NUREG-1144, Forwards Comments on Draft Rev 1 to NUREG-1144, Nuclear Plant Aging Research (Npar) Program Plan, Per J Vora Request.Draft NUREG-1144 Encl
| ML20238E585 | |
| Person / Time | |
|---|---|
| Issue date: | 04/27/1987 |
| From: | Novak T NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | Arlotto G NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20238E398 | List:
|
| References | |
| FOIA-87-465, RTR-NUREG-1144 NUDOCS 8709150085 | |
| Download: ML20238E585 (161) | |
Text
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!' a g NUCLEAR REGULATORY COMMISSION j g s t W ASHINGTON, D. C. 20$$5 I \.[...+ Apg n gy MEMORANDUM FOR: G. A. Arlotto, Director i Division of Engineering Safety l Office of Nuclear Regulatory Research l I
FROM: Thomas M. Novak, Director 4 Division of Safety Programs 1 Office for Analysis and Evaluation of Operational Data
SUBJECT:
DRAFT NPAR PROGRAM PLAN - NUREG-!!44 REVISION 1 (AEOD ACTION LOG f 87-46) In accordance with your request of March 22, 1987, we are providing the enclosed comments to Jit Yora (via copy). A marked up copy of the draft has also been attached which provides editorial comments. Based on the need for close coordination with the industry, we recomend that a letter be sent to the industry whose . contents are similar to the NRC's Technical Integration Review Group for Aging and Life Extension (TIRGALEX) memorandum to Eric S. Beckjord from L. C. Shao, April 6,1987. This memo presented the TIRGALEX plan which defined the major technical r,afety and regulatory policy issues associated with aging / life extension. We note that the draft NPAR program document is lacking information in many areas, such as electrical cabling issues; system / component selection; balance
; of plant components whose failure may challenge plant safety; and variances in similar products by different vendors. Although it may be difficult to include into a research program all of these and/or other such items which might affect plant life extension, they should be addressed in some manner as part of the overall life extension program. Also, we recognize your need for component aging data which the data bases accessible by the NRC currentl do not contain (e.g., adequater age tracking, stressor, and root cause data)y .
Therefore, an understanding of your data needs could possibly help AE00, through its NPRDS member role, ensure that your data needs are satisfied. In sumary, the enclosed coments: (1) request your response as to the NPAR program NPRDS data needs, and whether or not risk significant balance-et'- plant system / components will be evaluated; and (2) provides some recommendations for your consideration in future revisions to this draft. Mr. G. L. Plumlee III of our office has been assigned as AEOD's aging / life extension coordinator. Please direct any questions / responses regarding our comments to Mr. Plumlee at X24492. Fo1 A49-%c 6/1/ 8709150005 870911 PDR FOIA CORDON 87-465 PDR
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- 6. A. Arlotto As' discussed with J. Vora, we would appreciate a written response to our !
canments onthe role of NPRDS in the NPAR program as soon as possible in order to meet NPRDS Users Group schedules. Thomas M. Novak, Director j Division of Safety Programs Office for Analysis and Evaluation ; of Operational Data
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Division of Safety Programs . Office for. Analysis and Evaluation of Operational Data SUBJEh DRAFT NPAR PROGRAM PLAN . NUREG-1144, REVISION 1 (AEOD ACTION LOG f 87 46) In accordance with your request of March 22, 1987, we are providing the l aglosed comments to Jit Vora (via copy). A marked up copy of the draft has also been attached which provides editorial comments. In summary, the enclosed comments: (1) request your response as to the NPAR program NPRDS data needs, and whether or not risk significant balance.of. plant sy? tem / components will be evaluated; and (2) provides some recomen64tions for your consideration in future revisions to this draft. Mr. G. L. Plumlee III of our office has been assigtsed as AEOD's aging / life i extension coordinator. Please direct any questions / responses regarding our . comments to Mr. Plumlee at x24492. t As discussed with J. Vora, we would appreciate a written response to our
%v i coments onthe role of NPRG" ' the NFAR as soon as possible in order to meet NPRDS Users Group schedules.
Original signed M Dcmas M. Novak / s t'
- Thomas M. Novak, Director J Division of Safety Programs Office for Analysis and Evaluation of Operational Data e
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PEMORANDUM FOR: G. A. Arlotto, Director Division of Engineering Safety Office of Nuclear Regulatory Research FROM: Thomas M. Novak Director Division of Safety Programs Office for Analysis and Evaluation of Operatichal Data
SUBJECT:
DRAFT fiPAR PROGRAM PLAN . NUREG.1144, REVISION I (AEOD ACTION LOG # 87-46) In accordance with your request of March 22, 1987, we are providing the ecciosed comments to Jit Vora (via copy). A marked up copy of the draft has also been attached which provides editorial comments. Pr. G. L. Plumlee III of our office has been assigned as AE0D's aging / life extension coordinator. Please direct any questions / responses regarding our comments to Mr. Plumlee at x24492. As discussed with J. Vora, we would appreciate a written response to our coriments enthe role of NPRDS in the NPAR as soon as possible in order to ' meet NPRDS Users Group schedules. - Thomas M. Novak, Director Division of Safety Programs Office for Analysis and Evaluation of Operational Data
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7 0 t y. . Draft NPAR Program Plan - NUREG 1144 Revision 1 Comnents
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- 1. Throughout,this document, references are made to failure rate, in-service age, failure cause, or failure mode data and subsequent data compilation as it relates to aging (e.g., pages 21,28,30,37,A-4,andC-20). It
.specifically refers to data needs on the following pages:
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4 On page 37.- the'dr6ft states, in part, "One of the sources available l for failure data ... is obtained from ... (NPRDS)."
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- On page A-3, khe draft states, in p?rt, "The information ... is I derived from ..'. (NPRDS) ... is collected using failure-category l- and cause-codes ...." )
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On page A-4, the draft implies that failure information can be identified from.the NPRDS failure records, and On page C-20, the draft implies that component in-service age data has been/is being collected, and a d6ta base of aging-related
. failure data is being developed to provide aging root cause informa-tion for various systems.
An NPRDS subcommittee has recently been formed to evaluate and report to j the NPRDS Users Group current NPRDS user data needs and whether or not the current NPRDS provides these needs. Therefore, as a member of this i subcommittee, AEOD is interested in knowing whether or not the NPRDS is adequately meeting tne NPAR needs. If not, how can the NPRDS scope be revised to meet future NPAR needs (i.e., are the NPRDS cause codes, e.g., H-wear out; cause description codes, e.g., BD-aging / cyclic fatigue; and ; failure or cause narratives adequate)? Our trends and patterns statisti-
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cal studies specifically note a concern about whether or not many of the i component long lifetimes (both censored and not censored) are valid; i.e., have all failures been reported. Also the determination of component in-service age is not always possible due to the fact that incipient failures and subsequent replacement of the component are not NPRDS i
2-reportable. Therefore, our understanding of your data needs/ problems j L could possibly help AE00, through its NPRDS member role, to ensure that 1 your user needs are satisfied.
- 2. In Section 3.7.3, you reference NPAR's integration of infomation for. f microbiological 1y'inducedcorrosion(MIC). If not already incorporated, j
, you may want to add Fort St. Vrain's tendon corrosion mechanism experience to your data. The results of Public Service Company of Colorado's analysis indicate that tendon corrosion.was caused by microbial action on the tendon grease creating acetic and forzic acids, which in turn caused general corrosion and stress corrosion cracking.
l' 3. . AEOD's NPRDS Trends and Analysis Program, as referenced in Section D.1, ,
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does incorporate analyses of balance-of-plant (20P) components (nonsafety-related " key components"), e.g., main feedwater valves and pumps and condensate pumps for.PWRs. However, based on our review of this draft, q the NPAR program scope is limited to certain safety-related components shown to be risk significant. -In general the NPAR program appears to be
,; oriented toward. systems for accident mitigation. However, BOP systems j that cause challenges to plant safety systems should, to some extent. De evaluated for ' aging sensitivity. These systems may be more sensitive to aging degradation due to their lesser qualification requirements. j Therefore, some consideration might be given to the more important B0P 'l l
systems. Will the NPAR program scope be expanded to incorporate those BOP systems / components which operational experience has shown to be , major sources of unplanned reactor scrams?
- 4. Section A.2.1.1, on Page A-6, addressed the environmental effects considered in the NPAR program. The draft specifically addresses primary coolant chemistry. However, it appears that other coolant chemistry might also be considered in the NPAR program, e.g., secondary coolant chemistry effects on steam generators, or service water and component cooling water chemistry effects on safe shutdown components.
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- 5. Page 10 and 11.
- a. Item I should include erosion.
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- b. A separate item should be included to cover plant layup induced de-gradation. This would cover issues separate from storage such as the steam generator tube cracking at TMI-1 during several years of layup, and the piping stress corrosion cracking'at H. B. Robinson when the steam generators were replaced.
- 6. Page D-10 refers to two EPRI/ DOE initiated pilot studies involving Surry 1 (PWR)andMonticello(BWR). As we understand it, the aging / life extension data as defined in the EPRI Report HP-5002, " LWR Plant Life ,
Extension," January 1987, is not available for NRC inspection. It will not be possible for the NRC to authorize a license extension whose basis is founded on a pilot program's data, unless the NRC has access to validate this data. . 7. Obviously all the existing components in Safety Systems are not in the list of items to be considered (Tables 5.1, 5.2 and 5.3). However, the draft NUREG does not appear to adequately address the system / component selection process (e.g., to what level of detail should we go for each component? What was the basis for selection or deletion?). Normally one does not start by first utilizing PRA results (as implied in the draft). Systems / components are normally selected by engineering evaluations based on operating experience and engineering judgement. Then PRA results are i utilized to narrow the list down to a safety significant and manageable list.
- 8. It is not clear how this program can be successful, unless someone has first: ;
Listed af the components.
- Deleted components not affected by aging,
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*- Deleted components which are regularly replaced,
* - Deleted components which will be replaced as part of the " extension"-
request,
'*' - Deleted components for good reason (PRA), and
- Justified the remaining components (PRA, tests: and/or surveillance).
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Also. it is not clear how this can be done without operational data; not only failure rate, but also. time elapse to failure. The draft program does not clarify how/whether this data is being collected.
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- 9. It is not' clear as to whether or not EPRI and INPO have connented on-and/or_ helped in the system / component selection. Afterall, the final aging / life extension Justification will have to come from the industry. ,
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- 10. Once system / component selection is_ finalized and a realistic research program has been established, the program should then ensure that accelerated aging tests, synergistic effects, and real life experiences do not. provide erroneous results. However, the draft program plan does
. not indicate that as part of.a validation process, RES plans on evaluating the adequacy of EPRI's research.
- 11. Electrical cables are very important to safety and will significantly affect aging / life extension efforts. However, the draft's discussion in this area (Section C.S.1) does not adequately address the electrical
' cable _ aging issues. The draft does not address how much is currently known about cable aging nor does it address the diversity in cable materials, e.g., many plants have unique cabling with an assortment of subsequently added new cables; PVC (most old plants), armored (Duke I Power), asbestos (Indian Point 2 and 3), etc.
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- 12. Instrumentation and controls are also key safety components. However, i the draft program appears to imply that we will be depending on Shippingport data (prehistoric 180) to support life extension requests.
- 13. The draft program does not address how the vendor identification variable 1 will be handled, e.g., the quality input into component varies with the different vendors. Also the issue as the whether or not vendors will continue to support the nuclear industry (i.e., a minority group) should be addressed as part of a life extension program.
- 14. The Table of Contents would be more useful to identify specific areas if !
it was expanded to the next lower heading, e.g., 3.7.1, 4.1.2, or 5.2.1. l
- 15. All sections of the program plan that refer to the Office of Inspection and Enforcement should be changed to reflect the NRC reorganization.
- 16. Item number 6. " Enforcement Action Index," on Page 24, has now been deleted from the NRC's performance indicators.
- 17. SNL should be added to the list of national laboratories on Page C-1.
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. , NUREG-1144 i
} REVISION 1
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Nuclear Plant Aging Research(NPAR) Program Pan i U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research p"***%,
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,s NURIG-1144, Rev. ;
NUCIER PIANT AGING RESEARCH (NPAR). PROGRM PIAN _ FtR CDGQGNIS, SYSTEMS AND SMXTJRES
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p; c 2 sU REDORD USNRC's hardware oriented engineering research py-u for plant aging and dy.daticri monitoring of wir c .t.s and systems was first d4==M in
~the initial' version of NURErr-1144 innwt in July 1985. It was stated in the ry plan that the NUREG-1144 plan wtrdd be a living dersment and would be revised periodically. The revisions would be made to reflect experience gained in implementing the plan, to incorporate erunants received frm within the NRC, and frm industrial codes and standards ocamittees and domestic and foreign organizations and institutions.
Since issuing the' original psy.iu plan, the RES staff has' received numerous crunants freci various offices within the NRC as well as frm individuals, organizations and institutions outside NRC, both dcanestic and foreign. 'Ihe Otunnission in its 1986 Polit; and Planning Guidance document (NUREG-0885)~ prtwided planning guidance for needed safety research on plant .. aging and licanoe renewal. The Executive Director for Operations provided specific s wamu guidance to the staff for FY-1986-1988 planning and program development. . The NRC staff provided their oarmaants cm) the current researth piwtoru and needs for additional research and prioritization by " user-need" letters to RES and through the TIPGLIEX review of the NPAR program.
.a turing the past sixteen months, as a part of the overall phased approach to research, significant progress has been made in ocupleting the I
Phase I engineering research for selected myciwts and systems. They in::lude: notor operated valves, check valves, electric motors, emergency diesel generators, diargers and inverters, circuit breakers and relays, i batteries, auxiliary feedwater punps, and reactor protection systems. Also, progress has been made in developing models and approaches to evaluate rulative !==r+= of aging cui risk. The Phase I segment of research for the evaluaticn of systans level aging effects, fr m operating experience and risk evaluation of aging phoruisce, has been ocupletai. In consideration i of plant life extensicrylioense renewal, swi.es has been made in identifying major technical safety issues and defining major INR wr,ents 11
- and structures accordirg to their itq:ortance to plant safety. Also, a preliminary study has been ocmpleted identifying degradation sites and life ,
limiting pr-maa for each major Wwit. Finally, more has been learned from operating experience and frcan expert opinions. j
-Reflecting all of the aforementioned inputs, this heent ra.a a ^
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revised researdt plan. It addresses the' identification and a plan for the resolution of tatnical safety' issues relevant to plant aging and bases for license renewal. 'Diis plan focuses cm plant safety systems, electrical ard mechanical + - is, civil structures, and the utilization of technical-data in the regulatory process.
'this program plan for %A.ts, systems and civil structures in f conjunction with its sister plan for primary system pressure boundary l
-pr- ts form the overall framework for Nuclear Plant Aging Research '
l within the Division of Engineering Safety, Office of Nuclear Regulatory Paamatt:h of USNRC. I coments en this h==nt are welcxane and will be cxansidered in the development of e3 W ant editions of this plan. 'Ibey need not be restricted to the research activities described herein; ocenents identifying I canissions and/or r=0--adirs additional researdt are also welcxane. J1tendra P. Vora, Fivparu Manager I James E. Richardson, Chief ) Engineering Branch J ( l l A; proved by: Guy A. Arlotto, Director ( Division of Engineering Safety Office of Nuclear Regulatory Research lii l-l
.,/ IIST OF ACRONYMS AcRS Advisory ceramittee en Reactor safeguards AEoD Office for Analysis and Evaluation of Operational Data, U.S.
NRC~ ASPS Accident Sequence Precursor Study 4 IIS -Division of Engineering Safety EDO Executive Director of Operaticrl' HPCI Hi$1 Pressure Coolant Injection System (in BERs) IE Office of Inspection and Enforcement, U.S. NRC IPRDS In-Plant Reliability Data Systen IS and Im Inspection, Surveillance and (candition) Monitoring Methods NCAC Nuclear Operations Analysis Center at GINL NPAR Nuclear Plant Aging Researth NPRDS Nuclear Plant Reliability Data System NRC Nuclear Regulatory Ocumission NRR Office of Nuclear Reactor Regulation, U.S.' NRC i i NSAC Nuclear Safety Analysis Center operated by the Nuclear Industry 9_==ted Electric Power Research Institute (EPRI) NUMARC Nuclear Utility Management and Resources Ocunittee NUPIEX Nuclear Utility Plant Life Extension OL Operating License ORNL Oak Rid;pe National Iaboratory - OSRR Operational Safety Reliability Researth RCIC Reacto- Core Isolation Cooling RES Offica of Nuclear Regulatory Paaaavch RG Regulatory Guides SRP Standart1 Review Plan TIRGAIZX Technical Integration Review Group for Aging and Life Extension O l l iv l
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'I The reclear plant aging researd (NPAR) program described in this plan is intended to resolve technical safety issues related to the aging of'
' electrical and nochanical +=.ts, systems, . and civil structures used in ocenercial nuclear power plants. 'Jhe aging period of interest includes the normal licensed plant operation as well as the extended plant life, that may be requested in utility applications for license runawals.
Euphasis has been placed on identification and daracterization of the-mechanism of natorial and +-.t. degradation during service and the utilization of researd results in regulatory process. The reemarch includes the evaluation of methods of inspection, surveillance, condition monitoring and maintenance as means of mitigating aging effects dich may , I inpact safe plant operation. Specifically the goals of the program are as , follows: FKmaiu Cala A. To identify and daracterize aging effects Which, if unchecked, could cause degradation of + fd, systems and civil structures and
- therinby impair plant safety.
B. To identify methods of inspection, surveillance and monitring, and to evaluate residual life of -g-d.s, systems and civil stmetures, which will assure timely detection of significant aging effiscts prior to less of safety function. C. 'Ib evaluate the effectiveness of storage, maintenance, repair and replacement practices in mitigating the rate and extent of degradation caused by aging. The NPAR Wain is had on a phased approach to researd. She objectives of the Phase I studies are: to identify failure wrian ard causes attributable to aging; to identify measurable performance parameters y
includirg functional irdicators for potential use in a== sing operational .j 1 readiness of a -g-.t, a structure or a system, in establishing ) degradation U.w. ids, and in detecting incipient failures. !
'Ihe objectives of the Ihase II studies are to: perform in-depth engineering studies and aging e,ssessments haw on in situ measurements; ;
perform post-service examinations and tests of naturally aged / degraded
-p.rdas; and identify methods for ielen, surveillance and monitoring, or for evaluatinJ residual life, and to make zw%tions for the utilization of resear'dt results in the regulatory sucm.s. l
'Ibe objective of the pIrjected Phase III or the extended portion of romand is to provie's for the resolution of issues that may be raised i durirq the "results utilization efforts." !
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~ R CCNTDTPS FORDORD................................................................. ii-ACRCRTDS................................................................. iv k1GIRAC'T................................................................. V
- 1. INTR 0(IX'TICH . . . . * * . . . =** *** .....*....*=..** - *.......... 1' 1.1 nackground ..................................................... 2
-1.2 A Framewcak for Identifying ard Resolving Technical aafety T==== ........................................ 5 1.3 Organizaticr1 of the Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
- 2. TECHtTICAL SAITTY ISSOES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.1 Nature of Aging Pmm . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 2.2 Potential Impact of Aging on Safety . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 2.3 %chnical CLrjectives of the Ra==t:h . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 I
- 3. UTILIZATICH OF RESEMCH RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.1 License Renewal ................................................ 13 3.2 Generic Tamm--Unresolved Safety Issues . . . . . . . . . . . . . . . . . . . . . . . 20
, 3.3 Maintenance and Surveillance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 a
3.4 Plant Performance Indicators (Involving Aging Considerations) ... 24 3.5 Inspection ..................................................... 25 3.6 Codes and Standards ............................................ 26 3.7 NPAR Interfaces with Other Pw: aa6............................. 26 3.8 Brief Synopsis of NPAR Results In Support of the Regulatory Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31
- 4. RESFARCH APF9CWH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37 4.1 Risk and Systen Oriented Identification of Aging Effects ....... 37 4.2 Phased Approads to Aging Assessment and In-Depth Engineering Studies ................................... .
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- 5. NPAR PROGRAM EESCRUTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 5.1 Cuyaients, Systems, Struct22res Studied in the NPAR h-@ tan . . . 41 5.2 h wi -u El ements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 46 6.- COORDINATICH WITH 01HER PROGRAMS, INSITIUI' IONS AND CRGANIZATICNS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49
- 7. SOEDUIES AND RESCURCE REWIREMENIS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51 APPENDIX A-NPAR Strategy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 APPENDIX B-99JCR NPAR hwimu Elements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 APPENDIX C--NPAR Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-1 APPENDIX D-H3ngoing h@tais and Activities Related to NPAR . . . . . . . . . . . . . . D-1 l APPENDIX E--Milestones and Schedules . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . E-1 RETERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . R-1 FIGURES 1.1 NPAR Coordination and 'Ibchnical Integration . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.1 NPAR Panaa mh Approach . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 TABLES
,. 2.1 Sarne Exanples of Nuclear Plant Aging and
' Life Extension 'Inchnical Safety Tas . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 3.1 NRR Generic Issues that would dizactly benefit from the NPAR h wsma Results ................................................ 22 3.2 Generic Safety Issues that would indirectly benefit frcan the NPAR h ws -u Results . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.3 ' Develop Recamandations to Revise IEEE Stan$ards for Electrical T4 ==9t for Nuclear Power Plants . . . . . . . . . . . . . . . . . . . 34 3.4 Develop Reconsnandaticris to Revise ASME Standards for Operation and Maintenance of Machanical ET ,==9t 4
............... 36 5.1 Cu+cier.ts of Current Interest in the NPAR EWi-u .................. 42 5.2 Systes of Q1rrent Interest in the NPAR R@t mh . . . . . . . . . . . . . . . . . . . . . 43
'5.3 Major IHR Plant Elements of current Interest in the NPAR Program .... 44 L 6.1 Selected h@imis Relevant to NPAR Agirg and Life Extnsion Programs. 50 l
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- 1. INIRODUCTICri Since the early 1980s it has boome clear that the current vintage of camercial ruclear power plants has gone beyond the development stage and is reaching a stage of relative maturity. 'Ihe prototype reactors of the late 1950s and early 1960s have led to the develognent of two types of camercial light water reactors (IKAs), the pressurized water reactor (WR) and the boiling water reactor (BWR). 'Ihe United States now has approximately 100 reactors in m ...r.rcial operation. A few of these rmar+rws have been operating for over 20 years. As the population of IMs nature and advance in age, the need for a raaamd piwscuu which would provide a systematic !'
assessment of the effects of plant aging on safety was recognized. In his wmata on the Ieng Range Pacaamh Plan, the Director of the Office of Nuclear Reactor Regulation, USNRC, identified a need for a research program j to irrastigate the safety aspects of aging p _- ccac in ommercial nuclear ,
)
power plants. Initiation of an aging research program was aira re-sded by the ACRS in their 1983 report to Careiass.
'Ibe m=1== ion in its Policy and Planning Guidance document (NURD3-0885) provided guidance for needed safety researth on plant aging and license renewal. Also, the Executive Director for Operations has provided specific gwscuu guidance to the staff for fiscal year-1986-1988 planning ,
and program develeg a nt period. f
'Ihe USNRC Office of Nuclear Regulatory R*=aad (RES) has developed and implemented a hardware oriented engineering research program for plant aging and degradatico monitoring of myeredits and systems. '!his program is called the Nuclear Plant Aging Research (NPAR) FiWm.= and was first described in the July 1985 issue of NURD3-1144 (Reference 1) and discussed at length, at the July 1985 International Ctnference en Nuclear Plant Aging, Availability Factor and Reliability Analysis (Reference 2). 'Ihis report describes the NPAR Fr% sam for wycEe.rfa, systems and civil structures.
'Ibe NPAR p@iom is being conducted by the Engineering Branch of the i Division of Engineering Safety (EB/IES). A similar program on aging khich 1
i l _ _-___ a
is fm1W on vessels, piping, steam generators and non destructive examination techniques is being ocriducted by the Materials Branch of the Division of Engineering Safety (MB/IES). 'Ihe yawiaiu plan developed by MB/ DES is a sistar plan to this NPAR plan. 'Ihe two plans form the overall
. L- . ark for aging researd within the Division of Engineering Safety, Office of Nuclear Regulatory Researth of US NRC.
Significant progress has Lwn made since the issuance of the original l program plan. 'Ihe Phase I engirnering research has been ocupleted for selected w ud.s and systems. 'Ihey include: motor operatad valves, check valves, atM14=7 feedwater punos, amergency diesel generators, electric motors, batteries, chargers and inverters, ard circuit breakers and relays in safety related systems and reactor protection systems. Also, en sita a==amaments of electrical cirulits have been performed arxl aged
-g isits and materials are being retrieved from the Shippingport Atcznic Power Staticx1. Fr% -s also has been made in developing Fv4ala and ,
approaches to evalusta relative inpacts of aging on risk. 'Ihe main objective of this haant is to revise the original research plan by ira.ur
- r. rating What has been learned frce the NPAR pawiaiu activities and frun the ocuments received frtan the various industry ard government institutions and organizations,nie=aatic and foreign.
'Ihe revised NPAR pr%iam plan describes the raamarch effort currently i being implemented to resolve the technical safety issues relevant to plant aging and operating license renewal, ard describes the utilization of the technical data in the regulatory process.
1.1 Backmun:1 ard Need l Aging affects all reactor structures, systems ard W-uds to various .q degrees. For the NPAR pr%iaiu, aging refers to the cunulative degradation I of a system, + wit or structure which occuls with time and Which, if ) unchecked, can lead to an inrairmant of continuing safe operation of a nuclear pcuer plant as it advances in age. It is r===_q to take naaanres 1 to assure that age related degradation does not reduce the operational j 2 1
readiness of a plant's safety systems w-tus and structures ard does not result' in tvunnn mode failures of redundant, kut aged safety-related a gli M . It is also tw=amary to assure that aging does not lead to an undetected reduction of the defense-in-depth features of a comercial power reactor which provide for a series of nultiple barriers against the potential release of fission products to the public. 7b establish a pru.Wdve for describing the NPAR program, it is of ! interest to avamine the current status of h.arcial operating nuclear power i plants. As of January 1987, there were 99 licensed ocamarcial power plants in operaticm in the U.S. The age distribution of these plants is listed below. nr=ratim Lifetime (Years since Ooeratim License) No. of Flants o greater than 20 years 2 , o between 15 and 20 years 13 between 10 and 15 years o 45 o between 5 and 10 years 12 o less than 5 years 27 a The two oldest operating plants, Yankee Rowe (OL Date - 7/9/60) and Big Rock Point (OL Date- 8/30/62) have been in operation for 26 and 24 years rev A ively. However, they are demcmstration plants with design power of less than 200 Mk. The next oldest plant is San onofre 1 (OL Date- 3/27/67) has a net capacity of 430 Nie. In addition to the plants currently in operation, there are approximately 20 sore plants under construction. Most of these are expected to be in operation within the next darwk. As the population of U.S. lic,$ht water reactors has natured, problems j have already occurred which are the result of time dependent degradation l mechanisms such as stress corrosion, thermal aging, radiation embrittlement, { 3 l
J fatigue and erosion. 1hese problems have inclu$ed failures in valves and relays, embrittlement of cable insulation ard m ?urs, ard cracking of the heat treated an6or heads for post-tensioning systems in containment. Although rw .es is being ande to mitigate the age related dag detion that has already been identified, significant questions still remin hamma of the variety of ber a.s in a ocenercial power reactor, the cx:mplexity of ] the aging process and the limited experience with prolcnged operation of j these power plants. The NPAR hws m has been developed to provide a systematic researth j effort into how aging affects the safety of the plants currently in operation. This program provides a wWeissive effort to: learn frm operating experience and apart opinion; identify failures due to age degradation; foresee or predict safety problems resulting frm age related degradation; and develop r- - =dations for surveillance and maintenm::e ptMares Wiich will alleviate aging concerns. . The aging program alma provides key information to enable Mac to resolve technical safety 4m=== and to define its policy and regulatory positico on plant life extensien and license renewal. License renewal in this dew refers to the runewal of an operating license. Under the j
.; current regulations, reactors are licensed for up to 40 years of operation.
Current regulations also permit license renewal. The TIRGAIEX develcped a l l working definit.icn for life extension. Life Extensicn is defined to include license renewal beycmd the original license term of 40 years and a rwtam l for systenatic hardware renewal of plant systems, equipment, mirc As and structures. i Utilities are currently planning to apply for license renewals and they have defined a tantative whedule for several key staps in the p 4 'No j representative 1 hrs have been the subject of an EPRI/ DOE / utility sponsored k pilot study cm plant life extansicn (Reference 7). The two plants, that are the subject of this project, are Monticello, a 545 M4e BWR (OL date 9/08/70) { and Surry 1, a 788 M4e PWR (OL date 5/25/72). The purpose of this project is to provide an initial evaluation of the offacts of aging on cxmercial f 4 l
.........____.__m._. . . .. .
i
p :. nuclear plants and to establish the scope of the effort that will be needed ] l to' extend the operating lifetime of these plants beyond their initial 40 I years of licensed operation. The first suknittal to the NRC is expected in during the 1991-1993 period. A large number of additional suknittals for license renewal can be expected shortly thereafter. 2b keep pace with these industry plans and prepare for the large ramber of sunnittals, the NRC will need to devote substantial efforts over the next several years to define the j requirements for license renewal. The first license for a large plant (>400 Nie) will not expire until about the year 2007 (a===niry the license tern is defined fran OL issue data). However, the utilities need to decide between j requesting a plant license renewal or planning new generating capacity apprAtely 10-15 years in advance of the and of the licensed period. This is to allow for the long lead times required for planning and w a.aoction. This will require a firm NRC policy w rrdng the terms and conditions for license renewal to be in place by early 1990. Raamd on this , guidance, the first applications for license renewal can then be prepared and sutnittad to NRC. Review of these applications at this early stage will provide indication of the viability of the life-extension option in sufficient time (well in advance of the license expiration date) for a utility to elect an alternative option, if maary. Thus, NRC needs to clearly define its policy and regulatory positions in the near future to assure the safe operation of aged plants during the currunt license period and for extended life. Clearly defined policies and i criteria are needed to assure that requests for license renewal address the ] primary regulatory concerns and i====. j 1.2 - A F1 ark for Identifyin: and Resolviner Technical Safety Issues l A technical integration review gro.1p was established in 1986 to facilitate the planning and integration of NRC plant aging and license 1.rn _1/ life extension activities. This group was established by the Dcecutive Director for Operations (EDO) and is called the hinical l l Integration Review Group for Aging and Life Detension (TIRGAIEX). The initial objectives of TIRGAIEX has been to clearly define the technical ( 5 l l
i- i i safety ard regulatory policy issues aa w iated with plant aging and life extension and to develop a plan for resolving the 4==== in a timely, well integrated and effective manner. I Figure 1.1 shows the ' framework recarmended by TIRi41EX, and adopted in the NPAR Ftws u, for the planning and integration of activities related to plant aging and license renewal / life extension. As can be seen en the left i side of the figure, . technical informatim on aging and licanes renewal 2 l already being developed by a variety of sources. 'Ihis infcontion is I ocupiled and it will be periMirally updated by RES to assure.that all NRC offices involved in aging and license renewal have current information on f ongoing related efforts. Using the TIRGAIZX Integration Plan, the technical data currently being developed in related projects and the regulatory user needs, identified by ' i IE and NRR, are the key inputs used to establish'the priority of the research rws m elenants. RES then has the responsibility fc'r carrying out
, the nar===aq researd programs.
Hardware oriented engineering research naadad to resolve the issues related to aging is being ocriducted in the Division of Engineering Safety
! (DES). Two swi.- are being conducted in IES. '1he NPAR F4wtoru for
- risord;s, systems and structures is being performed by the Engineering Branch of IES. 'Jhe aging research piwtain on the vessels, piping, steam 9.r tator and rc.ies-Lative examination techniquas is being performed by the Materials Branch of IES.
As the principal technical elements in these research programs are ocupleted, the data and information is made available for use in the l regulatory process. RES will also make use of researd findings as they impact the RES responsibility for developing regulatory critaria, ' guides and stardards and ruview r e ares. I 6 l
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y 1.3 organization of the NPAR Plan Section 2 ocritains a die == ion of the technical safety issues related I to aging and safe operation of plants of all ages. 'Ihe nature of aging s c-r- rr are dian==ad first. 'Ihis is followed by a diam == ion of the potential impact of. aging cui safety. ard the technical tbjectives of the l l I
, researd) ocnsidered in the NPAR rwt.ie.
Section 3 contains an overview of the utilization of the NPAR swtam. results.in regulatory process. '!he di===ien of utilization is divided into seven categories: . License Renewal, Generic Issues, Maintenance and q surveillance, Plant Perfemmance, Inspection of Safety Systems and cu p -rd.s, codes and standards, and other Pt %s 6.- Also, a brief synopsis of the utilization of researd results in support of the regulatory process is included. -
]
. 1 Section 4 contains an outline of the systematic approns used in NPAR sws.un for amma==ing the effects of aging on plant safety systems, w w .uts and structures. 'Ihe critaria used to identify systems, ww rits and structures important to safety are dian=amd; as is the I phased approach developed to study the effects of age related degradation.
a Section 5.contains the description of major program elements and the scope of work for the subjects related to the systaas, + ta and structures included in the NPAR research paws.uu. i I l section 6 contains a description of the program coordination ard technical integration perforned within the NRC, other go% A. agencies, and with external instituticris and organizaticos, dcznestic and foreign. l Section 7 contains a d4=== ion of the schedules developed for the various NPAR activities. 'Ihe schedules include Nd activities in consideration of license extension efforts to be crupleted by early 1990s 1 and the continuation of age related confirmatory researdi. J 8 l
1 j r , Appendix A contains a description of the NPAR strategy and a phased approacts used in' conducting researtfl. Included also, are the method used for the initial selection of. systems and m w e h for aging studies and the devel+ d. of application guidelines and recommendations. A l i Appendix B ocritains a description of the major raaaatt:h areas being addmssed by the NPAR Fxwiwn. Appendix C ocntains a description of the researtti activities being 1 performed as part of the NPAR program. 'Ihe scope and current status of each of the major projects is dim =M. Appendix D contains an overview of other cngoing paws.n related to aging and life extension. 'Ihe coordinatico required between the NPAR effort and other engeing activities is dia,*=ami, with emphasis on the need to , cp imite the use of available resources. Appendix E ocntains the milestones and schedules for the various r==aatt:h activities in the NPAR pawiam. 2 Finally, the list of the references, cited in the prwiam plan arx:1 the five Appendices, is presented. l 1 9
- 2. TEOCUCAL SATEIY ISSUES A broad set of technical safety issues has been developed to provide q focus ard direction for the NPAR program. 'Ihese issues are based on operating experience, expert judgement, and risk significance. 'Ihese technical issues include the questions that need to be arswered, the problems that need to be solved, and the measures that must be taken to assure that safety levels are maintained as the present generation of reactors age. 'Ibe technical safety issues will be developed further and prioritized by first examining the nature of the aging pwr ss and then examining the potential role agirg plays in plant safety and the agency's mission to address plant agirg ard life extension / license renewal.
'Ihe specific technical objectives of the research plupcuu have been developed to address this broad set of technical safety 4*===. 'Ibe program technical objectives and the technical issues then provide the framewrk required for the development and guidare of the individual research ,
projects in the program. 2.1 Nature of Acina Processes 02nmercial nuclear power plants are large engineered complexes aid are . cmprised of many different systems, caponents ard structures which cover a j broad spectrum of materials and designs. 'Ibey operate in a variety of different environments ard falst meet different functional requirements. 'Ihe various cuwwnts, systems and structures are inspected and maintained by a variety of methods ard general approaches. Consequently there are a number of factors that can cause degradation of the functional capability of a cmponent, system or structure. 'Ibey include:
- 1. Material degradation mechanisms are active durirq operation ard storage. Exa::ples of typical causes of degradation mechanisms irx::1ude: neutron embrittlement, fatigue, corrosion, oxidation, thermal embrittlement, ard chemical reactions.
10 j
j
- 2. Stressors can be intMM by the storage, operating envirement, or external envisu w t. Irradiation, primary and secondary coolant chemistry, and vibratory loads are the typical examples of stressors introduced by the operating envitu.ur.nt. Freezing and thawing, brackish water, ard humidity are typical exanples of stressors introduced by external envisu =u t. Synergistic influence of electrical and mechanical stressors in ocabination ;
with other internal ard external envisu_,t also contribute to di.p.".ation processes.
- 3. Service wear. Am=0ation of fatigue damage due to plant operational cycling, service wear of rotating equtipment and wear of the drive red analy in a control rod drive nachanism are typical exanples.
- 4. Decessive tasting. Frequent testing of emergency diesel ,
generators is a typical example,
- 5. Iwtw installation, application, or maintenance. Investigation by HRC (Reference 32) has indicated that thirty percent of the nuclear plant abnormal occurrences can be attributed to faulty and
; inproper maintenance.
'Ihese factors, and others, can act with time either singly or synergistically to degrade a +->t, system or structure. !
"Agincf' is defined in this report as the ountlative degradation which .
m,ns with the passage of time in a u w nt, system or structure. 'Ihis degradation takes place he=n== of one or more of the factors listad abcVe.
'Ihis degradation can, if unchedkod, lead to a loss of function and an j impairment of safety. Aging is a couplex pv-m that begins as soon as a i u.mp-it or structure is praW and continues throughout its service life. Aging plays a significant role in the operation of a nuclear plant and must be factored into the determination of safe operating lifetime 11 1
l
i () I 1 [ Llimits. 'It'is also', important in the evaluation for license renewal. No nuclear plant,-inc M ing those still under construction or bsing. i noth-balled, should be' considered imaIne 'franjits effects. '
.i 1
2.2 Potentiar T=act of Aaim en sagggy jj i i 0
'Ihe main cancern addr====4 by this researd) yivviou is that: plant A :
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=
safety oculd be wuvi - 4w if degradation of key +->ts, systeim or y E structures is not detected prior to a less of functional capability and' , timely w .. dive action is not taken. In this way aging can result in an-i 4 undetected reduction in the defense-in-depth ww yt. 'Ibe defense-in-depth
, w yi requires that the public be protected from the accidental release of o fission products by a series of multiple barriers and engineered safety systans.- ,y
.x Age degradation of the reactor u.mycrets, g4==nt and structures can . -
rMrw the overall level of' safety. Experience at operating power plants
, provides exanples Where age degradation of vital v.mycieus could laad to a less 'of the margins provided by the defense-in-depth m,.yt. These exanples include items such as failure of emergency diesel generators, .
de.godation of valves and stress corrosion cracking of heat treated antor
- heads in .uu n La.ssed concrete containments.
c \ In the case of the major up d.s, age degradation zust be evaluated wban uensidering plant life extension and license renewal. 'Ibe major ! u.mycs As are the large, expensive, permanent parts of the reactor system i which are not routinely replaond or refurbished. Age degradation must be anaamasa and an evaluation of the residual life of the major wpsts is required if plant safety is to be assured during extended life operation. Age degradation can also cause a loss of operational readiness in engineered safety systems. 'Ihese are the systems that are required to mitigate the ocnsequences of a failure of a vital u wst; sud) as an 12 ___-----_..___m_ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
h.- r. . s i e ae=twel break in the primary system boundary. Examples of saiecy system are the amargency core cooling system, the reactor probacticn system, and
-the containment spray systam.
OE A survey (Reference 3). of IERs conducted by CRNL, as part of the planning for this' aging research plan, shows that numerous instances of
' aging induced failures of M'ir==nt have been m iad. 'Ihe reported events indicate that essentially all types of safety-related systans bsve been affected by a variety of G. htien processes. Also, GtNL described the background of selected age related IERs, in more detail to proride a better perspective regartiing the safety significance of age d. ht.icn -
(Reference 4) . Based on these studies, aging effects can contritute to
- n. both: 1) the probability of initiation of transients and arv-ir%ts, and l' 2) the probability of failure of the mitigating S'4 =arit ! during operation.
l . . , Aging can also lead to a higher probability of ocmimon mode failures. - This is the area of great concern. Aging can lead to wide meaTe or sindtaneous degradation of one or more of tlw physical barriers which prevent the release of radioactivity into the envi.s . A. In adctition, age related degradation can occur in redundant aq'ir==nt and safety systems intended to mitigate the effects of transients and accidents. If this degradatico is undetmeted, a trigger event such as an operational transient or accident cxuld be followed by sindtaneous failures of rwhrdant safety systems and sequential failures of back up systems and aq'ir==nt. An exanple of this, is the degradation of the insulation in the safety cables within the containment. In the case of a seismic event, there could be a siruitaneous loss of redundant control systems intended to sinrt down
. the reactor and maintain core cooling. Both, seismic events and fires (a second likely contributor to sucts events) rave been the subjects of research pmgoms to devalep better data and methodologies to reduce the uncertainty in the risk associated with these events.
13
Another potential cause of ocrimon rode failures is the siruitaneous failures of redurdant safety equipnant or systens in harsh envitumnts.
'Ihis can result frm a loss of coolant accident or from a high energy stem line break accident. Prototype electrical equignent has been testad to
<t,w. Late that it will function in these accident envitumd.s. Also, the K N equipnent usedhscrne of tLese tests has been artificially aged j to simulate service degradation. However, there is scrae doubt that such j techniques realistically represent the effects of in-service degradation. !
Ftxr example, it is known that accelerated radiation aging at the high dose rates, typically ecployed by ccanmercial testing laboratories, does not prcduce the degree of embrittlement of cables as may be caused by radiation at the actual dose rate encountered inside containment during operation. Also, the polymeric materials, used in certain types of solenoid valves, have been observed to _h' more vulnerable to failure under IDCA cmditions; due to natural eging rather than artificial aging. Because of the evidence that artificial or accelerated aging technigaes may be inadequate, at this time it is difficult to a- the increased degree of vulnerability of safety equipnent. This equipment, degraded by age related service and wear, ray be vulnerable to cx2rrn m:de failure during accidents ard transients which involve abnormal stresses and chaands on the equipnent.
- 2.3 Technical Obiectives of the Research 7he NRC has the responsibility for assurirg that licensed reactors can continue to be safely operated during their initial licensed lifetime and for the potential for exterded life. Because of the cocplexity of age related degradation and the diversity of the degradation g&m, a coordinated researth program is remrr to
- (1) identify the measures which are available to mitigate age related degradation; ard (2) identify anticipated prtblems which may result frcan plant aging. Using these two general critegia
+&
as guidance, a set of technical safety ins *e have baan K developed for3 NPAR prcgram. 'Ibese are listed in Table 2.1. The safety issues form the basis for establishing the technical cbjectives of the NPAR Program. 14
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TABLE 2.1. 4GE4MhMFHB NOCIEAR PIANI' AGING AND 1HE DCENSICH X
'IB3NICAL smell ISSUES o What structures, systems, ard ww-ta are sunomptible to aging effects that could adversely affect public health and safety?
Whim of thane structures, systans and ocspcr. sits are maintained , and are replaonable? o What are the degradation processes of materials, w-ia and structures ietich could, if unchecked, . (iewly maintained an4/or not replaond) affect safety during nootnal design life and during extended life? o How can operational readiness of aged structures, systems, and 3 mig is be assured during 40 year design life ard during extended life, o Are currently available examinaticrts and test nothods adequate to identify all relevant aging mechanisms before safety is affected? If n:rt, idiat efforts are underway to improve them?
~
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o Ehat critaria are required to evalteta residual life of myor,ents J and structures? Wat supporting evidence (data, analyses, ! inspections,.etc.) will be needed? o How should structures, systems, ard my.Sta be selected for comprehensive aging assessments and residual life evaluations? lelich structures, systems, and myrs-/J should be selected?
- o How effective are current gwme for mitigating aging (e.g.,
maintenance, replacanent, and repair)? o What kinds of reliability assurance ard maintenance gws-s will be needed to assure operational readiness of aged safety systems ard mycr a.s? o What cdditional charges will be remim9 in codes and standards to address aging? What schedule should be followed? I l l l I l i i 15 l l 1 i
me technical objectives of the ruclear plant aging research (NPAR) r wianiare: A. Ib identify and characterize aging effects which, if unchecked, oculd cause degradation of structures, mp As, and systems and thereby inpair plant safety. B. Tb identify methods of inspection, surveillance and monitoring; and to evaluate residual life of mp-tas, systems, and civil structures, which will assure timely detection of significant aging effects prior to loss of safety function. C. Ib evaluate the effectiveness of storage, maintenance, repair and replacement practices in mitigating the rate and extent of degradaticx1 caused by aging. The aging raaammh rwtasi has been developed to meet these objectives. The ptwiaru involves: (a) risk orinnted identification and selection of ocroponents, systens or structures for Which ====Semants of the inpact of aging on safety performance are to be o@M; (b) review of dasign base safety nargins, qualification tasting, operating experience, and methods for
- surveillance, inspection, monitoring and maintenance, leading to the development of re---- -rdations for irm$epth engineering studies; (c) engineering studies including verification of inspection, surveillan ,
3cnitorirg and maintenance methods, evaluation of residual life nod .!4, irt-situ examinations, collection of data 1; tun operating equipment, and cost / benefit analyses. The swiaru developed to meet the above objectives includes a variety ofprojects. Because of the multi-disciplinary scope of the researth ard need to make the best use of the available resources, the research effort is fem _c_ad on key ww-is ard structures in the systems of risk significance. The priority of the research effort has been established by taking into account: (1) information gained frcxn the 3-day workshop that was attended by over 300 pecple from the U.S. and other countries representing a 16 j
- i wide spectrum of interests and expertise (References 5, 6); (2) information I' gained frun the EPRI/ DOE workshop on plant life extension held in 1986 in Alexandria, Virginia in which results were presented frun the pilot projects at Surry-1 and Mcriticello (Reference 7); (3) insights gained frun the risk amaaammants empleted to date (References 8, 9); (4) advice frun a cross section of knowle& Jeable people; and (5) 21 ant operating experience, includirg Licensee Event Reports and INPO's Nuclear Plant Reliability Data System (References 3, 10). Other on-goirg NRC pas ==, industry spor.sored )
i researt:h, and swi-= belfg ccnducted in foreign countries are also ' considered in dsNelcping the r 4 rom plan. In those maan where relevant information is available or is being developed, the NPAR research has been planned to avoid duplication of effort. l 1 I l i l i l l [
)
l \ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
i-
- 3. UTILIZATION OF RESEARO{ RESULTS l
'lhe NPAR Fty-u aims are to obtain a better uid=t.'W of the aging and degradation pr-=== ard provide improved ocnfidence in available methods for detecting and mitigating aging degradatico. 1his Fawyt-. will J provide a basis for timely and sound regulatory decisiens regarding !
continued safe operation of ranclear plants of all ages as well as far the anticipated requests for license runawals. Detaction and mitigation of degradation damage at an early stage, before functional capability is impaired and before continued safe operation hammam cpestionable, will avoid urplanned and costly plant shutdowns. Also, use of researti results will-make operating plant maintenance more effective. Wear frtzn evn===ive tasting can be minimized through the use of more effective surveillance tediniques and result in the improved reliability of v4M. In addition to the general benefits mentioned above, the NPAR Fawg.a ,
=
is structured to respond to the follcwing specifie user-oriented needs:-
- 1. Develop data for identifying and resolving technical safety issues related to plant aging and license renewal,
; 2. Support NRR in resolving generic safety issues involving aged plant safety systams, support systems and electrical and mechanical - p eats,
- 3. Evaluate and recomend surveillance and maintenance methods needed to nonitor age related degradation and to support license renewal,
- 4. Develop technical data and provide r+ - =4ations useful far the clevelopment of plant performance irdicators (useful to II for plant inspections and to NRR for review of applications for extended license regmsts),
- 5. Provide information for the development of inservice ire _ ion sucedures suitable for aged wpents, systems and structures, 18 l
)
O, :'.
~6. Develop recaneandations for revisicais of app:rpriate indstry'
- 7. Develop tactinical' data useful to RES for the CRSS program and for I
NRR to evaluate the status of " mothballed" v4M. l 1
'the following subsections contain a brief description of the NRC staff
's defined needs and how the results of the NPAR program can be utilized in the regulatory process.
3.1 License Ranswal
'the NRC needs to clearly define its policy and regulatory positions in the near future so that utility planning for plant life eransion can ]
r M in an orderly manner.. q l
. :i
'the extension of. the period for a nuclear power plant license is
' p:xwided for in Part 50, para va $ 50.51' of the Code of Federal Regulations. 'this parays.$ states that a license is. issued for a fixed f l
period of time, not to exceed 40 years frun the date of. issuance. . It also
'l states that " Licenses may be renewed by the n=n4==ian upon the expiration - l
. . of the period." Alth:: ugh specific plant requirements for a license renewal -{
. are not specifically defined, it is clear that the " aged" cordition of the plant will have to be considered in any utility request for a license j i
,enewal. ' A pressing need for the NRC at the prenant time is to develop ]
guidance far irdustry on regalatory policy and r.- bres for the life extension. By 1991, utilities will need to know NRC requirements for license renewal rh. 'than, based gen the resolution of the tec:hnical safety issues, RES will need to develop appropriate regulatory guides and review r M ares by 1992-1993 for NRR to use in the evaluation of license renewal applications.
. the NPAR ectivities include the review of Sections 3-10 of the Standard Review Plan and associated guidance to. identify technical issues that may need to te addressed for license renewal. Technical specifications are
'inportant fran the standpoint of license to require methods for )(
19 l I I
l
'i l
Ldetecting' aging ard controlling it. 'Ihe NPAR Frw&on, will assre that this j aging / license renewal perWive is factored into in its ongoing programs and activities.. S
)(' 'Ihe NPAR Pawi. devalgand integratas the vast amount of agirg related knowledge so that the te&nical safety issues ===~4= tad with license runswal are identified and resolved in an effecti'se ard timely manner. We plan to aoocglish this integration by maintaining, evaluating, and updating the state-of-the-art information obtained frtan ongoing rw os
.related to agirg and license rensual. These programs are sponoceed by NRC, industry and' foreign organizations.. Fa%i-u ocordinaticri and te&nical integration are important program elements and are required to address the couplexity and diversity of the 4 dation pre ===== having a significant impact en plant safety. We have prioritized selected conpanents, and structures that are considered inportant to evaluate requests for license renewal. ,
Any additional researtti projects Whi& may be needed to resolve the . technical safety issues in consideration of license renewal will be added to the NPAR Fawa-u as they are defined. 3.2 Generie safety Issues a One of the objectives of the NPAR Fr%imu is to sqpport ER in the resolution of aging related generic safety issues identified in NURIr-0933, A Prioritization of Generic Safety Issues. NUREG-0933 contains a I*:--.-rded priority list to assist in the timely and efficient resolution of safety issues that have a high potential for reducing risk. The NPAR Fa@ tom results, Which can ba used by NRR in resolvirg several of these generic safety 4====, are listed in Table 3.1. Pbr example, the NPAR Fr%s.E. can support NRR in the resolution of the generic safety issues A-30,
" Adequacy of Safety Related DC Power Supplies" and B-56, " Diesel Reliability," by evaluating aging and service wear of emergency diesel generators. !
L h 20 l' i _________________________ _ __ _ _ 0
o, NRR has prwided " users need rupests" to RElS for the resolution of-some specific issues which are also listad in Table 3.1. !Ihe NPAR hwtain is supportig NRR in the resolution of the Generic Tasues II.E.6, "In Situ Testirug of Valves" and II.E.6.1, " Test Adequacy Sta. d' by assessing nethods for monitoring notor operated valves. 'the third is; ,e in'this category is GI-70, "PCRV and Block Valve Reliability."
'Ihe results of the residual life assessment task of the NPAR F% ani indirectly would be s@portive to NRR in resolving several generic safety issues related to the primary reactor coolant system +. id. 'Ihese 4 == ==: are identified in Table 3.2.
3.3 Maintenance and Surveillance Maintenance and surveillance sws-- at nuclear plants are significant ocntributors to system and plant reliability. 'Ihe NPAR h@tmu supports the . NRR - Maintenance and Surveillance huyt-- by prwiding evaluation of the
. . role of maintenance in mitigating aging effects. 'this evaluation consists of the follwing activities: (a) Raview current practices and procedures, carried out by nuclear utilities, to maintain gd - it. (b) Review nuclear mad P st vendor's recotenendations for maintenance of qw.ts or
- --: -,ts selected for aging ===a==mants. (c) Perform an evaluation, o
including a ocmparative analysis, of the relative serits of: (1) performing maintenance den a ocuperant has been discovered to be malfunctioning (vuttedve maintenance); and (2) performing maintenance when an observation has been nada throucjh surveillance, iWien, or monitoring, that a qwat say not function when required å a design basis or " trigger" event (preventive maintenance). Euphasis is placed on the relation between , failures (causes or modes) expected to be experienced during operation and thcee whi& would potentially cocur under the stresses associated with design basis or trigger events. (d) Identify, where pessible, those ;
+w.t failure me&anians likely to be induced through preventive or cmtative maintenance, specifically, look for those failures which might be detectable through short-tern, post emintenance surveillance, inspection, ,
or monitoring. (e) Develop h.-weations, for acceptable or preferred maintenance practices, based on the prwmiirig activities. i 21 f 1 l l t-
~
TABLE 3.1: ~ NRR' GDERIC ISSUES 71RT WOUID DIRECTLY BENEFIT FRCH 'DE NPARL ~ PROGRAM RESUISS l Issue Number Title Beactor Coolant Pump Seal Failures 23 ? 29 Bolting Degradation or Failures in Nuclear Power Plant 36- Isas of Service Water
~
46 Icos of 125 Volt DC has 47 .Imss of Off-Site Power
'48 IID for Class IE Vital Instrument Bases in Operating Reactors 49 Interlocks and IIDs for Class 2E Tie Brealours i 51 Proposed Rapirements for Inproving the the Reliability of open Cycle Se mice Water System 54 Valve Operator-Ralated Events Occurring During 1978, 1979, ]
and 1980. 65- Cm A. Cooling Water System Failure ) 68 loss of AME Due to APW Steam HEIB l 70 RRV and Block Valve Reliability . I
*75 Generic Implications of AIWS Events at the Salem Nuclear '
j Plant 84 91 CE PGWs Main Crankshaft Failure in Transamarica Delaval' Emergency l
. Diesel Generator i 93 Steen Binding of Auxiliary Feedwater Puups 118 'Dundan Anchorage Failure 122.la Canann Mode Failure of Isolation Valves in Closed Position i 124' Auxiliary Plandwater System !
; A-11 Reactor vessel Materials 7bughness l A-17 Systems Interaction A-30 Adequacy of Safety Related DC Power Supplies A Seismic Design Criteria-A-41 Iong 'Mrs Seismic P%=u A-44 Station Blackout .;
A-45 Shutdown Decay Heat Removal Requirements .l A-46 Seismic Qualification of T4M in Operating Plants A-47 Safety Implications of Control Systems ]1 B-6 Iceds, Inad Combinations, Stress Limits B-55 Imprwe Reliability of Target Rock Safety Dalief Valves B-56 Dianal Reliability C-8 Main Steam Line Isolation valve Isakage Control Systems
*I.D.3 Safety System Status Manitoring
*I.D.5 on-Line Ranctor Surveillance Systens g 7 *$ s 1 r. f f 5+$$j ** v* '"
Nididficada'n#[' o Y2Ian, Control, and Electrical i T 4.""'it j
*II.H.2 Obtain Technical Data on the Conditions Inside the 7MI-2 '
containment Structure
*Non-NRR 7MI and Generic Safety Issues )
l 22 ) 1 i
. i
'IABIE 3.2 GDIERIC SAFEN ISSUES THAT WDUID D1DIRECTLY BDEFIT FRCtd
'DE NPAR PROGRAM RESUIIIS !
3 i 1 Issue Nurber Title { i l 61 SRV Line Break Inside the BRR Wetwell Airspace of MARKI !
. arxS II Cantainnants 67.7 Steam Generator Staff Actions - Inproved Eddy Current Tests 73 Detached 'Ihermal Sleeves j
'J Iong Ranje Plan for Dealing With Stress Cerrosion Cracking 86 in BWR Piping
. l' 119.4 IER Piping Materials )
1 A-3,4,5 Steam Generator Tube Integrity A-49 Pressurized 'Ibernal Shock . i i E i
-l i
1 4 1 l i l i 23 i \ \ I l i 1
( 2e major anphasis of all of the above activities is on the technical aspects of maintenance rather than on institutional, organizational, pmt=satic, or hunan factor considerations. S e NPAR F%imu has been structured to define maintenance and surveillance needs to assure the operational readiness of aged power plant safety systems and canponents and provide support to the NRR staff in their review of the requests for license renewal. Se NPAR Favemu also provides the aging related information for the development of maintenance g%om criteria and standards, and maintenance indicators that NRR staff can monitor for specific v.mp r is and systams. 3.4 Plant Performance Indicators (Involvina Amino Considerations) De opting performance of nuclear power plants, especially in the 10 to 20 years prior to the and of a plant's operating license, is a , significant factor in evaluating requests for plant license renewals. Se
, term " performance indicators" refers to a set of data that may be correlatei with individual plant safety performance. Periodic review of the trends indicated by the plant performance indicators can aid in the evaluation of plant performance.
Dese indicators may be divided into two categories: direct indicators of current plant performance, i.e. safety system failures, and irxiirect or p m tanmatic indicators, i.e. an enfoi s .=nt action index. S e NRC staff M [ (IE) has selected a4 optimum set of seven indicators cm the basis of the p h k for >ce11Maticn and Ma'_mions with irdustry twi=sentatives. Se p ' selected irxiicators are:
.MKZbw&?
- 1. Automatic L.m While Critical,
- 2. Safety System Actuations,
- 3. Significant Events, j
- 4. Safety System Failures,
- 5. Forced Outage Rate, i p(,,:=y6. Enfuis ut Action Irdex, and
- 7. Equipment Forced Outage per 1000 Critical Hours.
24
S e third indicator, Significant Events, include degradation o2 inportant safety equi;nents, primary coolant pressure boundary w.wents i and important = = v iatad structures. S e research results enanating from l the NPAR Program oculd be used to evaluate the effectiveness of the first four of the (abcue mentioned) seven plant performance indicators. [ Af 3.5 Insoection h h r-2 NPAR P , _ aajs fotentia1 to support several ongoing m/rograns f x ; which guide the regional activities relevant to aging, aging detection, and mitigation of aging degradaticri. m ese et wiams include the Safety system Fun::tional Inspection PrwYam and the Generic Communication Program. In general, the Safety System Functional Inspection Pr@icuu assesses whether plant Mfications of selected safety systems have degraded the design margin to the point where the system's ability to mitigate design , basis events is inpaired. This program consists of an in-depth review of a small number of safety systems and is usually conducted at older plants. 7 Se objectives of the Generic Ctanmunication PiWicuu are to: )( h o inform licensees of problems, including tnose due to aging, th::,t have developed in irxiividual plants, ard a o require action when these problems are shown to be significant and generic. Rese etw r o-. apply to the pressure h urdary hardware, drivers, actuators, electrical power, ard the inst amentation and wd.wls of engineering safety features. i j ne NPAR Program will IE ' establishing inspection procedures that are relevant to aging. IE ncludes these procedures in the Inspection , Enforcement Marual issued to guide the activities of the regions. For example, scrne inspection pwculares establish guidance for ascertaining that inservice inspection ard testing activities are piwim.um, planned, ! l 25 l I j i l l
- conducted, recorded, and reported in accordance with Section XI of the ASME
- Boiler and Pressure Vessel Code..
Also, the NPAR Lyain has the potential' to support the engoing routine inspection effort conducted by the regional offices in accordance with the ' IE inspection programs. The objective is to assure that systems and. w .ts have not been measurably degraded as a result of any cause,
.ircluding. aging.
3.6 Codes and Standards codes and standards help define the inservice inspection requirements to assure operational. integrity of selected power plant electrical and . mechanical - p TJ. The NPAR hwa.m will develop rammendations to i revise relevant ASME and IEEE Codes and Standards to assure safe operation with aged ww a.s and syntams. These roccanandations will be developed , through active participation in the relevant technical ocanmittees. The special Workirg Group on Life Extension.- ASME Section n is ocordinating activities related to Codes and Standards of interest to the NPAR program. Several Working Grcqps in ASME Section H are involved in the
.- revision and development of the codes and standards for relevant life
~" extension 4 ====. A special IEEE working group was established to investigate the codes and standards aspects of plant life extension as it i may.be affacted by instrumentaticm and electrical control v 4= ant. The wp ira and systems currently of interest and being considered in ASME standards are listed in Table 3.f. Some of the Nevant IEEE standards are listed in Table 3. 3.7 NPAR Interferme with Other Fiwi-ia A nunber of additional NRR/RES programs and activities that have. potential to utilize the NPAR hvars results are listed below. 26
-c 1
a.. Egaipment qualification .j
- Reliability To&nology l
- b. -
y
- c. Evaluation of Mothballed Plants .-
- d. Imovative Materials and IHR Designs '
l
, e. ; Frantic III-
- f. PEIS
'g..- NtISM
- h. - NL5tEG-1150:
1 3.7.l' hi- ,
.t n=11ficattrwt
'me NPAR huwamu results sqpports implementation of 10 CFR Part 50.49, "Enviaw Aal Qualification of Electric hi,W Important to Safety for
. Nuclear Power Plants", Which ircludes the requirement:
4-
" Equipment qualified by test must be p uditioned by natural or '
artificial:(aooelerated) aging to its 'and-of-installei life condition.. l cannitieration must be given to all'significant types of degradation j
- Whis can have an effect on the functional capability of the '
equipment. If geditioning to an end-of-installed life condition is not practicable, the v4M nust be replaced or refurbished at the and of this designated life unless ongoing qualification demonstrates I that the item has additional life."
'me evaluation of actual aging processes throu$1 the research program provides a basis for assessing the adequacy of irdastry nurthods for i preconditioning prior to qualificaticm tasting or may lead to j
' rm. 4.tions for surve' llance i or monitoring. ' mis may involve
- h. 4 tions for zwvisions of the IEEE standards related to environmental qualification through participation of researchers in the aging research pw , in the relevant IEEE standards committmar* and develop of industry consensus based on the results of thf rectit. Sczne of the f(
relevant IEEE standards are listed in Table 3.3. j 27 i i
3.7.2 Reliability 'Ib&noloav
'Ihe hardware oriented NPAR Fi@tmu has the potential to prcuide support to the major elements of the NRC's Reliability Pacaarch and Technology Frwtaru. 'Ihe NPAR Ftwi.o is evaluating causes of ww.t and structural aging at nuclear power plants, the safety and risk implication of this l aging, and methods for detectirg and w, Lulling significant aging effects.
As part of its efforts, the NPAR Prwsma is collecting failure rate data on aging and is developing quantitative techniques Wtich can be used to l quantify the risk and reliability effects of agirg, usirg PPA p1t tree and
)( system models. M)$sults frm the NPAR Frwima can assis liability
'Ibchnology Pr%5mu to: mcr11 tor plant / equipment performance; ocupare plant j and equipment performance to acceptable or desired levels to help early f dete: tion of degradation; help identify causes of important problems; and help evaluate corrective action and verify effectiveness through performance ;
monitorirg. , j
'Ihe Reliability program has developed a framework and rwi-5 which can X be applied to maintain IHR[fety. An informal report, 'Ibe Organizaticr1 of JC Risk Analysis Ctdes for Living Evaluation (CPACIEp been otmpleted by the Division of Risk and Systems Safety of RES. It discusses several IRSS/RES
+4 e-
)( Pztgrams. 'Ihey irrtude: PFA Information for 3Plant Risk Ststus Information Manageme1t (PRISM) System, PRA-NURD31150, Probabilistic Evaluation of Technical Specification (PEIS) Fr%imu, Operational Safety Reliability Research (CSRR) FAWimu, and FRANTIC III Ctmputer Code for Time Deperdent Reliability and Risk Evaluation.
3.7.3 Evaluation of )bthhalled Plants
'Ibe NPAR P2tgram results could support NRR in development of criteria for evaluating plans foram:,thballinf' plants durirg wwLocticri and for ]
reactivation of mothballed aqd !=nnt. 'Ihe NPAR Prtgram is integrating information related to the degradation pr0: ==a= active in the nuclear power plant empaner;ts and structures, havire safety and -isk implications. Some \ 28 l l l ' L
7 of this integrated information would be useful in evaluating the integrity of the nothballed equipnent. For exanple, the NPAR Piwiam is integrating information for microbiologically induced corrosion (MIC) which is also hetive in nothballed equipment filled with licpid. 3.7.4 Innovative Materials ard nm Desians l I
'!he NPAR Fiwuuu is identifying and evaluating the critical degradation j sites ard mectanisms active in the m.gr=its and structees that are
~itical to plant safety. .hw# hts from the NPAR Prograd. chould be
- msidered in the design of the advanced light water reactors to assure higher safety nargins. Specifically, mp=,ts and systems specific j inspections, surveillance and condition monitoring methods identified in the f NPAR swiam oculd be irs.urporated as a built-in diagnostic system in the advanced IRR designs. 'Ihis design feature would assist in establishing ]
early base line data and trending of performance parameters and functional , { l indicators. ( 3.7.5 M WTTIC III l
'Ibe carputer code FPRTTIC III is used for time-dependent reliability
- and risk evaluations. It was developed by NRC and is particularly useful in technical specification evaluations. Technical data generated in the NPAR prwiam will be utilized in the development ard evaluation of time dependent
=vlaim and in determining risk significance of aging effects. !
3.7.6 PET 3 - Prteilist Evaluations of Technical Specification Procrram a
'Ibe PETS program is developing methcds for using reliability and risk analyses to impr1:ne technical specifications. 'Ihe development is focused on approaches for nodifying allcued outage periods and surveillance test intervals, on a plant specific or generic bases. The aim is for PETS to be available on software for personal corputers for NRC ard industry use.
29
i For specific wiria.dpiand systems to be stud' d the NPAR kwain, f surveillance and monitorirg methods will be reocamanded to alleviate agirg j concerns. 'Ibe x=v=nardations will include the identificatica of he ,e.r,t r ! performance parameters and functicmal indicators and optimin surveillance
.,K intervals. '!herefore, fsr thesie specific wpsisiits and PEIS f
-l ry.m could utilize NPAR results.
3.7.7 ppm 4 - Plant Pink Status Inferanation Manaamment Svstam , d
'Ihe PRISM system is a ocmputar software package written for an IIM-XT personal ocmputar. Its purpose is to provide plant-specific tools for plant i inspectors and others those jobs require little or no PRA '44. It is a decision-oriented, user-friendly, menu-driven program that ocritains data base % ~ ,t and interactive routines to aid NRC inspectors allocate their efforts toward thcee areas that have greatest inpact en plant safety.
Again, for specific +.=,ts and systems addressed in the NPAR pyam, Yb@ . K PRISM project oculd benefit frtan NPAR results. : 1' 3.7.8 NUREG-1150'- Psarter Risk Reference Ibcument
'Ihe NPAR sw m has potential to support the data base input for the ;
. development of NUREIl-1150 to account for aging and time dependent failures.
NUREG-1150 provides the results of major risk analyses for five different U.S. IHRs, using the ntate-of-the-art methods. It is intended that this document provide a' data basc and insights to be used in a nLabor of regulatory applicaticms including: (1) licensing - to evaluate the risk relevance of pW plant licensing ctanges, risk effectiveness of existing regulations, and risk priorities of generic technical issues; i (2) inspection - to develop the methodology and data to set priorities for-inspection activities, ard (3) reaanyth - to establish programs that directly address the analytical and experimental uncertainties identified in NUREG-1150. '!he " draft copy for connent" of NURID-1150 was ima==4 in February, 1987. l..
*fhe technical data integration with the aforementicsind raaaarth projects will be r- - =ded to the 7IRGAux for implementation, f
30 l-
- g. j o
3.8 Brief Syrmais 'of NPAR Results In Sinvvwt of the Raaulatory Prtx:ess sn o . Issued NUREG/m-4302._ Operating experience review and analysis were ocupleted to determine failure modes ard causes' due to agirg of check valves in plant safety systems. This researth supports NRR in resolution of Generic Issue II.E.6.1,'"In Situ 7tsting of l valves." ASME Operation and Maintenance (CEM) Ctanittee has been made aware of the results of the study. 'this NPAR_ effort related - to check valves has been referred to the NSSS Owners Groups' representatives regarding industry actions in response to the check valve and water hamer event at Gan Onofre, Unit 1. (IE has-the land.) l o Issued NURED/m-4597. A study was ocupleted to characterize agirg { of Aux. Feedwater Punos (AFP) and evaluate inspection and degradation'menitoring methods. Potential failures of the AFP , have been attributed to the presence of large hydraulic fortes, j particularly at lw flow ratas which are substantially different than the best efficiency flow. Methods for detecting failure modes and differentiating between failure causes were defined The 1 i researth will support the upgrading of R.G.1.147 and Inservice
- Inspection Code Case Acceptability - ASME Secticn XI, Division 1.
9 o Issued Draft NURBG/G-4590. The evaluation of operational experience and expert opinian indicated that the aging of nuclear servios emergency diesel-cenerators ist observable; follcus j recognizable patterns; shcws changes in the un4an of aging l j degradation with time; is cxmfined to few. relatively major wp As; increases as piuv Aage of all failures with time; and I is caused by ncremal operational stressors. The primary causes of diesel generator aging are: vibratien, adverse environment and human errors. 'Ihe results of this research have been conveyed to NRR and they can be used in the resolution of generic issue B-56,
" Diesel Reliability," and to Wade ASME Sections III and XI perraining to die.el term. .;
i 31 l 1 i
r j o Issued NURD3/G-4564. . Operating experience rwriew ard analyses were ocupleted to determine failure modes and canases, due to 'the I
~ aging of batterv-cfarners and invertars that are used in plant safety systans. 'Ihe identified major ocotriherwis to failures are -
fuses-and capacitors, and involve overheating and aging /wearout. ; 2he results of this researth in ocabination with the output of
>-Ocrs to upgrade mase II shwHam will be utilized via r+3 IEEE Standard 650, Qualification of Class IE Static Battery Chargers and Inverters for Nuclear Plant Generating Stations.
1 o Issued N.1RB2/m-4380. A field test program was carried out to 4 evaluate a technique'of valve signature analysis to detect and differentiate abnormalities including time-dependent degradation L (aging), and iram .ct adjustments in notar mtad valves. Measurusents were made at four operating plants to verify l
-l monitoring techniques and to obtain charat+*le " signatures" ,
indicative of G. .dation and misadjustments of meter,7perated valves. 'Ihis researth supports NRR in the resolution of Generic Issue II.E.6.1, "In Situ Testing of 'alves." 'Ibe research results emanating from the NPAR effort were utilized by IE in Bulletin No. l 85-03: M: tor Operated Valve-Carmnen M:de Failures During Plant i Transients Due 2b Iwsw Switch Settings.
.a ,
l o Issuen NUREG/m-4279. A study was oczploted to identify aging of { i hydtuulic and mechanical snubbers used on safety-related piping and w-&s of rmaclear power plants. 'the ASE Section XI code Danmittee ard ANSI /ASME/CH4 ccanittee have been made aware of the results of the study, and a value-inpact analysis reflecting the J reductica in the number of snubbers in existing plants is being in r..ted in a regulatory guide Sc-708-4, asy.1,
" Qualification and Acomptance Tests for art.iters used in systems y lwtarit. to Safety."
l ! I j \ 32 , I I
.t .4 . ,
io. Issued Draft NUREG/m-4692. A study was empletad for NRR using
- NPAR data in the resolutit:m of Generic Issue' No. 70, . "PORV and '
Block valve Reliability." 'the report 'contains a review of nuclear
~
power plant operating events ' involving failures of power operated relief valves (PCRVs) and associated block valves (BVs). Aging related data include:~ failure mode, failure mechanism and severity. < 'Jhe report also addresses questions sus as: (I) how do operator / maintenance actions contribute to valve' failures?; (II) are certain designs more prone to failures than others?;
~ (III) to what extant would qqrading. (valves, operators and > control systems of safety-related systems)' have prevented the failure?
o - Issued NURED/G-4234. Review and analysis of operating experience data was accomplished to determine the failure arv9a= and causes A (due to aging) for potor-qperated valyps. 'Ihis researth supports
. the NRR efforts to resolve Generic Issues II.E.6.1, "In Situ Testing of Valves." 'Ihe ASME Operation and Maintenance (0&M) -
comittaa has been made aware of the 2esults of thia study. Also, this NPAR effort has been referred to the NSSS Owners Groqp's representatives involved in responding'to IE Bulletin No. 85-03:
,$- Motor Oparated valve Coman MoSe Failures During Plant Transients Due to Ir.pp Switch Settings.
l 33 l i
TABLE 3.3 IEVEIDP RECOMEMATIONS TO REVISE IEEZ STANDARDS FOR ELECTRICAL ~ EQUIPMENT FOR NUCIEAR POWER PLANTS Standard Ntzter Title IEEE 308 Critaria.for Class 1E Power Systems for Nuclear Power Generatig Stations. IEE 317 Electrical Penetraticri Assemblics in Cantalment Structures for Nuclear Power Generating Stations. IEEE 323 Qualifying Class 1E M $4E ~.t for Nuclear Power Generating Stations. IEEE 334 Standard for Type Test of Continuous Duty Class IE Motors for Nuclear Power Generating Stations. IEEE 336L Installation, Inspection, and Testing Requirements for .l Class 1E Instrumentation and Equipment at Nuclear Power ! Generating Stations. IEEE 344 Roccananded Practices for Seismic Qualification of Class' 1E Mt4M for Nuclear Power Generating Stations. IEEE 382 Standard for Qualification of Safety-Related Valve Actuations. IEEE 383 - Staldard Types 14sts of Cle.ss 1E Electric cables, Field Splions, and R. d.icns for Nuclear Pcwer Generating
. Station.
IEE 387 Critaria for Diesel Generator Units Applied as Standby Power Supply for Nuclear Power Generating Stations. IEEE 501 Seismic Testing of Relays for Nuclear Power Generating Stations. IEEE 535 Qualification of Class 1E Imad Storage Batteries for Nuclear Power Generating Stations. I I IEEE 572 Qualification of Class 1E G fden Assemblies for Nuclear Power Generating Stations. IEEE 549 Qualify Class 1E Notor control Centers for Nuclear Power Generating Stations. 34 q l _____._-____________L
TABLE 3.3'(Continued) Stardard Number Title IEEE 650 Qualification of Class lE Static Battery Omrgers and Invertars for Nuclear Power Generating Staticns. IEEI 944 Application and Testing of Unintarruptable Power Supplies for Nuclear Power Generating Stations. Irrr/ low voltage AC Power cirt:uit Breakers Used in ANSI C37.13-1981 Enclosures. 4 4 9 C l 35
TABLE 3.4 DEVEIDP RDOCMG2MI'ICtIS 'IO REVISE ASME SIANDARDS FUR OPERATION AND MAINIDENCE OF MECHANICAL EQUIPMENT Stardard Number Title ASME CH-1 Inservice Performance 'Itstirg of Nuclear Power Plant Pressure Relief Services. ASME CM-2 Requirements for Performance 'Ittstiry of Nuclear Power Plant Closed Cooling Water Systems. ASME Of-4 Examination and Performance of Nuclear Power Plant Dynamic Restraints (Srnihhars) . ASME CH-5 Inservice Monitoring of Core Support Barrel Axial Preloads in PWRs. ASME GG-6 Requirements for Performance 'Ittsting of Punps in Light Water Cooled Nuclear Power Plants. l ASME Gi-8 Requirements for Preoperational and Pericdic Performance ' Testire of Motor-operated valve Assemblies. , ASME Gi-10 Requirements for Irservice 'Ittsting of Valves in Light Water Cooled Nuclear Pter Plants. ASME Gi-13 Requirements for Periodic Testirg ard Monitorirg of i Power Operated Relief Valve Assemblies. ASME Gi-14 Requirements for Vibration Monitorire of Rotatirg ,
*; Equipment. i l
ASME Gi-15 Requirements for Performance 'Ittsting of Nuclear Power j Plant Emergency Core Cooling Systems, j i Inservice Performance Testing of Nuclear Pmer Plant l ASME Gi-16 Diesel Drives. ASME Gi-19 Startup and Periodic 'Ittsting of Electro-Pneumatic Operated Valve Assemblies Used in Nuclear Power Plants, j j i I l 1 I l I 36 I 1 l - - - - - - - - - _ - - - - - _
gy i n, , . h
;.g 4.- IEEEAROi APPROhCH p
'the NPAR p,-o.is intended to achieve: (a)~ an ui-dia.'anding of melaar plant aging, (b) its potential effects on safety and (c) methods for detaction and mitigation of aging effects. 'Ihe program is structured to develop a technical data base sufficient to address tatnical safety issues and to support regulatory decisions on such issues.
4.1 Bipk and Systans Oriented Identification of Anim Effects
. !he research projects in the NPAR program use the phased eR: roach to research as shcun in Figure 4.1. An initial selection process is followed to establish priorities for detal. led aging assessments of specific systems and w w A.s. 'Ihe ==1=rtion criteria include nuclear plant experience,
" user" defined needs,wt judgment regarding susceptibility of the. system or ww-it to agingh 3ation, and the potential contribution to risk ,
from failure of systems or ww ia. 4.1.1 Operatim ExDariance and Opinions - one of the sources available for failure data of -g is and systems
, is the operating experience currently being obtained from dal light water zwactors. 'this information is obtained frcan the Licensee Event Reports (IZRs) and the Nuclear Plant Reliability Data Systaan (NPRDS). 'Ihe infcvInation is being analyrmi to identify systars and ww 4.s which are
=1=r=:tible to' aging related failures (Reference 11) . 'the carrrent effort has followed the initial soeping studies of plant operating experience (References 3,10) . Also, anoe the +-/s and syntans are selected for an in-depth engineering stanty, data bases are evaluated further to identify failure modes, failure frequencies, failure causes and methods used to l remedy the causes of the failures.
Expert opinion has also been used for the selection of systems and ww d.s. Expert cpinion was used early in the program to esttlish the priority of the researth affort (References 5, 6) . Since that ti]me 37
rherdations ard advice have been sought frm experts throughout the industry. Most recently expert opinion has been used by the TIEMD: to acca== the priorities which will be used in the NPAR program. 4.1.2 Risk Evaluation Risk studies have been used to evaluate the potential consequences of
+#.nt or systan failures due to aging %.dation. 'Ibe results of the initial effort are given in References 8,9. Mare recently, work has began on a time @ dent calculation of risk that can evaluate the effects of age relata:1 failures throughout a plant's service life.
4.2 Phase $ acornach to Acine Acca== ment and In-Detth Encineerina Studies 4.2.1 Thase I
'Ibe aging a==*==ments of the ww-rits or systems selected for evaluation involve two stages. 'Ihe first stage, Phase I, is hamad on raaM1y available information frm public and private data bases, vendor information, open litarature, utility sources and expert opinions. 'Ihe product of the Ihase I analysis includes a preliminary identification of the
. significant modes of degradation and an evaluaticri of current inspection, surveillance and monitoring methods. namaa on these evaluations, rv--srdations'are developed to identify detailed engineering tests and I analyses to be conducted in Phase II. 'Ibe Pha.se I evaluation is used to decide if a Phase II maaaa-nt is warranted aid on occasion may lead to a r+_--andation of a Ihase I amaanwnt of a +=nt or system not yet selected for evaluation.
4.2.2 Phase II
- In those cases where Phase II mem i.ts are needed, they generally j involve some cmbination of: (a) tests of naturally aged equipnent or of equipnent with sinulated degradation; (b) laboratory or in-plant verification of methods for inspection, nonitoring, and surveillance; 38 l
1
)
(c) development of r-_-_- =dations for inspection or nonitorirg techniques in lieu of tests 41cti cause excessive wear; (d) verificatim of methods for evaluation of resMm1 service lifetime; (e) identification of effective maintenance practices; (f) in-situ examination ard data gathering for operating v4M; ard (g) oost/ benefit analyses. With the ocupletion of the agirxJ mamaamment technical basis A is available for utilization in the regulatory prmana. Exanples of utilization i!clude: implementation of improved inspectism, surveillance, maintenance and nonitoring methods; modification of r_.a. codes and standards; developirrj guidelines for plant life extensist; ard resolution of gernric safety issoas. A detailed description of the individual steps. in the Phase I and Phase II assessments is given in Appendix A. . e 0 i l 39 i 1 i
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- 5. PROGRAM IESCRIPIICH 2 e NPAR r e am elements are listed in this section along with a description of their role in plant agig assessment ard license renewal.
De -parents, systems and civil structures whicts are of current interest were selected cri the basis of the reviews of safety significant items, users naaac and experts opiniens, and may be expanded in the future. 5.1 Cuwr=ds. Systems and Structures Studied in NPAR me -persats, systems, and civil structures of current interest have been selected by the risk oriented identification, users needs and expelts opinions. Se selection pr===a= are d4=mecad in Sectico 4, ard in more detail in Appendix A. S e currently selected items are listed in Tables 5.1, 5.2 and 5.3. Table 5.1 contains a listirg of the -parw4ts which have been re*#h identified as having an aging related impact on plant safety systens C availability and margins. Se c yorents in Group 1 were evalua early [ effort is the NPAR program. Phase I rerth has been expleted for the Group I m ycr=,ts. Se Phase II effort has been initiated ca motor operated valves, check valves, auxiliary fee @ater purrps, electric motors, batteries, chargers / inverters, snubbers, and solenoid operated valves. Additional research effort will be definal, when or if, issues are identified during the Phase II efforts. Se -rcents listed in Group 2 are cunently included in the NPAR scope of work. Dependig upon the availability of funds, Ihase I engineering evaluations on these cwparents are sctaduled to begin in FY 88. Se Grup 3 wycrwats have been rim.w..crded for aging studies by various sources. Engineering evaluations of ths mge.rits in Group 3 are not included in the NPAR Frupam at this time. Table 5.2 is the list of ruclear plant systens which are of current interest. Rese systems are considered important for accident prevention or mitigation. l 1 41
1 1 TABIE 5.1. COGUENIS OF CURRDTT DfrERES"r IN DIE NPAR PROGRAM Group 1 o Motor Operated valves o ChecA Valves o Auxiliary Feed Water Punp o Electric Motors o Batteries o Chargers /Invertars o Sru ***rs o Circuit Breakers and Relays o Solanoid Valves o Power operated Relief Valves Group 2 o cables (power, control, instrument) o Electrical Penetrations o Cuistors, Tar 2ninal Blocks o Heat Dechangers , o Compressors o Transformers o Bistables/Switdwe Group 3 o Fans Chillers o Purge and Vent Valves
, o Safety Relief Valves o Service Water and Cupaud. Cooling Water Pumps o Air Operated Valves o Main Steam Isolation Valves o Am=ilators o Surge A W Mrs o Isolation Condensers (EHR) l 42 l
TAPlE 5.2. SYSTDG OF CURRDTI Duuc5T IN THE NPAR Pf0 GRAM I Group 1 o High Pressure Daupcy Core Cooling System o Im Pressure Dnargency Core Cboling System o Service Watar System o C+4at Cooling Water System o Reactor Protectics) System o RasMm1 Heat RamcNa1 SystanVAuxiliary Heat Rancnal System o Mm 1E Electric Distribution System o AtM 14e y Feedwater System Group 2 o Control Rod Drive System Group 3 ) o Engineered Safsty Feature Actuation System o Recirculation ?unp Trip Actuation Instrumentaticm (BWR) , o Reactor Core Isolation Cooling System o Stanty Liquid Control System (BRR) o containment Isolation o Containment Cooling SyStans 43
)
l TABIE 5.3 MAJOR DR PLANI EUMENTS OF CURRENT INIEREST IN 'DE NPAR PROGRAM PWR i
*o Reactor Pressure Vessel o containment and Basemat
*o Reactor Coolant Piping and Safe Ends
*o Stamm Generator o Reactor Coolant Peip Body o Pressurizer
**o control Rod Drive Medanism
**o cables and connectors
**o. Emergency Diesial Generators o D "ilator o Reactor Pressure vessel Internals o Reactor Pressure Vana*1 Sqpport o Biological Shield o Pressurizer Line BWR
*o Reactor Prossure Vessel o containment and Basemat
*o Recirculation Piping, Safe Ends, Safety System Piping <
o Recirculation Punp Body
**o Centrol Rod Drive Medanism
**o cables and Ornectors
**o Emergency Diesel Generators o Reactor Pressurs Vana*1 Intervals o Reactor Pressure Vessel Support o- Biological Shield 1
l l a
* 'Ibese INR primary system wtW- As are the subject of in-depth engineering studies sponsored by the Materials Branch of the Division of Engineering Safety.
** 'Ihase IHR mp=As are the subject of in-depth engineering studies (as part of NPAR pu: mu) sponsored by the Engineering Branch of the Division of Engineering Safety.
44
l Phase I and Phase II evaluations are ncw s c=%g on the syrtams in l- Greup 1. Se Phase I' evaluation of the control' rod drive system (Group 2)- is planned for W 88. Se systems listed in Groqp 3 have recently been id!sntified as having i importance to the evaluaticri of plant performance. Due to limitad availability of resources, the systems in Groqp 3 are currently outside of the NPAR s vytom scope. De formal study of the residual life of the major IHR plant wp,=nts and structures listed in Table 5.3, was initiate:1 in n-1986. M ose major
-[
+ - ta and structures are the large, relatively expensive, parts of a ruclear power plant whicts are not routinely or frequently replaced. Se major mies .ts to be studied were selected on the basis of the safety critaria that the release of fission products, that may occur during an
- accident should be contained in the plant (Reference 23).
2e major up-its selected for evaluation include the pressure boundary -p,Ents, containment and supporting structures. Also included are mp its related to reactor control systens and reactor safety systams. me. reactor internals are incitzled here also, as their failure may prevent control rod insertion or may cause fuel failure. Major ww.ents are identified from both B4Rs and BWRs and are listed in order of priority in Table 5.3. N7TE: It is Dgt, the intent of the NPAR program to do indepth evaluations of: engineering, aging and defect daracterization, and methods for inspection, surveillance and monitoring of au significant 1 plant elements. Se NPAP pivytas efforts have to be focused since we have to ocrisider: (i) PWRs frtan three different NSSS vendors, (ii) BWRs, (iii) plants with WJmerous variations in design, I applications, and su g liers, as well as (iv) creraticn and maintenance with differing practices and philosophies. he intent of the NPAR ' effort is to study a few selected electrical and mechanical cup,ents, 45 l
l and a few representative safety systans ard support systems. 'Iben our purpose is to'danmstrate how the NPAR strategy can be applied to other . s v -rits, systems or structures of interest. Also, the applications . will involve different designs, operating and maintenance practices and l philosophies. It is the industry's responsibility to daracterize and evaluate their own plant systes, mwi-A.s and structures; and assure their operational safety as they advance in age. Evaluation of reactor pressure vessels, the reactor coolanc piping ard safe ends, and steam generators are being studied in the m mL sponsored by the Materials Brancil of IES. 5.2 F1 wiam Elements
'Ihe NPAR program has been implemented to develop technical data related to plant aging and license renewal. 'Ihe research being performed in the ,
NPAR program can be categorized in the following major subjects or technical areas. 5.2.1 Risk Significance of Acino-Systems Interacticri Studv. In this effort aging related failures, identified from plant operating data, are evaluated to determine their risk significance to plant safety. 'Ihis study has provided
. information W to select wpr ra, systems and structures for detailed Phase I and 3hase II evaluations.
5.2.2 Aaira Aem=mant of Arw-ific G---r-its and Systems. Aging a==== ants of mp-d.s and systans are in progress and consist of the 3hase I and Phase II ergineerirg evaluations.
'Ibese evaluations are being made on the w=/s ard systens considered vital to plant safety durirg normal operation as well as during accident and post-accident situationn. Safety i==== that need to be a&ir*==wi in 46
rwiewing applications for license renewal (involvirg specific wiwet.s and systems) will be identified and resolved. 5.2.3 Acim >= mci d. of Civil Structures. 'Ihis study involves an in-depth a==a===nt of the aging degradation of the mnete civil structures in nuclear plants. It incitdes identification of the principal structural safety issues; development of materials a u.ges data for aged civil structures; and evaluation of the functicnal capabilities of aged structures in a post-accident environment. The data generatcd will be useful in evaluating applications for plac life extensierg/ license renewal. 5.2.4 Iria_rmetion, Surveillance ard Manitorim Methods. 'Ihe methods used for inspection, surveillance and monitoring of each of , the various hwets, systems and structures are reviewed as part of the engineering evaluation. Current industry practices and prona+mes are a-=M and r+: -- =4ations developed for imprwed or preferred methods. 'Ihis review includes the identification of performance parameters and functional irxiioators M11ch can be used for the early identification of age-zulated degradation. 5.2.5 Role of Maintenance in Miticatim Acrim. Evaluations are made of the role of maintenance in mitigatirg aging effects. A rwiew, of present practices, in terms of a comparative analysis of corrective versus preventative maintenance, and r+,-. =dations for preferred practices are included in this study. 5.2.6 cu us,ent Lifetime Evaluation. 'Ihe current methods for predicting service life of electrical corporumts are reviewed and an initial lifetime evaluation started for mechanical wpsts and structures in this effort. Consideration is 47 1 I 1
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. given to the inpact of inspection and acnitoring of sane w iJ and structures; Whi& can be prohibitively high in resources required and occupational exposure. In these 1 cases, evaluaticms will' be made to deramine if there are - -]
ta &nically moonptable nethods for predicting service life. 5.2.7 *=r*4=1 Taoics
- 1. Aoina/se4=mic shock Ir*=vact4rm. 'Diis is a study to determine the vulnerability of age-degraded mp7 ta to aniemic events. seismic qualification of electrical equipnent already requires consideraticm of pre-aging Hourver, no requi.e.t currently exists for pre-aging the mechanical n'4_% to be qualified. 'Dtis effort is aimed at assessing how aging G6,t.dation will affect the-performance of electrical and am&anical n'i==nt during or after a seismic event.
- 2. . Wificatice of Anim.- 'this topic includes developing a practical a;proa& to the quantification of agin; and a residual life assessment of anjor +-ta. The latter i inclu5es an effort to-identify the life limiting pmces of es& of the major wwwta, trending of performance parematars and functional indicators, and the determination of margins. Methods are then developed for determining major coqxtent lifetimes.
Y g 3. gap 4==lenim of the sbirmimoodt Plant. 'this effort involves in-situ ==amman=nts, acquisition of selected data / records and specimen manples and + era from the plant, whi& has ocripleted 25 years of service. Also, post service tests and examinations are planned and l coordinated with NRC staff and NRC ocatractors. Program i details are described in the Appendix, starting on page C-2. 1 48 (-
- 6. . 000RDDUCION hTIH ODER 390 GRAMS, INST 1WIIONS AND CEANIZATIQ3 Various institutims and industry organizations have performed snaiies )
and instituted prograns relevant to the aging researtti. Results of the more important activities are reported in the References (1, 3, 5, and 7). Also, there are a rannher of ongoing p,- whicts are prciacing significant results whicts cannot and should not be duplicated. A major emphasis in the NPAR p % 5=Ti plan is that proper cooniination and integration, of plant ! ? aging research activities, is obtained at various levels. 'Dris apre ! will help actaieve overall pro;p1m goals and objectives and assure the efficient use of available resources. Ongoing NRC programs related to aging and license renewal are being ocmducted by AEDD, IE, HRR and RES. 'Ihe NPAR rwtwu ooortiinaticri and technical integration with other agency programs is in place or it will be implemented thrixgh the direction of the cuersight group, the 'Dachnical , Integration Review Group for Aging and Life Extension ( ), established by the Executive Director for operations ( .In tion to F )y( NRC sponsored raaaad, aging and nuclear plant life on prograns are being sponsored in the U.S. by nuclear industry groups including EPRI, NSSS Vendors, utilities, architect engineers and the DDE. Also, nuclear plant aging and life extension pa,== are being conducted in a number of foreign ocuntries. A listing of the more relevant r,-ts, external to the NPAR progran, is given in Table 6.1. Appendix D oontains a detailed description of cngoing aging-related ryous. NPAR sws u interfaces with ongoing NRC pro;rcans have been established and will be maintained. External prograns, involving both dcanastic and foreign organizations, also have been ocatacted. It is very likely that new aging-related prwim. will be spcewored in the future by outside 1 organit.ations. An important and continuing activity in the NPAR program is I I- to identify new projects and establish appropriate interfaces. 49 f
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l p 7. . SOEIUIES AND RESOURCE REQUIREMENTS The currently estimated schedules'and milestones for ocupleting specific researd :,stivities for aging and license renmal are prwided in l Appendix E. These general s&soules,. and partia.11arly schedules for evaluation of specific +.-,ts, systems, and structures will depend on funding, the assigrunent of priority, and the degree of coordinatica and participation by other institutions and organizations. . She NRC/RES staff l and its contractors will actively pursue the solicitaticr1 of such participation from dcznestic and foreign institutions and organizations. The active interest in requirements for license renewal (plant-life octansions) should facilitate industry cooperation and active participaticm in aging r==aartdt. Two critical sws-u requirements identified 2mr the NPAR program are: s ,
- 1. The timely availability of data for naturally aged v4.M frtrn operating power plant facilities. .
n
- In situ aging ======nents and trunding of e :- 4 and system performance are twmamary to aging characterizaticr1 rnd detectim of l defects. Also, post earvice examination and t-,Qq of naturally aged equipment are essential to relate artificial aging (pre-agirg/ accelerated aging) to normal aging. Evaluaticn and analysis of naturally aged equipment is intended to generate rw_4ations for critaria and guidelines for decisicris, cooperative prtxprams among the industry and researd organizations should be carried out to facilitate the availability of naturally aged equipmar.t for aging research. Schedules and resource requirements will need adjustaants to reflect the extent of these cooperative p w r -s,.
- 2. To identify and resolve the tactinical safety issues to support the NRC's definition of its policy and regulatory positions for license renewal. The major NPAR p.ogram elements in support of this objective include:
51 l
4
)
Prioritization of technical safety issues; reviews of industry's topical reports; reviews of applicable codes and standards; and develop iavposed regulatory policy review pivc.iares.
)
2e current NPAR schedule has been adjustad to support the regulatory activities now anticipated for nuclear plant life extension. NRC currently plans to have a policy in place by the early 1990s. To support the schedule for the NRC renewal policy, the reseant activities related to plant life f extensions are currently scheduled to be completed by early 1990s. me figures showing the milestmas and schedules for the NPAR program elenants are contained in Appendix E. For certain w w . ds and systems, i an additional activity is indicated by dashed lines going beyond .the Phase II effort. % e purpose of the dashed lines is to indicate that we q anticipate that additional tasks may be m"y in the vivc-ss of applying the Phase II results in the Regulatory pt w. , NRC fundinJ available to the pr@ tam for FY 87 is $5.2 million dollars. De FY 88 funding level has not been firmly established. In addition NRC staff participation at an average level of five full-time-professionals per year (ITE) will be W for the duration of the . program. When participation by outside organizations is achieved, the resources provided frun oatside NRC will be identified in future revisions
- of this plan.
l l l 52
l l. (. l . \ APPINDIX A NPAR RESEARQi S'IRATB3Y TABIE OF CONIDfIS l- \ l A.1 Selection of G p uts, Systems, Structures for Aging Evaluation ....................................................... A-1 A.1.1 Risk ard System Oriented Identification of Aging Effects .............................................. A-3 A.1.1.1 Operating Experience ard Expert opinion. . . . . . . . . . . . . A-3 A.1.1.2 Risk Evaluation .................................... A-4 A.2 Ihased Approach to Aging Assessment ard In depth Engineerire Studies . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-5 A.2.1 Phase I .................................................... A-6 A.2.1.1 Review of Design Information ard Applications. . . . . . . A-6 , i A.2.1.2 Survey of Operating Experience ard Failure Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-8 , 1 A. 2.1.3 Screenire 4xahtien ard Testirg . . . . . . . . . . . . . . . . . . A-9 A. 2.1.4 ISDH keview ard Evaluation . . . . . . . . . . . . . . . . . . . . . . . . A-9 A.2.1.5 Interim A==etent ard R+x-m---dations . . . . . . . . . . . . . A-10 O A . 2 . 2 Ihase II . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A- 11 A.2.'2.1 Review and Verification of Iltaoroved ISME ard In Situ A - emants....................... A-11 A.2.2.2 'Nsting of Naturally Aged Camponents . . . . . . . . . . . . . . . A-12 A. 2.2. 3 Resichm1 Life Evaluaticms . . . . . . . . . . . . . . . . . . . . . . . . . . A-13 1 A.2.2. 4 Service Life Prediction Methods . . . . . . . . . . . . . . . . . . . . A-13 A-i
.q
. ,1 A.3 Application Guidelines ........................................... A-13 fj 1 A.3.1 Value Inpact Study and Cooniination with Users . . . . . . . . . . . . . A-14 1 1
>A.3.2 Support Resolution of Generic Safety Issues ................ A-14 I A. 3. 3 considerations for Life Extension . . . . . . . . . . . . . . . . . . . . . . . . . . A-14 A.3.4 Guidelines for Inspecticm, Surveillance, and Maintenance ... A-15 ' ,;
J A.3.5 Guidelines for Service Life Predicticos . . . . . . . . . . . . . . . . . . . . A-15 A.3. 6 Roccanendations for Standards and Guides . . . . . . . . . . . . . . . . . . . A-15 t 1 A. 3.7 Dissernination of Tactinical Results . . . . . . . . . . . . . . . . . . . . . . . . . A-16 4 A.3.8 Innovative Materials and Design ............................ A-16
.I FIGURES
+ i A.1 NPAR Fis, cuu Strategy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2 ;
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. APPDOIX A
' NPAR RESEARG SIRATBn The NRC aging researdi pr%i-u is carried out in' discrete stages as ,
I shown in Figure A.1. The phased approat shown here is applied to all of l l the projects. in the NPAR Fauwa-u, except for the special topics. The use of q this formal process is pa=4a4 in the NPAR Fs,-o, boosuae of the wide variety of types of systems and v4==nt that are analyzed, and because of the multidie4NMary nature of the various research projects. The NPAR g,-u a two-phase approach to perform detailed aging assoasmant. In interim aging amma==aant is performed using )( readily available tion. Where warranted, this is followed by an in-depth Phase II wiv d-rsive aging maman=nant. The te&nical information generated in Phase I and Phase II is then used in developing application guidelines and te&nical recamardations. ' A 1hase III has been added to < provide for the resolution of issues that may be raised å the results utilization afforts. A.1 Selection of r w r ?J. Systems and Structures for Agim Evaluatico O The first stap in the approad used in NPAR is the selection of - OvupcE-ats, systems and structures for in-depth engineering studies. The selection critaria include: the potential contribution to risk from failures of s ycs As systems and structures, experience obtained from operating plants; expert judgement over the t--derg to aging degradation; and user , i needs. The user needs include resolution of generic 4a==, plant performance indicators, and plant maintenance and surveillance. The selection prmana includes: establishing a boundary to define what is to be included in the owycs-TJ, system or structure under consideration; the location of important irrarfaces; a d what is external to the boundary and not included as part of the hardware under evaluation. A-1 1
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F. 4 9 A.1.1 Risk and System Oriented Identification of Aaim Effects An initial evaluation is made of the effect of aging en plant safety systems, support systans andW==nt and its inpact en plant safety. 'Ibe
, historical operating experience of li@t water reactors, expert opinion and system risk evaluations are all used in the initial evaluation.
A.l.1.1 creratim Entrience and Dgpart Opinitm Failure data, derived from the operational experience of li@t water reactors, prwides information valuable for the evaluation of the inpact of aging related failures cm operating plants. 'the data are used to categorize systems or +- ts within specific syntams that are susceptible to aging related failures. General trerris of aging failures are identified by
- categorizaticm of failures and actual aging mechanisms are identified by the -
use of root cause analysis techniques. In general, the failure category , informaticm is used to identify which systems, and the r+tive system cuig d.s, are susceptible to age related failures and warrant indepth engineering studies. [ Plant Operating Data has been used'since the start of the NPAR Program to identify aging related degradation. Initial scoping studies (References 3, 4,10) were made to identify the effect aging has on cuig-,ts, syytams and-structures; and how aging C-podes the required functional performance.
'Ibe information, currently being gathered for system level aging degradation, is derived frun the Licensee Event Reports (IERs) and the Nuclear Plant Reliability Data System (NEDS). 'this information is collected using failure-category and cause-codes which allows identification of the failure as belonging to one of several broad failure categories. Analysis of this failure category information is contained in NUREG/CR-4747 and NUREG/CR-4769 (Reference 11,12). 'Ibe data shows that system dependencies of aging exist and can be readily identified. 'the systems covered in the NUREG reports are a subset of those delineated in NUREG-1144. Efforts are A-3
under way to obtain aging failure infomation en the remainder of the - l systems. '!he UR and NPRDS data will be supplemented with the data frm the j In-Plant Reliability Data 'Systen (IPRDS) . I i In addition to the systems covered in NUREG/CR-4747, information on the. ! root cause of migi A. failure was obtained for several service inter and class lE electrical power distribution systems. 'Ihase systems were chosen as a result of their safety siptificanos detaminations in past PRA studies. Failure information was also derived from the NPRDS. 'the aging anchanis;.3 can be identified to the level of resolutim provided in the failure racords. 'Ihis infomation is used in a=====ing the risk implications of aging nachanisms. 4 surveys of expert opinion have also been conductad to identify age related problems in rmelaar plant hardware, that inpact safety. Workshops were held early in the rwi.- (References 5, 6) to identify the i==== , m =ning aging and to review the state of knowledge on aging degradation. 3 Aging problems in miw-.ts, systems and structures were identified and a consensus was reached concerning which c-g-ids were the nest inportant, I in terms of aginJ related degadation. ! . A.l.l.2 Risk Evaluation !
'Ibe objective of this task is to identify mcg-ta, systems and structures which will significantly inpact nuclear plant risk if aging related degradation decreases reliability and availability or results in degraded performance. 'Ibe initial approach taken was to use the results of !
existing probabilistic risk analyses (PRAs) to investigate the relationship [ between risk and aging related de.podation (References 8, 9). A sensitivity study was performed to detamine the effect that increases in failure rates ! of mi g A.s and systems had on overall plant risk. Ra w m the aging . 1 sensitivity study, a large ruh of the risk significant w as were i identified in the auxiliary feedwater system and the reactor protaction A-4 i
r . system.: Punps, check valves, notor operated valves, ci2ntit breakers and actuating circuits were the - w ,t types found to have the most potential risk inpact. Time w--n. calculations, that propagate the effects of aging throughout plant design life and beyond, are tw=amay to prwide a reasonable estimate of aging effects. 'Ihe reason for this analysis is that the risk ranking of systems and +-,^a can ctange with time een one considers aging effects. Te&niques for parforming time dep-.". d. risk or oore melt probability calculaticos have been developed (Reference 12). Current efforts femari on the propagation of aging impacts, at the systen/-g at level, using the aging failure information frczn root cause analysis. 'Ihese results are not directly relatable to plant effects until the sequence calculations from PRA studies have been ocanbined with the aging impact propagation te&niques. Based on the results of the current evaluations and aling development, it is expected that future efforts , will focus on the larger aspects of plant safety or risk. Risk evaluation of aging mechanisms are used to gage the importance of aging on plant safety. In general, the calculaticos are carried only to core melt to provide a relative estimate of the dange in risk due to
.. aging. Where situations warrant, the effects of aging are carried through to the actual calculation of. risk (Inval 3 PRA'results).
9
- A.2 maw Ammad to Anim A=-- m-J. and In-deeth Encrinneriner Studies
'Ihe NPAR Psejima essentially uses a two phase approad for detailed aging ====ments to make the best use of the available resources. 'Ihe two phase effort assuras that the work being done in the program is fmmari on the acst significant research elements, modes of age related degradation and utilization of resources.
A-5 I . . . .. . . . . .. .
i l A.2.1 Phase I
%e mase I analysis of a selectM c went, system or structure includes a review of three elements: (1) the hardware design, operatirg j envitu er4t and performance requirements, (2) a survey of operating experience, and (3) the current methods used for inspect. ion, surveillance, monitorirg and maintenance and for qualifying and of life perfonnance.
A.2.1.1 Review of Nitrn Infomation and Anolications o Desian and Specifications. De first of the three elements of the Phase I assessment begins with a review of the design data and specifications for the hardware being studied. Eis includes such items as rueeietary design &w'ments, final safety analysis reports, operating and maintenance manuals and product literature. Additional sources of infomation is , also pursued. Dese include vendor surveys, utility contacts, published reports and expert opinion, o Materials. An important aspect of part of the a==a==mnt is the identification of all the significant materials which ocr: prise the hardware boundary under review. A list of all significant materials is generated for the harthare. Se
. materials and parts judgM most susceptible to aging are identified.
o ooeratirn and Dwiiuu d.a1 Stressors. De age related degradaticm of the e p-tus, systems arx3 structures is a time de+=4 erit pherar cgi arri anong other things, depends on operatirg envitu==r.t and operatirg history. De enviiu tal effects Which are considered include stressors such as taperature, radiation, chemicals, contaminants, atmospheric coniitions, humidity, and, in the case of primary system -e ients, c coolant chemistry. Se err /ircrnental effects considered include the conditions prevalent during A-6 1
t. operation and also the enviiw.- tal' canditions that prevail durig other periods, such as during shutdowns, storage periods, accident conditions, and post accident situations.
'Ibe operatig history ammaamment should include the thermal, me&anical arxi electrical stressors that +-as, . systems and structures experience durig their operating lifetimes.
n Here one has to consider normal operating conditions,. anticipated transients, off-norinal canditions, .and accident and post-accident conditions. Typical examples of electrical stressors include: ' slow switding transients, fast transients of the licAtning variety, and low occ d frequency 50-60Hz signals.- These canXsingly or in various X combinations. Exanples of mechanical stressors are: static loading stresser, dynamic loading stresses, seimmic and vibrational streces. . o Parfomance Parameter 1s and Functional Indicators
..Ihe performance parameters of the hardware are reviewed to a===== if aging degrades the ability of. the ecsponent, system ,
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. or structuzz to perform its required safety function dur
' normal, abnormal and accident canditions. , is possible y [
~to identify critical perfnmance and functional iniicators. l
'Ibese consist of indicators that are practical to monitor; and whi& provide cost effective means to identify age-related % 6.dation. Finally, ongoing research is zwiewed and applicable results are included in the amaa==mant of hardware under study.
'Ibe aforementioned review and analysis of materials, designs and L specifications, stressors and envitu. it, and operational parameters is attanpted on all + rus selected for the u M.misive aging a==== ament. It is recognized that sczne of the data seard and analysis may not be te&nically as well as practically feasible. Nevertheless, an effort A-7
is mde to acquire as much of this knowlete on a given m. gent as possible. If safety issues warrant such detailed assessments and adequate resources are available, then the logic for reviews and analysis described above are followed. (All NPAR research contractors follow this same strategy.) A.2.1.2 Survey of htim Bcoerience ard Failure Evaluation A second element of the Phase I ammanm=nt is a criHrm1 survey of the operating experience obtaired to-da^a cri the u.mp fa, systems and structures being evaluated. mis rwiew provides information en the reliability that can be expected and the failure modes and causes that have been experienced. Information on the failure no$es and causes that hem been experienced in nuclear power plants are obtained from a variety of sources. mese , include data from cm going supcus, sucti as: the In-Plant Reliability Data System (IPRDS) sponsored by NRC; and the Nuclear Plant Reliability D!Lta System (NPRDB) spcidwai by INPO. Other sources include Licensee Event Reports (IERs), Nuclear Plant Experience (NPE), Plant Maintenance Records and In-service Inspection Reports. - tw In this element of the Phase I assessment, 4he evaluation is made of g the hardware failures that have occurred to identify the following, o Pailure MechaM cms. % ese are established through the process of: identification of dominant stressors; study of materials and designs of mpts and parts; and review of service g envim-ut.s an1 applications. the nature o[and the factors ccritributing to age-related degradation and failures are evaluated. o railure modes. Se indicators of failures (i.e., voltage collapse or disturbance in current signature, etc.) are ==cai, and critical age-related failure nodes are identified. A-8
o Failure causes. De carditions of design, manufacture,~ servloe ernim.. .^J ard applications that may lead to failures, are detamined.4 A.2.1.3 Germanim F-n%atim and Testim In additicri to the evaluations of operational records, it is remmary to perfom same screening type exaaninations ard tests, on selected
-ir -,ts, m to sqplement or 'cenfim Mr=9 failure mechanisms. In Ihase I these are limitad to screening type examination and testing. 'this work also assists in the identification of key performance parameters which are monitored to detamim the ongoirug effects of aging.
Test samples may include equipment ard -vcc su,t. runswed from service at operating IHRs, and from mothballed or decommissioned reactors. Depending on circumstances, the examinaticais or tasts are conducted in-situ, on-site, after v4==t is rencwed; or at various laboratories with appropriate test and examination capabilities. Se Shippingport PWR, now in dacczamissioning, is a source of test ' specimens for the NPAR rug-. 'Ihm candidata wycc=,ts for examinaticri and tasting have been identified throupi sita visits by NRC ard contractor experts, reprasanting a range of hiplines and interests. Detailed infomation for each specific wiri-,t are developed ard used to further assess its relevance to commercial INR systems.
-A.2.1.4 IsI-Review and Evaluaticm me third element in the Phase I assessment is the review of the inspection, surveillance, monitoring ard maintenance practions. It also involves a review of artificial or accelerated aging techniciums used to qualify hardware for erd of life performance. Dcisting methods-for inspection, surveillance and nanitoring are evaluated to determine those .
l methods likely to be affactive in detecting aging degradation in an A-9
incipient stage, prior to loss of safety function. Also, it is iqportant that the methcxis rr>t be unreasonably expansive to implement; ard will not result in unacceptable levels of occupational wre. Surveillance and monitoring methods being evaluated include periodic l inspections, both visual and instrunant-aided, and on-line instrumented l techniques. Also, the evaluation seeks to identify performance parameters l and functional indicators that are capable of + :: 4.ing the functional J I capability of 5 , % . It should be possible to monitor the selected parameters and indicators at operating plants for reasonable costs. A ! review of artificial or accelerated aging techniques is made and ocmpared to data available fran naturally aged hardware to demonstrate the applicability of current practices. A.2.1.5 Interim A=== ment and h.m.ardations
'1he result of a Phase I evaluaticn of a mp-ut, system or structure is an interim aging a=== ment, a defect duracterization, ard an evaluation f of the safety significance of the probable failure modes. An interim evaluation of current ISGH 'IW:hnology is given and lists of potential I
performance parameters and functional indictors are developed. Interim r+>_- .m-daticos are made for Ihase II studies based on the results and a reviews of research activities ocmpleted in Phase I. '!he results of the Phase I study are issued in a technical pag 4&=Es report or a milestone report. Continuation to Phase II activities, for a given ww.nt or system, i is haltad if: (a) an adequate data base and e:perience exist within the industry; (b) industry sponsored gw.s adequately address the research needs; ard (c) resources can be utilized better for other research activities. A-10 1
A.2.2 phase U The Ihase II ====1eit is a larg term effort. It includes validation of advanced inspection, surveillance and monitoring nethods thra.gh laboratory and field testing of marrples; and validation of accelerated agirg - tactiniques. It also includes development of nodels to siza11 ate degradation, in-situ aging ===a==nant ard testing of naturally aged equipnent frm operating nuclear power plants. A.2.2.1 Review and Verification of Imoreved ISDN and In-Situ Assessments The Ihase II research en Ises involves a review of advanced wwUsds and techiclogy, for eacti catAgory of @mits uTder stLEfy. In this effort, advanced techniques and tactinologies, either in use or under development, are investigated. The sources of techrology bcfJ1 within the . nuclear industry and outside the ruclear industry are utilized. The scx.irces
outside the nuclear industry include fossil plants, the p Lv-ct=4m1 industry, the ae::ospace industry and various branches of IrD and other government agencies. Also, the practical feasibility of applying these technologies.to nuclear plant werits are explored.
O laboratory and field application and verification tests of ISDH candidate techno1cgy are carried out. The objective of the tests is to demonstrate that: the methods are appropriate to folicw the dynarles of the perfomarce parnetars an$ functicnal indicators of interest; the methods have adegaate selectivity (will not give false indications) and sensitivity (will detect in the incipient stage); and suitable acceptance / rejection criteria are available so that maintenance needs can be correctly identified. The laboratory tests involve sim21ation of defects, of varying % es of intensity in prototype hardware, to determine sensitivity and dete:: tion criteria. Various defect and envim wit combinations are usal to determine selectivity. These laboratory tests are carried cut to verify that tim A-11
nethods are applicable for in-situ use at pcuer plants. The field tests are reocrnarded at cooperating utilities in order to: confinn the laboratory results; prtnide information about the fre;;uerry and methcd of data collection and analysis; and estimate cost effectiveness and practicality of application. A.2.2.2 Testino of Naturally Acad Cuar.rsts A namrxi alament in the Ihase II assessment is the examination and testing of naturally aged u.yr ta obtained fzun operating power plants.
*2his researth element is perhaps the nost oost intensive and difficult element of the NPAR pupaa. Yet, it is essential to quantify aging and determine that adequate safety 2nargins exist to assure the operational readiness of naturally aged -p-its ard systans during design basis accident situations. ,
considerable effort is needed to aogaire naturally aged equipment for examinations and tests. Equipment which has experienced significant operatirg and envim.-tal stressors are being soucAt frun vazious sources. The ocuroes under consideration incluse connercial operating plants, decamissioned facilities and reaaarth reactors. A major thrust of
- this researth elanent is to evaluate performance of aged equipnent before and after it is subjected to the stressors ard eruim...crr21 conditions expected under accident conditions. 7he evaluation is based on follcuing the dyntnics of performance parameters and functional indicators, which were identifia$ in ihase I activity.
In-situ sonitoring of operating agaipment at IHRs is reccmended to gain an understanding of the interaction between agirq ard service wear defect characterization ard inspection, surveillance and maintenance. When available, agirg um%ts are performed on eqaim which has failed during operation, as well as on egaipnent which has survived extensive periods of operation. This is done in order to gain an understanding of those agirg effects which would only be excited durirg a tricyger event acompanied by abnormal stresses. A-12
i 1
, .: 1 A.2.2.3 Residual Life Evaluations j
')
Residual life evaluations will also be performed using data generated for major wgs TJ or test specimens from major w-As, if available. Work is already underway in other programs to evaluate the 4.e= tion of the steam generators removed fr m Surry (Reference 13). The structural i integrity of the reactor pressure vessel has also been the foms of a large and ocntinuing effore (Referunos 14). Results fra these researth efforts, and similar offorts, will be integrated into the residual 1ife evaluation of the major 1HR -ycimiJ for use in considerations of plant life extensiorVlicense runsual. ! 1 A.2.2.4 Service Life Pr=H etion Met W e A third element in the Base II assessment is the development of service life prediction el a. 'Jhis includes a ocmpilation of mrrently , used methods and an evaluation of their applicability. The effort will include testing and examination of aged ocrrponents, fra a participating power plant, and ocmparing these results to the service life predictions hammi m the actual service history. 'Ihis effort also includes the development and qualification of residual life models, for the major w yor- tJ of nuclear power plants. a A.3 Application Guidelines As can be seen in the NFAR Strategy flow cinrt, the researth performed from the Base I ard mase II ======ments will lead to the develepnent of application guidelines for standards; and to reocrenendations for improved Ist.MM practices. It will also provide a systematic collection of historical base line data and trending informatim for eva'tuating wayci i and system aging effects. 'Ibe specific areas of application, shown in Figure A.1, provide highlights of the and uses of the NPAR rs-i. A-13
.__________--__O
A.3.1 value Imonet Study and coordination With Users In.the development of application guidelines, a value igact study will be performed and an interchange with the various erd users of the information is planned. For exagle, prior to generating application guidelines for igioved ISOM te&nology, a u Md effort will be made to interact with the NRC staff, code .and standards muittees and industry groups, as indicated in Figure A.l. Value inpact stuides will be made of the cardidate surveillance and acnitoring methods for %iads2 ccmponents considered to have potential for arventual iglementation at operating plants. 2ese will be evaluated to detemine the occupational exposure likely to occur in conjunction with such methods. De study will assist identification of those methods whid are cost effective and practical, for application in a camercial plant envitu.. at. A.3.2 SsW M1ution of Generic Safety Issues
'he NRC report, NUREG-0933 (Reference 30) contains a Ir - = dad priority list to assist in the timely ard efficient resolution of safety i
issues that have a hicgh potential for rMw-ing risk. De NPAR gwims will generate guidelines, develop criteria and agport NRR in the resolution of the generic safety issues. De NRR generic safety inan== which would d4nctiv benefit frtzn the NPAR swimu results were listed in Table 3.1 (see Section 3.2). De NRR safety issues whid would irviinctiv benefit frczn the NPAR g%imu results were listad in Table 3.2 (also see Section 3.2). A.3.3 &nsiderations for 1.4fe Extension 1 1 An imortant objective of the NPAR Fi,-u is the identification and a
)( resolution of the ta&nich./ safety ins === involved in requests for license renewal of ruclear power plants. De and product of the NPAR research will be guidance or tv-- adations to NRC users on subjects such as: revisions A-14 i i
j
- to Is&MM methods, resMm1 lifetimes of major -p ra, key technical information required in applicaticms for license renewal, and assuring' contirmed safe operatial of plants with license renewals.
A.3.4 Guidelines for Wien. Surveillance, and Maintenance l
- Another and product of the NPAR Fa% om is the evaluation of the role L of maintenance in mitigating aging effects and development of gnihlines for revised or preferred maintenance practices. 'Ihis effort consists of the following activities
- review carrant practices and rh, review vendor's m--Mations,.' evaluate merits of performing mair tanance un$er varying conditions to anticipate malfunctions, identify failures caused by maintenance sw.iares, and develop r+>- =&tions for a preferred maintenance approach.
A.3.5 ouldelines for service Life Prediction 6 , 9_tih14nes will be developed for service life predictions of aged Aw ta and systems. Inprtnad service life prediction methods for electrical wp ta and residual life mials for major creper. ants are to be developed and qualified in Phase II. Using these methods, guidelines will be developed taking into special account the effects of plant operating history. 'Ihese guidelines will be of particular use in the areas of license renewal and maintenance and surveillance. A.3.6 Revwendation for star =4a%and Guides
% NPAR swys.- will develop reactanandations for the revision of relevant industry codes and standariis for continus' aged p3mt operation.
'Ibe NPAR rvvi.m will also provide a tactinical basis for paparing NRC l regulatory guides and review rue Cares concerned with the continued operation of aged plants and also for license renewal considerations.
A-15
A.3.7 Dissemination of Technical Re mits
'the romaad infomation developed in NPAR is Iming d4===minated thztugh preparation of technical papers and reports, sponsorship of workshops and synposia and information exchange swi==. 'Ihe program also includes the establishment of an information data bank Whis will be available to NRC user organizations and to other user groups (such as utilities, manufacturers ard laboratarias).
A.3.8 Innovative Materials and Desian
'Ihe last application shown for the NPAR prcgra:n strategy (in Figure A.1) is inrovative materials and designs. Here, r+> -< ardations would be provided (and up to the industry to inplement) to evaluate design charges to existing -pre ts, system and structures, Which would make them '
less 91=,aptible to aging induced degradation. It would also firx1 end uses in the other NRC swtams (such as Timant Qualification and Advanced Light Water Reactor Designs). O 4 A-16 4 I l
APPENDIX B M7L70R NPAR PROGRAM EIDENIS
'IABE OF O2CDES i
B.1 Agirg - system Interaction study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-1 B.2 Agirg Rama==mant of Cearpments and Systems . . . . . . . . . . . . . . . . . . . . . . . B-2 B.3 Aging Assessment of Civil Structuras . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-2 B.4 Inspection, Surveillance and Monitoring Methods . . . . . . . . . . . . . . . . . . B-4 B.5 Role of Maintenance in Mitigating Agirg . . . . . . . . . . . . . . . . . . . . . . . . . . B-5 B.6 kgent Lifetine Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B-6
- B.7 Investigaticri of Aging / Seismic Shock Interactions . . . . . . . . . . . . . . . . B-8 O
a f B-i i
APPDOIX B 19JOR NPAR PROGRAM E12MENTS Appeniix B contains a Won of and of the eight major subjects or technical areas beig addressed in the NPAR Ftws-u. B.1 Aoire->#- Interaction Studv 1 A,$m resear& oas initat.d in tha ama to detemsne w agim is . affecting the levels of plant safety at operatig plants. m is problec is being addressed by establishing the relative contribution to risk fra age-related system failures. 'Jhe initial approad is a data collection ; effort to identify agig failures over a broad category and the identificaticm of root causes or specific nechanisms of failures for specific systams and ww-as. While the data collection prtwides information about aging failures of systems and w 2.s, the data is .not particularly useful by itself. Aging is important only when it changes the overall levels of plant safety. m is type of determination is accomplished through the use of FRA techniques
.; ocupied with time C-i A t propagation of aging effects.
he results of the researd effort is to establish the importance of various aging related failures or mechanisms to core melt frequency cr plant risk, his pnnidas a method to identify where (in plant [fetyk.mos, y gp % and he E-.ts)c agig is significant to risk, and prioritize systems and +=r.ts for ii.LyG4 aging studies. A second application of this work is in inspection and maintenance. Systans and conponents, ! susomptible to age related failures, can be identified so that correcedve action can be takan. 1
)
B-1 e
. l
1 i B.2 Amirxr Ac====mant of cuversts ard Systems ; Detailed engineering studies of agirg effects in selected systems and u p is are being conducted. She objectives of these studies are: to identify age related failure sectanisms in systmas and %-.ts; and to J investigate the ability of aged ww-ia and systams, omsidered vital to plant safety, to perform their required safety functions &rring or after transients and =~4%. These efforts include an evaluation of operating experience, aczioning type agirg assessments and mMsite and laboratory testing of naturally aged aq$4==nt; and in sme cases, laboratory aged equipment. I The engineering study of aging effects on -p ra and systems will result in the following information:
)
o Identification of the age related failure mechanisms for the major , . systems and u p ia. o Identification of stressors due to envim.iat, maintenance and operaticn.
. 5 Identification of aging sites, failure modes and effects on plant ~'
safety or operation. o Identification of methods to &ttact and control aging degradation. B.3 Aains Aa=== ament of Civil Structures 1 7he ctdoctives of this research are to: identify structural safety issues, that the NRC will need to M an plant license extension K applications are reviewai; conduct confirmatory rieaarch to validate long term linhavior of structural material [ Man subjected to internal and g external stresacts; and determine the short and long term impacts of a-g post-IDCA (or SMI-2 like) containment envim..et on structural materials.
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An easy to use data , en aged structural materials' pipi ,' I will be generated frt:en available data, includirq data cbtained using samles tram f =- --- A =ioned facilities. Inservice inspection programs will be used to provide deterioration trends. A methodology will be developed to quantitatively assess structural reliability, as affected by aging. From these, NRC license reviewers will be provided with review issues and acceptance criteria for use in structural reviews of plants, for license extensicm riequests. Preliminary work, already empleted in other researt:21 programs, is prt:viding irput to this resear1:21. These include: NUREG/CR-4652 (Refererum 15), "Curn.i=te Cuyc. uit. Aging and Its Significance Relative to Life Extension of Nuclear Power Plants", September 1986; and a report (Reference 31) by the U.S. Army Corps of Engineers, Vicksburg, on empleted tasting of a concrete sample fran the shield wall of the '4a~==4=ioned reactor at Gui.L-.migen, FRG. The sanple (frun the 3atter report) shows no , significant effect en its structural properties after about eleven years of operational irradiation. i
'Ihe results of this raamd1 effort will include:
- 1. More advanomi understanding of structural material behavior, including envitu. ; ital effects of time, tanperature, radiation, noisture, and cha@21 interactions.
- 2. Improved understanding of the presence, or absence of synergistic effects on cyclic or fatigue loads; and their effect on the useful life of structural elements when ocabined with service or accident ;
t sirg levels) . loads (e.g., effect on g ^u
- 3. Develagnent of inprove$ icre^Mye techniques for examinirg concrete structures and yi-u ssing systems for defects,
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deterioration, or ria n . B-3 f
i s
- 4. Definirg and examining the unique envim.w.atal conditions that exist in a post-lLG contairamnt, ard the effect on structural materials of unforeseen streso or corrosion, at a time when fun:: tier:al reliability is essential; but access for inspecticn and maintenance are not pract.ical.
Regulatory applications of this researdi are as follows: I o Imprermd predictions of 1 cog term structural deterioration. o Imprtwed predicticrs of available safety margins at future times. o Limits on hostile erWimmuunt&1 exposuIts. o Baduction of licensirg reliance rm inspection and surveillance. - o Informed rwiew and approval of plant license extension applications. o Incorporation of r-rch results into national design and inspection staniards referenced by the SRP. a B.4 Insoection. Surveillance and Monitorim Methods A principal element in the engineering waluation of aging related degradation is the a==*= writ of methods used in inspection, surveillance and monitoring of nuclear ruclear plant systems and w,ents. In Phase I
/ ,
a rwiew is made of methods c2rrently in use. Also, an evaluation is provided of their effectiveness in detecting agirg degradation, in an ! incipient stage, prior to a loss of safety function. In Ihase II, a review is made of advanced technigaes. B-4
The Phase II efforts include a review of: sources, both inside and L outside of the industry; and advanced techniques or te&nology. Following this, candidate methods are chosen for laboratory and field testing. The objective of the testing is to show that the methods have adequate selectivity (will not give false indication) and sensitivity (will detect Nr dation in the incipient stage). Laboratory testing is carried out to demonstrate how advanced methods work in a omitrolled testing envim.w. int. Successful candidate methods are then subjected to field testing to confirm the laboratory results and to provide information en practical field applications of advanood methods. In both the Phase I and Ihase II anm=ments, performance parameters and functional indicators, that can be used to identify age related degradation at an incipient stage, are identified. In some cases they can be used to assess the severity of a problem ard its specific cause.
. 1 B.5 Role of Maintenance in Miticatirn Acina l Both the Phase I and the Phase II efforts,will include an evaluation of l the role of maintenance in mitigating aging effects. 7his effort will consist of the following activities:
- a. Review current practices and pr M ires carried out by nuclear utilities to maintain equipnent. Otrisider eas ue@#nt selected for aging mana==ments and r-+- =rd maintenance methods to assure safety. For completeness, also include additional un p rw As considered important by the utilities.
- b. Review nuclear vWnt vendors' r+ - =daticris for maintenance of u ,ss,ts or sA 2 :-:-r==As selected for aging a==*==ments.
B-5
- c. perform an evaluation, including a cxrparative analysis, of the relative merits of: (a) performing naintenaroe When a cuw,ent has been disocuered to be m1 functioning (cunedve mintenance);
and (b) perinruing maintenance When an observation has been m de through surveillance, inspection, or nonitoring, that a -ex=it may not furction when required during a design basis or " trigger" emnt (praventive maintanarce). Enphasis is placed on the relation between failures (causes or modes) P=1 to be experienced &xring operation and those Which would potentially occur urder the stresses associated with design basis or trigger l events.
- d. Identify, tere possible, those wwent failure mechanisms likely to be in+v=1 through preventive or cuit ctive -
mintenance. Specifically, look for those which might be . detectable through short-term, post-maintenance surveillance, inspection, or monitoring.
- e. Develop rz-- rdations, for acceptable or preferred maintenance practices, haeai on the foregoing activities.
; In all <acac, the euphasis is en the technical or hardware aspects of J
maintenance rather than cm institutional, organization, pwp-atic, or I human factors considerations. B.6 Cuwimit Lifetime Evaluatien l An evaluaticn of the age degradation ard residual life of major INR J
, -prents is being performed in the NPAR pwpcun. 'Ihe information I
generated in this reenarch effort has two principal objectives. One is to assist in developing criteria Which assure age reinted degradation of major w.e ,ents x does not impair safe plant operatico. 'Ibe second is to generate a technical basis for establishing criteria and develeping guidelines to be B-6 l 1 I I
used in licensig review procedures and plant life extensierVlicarse runawal. '!he effort is ceriplementary to the on-golig industry spczisored pilot projects on plant life. extension. However, the NPAR effort is fundamentally safety oriented.
'!he approach used in the residual life mana==mant proj , is to first X identify and prioritize the major % d.s with r+t to safe plant.
operation. 'this is followed by an initial effort to establish the life . limiting processes for and of the major cxmparants. Included in this effort is the identification of G-p. 2=ticri sites and failure modes during normal operation and accident conditions.
'1he initial effort also includes an amma=M of current and potential methods for inspection, surveillance and monitoring. For this phase of the effort, the work is focused cri integrating currently available technical information which is relevant to aging and life extension. .
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'!he results frtzn the initial assessment are then used in the develepnent of sirple me&anistic natals for determinig the residual life of selected major v ,c id. Developing these unia1= is of particular interest for + la which are not ream 1y arv===ible for routine
- maintenance and inspection. As these models are developed, resichaal life evaluation are planned for the major u ,g ta using actual plant cperating data. In this phase of the project,. key plant operating data will be identified. 'this is the operating data which is mt9f for a realistic (rather than conservative) estimate of the mechanical and thamal 1 ceding of the -es -its , as well as other envim d.a1 stressors. 'Ihis segment of the NPAR program is coortiinated with the ongoirq researt:h programs involving vessels, piping, steam generators and rgr(4-Lactive examinaticri techniques. B-7 l 1
B.7 Investigation of Acira/ Seismic Shock 1 teractions k An understarxiing of the vulnerability of age %.kd equipment to seismic disturbances is r* *===_q for the design life of a ruclear power plant. 'Ihis includes the original plant life of 40 years and any extended life period. Current industry standards (IEEE 323 and 344) require d i pre-aging before =4=ic qualification of electrical M mant. However, the NRC has not determined such a need for mechanical +=as and is currently evaluatirug the significance of aging as a factor in the qualification of regulatory guides. 'Iherefore, an assesseent is needed of l the potential importance of aging in degrading seismic performance of equil mat. ( Both the ruclear industry and the regulatory agency have on-going pwgans to assess the aging-seismic effects. 'Ibese include: o Iaboratory testirg of naturally, as well as artificially, aged mp= a.s. l o QMification of equipnent usirs existing test data, j
; o Usirs experimental data for qualifyirg wp-is, o Development of seismic fragilities for different wwents, o Identification of weak links in certain eqdreant accamblies.
o Development of surveillance and maintenance rup-s:, to alleviate l the agirg effects on seismic performance of Sd ment. In order to avoid duplication efforts, some studies involve both industry arri NRC. Recently, it was detemined that the qualified life of some equipment may have to be extended in the event that a utility submitted an application to the NRC for a license rernwal or plant life extension. In that case, the wp-ta originally galified for a 40 year period would have to be r==W for extended life. B-8 i
APPENDIX C NPAR RESEARCH ACTIVITIES TABLE OF CCtTIDfIS C.1 NPAR Paaand - PNL Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-1 C.1.1 Shippingport Reactor Aging Evaluation . . . . . . . . . . . . . . . . . . C-2 C.1.2 Aging Aaaaaenent and Analysis of Snubbers . . . . . . . . . . . . . . C-5 C.l.3 Diesel Generator Aging and Life Extension Aaaa==mant ............................................. C-6 C.1.4 Service Water System Aging Studies ..................... C-8 C.1.5 Aging Aaaaaenent of Rocn Coolers . . . . . . . . . . . . . . . . . . . . . . . C-9 C.1.6 A Practical Approach for the Quantification of Aging (QOA) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-9 C.2 NPAR Panaamh - ENL Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C C.2.1 Electric Motors ........................................ C-11 C.2.2 Battery Chargers and Inverters . . . . . . . . . . . . . . . . . . . . . . . . . C-13 C.2.3 Circuit Breakers and Relays . . . . . . . . . . . . . . . . . . . . . . . . . . . C-14 C.2.4 Motor Control Centers . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-15 C.2.5 Residual Heat Remcual and Cbnponent Cooling Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-16 C.3 NPAR P*amarch - INEL Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-17 C.3.1 Evaluation of Aging contribution to Plant Safety ................................................. C-18 j C.3.2 In-Depth Engineering Studies of Selected Systems - l Reaction Protection Systems, Class 1E Distribution Systems, High Pmssure Cbolant Injection Systems . . . . . . . .C-21 C.3.3 Residual Life Assessment of Major Cup =nts . . . . . . . . . . . C-23 1 i C.4 NPAR Researth - CRNL Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-25 { 1 C.4.1 Aging Assessment and Analysis of Auxiliary Feedwater System (A ns) ................................ C-25 C-1 j 1 1
F.. C.4.2 Aging Ammanament and Analysis of Motor Operated Valves, Check Valves, AFW Pungs and , Solenoid operated valves ............................... C-25 ! C 4.3 7% sting of Naturally Aged Solenoid Valves . . . . . . . . . . . . . . C-26 C.4.4 Diagnostics and Mcnitoring of Reactor Internals - Structural Integrity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-2 6 C.5 NPAR prch - Other Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-27 C.S.1 NBS Stufy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-27 l C.S.2 SFA Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-2 8 C.5.3. SNL Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-2 9 nsIzs l C.1 Shippingport Station Cwpau ds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-4, - 1 C.2 Systan Analysis Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C-19 t l C-il
APPENDIX C NPAR RESEMOI ACTIVITIES
'Ihis apperdix cantains a description of the varicus researd studies being performed as part of the NPAR piu mu. '!he scope of ead of the '
elements in the program ard their current status is Mammaad. 'Due major NPAR work underway at this time is being done at the following National laboratories. o Battelle-Pacific Northwest laboratory (PNL) o Brockhaven National Imboratory (INL) o Idaho National Engineering laboratory (INEL) , o Oak Ridge National laboratory (ORNL) k- __ Support for the NPAR Pingmu is also being provicht by the Franklin Pacmarch l Center (FRC) as a subcuhactor to CRNL and INL, Systems Engineering Associates, the National B.treau of Standards and Sandia National laboratory.- t In addition, work is =*- -hocted to various private engineering firms and private and academic consultants to make use of special expertise.
'1he following sections describe the major research activities of the NPAR program.
C.1 NPAR Pa=mard - PNL Activities l' 1 Paaaatt:h is being perfemnad a1 six major elements of the NPAR program by PNL, 'Ihe six elements aIn (1) Shippingport Reactor Aging Evaluation, (2) Aging A==a== ment and Analysis of Srnhhats, (3) Diesel Generator Aging and Life Extension Assessment, (4) Service Water System A==a==marit, C-1
i (5) Aging Amaa== ment of Rocn Coolers, and (6) Quantification of Aging. We following describes the scope of work of each of these elements. C.1.1 Shicolium Panctor Actinct Evaluatign
'Jhe Shippireiji t Atomic Power Staticm, now urduyving C---- 4==ioning, is a major source of naturally aged =?'4=='it for NPAR ww.t and system evaluations. As the first U.S. la% :: ale, central-station nuclear plant, the Shippingpart Station parallels ocmercial IMRs in reactor, steam, I a w414=vy, support and safety syst a s. With its 25-year service life
,j (1957-1982), it covers almn=t the entire time span of currently operating l reactors. Also, because of substantial modifications during the mid-60s and 1
{ 70s, it offers unique examples of identical or similar aq$4=arit used ] side-by-side, but representing different vintages and degrees of aging. { l
'Ibe objective of this NPAR task is to perform in-situ ama*==ments, acquire selected -yci Ts and sanples, obtain data and records, and ocnduct post-service examinations and tests of Shippingport Station aq'4=arit and materials in support of NPAR and other NRC yswuaus. A summary description of this effort including accanplishments and status is given below:
G PNL ocordinatas closely with the DOE Shippingport Station W4==ioning Project Manager and designated site personnel to incorporate the activities of programmatic interest to NRC in the overall site deocznmissioning plans. Work is in progress to obtain all available data and records for the systems, ww,.nts and materials selected frtu 4 l
'Shippingport. 'Ihis includes designs and specifications; operation and {
maintenance manuals; operation and post operation histories; maintenance history / record; inspection, surveillance and periodic test procedures / data; etc. F4wt.ss includes the acquisition of more than 50 technical manuals for plant -per ta, many original aq$4=arit and natarials specifications, and the maintenance history and record of ctanges for key +,ents. Se C-2 l i;
information for several selected systems aM -psts has been cer: piled and distributed to the assigned NPAR contractors to sLpport their evaluation studies. Arrar gsats are coordinated and site services are obtained as required f 1 to support the in-situ ==amaamant of systems, c , - ta and materials prior to their removal by the -f+ --- 4=1oning operations contracter. These in-situ assessments include visual and physical examinations, testing of electrical circuits and -,cr As, response studies of varicus el Lv ianical devices, and different types of re,1i-Lwiive examinations and tests. BMG for exmple, has conducted a etsprehensive
'in-situ evaluation of 46 Shippingport Station electrical % ta and circuits representing more than 1600 individual measurements of insulation resistance, DC loop resistance, total capacitance, total, indactance, and 4= =4ance. In additional, the ferrite centent of cast austanitic stainless steel primary system main valves and coolant ptmp volutes has been measured .
in-situ to identify caMWte materials for I:ES-MB thermal embrittlement
, studies. 'Ihese in-situ maasta-=(a indicated that nine of the 24 cast primary system u w As had sufficiently high ferrite levels to make them of interest or acquisition for detailed materials studies. - Arrargs-As are coordinated and site services are obtained as required 6
to support the acquisition of wycrnts selected by NRC and its contractors for off-site evaluation. '!his includes the identification, removal, packaging and shipment of wversta obtained in conjunction with the v:---,4mmicriing operations and also the retrieval of selected mycrsts after shipment to Hanford for M&1. More than 100 Shippingport Station wver As have been selected for NPAR evaluation and testing through site visits by NRC and contractor experts representing a range of disciplines and interests. Table C-1 oontains a listing, grouped by contractor, of these NPAR itans and other u g As that are being acquired for the Materials Branch as part of the NPAR site coordination effort: C-3 i j
( 7ABLE C-1. SHIPPINCGORT S'IATICH CDGWDTIS Offsite Evaluation Number of Items: Selected Removed Shicoed Pacific Northwest T
- _+etriv (PNL) o PV Nozzle C12 touts - 5 5 5 o . Coolant Purification Piping 2 o Rad Wasta Piping 2 o 4/D Instrment Piping 2 o Fuel Pool Piping 2 o Main Steam Piping 1 1 o Feedwater Piping -1 1 ;
bus National TaMatorv (ANL) o Main Coolant Pmp. 1 1 o Check Valves 4 2 o Manual' Isolation Valves 3 '
<. Hot lag Pipe Section 1 -
1 o Cold lag Pipe Section 1 1 Brooldviven National TmMratory (EtE) o Motor <ienerator Set 1 1 1 o Battery Chargers 2 1 1 o Inverters 3 2 2 o Motor Control Canter 1 1 1 o Differential Relays 2 o Protective Relays 4 o Agastat Relays 5 1 1 o Scram Brunkers 2 o W,-6 Relays 4 2 2 o DB-50 BrquW.ars 2 o' Circuit Breakers 8 o Current Transformers 2 j o Potential Transformers 2 o 480/120 Transformers 2 o constant Voltage Transformers 2 o Relay Panel 1 o Spare Parts for O)arger/ Inverter 1 box 1 box 1 box HMG Idaho, Inc. (INEL) o Motor Operated Valves 2 2 2 o Limit Switches 8 8 o Battery cells 6 o Nuclear Instnamentation Channels 2 o Electrical Panel 1 o Thermocouple Signal Box 1 C-4
TABII C-1. (Continued) Offsite Evaluation Nime=r of Items: Selected Russ:Ned Shicoed E&G Idaho. Inc. (INf1) (continued) o 'Ihsams.s@le Junction Box 1 o' 14: war Imad Junction Box 1 o M rminal Strip 1 o Power Cable 1 o Instrumentation Qible 2 o ' Rod control Junctimi Boxes 2 o' Selector Switch 3 2 2 o Pressure Switches 7 o Rosemount Trprwrs 4 o RIDS 16 o D/P Onlis 6 1 o Transmitters 6 1 o Invel Indicator 1 o G.p.ating Icn Chamber Detectors 4 , o BF3 Detectors 4 Oak Ridae National TmMratory (ORNL) o Solenoid Valves 7 4 o Motor. Operator 1 o Motor Operated Valves 5 o Check Valves 4 1 1 Acquisition of many of these ww-d.s should be ocupleted by the end of FY-1987. F6,6 will be reported in a milestone report. C.1.2 Acincr _1= cama = ant and Analysis of Sn*+*vs An aging assessment is being performed on the hydraulic and mechanical sru*+*vs used in nuclear power plants. 'Ihis a==a== ment is being done to establish failure medianhma and causes and to provide reconnendations for practical cost effective inspection surveillance and maintenance methods.
'Ihe Phase I assessment has been otmpleted and the results are @manted in the Phase I report entitled " Aging and Service Wear of Hydraulic arx1 Mechanical Smbbers Used on Safety-Belated Piping and C+-As of Nuclear Power Plants" (Reference 16) .
l C-5
The~ 1. putt p.e=Ws an overview of hydraulic and mechanical snubbers based on information from the literature ard other sources. Snutber operating experience is zwviewed using licensee event reports and other historical data for a period of more than tan years. Data are statistically ] I analyzed using arbitrary snubber populaticos;.the implications of the observed trends are assessed, and recxzmendations to modify and inprove examination and testing gue-lures to enhance snubber reliability are determined. Value-impact was 21so considered in tanns of exposure to a I
' radioactive envim_.t for examinaticrVtasting and the influence of lost snubber function and subsequent tasting rvys o expansion on the costs and operation of a nuclear power plant. Last of'all' optimization of stu***r populations by selective renoval of ure=mry arn***vs was also considered.
The current Phase II r==aardi is being conducted in accordance with the NPAR pivysmu strategy starting with a comprehensive aging anua== ment, post . 1 service examination and laboratory testing of aged v4.= ant. She effects of accident ocnditions, e.g., seismic and IDCA effects are to be included. A significant effort is planned during the testing phase to evaluate performance indications for snubbers. The results of this last effort are intended to assist in identifying practical and effective performance indicators. Advanced methods for snubber maintenance arx!! in-service evaluation will K be a=an==ai. These results along with the Ihase I M+- [ current results will be used in develeping application guidelines for inspection i surveillance and monitoring methods for ar=*+=r. This last activity will AP p include (ocnditio)nsith NRC, operating utilities, and appropriate codes and (ad s f **# stande.rds carnittees. C.l.3 Diesel Generator Aaim and Life Extension Amea== ment 1 A nulti disciplinary evaluation is being ocricluded on nuclear service i diesel generators. This task consists of a limitad aging assessment of i f C-6
f l
. diesel generators and related -pis.nts.- Also included is the evaluation of surveillance, inspection and maintenance nethods used on diesel generators and their role in mitigating aginy effects.
'Ihe Phase I effort en this element has been ompleted. In Phase I, an
' evaluation was made of current operating experience and an interim aging
====== ment was parfanned. A large datahnam was established and placed into L a amputer managed format for development analysis. 'the draft Report on l
Interim Aging Assessment was subjected to an industry wide peer review in a Diesel Generator Aging Seminar. After the peer review, selected results of the workshop were ik.v+roted into the Phase I report and a revised final i version was inand (Reference 17).
'Ihm gt----Mngs of the diesel generator aging seminar /%disq have been documented in a separata e t.. A draft ist. has been issued for a technical review by the participants. Once the review is ocmplete, the -
results will be issued in the final form to the NRC and participants. 'Ibe remaining work includes the Phase II aging evaluation and the development of application gnWHnes.
'Ihe Phase II effort extends the evaluation of aging ===a== ment started ; in Phase I, including selected studies of specific installations and the testing, inspection and diagnostic instrumentation in use. It will include a review of racticas and r+ -- =-4ations for inproved X
surveillance and nonitoring for early detection of aging effects. Ihase II will also include a study of current and potential service life prediction methods.
'the Phase II effort will lead to guidelines for inp ISPAM thods y leading to mitigation of aging effects. It will also provide a cal basis for establishing the socpe and documentation requirements for license renewal sutaittals. Another benefit from this effort may bis improved reliability of aayiucf diesel generating systems. Scne of the current testing and inspectico requirements for these systems have been identified as having an adverse effect on reliability and availability. 'Ihis effort C-7 i
1
includes an investigation of alternate methods of inspection ard altemate testig schatules which will result in better overall performance. c.1.4 Service Water Systen Aoirn Studies
'Ihis element ocnsists of an aging mana==mant of the safety related armas of the Service Water System of a nuclear plant. ;
'Ihe function of the Service Water System (SHS) is to transfer the heat' loads frm various sources in the plant to the t:Mlmate heat sink. The three safety related heat sources sa_rved try this system are identified to be:
o ccre decay heat j o Decay heat removal -i@-,Us o Emergency power scuroes Ibe to the wide variation in the nature of each plant's ultimate heat sink and the application of a m21tiplicity of system design approached, the system is defined from a functional standpoint as: All m ycr= ds, their
, associated instrumentation, miwls, electrical power, cooling and seal ; water, lubrication and other =~414aq equipment which caprises the final heat transfer loop between the safety related heat sources and the ultimate heat sink.
A Phase I assessment is currently being performed cn this system. A detailed task plan is being sq-M. 'Ihis will be followed by two I
. site visits to acquire actual SHS operational data. 'De data will be analyzed to produce interim ruoamerdations with respect to:
o System inspection guidelines o Cwpsiit monitoring methodology (surveillance) o Systematic approach to cw p ent maintenance l o System life extensien regulatory requirements C-8 l L _ _ _ _--- - _ _ _
'Ihe results will be documented in a NURm/m Phase I report.
Additional work en this element will follow the NPAR phased approach. - Ihase I ocnclusiens Which require additicmal verification or investigation will be ptrrsued, along with any further analysis required. She and result
' of this task is to produce a set of recamanded guidelines and to support the specific NRC license renewal regulatory uses. All-Phase II data and
. Agdng requirements will be satisfied prior to the end of FY-89 to insure timely support of the NRC life extensien effcet.
C.1.5 Mim Assessi=nt of Room Coolers
'Ihis element consisted of an aging amamnenarst of room coolers. An NPAR Ihase I study on roca coolers was ocupleted in FY-1986. 'Ihe results are sumarized in a report entitled " Operating Experience 'and Aging Amaaaament -
of ECCS Pimp Boom coclers" issued in October 1986 (Reference 18). Roan cooler operating experience data obtained during this sttriy suggest that punp room cooler operation has been relatively trouble-free. Licensee event report (IER) data indicate that room cooler failures azu
; rare. Room cooler failures usually develcp outside the roan cooler boundary in the notor ocntrol center electrical + sis or in the service water system driller, valves and punps. These + e ra are subjects of aging aaaa==nants in other NPAR tasks and, therefore, no duplication of effort was made. l It was reconnended that Ihase II of the ECCS punp rocan cooler aging
- mana==nant be delayed until NPAR system amamaanants ncu underway have further addressed the significance of room ooolers and more extensive operating experience is available.
C.l.6 A Practical Amroach for the cuantification of Mim (00M l
'Ihe objective of this Task is to develop and As.trate a practical approach to quantify the effects of aging on the safety margins for j C-9
safety-related reactor and==nt. A safety maztjin is a maa=N of a ww d.'s level of safety ompared to a f:hosen operatig limit, and is useful when data are insufficient to a===== the remaining. life. 'Ihe primary activities in this task are to develop a general methodology for a=====ing safety margins, demonstrate the method on a selected reactor safety ww-it, and integrate the results with the NPAR Faws.- Plan.
'Ihis task belongs to the catagory of special'topim and does not' follcu the usual NPAR Ihased Approach.-
'Ibe initial develcimient of the practical approach for 01 was ocmpleted in FY-1986. One Inajor result of the early work was the identification of a )
general definition for a safety margin: the difference between a selected operating. limit for a chosen performance indicator and the current value of that performance indicator. 'Ihis definition is applicable to a wide range of operating limits, performance iniicators, and v4==nt types. .
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'Ibe current work on this task involves refining the methodology developed to data. Along with this, several electrical and anchanical q u w J.s will be selected and an evaluation will be made to dennstrate on a preliminary basis how this approach will be applied, including the
. development of an engineering model. 'the latter may include statistical data fztra in-situ a=====mants, post-service examinations and tests, trending of performance parameters (functional indicators), and ocntrolled laboratozy testing. Such a model is needed to determine if a foundation for I extrapolations of trending. data can be established. Ramad on thes.t results a generalized methodology to quantify aging of reactor equipment will be generated for caparing v4==nt cperating critaria to expected performance. Future work is expected to address refinements and applications of the Q3A methodology. C-10
C.2 NPAR Ramaart:h - BE Activities Brookhaven Naticmal Iaboratory is currently acrducting a research i effort on the daractarization and detmeticm of age-related failures of j l selected mycs As and systems. The main objectives of this research effort are sunmarized as follows. The first is to identify ani characterize the aging and service waar effects associated with selected safety synam L + As. Next -is to determine how aging and service wear affects the capability of selected =?'imant to w.M during or after =*4=mic events. 1he third is to determine what methods any be effective in detecting significant aging and service wear detaricration which may omprtaise performance. i l' The systems and hycs-its being evaluated include electric motors, bettery chargers and invertors, circuit breakers ard relays, residual heat - - rumcual systems, and canponent cooling water systems.
'C.2.1 Electric Motors Electric motors of all types and sizes are used to drive punpe, valves, -
fans, and ocntrol devices in a typical ruclear power plant. These
.: ccuponents serve important roles for the performance of normal operations, as well as for the acocmplishment of safeguard functions during and/or after an abnormal or accident event.
During stor operation various parameters such as taperature, vibration, current, voltage, and application agi4=ent output can be utilized to generally ama*== motor integrity. When normal values for these parameters are observed to adversely dange, an incipient stage of l
.i degradation potentially leading to ultimate failure is occurring. ,
therefore, characteristic parameter performance can identify failure podes, .I mechanisms, and causes representative for all types of notors. l i C-11
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' The Ihase I aging assessnent of motors in ruclear. power plants is
@ ocuplete and the results documented in NUREG/CR-4156 (Reference 19). In addition to failure mode assessment, the report identifies functional indicators suitable for monitoring the motor dielectric, rotational, and I mechanical integrities.
A Phase II effort en motors is underway to assess the industry and . regulatory standards /gidd-- for motor performance in power plants. As part of this, a survey on current industry practions was performed to assess the W_M in the 4= ting surveillance and maintenance activities. Also included in Phase II are two test programs, one en a 12 year naturally aged 10'hp motor and the other on a failed 400 hp motor. These were conducted to evaluate the suitability of various test methods whi& can be used to monitor the motor health. The AWJ on these activities are being prepared. The Ihase II efforts whi& includes the above researt:h activities are being dmmat.ad in a NUREE/CR and the report (s) will be pablished shortly. 1he report will provide roccamendations for developing motor maintenance programs in nuclear facilities in order to improve its reliability both . under normal and accident canditions of the plant. Reconnandations for Maintenance Guidelines are now being propered. Periodic tasting, surveillance techniques and continuous monitoring methods are reviewed and assessed. Methods are r- tad for pmIw=nce evaluation and trend analysis, as well as for a value inpact analysis. Current industry maintenance practices were amaammad by reviewb1g the motor maintenance requirements at four ruclear power stations. _ Insulation resistance is always measured, whereas motor operating carent is only recorded in scne cases. Most testing is done for the driven Wmant, such . as MW stroke time or punp speed, whi& gives little indication of motor condition. Trend analysis is virh=11y non-existent for motor condition. C-12 _ _ _ _ _ _ _ _ - - _ - _ = _ _
i*ny 1he maintenance guidel lupa discussion of reliability cente ed X I
, maintenance g presented, alcrq with a logic & art to make motor maintenance deciElcra using RCM philosophy. This logic is' applied to the specific application of ocritainment fan cooler motors. 1he resultant maintenance reocznandations do not require any hardware modifications, but do require more tasting and trend analysis than currently socists.
C.2.2 Bat +=w amraars and Inver+m Nuclear power plants use battery dargers and invertars to sqpply power to safety-related Miimar.t, instrumentation, and controls. A battery
&arger converts alternating current (AC) to direct current (DC) to provide power to DC driven n'i==it and + as as well as to keep the standby batteries fully &arged. . On the other hand, invertars are used to supply Acw to safety related equipment and M'i==nt important to plant '
operation after ocriverting the DC-power source to an AC output. Plant systems such as the Reactor Protection System '(RPS), Emergency Core Cooling System (ECCS), Danctor Core Isolation Cooling (RCIC) System, and the AC/DC distribution system use these devices to satisfy certain nuclear power staticrs safety requirements. 1he Phase I report c:nering the aging assessment and review of operating experience of battery chargers and invertars was published in June 1986 (Reference 20). This report describes the aging and service waar characteristics of &argers and invertars based on a comprehensive review of their operating experience in nuclear power plants. It concludes that eM---x A. failures due to aging and waarcut ocritribute to n'4= ant reliability with potential impact cas plant safety.
. The current effort is fmiami en the tasting of naturally aged invertars and a naturally aged battery charger to determine the practicality and viability of ocrdition monitoring methods cited in the Phase I report.
Researds into advanood inspection, surveillance, and monitoring methods employed in other industries is also being conducted to determine applicability to ruclear plants. Because invertars are extensively used in C-13
ocrputer applications, it is expected that investigation in this area will provide useful information. Inputs to regulation and industry stanciards are expected in this area, especially for the irrverter. I C.2.3 giro11t Breakers and Relavs ( 1
'Ihree types of relays are used in nuclear safety-related systems:
protective, ocritrol and timing relays. Protective relays detect abnormal conditions an the plant's power system and initiate opening of circuit ] breakers to prevent damage to the prctacted VM sudi as motors, buses, { and tranatoreers. Centrol relays are used in the logic and protective I' ' a.ction initiation syicams and are basically te position relays with
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contacts that.tttnsfer position when the relay's cell is energized. Timing relays are und to relay or sustain a signal for a specific period in
\ accordade with system operating requirements.
fwo tfpts of circuit breakers are usad in nLx: lear safety applications:
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molded-case ani metal-clmi. Molded-case CBs are used in lw-voltage
.. applicaticas q.) to 48W for lower level distributicri systems and low power loads.' 'We metal-clwi switchgear CBs t.m used in applications where large u
loads and hiper fault currents are irrd.ved. '1 hey are more sophisticated thaa trolded case CBs and are usi on safety systemswith voltages ranging
,trem 480 V to 6900 v.
Se agirg irAeraction relating to CDs and relays used in the safety injection system will hidescribed in NURED/CR-4715 (to be issued). 'Ihe
.s.it prevalent failure em for the relays and CBs am rariewed and their '
effect on system cperation evaluated. 'Ibe study ccnclutas that failure of a e safety int;cticn train is possible from CB ani relay faihire if adequate 1 maintenaaoe and testing is not performed. Failure of redundant trains is rx:rt expecc.ed frac ocnman trode failure of a particultr type of CB or relay. : However, a sufficient number of different types of failures vern found and 1 this supports the ntsi for a strcng maintenance and test program to prevent j m11tiple egn relaud faiWes. l l
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C-14 j b
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"! The Phase I assessment of circuit breakars and relays has been j concluded and the'rupo:tW% the results is near ocepletion. A -)
Phase II waluation of circuit breakars and relays is planned. Phase II f j vill rwiew inspection, surveillance, and sanitoring methods that can and ( should be applied to specific raaclear power plant' circuit breakers and
-relays. The role of maintenance in a&iwing~ circuit breaker and relay reliability in nuclear safety syst e s will also be finalized.
1; l C.2.4 Motor centrol Centers ; on
, A Ihase I study aging assessment is being performedgnoter armtrol X cantars (M00s) including an investigation into the aging of and of the shp-,ts that constitute a MCC. She boundary that has been defined for the M00 analysis consists of the cabinet enclosing the various devices.
Since the shp-is of M00s can vary depending on the par +4m1mr application, the aging m-===t. includes a runhar of devices that were - identified during the rwiew of various failure occurrences descdbed in the literature. Each subp-it identified within the defined MCC boundary is then analyzed according to the general NPAR strategy. The Phase I researd for Moos includes an assessment of the current
; methods used by industry. for in==+4cri, surveillance and monitor (IS&M) X of MCC performance. 'T.nis effort has included a trip to the Square D y for diamamions with manufacturing personnel on the methods that are currently aspicyed for evaluating the operating performance of Kos and their various u p -,ts.
In addition, a plant tour and rwiew of the manufacturing and quality -nul im=*+1cn methods was ocupleted. It is
=ww+=d that visits to other manufacturing facilities will provide further data for evaluat EM .
y Additional information in this area, and in the waluation of maintenance practions will be obtained throups a survey of various utilities. She results of this Phase I evaluation will inchule reocenendations on potential performance parameters and functicrial C-15
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- irdicators that could be nortitored throughout the life of an MCC afd be useful in detmeting aging % ..dation. 'Ihese rectenandations will be zwviewed'and verified in detail during the Phase II work.
l C.2.5. posidual Heat Tamwal and Carmonant Cbolinqr Water Systems A fifth elenant. in the BE scope is the lhase I aging evaluation of X Wehe ruidual heat rumval (RHR) system and the +..t omling water (00t) systam. The IHR residual heat remwal system is a &aal purpose system at 54Rs which is used for: (1) remwal of resiami or decay heat from the reactor by transferring the heat to service Water in RHR Heat Exchargers, and (2) injar+im of water at low pressure and high volume into the reactor in the event of a IDCA (Iow Pressure Coolant Injecticm-IPCI mode of RHR).
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'the DR +- t cooling water system is a cooling water ' system used K at PWRs to cool a variety of or)lanth- ts sucts as: the Reactor Cnolant Ptaps, Basi &ml Heat Ramwal Heat Dcchangers, IXX:S pumps, and c e /e l K Istdcun Heat Dcchangers. It is a pure water system that is, in turn, sedai by Service Water.
. '!he iget of -vis-i failures on plant system performance is being evaluated utilizing results frm the -,cs-.i level studies, and work perforned by all pertinent NRC contractors for systems data ammanament and systems level rijsk analysis. 'Ibe study is performing in-depth systems level failure data reviews, reviews of current industry practices for system maintenance, tasting and operation, and probabilistic risk amammament (PRA) techniques to understand and to predict the 4M of aging on system availability. Barn =nandations for improving the system performance by means of %ue tion monitoring and timely preventive and m.acdve maintenance will be addressed. 'the pro $ect will intacrrate its products with other Ba,
'7 g,- f(operational safety Reliability Researctpi Performan]oe -
zndic.t.s. Y Afr:n*g u ftf-C-16 l
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i C.3 NPAR Research - DEL Activities Several studies have already been canpleted at INEL in support of the NPAR program effort. A preliminary aging assessment of batteries, cables, murwrs, terminal blocks, and per=^uations was ocmpleted. In this
===a== ment, the materials e=wtible to aging, stressors ard erwironmental envelopes, failure nedals and causes, functicnal indicators, ard current y etioes, were identified and evaluated. They also completed an early avaluation of the susceptibility of materials in pressure, ' capture, and i level sensing systems to aging degradation.
In-situ electrical maasta-era have been made en the plant safety systems at the shippirgport Atmic Power Station prior to plant
? ? s - Jamiening. Using the IXMG Idaho devel ECCAD , five cable circuit types were evaluated: pressurizer heaters, control rod position W1 r Z y dtE'# indicators, various primary system RIDS, motor operated valves, ard nuclear - '
instrumentation. The test results were h=nted in Reference 21. The remaining ===a== ment effort on these wwsis is being corw+w+M as part of the evaluation of selected systems. 1 Finally, an aging ==aa==wnt ard defect characterization was made of
; selected valves from Shippingport Atomic Power Station. This task was a joint activity between the NPAR prv: am and the Equipnent Qualification Program. In this task two valves were examined, refurbished, and operationally tasted. One was an 8-inch motor operated gate valve in service for 25 years. The other was a 2-inch motor operated globe valve in service for 5 years as a high pressure injectico punp throttle valve. The results of this effort are being LW=nted in a NURD3/CR report.
A review of the Stardard Review Plan was initiated to identify any age related technical issues. The critical reviews includes Chapters 3-10. The evaluation is wh'uated on the operating experience gained to date. 7he remainirg work on this task is included in the current effort. l l C-17
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'!he key results of the' DEL effort are beirg h==ntad in a series of NURED/m reports.- Several reports have beer, issued to data (References 10, 11, 12, 21,.22, 23). l 3
01rrantly research on three major NPAR tasks is being conducted at the ] DEL. ;'Jhe three tasks are: 1) Evaluation of Aging contribution to Plant Safety 2) In-Depth Engineering Studies of Selected Systans; and 3) Residual Life Assessments of Mechanical G.p ia. C.3.1 Evaluaticm of Acim Contribution to Plant Safety
'Ihe work in this task is being perfomed to identify were there are risk and safety ooncerns due to aging d. dation of -p.-As, systems, and structures.
'Ihis task belorgs to the special topics category and departs from the ,
usual NPAR phased approad. . 'the appread followed in this task is to first review operating experience and then develop a model Wi& irmuriotes aging effects an system and reactor risk analysis. Using these models an evaluation is made to identify in Wich systems aging has a significant
; effect on risk. Once the systems are identified, reocumendations are made-for in-depth engineering studies, i In the first part of this task, an evaluation is made to determine the extant to Wich aging has affacted 1HR safety system performance based on the operatirg experience contained in the Naclear Plant Reliability Data System (NPRDS) data base. 'the systems being evaluated are listad in Table C.2. 'Ihe evaluation efforts empleted through W-1986 are indicated in parenthesis.
C-18
l 1.
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TABII C-2. SYFTDi ANAIXSIS SDCUS 1HR Safety Systes (Fcr additional systams to be studied in FY-1987, { priority will be placed on mWk and Wilcox (B&W) plants.) J I
- 1. Reactor Protecticm Trip System (completed Westin;h:xase)
- 2. Chemical and Volume control System. focmoleted B&W). i
- 3. Engineered Safety Phatures Actuation System
- 4. Residual Heat Renoval System
- 5. Power Ocrwarsion System
- 6. Emergency Core Cooling S'/ stem
- a. HicA Pressure Injection and Facirculation focmoleted Westimhouse and B&W)
- b. Icw Pressure Injection and Recirmlation 1
- 7. Auxiliary Feedwater Systan (_ocroleta$ Westimhouse and B&W)
- 8. Pressure Control System. (e.g., Power Operated Relief Valves, etc.) -f
- 9. Safety-Belated Reactivity Control System PWR Succort safety Bvstems ,
- 1. Class lE Electrical Power Distribution System focmoleted Westimhouse and B&W)
- 2. Service Water System (ccroleted Westimhouse and B&W)
- 3. C-pr-it Cooling Water System focmoleted B&W)
IHR Safety Systems
- 1. Reactor Prote: tion System (ocmpletad)
- 2. Standby Liquid contzel System (oortpleted)
- 3. Engineered Safety Feature Actuation System
- 4. Empay Core cooling Systems a, coolant Injection-RicA and low Pressure (ocroleted RRR)
- b. Automatic Depressurization
- c. Cere Spray-HicA and Im Pressure
- 5. Reactor core Isolation cooling System )
- 6. Pressure Cbntrol System i
- 7. Safety-Related Reactivity Cbntrol System BWR Sunoort Safety Cvstems
- 1. Class 1E Power Distributico System (completed)
- 2. Service Water System (careleted)
- 3. Cup 4-4. Cooling Water System (garoletad) 1 C-19 I
For these systes, the ww=,t failure contributions to five major categories (the " broad brush" analysis) are to be determined. '!hese categories are: (a) aging and serdce wear, (b) design and installation, (c) testing and maintenance, (d) hunan related, and (e) other. 'this will identify -get/ failure category oczubinations that will be used to j establish relative impacts on system unavailability. Additionally, events in these failure categories which resultad in initiation of a transient or accident, loss of system function, loss of redundancy, or degradation of 1 system availability are to be identified. 'the informatim gathered should include, eere obtainable, the age of the plant involved, the system, wet type, and in-service age of the -gst. 'Ihis information is cataloged and filed for future evaluation. 'Ihe results of this effort are g[ h =ritad in a h Repo 'Ibe i@t includes categorized failure results by systes and -g-d.s, for 5-year irewts of service age. A second part of this task is the development of aging mials. Aging , ( mials are being developed to provide quantitative determination of the effect aging has cri plant safety. In addition to developing aging models, a data base of aging-related failuru data is being developed to provide aging l l root cause information for various systems. In FY-1986, DGt Service Water Systams and sczne PWR Class 1-E power distribution systems were analyzed. - 'Ihe work remaining includes root cause evaluation of the Westinghouse O
. auxiliary feedwater system (AINS) and hiW1 pressure injection system (HPIS). 4 With the aging models and the aging related root cause data, an evaluation will be made to assess the relative effects aging has en reactor systems. 'Ihis analysis will be made on the plant / systems miels used in the Surry PRA performed for NLRED 1150. 'Ihis evaluation will pzwide a best !
, estimate of the risk and safety implications of aging using the time l dependent models and best available failure data. Use of the Surry models l will also pzwide a direct cartparison between current PRA results and the j time dependent agirg rd*1m. :
1 C-20 l
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,o C.3.2 In-Depth Engineerin:r SMes 'of selected Systen
'Ibe in depth engineering studies of systems are being ocnducted in accordance with the NPAR phased e.pproach to resehrth dmacribed in NURIr,-1144. The systems of interest for this task are those that provide the protaction function for the major boundaries to the release of radioactivity (n:s, contalment). Cartain Class 1E power W-As also-will be studied as-subsystems to determine their iwW to safety and as part of the systans of interest. '!he reeaarch results emanating from this task will provide at the systan level:
- 1. Identifica,ticm of aging degradation effects.
- 2. Identification of trend G-p.datien.
K Identification of methods to mitigata degradatica (maintenance and 3. replaoament).
- 4. Reoamendaticos for modifications to weinte codes, standards, and . '
regulatory guides.
'Ihree systems are currently being evaluated in this task. 'Ihey are the High Pressure Eh-ty cy core cooling System (HP-ECCS), the In-Plant Class 1-E .
Power Distributica .%stam and the Reactor Protection System (RPS). e A maae I evaluation of HP-ECCS is being perforund. The evaluation planned for this system includes an evaluation of aparating experie. 7 (Which will ir.-r.,ste the results already obtained in part Task 1 a swt Phase I aging ====== ment, and' a review and rwrunaridation for ISEMM. The ) dN#y Phase I aging assessment of the HP-ECCS also includes in-plant aging studies ] 7 a Evaluations are being made of internal event and transient j g geb.d amammament a. rats, inservice inspectico (ISI) and tasting twoords, performance tasting and station modification records, rnaclear maintenance and equipnent data bases, and results of trend analysis and predictive maintenance programs. 'Ibe results of the Phase I HP-EOCS evaluation will be h==%sd in a NURES/CR i ru.t.. i C-21 l
Task II ' the Class 1-E Pmer
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g.,4 The.smoond system being evaluated Distribution Systam. A Phase I evaluation in p 4s followirg the NPAR stratagy. She work being perfrom4 includes an evaluation of operatirg performance' to' determine what aging related problems have been experienced i and a Phase I aging assessment Wiich will include a review, evaluation and : I reocanandations for ISM. Also included in the evaluation of this .
.f is an in-plant study of the Class 1-E PcWer Distribution System a CNS-3 A three part study will be performed of: (a) the direct current (de) power system, (b) the vital instrument power subsystem, and (c) the hicjh veltage power subsystem. This work will be h=aritad in a NUREG/CR awi..
A third system currently bairug assessed is the Reactor Protection System. A Phase II evaluation is being performed in this subVsk. She work being performed consists of a omgrehensive Phase II aging assessment including post servios examinations of naturally aged Wr=arst. A f significant effort is planned'here on waluation potential performance , pr J- indicators for the RPS. With the ocupletion of the mase II evaluation, application gnirWines will be prepared for surveillance and maintenance and for conditien determination (to assare operational readiness of aged RPS). Rocarnandatihas for modifications in testirg and inspection methods will be made. 7 1he ask 2 ffort includes plans for future studies of additional j j systams. II evaluations are projected for the HP-E03 arxi Class 1-E Power Distribution Systems. This work should include evaluation of do i performance indicators, service wear effects (ISW) advanced methods a:xi
. application guidelines.
Phase I studies are planned cr1 the service water, residual heat renoval I i (RHR) and wpiist coolirg systems for a PWR. Aging studies en the RPS or 1 HP-EOs will be extended to a ENR or Westirxjhouse plant. C-22 I l
4 C.3.3 Beg.idual Life Assessment of Maior cwreta An evaluation of the age degradation and resMm1 life of mjor IWR mps=&s is being performed at the INEL for the NPAR gwieu. The NPAR effort is safety oriented and is cartplementary to the crt-going irdustry sponsored pilot projects on plant life extension. She approad used in the raidual life ==aa====qt task is to first identify and prioritize the major miwera with r==r=+ to safe plant operation. '!his is folicwed by a Phase I effort to establish the life limidng prma==a= for ea& of the major wip=is. Included here is the ident*.fication of degradation sites and failure wrria= during normal operation ard accident ocnditions. The Ihase I effort also includes an n=== ment of current and potential inspection, surveillance and monitoring methods. Ibr this phase of the - l effort, the work is fe=M on integrating currently available technical informatico dich is relevant to a?ing and life extension. This is mainly information which has been generated or is now being generated by other NRC and industIy pwtam. ] The results from the initial accacemant of life limiting processes will then be ussi in the development of simple mechanistic models for determining 4 the residual life of selected major e.mpor-is. Develeping these models is of particular interest for c.uweAs Wich are not readily ama==ible for routine maintenance and inspection. l l As these models are developed, residual life evaluations are planned for the major 4eds using actual plant operating data. In this phase l of the project, key plant cperating data will be identified. This is the 1 n
)( operating data 56ich is r=,auey for a realistic (rather than a enveloping 3
design or conservative) estimate of the mechanical and thermal loading of I the s.mycimis as well as other envimanal aiisr. I C-23
I ' y With the coqpletion of the Phase g evaluation effort, the results will be reviewed and areas Were additienal research is W will be identified. Where additional work is required, r s = 4ations will be made for Ihase II racaech. As of FY-1987, the major www.ts important to plant safety have been identified vd prioritir.ed. An initial evaluation has been made of five PWR-ww ia--the contairmant, pressure vessel, primary piping, steam generator and vessel support--and of three EHR +-,'J--the pressure vessel, recirculation piping and vessel a pports. In this evaluation, the degradation sites, degradation medianisms, stressors and failuru modes have been identified. 'D11s evaluation also includes a review of the current methods used for inspection and surveillance of these wwwid. The results of this effort have been rumanted in NUREr./CR 4731 (Reference 23). The identified scope of wort tidi remains is as follows. The initial , evaluation of the rtraining ww-its will be cxmpleted and a more detailed investigation of selected major ww ds such as BRR and PWR containments will be undertake.. A detailed anamnenent of current and emerging inspection surveillance
,; and monitoring methods will be performed. This will include methods being developed for use in nuclear power plants and methods being developed in allied industries such as fossil power, chemical plants and aerospace.
Ree m -rdations will be made for additional researds or engineering qualification efforts for newly developed techniques. i C-24 U- _ - _
m C.4 FPAR Resegh - ORE Activities
'Ihe NPAR p,.iu is spcrisccirg research m four major tasks at CRE. l
'Ibe first task is an evaluation of the auxiliary feedwater systam. 'Ibe other three tasks involve aging assessments of varicus w -nts. 'the results have been r h nnanted in References 24, 25, 26 ard 27.
C.4.1' Aerirn A ---mant and Analysis of Auxi14 mrv Faa4 water System (AfWS)
'this task includes the aging emmanamant and analysis of wilhary X (Ants) . '!he main objective of this task is to apply the NPAR Fa,-u strategy and provide rw=mandaticns and guidalinesras.-
for inspecticn, surveillance, monitoring, and maintenance of g AFus. Phase I of the task on )(
.f Ac,Afus nc i uldes eva ul ation of operating experience, determination of agirg d inpact upon its operability, review of inspectig, surveillance, and M condition monitoring methods, and evaluation of 3role of maintenance practices in counteracting aging. CR E is planning to perform the Phase I study in conjunction with a cooperating utility. '!he Phase I will A X.
provide r+ - =4ations for cceprehensive 3 aging a==anamant o3AYwstobe followed in Phase II study.
; C.4.2 Asina A==a== ment and Analysis of Motor Ooerated Valves.
Check Valves, A7W B- and Solenoid coerated Valves
'Ibe secord task includes the aging mana.mamant ard ana}ysis of motor operated valves, check valves, ATW punps and solenoid operated valves used in nuclear power plants. Phase I of this task is ocarplete afd CRE is X performing a O wrerensive H)ase II aging assesfenent. In a separate task, detailed evaluations of the role of maintenance in counteracting agirg effects in these valves will be done. Here the relative benefits of various predictive, preventive, and in -tive maintenance practices will be evaluated and lugvye.r maintenance practions causing valve degradation identified. 'the evaluation of motor operated valves was octrpleted in FY-1986.
C-25
C.4.3 'nastire of Naturally _M Solanoid Valves hstirg of returally and artificially agmi solenoid valves is beirg perfomed usirg IEEE 382 ard 323 as a guide. ' 2ezmal' agirg and cyclirg will
- be used for artificially aged valves. Bene valves will be subjected to gamma radiation to sina21 ate the accident exposure, and than pat through a 30 day Loch test. Se valves will be functionally and electrically tested and physically examined at the and of ands phase of testirq to determine the degree of degradation.
C.4.4 Diarmastir= ard Manitorina of name tor Internals - struct22ral Ir*=rvrity In-core and.ex-core neutron noise nonitoring is being evaluated to detect
%ndation in PWR reactor vessel internals. S e results of this task will ~
provide iriput in revising a standard for the use of neutron noise to nonitor - core barrel vibrations and preparing a standard to monitor loose parts..
'Ihis task will attempt to predict the effects of various types of degradation' noise and vibration signatures.
X &:ha:~y CRNL has analyzed ex-core neutron detector noise' data to deterrine the
; feasibility of detecting incipient thermal shield degradation in tuo danestic FAR reactor pressure vessels. Results of the noise data, analysis indicatas that thernal shield sqpport degradation probably began ee.rly in ;
the life of both plants. Se degradation was characterized by the appearance of new resonances in the ex-core neutron detector noise. 'Ihis study shcus that neutron noise analysis program can be used to monitor degradation of reactor internals. 1 l' \
- i l C-26 o
C.5 NPAR Pa=am d - Other Activities The National Bureau of Stardards (NBS), Systers Engineering Associates (SEA) and Sardia Naticral laboratories (SNL) are conducting romaad studies and providing support for the NPAR y%-n. C.5.1 FIE Study h
'!he NBS is conductingga irdepiu-4. t review of the techniques that have y
been used for in-situ testing of electrical cables inside the containment. The techniques being evaluated are the diagnostic methods and
=am e n at approaches used for detecting incipient defects die to the aging of both electrical and nachanical ww fa in plant safety systems.
- 'Ibe safety of operating nuclear power plants depends upon the ability -
to identify and replace those ww-is Which, due to ruutine or abnormal aging, may fail to perform their intended function during nnrmal operation, during or after trigger events, under accident conditicris, or during extended life. 'Id identify the ww tm which have propensity for aging, it is rww==ary to develop the measurunent techniques to determine their
; operational readiness and Whether or not failure is likely.
'Ihe NBS ' investigation fe- on the aging of Class IE electrical cables in nuclear power plants. Previous work has identified three characteristic failure =vh Which are of concern in nuclear power plant applications. 'Ibey are: dielectric failure, localized changes in characteristic iw=hnoe, and localized increase in the resistance of the conductor.
While there may be other failure nevk for the electrical cables in question, the failure
- listed underscore the fact that the failure criteria are diverse. It is unlikely that a single simple test can be used to evaluate all relevant properties of electrical cables. It is also true C-27
i that the physics and chemistry of aging are not well enough urderstood to ! permit a unique and unambiguous identificatim of incipient failures of cables inside contairnent. '311s evaluation pr%.a at Naticnal Bureau of Standards will provide a technical focus for basic investigations into aging and failure mechanisms and to provide NRC a capability for an irdepen$ent f j evaluation of measurement methods and approaches for detecting defects in i aged cable systems inside ocntainment, during normal operating life ard during extended life. C.5.2 SEA Study
'Jhe research effort by SEA is aimad at evaluating the consequences aged w=uis have on vital IHR systems and the resulting effect on plant safety.
The Ihase I effort, " Method to Analyze and Urdundand Aging Effects" . has denc<nstrated the application of the N-square diagram nodeling of the systsia interactions to identify ww. ara ard parts within ocmponents with agirq significance. 'Ibe method involves proper characterization of the functional and spatial systems interactions ard information pertaining to: . o . 'Jhe relationship (and effect) of parts to caponents perforrrance, a o 'Ibe relationship (ard effect) of systems to plant performance.- o 'Ibe effect of :ging and service stresses at the part and ccrnponent level. o critical specification parameters that must be maintained ard are affected by age and/or services stress. l o Presentation of simultaneously occurring interactions for evaluation. C-28
/,
, . l J
'the current researth effort is directed towards applying system j interaction nodal w- inu, developed in Ihase I, to selected safety -
systems and support systems; irwestigating the systems ability to mitigate effects of aging leading to carsnan mode failures; identifying weta and- k parts whicti have propensity for aging % .dation; generating recrJanendations for maintenance of the systans to alleviate aging concerns. l
'Jhe systans interaction model im-- tmas are being applied to the i
followig 3HR and BGR fluid-mechanical and electrical anfaty systems. 1 (1) IHR Safety Systems ard SuDoort Systems (a) Hipi Pressure Emergency Core Coolig Systan (b) Class lE Electrical Power Distributico System (c) Service Water System (d) A wil4avy Feedwater System * (2)' B4R Safety Systems (a) Iow Pressure Emespcy Core cooling System ; (b) Reactor Protection System ! o C.S.3 SNL StudV The SNL is conductig an assessment of Class lE electric cables for aging and their gaalification for useful life extension. h is assessment will focus on: (1) determining how the various monitoring indicators, for cable aging, change with time; and (2) determine by 10Ch Test whether aged
. cables can be shown to be qualiiled for life extension (beyond 40 years).
Selected Class lE qualified cables, i.p.eardative of those currently used in nuclear power plants for safety-related functicns, will be used in the researth. 'the degree of cable %iodaticm will be determined, by measurement, for various usage periods (i.e., 20, 40 and 60 years) . The i l C-29
cables will be' artificially aged using: a long time period, tanperaturns based on the Arrhenius Theory a-d a reasonably.Iow dose rata (to assure that .
- - dose rata effects are properly considered). .The pt w tom tasks include:
(1) A review of lists of Class lE electric cables (EPRs, polyelefines, etc.) in general use for safety-related functions and reocenand, to i NRC, a representative selection of cables for testing. j (2) Devaleping an aging test plan. This plan will be based on using. standard test spools, around idhich selected cables can be EM and ' aged,:to perform a IDCA Qualification Test. The IDCA test would be run for cables which have been aged to the equivalent of 20, 40 and 60 years.- The test results are to be used to evaluate the cables' survival and qualified life as a function of age. Also, the test results will be studied for eviderce of potential problems that might ecist if the cables use were extended beyond the normal 40 years . (qualified) life. (3) Developing a method of, and performing an acoalerated aging test of the
' cables en the test spool frames. The aging period should be about six acnths and use the tanparature equivalent, ard radiation dose- - equivalent of power plant operation of 20, 40 and 60 years.-
a (4) Periodically tasting the ptWies of short lengths of the same aged cable materials that are on the spool frames. She cable sanples are to be removed from the aging program at selected intervals to maa=ne cable degradation ever the equivalent 60-year aging period. Mechanical l rwties are to be measured, including tensile si tv 3th, alcmgaticr2, hardness and density. Electrical r%ty measurements are to be made ] for capacitance, resistance and voltage discharge. The sanplirg periods will be at approximately 10-year ir A of age. The degradation maaeniments will be plotted and correlated. nacai on the analyses of the data, reconnendations will be made for actual in-plant 3 cable monitoring measurements to amma== cable aging in nuclear power plants. C-30 . l
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- - - - - --------- ___ _ -______ ___J
y- . APPENDIX D CNGODG PROGRAMS REIATED TO NPAR ., 1ABIE OF CENIENIS D.1. Office for Analysis ard Evaluating Operational D-2 Data (AEOD) ...................................................... D.2 office of Inspection ard Enforcement (II) . . . . . . . . . . . . . . . . . . . . . . . . D-3 D.3 office of Nuclear Ranctor Regulation (NRR) . . . . . . . . . . . . . . . . . . . . . . . D-5 D-7 D. 4 - Office of Rasaardi (RES) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D.5 office of Nuclear Material Safety and Saf eguards (NMSS) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D-7 D.6 7W:hnical Integration Review Group for Aging ard Life " D-8 Extension (1RIGALEX) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . D.7 Ongoing Aging and Life-Extension Fi%tous at EPRI . . . . . . . . . . . . . . . . D-9 D.8 Ongoing Aging and Life Extension Fau:g-a in Irdustry . . . . . . . . . . . . D-10 D.9 Ongoing Aging ard Life-Extensien Faw:r-a at DOE . . . . . . . . . . . . . . . . . D-12 D.10 Ongoing Life Extension Activities in Ocdes/Stardards . . . . . . . . . . . . . D-13 D.11 Ongoing Aging and Life-Extension Figt-a in Foreign Countries ........................................................ D-14 O d l J I i l l l l [ l D-i l.
APPDOIX D J ONGODG HOGRAMS REIATED TO NPAR Aging of nuclear power plant wp Tus have significant inpact cri reactor safety and economy of the plant. Aging my reduce the safety margins for critical +-74, structures, and safety systems,and thus emprmise the defense-in-depth uawi.. ' Therefore, the USNRC and the regulatory agencies in the foreign countries are sponsoring reseerdi and development programs to evalusta the impact of aging on the safe operation of nuclear power plants. Nuclear industry incitding EPRI, NSSS venders, utilities and architect engineers are' int =M in aging because of the significant economic advantage that can be derived frm extending the operating licenso of the aged power plant. Uherefore the naclear industry ard esprcially IFRI/ DOE, is sponsoring pilot studies to evaluate the potentiati of plant life extension (PIEX), i.e. operating license extension, for a typical EHR ard IMR plari. - There are several NRC and industry sponsored aging related pi@uus that are currently being carried out. Many of these programs are ocmplementary to endi other and therefore it is essential that they are coordinated. Coordination of these r%i-s will eliminate duplicaticri cf
- efforts, and provide a more ocupleta set of aging related results in a timly and cost efficient manner. 1his coordination and integration effort has a special significance for NRC because it is expected that the criteria related to license extension will be required by early 1990s.
The Dcecutive Director for Operaticris (EDO) has established a 'hchnical Integration Review Group for Aging and Life Dctansion (TIRGAIEX) to ensure effective utilization of HRC resources. TIRGA1EX has 1iwf aized the importance of the coordination effort and ocnducted a preliminary review of relevant swi s and activities already underway. 4 D-1 k-----__.---_ __m -______._
i \ interfaces between NPAR PtWimu and other argoing NRC pWims have
)
l been established and will be maintained. similarly, external swims involving botn dcnestic and foreign organization have been contacted. This sectico cutlines the aging related piwims FNM by NRC. It also i cutlines the similar g%ims sponsored by irxtustry in the United States, i l Japan ard Europe. De outline of these rws== is structured into follcuing six 51Mians. I i
- 1. NRC
- 2. EPRI A
- 3. Industry
- 4. DOE l
- 3. Cb6ns and Standards Otanittees
- 6. Foruign countries.
1 Se current NRC kWime identified for coordination and technical integration with the NPAR piWema include, (i) Unresolved arri Generic Safety
. Issues (NRR), (ii) Maintenance and Surveillance Ptwsma (NRR), (iii) Plant I
Performance Indicator h @tma (IE), (iv) Safety System Functional Inspection Prwsmu (IE), (v) Development of Regulatory Policy, Guidelines, and Review Procedures for Life Extensian/ License Renewal (NRR in conjunction with RES),
; and (vi) Reliability Assurance Pt%Ama (RES).
I D.1 Office for Analysis and Evaluation of Operational Data (AIDD) l AIDD has developed, through the Oak Ridge Naticnal IaboratoIY (ORNL), a cuprlbersive ocmputerized dataham to aid in collecting and evaluating licensee event reports (IIRs). This dataham, the Sequence 0:xiing and ) Search System (SCSS), contains established prrr*bres and
- for collecting operational data. ItalsohayJOresultad A
in codes specifically X identified for aging degradation. ADOD ghas an ongoing project to analyze the Nuclear Plant Reliability Data System (NPRDS) dataham. In s ce.It with the Institute of Nuclear Power Operatyms (INIO), which maintains the j database, AEOD has developed a list of critical -ements en which to focus K attention. The trends and patterns analysis of NPRDS data focuses on key D-2 l I 4
-n ..
. m p -nts. AEDD's analysis of NPRDS data results in statistical and ergineerig evaluations of -p- A failure modes, times to failure, operatig conditions that affect failure,. and chemical ard physical
,, conditions affecting +-a wearcut ratas.
p ! A major activity of the NPAR program, in Phase I studies for -Wides l and systens, involves' reviews of operating experience and analysis of the data base (IERs and NPRDS). 'Ihese reviews of the data base will continue. Also, the data that any amenate from the ABOD's trends and pattern analysis and operaticral event and root cause programs will be factored into the aging data bank. vo D D.2 office of Insoection and Enf m A (IE) Several IE programs guide crgoig regional activities relevant to c([ L < , aging, aging detection, and mitigation of agig oansequences. 'Ihase . programs include the Safety System Functicnal Inspection Fr,-i, the Safety System Outage Modificatisms Inspection FAws n, and the Generic Oceanunication F.,- .
'!he Safety Sys+=a Punctional Inanne+% Fhmu, in general, ammaan.c
- whether plant modifications of selected safety systems have C .LT.d the design margin to the point where the system's ability to mitigste design basis events is impaired. 'Ihis program consists of an in-depth review of a small number of safety syste s and is usually conducted at older plants.
'Ihe major objectives of the psy m are to assure that safety systems are:
o capable of performig the safety functions required by their design bases ! o Testing is adequate to demonstrate that the systems would perform all of the reglized safety functions o System maintenance (with acphasis on punps and valves) is adequate to assure system operability under postulated accident oorditions ' D-3
.o- operator and maintenance technician training is adequate to ensure sw operaticms and maintenance of the system o manan factors oorsiderations relating to systems and supportirg r< ewes are adegate to ensure proper system operatico under normal and accident ocoditions.
De cbjectives of the safety sys+= outacie WWications Wat Pm;tries are to verify, through sampling i,Fdans, that o
~
Linansees havat effective controls for conducting modification ard repair activities durirg cutages l o Activities are accomplished in accordance with established r J ares and oc m itments o completed repairs and modifications inve been properly designed, j installed, inspected, ard tasted j o Affected systems are ready for safe startup and operation of the , I plant follcwing an outage. I De objectives of the Generic mwiication Fimme are to: o Inform licensees of problems, including thcee due to eging and wear, that have developed in individual plants ) { i' o Require action dan these problems are shown to be significant and generic. 1 IE also guides the activities of the regions by issuing the Inspection i and Enforcement Manual. Portions of this manual establish inspection j rMares that are relevant to aging and life extension. 7br exanple, some inspection procedures establish guidance for ascertaining that inservice 4 F4 p' [
c I iedon and testing activities are rug-mmad, planned, conducted, re,ivided, and reported in accordance with Section XI of the ASME Boiler and Pressure Vessel Code. Where applicable, these s.- awas prescribe inspection of the licensee's recordkaapig for modification, maintenance, and repair activities. An ongoig reutine inspection effort is conducted by the regional offices in a Csi. with the E inspection program. 'Dae objective is to assure that systans and +. iJ have not been measurably J g d as a result of any cause includig agig. As a pstt of its Plant Performance Indicator F&ugmu, IE has identified a set of performance irdicators. '!he results of the 2 activity related to 1 direct indicators of current plant performance such as safety system failures will be factored into the overall NPAR program. At the same time, j it is vi==14*M that the results amanating' frtan the hardware oriented NPAR - l piup-u, for specific -pr cits ard systems, oculd be WDi'M in evaluating the perfemance of older plants as they reach maturity and continue into extended life. f D.3 office of Nuclear Reactor Raoulation (NPR) l A substantial manber of aging-related programs are in r ugoss in NRR.
'Ibey are found'in neveral of the broad supou categories that NRR tracks:
Casework, Operating Beacta.1s, and Safety Technology. '!he major agirg-related programs are di-e=M below. Casewrk. NRR is respcruible for licensing actions and safety ! e===nants of both currently operating reactors and new reactors ocaning on line. While aging ccricarns are built into the licensirg rvc-ss (e.g., through envitu.wAm1 qualification) to sczne extant, the licensirg ru =ss is not geared toward characterizing the aging prer=== as it occurs or as it nicAt exist at the time of a license renewal request. 1 D-5 , I I l
ooeratirn Reactors. 'Ihis is one of the larger prcgram categories in NRR ard represents projects that are conducted to support individual licensirg actions. 'Ihese projects may be initiated by licensee requests for license 8Hr_rdirud.s or by events that emw at the plants that may require a regulatory response frm the NRC. Some of these events are age-related and contribute to enhancing the data hamac. Safety 'IWincleav. NRR divides this s wimu category into five subgroups; unresolved ard generic safety issues; risk a==a==mant; regulatory requirements; oods analysis and maintenance; and human factors program ice == . Scne of these swtan have aging-related aspects. For exa::ple, unresolved and generic safety issues are scnetimes related to aging issues, e.g., pressurized thermal shock, in situ-testing of valves, and diesel reliability. Current projects in risk == 1. are not specifically related to aging. It is recognized that the ability of PRAs to model the cffacts of aging is limited at present. . Within the human factors category, the crucial role of maintenance in predictire aM correcting aging degradation has been reflected in the Maintenance and Surveillance Ft%imu. Recently, Ihase I of the maintenance and surveillance p@uam (NURD3-1212) was ocmpleted and was reviewed by the - Etc. Ihase II is currently underway. Ihase I was designed to survey 9 current maintenance practices in the U.S. Nuclear Industry ard to evaluate their effectiveness. Phase II is working tward resolutions of the issues identified ard a-W in Phase I. NRR/DSRO/EIB is planning an activity, called License Renewal Policy Develosnent, whose objective is to fmmlata and preFre methods of resolution for PIEX (Plant Life Detansion) issues. 'Ibe NPAR program in conjunction with TIRGAIIX has identified broad technical safety issues which will require timely resolution in consideration for plant license extension. An implementation plan to resolve the technical safety issues has been developed for the EDos apptwal. D-6 I
e NPAR program coordination ard technical integration with NRR en aging related technical safety issues is in place and involves, (i) the resolution of specific generic issues, (ii) maintenance and surveillance ytwswu, and
-(iii) plant licanoe extension.
D.4 Office of Daman d (RES) RES sponsors a o mber of large aging-related research programs. 'Ihe general objectives of the programs are:- identifying aging mechanians, evaluating their safety and regulatory inpacts, assessing detection methods ' for particular aging degradation mechanisms, and developing mitigative ard a corrective actions. } l Major aging related RES sponsored programs include Mechanical and Electrical Eqs4.-rit Qualification; Primary System Integrity, which includes the Heavy Sectim Staal Schnology (HSST) program, Degraded Piping and Steam - Generator Integrity projects; W4==1t Operation and Integrity, which includes the Nuclear Plant Aging Researth (NPAR) program; and Non-Destructive Examination. A number of researdi programs sponsored by the Divisicm of Reactor System Safety interface with the NPAR yiw&-.. 'Ihese programs are (i) Operational Safety Reliability Research (OSRR) PWi-u and (ii) FRANTIC III Ocuputar Code for Time Dependent Belinhility and Risk Analysis. Additicmal interfaces are planned involving Risk Analysis for NUREG-1150, Mlant Risk Status Information Management (PRISM) System, and g Probabilistic Evaluation of Technical Specifications (PETS) Fawamu. D.5 Office of Nuclear Material Safety and Safecnm@ (NMSS) From the safeguards standpoint, the instrumentation and wi,-nt maarv iated with physical security systems for reactors are subjected to routine tests, surveillance and maintenance resulting in repair or replacement to assure performance such that aging is not an important issue. 'Ihe w-nts and cystems amarriated with fuel fabrication facilities are far less emplex, and they are rcutinely maintained, refurbished, and replaced. Also, frequent opportunities exist for review of I D-7 l l
)
the status of the facilities by the five-year license term. NMSS will j ocotinue its participation in TIRCAIZX to report on any further NMSS ytwtans that may relate to reactor aging issues and to maintain cognizance
/
of goig agig ard life extension work that may have benefits to }tES ,- licensing activities. y l (1 D.6 Tedinical Intamation Review Gm2D f r Aaing ,, #, { ard Life Extensive (TIR@TrX) l[j'f
'1he Executive Director for Operations (EDO) has established a 'Aschnical Integration Review Group for Aging and Life Extension'(TIBGAIZX) to 3 facilitate the plannig and integration of HRC activities related to plant <l agig and life extension. 'Ibe objectives of TIRGAIEX are essentially to:
- 1. Clearly define the technical sahecy and resplatory policy issue associated with plant agig ard life extension, and ,
- 2. Develop an implementation plan for resolving the issues in a tirely, well integrated, and efficient manner.
A draft implementatiht kwtam plan has been prepared ard it is under t \ review. 'Ibe document presents the managemnt plan developed by TIEM.EX to i
; acconplish the aforementioned objectives. In developig the plan, TTAIAIIX has fecai its attention on three principal areas:
1
- 1. Definig the major safety and regulatory issues associated with ,
plant aging and life extension.
, 2. Identifying agoing gwtairs ard activities that address the j i
issues. 'Ihis inclu$es work performed by NRC, irdadtry, the e I l Department of Energy (DOE), the Electric lbwer Scsearch Institute (EPRI), codes ard stardards cxxamittees, ard foreign countries. s a i D-8 j l i
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- 3. Ri=.w. walinj further work, where ews wilate. This includes r+: -. =4atJ.crs on how the work abould be integrated with .
effective utilization of NRC resources. D.7 Oncroirn Aairn and Life-Extension Fiwi-s at EPRI 7he Electric Power paamatth Institute (EPRI) has a number of yawtaas Which are considered related to aging. Some of them directly address aging
- ard life-ectansion:
k e o Natural Versus Artificial Aging of Materials in Nuclear Plant
;e M34= ant
{ .V.. V o Nuclear Plant Life Dctension , 'fp e Cbrrosion control 9' ,
. f.,. o' Q-kv/ tier:t Reliability
, ?
,.The projs::t on Natural vs. Artificial Aging of Materials addra==*=
the aging of polymeric materials in' ruclear reactor containments. M
; ( Specifically, it is designed to look for differences between the 7 aging r..m in the contairnent envim _,t and those occurring durirg acceltnted testing as used for cpalification of electrical
'v ,
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s 1 , j 1he NoelearPlant Life yxtension Program approaches agirg ard life j extersion from the technological pu.gdve of tAmb.umding the aging
, prrr=== for systems and $stprac,ts. . She Cbrrosion control Fawtain acktresses
- envim. a. ally caused cdackirs and pitting. It specifically suphasizes boiling water reactor (BRR) water chemistry and ts.h.Umding and mitigatirg o.
pip cracking due to corrosion. The Cu+cie,t Reliability Program is related vto, but more general than, the currosion G hul Pawtam. y'f Structural' reliability aid safety ard safety inprcuement are the main V.
.i " p9 h
U i 2L,
' 1 t'
thrusts of the Goyei=,d. Reliability Fr%swa. Specific exanples of activities include material c:haracterir.ation, flaw detmetion and n==aca"ent, inspection hardware development, and reliability methodology. 1elile the central pwten focus is materials, the ptwteu also a&iresses piping, l reactor pressure vessel, and steam generator tubes. l
'No other ytwt- , the Plant Availability Pi@t-u and the Safety control 'hsting FiWi-u, irdirectly address aging concerns. Both address or L
l are related to online monitoring, plant availability, and plant performance. Diagnostic systems, human engineering and the man eactrine l interface form the rucleus of these programs. 1 EPRI also ccesyci ors a joint program with irdustry and IDE to address life e lan issues. 'Ihis piwt u is M==M in more detail in section 3.0. . An NPAR-EPRI interface has been established through its Eqaipment l , Qualification Advisory Group (EQM). A utility review of selected NPAR l Ihase I reports is in ptwt.ss ard it is expected to continue in the future. Close NPAR gws-a coordination with DEI-IQM is planned. D.8 Onacina Acirn ard Life-Betension F1hs in Irdustry A joint irdustry, EIRI, and DOE ptwinu was initiated in 1984 to identify issues associated with IHR life extension. In 1984, DT ard EPRI agreed to co-fund studies. 'Ihey developed a joint R&D plan in 1985. Two pilot studies involving Surry 1 (IHR) ard Monticello (ERR) were initiated. h An AIF/ National Envitu -ital Studies Project regulatory study gas also K
- started. Results to date, of the studies were yt-b=iund at a seminar Airyast 25-27, 1986, in Alexardria, Virginia. 'Ibe presentations delineated the roles of program participants, as follows:
D-10 j
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[ o 351,J4fe Extesion Utidly_Stearirn C3,g.ry. 'Ihis group's prhnary s A "d(getive is to preserve and enhams tM ,optico for extendid).the
'( ' ljfe of its aanbers' rO:lasr js:ser plants beyond the initial liderxM 11&rtir.e et 40 years. 'Ib actdene this objective, the
[',, .1 c.anittu ranst: (1) help uses the avellability of technology ) f5' 5 #' and axhods to enablsa Imdivirlaal.ownr.rs to make informed decisions
, M about life' extension and (2) assist in est%ishing a proven
- regulatory process for. seoaring renswai licnes frun the NRC..
f6 a , ' > i
, o glg$ric A3.g;h% Institute lEPRil. EPRI's primary function is to trasfer technology aid lensars learned to its ers. It stppora ongoinr/ life extavsion wcck by.co-sponsoring tM pilot 5 plant stalles'and providirq the remilta of other EPRIPM projects for use in the eqcirrj Pm nseanfl. Several years ago, EPRI sparta.ed' initial stalles of tM feasibility of extending the
^ ~
cperating life of ruclear power plants. - t 3 ELiDepartment of Enerov i2013 DOE is concerned with o raintainixxJ cptions for meeting future energy ocunands ard views nuclear power plant life extenian an a viable ohtion. DOE is , cxMr.pansoring the pilot plant stuiics and is performirxJ several asearch Frejects that will enharm the life extension option.
"c
'Jhe other projects inclub rick ayanents, cable agirg evalbatirms, and accraaic assesammth. DOC has satered a
[i a:x:r.erad.wn wn wnment with EPKI to r,upport R&D that wiir enhance the feasibility of extnnded life.
, 3. ,
. o grthern ctrtes_B;ddg_E9"d. NSP is conductirg the BWR pilot
, plart: study 6t ito Rnticello Nuclear Generatire Station. 'Ihis
- star'fis co-funhf by DEI ard Drx, n
w yj;d@_B;!Ug. ' hiuJ n.ta i Rear is condoctirrj the IMR pilot plant
, , sudy at its surry pWt i Plant. "Diic study is also ors-funded by EPRI ard DOE.
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o . Atomic Irdastrial Forum (ATF) . 'Ihe AIF has two life-extension s wtons in s wi=ss. 'Jhe first program, sponsored by the AIF Naticnal Envimur.ut. Studies Project (AIF/NESP), is a study to 4 1 determine the regulatory issues that will affect life extension. j I
'Ibe second program, the AIF Life Extension 9_hittee, was created in 1986 to review licensirg issues. ;
NPAR swtaru cocedination exists with the aforementioned ongoing agirg and life extensicn programs implemented by the industry. D.9 Onacirn Acirn and Life-Extension F1-ans at IDE Plant life extersion is being pursued by DOE through its nuclear energy organizational element. One of three major efforts within the light-water reactor (INR) safety area is the cq=rative EPRI/ industry / DOE swimu to extend the productivity of existirg and future IHR's. DOE s wiaus are , intended to support relicensing by influencing the federal regulatory process and by evaluating the safety and ecx:namic impacts of improved plant performance. A joint DOE-NRC Workirg Gray) on Plant Aging / Life Extension is being considered to facilitate information excharge and ccq=tative efforts.
. As a part of the joint IDE-NRC coordination and information exchange, members of the RES and NRR staff participated in a 1 day workshop on plant aging ard life extensicrVlicense renewal. 'Ibe intent of the workshop, considera$ as a kick off meeting, was for an informal information exchange with the hope that same consensus can be reached to improve encpirg DOE and NRC research programs related to aging / life exte.nsion, order priorities, ard facilitate ocrplementary activities between agencies and contractors.
'Ihe DOE swimu managers and contractors associated with the N-reactor, EBR II, A'IR, HFIR, HFER, and FFIT provided an overview of their experiences, crgoing activities, ard future swtous related to agirg and life extension. 'Ihe NRC staff reviewed the hardware oriented engineering researth swtous related to, (i) primary system boundary material aging, (ii) safety systems and w#nts ard (iii) agirg of civil structures. 'Ibe D-12
___ _ _______ _ __ __________________________l_______
'l dialogue and sharing of data between DOE and NRC is expected to omtinue if-apprtwed b; the respective management, perhaps in a more formal manner with special subgroups directed at specific ta&nical areas / tasks. An innadiata actice item that was identified at the conclusicm of the W hcp was to assigrVnminata three or four manhers fzten eat agency to form a working group. So working group's first task will then be to prepare a marter, define the goals and objectives and foam 21 ate scune early . 1 tasks. D.10 Ornoim Life-Extension Activities b */Starwh'ds me following is a uumary of status with regard to code activities related to life auctansion. Institute of Electrical and Electmnics Encirmats (MP) , IEEEVNPEC Working Group 3.4, " Nuclear Plant Life Detansion," has held several meetings. Its purpose is to investigate the codes and stan$ards aspects of plant life extension as they pertain to electrical M d r= ant. An action plan has been developed. A primary report is to be published in 1988. , e J=vican Society of Mechanical Enaineers (ASME) Under the auspices of the ASME Section XI =hrv=nittee, the Special Working Group on Plant Life Detansion has met seven times. B ey have active i participation frtun EIRI, NRC, DOE, utilities and NSSS suppliers. 'Ib date, the special working group has reviewed a broad spectrum of aging and life-extension programs. De latest activity is a request to all ASME Section XI Subgroup to begin to develop &anges to the code factoring in life )( j extansion. 21s probably wi13 require at least cne year to produce r%- a m38tions for sCune meaningful danges. D-13 j f
D.11' Ongoim Mim and Life-Extansion Fiwi- in Forelan Cbuntries Extensive programs in materials degradation, NEE, fracture andhanics, l structures, and other aging-related areas exist worldwide. ] Details cm fertign activities related to agiruJ were provided at an IAEA tadmical cranittee meeting held in Vienna, Austria, ta " Safety Aspects of Nuclear PoWar Plant Aging". 'Nelve countries attended the cxamittee i meeting: Canada, czechoslovakia, Finland, France, West Germany, Italy, j Pakistan, Sweden, switzerland, the United Kingdczn, the Uhited States of America, and Y W avia. ] l of the twelve participants, West Germany, Canada, France, Italy, and j the U.S. have aging programs although the scope of the programs varies l widely. Japan has a program plan for nuclear plant life extensicn MD, but , it was not M=* at the IAFA Meeting. A sumary of the aging-related p@t-6 of West Germany, Onnada, France and Italy as well as Japan's MD
. rws-u for plant life extension has been included in the text that follows.
o West GermanV been engaged, on behalf of the Federal Ministar of the 7 "uv [g,j A Enviu t, Nature conservation and Regulatory Safety, in a data collection rwi-u at power plants for 15 years (since initial operations) and has domenstrated how 1cng-term data collection and trending of +-i,t performances parameters and functional indicators (a strategy /approm:h similar to that recommendation in the NRC's program plan for Nuclear Plant Aging Research - 2 re
- NUREG-1144) provide protecticn against failures resulting frt:n increased aging.
Aging degradation has been observed in diesel, pruer p ,,,,,, , - measuring and s.Lul cables, punp motors, vel , a variety of electric and electronic aqd= ant. The safety and availability D-14 1 l
L .. prelams Wii& say result from aging are. counteracted by a systen of. inspection'and planned p. M.ve maintenance measures. Gts has implemented a long-term systematic collection of data to make it possible to r w .ize the - - 1ation of defects on certain w is. j M A #h;,.d '1he Canadian awsMa to managing aging degradation o NPP
.i-is and maintaining adequate plant' safety is part of the cmarall canadian apprea d to reactor safety. Major features of this appread include:
g o *Ihe design o CANDLT NPPs ives to arvamiate various aging effects by using awaysiate design features su& as diversity, physical separation, and testability. . o operating policies e.nd rh prescribe practices Wii& are designed to minimize aging effects.. G.g-ia, particularly preseraiaining mg-As, important to safety are subject to in-service inspection.
. o Deficiencies are systematically detected and receeded; causes (including aging degradation) are determined, and appropriate
- w. stive actions are taken.
.o Licensees are required to senitor and periodically evaluate equipnent and systan performance against the reliability targets set by ABCB, lowing dewim of aging trends.
f o Researd and devalepnent, by both the industry and K provides the basis for predicting the behavior of critical plant
-r -,tss during plant operaticm, develop better w As, and develop armitoring methods capable of detecting -r s A.
degradation before loss of safety function occurs. 1 i D-15 l
- o. Aggressive monitorirsy is utilized, employing monitoring instruments, periodic testing, inspection, maintenance, and field s patrols.. It is recognized that aging 4 :r dation can be detA%
anly if appropriate time-variant paranstars are acnitored. o Significant Event **ts (SERs) are popu.d for events that have a significant negative effect en reactor safety, worker or public
. safety,'and cost. 'Diis system has the folloring important features:
- equipment and operation deficiencies are recorded in a specified, systematic manner to allow event review and analysis,
- nulti-level diverse screening process,
- trend lines of equipref; and system deficiencies, and Y'd n
- Innsons learned are ocumunicated to CANDU tions, owners, designers, and aq'4M annufacturers.
o Anactor Safety Reliability amamannant is an part of the y annual wessive and systematic review o NPP tion and maintenance by both the lionnees and AECB. It gives the X actual past-year performance and the eted future performance in terms of system unavailabilities and sericus process failure occurrences, dtich can be compared to the AE2 reliability targets mentioned earlier. < France Frtan the cutset of the ruclear program, the French Safety Authorities K and Licensee) took into consideration the effects of aging on the installed =?ti.= ant. D-16 l l
In Franch designs, gaalification is one of the means used to check equipner.t design. In most r==*= it includes testing that is designed to evaluate the behavior of the equipment with time. j 4 Dcarples of aging experiences in Frent reactors include- ) l o for mechanical Wmant, diesel generators are exposed to fast start omditions and steam generators are subject to unforeseen corrosico or erosion due to foreign matter o for electrical equipment, isolating switches are operated under i unscheduled loading conditions ard battcries whose autcrmy sometimes changes unexpectMly. In addition, materials such as coatirys (paints), lubricants (oils, greases, etc.) have a great effect en the behavior of the equipment with , 1 which they are associated. In conjunction with the safety authorities, Electricity de France has initiated a sugain of investigations to
. o develop means of mair.nt, in theory ard practice, of the aging of the equipment installed o determine the influence of NPP cperating rvoi:dares on aging.
In addition, a file of events is kept up to date for each NPP and analyzed to evaluate the behavior of the installation. Since the beginning of the French nuclear r ug am, means have been set j up which, among other things, ensure that the periodic monitoring of operation of the intrinsic properties of the equipnent or systans. Scre typical exanples are: D-17 i .. . .
o periodic testig ; o in-sezvice inspection o preventive maintenance Finally, ylh have been set up to identify primary circuit it of primary circuits pressure ard ] situations for each plant (m t-wature values above a certain threshold) and allow ocuparison to design features. I Italy !
'Iwo Italian ylws== have been identified to understand ard manage plant agirg. 'Ibey are (1) Italian Agig Research Fi@tmu and Electrical and Instrumentation Equipoent and (2) Cycle of Preventative Maintenance. .
'Ibe Agirg Research Fr%&ma en hectrical ard Instrumentation Equipment,
/ a limited piwiws intended to resolve issues related to the agig ard !
service wear of equipnent at reactor facilities ard their possible inpact on plant safety, has been started in Italy. 'Ibe main goal of the program is to
- provide a basis for amaaa=irg the adequacy of irdustry methods for
)( pt-.litienirg/ prior to qualification testing.
'Ibe ===W will include the examination ard testirg of equipnent removed frtan the Garigliano reactor, tru awaitig 4 -- Acionirg. 'Ihe candidate Garicliano u w =nts that have been selected through the site
)( visits DISP ard DEI, power and signal cables and theiwmyles.
[h:d 'Ibe cable activities will be hansa en the implementation of qualification test plans on nat:urally aged cables and en cable availability at the plant warehouse. 'Ihe qualification test plan will include definition of electrical characteristics, aging under u.ui.wlled enviium=nt conditicos, humidity absorption after agirg, experiW determination of D-18
activation energies,. fire propagation tests, IDCA tests, and radiation v m damage tests.
. the activities relevant to tham;1es will be based on detmetion of the systematic error affectdig the naasurements en all the ti-- =--:9glas presently imersed in shielding water; measurment of time constants on most of the tr- =-'-19ples, dare feasible; neta11%s4dc study of the hot junction; qualification of the cold junction in ocupliance with the gMares presently in use; and dynamic brittleness tests. l
\
'Ibe ytws.o will be ocmducted in cooperation with DGA and DEL. A dialogue for wative raamad with the ongoirg RES/NPAR program has been initiated.
J.asaD -
'hx:hnology development for rmaclear power plant life extension is a priority effort in Japan, with a seven-year technology development plan implemented in Fiscal Year 1985.
'Ibe major tasks in plant life extansion technology development are related to the diagnosis of nuclear power plant aging deterioration, prediction of remaining plart life, and replacanent and ivv i _;it cf plant equipment. It is expected that the following tasks will be zw- =rded for ocmprehensive evaluaticrs:
o develepnent of life diagnosis and prediction methodology (develepnent of life evaluation methods, creaticn of a datahama on aging and degradation phencunena, devel,-.t of monitoring techniques, etc.) o develepnent of technology for replacing and improving large I equipnent (development of methodology, verificatim, etc.). D-19
1 Dialogues have been initiated for inforation exchange and for bilateral cooperation for aging researtin with Japan, Taiwan, France ard Italy. opportunities for cucpatative research with West Gereny will be explored in the near future (1987-88). International vvvy raticri, through the International Atanic Energy Agency, (IAEA) is being sucpgested and my evolve into a allti-national vucpe.r-a tive program. In this regard, developing consensus among the participants is a key to the bdWng of an effective international pr@ tam. Participants mast be oarwinced that the and pr: duct emanating frtan a wative r wtom will be useful in managing aging of their nuclear power plants, and that the safety of their operating plants will be enhanced, and reliability and availability of the plant systems will be imprened. cooniination and technical integration with NRC's internal pr%saus and extermi organizations and institutions are important parts of the overall , NPAR prwtam. A mjor a W is in the NPAR prwimu is that proper coordination and technical integration of plant agirg research activities
- shall be attained at various levels to achieve overall ptwtam goals and objectives and to assure the efficient use of available resources.
i I l D-20 ) 1
. q
_ _ _ _ _ _ _ - _ _ _ _ . l
APPDOIX E SOEWIIS AND MILESTONES
'Ibe schedules for the effort on major elements of the NPAR are given in Figures E.1-E.7. As an exanple of a detailed s&edule with milestones, the Insearch plan for the assessment of the residual life of anjar reactor w w is is shown in Figure E.7. 'Ihe s&eduling of the Phase I and i
Phase II assesments for specific W-,^J and systans are shown for each j of the major activities. Also shewn is the s&edule for utilization of the j researth effort. An additional activity, indicated by the dettad lines, is l to tentatively plan for tasks whis may be r==4ad for the resolution of issues that any be raised during the results utilizatim efforts. '!he vertical dettad line in eat figure on the beginning of the 4th quarter of FY-87, indicates the expected issue date for this rwined plan. '!he schedules and major alaments are based on the current NPAR researd I priorities. It must be %dzad that the activities and s&edules can .
)
change as information is devaleped in the new p.w. and as additional irputs are prwidad and program needs are identified. t { l It should be emphasized that the number'of systes and +-is and the degree and depth of maammm=nts ard analyses Whis can be carried cut
, effectively will depend ypen the availability of funds and the period of time over Whis the results are required. 'the timely availability of naturally aged W= ant fztzn operating ard darwenimaioned facilities and the opportunity for in situ assessments will determine, in a significant manner, the resource requirunents ard the ocupletion schedule for the ,
various activities. l E-1 i 1
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REFERENCES
- 1. NUREG-1144, Nuclear Plant Agirg Reseant (NPAR) Fi,-u Plan, July 1985, B. M. Morris and J. P. Vora (USNRC) .
- 2. Understanding Aging - A Fay to Ensuring Safety, Guy A. Arlotto, International Conference en Nuclear Plant Aging, Availmhility Factor arxi Reliability Analysis, July 8-11, 1985, San Diego, QLlifornia,
- 3. NUREG/CR-3543, Survey of Operating Experiences frtzn IERs to Identify Aging Trends, January 1984, G. A. Murphy, R. B. Calla @er, M. L. Namaa arxi H. C. Hoy (CENL) .
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- 5. NUREG/CR-0036, PImmad4_ngs of the Wof,widsp en Nuclear Plant Aging, November 1982, Otmpiled by B. E. Bader and L. A. Handley (SNL) .
- 6. FUREG/CR-3818, Report of Results of Nuclear Power Plant Aging E.,r,wrJ+, May 1984, N. H. Clark and D. L. Berry (SNL)
- 7. Premad_ings, " Seminar on Nuclear Power Plant Life Ertension," -
Alexandria Va, August 1986, Cosponsorr,: EPRI, Northern States Power, l USDOE, Virginia Power.
- 8. NUREG/CR-3385,. Measures of Risk Importance and their Applications, July 1983, W. E. Vesely, T. C. Davis, R. S. Dennig, N. Saltos (BCL) .
- 9. NURB3/CR-4144, Inportance Ranking naamd on Aging Consideration of G=p-rd.s Ireluded in Probabilistic Risk humanamants, Jpril 1985,
. T. Davis, A. Shafagi, R. Kurth armi F. Inverenz (PNL).
- 10. NUREG/CR-3819, Survey of Aged Power Plant Facilities. July 1985, J. A. Rose, R. Steel, K. G. DeHall, B. C. Cornwell (INEL) .
- 11. NURB3/CR-4747, An Aging Failure Survey of Licjht Water Reactor Safety Systems and G.W-As, December 1986, B. M. Maale and i D. G. SattarWhite. (IRAFT) . l
- 12. NURDG/CR-4769, Risk Evaluations of Aging Phenanana: 'Ihe Linear Aging -)
Reliability Model and Its Extensions, October 1986, W. E. Vasely. (IRAFT) .
] ,
i
- 13. NURErr-1155, Vol. 2, Paaanttil Fay uu Plan-Staatn Generators, June 1985, ,
J. Mena, C. Z. Serpan. l
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R-1 , i i l i - _ _ _ _ _ _ _ _ _ _
E 4
- 15. NURD3/m-4652, Ccruote Che5=nt Aging and Its Significance Relative i to Life Extension of Nuclear Power Plants," Septecber 1986, D. J. Haus. l
- 16. NUREG/CR-4179, Agirg ard Service Wear of Hydraulic and Maianical Srm*+*rs Used on Safety-Related Piping and Cwyx a= of Nuclear Pomr Plants, S. H. Bush, P. G. Haasler, R. E. Do&Je (PNL) , (to be published) .
- 17. NUREG/m-4590, Aging of Nuclear Station Diesel Generators: Fvaluation of' Operating ard Expert Experience. Volume 1, June 1984, F. F. Dyer, M. P. May.
- 18. PNL-5722, Opezating Experience and Agirq Assessment of EOCS Punp Room Coolers, Ct+hr 1986, D. E. Blahnik, R. L. Goodman.
- 19. NURD3/m-4156, Operating Experience and Aging-Se4=mir Am====marit of Electric Motors, M. Subudhi, E. L. Burns, J. H. Taylor (RE) (to be published). ,
- 20. NURD3/CR-4564, Operating Experience and Aging-Seismic Anaamament of Battery Charge 1n and Invertors, June 1986, W. E. Gunther, M. Subudhi, and J. H. Taylor (BE) . ,
- 21. NUREG/CR-3956, In-Situ % sting of the Shippingport Atanic Power Station Electrical Circuits, February 1986, N. R. Dinsel, M. R. Donaldson, F. T. Soberano (INEL) . (NAFT).
- 22. NURD3/CR-4740, Nuclear Plant Aging Researdi en Reactor Protection System, September 1986, L. C. Meyer (INEL) . (NAFT).
- 23. NURD3/m-4731, Residual Life Aa== ment of Major Licht Water Reactor
, CwpT=Lts, August 1986, V. N. Shah ani P. E. MacDonald (INEL) .
(WAYT).
- 24. NURD3/CR-4234, Vol.1, Aging and Service Wear of Electric Motor-operated Valves Used in Engineered Safety-1%ature Systems of Nuclear Power Plants, July 1985, W. L. Greenstreet, G. A. Murphy, D. M. Eissenberg (CENL) .
- 25. NURD3/CR-4302, Aging ard Service Wear of Check Valves Used in Engineered Safety-Feature Systams of Nuclear Power Plants, Decenber 1985, W. L. Giser=Loet, G. A. Murphy, R. B. Gallaher, !
D. M. Timaanberg (CENL) .
- 26. NURD3/CR-4380, Evaluation of the Motor-operated Valves Analysis and Test System (MNATS) to Detect Degradation, Irwinct Adju=Lera, ard other Abnormalities in Motor-Operated Valves, January 1986, J. L. Crowley, D. M. Eissenberg (CENL) .
R-2
)
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