen, "The impact of Ther-f the 16-foot vertical cable trays was observed in all mal Aging on the Flammability of Electric Cables,"
cases. Ilowever, upon examination of the test data,it Sandia National L_aboratories, SAND 90-2121, was found that, for both cable types. the aged cable March 1991.
samples displayed a reduced flammability as compared This report describes tests on the fire vulnerability to the unaged cable samples.This was reflected in re-of aged electrical components performed for the NRC ductions in both the rate of rise and the peak value of NPAR program.1he objective was to identify and in, the measured fire heat release rates for the aged cable vestigate issues of plant aging that might result in an in-samples as compared to those for the unaged cable i
creased fire risk at commercial nucicar power plants.
samples.
The particular issue investigated in these tests is the These results indicate that, at least for the two cable impact of thermal aging on the flammability of electri-types tested, thermal aging resulted in a decrease of cal cables.
material flammability. Ilence, for these two cable
'Ihe cable insulation represents the dominant types, the issue of material aging and cable flammabil-source of combustible materials in most nuclear power ity is not of concern.The use of material flammability plant areas. Current USNRC standards require the use parameters obtained from tests of unaged cable sam-of low-flame-spread cables, as certified by the ples will therefore provide conservative assessments of IEEE-383 qualification standard, in all new installa.
material flammability in a thermally aged condition.
tions liowever, should these cables lose their fire-1hese results are consistent with results of previous retardant properties as a result of material aging, an in-cable aging studies. It has been obsen ed that the proc-crease in fire risk could result based on the role cable ess of thermal aging tends to drive off certain of the installations have played in past fire risk assessments.
more volatile constituents of the cable insulation mate-
'Ib assess this issue, four large-scale cable flammability rials. 'lhis will leave less of these compounds available tests were performed.TWo commonly used types of nu-during a fire to support combustion; hence flammabil-c! car grade electrical cables were tested in both the new ity is reduced somewhat. Although other cable types (unaged) and a t h ermally aged (th rough accelerated ag-have not been tested, it is expected that similar results ing) conditiom would be obtained. No further investigation of this is-1.
Rockbestos FIREWALL III,3-cmductor,12 sue is recommended.
AWG, Neoprene jacketed, cross-linked poly-ethylene (XPE) insulated light power or con' NUREG/CR.5643. D. E. Illahnik, D. A. Casada, J. L trol cable, and Edson, D. L Fineman, W. E. Gunther,11. D.
2.
Boston Insulated Wire (filW) Bostrad 7E, Ilaynes, K. R. Iloopingarner, M. J. Jacobus, D.11.
Jarrell, R. C. Kryter, II. L Magelby, G. A. Murphy, 2-conductor with shield and drain,16 AW,G, and M. Subudhi, " Insights Gained from Aging ife-11ypalon Jacketed, ethylene-propylene rubber search," Brookhaven National Laboratory, llNL-(EPR) insulated instrumentation cable.
NUREG-52323, March 1992.
Both of these cables are certified nuclear grade ca-Aging, if it is not properly managed, affects the op-bles, including certification as low-flame-spread cables.
erational safety of all reactor structures, systems, and They are among the most commonly used cable types in components, and it has the potential to increase risks to U.S. commercial reactors.
public health and safety. It is therefore essential to un-Since these cables were ' certified as low flame derstand the aging processes that occur in a system or spread, they were exposed to a fire that was more component so that they can be effectively managed.
severe than the standard exposure test.This exposure The NRC's Nuclear Plant Aging Research (NPAR) was based on work performed by Factory Mutual Re-Program has identified those components and systems search Corporation (FMRC). FMRC found that a dif-that have a propensity for age-related degradation and 55 NUREG-1377
Main Citations and Summaries I
has eviduated methods for detecting and mitigating ag-search has shown to be beneficial for managing the ag-ing effects.
ing of that component or system.
His report was developed to consolidate the NUREGiCR-5646, R. Steele, Jr., and' M. E. Nitzel, I
research results from the assessments of component
" Piping System Response During High Level and system aging sponsored by the NRC for use by in-Simulated Seismic Tests at the Heissdampfreaktor dustry and by NRC in understanding and managing the Facility (SHAM 'Ibst Series)," Idaho National aging of systems, structures, and components in nu.
Engineering Laboratory, EGG-2655, July 1992.
clear power plants. %c report discusses aging-related This report describes the analysis and results from problems, operating experience, solutions to aging the "Servohydraulische Anregung Mashinetechnik" problems, and reference documents. The input for this (SH AM) seismic research program, in which the NRC report was provided by the NRC contractors who were and the Idaho National Engineering Laboratory responsib!c for the research, including:
(INEL) participated. The program was conducted by Kernforschungszentrum Karlsruhe (KfK, the Nuclear 1.
Brookhaven National Laboratory Research Center) at the decommissioned Heissdamp-Components: Battery chargers, inverters, mo.
freaktor (HDR, the superheat reactor) located near tors, and motor control centers Frankfurt, Germany. The SHAM experiments con-Systems:
Compcment cooling water, con-sisted of the direct excitation of a piping system called the Versuchskreislauf (VKL) that was modified to m, -
trol rod drive (Westinghouse), in-strument air, and residual heat a naturally aged U.S.-made 8-m. motor-operated gate valve. I'he piping system was excited at seismic lev-remmat els of 100% of the safe shutdown earthquake (100%
2.
Idaho National Engineering Laboratory SSE), increasing up to 800% of the SSE using two large 40-ton servohydraulic shakers mounted to the HDR I
Component:
Batteries containment building and attached to the piping sys-Systems:
lE distribution, reactor protec-tem. Experiments were conducted with the piping sup-tion, high-pressure coolant injec-ported by six different piping support systems. These tion and core spray (BWR), and included support configurations typical of those com-high-pressure safety injection monly used in European power plants,' a typical stiff (PWR).
U.S. system, and a very flexible system.This report spe-cifically addresses the tests performed with the U.S.
3.
Oak Ridge National Laboratory stiff support configuration.The objectives of the INEL Components: Auxiliary feedwater pumps, check portion of the research included determining the safety valves, motor-operated valves, margins and failure modes of nuclear grade snubbers power-operated relief valves and.
and other support components, determining the effects block valves, - and solenoid-of support failurcs on piping response, and determining operated valves the effects of the dynamic loads on gate valve operabil-Systems:
Results from tests at input levels of 200% SSE, 4.
Pacific Northwest Laboratory 600% SSE, and 800% SSE were examined. The 100%
tests were not compared to design predictions because Components: Emergency diesel generators and of support failures at this level of input.The 200% SSE snubbers tests, with all dynamic supports operable, showed that Systems:
the design analysis predicted maximum stresses at the same kications where the maximum strains were re-5.
Sandia National Laboratory corded during the tests; also, the accelemtion histories Cornponents: Cables.
showed that the piping responses were generally in the The document includes a Summary of Research Re-same frequency bands as the predicted natural fre-sults and an Aging Assessment Guide. The Aging As.
- quencies, sessment Guide is more concise than the Summary of The 800% SSE h>adings caused overload failures of Research Results, and it focuses on the specific inspec-several snubbers with measured strains greater than tion activities to be considered when assessing the op.
yield.The timingof the failuresof threcof the snubbers crational readiness cf the component or system. De and the force and displacement data indicated that a Guide contains visual inspection techniques for detect.
ripper-effect failure phenomenon occurred. However, ing aging degradation, including external and internal even with the large displacements and strains, no physi-indicators and important operating parameters. In ad.
cal failure of the piping occurred.
dition, the Guide lists those activities associated with Except for two cases, all snuboer failures occurred maintenance, operations, design, and testing that re-at loads well above their design loading. In one case, a l
'NUREG-1377 56
Main Citations and Summaries load of 8.67 times the design rating was sustained prior not involved in either the submergence testing or the to failure. One snubber and its replacement from the high-temperature steam testing and are therefore not same manufacturer failed well below their design discussed in this report.
loads. Hoth snubbers were returned to the manufac-
'Ihe submergence test used a solution close to that turer for inspection and analysis. No further informa-specified by IEEE 383-1974 for chemical spray during -
tion had been received on this at the time this report LOCA simulations. The solution was maintained at was prepared.
about 95*C during the exposure, which lasted a total of The U.S.-made 8-in. motor-operated gate valve op-1000 hours. The high-temperature steam test involved crated smoothly during all tests in the SHAM series.
exposure to steam at temperatures as high as 400*C Some limit switch chatter was observed; however, the (750*F). Cable insulation resistances were monitored limit switch c(mtacts did not stay open long enough to throughout the high-temperature steam test and at dis-cause the motor controller circuit to interrupt the cur-crete times during the submergence test. Dielectric tent flow to the motor.1he data showed that even un-withstand testing was performed before the submer-der the most severe structuralloading experienced dur-gence and high temperature steam tests and at the end ing the tests, the valve operated smoothly.
of the submergence test. The cables that passed the lhe test results indicate that sufficient safety mar.
post-submergence dielectric test were subsequently gins exist when commonly accepted design methods are wrapped around a mandrel with a diameter 40 times applied and that piping systems will likely maintain that of the cable and exposed to a final dielectric with-stand test.
their pressure boundary in the presence of severe load _
ing and with the loss of multiple supports.
The conclusions from this study are:
1.
The results of the high-temperature steam NUREGICR-5655, M.J. Jacobus and G.F. Fuchrer,
" Submergence and liigh lemperature Steam test indicate the approximate thermal failure 1bsting of Class IE Electrical Cables." Sandia thresholds for each cable type. EPR cables National Laboratories, SAND 90-2629, May 1991.
generally survived slightly higher tempera-tures (370-400*C) than XLPO cables Many types of cable are used throughout nuclear power plants in a wide variety of applications. Cable (299-388'C) during the high-temperature qualification typically includes thermal and radiation steam exposure. The XLPO-insulated con-ductors had no insulation left at the end of the aging intended to bring the cable to a defined "end of.
life" condition before exposure to a simulated design.
high-temperature steam test. Silicone rubber basis accident. In some instances, cables must be quali.
failed in the range of 396 to 400 *C, Kerite FR fied for submergence conditions. High-temperature at 372* to 382 C, and polyimide at 399"C.
steam testing of cables (beyond the design basis) is not 2.
The results of the submergence test indicate currently required for qualification.
that a number of cable types can withstand 1his report describes the results of high-tempera.
submergence at elevated temperature, even ture steam testing and submergence testing of 12 dif.
after exposure to a loss-of-coolant accident ferent cable products. The cable products tested are simulation. XLPO cables generally performed j
typical of cables used inside containments of U.S. light better than EPR cables in the submergence water reactors and inelude primary insulations of cross.
test and in the post-submergence dielectric linked polyolefin (XLPO), ethylene propylene rubber testing. By the end of the final dicIcctric test (EPR), silicone rubber (SR). polyimide, and chlorosul.
(after the mandrel bend), only 1 of 11 (9.1%)
fonated polyethylene (CSPE).
XLPO-insulated conductors,17 of 20 (85%)
EPR-insulated conductors, and 6 of 8 other These cables were part of a larger test program in cables (silicone, Kerite FR, and Rockbestos which four sets of cables were subjected to simultane-co xial) had failed.
ous thermal and radiation aging for 0 (unaged),3,6, and 9 months. Following the aging, each set of cables was 3.
A number of cables that performed well dur-exposed to a simulated loss-of-coolant accident ing the submergence test failed post-submer-(LOCA).
gence dielectric withstand testing (either be-fore or after the mandrel bend). This indicates The submergence test was performed on the cables that the IEEE 383 dielectric withstand tests that had been aged for 6 months and then exposed to and mandrel bends can induce failure of other-the simulated LOCA, and the high-temperature steam
_1 wise functional cables. Note that this conclu-test was performed on the cables that had been aged for 3 months and also exposed to the LOCA. Both of these sion does not imply a criticism of the IEEE 383 tests were added to the scope of the test program be-requirements, which are intended to provide a level of conservatism in the testing.
cause the aged cables had completed all planned test.
ing and mary of the cables had not yet failed._ The 4.
The IEEE 3M dielectric withstand tests are j
unaged cables and the cables aged for 9 months were very severe even if a mandrel bend test is not i
57 NUREG-1377
Main Citations and Summaries performed. This is evidenced by the failure of attributed to the control rod drive system, and (4)infor-nine conductors and the near failure of three mation exchange with nuclearindustry CRDM mainte-more in the post-submergence dielectric with-nance experts.
stand test, only two of which were showing a Nearly 23% of the NPRDS CRD system component strong indication of degradation during the failure reports were attributed to the CRDM..He submergence test.
CRDM components most often requiring replacement em a na an a gae ar -
NUREG/CR-5693, R. Lofaro, W. Gunther, M.
s als. y Subudhi, and B. Lee,"Agin Assessment of The predominant causes of aging for these seals Component Cooling Water stems in Pressuriled are mechamcal wear and thermally mduced embrittle-Water Reactors-Phase II,"
rookhaven National ment. More than 59% of the NPRDS CRD system fail-Laboratory, BNL-NUREG-52283, June 1992.
ure reports were attributed to components that make A two-phase aging analysis of component cooling up the hydraulic control unit. He predominant hy-water (CCW) systems in pressurized water reactors draulic control unit components expenencing the ef-(PWRs) was performed. In Phase I (NUREG/
fects of sersice wear and agmg are valve seals, discs, CH-5052, July 1988), the effects of aging were charac-seats, stems, packing, and diaphragms.
terized, and the predominant failure modes, aging Since CRDM changcout and rebuilding is one of mechanisms, and components susceptible to aging deg.
the highest dose, most physically challenging, and most radation were identified. Failure rate trends were ex.
complicated maintenance activities routinely accom-amined, and their effect on time-dependent system un-plished by BWR utilities, this report also highlights re-availability was investigated.
cent innovations in CRDM-handling equipment and In this Phase II study, the methods used to manage rebuilding tools that have resulted in significant dose aging degradation in the CCW system were studied. In-reductions to the maintenance crews using them.
formation was collected and analyzed on inspection, NUREG/CR-5700, A. C. Gehl and E. W. IIagen, surveillance, monitoring, and maintenance (ISM &M)
" Aging Assessment of Reactor Instrumentation and techniques. Also investigated were advanced tech.
Protection Systems Components." Oak Ridge niques for detecting and mitigating aging degradation National Laboratory, ORN1/FM-11806; July 1992.
of construction materials and their relationship to ag-A study of the aging-related operating experiences ing mechanisms, as well as codes and regulatory re.
throughout a five-year period (1984-1988) of six generic quirements, instrumentation modules (indicators, sensors, control.
Results of this aging study show that there are spe-lers, transmitters, annunciators, and recorders) was.
cific basic ISM &M practices that all plantt perform.
performed as a part of NRC's NPAR Program.These These practices are typically required by Code or plant.
six categories were selected because of their impor-technical specifications, and they are capable of detect.
tance m all the operations of safety-related instrumen-ing degradation arising from many aging mechanisms.
tation and control (I&C) systems and because they llowever, they are not comprehensive enough to com-have not been previously reviewed within the NPAR pletely control all aging degradation. 'Ib more effec _
pmgram.
tively control aging, ISM &M programs should include The issue of aging of safety-related control systems a cambination of basic and supplemental practices that in a world of mereased performance demands is a rela-are selected on the basis of plant-specific conditions.
tively new one. If left unchecked, components of that The report presents listings of supplemental practices, system can lead to an impairment of continued safe op-I correlated with the respective aging mechanisms each eration of a nucicar power plant.
of the practices helps to detect or mitigate.
In this report, the effects of aging from operational and environmental stressors were characterized by the NUREGICR-5699, R.it Greene, " Aging and Sem.ce Wear of Control Rod Drive Mechamsms for BWR results depicted in Licensee Event Reports (LERs).
Nuclear Plants " Volume 1, Oak Ridge National
.lhe data are graphically displayed as frequency of-
- Laboratory, ORNL -6f((>/V1, November 1992, events per plant year (on the vertical axis) for various His Phase i Nuclear Plant Aging Research (NPAR) operating plant ages (on the horizontal axis, ranging study examines the aging phenomena associated with from 1 to 28 years). Such graphs help determme agmg.
re fa um n s an patterns of events.
BWR ccmtrol rod drive mechanisms (CRDMs) and assesses the merits of various methods of" managing" Ihree main conclusions were drawn from this this aging. Informatin for this study was acqw ed from study.
(1) the results of a special CRDM aging questionnaire
- 1. I&C modules make a modest c(mtribution to distributed to each U.S. BWR utility, (2) a first-of-its-safety-significant events.
kind workshop held to discuss CRDM aging and main-17% of LERs during 1984-1988 dealt with e
tenance concerns,(3)an analysis of the Nuckar Plant malfunctions of the six I&C,uodules Reliability Data' System (NPRDS) cases of failures studied.
NURI G-1377 58
Main Citations and Summaries e 28% of the LERs dealing with these I&C ance standards or guidance documents for acceptable module malfunctions were aging-related design basis tests.
(other studies show a range of 25-50%).
In order to perform an in-depth study, the authors
- 2. Of the six modules studied, indicators, sensors, have reviewed all of INEL past, current, and ongoing and controllers account for the bulk (83%) of valve research, including test data. This review re-aging-related failures.
vealed that the use of in situ test results to estimate the
- 3. " Infant mortality" (during the early life cf in.
msponse of gate and butterfly valves at design basis struments) appears to be the dominant aging-conditions is possible, but some caveats are necessary.
related failure mode for most I&C module cate.
It was found that the methods used by industry to pre-gories (with the exception of annunciators and dict the required stem force for a gate valve and re-recorders, which appear to fail randomly).
quired stem torque for a butterfly valve are flawed.
Also, it was observed that satisfactory performance of NUREG/CR-5706, D.A. Casada, "NRC llulletin 88 04:
MOV diagnostic systems is possible, but very few of the Potential Safety-Related Pump Loss-An Assess-currently available systems measure enough parame-ment of Industry Data " Oak Ridge National Labo-ters to be useful' ratory, ORNL-6671, June 1991.
Additionally, this report discusses INEL participa-Nuclear utility plants are required to periodically
""""'8""
8" test safety-related Pumps to demonstrate Pr Per func-ance documents for acceptable design basis tests. Such tionmg of the pump. Historically,a substantial numher participation includes an extensive information ex-of these pumps have been routinely tested at the flow change with the American Society of Mechanical Engi-rate available through the pump's minimum flow recir-neers standards writing committees and attendance at culation flow path, which in many cases was sized t working group meetings.
avoid overheatingonly. It has become more widely rec-ognized that operation of a pump under low-flow con.
NUREG/CR-5754, K.H. Luk, "lloiling-Water Reactor ditions can result in hydraulically unstable conditions Internals Aging Degradation Study, Phase 1,'" Oak Ridge National Laboratory, ORNL/FM-11876, that can damage the pump, even though the rate of Septemb E.
flow is adequate for heat removal.
Th,is report documents the results of a study on the -
Nuclear Regulatory Comm.ission (NRC) Bulletin effects of aging degradation on 25 selected boiling-88-04 required utilities to examine (1) the potential for water reactor (BWR) internal components. The oper-dead-headmg of pumps due to parallel pump competi-ating environment inside a BWR pressure vessel pro- -
tion and (2) the adequacy of the minimum flow rate duces stressors that could lead to the development of provided for each safety-related pump. Utihties have aging-related degradation mechanisms. A data base reviewed the currently recommended mimmum flow containing aging-related failure information for the se-rates with pump vendors and have examined existing lected internal components is established using data system design provisions, operating controls, and his-from Licensee Event Reports. Results of the failure in-torical maintenance expenence*
formation survey identified two major aging-related Under the auspices of the NRC's Nuclear Plant Ag-degradation mechanisms for reactor internals: stress ing Research Program, Oak Ridge NationalI aboratory corrosion cracking (SCC) and fatigue. SCC includes in-i has reviewed utility responses to Bulletin 884. An as-tergranular SCC and irradiation-assisted SCC sessment of the industry response and resultant con-(IASCC).
clusions and recommendations are presented.
Strategies for controlling and managing aging deg-radations are based on understanding the relationship NUREG/CR-5720, R. Steele, Jr., J. C. Watkins, K. G.
DeWall, and M. J, Russell, " Motor-Operated Valve between stressors and the associated aging-related Research Update, Idaho National Engineering degradation mechanisms. The implementation of a-Laboratory, EGG-2643, June 1992.
plant hydrogen water chemistry (HWC) program is The U.S. Nuclear Regulatory Commission (NRC) is considered to be a promising method for controlling -
3 supporting ' motor-operated valve (MOV) research at SCC, which is the m. ore prevalent problem for BWRs.
the Idaho National Engineering Laboratory (INEL).
Flow-induced vibration (FIV) is the major cause of fa-The MOV tests provide the basis for assessing the ef-tigue problems in BWR internals. FIV problems are re-fects of various factors on the valves and for evaluating solved either by eliminating the excitation sources or by -
the current industry standards. His report addresses detuning the structure from input excitations. Ques-sew tal research itemn including _ (1) the use of in situ tions xmain concerning the effectiveness of HWC in test nsults to estimate the response of a valve at design mitigating SCC in internals and in the assessment of i
basis conditions, (2) the methods used by industry to high-cycle fatigue in a corrosive emironment;
)
predict required valve stem forces and torques, (3)
Vibration monitoring, based on neutron noise guidelines for satisfactory performance of MOV diag-measurements and trending studies, is an inspection nostics systems, and (4) the development of perform-method that can provide early. failure detection 59 NUREG-1377
Main Citations ano Sc. conics capability and can improve the effectiveness of current experiences of Ihe research team, and requests from plant inservice inspection programs. Ilowever, the the utilities an manufacturers.
hrge water gap and the lack of existing ex-core neutron These evaluations of degmdation conditions
- flux monitors may hinder the use of neutron noise showed that generally accepted current nuclear plant vibration measurements in llWRs.
maintenance practices do not always detect the effects of significant aging mechanisms. This provides insight NUREG/CR-5762, J. E,Gleason, " Comprehensive Ag-into the reason failures of safety-related relays have oc-1 I at ri s, 1 1 hia ch 19 curred in service in spite of a comprehensive mainte-nance program.
This report describes the results of a comprehen-sive aging assessment of relays and circuit breakers that At specific plants (Catawba Nuclear Station and wascompletedaspartof the NRCNuclearilant Aging Nine Mile Point Unit 1), the research team witnessed Research (NPAR) I rogram. Ihis is a Phase II report plant maintenance personnel performing routine and the research has followed the established Ni AR m intenance on relays and circuit breakers. Cop.ies of strategy described in Revision 20f NUREG-1144."Nu-pmcedures were obtamed, results of plant clear Plant Aging Research (NPAR) Program Plan" maintenance tests reviewed, and engmcenm,g and m intenance perse:.act mterviewed. Additionally, (June 1991). Relays and circuit breakers are important safety-related equipment that perform critical fune-nonin trusive lSM procedurcs of infrared pyrometry, tn-tions in the operation and control of nuclear power frared scanning, and vibration testing were demon-strated.
plants.
The research of this report has particular signifi-It is a challenge to a good preventive majntenance cance with resocct to Generic Letter 83-28," Required program to be sensitive to the effects of agmg. Early Actions liase'd on Generic Implications of Salem identification of age-related degradation increases the ATWS Events," Information Notice 84-20, " Service pnybability that the safety signtficance of the identified Life of Relays in Safety-Related Systems," and IE umg process can be mmamized.
Ilulletin 84-02, "Fa ures of General Electric 'lipe n
A comprehensive effort was undertaken to verify liFA Relays in Use in Class IE Safety Systems " These improved inspection, surveillance, and monitoring documents require licensees to have preventive (ISM) methods. The Phase 11 effort was accomplished maintenance and surveillance programs for circuit in four major elements: an investigation into current breakers and relays.De research providesinformation and advanced ISM methods, tests of aged relays and on the effectiveness of the required ' preventive circuit breakers, tests of degraded relays and circuit maintenance methods and demonstrates that im-breakers, and in situ tests-proved ISM procedures were more effective than cur-Current and advanced ISM methods were ascer-rent industry practices for detecting aging and mitigat.
tained by soliciting information from nuclear and non-ing its effects on specific devices and components.
nuclear utilitics, relay and circuit breaker manufactur-Finaby, specific recommendations were made for -
ers, and maintenance facilities. Testing of naturally the introduction of changes in current nuclear industry aged devices was performedJIbst specimens for each of practices of inspection, surveillance, and maintenance the five relay types (auxiliary. control, electronic, pro-on relays and circuit breakers.These recommendations tective, and timing) and two circuit breaker types were based on significant research results (described in (molded case and metal clad) were obtained from nu-this report), that. identified improved inspection, sur-clearand nonnuclear utilities and manufacturers. A to-veillance, and monitoring methods, Implementation of l
tal of 39 specimens were tested, utilizing current and these methods could minimize the impact of aging and improved ISM methods.
result in more cost-effective maintenance of relays and Eleven specimens of relays and circuit breakers circuit breakers.
were purposely degraded and the ISM methods imple-NUREC/CR-5772, M.J. Jacobus, " Aging, Condition mented after each degraded condition. The purpose of Mor itoring.and Loss-of roolant Accident (LOCA) these degradation tests was to evaluate the effective-Tests of Class IE Electrical Cables," Volume 1, 1
ness of the method to detect or predict the level of deg.
Sandia National Laboratorics, SAND 91-1766/1, radation. This also provided some quaniifiable parame-August 1992.
ters of the extent of degradation. The degradations -
This report describes the results of aging, condition chosen for each relay and circuit breaker type were pur-monitoring, and accident testing of crosslinked polyole-posely severe, but for the most part, did not cause total fin (XLPO) cables. Three sets of cables were aged for l
loss of operability of the device. Dus, an attempt was up 10 9 months under simultaneous thermal (cie 100*C) -
')
made to simulate the worst state of deterioration or and radiation (cie 0.10 kGy/hr) conditions. A sequential degradation prior to failure to operate. The degrada-accident consisting of high dose rate irradiation (ce 6 tions were chosen based on a review and evaluation of kGy/hr) and high-temperature steam followed the the failure modes and mechanisms reported in Phase I, aging. The test results indicate that most properly NUREG-1377 60
- -=
~ -
l Main Chations and Summaries installed XLPO cables should be able to survive an ac-NUREG/CR.5779, J. C. Moyers " Aging of Non Power-cident after 60 years for total aging doses up to 400 kGy Cycle Heat Exchangers Used in Nuclear Power and for moderate ambient temperatures on the order Plants," Vol.1. Oak Ridp,2.e National Laboratory, of 50-55'C (potentially higher or lower, depending on ORNL-6687/VI, July 199 material-specific activation energies). Mechanical lhis report presents the results of the Phase I as-I measurements (primarily elongation, modulus, and sessment of the time-related (aging) degradation of density) were more effective than electrical measure-non-power-cycle heat exchangers used in safety-ments for monitoring age-related degradation.
related systems or that provide normal operating capa-I NUREG/CR-5772, M.J. Jacobus, "Ag'
, Condition sp n ce of Nudcar Regulatory Ro Monitorin and Loss-of-Coolant cident (LOCA) scarch of the USNRC as an element of the ongoing Nu-Tests of Cfa,ss 1E Electrical Cables," Volume 2, clear Plant Aging Research Program.The objectives of Sandia National Laboratories, SAND 91-1766/2, November 1992.
this Phase I research effort were to review operating -
experience and other information, to identify failure This report describes the results of aging, condition m des and causes resulting from aging, and to identify monitoring, and accident testing of ethylene propylene measurable parameters that might provide a better m-rubber (E PR) cables. Three sets of cables were aged for dication of equipment condition.
up to 9 months under simultaneous thermal (m 100 *C) and radiation (u 0.10 kGy/hr) conditions. A sequential lhe report briefly reviews the design and applica-accident consisting of high dose rate irradiation (= 6 tion of the heat exchangers in both PWR and HWR kGy/hr) and high-temperature steam followed the plants. lypical design characteristics and materials of aging. Also exposed to the accident conditions was a construction are given for the various applications. Op-fourth set of cables, which were unaged. The test erational stressors are categorized and discussed.
results indicate that most properly installed EPR cables Operating events described in data bases for' nu-should be able to survive an accident after 60 years for clear power plants and in nuclear industry reports were total aging doses on the order of 150-200 kGy and for examined. These data bases included (a) the licensee moderate ambient temperatures on the order of event report file as cataloged in the Sequence Coding 45-55"C (potentially higher or lower, depending on and Search System maintained by ORNUs Nuclear Op-material-specific activation energies and total radiation crations Analysis Center, (b) the Nucicar Plant doses). Mechanical measurements (primarily elonga-Reliability Data System compiled by the Institute for tion, modulus, and density) were more effective than Nuclear Power Operations, (c)- Nuclear Power electrical measurements for monitoring age-related Experience published by Stoller Power Inc., and (d) degradation.
maintenance records for a two-unit PWR plant as fur-nished by a cooperating utility. A total of 710 reported NUREG/CR-5772, M.J. Jacobus, " Aging, Condition events were examined. Of these,279 events involved
, Monitoring, and Loss-of-Coolant Accident (LOCA) interfluid leakage, 217 involved external leakaEe,156 Ibsts of Class 1E Electrical Cables, Volume 3, San-involved tube-side flow hkickage, and 25 m.yolved im-dia National Laboratories, SAND 91-1766/3, paired heat transfer; the remaining 33 reportings re.
- November 1992.
lated to miscellaneous events.
This report describes the results of aging, condition monitoring and accident testingof miscellaneouscable There are only minimal regulatory or Technical types. Three sets of cables were aged for up 109 months Specification requirements for inservice inspection and under simultaneous thermal (= 100 *C) and radiation testing, and they are limited primarily to those inspec-(= 0,10 kGy/hr) conditions. A sequential accident con.
tions and tests required to maintain the integrity of the sisting of high dose rate irradiation (u 6 kGy/hr) and pressure-containing boundary.The general philosophy high-temperature steam followed the aging. Also ex.
of plant operators regarding flow blockage and leakage posed to the accident conditions was a fourth set of is that repairs and maintenance will be done as re-cables, which were unaged. The test results indicate quired. Improvements to this philosophy that could that, properly installed, most of the various miscella.
lead to enhanced reliability are not apparent, and no -
neous cable products tested should be able to survive such improvements are suggested in the report.
an accident after 60 years for total aging doses of at Inservice testing to determine the heat transfer ca-least 150 kGy or higher (depending on the material) pability of the heat exchangers has normally been done and for moderate ambient temperatures on the order only when possible degradation was indicated from ob-of 45-55 *C (potentially higher or lower, depending on servation of process parameters; scheduled perform-material-specific activation energies and total radiation ance testing was not normally done. Ilowever, largely doses). Mechanical measurements (primarily elonga-as a result of NRC concerns for the capability of safety-tion, modulus, and density) were more effective than related service-water-cooled heat exchangers to per-electrical measurements for monitoring age-related form as required under accident c(mditions and the re-i degradation.
sulting issuance of Generic Letter 89-13 (" Service 61 NUREG-1377
Main Citations and Summaries Water System Problems Affecting Safety-Related valve has been developed from the basic prin-Equipment," 10ly 18,1989), plant-specific inservice ciples (balance of forces equation). Compari-performance testing programs are being developed by sons against data supplied by Duke Power plant owners. In addition, the Operation and Mainten.
Company have confirmed that the methodol-ance Committee of the Americtm Society of Mechani-ogy is sound, and there is good quantitative cal Engineers is developing standards to address both agreement between analytical predictions and vibration monitoringand inservice performance testing actual test results.
of heat exchangers.
e
'Ihe results of a comprehensive review of fric.
NUREG/CR-5783, E. Grove and W. Gunther, " Aging tion and galling data are documented in this Assessment of the Combustion Engineering and report to provide a rational basis for selecting Habcock & Wilcox Control Rod Drives," Brook.
an appropriate coefficient of friction for a haven National Laboratory, UNL -NUREG-52299, given application.
I""" '7 IM The concept of index of contact stress severity e
The effects of aging upon the Habcock & Wilcox has been introduced to determine whether or (B&W) and Combustion Engineering (CE) control rod not a gate valve will behave predictably under l
drive (CRD) systems have been evaluated. For this fluid-flow forces. Preliminary analyses to cal-study, the CRD system boundary included the control culate localized contact stresses at the disc-to-l rod assemblics, guide tubes, control rod drive mecha-guide contact and at the disc-to-downstream nism, control system components, rod position mdica-seat contact under disc-tilting conditions leave tion components, and cooling system. Detailed opera-been developed.
tion experience data for 1980 to 1990 was evaluated to Significant factors that affect the opening e
identify the predominant failure modes causes.and ef-fects. The results of this evaluation, along sith an thrust requirements of a gate valve have been assessment of component material and operating envi.
identified, and quantitative methods that can ronment, lead to the conclusion that both the B&W be used to diagnose valve opening problems and CE CRD systems are susceptib!c to age degrada.
have been documented.
(.
tion. Failures of the CRD system have resulted in sig-Improvements in gate valve designs to make e
nificant plant effects, including power reductions, plant them less sensitive to pressure / thermal tran-shutdowns, scrams, and ESF actuations.
sients and external pipe loads have been iden-l Information on current plant system inspection and tified, and some quantitative examples are in-maintenance practices were obtained from two H&W cluded in the report to show the achievable plants and four CE plants through an industry survey.
degree of improvement.
The results of this survey indicate that some plants In summary, the Phase l research has been success; have modified the system, replaced components, and fulin completing the preliminary development ofim-established preventive maintenance programs, some of proved gate valve operability models.This can serve as which effectively address the aging issue while others an excellent foundation to continue further analytical do not. The potential application of some advanced and experimental development that is necessary to pro-monitoring inepection techniques are discussed.
vide reliable and proven gate valve operability models NUREG/CM-5807, I K. Wang and M. S. Kalsi, to the nuclear power industry.
" Improvements M Motor Operated Gate Valve Design and Predic6nn Models for Nuclear Power NUREG/CR-5848, J.S. Dukelow,."Recordkeeping Plant Systems," Kalsi Engineering, Inc., KEI No.
Needs'Ib Mitigate the impact of Agmg Degrada-1721, May 1992.
tion, Pacific Northwest Laboratory, PNL-7987, October 1992.
This report documents the results of Phase I re-
,Th.is report discusses techm. cal issues associated search proposed and conducted by Kalsi Engineering, Inc., to improve the operability of motor operated gate with the role of nuclear plant records systems m under-valves in nuclear power plants. Phase I research, standing and managing the a6g of nuclear plant com-l funded by the Small Business Innovation Research ponents, systems, and structures. It considers both the I
(SUIR) program, resulted in the following major ac-types of techmcal data useful for verifying continued complishments:
safe operation and the use of new technoiogy for up-grading records systems. Specific topics reviewed in-Opening and closing thrust equations for the elude the need for maintenance and reliability data, op--
e l
common types of gate valves used m U.S.
crational history data to support the assessment of re-nuclear power plants have been developed and maining fatigue life, comprehensiveness and usability documented.
of the engineering design basis, improvement of the An analytical methodology to predict inertial data input process, and conversion of existing records e
L thrust overshoot in a motor operated gate into machine-readable forms.
NUREG-1377 62
l:
Main Citations and Summarien
'lhe report concludes that successful management NUREG/CR-5941, D.A. Casada and M.D. Tbdd, "A of nuclear plant aging will require improvement of ex-Characterization of Check Valve Degradation and isting plant records systems; several generic and specif-Failure Experience in the Nuclear Power Industry,"
ic recommendations are provided. The computer-Oak Ridge National Laboratory, ORNle6734, based technology for meeting this need and imple-September 1993.
menting these recommendations already exists and can Check valve operating problems in recent years be implemented at a reasonable cost.
have resulted in significant operating transients, in-creased cost, and decreased system availability. As a re-NUREGICR-5870, D. P. Brown, E. V. Werry, and D.
sult, additional attention has been given to check valves E. Blahnik,"Results of LWR Snubber Aging Re-by utilities (resulting in the formation of the Nuclear search," Pacific Northwest Laboratory, May 1992-Industry Check Valve Group), as well as the NRC and Snubbers are safety-related devices used to restrain the American Society of Mechanical Engineers Opera-undesirable dynamic loads at various piping and equip-tion and Maintenance Committee.1hese organiza.
ment locations in nuclear power plants. Each snubber tions have the fundamental goal of ensuring reliable must accommodate a plant's normai thermal move-operation of check valves.
ments and be capable of restraining th; maximum off-A key ingredient to an engineering-oriented reli-normal dynamic loads postulated for its specific loca-ability improvement effort is a thorough understanding tion. Snubbers are subject to the effects of aging; the of relevant historical experience. A detailed review of factors that contribute to the degradation of their historical failure data, available through the Institute safety performance need to be better understood.
of Nuclear Power Operation's Nuclear Plant Reliability Snubber operability is mandated by the regulations.
Data System, has been conducted. The focus of the re-which stipulate that systems, structures, and compo-view is on check valve failures that have involved signif-nents, including sn ubbers, be designed to withstand the icant degradation of the valve internal parts. A variety effects of normal and off-normal dynamic phenomena.
of parameters are condidered, including size, age, sys.
In the mid 1980s, the NRC recognized the need to en-tem of service, method of failure discovery, the af.
hance snubber performance through aging studies and fccted valve parts, attributed causes, and corrective ac-improved service-life monitoring techniques. The tions.
NRC's Nuc! car Plant Aging Research (NPAR) Pro ~
gram Plan of 1987. provided the vehicle and the NUREG/CR-6001, G.D. Buckley, R.D,0 ton, A.B. Johnson, Jr., LL Larson, " Aging Assessment sponsorship to undertake prehmmary mvestigations of BWR Standby Liquid Control Systeres," Pacific into sn ubber performa nce a nd aging. Pacific North west Northwest Laboratory, PNL-8020, August 1992.
Laboratory (PNL) staff and its subcontractors, Lake Pacific Northwest Laboratory conductM a Phase I Engineering and Wyle Laboratories, visited 13 plants, aging assessment of the standby liquid cor.*.rol (SLC) conducted interviews, collected relevant data, and per' system used in boiling-water reactors. The study was formed the snubber research.
based on detailed reviews of SLC system omponent This report describes the Phase 11 NPAR in-plant and operating experience information obtained from aging research conducted to enhance the understand-the Nuclear Plant Reliability Database Systert. the Nu-ing of sn ubber aging and its consequences. The in-plant clear Document System, Licensee Event Reports, and aging research was based on a research plan by Brown other databases. Sources on sodium pentaborate, re-et al., which clarified the relationship of snubber aging rates, and boric acid, as well as the effects of environ.
to snubber degradation and identified additional infor-ment and corrosion in the SLC system were also re-'
mation on aging that requires further investigation and viewed to characterize chemical properties and corro-analysis for both hydraulic and mechanical snubbers.
sion characteristicr d borated solutions.
This report presents the results of snubber aging re.
Relatively few SLC component failures were attrib-search, including methodology, evaluation, testing and uted to sodium pentaborate buildup or corrosion. The failure data, as well as service-life monitoring recom-leading aging degradation concern to date appears to mendations that emphasize distinguishing between be setpoint drift in relief valves, which has been discov-aging-and nonaging-related snubber failures. The cred during routine surveillance and is thought to be graphics, tables, and supporting text clearly illustrate caused by mechanical wear. A highersetpoint results in -
this distinction. The results of reported work support loss of system overpressure protection, and a decrease the perspective that snubber failures are closely related in setpoint results in a reduction of boron injection rate.
to age-related degradation caused by inservice opera.
Degradation was also observed in pump seals and inter-tional environmental influences, e.g., vibration and nal valves, which could prevent the pumps from operat-elevatec temperat ure. Because there is a lack of infor-ing as required by the technic d specifications in gener-g mation on service to mechanical snabbers, special em-al, however, the results of the Phase I nudy indicate phasis was placed on gathering such information for that age-related degradation of SLC systems has not these devices.
been serious.
1 63
. NUREG-137/
j Main Citations and Summaries NUREG/CR-6029, W.K. Winegardner, " Phase ! Aging of adequate monitoring. Omission of scheduled main.
l Assessment of Nuclear AirJIreatment System tenance and human errors also contribute to failures.
HEPA Filters and Adsorbers," Volume 1, Pacific Northwest Laboratory, PNL-8594, August 1993.
NUREG/CR-6048, K.H.;Luk, Pressurtzed-Water Reactor Internals Agmg Degradation Study, Phase t
A Phase I aging assessment of high-efficiency par-I." Oak Ridge National Laboratory, ORNU ticulate air (HEPA) filters and activated carbon gas ad-TM-12371, September 1993.
sorption units (adsorbers) was performed by the Pacific This report is a summary of the results of a Phase 1 J
Northwest Laboratory as part of the NRC's Nuclear study on the effects of aging degrada! inns in
]
Plant Aging Research Program. Information concern-pressurized-water reactor (PWR) internal compo-j ing design features; failure experience; aging mecha*
nents. Westinghouse, Combustion Engineering, and i
nisms, effects, and stressors; and sm veillance and mon-Babcock & Wilcox reactors are included in the study, itoring methods for these key ai -treatment system Stressors associated with the operating environ-components was compiled. Over 1100 failures, or 12%
of the filter installations, wera reported as part of a ment inside the reactor pressure vessel provide condi-DOE survey. Investigators from other national labora-tions that are favorable to the development of aging-related degradation mechanisms. The dominant stres-tories have suggested that aging effects could have con-sors are flow-induced oscillatory hydrodynamic forces tributed to over 80% of these failures. Tensile strength generated by the reactor primary coolant flow. Results tests on aged filter media specimens indicated a de' of a survey of the component failure information iden-crease in strength. Filter aging mechanisms range from tified three major aging-related degradation mecha-those associated with particle loading to reactions that nisms: fatigue, stress corrosion cracking, and mechani-alter properties of scalants and gaskets. Low rad, ' ~
cat wear.
dine decontamination factors associated with the Three Mile Island acdent were attributed to the pre-Strategies for controlling and managing aging deg-mature aging of the carbon m the adsorbers. Mecha-radations are formulated based on the understanding nisms that can lead to impatred adsorber performance of the linkagebetween stressorsand agingdegradation mclude oxidation as well as the loss of potentially avail-mechanisms. Flow-induced vibration problems are re-able active sites as a result of the adsorption of pollut-solved by conventional engineering practices; by climi-ants. Stressors melude heat, moisture, radiation, and nating excitation sources or by de-tuning the structure airborne particles and contammants.
from input excitations. Uncertainties remaining in the.
assessment of aging effects on PWR internals include NUREG/CR-6043, D.E. lilahnik and R.E Klein, long-term neutron irradiation effects and the influence
" Phase I Aging Assessment of EssentialIIVAC of environmental factors on high-cycle fatigue failures.
Chillers Used m Nuclear Power Plants," Pacific An effective plant inservice inspection program will Northwest Laboratory, PNL-8614. September 1993.
ensure the structural integrity of reactor internals.
The Pacific' Northwest Laboratory conducted 'a Reactor internals can be replaced if it is deemed neces-Phase I aging assessment of chillers used in the essen-'
sary. Therefore, an inspection method with early fail-tial safety air-conditioning systems of nuclear power ure detection capability will further enhance the safety plants. Centrifugal chillers in the 75-to 750-ton refrig-as well as the efficiency of plant operations.
eration capacity range are the predominant type used' ORNL/NRC/LTR-91/25, D.A. Casada, " Throttled The chillers used, and air conditionmg systems served.
vary in design from plant to plant, it is crucial to' keep
- Valve Cavitation and Erosion," Oak Ridge National Laboratory, December 199L Available from the chiller internals very clean and to prevent the leakage NRC Public Document Room.
of water, air, and other contaminants into the refriger-In November of 1988, Brunswick plant mainte-ant containment system. Periodic operation on a week-nance personnel discovered significant k)calized cro-ly or monthly basis is necessary to remove moisture and sion of the valve body of a Unit I residual heat removal noncondensable gases that gradually build up inside (RiiR) valve,1-Ell-F017B (the maintenance was be-the chiller.This is especially desirable if a chiller is re-ing performed to repair the valve stem and back seat).
quired to operate only as an emergency standby unit.
'Ihe F0178 valve is a 20-inch Rockwell angle globe The primary stressors and aging mechanisms that valve that has historically been used to throttle RilR affect chillers include vibration, excessive tempera-flow. Excessive throttling of the valve had resulted in '
tures and pres et:res, thermal cycling, chemical attack, cavitation-induced erosion damage to areas immedi-nnd poor quahty cooling water. Aging is accelerated by ately downstream of the seat. Subsequent investigaikn
}
moisture, noncondensable gases (e.g., air), dirt, and indicated that erosion of valve bodies wasa generic con-i' other contamination within the refrigerant contain-cern for the other RHR valves used in the same service l.
ment system; excessive start /dop cycling and opera-
- (F017A on Unit 1 and F017A and F017B on Unit 2).
tion below the rated capacity. Aging is also accelerated The November 1988 event led to the issuance of N8C by corrosion and fouling of the condenser and evapora-Information Letter 89-01. The RHR valves used for tor tubes. The principal cause of chiller failures is lack -
suppression pool cooling (16-inch Anchor Darling l
l NUREG-1377 64
i Main Citations and Summaries globe type) were also found to have been damaged by The material used in valves and piping is significant
.avitation crosion.
in determining the rate of erosion. In general, hard-facing materials such as stellite have been used to mini-This study was conducted to identify the causes of crosion, valves most susceptible to erosive effects, his-emywe e cts Non-cobalt-bearing materials nm such as mckel alumm, ides have been demonstrated as torical crosion-related experience of valves and adja-I *Viding emsion resistance without "E C#P P
cent piping, and potential means of correcting the the potential radiolog.ical effects of cobalt-bearing ma-problems.
terial use.
The principal source of erosion damage in nuclear Besides erosion, another significant result of cavita-plant valves and piping is cavitation. Other contributors tion is vibration. Within a valve, it can result in cracked to erosion damage are corrosion and impingement by welds or loosened parts of the valve or other adjacent high-vehicity liquid and by abrasive particles. As fluid equipment.
passes through the minimal flow area within a valve The use of appropriate control valve trim is a pre-txidy, the pressure drops. Depending upon the system ferred means of addressing cavitation concerns, since it conditions and valve design, the pressure may drop be-can help minimize cavitation-caused crosion and resul-low the fluid-vapor pressure. If this occurs, the subse^
tant vibrations.
quent pressure recovery downstream of the valve causes the vaporpockets to collapse.The process of the PNI 5722, D. E. Blahnik and R. L. G<xximan, vapor pocket collapse (cavitation) results in substantial
" Operating Experience and Aging Assessment of energy dissipation. The energy dissipation may damage ECCS Pump Room Coolers," Pacific Northwest materialin the vicinity of the cavitation site and result Laboratory, October 1986.
in substantial component and system vibration.
This report provides a preliminary aging assessment of s fety-related room coolers for the emergency core Valves used in heavily throttled conditions are fre-cmHng system (ECCS) pump rooms m nuclear power quently the most susceptible to damage, mainly be-plants. N assessment conforms to the NRC NPAR cause of the fact that the system conditions that dictate progr m strategy and is based on limited the throttling are more likely to result in cavitation inf rmation obtamed through public and private data than are those for which throttling is not required or is bases, equipment vendors, utility contacts, literature minimal. Valves used in heat exchanger outlet or by-scarches, and expert opinion.
pass servicc orin other control functions are most likely to experience cavitation.
Description of the ECCS pump room cooler systems were based on FSARs and vendor-supplied m-Test indicated that cavitation existed throughout formation. Data from LERs, review of maintenance the range of 4,000 to 16,000 gpm with F017 valves, while requests at a reactor plant, and discussions with perscm-F024 valves started to indicate cavitation to 4,500 gpm.
nel that do utility repair and maintenance work were Cavitation was most prevalent at higherflow rates, and
' used to determine the operating experience of pump i
it was also noted that the k) cation of the cavitation room coolers. Failure modes, causes, frequency rates, moved throughout the valve's body as flow changed.
and methods of detection are summarized from the op-Subsequent investigation of seven other valves used crating records Maintenance actions and modifica.
in safety-related throttling service, including core spray tions needed as a result of the operator experience are (CS), high-pressure coolant injection (HPCI), and re.
addressed. Operational stressors are summarized, j
manufacturer recommendations for mamtenance and actor core isolation cooling (RCIC) systems revealed that one other valve, thc HPCI system full-flow test iso.
surveillance are listed, and aging and senice-wear i
lation valve F008, had experienced notable crosion, monitoring are briefly evaluated.
Hutterfly valves are particularly susceptible to cavi-PNie6287, K. R. Iloopingarner, B. J. Kirkwood, and tation because of the fact that the minimum pressurein P. J. Lonrecky, " Study Group Review of Nuc! car them is substantiallyless than that for a globe valve un-Service Diesel Generator'lestmg and Agin arch der the same overall pressure drop.
hation, Pacific Northwest 1. boratory, The senice water system was found to have experi-As part of the NPAR program,'the Pacific North-enced the most crosion-related damage to valves and west Laboratory is performing a diesel generator aging pipmg. l'he condensate and feedwater system has also assessment study. In the on-going NPAR Phase II of expero rd a significant nt nMr of crosion-related the aging stu6 efforts have been focused on aging failures.
mitigation and other success strategies for improving.
The Oak Ridge National Laboratory performed an nuclear plant diesel generator operation and mainte-assessment of the significimcc of valve body crosion, fo.
nance and also increasing its reliability.
cusing on the identification of valve types and applica-A study group of diesel experts, the authors of this lions susceptiW to crosion, report, met on April 29 and 30,1987, to resolve issues 65 NUREG-1377
.~.
4 t
Main Citations and Summaries P
on mitigatmg diesel generator aging and improving op-7.
Eliminate many unnecessary and partially re-erations, testing, and maintenance. 'Ihe focus of the dundant tests and engine starts in the study group was to (1) address the diesel generator ag-18-month test period (including those due to -
ing stressors resulting from the present periodic testing false signals),
practices of the nuclear industry and (2) propose 8.
Eliminate, where possible, short engine run potential mitigating measures. A new recommended times and excessive idle times, testing program was developed and is documented in The technical bases for such changes to the specifi-this report.The report lays out the conclusions and rec-cations are obtained from the NRC N PAR program and ommendations of the study group.The experts agreed from research sponsored by the Electric Power Re-that, if these recommendations are put into practice, scarch Institute (EPRI) and the Nuclear Safety Analy-many of the engine aging stressors (e.g., those due t sis Center (NSAC) operated by EPRI. Fast starts, fast fast start) could be reduced or chmmated; another con-loading, and the large number of test runs are cited as sequence could be a reduction of failures and an im-acting to increase diesel generator stress and wear. 'lhe provement in operability and reliability' results from this study confirm these stressors and add -
PN1-7516, K.R. Iloopi arner,"Emer ency Diesel excessive testing k> ads as another important stressor, Generator Technical pecifications Study Results,"
PNI 7823, A.D. Chockie, K.A. Bjorkelo, T.E. Fleming, Pacific Northwest Laboratory, March 1991.
W.B. Scott, and W.I. Enderlin " Maintenance
'lhis report covers a study in support of the NRC Practices'Ib Manage Aging: A Review of Several NPAR program on the effects of aging on emergency Technologies," lhetfic Northwest Laboratory, October 1991. Available from the NRC Public diesel generators (EDG). The research was performed cument Roont in two phases. PhaseI used plant operating experience, data, expert opinion, and statistical methods to produce The quality of a maintenance program directly af-a new data base related to aging, reliability, and opera _
fccts the abih,ty of a nuclear power plant to detect and tional readiness of nuclear sesvice diesel generators.
mitigate the effects of age-related degradation. The Phase II was chiefly concerned with measures for miti-Nuclear Plant Aging Research (NPAR) Program, which gating the effects of aging.
is sponsored by the NRC Office of Nuclear Regulatory Research, has contracted this research m order to ana-Insights from a number of sources indicate that lyze effective maintenance activities for the manage-i there are many opportunities for improving the man.
ment of aging of systems and components.
~
agement of EDG systems. Existing technical specifica-tions, for example, could be modified to yield sigmfi-The maintenance programs used by two commer-cant safety benefits by reducm, g direct effects of agmg cial industries and two military organizations were se.
4 lected for this study-the U.S. commercial airline in-l and mcreasmg system reliability. Thus techmcal speci-fications related to the management, testing, and reh-dustry, the Japanese nuclear power industry, the U.S.
Air Force B-52 bomber, and the U.S. Navy llallistic ability of emergency diesel generators were reviewed.
submarine.
lieneficial specifications were identified as were those that could adversely influence aging and reliability.
The maintenance programs of these four entities were examined in this report since they offer valuable Potential improvements in techm. cal specifications lessons for managing aging in the U.S. nuclear power and engine and system management aimed at reducing ndustry. Specifically, they indicate the need for an ef-l agmg effects and increasing reliability would:
fcctive maintenance program to manage the aging deg-1.
Significantly reduce the number of total en-radation of critical systems and comp (ments. Such a -
gine starts, maintenance program should include three basic ele.
2.
Reduce the load application rates for testing ments:
purpses by gradually adding load,
- 1. A systematic approach to the - conduct of ma ntenam tash, 3.
Reduce the EDG testing k> ads to 90% of the 2 Methods for monitoring and assessing mainte-continuous load rating or to the plant emer-t nance activities, and gency unit load, whichever is less, r
- 3. Mechanisms for feedback and corrective actions 4.
Increase the maximum EDG start time to25 t to improve maintenance effectiveness.
30 seconds, A systematic approach to maintenance includes a 5.
Make necessary changes to support the reli-comprehensive maintenance policy, clear maintenance ability emphasis of Regulatory Guide 1.9. Re' program objectives and goals, and the physical conduct vision 3, and delete statistical emphases, of maintenance based on the overall policy, objectives, 6.
Address fuel oil storage management to per.
and goals.
mit flexibility and the use of a large fraction of The structure of the maintenance program is im-stored fuct before replacement, portant to ensure that aging issues are addressed.This
(
NUREG-1377 66 l
Main Citations and Summaries i
analysis identified four elements inherent in an effec-of an overall untimely degradation of the plants and,in tive maintenance program that are also important to an particular, of nuclear safety-related systems and com-aging management program. He elements are select-ponents, often resulting in reactor shutdowns and ex-ing critical systems and components, u nderstanding ag-tended outages. Examples of major corrosion-induced ing through the collection and analysis of equipment degradation include intergranular stress corrosion performance information, mitigating aging by conduct-cracking (IGSCC) of piping in boiling water reactor
. ing necessary maintenance, and the use of feedback to recirculation bypass systems and denting, pitting, inter-improve the aging management program.
granular attack (IGA), and IGSCC of steam generator Critical systems and components can be selected by tubes in pressurized water reactors. These types of several methods. For example, the aviation industry's degradation involve signifiumt phenomena that have reliability-centered maintenance program uses a risk.
been widely recognized and investigated. Major invest-based approach to identify safety-significant equip _
ments have been made to detect and mitigate the ef-
- ment, fccts of these corrosion mechamsms. Numerous other mechanisms and phenomena have been observed in Understanding of aging processes is accomplished nuclear systems and components; some are obvious, by collecting and analyzing relevant operating charac-some subtle.
teristics and performance data on the equipment.De maintenance program should be designed to detect, The failure of a valve to open or close on demand, ident/y, and correct problems caused by aging mecha-for example, may be a consequence of corrosion. His nisms, such as corrosion, wear, and fatigue, before the and similar conuderations increase in importance as safety or reliability of the plant is impaired.
plants age but take on an additional dimension with i
Once the aging process of critical equipment is un, consideration of extending the licenses of the plantsbe-derstood, maintenance tasks designed to detect and y nd the current 40-yearpenod. Both the nuclearutik-correct equipment degradation can be selected and ties and the NRC are considering in detail the degrada-scheduled, includeing inspection, surveillance, equip-tion mechanisms that may have special significance in ment monitoring, replacement, and overhaul.
Feedback mechanisms ensure continual mainte.
Mitigation of corrosion impacts in nuclear facilities nance program refinement and improvement. Dese must involve more than technical considerations. It mechanisms may consist of specific maintenance activi-must involve attitudes of alertness and commitment on ties that serve as a basis for establishing and scheduling the part of regulators, plant management, and t_he future inspection and maintenance tasks. Feedback is plant work force. It also requires timely and ample al-also obtained from groups of personnel who evaluate location of resources.
-i and improve the maintenance programs. Feedback-mechanisms are vitalif the maintenance program is to There is a trend in the nuclear m. dustry to advance address changing conditions such as the degradation of 1he management of agmg phenomena, mcludmg corro-1
. plant equipment from aging..A summary of the si n. The fact that newcorrosion phenomenacontinue
' maintenance-related activities to address system and to emerge, however, provides evidence that principles component aging is presented in tabular form.
of corrosion control still need to be aggressively ap-plied. Lessons learned m, current reactor operation PNieSA-18407, A.lL Johnson, Jr., D.H. Jarrell, U.R need to be systematically and effectively applied to ex-Sinha, and V.N. Shah, " Understanding and tended operation and advanced reactor designs.
Managing Corrosion in Nuclear Power Plants,
Pacific Northwest Laboratory, August 1990.
This report defines the concept of understanding and managing corrosion,' references relevant regula.
, The concept of understanding and managing corro-tory and industry initiatives, and focuses on an overview sion m nuclear power plants is not new-m vanous of how the concept is being applied, drawing on results
'l forms, this main theme of the report has been applied from the NPAR program. The overview includes a throughout the development and maturing of nuclear brief survey of corrosion impacts on major structures, technology, %o often, however, understanding corro-systems, and components, including service water, sion has been based on reacting to it rather than on steam gererators, piping, and containment. Mitigation anticipatmg its occurrence. Regulatory and utility methods are briefly reviewed. He overview is refer-initiatives are creating a climate and framework for enced to a major data base that is being developed to -
more effective application of the concept. His report assist both utilities and regulators in the important and characterizes the framework and provides some illus-responsible task of understanding an_d managing corro-trations of how the concept is being applied in support sion and other degradation mechanisms in nuclear of the NRC NPAR.
plants. An effective application of understanding and
- Although corrosion has not caused a majoraccident managing these mechanisms is crucial not only to safe' in a nuclear power plant, it has been a continuing cause and economic operation of the nuclear plants, but also
. 67
i.
Main Citations and Summaries to public perception of a soundly designed, managed, years to make thermal aging predictions during experi-and operated technology.
mentally inaccessible times. Given the historical suc-cess of time-temperature superposition, the authors PNL-SA-20219, D. P. Brown and D. E Blahnik, have expanded this approach for combined radiation-
"ASME Subsection ISTD Recommendations Based up(m NPAR Snubber Aging Research Results, thermal environments yielding an empiricaltime-tem-Pacific Northwest Laboratory, December 1991.
perature-dose-rate shifting procedure..Ihe procedure As a result of information obtained through the Qes an Whnnahe for a given amount of ma.
tenal damage vems dow rate at a sched reference NRC's Nuclear Plant Aging Research (NPAR) Pro-temperatum.hs ts don @n&ng h Ms ac-gram and from snubber task research, recommenda-tivation energy that causes higher-temperature dose-tions were made in three areas for the next revision of rate data to supupow wMn sh to k refeme the American Society of Mechanical Engineers temperature.-The resultmg superposed curve at the (ASME) Operaticms and Maintenance (OM) Code, reference temperature extends to much lower dose Subsection ISTD (In-Scrvice Testing-D): (1) service-rates that are expenmentally tnaccessible because of life monitoring, (2) visual examination attributes, and the long time penods that would be required to simu-(3) failure grouping and corrective actions.
late aging.This procedure therefore allows meaningful A service-life monitoring program will be most ef-predictions to be made for long-term, low-dose-rate ra-fective if it distinguishes betw;en service-related and diation aging conditions. Using historical data from non-service-related failures. It is important that the Sandia's radiation-aging program on nuclear power mot cause of snubber failure or degradation (e.g., dy-plant cabic materials, the authors have successfully ap-namic transient, vibration, excessive temperature), be plied the time-temperature-dose-rate superposition identified along with the failure mode (e.g., high drag approach to four different materials: hypalon, neo-force, low activation)'and the failure mechanism (e.g.,
prene, polyethylene, and PVC jacket material. For two deformation of screw or ball shaft, solidification of of these materials, extrapolated predictions based on grease). This information provides the basis to take ef-the superimposed data were found to be in excellent fcctive countermeasures.
agreement with 12-year, low-dose-rate nuclear power it is suggested that failure evaluation data sheets plant results.
provide key information, including failure mode, MTLE 60103-X, J. E Gleason, R. A. DeFour, J. M.
failure mechanism, failure cause, environment, service llammond, and P. A. Lubeski, ' Test Plan for the time, abnormal conditions, test data, visual observa-Comprehensive Aging Assessment of Circuit Break-tions, and other test-related observations.They include ers and Relays for Nuclear Plant Aging Research trending and diagnostic tests and post-service as-found Program, Phase II," Wyle Laboratories,
. (NPAR)lle, AL, July 1989.
Huntsvi tests. Some of them might be included in ISTD. If in.
cluded, an appropriate statement should be written.
This entry refers to seven individually bound re-For effective failure cause and root cause determi-ports, each presenting the test plan for a specific type of '
i' nation, it is important t hat personnel involved in failure circuit breaker or relay:
l cvaluation have adequate experience. Failure evalu-60103-1 Molded-case circuit breakers i
ation data sheets should not be formatted in a manner
_60103-2 Metal-clad circuit breakers 12 that might lead the examiner toward a potentially in-60103-3 Auxiliary relays correct failure cause.
60103-4 Control relays Suggested service-life monitoring techniques can 60103-5 Protective relays be added to those in the ISTD, Section 7.0. They take 60103-6 Timing relays into consideration the capability of the various snubber 60103-7 Electronic relays models to endure the full range of plant environments The purpose of these reports is to provide details of (from benign to severe). Previously unrecognized se-the tese, planned for the types of circuit breakers and.
vere environments may often be identified by root relays under investigation in Phase 11 of the Compre.
I cause evaluation of failed or degraded snubbers. Infor-hensive Aging Assessment of Circuit Breakers and mation regarding the snubber endurance capability is Relays. This work is being performed by Wyle Labora-often obtamed from operating experience (i.e., from tories for the NRC NPAR program, which is intended i
failure data or by monitoring the degradation).
to resolve technical safety issues related to the aging -
g a n elch and mdanica safety systems, support systems, and c} compo SAND 88-0754 UC-78, K. T. Gillen and R. L. Clough, ml structures
" Time-Temperature-Dose Rate Superposition: A Methodology for Predicting Cable L)cgradation Un.
used in commercial nuclear power plants. The aging der Ambient Nuclear Power Plant Agmg Condi-period of interest includes the period of normal tions," Sandia National Laboratories, August 1988.
licensed plant operation as well as the period of ex-j Time-temperature superposition is an empirical ap-tended plant life that may be requested in utility appli-proach that has been used in polymers for more than 30 cations for license renewals.
NUREG-1377 68
l l
Main Citations and Summaries i
The Phase I report, NUREG/CR-4715, "An Aging
- 3. Tb evaluate the effectiveness of storage, main-Assessment of Relays and Circuit Ilreakers and System tenance, repair, and replacement practices in i
Interactions," showed that relays and circuit breakers mitigating the rate and extent of degradation-t are important nuclear plant components that are sus-caused by aging.
ceptible to degradation with time. Thus Phase II, a comprehensive aging assessment of relays and circuit Methods are available to detect and mitigate agmg breakers, was implemented to provide (1) a review and degradation and thereby to mmimtze its impact. The verification of improved inspection, surveillance, reports describe the background of the research monitoring, and maintenance methods; (2)in situ ex-strategy, list and elaborate on the objectives of the aminations and data gathering for operating equip _
research, and define the testing to be performed on ment; (3) postservice examinations of naturally aged naturally aged and degraded equipment in order to de-components or components with simulated degrada, termine the methods most effective for detecting age tion; (4) an evaluation of the role of maintenance in degradation. Emphasis has been placed on identifying mitigating the effects of aging;(5)cvaluations of meth, and characterizing the mechanisms of material and ods for predicting residuals and. sersice life; and component degradation during service and using re-(6) recommendations for using research results in the search results m the regulatory process, regulatory process. Specific goals of the program are:
The testing consists of performing and evaluating i
- 1. 'Ib identify and characterize aging effects that, if various methods of inspection, surveillance, condition unchecked, could cause degradation of compo-monitoring, and maintenance, including simulated nents and subsystems of circuit breakers and degradation, to aid in determining the usefulness of relays and thereby impair plant safety.
these methods for managing the effects of aging on safe
- 2. 'Ib identify methods of inspection, surveillance, plant operation.The devices chosen for testing are rep-and monitoring and to evaluate the residual life resentative of circuit breakers and relays that have of components and subsystems of circuit break-been in use in nuclear power plants. New, used, and ers and relays that will ensure timely detection aged specimens up to 40 years old have been located '
of significant aging effects before loss of their from a variety of sources, including Wyle stock and nu-safety function.
clear plants.
l i
l 1
69 NUREG-1377
PERSONAL AUTIIOR INDEX This index lists, in alphabetical order, all participating authors of each report listed in the main citation listing. 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 main citation by the report number.
Adams, M.L.
clear Power Plants " Vol.1," Operating Experience and Failure Identification."
NUREG/CR-4597, " Aging and Service Wear of Auxil-iary Feedwater Pumps for PWR Nuclear Power NUREG/CR-4992," Aging and Service Wear of Multi-Plants." Vol.1, " Operating Experience and Failure stage Switches Used in Safety Systems of Nuclear Identification."
Power Plants," Vol.1.
NUR EG/CR-5141, " Aging and Qualification Research Ahmed, S.
on Solenoid operated Valves."
NUREGICR-4257, " Inspection, Surveillance, and Monitoring of Electrical Equipment inside Contain.
Bader, B.E.
ment of Nuclear Power Plants-With Applications to NUREG/CP-0036, " Proceedings of the Workshop on Electrical Cables."
Nuclear Plant Aging.
Allen, R.P.
Beranek, A.E l
NUREG/CR-5491, "Shippingport Station Aging NUREG/CP-0100, " Proceedings of the International Evaluation.
Nuclear Power Plant Aging Symposium."
Anderson, R.L.
Berry, D.L.
NUREG/CR-4928 " Degradation of Nuclear Plant NUREG/CR-3818. " Report of Results of Nuclear Temperature Sensors "
Power Plant Aging Workshop "
Arendts, J.G.
Beverly, D.D.
NUREG/CR-4977, "SilAG Test Series: Seismic NUREG/CR-5560, " Aging of Nuclear Plant Resis-Research on an Aged United States Gate Valve and on tance Temperature Detectors."
a Piping System in the Decommissioned Heiss-dampfreaktor (HDR): Summary," Vol.1.
Bjorkelo, K.A.
NUREG/CR-4977, " SHAG Test Series: Seismic PNie7823. " Maintenance Practices'Ib Manage Aging:
Research on an Aged United States Gate Valve and on A Review of Several Technologies."
a Piping System in the Decommissioned Heiss-dampfreaktor (HDR): Appendices," Vol. 2.
Blahnik, D.E.
NUREG/CR-5643," Insights Gained from Aging Re-Atwood, C.L.
search."
EGG-SSRE-8972, " Estimating Hazard Functions for -
NUREG/CR-5870,"Results of LWR Snubber Aging Repa,rable Components."
Research."
i EGG-SSRE-9017,
- User's Guide to Pil A7.E a Corn-NUREG/CR-6043," Phase ! Aging Assessment of Es-puter Program for Ibrametric Ilazard Function Est'-
sential HVAC Chillers Used in Nuclear Power Plants."
mation."
PNL-5722," Operating Experience and. Aging Assess-NUREG/CR-5378 " Aging Data Analysisand Risk As-ment of ECCS Pump Room Coolers."
sessment-Development and Demonstration Study."
PNieSA-20219,"ASM E Subsection ISTD Recommen-d"II "5 I1 sed upon NPAR Snubber Aging Research Bacanskas, V.P.
Results.,
NUREG/CR-4715, "An Aging Assessment of Relays and Circuit Ilreakers and System Interactions."
Bohn, S.J.
NUREG/CR-4819, " Aging and Service Wear of Sole-NUREG/CR-5379, " Nuclear Plant Service Water Sys-noid-Operated Valves Used in Safety Systems of Nu-tem Aging Degradation Assessment," Vol. 2.
71 NUREG-1377 -
Personal AuthorIndex i
Borkowski, R.J.
ORNL/NRC/LTR-91/25," Throttled Valve Cavitation and Erosion."
NUREG/CR-2641, "The In-Plant Reliability Data Ilase for Nuclear Power Plant Components: Data Col-lection and Methodology Report."
Casada, M.L.
NUREG/CR-3543, " Survey of Operating Experiences NUREG/CR-3154, "The In-Plant Reliability Data rm enMy Ag ng nds Ilase for Nuclear Plant Components: Interim Report-The Valve Component."
g,
}
l Brown, D.P*
NUREG/CR-4597, " Aging and Service Wear of Auxil-iary Feedwater Pumps for PWR Nuclear Power NUREG/CR-5386, "Hasis for Snubber Aging Plants," Vol. 2, " Aging Assessments and Monitoring Research: Nuclear Plant Aging Research Program."
Method Evaluations."
NUREG/CR-5870, "Results of LWR Snubber Aging Research."
Chockie, A.I).
PNL-SA-20219,"ASME Subsection ITI'D Recommen-PNL-7823," Maintenance PracticesTo Manage Aging:
dations Ilased upon NPAR Snubber Aging Research A Review of Several Technologies.
Results."
Chn. tensen, J.A.
s Buckley, G.D.
NUREG/CP-0105, Paper by J. A. Christensen,"NPAR Approach to Controlling Aging in Nuclear Power NUREG/CR-6001, " Aging Assessment of IlW R Plants.,,
Standby Liquid Control Systems."
Cifuentes, E Burns, E.L.
HNLTbchnicalReport A-3270-11-85."SeismicEndur-NUREG/CR-4156. " Operating Experience and ance Tests of Naturally Aged Small Electric Motors."
Aging-Seismic Assessment of Electric Motors."
Clark, N.lt Bush, S.H.
NUREG/CR-3818, " Report of Results of Nuclear NUREG/CR-4279, " Aging and Service Wear of fly-Power Plant Aging Workshop."
draulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Clauss, D.B.
Plants," Vol.1.
NUREG/CR-5334," Severe Accident Testing of Elec-trical Penetration Assemblies."
NUREG/CR-425'/, " Inspection, Surveillance, and Cletcher, J.W.
Monitoring of Dectrical Equipment Inside Contain-ment of Nudcar Power Plants-With Applications to NUREG/CR-4692, " Operating Experience Review of Electrical Cables."
Failures of Power Operated Relief Valves and Illock Valves in Nuclear Power Plants."
NUREG/CR-5141," Aging and Qualification Research on Solenoid Operated Valves."
Clinton, J.
NUREG/CR-4985, " Indian Point 2 Reactor Coolant Casada, D.A.
Pump Seal Evaluations."
NUREG/CR-5404, " Auxiliary Feedwater System Ag-ing Study," Vol.1.
Clough, R.L.
NUREG/CR-5643, " Insights Gainu! from Aging Re.
SANDS 8-0754 UC-78, "TimeIIbmperature-1)ose t
search."
Rate Superposition: A Methodology for Predicting Ca-ble Degradation Under Ambient Nuclear Power Plant NUREG/CR-5706. "NRC llulletin 88-04: Potential Aging Conditions."
Safety-Related Pump Loss-An Assessment of Indus-try Data."
Collins, E.P.
NUREG/CR-5944. "A Characterization of Check NUREG/CR-5248, "Prioritization of TIRGALEX-Valve Degradation and Failure Experience in the Nu-Recommended Components for Further Aging Re.
clear Power Industry "
search."
1 NUREG-1377 72
Personal AuthorInder Cornwdl, B.C.
NUREG/CR-5406, "BWR Reactor Water Cleanup System Flexible Wedge Gate ! solation Valve Qualifica-NUREG/CR-3819," Survey of Aged Power Plant Fa-tion and High Energy Flow Interruption Tbst," Vol. 2 eilities."
" Data Report."
Corugedo, J.J.
NUREG/CR-54%, "BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation Valve Qualifica-NUREG/CR-5515, " Light Water Reactor Pressure tion and High Energy Flow Interruption Test,, Vol. 3, Isolation Valve Performance Testing."
" Review ofIssues Associated with BWR Containment Isolation Valve Closure."
Crowley, J.L.
NUREG/CR-5558, " Generic issue 87: Flexible Wedge NUREG/CR-4380, " Evaluation of the Motor-Gate Wlve Tbst Program: Phase II Results and Analy-Operated Wlve Analysis and Test System (MOVATS) lb Detect Degradation, incorrect Adjustments, and Other Abnormalities in Motor-Operated Valves."
NUREG/CR-5720, " Motor-Operated Valve Research Update."
Curreri, J.
Dingee, D.A.
BNLTechnicalReport A-3270-Il-85,"SeismicEndur-ance Tests of Naturally Aged Small Electric Motors."
NUREG/CR~1590, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating and Expert Expe-rience," Vols. I and 2.
Czgikowski, C.J.
NUREG/CR-4985, " Indian Point 2 Reactor Coolant 0 " SCI' M.R.
8 Pump Seal Evaluations."
NUREG/CR-3956, "In Situ Testing of the Ship-pingp rt Atomic Power Station Electrical Circuits."
Damiano, B.
NUREG/CR-5479, " Current Applications of Vibra-Dodge, R.E.
tion Monitoring and Neutron Noise Analysis: Detec-NUREG/CR-4279, " Aging and Service Wear of Hy-tion and Analysis of Structural Degradation of Reactor draulic and Mechanical Snubbers Used on Safety-Vessel Internals from Operational Aging.
Related Piping and Components of Nuclear Power Plants," Vol.1.
' Davis, T.
NUREG/CR-4144, "importance Ranking Based on Donaldson, M.R.
Aging Consideration of Components Included in NUREG/CR-3956, "In Situ Testing of the Ship-Probabilistic Risk Assessments."
pingport Atomic Power Station Electrical Circuits" DeFour, R.A.
Drago, J.R WYLE 60103-X,
- Test Plan for the Comprehensive NUREG/CR-2641, "The In-Plant Reliability Data i
Aging Assessment of Circuit Breakers and Relays for Base for Nuclear Power Plant Components: Data Col-Nuclear Plant Aging Research (NPAR) Program, lection and Methodology Report."
l Phase II."
Dukelow, J.S.
DeWall, K.G.
NUREG/CR-5848, "Recordkeeping Needs 1b Miti-EGG-SSRE-9777,"IsolationValve Assessment (IVA) gate the Impact of Aging Degradation."
'i Software Version 3.10, User's Manual."
Edson, J.L.
EGG-SSRE-9926," Evaluation of EPRI Draft Report NUREG/CR-4457, " Aging of Class IE Battenes m NP-7065-Review of NRC/INEL Gate Valvelbst Pro-Safety Systems of Nuclear Power Plants.,,
gram."
NUREG/CR-5181, " Nuclear Plant Aging Research:
NUREG/CR-3819," Survey of Aged Power Plant Fa-The IE Power System.
I g;g,.
NUREG/CR-5448." Aging Evaluation of Class IE Bat-NUREG/CR-5406 "BWR Reactor Water Cleanup teries: Seismic Testing "
System Flexible Wedge Gate Isolation Valve Qualifica-
. tion and High Energy Flow Interruption Test," Vol.1, NUREG/CR-5643," Insights Gained from Aging Re-search,"
-l
" Analysis and Conclusion."
73-NUREG-1377 f
Personal AuthorIndex Eissenberg, D.M.
NUREG/CR-5268, " Aging Study of Boiling Water NUREG/CR-4234, " Aging and Service Wear of Reactor Residual Heat Removal System."
Electric Motor-Operated Valves Used in Engineered NUREG/CR.5419 " Aging Assessment ofInstrument Safety-Feature Systems of Nuclear Power Plants," Vol.
Air Systems in Nuclear Power Plants."
L NUREG/CR-4302," Aging and Service Wear of Check aHab, R.R Valves Used in Engineered Safety-Feature Systems of NUREG/CR-3543, " Survey of Operating Experiences Nucicar Power Plants," Vol.1.
from LERslb Identify Aging 1 rends" NUREG/CR-4380, " Evaluation of the Motor-Oper-NUREG/CR-4302, " Aging and Sen' ice Wear of Check
]
ated Valve Analysis and Test System (MOVATS)1b V Ives Used in Engineered Safety-Feature Systems of
?
Detect Degradation, Incorrect Adjustments, and clear Nwn Mants," Vol.1.
Other Atmormalities in Motor-Operated Valves."
Gehl, A.L.
Enderlin, W.I*
NUREG/CR-5700," Aging Assessment of Reactor In-strumentation and Protection Systems Components.
PNL-7823. " Maintenance Practices 1b Manage Aging:
A Review of Several Technologies."
Gillen, K.T.
SAND 88-0754 UC-78, ' Time-Temperature-Dose Fa.in, R.E.
Rate Superposition: A Methodology for Predicting Ca-
.i NUREG/CR-5383, "Effect of Aging on Response ble Degradation Under Ambient Nuclear Power Plant i
lime of Nuclear Plant Pressere Sensors "
Aging Conditions "
Fineman, D.L.
Gingrich, JJ.
i NUREG/CR-5643," Insights Gained from Aging Re-NUREG/CR-5383, "Effect of Aging.on Response Time of Nuclear Plant Pressure Sensors.
Gleason, J.E Fleming, T.E.
NUREG/CR-5762, " Comprehensive Aging Assess-T PNL-7823," Maintenance Practices 10 Manage Aging:
ment of Circuit Breakers an i Helays."
A Review of Several Technologies."
WYLE 60103-X, " Test Plan for the Comprehensive Fragola, J.R.
Aging Assessment of Circuit Breakers and Relays for l
Nuclear Plant Aging Research (NPAR). Program, NUREG/CR-3154, "The In-Plant Reliability Data Phase II."
Base for Nuclear Plant Components: Interim Report-i The Valve Companent."
Goldberg, EE NUREG/CR-2641, "T11e In-Plant Reliability Data Fresco, A.
Base for Nuclear Power Plant Components: Data Col-HNL'Ibchnical Report A-3270R-2-90, " Aging Effects Icction and Methodology Report."
of Important Halance of Plant Systems in Nuclear Power Plants."
Goodman, R.L.
PNL-5722 " Operating Experience and Aging Assess-Fuehrer, G.E ment f ECCS Pump Room Coolers."
NUREG/CR-5655, " Submergence and High Tempera-Gore, M.L.
ture Steam Testing of Class 1E Electrical Cables."
t NUREG/CR-5379, " Nuclear Plant Service Water Sys-tem Aging Degradation Assessment: Phase I," Vol.1.
Fullwood, R.
l HNLTechnical Report A-3270-12-86," Aging and Life NUREG/CR-5379, " Nuclear Plant Service Water Sys-l Extension Assessment Program (ALEAP) Systems tem Aging Degradation Assessment," Vol. 2.
Level Plan."
Greene, R.H.
NUREG/CR-5052," Operating Experience and Aging NUREG/CR-5699, " Aging and Service Wear of Con-i Assessment of Component Cooling Water Systems in trol Rod Drive Mechanisms for BWR Nuclear Plants,"
i Pressurized Water Reactors."
Vol.1.
74 r
i Personal AuthorIndex
-l Greenstreet, W.L.
NUREG/CR-5693," Aging Assessment of Component '
NUREG/CR-4234, " Aging and Service Wear of Cooling Water Systems in Pressurized Water Reac-tors-Phase IL,
Electric Motor-Operated Valves Used in Engineered Safety-Feature Systems of Nuclear Power Plants,"
NUREG/CR-5783, " Aging Assessment of the Com-Vol.1.
bustion Engineering and Babcock & Wilcox Control NUREG/CR-4302," Aging and Service Wear of Check Valves Used in Engineered Safety-Feature Systems of Hagen, E.W.
Nuclear Power Plants," Vol.1.
NUREG/CR-5700, " Aging Assessment of Reactor In-strumentation and Protection Systems Components."
Grove, E.
UNL Technical Report A-3270 6-21-91, " Degradation llammond, J.M..
Modeling: Extensions and Applications."
WYLE 60103-X " Test Plan for the Comprehensive Aging Assessment of Circuit Breakers and Relays for BNL lbchnical Report TR-3270-9-90, "An Opera-Nuclear Plant Aging Research (NPAR) Program, tional Assessmen'. of the Habcock & Wilcox and Com-Phase IL" bustion Engineering Control Rod Drives "
Hanchey, L.A.
NUREG/CR-5783, " Aging Assessment of the Com-bustion Engineering and Habcock & Wilcox Control NUREG/CP-0036, " Proceedings of the Workshop on i
Rod Drives.
Nuclear Plant Aging "
Hardin, J.E.
Gunther, W.E.
NUREG/CR-4457, " Aging of Class 1E Hatteries in UNL Tbchnical Report TR-3270-6-90, " Maintenance Safety Systems of Nuclear Power Plants."
l Team Inspection Results: Insights Related to Plant Aging."
Hashemian, H.M.
NUREG/CR-4928. " Degradation of Nuclear Plant BNL Techm. cal Report TR-3270-9-90, "An Opera-Temperature Sensors."
tional Assessment of the Habcock & Wilcox and Com-bustion Engineering Control Rod Drives "
. NUREGICR-5383, "Effect of Aging on Response i
Time of Nucicar Plant Pressure Sensors."
NUREG/CR-4564, " Operating Experience and NUREG/CR-5560, " Aging of Nuclear Plant Resis-Agmg-Seistmc Assessment of Battery Chargers and tance Temperature Detectors."
Inverters.
NUREG/CR-4939, " Improving Motor Reliability in Haynes, H.D.
Nuclear Power Plants"; Volume L " Performance NUREG/CR-4234, " Aging and Service Wear of Elec-Evaluation and Maintenance Practices"; Volume 2, tric Motor-Operated Valves Used in Engineered
" Functional Indicator Tests on a Small Electric Motor Safety-Feature Systems of Nuclear Power Plants: Ag-Subjected to Accelerated Aging"; Volume 3 " Failure ing Assessments and Monitoring Method Evalu-Analysis and Diagnostic Tests on a Naturally Aged ations," Vol 2.
Electric Motor "
NUREG/CR-4302 " Aging and Service Wear of Check NUREG/CR-5051, " Detecting and Mitigating Hattcry Valves Used in Engineered Safety-Feature Systems of
.j Charger and Inverter Aging "
Nuc! car Power Plants,"Vol. 2 " Aging Assessments and -
Monitoring Method Evaluations."
NUREG/CR-5192, "7bsting of a Naturally Aged NUREG/CR-5643," Insights Gained from Aging Re-Nucicar Power Plant Inverter and Battery Charger.
scarch."
NUREG/CR-5268, " Aging Study of Boiling Water Hensler, Es.
Reactor Residual 11 cat Removat System."
NUREG/CR-4279, " Aging and Service Wear of Ily-NUREG/CR-5507, "Results from the Nuclear Plant draulic and Mechanical Snubbers Used on Safety-
)
Aging Research Program:1 heir Use in Inspection Ac-Related Piping and Components of Nuclear Power tivities."
Plants," Vol 1.
NUREG/CR-5555, " Aging Assessment of the West-Hebble, T.L.
inghouse PWR Control Rod Drive System."
NUREG/CR-3154, "The In-Plant Reliability Data NUREG/CR-5643, " Insights Gained from Aging Base for Nuclear Plant Components: Interim Report-Research."
The Valve Component."
75 NUREG-1377
Personal AuthorIndex i
Ifiggins, J.C.
IIunt, T.II.
UNLTechnical Report A-3270-Il-85," Seismic Endur.
EGG-SSRE-10039 "An Evaluation of the Effects of ance Tests of Naturally Aged Small Electric Motors."
Valve Body Erosion on Motor-Operated Valve Oper-BNLibchnical Report A-3270-12-86," Aging and Life Extension Assessment Program (ALEAP) Systems Jacobus, M.J.
i Level Plan."
4 NUREG/CR-5461, " Aging of Cables, Connections, i
NUREG/CR-5052," Operating Experience and Aging and Electrical Penetration Assemblies Used in Nuclear Assessment of Component Cooling Water Systems in Power Plants "
Pressurized Water Reactors."
NUREG/CR-SM3," Insights Gained from Aging Re-search."
IIolbert, K.E.
NUREG/CR-5655," Submergence and High lbmpera-NUREG/CR-4928, " Degradation of Nuclear Plant ture Steam Testing of Class 1E Electrical Cables."
]
Temperature Sensors."
NUREG/CR-5772, " Aging, Condition. Monitoring, and Loss-of-Coolant Accident (LOCA) Tests of Class Iloopingarner, K.R.
IE Electrical Cables," Vol.1.
" Safety Implications of Diesel Generator Aging "Nu-clear Safety, December 1990.
NUREG/CR-5772, " Aging, Condition Monitoring, and Loss-of-Coolant Accident (LOCA) Tests of Class NUREG/CR-4590, " Aging of Nuclear Station Diesel 1E Electrical Cables," Vol. 2.
Generators: Evaluation of Operating and Expert Expe-NUREG/CR-5772, " Aging, Condition Monitoring, rience." Vols. I and 2, and Loss-of-Coolant Accident (LOCA)1bsts of Class IE Electrical Cables," Vol. 3.
NUREG/CR-5057, " Aging Mitigation and Improved Programs for Nuclear Service Diesel Generators "
Jarrell, D.B.
NUREG/CR-5643, " Insights Gained from Aging Re-NUREG/CR-5248, "Prioritization of TIRGALEX-search." '
Recommended Components for Ftirther Aging Re.'
search."
PNL-6287, " Study Group Review of Nuclear Service Diesel Generator Testing and Aging Mitigation."
NUREG/CR-5379, " Nuclear Plant Service Water Sys-tem Aging Degradation Assessment: Phase I," Vol.1.
PNL-7516, " Emergency Diesel Generator Techm. cal Specifications Study Results."
NUREG/CR-5379, " Nuclear Plant Service Water Sys-tem Aging Degradation Assessment," Vol. 2.
Ilorst, C.L.
NUREG/CR-5643," Insights Gained from Aging Re-NUREG/CR-5159, " Prediction of Check Valve Per.-
scarch."
formance and Degradation in Nuclear Power Plant Sys-PNL-SA-18407, " Understanding and Managing Cor-t ems."
rosion in Nuclear Power Plants."
NUREG/CR-5583, " Prediction of Check Valve Per-formance and Degradation in Nuclear Power Plant Sys.
Jaske, C.E.
tems-Wear and Impact 1bsts."
NUREG/CR-5314," Life Assessment Procedures for Major LWR Components," Vol. 3, " Cast Stainless lioy,II.C.
St el Components."
NUREG/CR-3543," Survey of Operating Experiences Jeanmougin, N.M.
from LERs1bidentify Aging Trends."
NUREG/CR-5515, " Light Water Reactor Pressure Isolation Valve Performance Testing."
l BNL Technical Report A-3270 6-21-.91," Degradation Johnson, A.B.
Modeling: Extensions and Applications."
NUREG/CR-5379," Nuclear Plant Service Water Sys-tem Aging Ngradadon hessment Mase U Vol L NUREG/CR-5612. " Degradation Modeling with Ap-plications to Aging and Maintenance Effectiveness NUREG/CR-5491, "Shippingport Station Aging Evaluation."
Evaluation."
NUREG-1377 76
Personal AuthorIndex i
NUREG/CR-6001, " Aging Assessment of UWR Nuclear Power Plants," Vol. 2," Evaluation of Monitor-Standby Liquid Control Systems."
ing Methods."
i PNL-SA-18407, " Understanding and Managing Cor-NUREG/CR-5479, " Current Applications of Vibra-rosion in Nuclear Power Plants."
tion Monitoring and Neutron Noise Analysis: Detec-tion and Analysis of Structural Degradation of Reactor Johnson, J.W.
Vessel Internals from Operational Aging."
NUREG/CR-3154, 'The In-Plant Reliability Data NUREG/CR-5643," Insights Gained from Aging Re-Base for Nuclear Plant Components: Intcrim Report-search."
The Valve Component."
Kueck, J.D.
Kahl, W.K.
NUREG/CR-5404, " Auxiliary Feedwater System Ag-NUREG/CR-3154, "The In-Plant Reliability Data ing Phase I Follow-on Study," Vol. 2.
Ilase for Nuclear Plant Components: Interim Report-The Valve Component."
Kurth, R.
NUREG/CR-4144, "Importance Ranking Based on Kalsi, M.S.
Aging Consideration of Components Included in NUREG/CR-5159, " Prediction of Check Valve Probabilistic Risk Assessments."
Performance and Degradation in Nuclear Power Plant NUREG/CR-5510. " Evaluations of Core Melt Fre-Systems.
quency Effects Due to Component Aging and NUREG/CR-5583, " Prediction of Check Valve Maintenance."
Performance and Degradation in Nuclear Power Plant Systems-Wear and Impact Tests."
Larson, L.L.
NUREG/CR-5379, " Nuclear Plant Service Water NUREG/CR-5807, *1mprovements in Motor Oper.
l ated Gate Valve Design and Prediction Models for System Aging Degradation Assessment," Vol. 2.
Nuclear Power Plant Systems."
NUREG/CR-6001," Aging Assesment of BWR Stand-by Liquid Control Systems."
NUREG/CR-4928 " Degradation of Nuclear Plant Lee, B.S.
'Ibmperature Sensors."
NUREG/CR-5280, " Age-Related Degradation of Westinghouse 480-Volt Circuit Breakers," Vol. 2,"Mc-Kirkwood, B.J.
chanical Cycling of a DS-416 Breaker. Test Results."
PNL-6287, " Study Group Review of Nuclear Service NUREG/CR-5693," Aging Assessment of Component Diesel Generator Testing and Aging Mitigation."
Cooling Water Systems in Pressurized Water Reac-tors-Phase II."
Kitch, D.M.
NUREG/CR-4597," Aging and Service Wear of Auxil-Leverenz, E iary Feedwater Pumps for. PWR Nuclear Power NUREG/CR-4144, "importance Ranking Hased on Plants," Vol. 2, " Aging Assessments and Monitoring Aging Consideration of Components Included in Method Evaluations."
Probabilistic Risk Assessments."
i Klein, R.E Levy, I.S.
NUREG/CR-6N3," Phase I. Aging Assessment of Es-NUREG/CR-5248, "Prioritization of TIRGALEX-t sential HVAC Chillers Used in Nuclear Power Plants."
Recommended Components for Further Aging.
Research."
Kochis, S.
NUREG/CR-5280, " Age-Related Degradation of-Le%is, R.
Westinghouse 480-Volt Circuit Hrcakers," Vol. 2, "Me-NUREG/CR-5051, " Detecting and Mitigating Hattery chanical Cycling of a DS-416 Breaker. Test Results."
Charger and Inverter Aging "
Kryter, R.C.
Lodlow, C.C.
NUREG/CR-4819, " Aging and Service Wear of NUREG/CR-4715, "An Aging Assessment of Relays Solenoid-Operated Valves Used in Safety Systems of and Circuit Breakers and System Interactions."
77 NUREG-1377
Personal AuthorIndex Lofaro, R.
Magelby, ILL.
NUREG/CR-4939, " Improving Motor Reliability in NUREG/CR-5643, <' Insights Gained from Aging Re-Nuclear Power Plants": Volume 1, " Performance search."
Evaluation and Maintenance Practices"; Volume 2,
" Functional Indicator Tests on a Small Electric Motor Makay, E.
Subjected to Accelerated Aging"; Volume 3," Failure NUREG/CR-4597, " Aging and Service Wear of Auxil-Analysis and Diagnostic Tists on a Naturally Aged iary Feedwater. Pumps for PWR Nuclear Power Electric Motor."
Plants." Vol.1, " Operating Experience and Failure cadon?
NUREG/CR-5052," Operating Experience and Aging Assessment of Component Cooling Water Systems in M
14' ED*
Pressurized Water Reactors.
NIS11R 4485, " Annotated Bibliography: Diagnostic NUREG/CR-5268, " Aging Study of Iloiting Water Methods and Measurement Approaches'Ib Detect in-Reactor Residual licat Removal System."
cipient Defects Due to Aging of Cables."
NUREG/CR-5693," Aging Assessment of Component NIS'11R 4487, " Detection of Incipient Defects in Ca-Cooling Water Systems in Pressurized Water Reac.
bles by Partial Discharge Signal Analysis."
tors-Phase II."
Meale, ILM.
Lonzecky, IU.
NUREG/CR-4747,"An Aging Failure Survey of Light Water Reactor Safety Systems and Components,"
PNL-6287, " Study Group Review of Nuclear Senice Vols. I and 2.
Diesel Generator Testing and Aging Mitigation."
Meininger, R.D.
Lubeski, P.A.
NUREG/CR-5008, " Development of a Testing and WYLE 60103-X," Test Plan for Comprehensive Aging
^"#WS
"'"EY E""
Assessment of Circuit Breakers and Relays for Nuclear Plant Aging Research (NPAR) Program, Phase II."
W yer,L.C.
NUREG/CR-4740, " Nuclear Plant-Aging Research Luk, K.li.
on Reactor Protection Systems."
NUREG/CR-5754, "Hoiling-Water Reactor Intern NUREG/CR-4%7." Nuclear Plant Aging Researchon -
Aging Degradation Study, Phase 1."
Iligh Pressure injection Systems."
NU REG /CR-6048, " Pressurized-Water Reactor inter-NUREG/CR-5181, " Nuclear Plant Aging Researce j
nals Aging Degradation Study, Phase 1."
The IE Power System."
l MacDonald, P.E.
MiHer,B.
^~
~
"E NUREG/CR-4731," Residual Life Assessment of Ma-t " Containment Purge and Wnt Vake Seal jor Light Water Reactor Components," Vol.1.
te i l
NUREG/CR-4731," Residual Life Assessment of Ma-jor Light Water Reactor Components-Oveniew,"
Mitchell, D.W Vol. 2-l NUREG/CR-5560, " Aging of Nuclear Plant Resis -
i tance Temperature Detectors."
\\
MacDougall, E.
Mopsik, EI.
NUkEG/CR-5280, " Age-Related Degradation of NISTIR 4787, "The Use of Time-Domain Diclectric i
Westinghouse 480-Volt Circuit Dreakers," Vol.1. "Ag-Spectroscopy lo Evaluate the Lifetime of Nuclear mg Assessment and Recommendations for improvmg Power Station Cables."
Hreaker Reliability."
l l
NUREG/CR-5280, " Age-Related Degradation of Morris, ILM.
l Westinghouse 480-Volt Circuit Breakers " Vol. 2,"Me-NUREG-1144, " Nuclear Plant Aging Research chanical Cycling of a DS-416 Breaker. Test Results."
(NPAR) Program Plan,"
)
)
78
=
Personal AuthorIndex Moyers, J.C.
Nowlen, S.P.
NUREG/CR-5519, " Aging of Control and Service Air NUREG/CR-5546,"An Investigation of the Effects of Compressors and Dryers Used in Nuclear Power
'Ihermal Aging en the Fire Damageability of Electric Plants," Vol.1.
Cables."
NUREG/CR-5779, " Aging of Non-Power-Cycle IIcat NUREG/CR-5619,"The Impact of Thermal Aging on Exchangers Used in Nuclear Power Plants," Vol.1.
the Flammability of Electric Cables."
n, R.D.
Murphy, G.A.
NUREG/CR-600), " Aging Assessment of IlWR NUREG/CR-3543, " Survey of Operating Experiences Standby Liquid Control Systems."
from LERs Tb Identify Aging Trends."
NUREG/CR-4234, " Aging and Service Wear of Elec-Paltner, G.R.
tric Motor Operated Valves Used in Engineered NUREG/CR-5386, "llasis for Snubber Aging Re-Safety-Feature Systems of Nuclear Power Plants,"
search: Nuclear Plant Aging Research Program."
Vol.1.
Perrey, A. G.
NUREG/CR-4302, " Aging and Semcc Wear of Check Valves Used in Engineered Safety-Feature Systems of NISTIR 4485, " Annotated Hibliography: Diagnostic Nuclear Power Plants " Vol.1.
yethods and Measurement ApproachesTo Detect In-cipient Defects Due to Agmg of Cables."
NUREG/CR-4692, " Operating Experience Review of Failures of Power Operated Relief Valves and Illock Petersen, K.M.
Valves in Nuclear Power Plants."
NUREG/CR-4928, " Degradation of Nuclear Plant NUREG/CR-5643," Insights Gained from Aging Re-Temperature Sensors "
search."
NUREG/CR-5383, "Effect of Aging on Response Time of Nuclear Plant Pressure Sensors."
Naus, D.J.
Pike, D.II*
NUREG/CR-4652. " Concrete Component Aging and Its Significance Relative to Life Extension of Nuclear NUREG/CR-2641, "The In-Plant Reliabil,ty Data i
Ibwer Plants."
llase for Nuclear Power Plant Components: Data Col-tection and Methodology Report."
Neely, II.H.
Reich, M.
NUREG/CR-5515. " Light Water Reactor Pressure llNLTechnical Report A-3270-Il-85,"Scismic Endur-Isolation Valve Performance Testing "
ance Tests of Naturally Aged Small Electric Motors "
Nehring, T.
IINL Technical Report A-3270-Il-85, "Seiam?c Rib, L.N.
Endurance Tests of Naturally Aged Small Electric Letter Report," Summaries of Research Reports Sub-J M otors."
mitted in Connection with the Nuclear Plant Aging Research (NPAR) Program."
Nesbitt, J.F.
Roberts, G.C.
NUREG/CR-4590, " Aging of Nuclear Station Diesel NUREG/CR-4819, " Aging and Service Wear of Generators: Evaluati(m of Operating and Expert E,xpe-Solenoid-Operated Valves Used in Safety Systems of rience, Vols.1 and Nuclear Power Plants," Vol.1, " Operating Experience and Failure Identification."
Nitzel, M.E.
NUR EG/CR-4992. " Aging and Service Wear of Multi-EGG-SSRE-10039, "An Evaluation of the Effects of stage Switches Used in Safety Systems of Nuclear Valve ikxty Erosion on Motor-Operated Valve Oper-Power Plants " Vol.1.
ability."
NUREG/CR-5646, " Piping System Response During Roesener, W.S.
High Level Simulated Seismic Tests at the NUREG/CR-5378," Aging Data Analysis and Risk As-licissdampfreaktor Facility (SilAM Test Series)."
sessment-Development and Demonstration Study."
79 NUREG-1377
Personal Author Index Rose, J.A.
Sharma, V.
NUREG/CR-3819," Survey of Aged Power Plant Fa.
NUREG/CR-5583, " Prediction of Check Valve Per-cilities."
formance and Degradation in Nuclear Power Plant Sys-tems-Wear and Impact Tests."
Russell, M.J.
Sheets, M.W.
NUREG/CR-5720," Motor-Operated Valve Research BNLTechnical Report A-3270-3-86,"7bsting Program Update."
for the Monitoring of Degradation in a Continuous Duty 460-Volt Class "B," 10-HP Electric Motor."
Samanta, RK.
NUREG/CR-4939, " Improving Motor Reliability in BNL Technical Report A-3270 6-21-91, " Degradation Nuclear Power Plants": Volume 1, " Performance Modeling: Extensions and Applications."
Evaluation and Maintenance Practices"; Volume 2,
" Functional Indicator Tests on a Small Electric Motor NUREG/CR-5612," Degradation Modeling with Ap-SuNectcd to Accelerated Aging _; Volume 3, Failure plications to Aging and Maintenance Effectiveness An lysis and Diagnostic Tests on a Naturally Aged Evaluation."
Electric Motor,,,
Satterwhite, D.G.
Shier, W.
NUREG/CR-4747,"An Aging Failure Survey of Light NUREG/CR-5053," Operating Experience and Aging Water Reactor Safety Systems and Components," Vols.
Assessment of Motor Control Centers" 1 and 2.
NUREG/CR-5268, " Aging Study of Boiling Water Re-actor Residual Heat Removal System."
Scalzo, S.M.
NUREG/CR-5280, " Age-Related Degradation of NUREG/CR-5510. " Evaluations of Core Melt Fre-Westinghouse 480-Volt Circuit Breakers," Vol.1,"Ag-quency Effects Due to Component Aging and ing Assessment and Recommendations for Improving Maintenance."
Breaker Reliability."
Schlonski, J.S.
Shook, T.A.
NUREG/CR-4715, "An Aging Assessment of Relays NUREG/CR-4992," Aging and Service Wear of Multi-and Circuit Breakers and System Interactions.,,
stage Switches Used in Safety Systems of Nuclear Power Plants," Vol.1.
gjgye p, y, y,
BNL Technical Report A-3270-12-85, " Pilot Assess.
Scott, W.B-ment: Impact of Aging on the Seismic Performance of PNL-7823," Maintenance PracticesTb Manage Aging:
Selected Equipment lypes "
A Review of Several Tbchnologies."
Shafaghi, A.
NISTIR 4487, "Deection of Incipient Defects in Ca-0#
"* ^""
NUREG/CR-4144, "importance Ranking Based on Aging Consideration of Components Included in Sinha, U.R Probabilistic Risk Assessments."
. PNL-SA-18407. " Understanding and Managing Cor-rosion in Nuclear Power Plants."
NUREG/CR-4731," Residual Life Assessment of Ma.
Skreiner, K.M.
jor lj ht Water Reactor Components," Vol.1.
NUREG/CR-4939, " Improving Motor Reliability in j
g Nuclear Power Plants": Volume 1, " Performance NUREG/CR-4731,"ResidualLife Assessment of Ma-Evaluation and Maintenance Practices"; Volume 2, jor Light Water Reactor Components-Overview,"
" Functional Indicator Tests on a Small Electric Motor Vol. 2.
Subjected to Accelerated Aging"; Volume 3," Failure NUREG/CR-5314, " Life Assessment Procedures for Analysis and Diagnostic Tests on a Naturally Aged Electric Motor."
l Major LWR Components," Vol. 3, " Cast Stainless Steel Components."
Soberano, ET.
PNL-SA-18407, " Understanding and Managing Cor-NUREG/CR-3956, "Iri Situ fresting of the Ship-rosion in Nuclear Power Plants."
pingport Atomic Power Station Electrical Circuits."
80.
Personal Author index j
-l Sowatskey, RJ.
Stratton, R.C.
NUREG/CR-4517," Aging and Service Wear of Auxil.
NUREG/CR-5379, " Nuclear Plant Service Water Sys-iary Feedwater Pumps for PWR Nuclear Power tem Aging Degradation Assessment," Vol. 2, Plants," Vol. 2, " Aging Assessments and Monitoring Method Evaluations."
Subudhi, M.
Letter Report, " Review of Aging-Seismic Correlation Studi s n Nuclear Plant Equipment."
Steele, R.
H ednical Report A-3270-ll-85," Seismic Endur-EGG-SSRE-9777," Isolation Valve Assessment (IVA) ance Tests of Naturally Aged Small Electric Motors."
Software Version 3.10, User's Manual."
HNL Technical Report A-3270-12-85, " Pilot Assess-EGG-SSRE-9926," Evaluation of EPRI Draft Report ment: Impact of Aging on the Scismic Performance of NP-7065-Review of NRC/INEL Gate Valve Test Selected Equipment 'Ippes."
Program.
1 HNLTechnical Report A-3270-3-86,"TestingProgram j
NUREG/CR-3819 "Suney of Aged Power Plant for the Monitoring of Degradation in a Continuous l
Facilitics."
Duty 460-Volt Class "H,"10-11P Electric Motor."
HNLTechnical Report A-3270-12-86," Aging and Life NUREG/CR-4977, "SiIAG Test Series: Scismic Extension Assessment Pmgram (ALEAP) Systems i
Research on an Aged United States Gate Valve and on a
Piping System in the Decommissioned l
licissdampfreaktor (IIDR): Summary," Vol.1.
HNL Technical Report A-3270R-2-90, " Aging Effects of Important Halance of Plant Systems in Nuclear NUREG/CR-4977, " SHAG Test Series: Scismic Re-Power Plants."
search on an Aged United States Gate Valve and on a HNLTechnical Report A-3270 6-21-91," Degradation Piping System in the Decommissioned Heiss, J
Modeling: Extensmns and Applications."
dampfreaktor (HDR): Appendices," Vol. 2.
NUREG/CR-4156, " Operating Experience and NUREG/CR-5406, "HWR Reactor Water Cleanup Aging-Seismic Assessment of Electric Motors."
System Flexible Wedge Gate Isolation Valve Qualifica.
NUREG/CR-4564, " Operating Experience and tion and High Energy Flow Interruption Test," Vol.1.
" Analysis and Conclusion."
Aging-Scismic Assessment of Hattery Chargers and Inverters."
NUREG/CR-5406, "HWR Reactor Water Cleanup NUREG/CR-4939, " Improving Motor Reliability in System Flexible Wedge Gate Isolation Valve Qualifica-Nuclear Power Plants" Volume 1, " Performance tion and Iligh Energy Flow Interruption Test," Vol. 2, Evaluation and Maintenance Practices";~ Volume 2.
" Data Report."
" Functional Indicator Tests on a Small Electric Motor Subjected to Accelerated Aging"; Volume 3, " Failure NUREG/CR-5406, "HWR Reactor Water Cleanup An lysis and Diagnostic Tests on a Naturally Aged System Flexible Wedge Gate bolation Valve Qualifica, Electric Motor.
tion and High Energy Flow Interruption Test," Vol. 3,
" Review of Issues Associated with HWR Containment NUREG/CR-4985, " Indian Point 2 Reactor Coolant Isolmion Valve Closure."
Pump Seal Evaluations?
JURECR-5558," Generic Issue 87: Flexible Wedge NUREGICR-5051," Detecting and Mitigating Hattery Gt 4alve "Ibst Program: Phase II Results and Charger and Inverter Aging."
aalysis?
NUREG/CR-5052," Operating Experience and Aging Assessment of Component Cooling Water Systems in i
N UP EG/CR-5646, "Pipm.g System Response During Pressurized Water Reactors "
High Level Simulated Scismic Tests at the Heissdampfreaktor Facility (SilAM Test Series)."
NUREG/CR-5053," Operating Experience and Aging Assessment of Motor Control Centers "
EG/CR-5720. " Motor-Operated Valve Research NUREG/CR-5268," Aging Study of Boiling Water Re-actor Residual Heat Removal System
- NUREG/CR-5280, " Age-Related Degradation of Steiner, J.R Westbghouse 480-Volt Circuit Preakers," Vol.1 "Ag-NISTlR 4487," Detection of incipient Defects in Ca-ing Assessment and Recommendations for Improving bles by Partial Discharge Signal Analysis "
Breaker Reliability,"
81 NUREG-1377 J
Personal Author Index NUREG/CR-5280, " Age-Related Degradation of NUREG/CR-5052," Operating Experience and Aging Westinghouse 480-Volt Circuit lireakers," Vol. 2, "Me-Assessment of Component Cooling Water Systems in chanical Cycling of a DS-416 Breaker. Test Results "
Pressurized Water Reactors."
NUREG/CR-5419," Aging Assessment of Instrument NUREGICR-5268, " Aging Study of Iloiling Water Re-Air Systems in Nuclear Power Plants."
actor Residual lleat Removal System."
NUREG/CR-5612. " Degradation Modeling with Ap-NUREG/CR-5507, "Results from the Nuclear Plant plications to Aging and Maintenance Effectiveness Aging Research Program:Their Use in Inspection Ac-Evaluation."
tivities."
i NUREG/CR-5643, " Insights Gained from Aging Re-g NUREG/CR-5944, "A Characterization of Check l
NUREG/CR-5693," Aging Assessment of Comp (ment Valve Degradation and Failure Experience in the Nu-l Cooling Water Systems in Pressurized Water Reac-clear Power Industry."
1 tors-Phase II."
l Toman, G.J.
Sugarman, A.C.
NUREG/CR-4257, " Inspection. Surveillance, and ilNLTechnical Report A-3270-3-86," Testing Program Monitoring of Electrical Equipment Inside Contain-for the Monitoring of Degradation in a Continuous ment of Nuclear Power Plants-With Applications to Duty 460-Volt Class "H," 10-1IP Electric Motor "
Electrical Cabics," Vol. L NUREG/CR-4939, " Improving Motor Reliability in NUREG/CR-4257, " Inspection, Surveillance, and Nuclear Power Plants" Volume 1, " Performance Monitoring of Electrical Equipment in Nuclear Power Evaluation and Maintenance Practices"; Volume 2.
Plants," Vol. 2, " Pressure Transmitters."
" Functional Indicator Tests on a Small Electric Motor NUREG/CR-4715. "An Aging Assessment of Relays Subjected to Accelerated Aging"; Volume 3," Failure and Circuit lireakers and System Interactions.
Analysis and Diagnostic 'Ibsts on a Naturally Aged Electric Motor."
NUREG/CR-4819," Aging and Service Wear of Sole-noid-Operated Wlves Used in Safety Systems of Nu-1 Sullivan, K.
clear Power Plants," Vol.1, " Operating Experience and Failure Identification."
NUREG/CR-5555, " Aging Assessment of the West-inghouse PWR Control Rod Drive System."
NUREG/CR-5141,"Agingand Qualification Research.
on Solenoid Operated Valves."
Taylor,J.H.
IINLTechnical Report A-3270-Il-85," Seismic Endur-Van Brunt, RJ.
ance'Ibsts of Naturally Aged Small Electric Motors."
NISTlR 4487, " Detection of Incipient Defects in Ca-bles by Partial Discharge Signal Analysis."
HNLTechnical Report A-3270-12-86," Aging and Life Extension Assessment Program (ALEAP) Systems Vasudevan, R.
Level Plan."
HNL Techmcal Report A-3270-12-85, " Pilot Assess-NUREG/CR-4156, " Operating Experience and ment: Impact of Aging on the Seismic Performance of -
Aging-Scismic Assessment of Electric Motors."
Selected Equipment 'I) pes."
NUREGICR-4564, " Operating Experience and Aging-Scismic Assessment of flattery Chargers and Vause,J.W.
Inverters."
NUREG/CR-4590, " Aging of Nuclear Station Diesel NUREG/CR-4939, " Improving Motor Reliability in Generators: Evaluation of Operating and Expert Expe-r ence," Vols. I and 2.
Nuclear Power Plants" Volume 1, " Performance Evaluation and Maintenance Practices"; Volume 2,
" Functional Indicator 'Ibsts on a Small Electric Motor Vesely, W.E.
Subjected to Accelerated Aging"; Volume 3. " Failure BNL Technical Report A-3270 6-21-91," Degradation Analysis and Diagnostic Tests on a Naturally Aged Mocleling: Extensions and Applications."
Electric Motor "
NUREG/CR-4'/69, " Risk Evaluations of Aging Phe-NUREG/CR-4985, " Indian Point 2 Reactor Coolant
- nomena: The Linear Aging Reliability Model and Its Pump Seal Evaluations."
Extensions."
NUREG-1377 82 c
Persota! Authorindex NUREG/CR-5510, " Evaluations of Core Melt Fre-Weeks, J.
quency Effects Due to Component Aging and Maintenance.
NUREG/CR-4985, " Indian Point 2 Reactor Coolant Pump Seal Evaluations."
NUREG/CR-5587, " Approaches for Age-Dependent Probabilistic Safety Assessments with Emphasis on Weir, T.J.
Prioritization and Sensitivity Studies."
NUREG/CR-5008, " Development of a Testing and NUREG/CR-5612. " Degradation Modeling with Ap.
Analysis Methodology To Determine the Functional plications to Aging and Maintenance Effective, ness Condition of Solenoid Operated Valves."
Evaluation."
Werry, E.V.
Villaran, M.
NUREG/CR-5386, "Hasis for Snubber Aging Re-NUREG/CR-5419," Aging Assessment of Instrument scarch: Nuclear Plant Aging Research Program."
Air Systems in Nuclear Power Plants."
NUREG/CR-5490, " Regulatory Instrument Review:
y
' y*p-Management of Aging of LWR Major Safety-Related Components." Vol.1.
NUREG-Il44, " Nuclear Plant Aging Research (NPAR) Program Plan."
NUREG/CR-5870. "Results of LWR Snubber Aging Research."
NUREG-1144, " Nuclear Plant Aging Research (NPAR) Program Plan," Rev.1.
Wilhelm, W.
I Wang, J.K.
NUREG/CR-5280, " Age-Related Degradation of Westinghouse 480-Volt Circuit lircakers." Vol. 2, "Mc-NUREG/CR-5159, " Prediction of Check Valve chanical Cycling of a DS-416 Breaker. 'Ibst Performance and Degradation in Nuclear Power Plant i
Results."
Systems."
NUREG/CR-5583, " Prediction of Check Valve Per-Winegardner, W.K.
formance and Degradation in Nuclear Power Plant Sys-NUREG/CR-6029," Phase I Aging Assessment of Nu-l tems-Wear and Impact Tests.
cicar Air.'Ireatment System HEPA Filters and NUREG/CR-5807 " Improvements in Motor Oper.
Adsorbers "
ated Gate Valve Design and Prediction Models for Nuclear Power Plant Systems "
Wolford, A.J.
NUREG/CR-5378," Aging Data Analysisand Risk As-Watk. ins, J.C.
sessment-Development and Dernonstration Study."
EGG-SSRE-9777," Isolation Valve Assessment (IVA)
Software Version 3.10, User's Manual."
Zaloudek, F. R.-
l EGG-SSRE-9926," Evaluation of EPRI Draft Report
" Safety implications of Diesel Generator Aging,"
l NP-7065-Review of NRC/INEL Gate Valve Test Nuclear Safety, December 1990.
- 8"*
NUREG/CR-5057, " Aging Mitigation and Improved NUR EG/CR-5558, " Generic issue 87: Flexibic Wedge Programs for Nuclear Service Diesel Generators."
Gate Valve Test Program: Phase 11 Results and Analysis,"
Zimmermen, P.W.
NUREG/CR-5720, " Motor. Operated Valve Research NUREG/CR 4379,"Nuc! car Plant Service Water Sys.
Update."
tem Aging Degadation Assessment: Phase I," Vol.1.
i 83 NUREG-1377
CORPORATE AUTHOR INDEX This index lists, in alphabetical order, the organizations that prepared the reports listed in this compila-tion. 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.
NUREG/CR-4715, An Aging Assessment of Relays NUREG/CR-4928, " Degradation of Nuclear Plant and Circuit lireakers and System Interactions."
lbmperature Sensors "
NUREG/CR-4939, " Improving Motor Reliability in u ear wer ManC Mnw 1, "Perfonnance NUREGICR-5383, "Effect of Aging on Response Time of Nuclear Plant Pressure Sensors."
va apon and Maintenance Practices,,; Volume 2, Functional Indicator Tests on a Small Electric Motor NUREG/CR-5560, " Aging of Nuclear Plant Resis-Subjected to Accelerated Aging"; Volume 3," Failure tance Temperature Detectors."
Analysis and Diagnostic Tests on a Naturally Aged i
Electric Motor."
Brookhaven National Laboratory (BNL)
NUREG/CR-4985, " Indian Point 2 Reactor Coolant Letter Report, M. Subudhi," Review of Aging-Seismic Pump Seal Evaluations."
Correlation Studies on Nuclear. Plant Equipment,"
N UREG/CR-5051, " Detecting and Mitigating Battery January 1985.
Charger and Inverter Aging."
BNL lbchnical Report A-3270-11-26-84, " Scoping NUREG/CR-5052," Operating Experience and Aging Test on Containment Purge and Vent Valve Seal Assessment of Component Cooling Water Systems in Material."
Pressurized Water Reactors."
BNL1bchnicalReport A-3270-Il-85,"SeismicEndur-NUREG/CR-5053," Operating Experience and Aging ance Tests of Naturally Aged Small Electric Assessment of Motor Control Centers "
Motors."
NUREG/CR-5192, " Testing of a Naturally Aged BNL Technical Report A-3270-12-85, " Pilot Assess.
Nuclear Power Plant Inverter and Battery Charger "
ment: Impact of Aging on the Seismic Performance of NUREG/CR-5268, " Aging Study of Boiling Water Re-Selected Equipment 'lipes."
actor Residual Heat Removal System."
BNL1bchnical Report A-3270-3-86,"1bsting Program NUREG/CR-5280, " Age-Related Degradation of for the Monitoring of Degradation in a Continuous Westinghouse 480-Volt Circuit Breakers," Vol.1,"Ag-Duty 460 Volt Class "11," 10-IIP Electric Motor."
ing Assessment and Recommendations for Improving I
UNL'Ibchnical Report A-3270-12-86, " Aging and Life Extension Assessment Program (ALEAP) Systems NUREG/CR-5280, " Age-Related Degradation, of Level Plan."
Westinghouse 480-Volt Circuit Breakers," Vol 2,"Me-chanical Cycling of a DS-416 Breaker, 1bst BNL1bchnical Report A-3270R-2-90, " Aging Effects Results."
of Important Balance of Plant Systems in Nuclear Power Plants."
NUREG/CR-5419. " Aging Assessment of Instrument Air Systems in Nuclear Power Plants."
BNL1cchnical Report A-3270 6-21-91," Degradation NUR. EG/CR-5507, *Results from the Nuclear Plant Modeling: Extensions and Applications."
Aging Research Program: Their Use m Inspection BNL Technical Report TR-32704-90, " Maintenance Activities "
Tham inspection Results: Insights Related to Plant NUREG/CR-5555, " Aging Assessment of the Wes-l
' A *8' E
tinghouse PWR Control Rod Drive System."
BNL lbehnical Report TR-3270-9-90, "An Opera-
. NUR EG/CR-5612. " Degradation Modeling with tional Assessment of the Babcock & Wilcox and Com-Applications to Aging and Maintenance Effectiveness bustion Engineering Control Rod Drives."
Evaluation."
NUREG/CR-4156, " Operating Experience :and NUREG/CR-5643, " Insights Gained from Aging
. Aging-Seismie Assessment of Electric Motors."
Research."
NUREG/CR-4564, " Operating Experience and NUREG/CR-5693," Aging Assessment of Component Aging-Scismic Assessment of Battery Chargers and Cooling Water Systems in Pressurized Water Reac -
Inverters."
tors-Phase II."
85 NUREG-1377 y
3 p- - - - -
A--w s
Corporate AuthorIndex NUREG/CR-5783, " Aging Assessment of the Com-NUREG/CR-4747,"An Aging Failtre S'evey of Light bustion Engineering and llabcock & Wilcox Control Water Reactor Safety Systems and Components," Vols.
Rod Drives."
1 and 2.
NUREG/CR-4769. " Risk Evaluations of Aging Phe-Energy Technology Engineering Center nomena: The Linear Aging Reliability Model and its NUREG/CR-5515, " Light Water Reactor Pressure Extensions."
Isolation Valve Performance Testing."
NUREG/CR-4%7," Nuclear Plant Aging Research on liigh Pressure Injection Systems."
Engineering and Economics Research,Inc.
NUREG/CR-4977, " SHAG Test Series: Seismic (EER)
Research on an Aged United States Gate Valve and on Letter Report, L N. Rib, " Summaries of Research a Piping System in the Decommissioned lleiss-Reports Submitted in Connection with the Nuclear dampfreaktor (HDR): Summary," Vol.1.
Plant Aging Research (NPAR) Program."
NUREG/CR-4977, " SHAG Test Series: Seismic Research on an Aged United States Gate, Valve and on Franklin Research Center Pipmg ' System m the Decommissioned Heiss-a NUREG/CR-4715,"An Aging Assessment of Relays dampfreaktor (HDR): Appendices," Vol. 2.
and Circuit Breakers and System Interactions."
NUREG/CR-5181, " Nuclear Plant Aging Research:
NUREG/CR-5141 "Agingand Qualification Research The IE Power System."
on Solenoid Operated Valves."
NUREG/CR-5248, "Prioritization of TIRGALEX-Recommended Components for Further Aging Idaho National Engineering Laboratory (INEL)
NUREG/CR-5314, " Life Assessment Procedures for EGG-SSRE-8972, "Estimatint, Mazard Functions for M j r LWR Components" Vol. 3 " Cast Stainless Repairable Components..
Steel Components.
EGG-SSRE-9017," User's Guide to PHAZE, a Com-NUREG/CR-5378," Aging Data Analysisand Risk As-puter Program for Parametric Ilazard Function Esti.
sessment-Development and Demonstration Study" mation.',
NUREG/CR-5406, "BWR Reactor Water Cleanup EGG-SSRE-9777 " Isolation Valve Assessment (IVA)
System Flexible Wedge Gate Isolation Valve Qualifica-Software Version 3.10, User's Manual.~
tion and High Energy Flow Interruption Test," Vol.1, EGO-SSRE-9926," Evaluation of EPRI Draft Report
" Analysis and Conclusion."
NP-7065--Review of NRC/INEL Gate Valve 'Ibst NUREG/CR-5406, "BWR Reactor Water Cleanup ogram.
System Flexible Wedge Gate Isolation Valve Qualifica-EGG-SSRE-10039,"An Evaluation of the Effects of tion and High Energy Flow Interruption Test," Vol. 2, Valve Body Erosion on Motor-Operated Valve Oper.
" Data Report."
ability."
NUREG/CR-M06, "IlWR Reactor Water Cleanup NUREG/CR-3819, " Survey of Aged Power Plant System Flexible Wedge Gate Isolation Valve Qualifica-tion and High Energy Flow Interruption Test," Vol. 3.
Facilities "
" Review of Issues Associated with BWR Containment NUREG/CR-3956, "In Situ Testing of the Ship-1 solation Valve Closure."
pingpon Atomic Power Station Electrical Circuits "
NUREG/CR-5448," Aging Evaluation of Class 1E Bat-NUREG/CR-4457, " Aging of Class 1E Batteries in teries: Seismic Testing "
Safety Systems of Nuclear Power Plants."
NUREG/CR-5558," Generic issue 87: Flexibic Wedge NUREG/CR-4731, " Residual Life Assessment of Gate Valve Test Program: Phase H Results and Major Light Water Reactor Components," Vol.1.
Analysis."
NUREG/CR-4731, " Residual Life Assessment of NUREG/CR-5646, " Piping System Response During Major Light Water Reactor Components-Overview,"
11igh Level Simulated Seismic Tests at the Vol.2.
Heissdampfreaktor Facility (SH AM Test Series)."
NUREG/CR-4740, " Nuclear Plant-Aging Research NUREG/CR-5720," Motor-Operated Valve Research on Reactor Protection Systems."
Update."
NUREG-1377 86
Corporate AuthorIndex Kalsi Engineering, Inc.
NUREG/CR-4234, " Aging and Service Wear of NUREG/CR-5159, " Prediction of Check Valve Per-Electric Motor-Operated Valves Used in Engineered formance and Degradation in Nuclear Power Plant Safety-Feature Systems of Nuclear Power Plants,"
Systems."
V"I I-
"8 ""
9 NUREG/CR-5583, " Prediction of Check Valve Per-
- 8 "E "# '#
formance and Degradation in Nuclear Power Plan t Sys-tems-Wear and Impact Tests.,,
Safety-Feature Systems of Nuclear Power Ilants: Ag-i ing Assessments and Monitoring Method Evalu-I NUREG/CR-5807, " Improvements in Motor Oper.
ations," Vol. 2.
ated Gate Valve Design and Prediction Models for NUREG/CR-4257, " Inspection, Surveillance, and Nuclear Power Plant Systems "
Monitoring of Electrical Equipment Inside Contain-ment of Nuclear Power Plants-With Applications to Lake Engineering Company Electrical Cables," Vol.1.
j NUREG/CR-5386, "llasis for Snubber Aging Re-NUREG/CR-4257, " Inspection. Surveillance, and search: Nuclear Plant Aging Research Program."
Monitoring of Electrical Equipment in Nuclear Power Plants," Vol. 2, " Pressure Dansmitters."
National Institute of Standards and NUREG/CR-4302, " Aging and Senice Wear of Check Technology Valves Used in Engineered Safety-Feature Systems of
' NISTIR 4485, " Annotated Bibliography: Diagnostic Nuclear Power Plants," Vol.1.
Methods and Measurement Approaches 1b Detect In-NUREG/CR-4302," Aging and Service Wear of Check cipient Defects Duc :o Aging of Cables."
Valves Used in Engineered Safety-Feature Systems of NISTIR 4487, " Detection of Incipient Defects in Nuclear Power Plants," Vol. 2," Aging Assessments and Cables by Partial Discharge Signal Analysis.-
Monitoring Method Evaluations."
NISTIR 4787,"Ihe Use of Time-Domain Dielectric NUREG/CR-4380, " Evaluation of the Motor-Spectroscopy 7b Evaluate the Lifetime of Nuclear Operated Valve Analysts and 7bst System (MOVATS)
Power Station Cables.,,
7b Detect Degradation, Incorrect Adjustments, and Other Abnormalitics m Motor-Operated Valves."
l Nuclear Regulatory Commission (NRC)
NUREG/CR-4597," Aging and Service Wear of Auxil-j Techm. cal Integration Review Group for Aging and Lif iary Feedwater Pumps for. PWR Nuclear Power Extension (TIRGALEX)," Plan for Integration of Ag-Plants," Vol.1, " Operating Experience and Failure Identification."
ing and Life-Extension Activities."
1 NUREG/CR-4597," Aging and Service Wear of Auxil-1 NUREG-Il44, " Nuclear Plant Aging Research iary Feedwater Pumps for PWR Nuclear Power (NPAR) Program Plan."
Plants," Vol. 2, " Aging Assessments and Monitoring -
NUREG-1144, " Nuclear Plant Aging Research Method Evaluations."
(NPAR) Program Plan," Revision 1.
NUREG/CR-4652. " Concrete Component Aging and NUREG-1144, " Nuclear Plant Aging Research its Significance Relative to Life Extension of Nuclear wer Mants, (NPAR) Program Plan, Status and Accomplishments,"
Revision 2.
NUREG/CR-4692, " Operating' Experience Review of Failures of Power Operated Relief %1ves and Block NUREG/CP-0100, " Proceedings of the International Valves in Nuclear Power Plants."
Nuclear Power Plant Aging Symposium."
NUREG/CR-4819,. " Aging and Senice Wear of Oak Ridge National Laboratory (ORNL)
S lenoid-Operated Valves used in Safety Systems of Nuclear Power Plants, Vol.1," Operating Experience
'NUREG/CR-2641, "The In-Plant Reliability Data and Failure Identification."
Base for Nuclear Power Plant Components: Data Col-lection and Methodology Report."
NUREG/CR-4819, "Agm.g and Service Wear of Sole-noid-Operated Valves Used in Safety Systems of NUREG/CR-3154, "The In Plant Reliability Data Nuclear Power Plants," Vol. 2, " Evaluation of Moni-Base for Nuclear Plant Componen ts: Interim Report-toring Methods."
The Valve Component."
NUREG/CR-4992, " Aging and Service Wear of Multi-NUREG/CR-3543," Survey of 0perating Experiences stage Switches Used in Safety Systems of Nuclear j
~
from LERs 7b Identify Aging Rends."
Power Plants," Vol.1.
j l
87 NUREG-1377
Corporate AuthorIndex NUREG/CR-5404, " Auxiliary Feedwater System Ag-NUREG/CR-5057, " Aging Midgation and Improved ing Study," Vol.1.
Programs for Nuclear Service Diesel Generators "
NUREG/CR-5404, " Auxiliary Feedwater System Ag-NUREG/CR-5379," Nuclear Plant Service Water Sys-ing Phase i Follow-on Study," Vol. 2.
tem Aging Degradation Assessment: Phase I," Vol.1, NUREG/CR-5479, " Current Applications of Vibra.
NUREG/CR-5379," Nuclear Plant Service Water Sys-tion Monitoring and Neutron Noise Analysis: Detec.
tem Aging Degradation Assessment," Vol. 2.
tion and Analysis of Structural Degradation of R,cactor NUREG/CR-5386, " Basis for Snubber Aging Re-i Vessel Internals from Operational Aging."
search: Nuclear Plant Aging Research Program."
NUREG/CR-5519," Aging of Control and Service Air NUREG/CR-5490, " Regulatory Instrument Review:-
1 Compressors and Dryers Used in Nuclear Power Management of Aging of LWR Major Safety-Related Plants," Vol. L Components," Vol.1.
NUREG/CR-5699, " Aging and Senice Wear of Con-NUREG/CR-5491, "Shippingport Station Aging trol Rod Drive Mechanisms for IlWR Nuclear Plants,"
Evaluation."
Y"b '
NUREG/CR-5848, "Recordkeeping Needs 1b Miti-NUREG/CR-5700," Aging Assessment of Reactor In-gate the Impact of Aging Degradation."
strumentation and Protection Systems Components."
NUREG/CR-5870, "Results of LWR Snubber Aging NUREG/CR-5706, "NRC Bulletin 88-04: Potential Research."
Safety-Related Pump Loss-An Assessment of Indus-NUREG/CR-6001, " Aging Assessment. of BWR i
try Data "
Standby Liquid Control Systems."
NUREG/CR-5754, " Boiling-Water Reactor Internals NUREG/CR-6029,"PhaseI Aging Assessment of Nu-Aging Degradation Study, Phase 1."
clear AirIReatment System HEPA Filters and Adsorb.
'#8' NUREG/CR-5779, " Aging of Non-Power-Cycle Heat Exchangers Used in Nuclear Power Plants," Vol.1.
NUREG/CR-6043," Phase I Aging Assessment of Es-NUREG/CR-5944, "A Characterization of Check Valve Degradation and Failure Experience in the Nu.
PNL-5722, " Operating Experience and Aging Assess-clear Power Industry."
ment of ECCS Pump Room Coolers."
NUREG/CR-6048," Pressurized-Water Reactor inter.
PNL-6287," Study Group Review of Nuclear Service.
nals Aging Degradation Study, Phase L" Diesel Generator Testmg and Aging Mitigation."
PNL-7516, " Emergency Diesel Generator Technical ORNL/NRC/LTR-91/25 " Throttled Valve Cavitation Specifications Study Results.
and Erosion.,,
PNL-7823," Maintenance Practices 10 Manage Aging:
Pacific Northwest Laboratory (PNL)
A Review of Several'lbchnologies."
" Safety Implications of Diesel Generator Aging" PNL-SA-18407, " Understanding and Managing Cor-L Nuclear Safety, December 1990.
rosion in Nuclear Power Plants."
NUREG/CP-0105, PaperbyJ. A.Christensen "NPAR PNL-SA-20219,"ASME Subsection ITFD Recommen-Approach to Controlling Aging in Nuclear Power dations Based upon NPAR Snubber Aging Research Plants."
Results."
1 i
QUREG/CR-4144, "Importance Ranking Based on Pentek, Inc.
Aging Consideration of Components included in NUREG/CR-5008, " Development of a Testing and Probabilistic Risk Assessments.
. Analysis Methodology lo Determine the Functional NUREG/CR-4279, " Aging and Service Wear of Hy.
Condition of Solenoid Operated Valves."
draulic and Mechanical Snubbers Used on Safety-Related Piping and Components of Nuclear Power Sandia National Laboratories (SAND) '
Plants," Vol.1.
NUREG/CP-0036," Proceedings of the Workshop on Nuclear Plant Aging."
r NUREG/CR-4590, " Aging of Nuclear Station Diesel Generators: Evaluation of Operating and Erpert Expe-NUREG/CR-3818. " Report of Results of Nuclear rience," Vols. I and 2.
Power Plant Aging Workshop."
NUREG-1377 88
Corporate AuthorIndex NUREG/CR-5334," Severe Accident Testing of Elec-Science Applications International Corp.
trical Penetration Assemblies."
NUREG/CR-5248, "Prioritization of TIRGALEX-NUREG/CR-5461, " Aging of Cables, Connections, Recommended Components for Further Aging Re-and Electrical Penetration Assemblics Used in Nuclear search.
Power Plants "
NUREG/CR-5510 " Evaluations of Core Melt Fre.
NUREG/CR-5546, "An investigation of the Effects of quency Effects Due to Compcment Aging and Maintenance."
Thermal Aging on the Fire Damageability of Electric Cables."
NUREG/CR-5587, " Approaches for Age-Dependent NUREG/CR-5619,"The Impact of Thermal Aging on Prpbabilistic Safety Assessments with Emphasis on Pn ntization and Sensitivity Studies.
the Flammability of Electric Cables."
NUREG/CR-5655," Submergence and High Tempera-Wyle Laboratories ture Steam Testing of Class IE Electrical Cables,"
NUREG/CR-5386, "llasis for Snubber Aging Re-NUREG/CR-5772, " Aging, Condition Monitoring, search: Nuclear Plant Aging Research Program."
and Loss-of-Coolant Accident (LOCA) Tests of Class NUREG/CR-5762, " Comprehensive Aging Assess-l IE Electrical Cables," Vols.1,2, and 3.
ment of Circuit fireakers and Relays."
j SAND 88-0754 UC-78,
" Time-Temperature-Dose WYLE 60103-X, " Test Plan for the Comprehensive f
Rate Superposition: A Methodology for Predicting Ca.
Aging Assessment of Circuit Ilreakers and Relays for ble Degradation Under Ambient Nuclear Power Plant Nuclear Plant Aging Research (NPAR) Program, j
Aging Conditions."
Phase 11,"
l l
89 NUREG-1377
1 SUBJECT INDEX i
In this index, each report is listed under one or more of the following subjects i
1.
Aging, including Plans, Surveys, Analyses, Methods, and Models.
2.
Dicsci Generators and Related Systems.
3.
Electric Power Systems, including Cabics,'Irays, Connectors, Circuit Ilreakers, Relays, Switches, Penetrations, and Related Components.
4.
Electrical Equipment, Including 'Iransformers, Motors, Hatteries, Chargers, and Inverters.
5.
Instrumentation, Measurement, and Control Systems.
6.
Maintenance.
7.
Major Components: Reactor Vessels, Reactor Coolant Pumps, Steam Generators, Pressurizers, I
and Structures (Including Ilasemat and Containment).
8.
Monitoring.
9.
Operating Experience, Field Results, and Related Data.
- 10. Piping, including Valves, Scals, Supports, Snubbers, and Related Components.
I 1.
Probabilistic Risk Assessment (PRA),
- 12. Safety and Protection Systems (including Injection Systems) and 'Dicir Components.
- 13..Scismic Effects and Aging.
- 14. Service Water, Auxiliary Feedwater, Instrument Air, and Other Fluid Systems, Including Their Pumps, !! cal Exchangers, and Related Components; llalance of Plant Systems and 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 knowledge-able personnel sccking published information on the various aspects of nuclear plant aging.
- 1. Aging, including Plans, Surveys, Analyses, NISTIR 4487," Detection of incipient Defects in Ca-Methods, and Models bles by 1*artial Discharge Signal Analysis" Letter Report, L. N. Rib, "Surnmaries of Research NISTIR 4787, "The Use of Time-Domain Dielectric Reports Submitted in Connection with the Nuclear Spectroscory Tb Evaluate the Lifetime of Nuclear Plant Aging Research (N!%R) Program,"
Power Statm i Cables."
NUREG/CP-0105, Paper by J A. Christensen, Technical Integration Review Group for Aging and "NI%R Approach to Controlling Aging in Nuclear Life Extension (IIRGALEX), " Plan for Integration Power Plants."
of Aging and Life-Extension Activities" NUREG-1144, " Nuclear ' Plant Aging Research
(
).
8 '" * ""'
HNL 'Itchnical-Report A-3270-12-86, " Aging and Life Extension Assessment Program (ALEAP) Sys.
NUREG-ll44, " Nuclear ' Plant Aging Research tems Level Plan."
(NPAR) Program Plan " Revision 1.
NUREG-Il44, "Nuc! car Plant Aging Research j
HNL 'Ibchnical Report A-3270 6-21-91, "Degrada-(NPAR) Program Plan, Status and Accomplish-tion Modeling: Extensions and Applications "
ments," Revision 2.
EGG-SSRE-8972,. " Estimating flazard Functions NUREG/CP-0036,"Proceedingsof theWorkshopon for Repairable Components "
Nuclear Plant Aging" NUREG/CP-0RK), " Proceedings of the Interna-EGO-SSRE-9017 " User's Guide to PIIAZE, a tional Nuclear Power Plant Aging Symposium."
put r Program for Iarametric flazard Function NUREG/CR-3818 " Report of Results of Nuclear Power Plant Agmg Workshop."
NISTIR 4485," Annotated Bibliography: Diagnostic NUREG/CR-4144,"Importance Ranking Based on -
Methods and Measurement Approaches 1b Detect -
Aging Consideration of Comp (ments included in Incipient Defects Due to Aging of Cables."
Probabilistic Risk Assessments."
)
91 NUREG-1377 1
I
Subject Index NUREG/CR-4652, " Concrete Component Aging Cable Degradation Under Ambient Nuclear Power and Its Significance Relative to Life Extension of Plant Aging Conditions "
Nuclear Power Plants "
WYLE 60103-X, Test Plan for the Comprehensive NUREG/CR-4731, " Residual Life Assessment of Aging Assessment of Circuit Hrcakers and Relays for Major Light Water Reactor Commments," Vol.1, Nuclear Plant Aging Research (NPAR) Program, Phase !!."
NUREG/CR-4731, " Residual Life Assessment of Major Ught Water Reactor Components-Over-
- 2. Diesel Generators and Related Systems view, Vol.2.
" Safety Implications of Diesel Generator Aging" NUREG/CR-4769," Risk Evaluations of Aging Phe-Nuclear Safety, December 1990.
nomena:The Linear Aging Reliability Model and Its Extensions NUREG/CR-4590," Aging of Nuclear Station Diesel a
Generators: Evaluation of Operating and Expert Ex-NUREG/CR-4819, " Aging and Service Wear of perience," Vols. I and 2.
Solenoid-Operated Valves Used in Safety Systems of NUREG/CR-4731. " Residual Life Assessment of Nuc! car Power Plants, Vol.. Esaluation of Mom-b1 j r Light Water Reactor Comp (ments-Over-toring Methods."
vicw," Vol. 2.
NUREG/CR-5008," Development of a Testing and NUREG/CR-5057," Aging Mitigation and improved Analysis Methodology'Ib Determine the I unctional Programs for Nuclear Service Diesel Generators."
Condition of Solenoid Operated Valves.
NUREG/CR-5490,"RegulatoryInstrument Review:
NUREG/CR-5248, "Prioritization of TIRGALEX.
Management of Aging of LWR Major Safety-Related Recommended Commments for Further Aging Re' Components," Vol.1.
search.
I PNL-6287, " Study Group Review of Nuclear Senice NUREG/CR-5314," Life Assessment Proceduresfor Diesci Generator Testing and Aging Mitigation."
4 Major LWR Components " Vol. 3. " Cast Stainless Stect Components."
PNL-7516. " Emergency Diesel Generator Technical Specifications Study Results."
NUREG/CR-5386, "Hasis for Snubber Aging Re-scarch: Nuclear Plant Aging Research Program."
- 3. Electric Power Systems, Including Cables, NUREG/CR-5491, "Shippingport. Station Aging Trays, Connectors, Circuit Breakers, Re-Evaluation."
lays, Switches, Penetrations, and Related -
Components NUREG/CR-5507, "Results from the Nuclear Plant in i escarch Program: their Use in Inspection g.Ibchnical Report A-3270-12-85," Pilot Assess-7 g
g; p
7 Selected Equipment Types." '
. NUREG/CR-5510 " Evaluations of Core Melt Fre-UNL lbchnical Report A-3270R-2-90, " Aging Ef-quency Effects Due to Commment Aging and fects of Important Halance of Plant Systems in Nu-Maintenance, clear Ibwer Plants."
NUREG/CR-5583," Prediction of Check Valve Per-NIS11R 4487, " Detection of Incipient Defects in Ca-formance and Degradation in Nuclear Power Plant bles by Partial Discharge Signal Analysis "
Systems-Wear and Impact 'Ibsts."
NIS11R 4787,'The Use of Time-Domain Dielectric NUREG/CR-5612, " Degradation Modeling with Spectroscopy 1b Evaluate the Lifetime of Nuclear Applications to Aging and Maintenance Effective
- Power Station Cables."
ness Evaluation."
NUREG/CR-3956, "In Situ Testing of the Ship-NUREG/CR-5807, "Impnivements in Motor Oper-pingport Atomic Power Station Electrical Circuits."
ated Gate Valve Design and Prediction Models for Nuclear Power Plant Systerns.-
NUREG/CR-4257, " Inspection, Surveillance, and Monitoring of Electrical Equipment inside Contam-PN L-S A-18407, "Understamling and Managing Cor-mentof NuclearPowerPlants-With Applicationsto rosion in Nuclear Power Plants "
Electrical Cables."
SANDSS-0754 UC-78, " Time-Temperature-Dose NUREG/CR-4715."An Aging Assessment of Relays Rate SuperpcJtion: A Methodology for Predicting and Circuit Breakers and System Interactions."
NUREG-1377 92
l Subject Index NUREG/CR-4731, " Residual Life Assessment of
- 4. Electrical Equipment, Including 'IYans-Major Light Water Reactor Components-Over-formers, Motors, Batteries, Chargers, view," Vol. 2.
and Inverters NUREG/CR-4747, "An Aging Failure Survey of UNL Technical Report A-3270-11-85, "Scismic En-Light Water Reactor Safety Systems and Compo-durance Tests of Naturally Aged Small Electric Mo-nents," Vol.1.
tors."
.N UREG/CR-4992, " Aging and Service Wear of M ul-HNLTechnical Report A-3270-12-85," Pilot Assess-tistage Switches Used in Safety Systems of Nuclear ment: Impact of Aging on the Scismic Performance of Power Plants," Vol.1.
Selected Equipment Types "
NUREG/CR-5181," Nuclear Plant Aging Research:
HNL Technical Report A-3270-3-86, " Testing Pro-The IE lbwer System."
gram for the Monitoring of Degradation in a Con-tinuous Duty 460-Volt Class "H," 10-IIP Electric NUREG/CR-5280, " Age-Related Degradation of Motor."
Westinghouse 480-Volt Circuit Breakers," Vol.1,
" Aging Assessment and Recommendations for Im.
NUREG/CR-4156, " Operating Experience and proving Hrcaker Reliability."
Aging. Seismic Assessment of Electric Motors."
NUREG/CR-5280, " Age-Related Degmdation of NUREG/CR-4457, " Aging of Class IE Hatteries in Westinghouse 480-Volt Circuit Breakers," Vol. 2, Safety Systems of Nuclear Power Plants.
" Mechanical Cycling of a DS-416 Hreaker. Test Re-NUREG/CR-4564, " Operating Experience and suits."
Aging-Scismic Assessment of Battery Chargers and NUREG/CR-5461, " Aging of Cables, Connections, and Electrical Penetration Assemblics Used in Nu-NUREG/CR-4939, " Improving Motor Reliability in clear Power Plants."
NucicJr Power Plants": Volume 1. " Performance Evalunion and Maintenance Practices"; Volume 2, NUREG/CR-5546, "An Investigation of the Effects
" Functional lndicatorlests on a Small Electric Motor of Thermal Aging on the Fire Damageability of Elec-tric Cables "
Subjected to Accelerated Aging"; Volume 3," Failure Analysis and Diagnostic Tests on a Naturally Aged NUREG/CR-5619, "Ihe Impact of Thermal Aging Electric Motor."
on the Flammability of Electric Cables" NUREG/CR-5051, " Detecting and Mitigating Hat-NUREG/CR-5655, " Submergence and fligh Tem-tery Charger and Inverter Aging "
perature Steam 'R sting of Class IE Electrical Ca-NUREG/CR-5053," Operating Experience and Ag-bles."
ing Assessment of Motor Control Centers."
NUREG/CR-5762, " Comprehensive Aging Assess-NUREG/CR-5192," resting of a Naturally Aged Nu-ment of Circuit Breakers and Relays."
clear Power Plant Inverter and Battery Charger."
.NUREG/CR-5772, " Aging, Condition Monitoring, NUREG/CR-5448, " Aging Evaluation of Class lE and Loss-of-Coolant Accident (LOCA) Tests of Hatteries: Scismic Testing."
Class IE Electrical Cables," Vol.1.
NUREG/CR-5643 "InsightsGainedfrom AgingRe-NUREG/CR-5772, " Aging, Condition Monitoring, scarch."
and Loss-of-Coolant Accident (LSCA) Tests of Class IE Electrical Cabics," Vol. 2.
- 5. Instrumentation, Measurement, and NUREG/CR-5772, " Aging, Condition Monitoring, and Loss-of-Coolant Accident (LOCA) Tests of HNL Technical Report TR-3270-9-90, "An Opera-Class IE Electrica1 Cables," Vol. 3.
tional Assessment of the Habcock & Wilcox and Combustion Engineering Control Rod Drives "
SAND 88-0754 UC-78, " Time-Temperature-Dose Rate Superposition: A Methodology for Predicting NUREG/CR-4257, " Inspection, Surveillance, and Cable Degradation Under Ambient Nuclear Power -
Monitoring of Electrical Equipment m Nuclear Plant Aging Conditions."
Power Plants " Vol. 2, " Pressure 'Iransmitters."
WYI.E 60103-X," Test Plan for the Comprehensive
. emp rature Senus.,fg a n uclear nt Aging Assessment of Circuit Hrcakers and Relays for Nuclear Plant Aging Research (NPAR) Program, NUREG/CR-5383, "Effect of Aging on Response Phase 11."
Time of Nuclear Plant Pressure Sensors."
l 93 NUREG-1377 e
J Subject index NUREG/CR-5546, "An Investigation of the Effects NUREG/CR-4597," Aging and Service Wear of Aux-of Thermal Aging on the Fire Damageability of Elec-iliary Feedwater Pumps for PWR Nuclear Power tric Cables "
Plants," Vol. 2, " Aging Assessments and Monitoring NUREG/CR-5555, " Aging Assessment of the Wes-tinghouse PWR Control Rod Drive System."
NUREG/CR-4939, " Improving Motor Reliability in Nuclear Power Plants". Volume 1, " Performance NUREG/CR-5560, " Aging of Nuclear Plant Resis-Evaluation and Maintenance Practices"; Volume 2, tance Ihmperature Detectors.,,
" Functional IndicatorTests on a Small Electric Motor NUREG/CR-5655, " Submergence and High Tem.
Subjected to Accelerated Aging"; Volume 3," Failure perature Steam 'Ibsting of Class lE Electrical Ca.
Analysis and Diagnostic Tests on a Naturally Aged Electric Motor."
bles."
NUREG/CR-5051, *8 et a a W Mitigui," Bat-NUREG/CR-5699," Aging and Service Wear of Con.
trol Rod Drive Mechanisms for BWR Nuclear tery Charger and Inve ar n,y e Plants," Vol. l-NUREG/CR-5057,"/,ir,
).. ei i y hweed NUREG/CR-5700, " Aging Assessment of Reactor Programs for Nuclear S:rvice D.W
. man. '
Instrumentation and Protection Systems Compo*
NUREG/CR-5181," Nuclear Plant Agii.g De mr.
nents."
' Die IE Power System."
NUREG/CR-5783," Aging Assessment of the Com-NUREG/CR-5280, " Age-Related Degradattor of bustion Engineering and Babcock & Wilcox Control Westinghouse 480-Volt Circuit Ilreakers," Vol.1, Rod Drives."
" Aging Assessment and Recommendations for Im-Proving Breaker Reliability."
NUREG/CR-6029, " Phase I Aging Assessment of Nuclear Air-Treatment System HEPA Filters and NUREG/CR-5280, " Age-Related Degradation of Adsorbers," Vol.1.
Westinghouse 480-Volt Circuit Breakers," Vol. 2,
" Mechanical Cycling of a DS-416 Breaker.1bst Re-NUREG/CR-6043, " Phase I Aging Assessment of suits.,
Essential liVAC Chillers Used in Nuclear Power Plants."
NUREG/CR-5519, " Aging of Control and Service Air Compressors and Dryers Used in Nuclear Power
- 6. Maintenance Plants " Vol.1.
" Safety Implications of Diesel Generator Aging,"
NUREG/CR-5612, " Degradation Modeling with Nuclear Safety, December 1990.
Applications to Aging and Maintenance Effective-ness Evaluadon."
BNL Technical Report TR-3270-6-90, "Mainte.
nance 'Ibam Inspection Results: Insights Related to PNL-5722, " Operating Experience and Aging As-Plant Aging."
sessment of ECCS Pump Room Coolers."
NUREG/CR-4234," Aging and Service Wear of Elec-PNL-7823, " Maintenance Practices To Manage Ag-tric Motor-Operated Valves Used in Engineered ing: A Review of Several Technologies."
Safety-Feature Systems of Nuclear Power Plants,"
Voi,1.
- 7. Major Components: Reactor Vessels, Reac-tor Coolant Pumps, Steam Generators,
- NUREG/CR-4302, " Aging and Service Wear of Pressur,zers, and Structures (Including i
Check Valves Used in Engineered Safety-Feature l
Systems of Nuclear Power Plants," Vol. 2, " Aging Basemat and Containment) l.
Assessments and Monitoring Method Evaluations."
NUREG/CR-4652, " Concrete Component -- Aging.
and Its Significance Relative to Life Extension of i
NUREG/CR-4457, " Aging of Class 1E Batteries in Nuclear Power i lants.
l Safety Systems of Nuclear Power Plants.
NUREG/CR-4731 " Residual Life Assessment of NUREG/CR-4564," Operating Experience and Ag-Major Light Water Reactor Components,, Vol.1.
ing-Scismic Assessment of Battery Chargers and In-verters."
NUREG/CR-4731, " Residual Life Assessment of a er eactor Components-Over-NUREG/CR-4597, " Aging and Service Wear of Aux-3 '
iliary Feedwater Pomps for PWR Nuclear Power Plat.ts," Vol.1. "O9erating Experience and Failure NUREG/CR-5334, " Severe Accident' Testing of Identificah" Electrical Penetration Assemblies."
j NUREG-1377 94 i
1 Subject Index NUREG/CR-5479," Current Applications of Vibra-NUREG/CR-4564, " Operating Experience and tion Monitoring and Neutron Noise Analysis: Detec-Aging-Seismic Assessment of Hattery Chargers and tion and Analysis of Structural Degradation of Reac.
Inverters."
tor Vessel Internals from Operational Aging."
NUREG/CR-4597, " Aging and Senice Wear of Aux-NUREG/CR-5490," Regulatory Instrument Review:
iliary Feedwater Pumps for PWR Nuclear Power-Management of Aging of LWR MajorSafety-Related Plants," Vol.1, " Operating Experience and Failure Components," Vol.1.
Identification."
NUREG/CR-5754, "Hoiling-Water Reactor Inter-NUREG/CR-4597," Aging and Service Wear of Aux-nals Aging Degradation Study, Phase 1.
iliary Feedwater Pumps for PWR Nuclear Power Plants," Vol. 2 " Aging Assessments and Monitoring NUREG/CR-6048, " Pressurized-Water Reactor In.
Method Evaluations."
ternals Aging Degradation Study, Phase I."
NUREG/CR-4819, " Aging and Senice Wear of PNL-SA-18407 " Understanding and Managing Cor-Solenoid-Operated Valves Used in Safety Systems of rosion in Nuclear Power Plants.
Nuclear Power Plants," Vol.1, " Operating Experi-ence and Failure Identification."
NUREG/CR-4939, " Improving Motor Reliability in
- 8. Mon. tor.ing i
Nuclear Power Plant,'- Volume 1. " Performance BNL Tbchnical Report A-3270-3-86, " Testing Pro-Evaluation and Maintenance Practices"; Volume 2, gram for the Monitoring of Degradation in a Con-
" Functional !ndicator Tests on a Small Electric Motor tinuous Duty 460-Volt Class "H," 10-HP Electric Subjected to Accelerated Aging"; Volume 3," Failure Motor "
Analysis and Diagnostic Tests on a Naturally Aged Electric Motor "
BNL Technical Report A-3270-12-86, " Aging and Life Extension Assessment Program (ALEAP) Sys-NUREG/CR-4%7, "Nuc! car Plant Aging Research tems Level Plan."
on High Pressure Injection Systems."
NUREG-1144, " Nuclear Plant Aging Research NUREG/CR-5008, " Development of a Testing and (NPAR) Program Plan," Rev.1.
M&dolog To Determine the Functional Condition of Solenoid Operated Valves."
NUREG/CR-3543, " Survey of Operating Experi-NUREG/CR-5051, " Detecting and Mitigating Hat-ences from LERs To Identify Aging Trends."
tery Charger and Inverter Aging."
N U R EG/CR-4234, " Aging and Senice Wear of Elec-NUREG/CR-5053," Operating Experience and Ag-tric Motor-Operated Valves Used in Engineered ing Assessment of Motor Control Centers."
Safety-Feature Systems of Nuclear Power Plants,"
yog, g, NUREG/CR-5057, " Aging Mitigation and Improved Programs for Nuclear Senice Diesel Generators."
j NUREG/CR-4234," Aging and Senice Wear of Elec-NUREG/CR-5181," Nuclear Plant Aging Research:
tric Motor-Operated Valves Used m Engineered l
Safety-Feature Systems of Nuclear Power Plants: Ag-The IE Power System'"
ing Assessments and Monitoring Method Evalu.
NUREG/CR-5192,' Testing of a Naturally Aged Nu-ations," Vol. 2.
clear Power Plant Inverter and Battery Charger "
NUREG/CR-4257, " Inspection, Surveillance, and NUREG/CR-5461, " Aging of Cables, Connections,
. Monitoring of Electrical Equipment Inside Contain, and Electrical Penetration Assemblies Used in Nu-l ment of Nuclear Power Plants-With Applications to clear Power Plants."
Electrical Cables," Vol. L NUREG/CR-5479," Current Applications of Vibra-NUREG/CR-4257, " Inspection, Surveillance, and tion Monitoringand Neutron Noise Analysis:Detec-Monitoring of Electrical Equipment in Nuclear.
tion and Analysis of Structural Degradation of Reac-Power Plants " Vol. 2, " Pressure Transmitters."
tor Wssel Internals kom Operadonal Aging.
NUREGICR-5519, " Aging of Control and Service NUREG/CR-4302, " Aging and Service Wear of Air Compressors and Dryt rs Used in Nuclear Power Check Valves Used in Engineered Safety-Feature Plants," Vol.1.
Systems of Nuclear Power Plants." Vol.1.
NUREG/CR-5612. " Degradation Modeling with NUREG/CR-4457, " Aging of Class IE Hatteries in Applications to Aging and Maintenance Effective-Safety Systems of Nuclear Power Plants" ness Evaluation."
95 NUREG-1377
Subject Index 4
PNL-6287, " Study Group Review of Nuclear Service NUREG/CR-4747, "An Aging Failure Suncy of Diesel Generator Testing and Aging Mitigation."
Light Water Reactor Safety Systems and Compo-nents," Vols.1 and 2.
- 9. Operating Experience, Field Mits, and NUREG/CR-4819, " Aging and Service Wear of Related Data Solenoid-operated Valves Used in Safety Systems of.
" Safety Implications of Diesel Generator Aging,"
Nuclear Power Plants," Vol.1. " Operating.Experi-Nuclear Safety, December 1990.
ence and Failure Identification."
BNL 'Ibchnical Report TR-3270-9-90, "An Opera.
NUREG/CR-4%7," Nuclear Plant Aging Research tional Assessment of the liabcock & Wilcox and on IIigh Pressure injection Systems "
Combustion Engineering Control Rod Drives."
NUREG/CR-4992," Aging and Service Wear of Mul-NUREG/CR-2641, "The In-Plant Reliability Data tistage Switches Used in Safety Systems of Nuclear Power Plants," Vol.1.
Base for Nuclear Power Plant Components: Data Collection and Methodology Report."
NUREG/CR-5052," Operating Experience and Ag-NUREG/CR-3154, "'i %: in-Plant Reliability Data ing Assessment of Component Cooling Water Sys-i tems m Pressurized Water Reactors.
Base for Nuclear Plant Components: Interim Re-port-The Valve Component."
NUREG/CR-5181," Nuclear Plant Aging Research:
NUREG/CR-3543, " Survey of Operating Expen.-
The 1E Power System."
ences from LERs'Ib Identify Aging'Irends."
NUREG/CR-5268, " Aging Study of Boiling Water Reactor Residual Heat Removal System."
NUREG/CR-3819," Survey of Aged Power Plant Fa-cilities."
NUREG/CR-5280, " Age-Related Degradation of Westinghouse 480-Volt Circuit Breakers," Vol.1, NUREG/CR-4156, " Operating Experience and
" Aging Assessment and Recommendations for Im-Aging-Seismic Assessment of Electric Motors."
proving Breaker Reliability."
i NUREG/CR-4234," Aging and Service Wear of Elec-NUREGICR-5280, " Age-Related Degradation of tric Motor-Operated Valves Used in Engineered Westinghouse 480-Volt Circuit Breakers," Vol. 2, Safety-Feature Systems of Nuclear Power Plants,"
" Mechanical Cycling of a DS-416 Breaker. Test Re-Vol.1.
suits."
NUREG/CR-4302, " Aging and Service Wear of NUREG/CR-5379, " Nuclear Plant Service Water Check Valves Used in Engineered Safety-Feature System Aging Degradation Assessment: Phase I,"
Systems of Nuclear Power Plants," Vol.1.
Vol. L NUREG/CR-4457, " Aging of Class IE Batterics in NUREG/CR-5383, "Effect of Aging on Response Safety Systems of Nuclear Power Plants."
Time of Nuclear Plant Pressure Sensors."
NUREG/CR-4564," Operating Experience and Ag-NUREG/CR-5404,"AuxiliaryFeedwaterSystem Ag-ing-Scismic Assessment of Battery Chargers and in.
ing Study," Vol.1.
m ters?
NUREG/CR-5419, " Aging Assessment' of Instru-NUREG/CR-4590," Aging of Nuclear Station Diesel ment Air Systems in Nuclear Power Plants."
Generators: Evaluation of Operating and Expert Ex-NUREG/CR-5461, " Aging of Cables, Connections, perience, Vols. I and 2.
and Electrical Penetration Assemblies Used in Nu-NUREG/CR-1597," Aging and Service Wear of Aux-clear Power Plants."
iliary Feedwater Pumps for PWR Nuclear Power NUREG/CR.5507,"Results from the Nuclear Plant -
Plants," Vol.1, " Operating Experience and Failure Aging Research Program: Their Use in Inspection 1 Identification."
Activities."
NUREG/CR-4692, " Operating Experience Review NUREG/CR-5519, " Aging of Control and Service of Failures of Power Operated Relief Valves and Air Compressors and Dryers Used in Nuclear Power Block Valves in Nuclear Power Plants."
Plants," Vol.1.
NUREG/CR-4715,"An Aging Assessment of Relays NUREG/CR-5555, " Aging Assessment of the Wes-and Circuit Breakers and System Interactions."
tinghouse PWR Control Rod Drive System."
NUREG/CR-4740, " Nuclear Plant Aging Rucarch NUREG/CR-5560, " Aging of Nuclear Plant Resis-on Reactor Protection Systems."
tance Temperature Detectors."
Subject Index NUREGICR-5706,"NRC Bulletin 88-04: Potential NUREG/CR-4380 " Evaluation of the Motor-Safety-Related Pump Loss-An Assessment of In-Operated Valve Analysis and Test System (MOVATS) dustry Data."
To Detect Degradation, Incorrect Adjustments, and Other Abnormalities in Motor-Operated Valves "
NUREG/CR-5848, "Recordkeeping Needs 7b Mite gate the impact of Aging Degradation."
NUREG/CR-4819 " Aging and Service Wear of Solenoid-Operated Valves Used in Safety Systems of PNL-5722. " Operating Experience and Aging As-Nuclear Power Plants," Vol.1, " Operating Expen-sessment of ECCS Pump Room Coolers."
ence and Failure Identification."
PNL-7516, " Emergency Diesel Generator Technical NUREG/CR-4819, " Aging and Service Wear of Specifications Study Results.'
Solenoid. Operated Valves Used in Safety Systems of Nuclear Power Plants," Vol. 2," Evaluation of Moni.
- 10. Piping, Including Valves, Seals, Supports, toring Methods "
Snubbers, and Related Components NUREG/CR-4977, "SIIAG Test Series: Seismic HNL Technical Report A-3270-11-26-84, " Scoping Research on an Aged United States Gate Valve and 7bst on Containment Purge and Vent Valve Seal Ma-on a Piping System in the Decommissioned Ileiss.
terial."
dampfreaktor (HDR): Summary," Vol.1.
UNLTechnical Report A-3270-12-85," Pilot Assess-NUREG/CR-4977, " SHAG Test Series: Seismic ment: Impact of Aging on the Seismic Performance of Research on an Aged United States Gate Valve and Selected Equipment Types."
on a Piping System in the Decommissioned Heiss-d mpfreaktor (HDR): Appendices," Vol. 2.
EGG-SSRE-97/7, " Isolation Valve Assessment (IVA) Software Version 3.10 User's Manual."
NUREG/CR-4985," Indian Point 2 Reactor Coolant
"
- E ""'
EGG-SSRE-9926, " Evaluation of EPRI Draft Re.
port NP-7065-Review of NRC/INEL Gate Valve NUREG/CR-5008," Development of a Testing and Test Program."
Analysis Methodology Tb Determine the Functional E# #
EGG-SSRE-10039,"An Evaluation of the Effects of Valve Body Erosion on Motor-Operated Wlve Oper.
NUREG/CR-5141, " Aging and Qualification Re.
ability."
search on Solenoid Operated Valves."
NUREG/CR-3154, "The In-Plant Reliability Data NUREG/CR-5159," Prediction of Check Valve Per.
Hase for Nuclear Plant Components: Interim Re-formance and Degradation in Nuclear Power Plant port-The Valve Component."
Systems."
NUR EG/CR-4234," Aging and Service Wear of Elec.
NUREG/CR-5386, "Hasis for Snubber Aging Re-tric Motor-Operated Valves Used in Engineered scarch: Nuclear Plant Aging Research Program."
Safety-Feature Systems of Nuclear Power Plants" NUREG/CR-5406, "HWR Reactor Water Cleanup Vol.1.
. System Flexible Wedge Gate Isolation valve Qualifi.
N UREG/CR-4234, " Aging and Service Wear of Elec.
cation and High Energy Flow Interruption Test," Vol.
tric Motor Operated Valves Used in Engineered 1," Analysis and Conclusion."
Safety-Feature Systems of Nuclear Power Plants: Ag-NUREG/CR-5406, "HWR Reactor Water Cleanup ing Assessment and ' Monitoring Method Evalu-System Flexible Wedge Gate Isolation Wlve Qualifi-ations," Vol. 2.
cation and High Energy Flow 1nterruptionTest," Vol.
NUREG/CR-4279," Aging and Senice Wear of Hy _
2, " Data Report."
draulic and Mechanical Snubbers Used on Safety-Re-NUREG/CR-5406, "HWR Reactor Water Cleanup lated Piping and Components of Nuclear Power System Flexible Wedge Gate Isolation Valve Qualifi-Plants," Vol.1.
cation and High Energy Flow Interruption Test," Vol.
3, " Review of Issues. Associated with HWR Contain-NUREG/CR-4302, " Aging and Service - Wear of men. Isolation Valve Closure.
Check Valves Used in Engineered Safety-Feature Systems of Nuclear Ibwer Plants," Vol.1.
NUREG/CR-5490, " Regulatory Instrument Review:
M nagement of Agingof LWR MajorSafety-Related NUREG/CR-4302, " Aging and Senice Wear of Check Valves Used in Engineered Safety-Feature Components,,, Vol.1.
Systems of Nuclear Ibwer Plants," Vol. 2, " Aging NUREG/CR-5491, "Shippingport Station Aging Assessments and Monitoring Method Evaluations."
Evaluation."
97 NUREG-1377
J I
Subject Index NUREG/CR-5515, " Light Water Reactor Pressure
- 12. Safety and Protection Systems (Including isolation Valve Performance Testing."
Injection Systems) and Their Components NUREG/CR-5558, " Generic Issue 87: Flexible NUREG/CR-3819 " Survey of Aged Ibwer Plant Fa-Wedge Gate Valve 'Ibst Program: Phase II Results cilities."
and Analysis."
NUREG/CR-4302, " Aging and Service Wear of NUREG/CR-5583," Prediction of Check Valve Per-Check Valves Used in Engineered Safety-Feature formance and Degradation in Nuclear Power Plant Systems of Nuclear Power Plants," Vol.1.
Systems-Wear and Impact 1bsts."
NUREG/CR-4302, " Aging and Service Wear of NUREG/CR-5643,, Insights Gam.ed from Aging Re-Check Valves Used in Engineered Safety-Feature search.
Systems of Nucicar Power Plants," Vol. 2, " Aging Assessments and Monitoring Method Evaluations."
NUREG/CR-5646," Piping System Response During NUREG/CR-4731, " Residual Life Assessment of High Level Simulated Seismic Tests at the Major Light Water Reactor Components-Over-Heissdampfreaktor Facility (SHAM lbst Series)."
view." Vol. 2.
NUREG/CR-5720, " Motor-Operated Valve Re-NUREG/CR-4740, " Nuclear Plant-Aging Research search Update."
on Reactor Protection Systems."
NUREG/CR-5807, " Improvements in Motor Oper-NUREG/CR-4747, "An Aging Failure Survey of ated Gate Valve Design and Prediction Models for Light Water Reactor Safety Systems and Compo-Nuclear Power Plant Systems."
nents," Vols. I and 2.
NUREG/CR-5870,"Results of LWR Snubber Aging NUREG/CR-4%7," Nuclear Plant Aging Research Research."
on High Pressure Injection Systems."
NUREG/CR-5944, "A Characterization of Check NUREG/CR-4992," Aging and Service Wear of Mul-Valve Degradation and Failure Experience in the Nu-tistage Switches Used in Safety Systems of Nuclear clear Power Industry."
Power Plants," Vol.1.
NUREG/CR-5558, " Generic Issue 87: Flexible GRNL/NRC/III11-91/25, " Throttled Valve Cavita-r j
tion and Erosion.,
Wedge Gate Valve lest Program: Phase 11 Res' tits and Analysis."
PNL-S A-l M07, " Understanding and Managing Cor*
NUREG/CR-5643,"InsightsGainedfrom AgingRe.
l<
rosion in Nuclear Power Plants."
- search "
l PNL-SA-20219, "ASME Subsection ISTD Recom-NUREG/CR-5700, " Aging Assessment af Reactor mendations Based upon NPAR Snubber Aging Re-Instrumentation and Protection Systems Compa-search Results."
nents."
- 11. Probabilistic Risk Assessment (PRA)'
- 13. Seismic _ Effects and Aging NUREG/CR-4144,"importance Ranking Based on; Letter Report, M. Subudhi, Review of Aging-Seismic.
Aging Consideration of Components Included in Correlation Studies on Nuclear Plant Equipment," '
Probabilistic Risk Assessments."
Brookhaven National Laboratory, January 1985.'
NUREG/CR-5268, " Aging Study of Boiling Water BNL Technical Report A-3270-11-85, " Seismic En-~
. Reactor Residual Heat Removal System.
durance Tests of Naturally Aged Small Electric Motors" NUREG/CR-5378, " Aging Data Analysis and Risk UNLTechnical Report A-3270-12-85, " Pilot Assess-Assessment-Development. and Demonstration '
. Study.
ment: Impact of Agingon the Seismic Performance of :
Selected Equipment lypes."
NUREG/CR-5510," Evaluations of Core Melt Fre-NUREG/CR-4156, " Operating Experience and Ag-quency Effects Due to Component' Aging and ing Seismic Assessment of Electric Motors."
Maintenance."
i NUREG/CR-4279," Aging and Service Wear of Hy-i NUREG/CR-5587, " Approaches for Age-Depend-draulic and Mechanical Snubbers Used on Safety-ent Probabilistic Safety Assessments with Emphasis Related Piping and Comp (ments of Nuclear Power j
on Prioritization and Sensitivity Studies."
Plants," Vol.1.
NUREG-1377 98
Subject Index NUREG/CR-4977, "SH AG Test Scrics: Seismic NUREG/CR-4985," Indian Point 2 Reactor Coolant Research on an Aged United States Gate Valve and Pump Scal Evaluations."
on a Piping System in the Decommissioned NUREG/CR-5052," Operating Experience and Ag.
IIctssdampfreaktor (IIDR): Summary, Vol.1.
ng Assessment of Componen-i'ooling Water Sys-NUREG/CR-4977, "SiI AG 'Ibst Series: Seismic Re-tems in Pressurized Water Ret.wrs.
search on an Aged United States Gate Valve and on a NUREG/CR-5268, " Aging Study of Iloiling Water Piping System m the Decommissioned lleiss.
g g ;gy g
g3 dampfreaktor (IIDR): Appendices, Vol...
I NUREG/CR-5379, " Nuclear Plant Service Water NUREG/CR-5448, " Aging Evaluation of Class IE ynem Ag ng egradation Assessment: Phase 1, llatterics: Scismic Testing "
Vol.1.
l NUR EG/CR-SM6. " Piping System Response During NUREG/CR-5379, " Nuclear Plant Service Water 11igh Level Simulated Seismic rests at the System Aging Degradation Assessment," Vol. 2.
IIcissdampfreaktor Facility (SIIAM Test Series)"
l NUREG/CR-5404," Auxiliary Feedwater System Ag-
- 14. Service Water, Auxiliary Feedwater, In.
ing Study," Vol.1.
strument Air, and Other Fluid Systems, NUREG/CR-5404," Auxiliary Fecdwater System Ag-Including Their Pumps, Ileat Exchangers, ing Phase I Follow-on Study," Vol. 2.
and Related Components; Ilalance of NUREG/CR-5419, " Aging Assessment of Instru-Plant Systems and Components ment Air Systems in Nuclear Power Plants."
IINL 'Ibchnical Report A-3270R-2-90, " Aging Ef-fccts of Important llalance of Plant Systems in Nu.
NUREG/CR-5519, " Aging of Control and Scnice clear Power Plants."
Air C mpressors and Dryers Used in Nuclear Power Plants," Vol.1.
NUREG/CR-4597," Aging and Senice Wear of Aux-NUREG /CR-5643," Insights Gam.edfrom AgingRe-iliary Feedwater Pumps for PWR Nuclear Power scarch.
Plants," Vol.1, " Operating Experience and Failure Identification."
NUREG/CR-5693, " Aging Assessment of Compo-nent Cooling Water Systems in Pressurized Water NUREG/CR-4597," Aging and Service Wear of Aux-Reactors-Phase 11.
iliary Feedwater Pumps for PWR Nuclear Power Plants," Vol. 2. " Aging Assessments and Monitoring NUREG/CR-5706,"NRC llulletin 88-N: Potential Method Evaluations."
Safety-Related Pump Loss-An Assessment of In-NUREG/CR-4731, " Residual Life Assessment of dustry Data."
Major Light Water Reactor Comp (ments," Vol.1.
NUREG/CR-5779, " Aging of Non-Power-Cycle IIcat Exchangers Used in Nuclear Power Plants,"
NUREG/CR-4731. " Residual Life Assessment of Major Light Water Reactor Components-Over-
~'
view," Vol. 2.
NUREG/CR-6001, " Aging Assessment of IlWR NUREG/CR-4747, "An Aging Failure Survey of Light Water Reactor Safety Systems and Compo-PN L-SA-18407, "Und :rstanding and Managing Cor-ncnts," Vol.1.
rosion in Nuclear Ibv cr Plants."
99 NUREG-1377
1 u
CHRONOLOGICAL LISTING (in order of publication) 1.
'NUREG/CR-2641, J. It Drago, R. J. Borkowski, 10.
N UR EG/CR-4156, M. Subudhi, E. L llurns, and J.
D. IL Pike, and E E Goldberg,"The In-Plant Reli-1L 'Ihylor, " Operating Experience and Aging-ability, Data Base for Nuclear Power Plant Compo-Seismic Assessment of Electric Motors," Brook-nents: Data Collection and Methodology Report "
haven National Laboratory, IINL-NUREG-51861, Oak Ridge National Laboratory, ORNL/IM-8271, June 1985.
July 1982.
11.
NUREG/CR-4234, W. L Greenstreet, G, A.
2.
'NUREG/CP-0036, (Compilation by) B. E. Bader Murphy, and D. M. Eissenberg," Aging and Service i
and L A. llanchey," Proceedings of the Workshop Wear of Electric Motor-Operated Valves Used in on Nuclear Plant Aging," Sandia National Lalmra-Engineered Safety-Feature Systems of Nuclear tories, SAND 82-2264C, November 1982.
Power Plants," Vol.1 Oak Ridge National Latwra-3.
NUR EG/CR-3154, R. J. Borkowski, W. K. Kahl, T.
L liebble, J. R. Fragola, and J. W. Johnson, "The 12.
NUREG-1144, H. M. Morris and J. P. Vora, in-Plant Reliability Data Base for Nuclear Plant
" Nuclear Plant Aging Research (NPAR) Program Components: Interim Report-The Valve Compo-Plan," U.S. Nuclear Regulatory Commission, July nent," Oak Ridge National Laboratmy, ORNL/
1985.
TM-8647, December 1983.
13.
NUREG/CR-4257, S. Ahmed, A. Carfagno, and G.
4.
NUREG/CR-3543, G. A. Murphy, R. IL Gallaher, J. 'Ibman, " Inspection, Surveillance, and Monitor-M. L. Casada, and IL C. Iloy," Survey of Operating ing of Electrical Equipment inside Containment of Experiences from LERs to Identify Aging' Rends,"
Nuclear Power Plants-With Applications to Elec-Oak Ridge National Laboratory, ORNL-trical Cables," Oak Ridge National Laboratory, NSIC-216. January 1984.
ORNL/SUB/83-28915/1, August 1985.
5.
NUREG/CR-3818, N. IL Clark and D. L Berry, 14.
UNLlechnical Report' A-3270-11-85, J. IL1hylor,
" Report of Results of Nuclear Power Plant Aging M. Subudhi, J. liiggins, J. Curreri, M. Reich, Workshop,"
Sandia National Lalmratories.
E Cifuentes, and T. Nehring, "Scismic Endurance 1
SAND 84-0374, August 1984.
Tests of Naturally Aged Small Electric Motors,"
l Brookhaven National Laboratory, November 1985.
,l 6.
IINL lechnical Report A-3270-11-26-84,
- 11. Miller," Scoping 1bst on Containment Purge and 15.
BNLTechnical Report A-3270-12-85, M. M. Silver, Vent Valve Seal Material," Brookhaven National R.Vasudevan,and M.Subudhi," Pilot Assessment:
Laboratory, December 1984.
Impact of Aging on the Seismic Performance of Selected Equipmentlypes," Brookhaven National 7.
_ Letter Report, M. Subudhi, " Review of Aging-Laboratory, December 1985.
Seismic Correlation Studies on Nuclear Plant Equipment," Brookhaven National Laboratory, 16.
NUREG/CR-4302, W.. L Greenstreet, G. A.
January 1985.
Murphy, R. B. Gallaher, and D. M. Eissenberg, j
" Aging and Service Wear of Check Valves Used in -
8.
NUREG/CR-4144, T.
- Davis, A.
- Shafaghi, Engineered Safety-Feature Systems of Nuclear -
R. Kurth, and E Leverenz, "Importance Ranking Ibwer Plants," Vol.1 Oak Ridge National Labora-Based on Aging Consideration of Components In-tory, ORNL-6193/V1, December 1985.
cluded in Probabilistic Risk Assessments," Pacific Northwest laboratory, PNL-5389, April 1985.
17.
NUREG/CR-4380, J. L. Crowley and D. M.
Eissenberg, " Evaluation of the Motor-Operated 9.
NUREG/CR-3819, J. A. Rose, R. Steele, Jr., K. G.
Valve Analysis and Test System (MOVATS) to DeWall, and B. C. Cornwell, " Survey of Aged Detect Degradation, Incorrect Adjustments, and Ibwer Plant Facilities," Idaho National Engineer-Other Abnormalities in Motor-Operated Valves,"
ing Laboratory, EGG-2317, June 1985.
Oak Ridge National Laboratory, ORNL-6226, j
January 1986.
NpnNhc a ly 18.
NUREG/CR-4279, S.11. Bush, P. G. Heasier, and N* E. Dodge, " Aging and Service Wear of Hydraulic 8 " ' " "
R t a n
101 NUREG-1377
__.m Chronological Listing r
and Mechanical Snubbers Used on Safety-Related ence and Failure Identification," Oak Ridge N1-Piping and Components of Nuclear Power Plants,"
tional Laboratory, ORNUSUll/83-28915/4/V1, Vol.1, Pacific Northwest Laboratory, PNL-5479, March 1987.
February 1986.
28.
NUREG/CR-3956, M.
R.
Dinsel,. M R.
19.
BNL Technical Rcport A-3270-3-86, A. C. Sugar.
Donaldson, and E T. Soberano, "In Situ Testing of man, M. W. Sheets, and M. Subudhi, "1bsting Pro.
the Shippingport Atomic Power Station Electrical gram for the Monitoring of Degradation in a Con.
Circuits " Idaho National Engineering Laboratory, tinuous Duty 460-Volt Class "B," 10-HP Flectric EGG-2443, April 1987.
Motor," llrookhaven National Laboratory, March l
29.
NUREG/CR-4769, W. E. Vesely, " Risk Evalu-ations of Aging Phenomena:The Linear Aging Re-20.
NUREG/CR-4564, W. E. Gunther, M. Subudhi, liability Model and Its Extensions," Idaho National and J. I L1hylor," Operating Experience and Aging.
Engineering Laboratory, EGG-2476, April 1987.
Seismic Assessment of Battery Chargers and Inver-30.
1bchnical Integration Rev.iew Group for Aging and ters," Brookhaven National Laboratory, HNL_
NUREG-51971, June 1986.
Life Extension (11RGALEX), " Plan for Integra-tion of Aging and Life-Extension Activities," U.S.
21.
NUREG/CR-4597, M. L Adams and E. Makay, Nuclear Regulatory Commission, May 1987.
" Aging and Service Wear of Auxiliary Feedwatc5 31.
NU R EG/CR-4715, G. J.1bman, V. P. Bacanskas, T.-
I umps for I WR Nuclear Power Plants, Vol.1
" Operating Experience and Failure identification,"
A. Shook, and C. C. Lodlow,"An Aging Assessment Oak Ridge National Laboratory, ORNL-6282/V1, f Relay and Circuit Breakers and System Interac-J Q 1986' tions," Brookhaven National Latmratory, Franklm Research Center, Philadelphia, PA, BNL-
)
NUREG-52017, June 1987.
22.
NUREG/CR-4257, G. J. lbman, " Inspection, Sur-veillance, and Monitoring of Electrical Equipment 32.
NUREG/CR-4731, V. N. Shah and P. E. Mac-m Nuclear Power Plants. Vol. 2: I ressure frans-Donald," Residual Life Assessment of Major Light mitters Oak Ridge National Laboratory, ORNU SUB/83-28915/3/V2, August 1986' Water Reactor Components," Vol.1, Idaho Na-tional Engineering L.aboratory, EGG-2469. June 1987*
J 23.
Letter Report, L. N. Rib, " Summaries of Research Reports Submitted in Connection with the Nuclear
- 33. NUREG/CR-4928, II. M.11ashemian, K. M. Peter-Plant Agmg Research (NPAR) Program," Engi-sen, T. W. Kerlin, R. L. Anderson, and K. E. Hol-neering and Economics Research (EER) Inc., Res-bert, " Degradation of Nuclear Plant "Ibmperature ton, VA. September 1986.
Sensors," Analysis and Measurement Services Cor-P"#"" "' " *"
' """ 9 24.
NUREG/CR-4652, D. J. Naus, " Concrete Compo-nent Aging and its Significance Relative to Life 34.
NUREG/CR-4457, J. L. Edson and J. E. Hardin, Extension of Nuclear Power Plants," Oak Ridge
.' Aging of Class 1E Hatteries in Safety Systems of National 12boratory, ORNLffM-10059, Septem.
Nuclear Power Plants," Idaho National Engineer.
ber 1986, ing Laboratory, EGG-2488, July 1987.
25.
PNL-5722, D. E. Blahnik and R. L. Goodman,"O -
35.'
NUREG/CR-4747, IL M. Meale and D. G. Sat-P erating Experience and Aging Assessment of ECCS terwhite,"An Aging Failure Survey of Light Water Pump Room Coolers." Pacific Northwest Labora-Reactor Safety Systems and Components," Vot-1, J
tory, October 1986.
Idaho Nati(mal Engineering Laboratory; '
l l
EGG-2473, J uly 1987.
26.
HNLTechnical Report A-3270-12-86, R. Fullwood,
~
{
J. C. Higgins, M. Subudhi, and J. IL Thylor, " Aging 36.
NUREG/CR-4590, K. R. Hoopingarner, J. W.
and Life Extension Assessment Program (ALEAP)
Vause, D. A. Dingee, and J. E Nesbitt, " Aging of Systems Level Plan," Brookhaven National Labo-Nuclear Station Diesel Generators: Evaluation of ratory, December 1986.
. Operating and Expert Experience,".Vols.1 and 2, l
Pacific Northwest Laboratory, PNU-5832, August 27.
NUR EG/CR--4819, V. P. Bacanskas, G. C. Roberts, 1987.
and G. J. Ibman, " Aging and Service Wear of Sole-noid-Operated Valves Used in Safety Systems of 37.
NUREG/CR-4985, M. Subudhi, J. H.1hylor, J.
Nuclear Power Plants," Vol.1:" Operating Experi-Clinton, C. J. Czajkowski,' and J. Weeks, " Indian NUREG-1377 4 102
Chronological Listing Point 2 Reactor Coolant Pump Seal Evalua.
Service Wear of Auxiliary Feedwater Pumps for tions," Brookhaven National Laboratory, HNL-PWR Nuclear Power Plants," Vol. 2:" Aging Assess-NUREG-52095, August 1987.
ments and Monitoring Method Evaluations," Oak Ridge National Laboratory, ORNL-6282/V2, June 38.
NUREG-1144, J. P. Vora, " Nuclear Plant Aging 1988.
Research (NPAR) Program Plan," Rev.1, U.S.
47.
NUREG/CR-4747, H. M/ Meale and D. G. Sat-Nuclear Regulatory Commission, September 1987.
terwhite,"An Aging Failure Survey of Light Water 39.
NUREG/CR-4992, G. C. Roberts, V. P. Bacanskas.
Reactor Safety Systems and Components," Vol. 2, Idaho National Engineering Laboratory, and G. J. lbman, " Aging and Service Wear of M ulti-stage Switches Used in Safety Systems of Nuclear EGG-2473, July 1988.
Ibwer Plants," Vol.1, Oak Ridge National Labora-48.
NUREG/CR-5052, J. C. Iliggins, R. Lofaro, tory, O RNIJSUH/83-28915/5/V1, September 1987.
M. Subudhi, R. Fullwood, and J. H.1hylor, "Oper-
"""8
" ^N"8 A"**"I
- P 40.
NU R EG /CR-5008, R. D. Meinin8er and T. J. Weir, nent Cooling Water Systems m Pressurized Water
" Development of ali stmg and Analysis Methodol-Reactors," Brookhaven National Laboratory, ogylb Determme the Functional Condition of So-IINL-NUREG-52117. July 1988.
lenoid Operated Valves, Pentek, Inc., Coraopohs, PA. September 1987.
49.
NU REG /CR-5053, W. Shier and M. Subudhi, "Op-erating Experience and Aging Assessment of Motor 41.
NUREG/CR-4692, G. A. Murphy and J. W.
Control Centers," Brookhaven National Labora.
Cletcher H," Operating Experience Review of Fail-tory, UNL-NUREG-52118, July 1988.
ures of Power Operated Relief Valves and lilock Valves in Nuclear Power Plants," Oak Ridge Na-50.
NUREG/CR-5051, W. E. Gunther, R. Lewis, and tional Laboratory, ORNi>NOAC-233, October M. Subudhi, " Detecting and Mitigating Battery 1987.
Charger and inverter Aging," Brookhaven Na-tional Laboretory HNL-NUREG-52108, August 42.
NUREG/CR-4939, M. Subudhi, W. E. Gunther, 1988.
J. H.1hylor, R. Lofaro, K. M. Skreiner, A. C. Sugar-man, and M. W. Sheets," Improving Motor Reliabil.
51.
NUREG/CR-5141, V. P. Hacanskas, G. J.1bman, ity in Nuclear Power Plants": Vol.1, " Performance and S. P. Carfagno, " Aging and Qualification Re-Evaluation and Maintenance Practices"; Vol. 2, search on Solenoid Operated Valves ** Franklin Re-
" Functional Indicator 1csts on a Small Electric Mo-search Center, Norristown, PA, August 1988.
tor Subjected to Accelerated Aging"; Vol. 3," Fail-52.
- SAND 88-0754. UC-78, K. T. Gillen and R. L ure Analysis and Diagnostic Tests on a Naturally Aged Electric Motor"; Brookhaven National Labo.
Clough, " Time ' Temperature-Dose Rate Superpo-ratory, HNL-NUREG-52031, November 1987.
sition: A Methodology for Predicting Cable Degra-dation Under Ambient Nuclear Ibwer Plant Agmg Conditions,"Sandia Nationallaboratories, August 43.
NUREGICR-4740, L C. Meyer, " Nuclear Plant-I988-Aging Research on Reactor Protection Systems, Idaho National Engineering 12boratory, EGO-53.
NUREG/CR-5192, W. E. Gunther, " Testing of a 3
2467, January 1988.
Naturally Aged Nuclear Ibwer Plant Inverter and I
Ilattery Charger," Brookhaven National Labora-44.
PNL-6287, K. R. Hoopingarner, H. J. Kirkwood, tory, BNL-NUREG-52158, September 1988.
and P J. Lonzecky, " Study Group Review of Nu-clear Service Diesel Generator Testing and Aging 54.
NUREG/CR-5248, L S. Levy, D. B. Jarrell, and Mitigation," Pacific Northwest Laboratory, March E. P. Collins, "Prioritization of TIRGALEX-1988.
Recommended. Components for Further Aging Research," Pacific Northwest Laboratory, Science 45.
NUREG/CR-5159, M. S. Kalsi, C. L Horst, and.
Applications International Corp., ; PNL-6701, J. K. Wang, " Prediction of Check Valve Perform-November 1988.
ance and Degradation in Nuclear Power Plant Sys-tems," Kalsi Engineering, Inc., Sugar Land, TX,-
$5. NUREG/CP-0100, A. E Heranck," Proceedings of KEl No.1559, May 1988.
the International Nuclear Power Plant Aging Sym-
- 46..NUREG/CR-4597, D. M. Kitch, J. S. Schtonski,
- $"NhIu$c#ar i cNr*p#"bigi s n$ti N$f$h ! b c
1a P. J. Sowatskey, and W. V. Cesarski, " Aging and NPAR pmgram.
103 NUREG-1377
Chronological Listing posium," U.S. Nuclear Regulatory Commission, Energy Flow Interruption Test," Vol.1, " Analysis March 1989.
and Conclusion," Idaho National Engineering Laboratory, EGG-2569, October 1989.
56.
NUREG/CR-5268, R. Lofaro, M. Subudhi, W. E.
Gunther, W. Shier, R. Fullwood, and J. H. *lhylor, 65.
NUREG/CR-5406, K. G. DeWall and R. Steele,
" Aging Study of Boiling Water Reactor Residual Jr.,"BWR Reactor Water Cleanup System Flexible Heat Removal System," HNL-NUREG-52177, Wedge GateIsolation Valve Qualification and High Ilrookhaven National Laboratory, June 1989.
Energy Flow Interruption 'Ibst," Vol. 2. " Data Re-port," Idaho National Engineering Latmratory, 57.
NUREG/CR-5379, D. B. Jarrell, A. B. Johnson, EGG-2569, October 1989.
Jr., P. W. Zimmerman, and M. L. Gore, " Nuclear 66.
NUREG/CR-5406, K. G. DeWall and R. Steele, Plant Service Water System Aging Degradation As.
sessment: Phase 1 " Vol.1, Pacific Northwest Labo-3r.,"BWR Reactor Water Cleanup System Flexib e ratory, PNL-6560, June 1989.
Wedge Gate Isolation Valve Qualification and Iligh Energy Flow Interruption Test," Vol. 3 " Review of Issues Associated with BWR Containment Isola-58.
NUREG/CR-5383, H. M. Hashemian. K. M. Peter-tion Valve Closure," Idaho National Engineering sen, R. E. Fain, and J. J. Gingrich,"Effect of Agin[,
Lalmratory, EGG-2569, October 1989.
on Response Time of Nuclear Plant Pressure Sen sors," Analysis and Measurement Services Corpo-67.
NUREG/CR-4731 V. N. Shah and P. E. Mac-ration, Knoxville, TN, June 1989.
Donald,"Residaal Life Assessment of Major Light Water Reactor Components --Overview," Vol. 2 59.
WYLE 60103-X, J. E Gleason, R. A. DeFour, J. M.
(Draft), Idaho National Engmeermg Laboratory, Hammond, and P. A. Lubeski, ' Test Plan for the EGG-2469, November 1989.
Comprehensive Aging Assessment of Circuit Breakers and Relays for Nuclear Plant Aging Re-68.
NUREG/CR-5334, D.B.Clauss," Severe Accident search (NPAR)I rogram, Phase II, Wyle Laborato-
.hipf NM PmmiMMd h rics, Huntsville, AL, July 1989.
dia National Laboratories, SAND 89-0327, November 1989.
6(L NUREG/CR-4234 H.D.Haynes,"AgingandServ-ice Wear of Electric Motor-Operated Valves Used 69.
NUREG/CR-5057, K. R. Hoopingarner and E R.
m Engineered Safety-Feature Systems of Nuclear Zaloudek, " Aging Mitigation and Improved Pro-Power Plants: Aging Assessments and Monitormg grams for Nuclear Service Diesel Generators," Pa-Method Evaluations," Vol. 2, Oak Ridge National cific Northwest L.aboratory, PNL-6397, December Laboratory, GRNL-6170/V2, August 1989.
- 3939, 6L NUREG/CR-4%7, L C. Meyer, " Nuclear Plant 70.
NUREG/CR-5386, D. P. Brown, Gi R. Palmer, Aging Research on High Pressure Injection Sys' E, V. Werry, and D. E. Blahnik " Basis for Enubber tems," Idaho National Engineering Laboratory.
Aging Research: Nuclear Plant Aging Research EGG-2514, August 1989.
Program, Pacific Northwest Latmratory, Lake Engi-neering Company, Wyle Laboratories, PNL-
- 62.. NUREG/CR-4977, R. Steele, Jr. and J. G.
6911, January 1990.
Arendts," SHAG Tbst Series: Seismic Research on an Aged United States Gate Valve and on a Piping 71.
NUREG/CR-5419, M. Villaran, R. Fullwood, and System in the Decommissioned Heissdampfreaktor M. Subudhi," Aging Assessment of Instrument Air (HDR): Summary," Volf 1, Idaho National Engi-Systems in Nuclear Power Plants," Brookhaven Na-neering Laboratory, EGG-2505, August 1989.
tional Laboratory, HNL-NUREG-52212, January 1990.
63.
NUREG/CR-4977, R. Steele, Jr. and J. G.
Arendts," SHAG Test Series: Seismic Research on 72.
NUREG/CR-5491, R. P. Allen and A. H. Johnson an Aged United States Gate Valve and on a Piping Jr., "Shippingport Station Aging Evaluation," Pa-System in the Decommissioned Heissdampfreakter cific Northwest Laboratory, PNL-7191, January (llDR): Appendices," Vol. 2, Idaho National Engi-1990.
neering Laboratory, EGG-2505, August 1989.
73.
HNL Technical Report A-3270R-2-90, A. Fresco.
64.
NUREG/CR-5406. K. G.'DeWall and R. Steele, and M. Subudhi, " Aging Effects of Important Jr.,"BWR Reactor Water Cleanup System Flexible Balance of Plant Systems in Nuclear Power Plants,"
Wedge Gate Isolation Valve Qualification and High Brookhaven National Laboratory, February 199(k NUREG-1377 104 i
Chronological Listing 74.
NUREG/CR-5479, H. Damiano and R. C. Kiyter, 84.
NUREG/CR-5461, M. J. Jacobus, " Aging of Ca-
" Current Applications of Vibration Monitoring and bles, Connections, and Electrical Penetration As-Neutron Noise Analysis: Detection and Analysis of semblies Used in Nuclear Power Plants," Sandia Structural Degradation of Reactor Vessel Internals National Laboratories, SAND 89-2369, July 1990.
from Operational Aging," Oak Ridge National Laboratory, ORN1/1%11398, February 1990.
85.
NUREG/CR-5515, II.H. Neely, N.M. Jeanmou-I gin, J.J. Corugedo, " Light Water Reactor Pressure k
75.
NUREG/CP-0105, " Proceedings of the Seven.
Isolation Valve Performance Testing," Energy teenth Water Reactor Safety Information Meet-Technology Engineering Testing, ETTiC 88-01, July 1990.
l ing," Vol. 3, U. S. Nuclear Regulatory Commission.
I lbper by J. A. Christensen, "NPAR Approach to 86.
NUREG/CR-5519, Vol.1, J. C. Moyers, " Aging of Controlling Aging in Nuclear Power Plants,"Ibeific Northwest Laboratory, PNL-SA-17487, March Control and Service Air Compressors and Dryers Used in Nuclear Power Plants, Oak Ridge Na-1990' tional Laboratory, ORNL-6607/V1, July 1990.
I 76.
NUREG/CR-54M, D. A. Casada, " Auxiliary Feed-87.
NUREG/CR-5448, J. L Edson, " Aging Evaluation water System Aging Study," Vol.1, Oak Ridge Na-of Class lE Batteries: Seismic Testing," Idaho Na-tional Laboratory, ORNL-6566/VI, March 1990.
tional Engineering Laboratory, EGG-2576, August 1990.
77.
EGG-SSRE-8972,C.L. Atwood "EstimatingHaz-ard Functions for Repairable Components," Idaho 88.
NUREG/CR-5583, M.. S. Kalsi, C. L Horst, J. K.
National Engineering Laboratory, May 1990.
Wang and V. Sharma, " Prediction of Check Valve Performance and Degradation in Nuclear Power 78-NUREG/CR-5181, L C. Meyer and J. L Edson, Plant Systems-Wear and Impact 'Ibsts," Kalsi En-
" Nuclear Plant Aging Research:The IE PowerSys.
gineering, Inc., KEI No.1656, August 1990.
"E 89.
PNL-SA-18407, A. IL Johnson, Jr., D. IL Jarrell, U.
C
-2 5*Ma 1 P. Sinha, and V. N. Shah," Understanding and Man-aging Corrosion in Nuclear Power Plants," Ibcific 79.
IINL Tbchnical Report TR-3270-6-90, W. Gun-Northwest Laboratory, August 1990.
ther, " Maintenance Team Inspection Results: In-sights Related to Plant Aging," Brookhaven Na-90.
HNL Technical Report TR-3270-9-90, E. Grove tional Laboratory, June 1990-and W. Gunther, "An Operational Assessment of the Habcock & Wilcox and Combustion Engineer-80.
NUR EG/CR-5510, W. E. Vesely, R. E. Kurth, and ing Control Rod Drives" Brookhaven National S. M. Scalzo, " Evaluations of Core Melt Frequency Laboratory, September 1990.
Effects Due to Component Aging and Mainte-nance," Science Applications International Corpo-91.
NUREG/CR-5507, W. G unther and J. Thylor, "Re-ration, SAIC-89/1744, June 1990.
sults from the Nuclear Plant Aging Research Pro-gram: Their Use in Inspection Activities," Brook.
81.
NUREG/CR-5560, H. M. Hashemian, D. D.
haven National Laboratory, BNL-NUREG-52222, B everly, D. W. Mitchell, and K. M. Petersen, " Aging Septembe.* 1990.
of Nuc! car Plant Resistance 1bmperature Detec-92.
NUREG/CR-5314 C. E. Jaske and V. N. Shah tor " Analys d Measurement Semces Corpo-
"W Procedures for Major LWid Components," Vol. 3, " Cast Stainless Steel Compo-82.
EGG-SSRE-9017, C. L Atwocx!," User's Guide to E
2 2,
c 1
PHAZE, a Computer Program for Parametric Haz-ard Function Estimation," Idaho National Engi' 93.
NUREG/CR-5490 E. V. Werry, " Regulatory In-neering Laboratory, July 1990.
strument Review: Management of Aging of LWR Major Safety Related Components," Vol.1, Pacific
- 83. NUREG/CR-5280, M. Subudhi, W. Shier, and E.
Northwest Laboratory, PNL-7190, October 1990.
~
MacDougall," Age-Related Degradation of Westin-ghouse 480-Volt Circuit Breakers," Vol.1, " Aging 94.
NUREG/CR-5280, M. Subudhi, E. MacDougal, S.
Assessment and Recommendations for Improving Kochis, W. Wilhelm, and H. S. Lee, " Age-Related Breaker Reliability," B rookhaven National labora-Degradation of Westinghouse 480-Volt Circuit tory, BNL-NUREG-52178, July 1990.
Breakers," Vol. 2, " Mechanical Cycling of a DS-416 105 NUREG-1377
('
k Chronological Listing Ilreaker. 1bst Results," Brookhaven National 105. EGG-SSRE-9777, J. C. Watkins, R. Steele, J r., and laboratory, HNL-NUREG-52178, November K. G. DeWall, " Isolation Valve Assessment (IVA) 1990.
Software Version 3.10, User's Manual " Idaho Na-tional Engineering Laboratory, June 1991.
95.
K. R. Hoopingarner and E R. Zaloudek, " Safety Implications of Diesel Generator Aging." Pacific 106. NUREG-1144, " Nuclear Plant Aging Research; Northwest Laboratory, Nuc/ car Safety. 31:484-489, (NPAR) Program Plan, Status and Accomplish-October-December 1990.
ments," Revision 2. U.S. Nuclear Regulatory Com-mission, June 199L NUREG/CR-5558 R. Steele, Jr., K.G. D'eWall, and J. C. Watkins, " Generic Issue 87: Flexible 107. NUREG/CR-5706, D. A. Casada,"NRC Bulletin 8004: Potential Safety-Related Purnp Loss-An Wedge Gate Valve Test Program: Phase H Results Assessment of Industry Data," Oak Ridge National and Analysis " Idaho National Engineering Labora-tory EGO-2600, January 1991.
Lalmratory, ORNL-6671, June 1991.
108. NISTIR 4485, E D. Martzloff and A. G. Perrey.
97.
NUREG/CR-5555, W. Gunther and K. Sullivan,
" Annotated Bibliography: DiagnosticMethodsand
" Aging Assessment of the Westinghouse PWR Con-Measurement Approaches 1b Detect Incipient De-trol Rod Drive System." Brookhaven National fects Due to Aging of Cables," National 1nstitute of Lalmratory, BNL-NUREG-52232, March 1991.
Standards and 1bchnology, July 1991.
98.
N U REG /CR-5612, P. K. Samanta. W. E. Vescly, E 109. PNL-7823, A.D. Chockie, K.A. Hjorkelo, T.E.
Hsu, and M. Subudhi," Degradation Modeling with Fleming, W.H. Scott, and W.I. Enderlin, "Mainte-Applications to Aging and Maintenance Effective-nance Practices 1b Manage Aging: A Review of ness Evaluation," Unokhaven National Latmra, Several Technologies," Pacific Northwest Labora-tory, UNL-NUREG-52252, March 1991.
tory, October 1991.
99.
NUREG/CR-5619, S. P. Nowlen, "The Impact of 110. EGG-SSR E-9926, R. Steele, J r., J. C. Watkins, and Thermal Aging on the Flammability of Electnc Ca-K. G. DeWall, " Evaluation of EPRI Draft Report bles, Sandia National Laboratories, SAND 90-NP-7065-Review of NRC/INEL Gate Valve Test 2121, March 1991.
Program," Idaho National Engineering Laboratory, 100. PNL-7516, K. R. Iloopingarner, " Emergency Die-sel Generator Technical Specifications Study Re' 111. ORNL/NRC/LTR-91/25, D.A. Casada," Throttled sults," Pacific Northwest Laboratory, March 1991.
Valve Cavitation and Erosion," Oak Ridge 101. NUREG/CR-4302, H. D. Haynes. " Aging and Sciv-ice Wear of Check Valves Used in Engineered 112. PNL-SA-20219 D. P. Brown and D. E. Blahnik, Safety-Feature Systems of Nuclear Power Plants "
"ASME Subsection ISTD Recommendations Based Vol.2," Aging Assessments and Monitoring Method upon NPAR Snubber Aging Research Evaluations," Oak Ridge National Laboratory, Results," Pacific Northwest Laboratory, December April 1991.
- 1993, 102. NU REG /CR-5546, S. P. Nowlen, "An investigation 113. NUREG/CR-5M3, D. E. Blahnik, D. A. Casada, of the Effects of Thermal Aging on the Fire J. L Edson, D. L Fineman, W. E. Gumher, H. D.
Damageability of Electric Cables," Sandia National Haynes, K. R. Hoopingarner, M. J. Jacobus, D. B.
Laboratories, SAND 90-0696, May 1991.
Jarrell, R. C. Kryter, H. L Magelby, G. A. Murphy, and M. Subudhi," Insights Gained from Aging Re-103. NUREG/CR-5655, M. J. Jacobus and G. E Fuch-search," Brookhaven National Laboratory, BNL-rer " Submergence and High 1bmperature Steam NUREG-52323 March 1992.
1bsting of Class 1E Electrical Cabies," Sandia Na-tional Laboratories, SAND 90-2629 May 1991.
114. NUREG/CR-5762,J. E Gleason," Comprehensive Aging Assessmentof CircuitBreakersandRelays,"
104. HNL1bchnical Report A-3270 6-21-91, E Hsu, W.
Wyle Lateratories, WYLE 60101, March 1992.
E. Vescly, E. Grove, M. Subudhi, and P. K.
Samanta," Degradation Modeling: Extensions and 115. NISTIR 4787, E I. Mopsik, "The Use of Time-Applications." Hrookhaven National Lainratory, Domain Diclectric Spectroscopy 1b Evaluate the June 1991.
Lifetime of Nuclear Power Station Cabics," Na-I NUREG-1377 106 l
c
Chronological Listing tional Institute of Stamdards and 7b:hnology, April Study," Idaho National Engineering Laboratory, 1992.
EGG-2567, August 1992.
116. NUREG/CR-5807, J. K. Wang and M. S. Kalsi, 126. NUREG/CR-5587, W.E. Vesely, " Approaches for J
" Improvements in Motor Opented Gate Valve De-Agt-Dependent Probabilistic Safety Assessments sign and Prediction Models for Nuclear Power Plant with Emphasis on Prioritization and Sensitivity Stu-Systems," Kalsi Engineering, Inc., KEI No.1721, dies," Science Applications International Corpora-May 1992.
tion, SAIC-92/1137, August 1992.
I17. NUREG/CR-5870, D. P. Brown, E. V. Werry, and 127. NUREG/CR-5772, M.J. Jacobus, " Aging, Condi-D. E. Blahnik, "Results of LWR Snubber Aging tion Monitoring, and Loss-of-Coolant Accident 1
Research," Ibcific Northwest Laboratory, May (LOCA)'Ibsts of Class IE Electrical Cables " San-
- 1992, dia National Laboratories, SAND 91-1766/1, Vol-ume 1. August 1992.
i 118. NUREG/CR-5693.
R. Lofaro, W. Gunther, M. Subudhi, and H. Lee, " Aging Assessment of 128. NUREG/CR-6001, G.D. Buckley et al., " Aging Component Cooling Water Systems in Pressurized Assessment of HWR Standby Liquid Control Sys-Water Reactors-Phase II," Brookhaven National tems " Pacific Northwest Laboratory, PNI-8020, Laboratory, BNL-NUREG-52283, Jun,: 1992.
August 1992.
129. NUREG/CR-5379, D.H. Jarrell et al., " Nuclear 119. NUREG/CR-5720. R. Steele, Jr., J. C. Watkins, K.G.DeWall,and M. J. Russell," Motor-Operated P1 nt Service Water System Aging Degradation Valve Research Update," Idaho National Engineer-Assessment," lbeific Northwest Laboratory, ing Laboratory, EGG-2643, June 1992.
PLN-7916, Volume 2, October 1992.
130. N UREG/CR-5848, J.S. Dukelow, "Recordkeeping 120. NISTIR 4487, E D. Martzloff, E. Simmon, J. P.
Steiner, and R.J. Van Brunt," Detection ofIncipi-Needs 7b Mitigate the Impact of Aging Degrada-ent Defects in Cables by Ibrtial Discharge Signal hc"'>ber1 Analysis," National Institute of Standards and
~'
Tbchnology, July 1992.
131. NUREG/CR-5699, R.II. Greene, " Aging and Service Wear of Control Rod Drive Mechanisms for 121. NUREG/CR-4819, R. C. Kryter, " Aging and Serv-HWR Nuclear Plants," Oak Ridge National Labo-ice Wear of Solenoid-Operated Valves Used in Safety Systems of Nuclear Power Plants," Vol. 2:
ratory, ORNL-6666/VI, Volume l, Novcmber 1992.
" Evaluation of Monitoring Methods," Oak Ridge National bboratory, ORNLfrM-12038, July 1992.
132. NUREG/CR-5772, M.J. Jacobus, " Aging, Condi-tion Monitoring, and Loss-of-Coolant Accident (LOCA) Tests of Class IE Electrical Cables " San-122. NUREG/CR-5646, R. Steele, Jr., and M. E. Nitzel, dia National Laboratories, SAND 91-1776/2, Vol-
" Piping System Response During Iligh Level Simu.
ume 2, November 1992.
lated Seismic 7ests at the Heissdampfreaktor Facil-ity (SHAM Test Series)," Idaho National Engi-133. NUREG/CR-5772, M.J. Jacobus, " Aging, Condi-neering bboratory, EGG-2655, July 1992.
tion Monitoring, and Loss-of-Coolant Accident (LOCA) Tests of Class IE Electrical Cables," San.
123. NUREG/CR-5700, A. C. Gehl and E. W. Hagen, dia National Laboratories, SAND 91-1776/3, Vol.
" Aging Assessment of Reactor Instrumentation ume 3, November 1992.
and Protection Systems Components," Oak Ridge National Laboratory, O RNL/FM-11806, July 1992.
134. NUREG/CR-5783. E. Grove, W. Gunther, " Aging Assessment of the Combustion Engineering and 124. NUREG/CR-5779, J C. Moyers, Ning of Non-Babcock & Wilcox Control Rod Drives," Brookha.
Power-Cycle Heat Exc:iangers Used la Nuclear ven National Laboratory, BNieNUREG-52299, Power Plants," Vol.1. Oak Ridge NationaI Labora-
-January 1993.
tory, ORNL-6687/V1, July 1992.
135. EGG-SSRE-10039, T.H. Hunt, M.E. Nitzel, "An -
-)
125. NUREG/CR-5378, A.J. Wolford, C.L Atwood, Evaluation of the Effects of Valve Body Erosion on W.S. Roesener," Aging Data Analysis and Risk As-Motor-Operated Valve Operability," Idaho Nation-sessment-Development - and Demonstration al Engineering Laboratory, May 1993.
107 NUREG-1377
i Chronological Listing i
136. NUREG/CR-5404, J.D. Kucck, " Auxiliary Feedwa-139. NUREG/CR-5944, D.A. Casada and M.D. Todd, ter System Aging Phase Follow-on Study," Oak "A Characterization of Check Valve Degradation Ridge National Laboratory, ORN14566/V2, Vol-and Failure Experience in the Nuclear PowerIndus-ume 2, July 1993, try," Oak Ridge National Laboratory, ORNie6734, September 1993.
137. NUREG/CR4029, W.K. Winegardner, " Phase I 140. NUREG/CR-6043, D.E. Blahnik and R.F. Klein, Aging Assessment of Nuclear Air-Treatment Sys-
" Phase 1 Aging Assessment of Essential HVAC tem IIEPA Filters and Adsorbers," Volume 1, Pacif-Chillers Used in Nuclear Power Plants," Pacific ic Northwest Laboratory, PNL-8594, August 1993.
Northwest. Laboratory, PN1 8614, September 1993.
138. NUREG/CR-5754, K.H.
Luk, "Hoiling Water 141. NUREG/CR-6048, K.H.Luk, " Pressurized-Water Reactor Internals Aging Degradation Study, Phase Reactor Internals Aging Degradation Study, Phase 1," Oak Ridge National Laboratory, ORNL/mi-1," Oak Ridge National Laboratory, ORNiffM-11876, September 1993.
12371, September 1993.
i
-1 s
k i
NUREG-1377 108
f AC f 04M 335 U.S NUCtliAR RrGut. ATOR v' COMMISSION
- 1. HEPORI NUMBE R
(?-891 NHCM 1102.
(Ass #gned by NRC, Add Vol.,
Supp., Rev, and Addendum Num-W m2 BlBLIOGRAPHIC DATA SHEET ta"-
" aav-)
isr.e mtrucuone on ine reverse >
NUREG-1377, P, T H LE Af JU SUtsllill
- 3. DAll HE POH f PUUUSHLO NitC Research Program on Plant Aging: Listing and Summanes of McNm I
YEAR Reports Issued 'through September 1993 i
December 1993
- 4. FIN 04 GRANT NUMnER 5 A V I MUH W
- 6. T YPE OF HE POHI J. P. Vora Technical l
- 7. PERIOD COVERED (irmluswo Datos)
{
1981 - 1993 8 PL HI OHM!NG OHO ant / A HOh - NAME AND ADOHE SS (if NHC, prcrnoe Omston, Othee or Region. U S Huclear flegulatory Comm#ssion, arv.i rnaibng a& Fess; if cavitractor, provice name and rnaihng address. )
Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 205554N t SPONSOHING OHQAN:/AllON - NAME AND ADOHf.5 S (if f 4HC, type
- Lame as attve"; if contractof. prowde NHC Dmlion. Offte or Region.
D.
- o. B. Na tear Occuletory Commimon. and ma.hng address. )
Same as above
- 10. SUI'PL EMLN I AH V NO T LS
- 11. ACS f R ACT (200 words or lesa)
The U.S. Nuclear Regulatory Commission is conducting the Nucicar Plant Aging Research (NPAR) Program. This is a comprehensive hardware-onented engineering research program focused on tmderstanding the aging mechanisms of components and systems in nuclear power plants. The NiWR program also focuses on methods for simulating and i
monitoring the aging-related degradation of these components and systems. In addition, it provides recommendations J
for effective maintenance to manage aging and for the implementation of the research results in the regulatory pusess.
This document contains a listing and index of reports generated h the NIWR program that were issued through Sep-tember 1933 and sumrnaties of those reports. Each summary describes the elements of the research covered in the report and outlines the sigmficant results. For the convenience of the user, the reports are indexed by personal author, corporate author, and subject.
- 12. xt v wonos /orscniPToHs <t i,.t wwot or p,vam, inn,,ii assi.i researcners 3n iucanno ine r,pmt )
ta. AvAILADluTY STATEMENT Unlimited
" munmtAsmCArION Nuclear Plant Aging Research (NPAR) erhis r..)
aging mechanisms agmg mitigation Unclassified compilation crhi. unro life extension Unclassified plant aging 14 NUMBEH OF PAGE.S 16 Pf vCE t#tC FORM IM r?49)
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