ML20140C734

Ack Receipt of Response to Providing Comments on Draft RG 1051, Monitoring Effectiveness of Maint at Npps

ML20140C734
Person / Time
Issue date:	04/15/1997
From:	Boger B NRC (Affiliation Not Assigned)
To:	Ortciger T ILLINOIS, STATE OF
References
TASK-*****, TASK-RE NUDOCS 9704180114
Download: ML20140C734 (3)

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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 4001

• • • • • ,o April 15, 1997 Mr. Thomas W. Ortciger, Director i Department of Nuclear Safety State of lilinois 1035 Outer Park Drive Springfield, Illinois 62704 1

SUBJECT:

RESPONSE TO OCTOBER 31,1996, LETTER PROVIDING COMMENTS ON DRAFT REGULATORY GUIDE DG-1051, " MONITORING THE EFFECTIVENESS t

OF MAINTENANCE AT NUCLEAR POWER PLANTS"

Dear Mr. Ortciger:

Thank you for providing comments on Draft Regulatory Guide 1051 (DG-1051), " Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," which was the proposed Revision 2 to Regulatory Guide 1.160 (RG 1.160) of the same title. In March 1997, the staff I

issued RG 1.160, Revision 2, a copy of which is enclosed. This letter provides the staff

response to the comments on DG-1051 provided in your letter dated October 31,1996.

Your letter provided general comments that stated that while the maintenance rule and regulatory guide were a step in the right direction, the rule and regulatory guide fell"short of.

the mark" and would not result in a consistent level of maintenance across the industry. The comments also noted disappointment that the NRC has not established standards for Individual Plant Examinations (IPEs) or probabilistic risk assessments (PRAs). Overall, the comments appeared to indicate that you desire that the NRC prescribe specific performance criteria for structures, systems, and components (SSCs) within the scope of the maintenance rule, and provide detailed guidance on how to implement the rule to ensure that SSC reliability and unavailability criteria and maintenance programs are consistent throughout the industry. The comments stated that this lack of consistency in program implementation would make consistent enforcement of the rule difficult.

The intent of the maintenance rule was to provide licensees maximum flexibility in implementing the requirements of the rule. The regulatory guide and NUMARC 93-01,

" Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants,"

were intentionally developed to provide only general guidelines for the activities that constituted a program that would comply with the requirements of the maintenance rule. This flexibility included the establishment of appropriate performance criteria and the extent to which the licensee uses IPE/PRA. It was recognized that this flexibility would result in different approaches to implementing the rule, and different performance criteria for similar SSCs at different plants. As noted in footnote 6 of RG 1.160, Revision 2, the staff is [_)

developing guidance that addresses the acceptable criteria for the use of PRAs in risk-informed regulatory matters. When completed and issued, the staff will use this criteria in this and other risk-informed regulations.

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T.W. Ortciger Due to the differences between the design of plants, the staff believes that it would be impractical, if not impossible, to impose standard reliability and unavailability performance criteria. During maintenance rule baseline inspections, the staff has observed variation in the values chosen by licensees for their performance criteria. This variation has not been a concem; however, where licensees have been unable to demonstrate that the performance criteria chosen were commensurate with safety, the staff has issued violations against the maintenance rule.

In order to allow licensees to use existing maintenance programs to the maximum extent practical, the staff intentionally did not provide detailed guidance on how to establish a program to comply with the rule. This has resulted in programs that are unique at each plant, although there exist common elements and definitions as described in NUMARC 93-01.

The fact that each maintenance rule baseline inspection must evaluate a different program does make the inspection challenging. However, the staff established a thorough training program for inspectors participating in the basefire inspections and has established a rigorous process for determining the appropriate enforcement for violations identified during the inspections.

The training program for inspectors it cludes a three-day course specific to the maintenance rule, the regulatory guide, NUMARC 93-01, and the associated inspection procedures. In addition, the Office of Nuclear Reactor Regulation provides a staff support person (and sometimes additionalinspectors) from the Quality Assurance and Maintenance Branch (the branch with programmatic responsibility for the maintenance rule) to provide guidance on staff positions related to the maintenance rule and to promote consistency between inspections.

The enforcement process established for the maintenance rule includes an enforcement panel that convenes specifically to review potential violations of the maintenance rule. The panel consists of representatives from the appropriate regional office, the Quality Assurance and Maintenance Branch, and the Office of Enforcement. The purpose of this panelis to provide consistency of enforcement across the industry by ensuring that similar findings at different sites are given similar assessments and result in similar enforcement actions.

In addition to comments on DG-1051, you also provided a copy of your comments to an earlier version of the regulatory guide (DG-1020) with a statement that many of these comments still apply to DG-1051. The staff did not specifically address the comments that I were originally provided on DG-1020 because they had been addressed at the time of their original submission, and the staff has not changed any positions since the earlier draft regulatory guide.

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T.W. Ortciger 3- l l

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If you have any questions regarding the staff response to your comments, please contact Mr. Richard P. Correia, Chief, Reliability and Maintenance Section, Quality Assurance and Maintenance Branch, (^*} 415-1009.

Sincerely,

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Bruce A. Boger, Director Division of Reactor Controls and Human Factors Office of Nuclear Reactor Regulation

Enclosure:

RG 1.160, Revision 2 l

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OFFICE OF NUCLEAR REGlJLATORY RESEARCH REGbLATORY GUIDE 1.160 (Draft was DG-1051)

MONITORING THE EFFECTIVENESS OF MAINTENANCE AT NUCLEAR POWER PLANTS A. INTRODUCTION consistent with the NRC's defense-in-depth philoso-The NRC published the maintenance rule on phy. Maintenance is also important to ensure that de-July 10,1991, as Section 50.65, " Requirements for sign assumptions and margins in the original design ba-Monitoring the Effectiveness of Maintenance at Nu- sis are maintained and are not unacceptably degraded. l clear Power Plants," of 10 CFR Part 50, " Domestic Li- Therefore, nuclear power plant maintenance is clearly l censing of Production and Utilization Facilities." The important in protecting public health and safety.

NRC's determination that a maintenance rule was Paragraph (a)(1) of 10 CFR 50.65 requires that needed arose from the conclusion that proper mainte- power reactor licensees monitor the performance or nance is essential to plant safety. As discussed in the condition of SSCs against licensee-established goals in regulatory analysis for this rule,I there is a clear link be. a manner sufficient to provide reasonable assurance tween effective maintenance and safety as it relates to that such SSCs are capable of fulfilling their intended l

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v such factors as the number of transients and challenges to safety systems and the associated need for operabil.

functions. Such goals are to be established commensu-rate with safety and, where practical, take into account ity, availability, and reliability of safety equipment. In industry-wide operating experience. When the perfor-addition, good maintenance is also important in provid. mance or condition of an SSC does not meet estab-ing assurance that failures of other than safety-related lished goals, appropriate corrective action must be Mk-structures, syr tems, at:0 components (SSCs) that could en. For a nuclear power plant for which the licensee has initiate or adversely affect a transient or accident are submitted the certifications specified in 10 CFR minimized. Minimizing challenges to safety systems is 50.82(a)(1) (i.e., plants undergoing decommissioning),

Paragraph (a)(1) of 10 CFR 50.65 applies only to the I

NRC Memorandum to All Commissioners from J. Tavlor on "Mamte. e D at me neme mm mem me peemance nance Rulemak mg." June 27,1991. Copics are ss ailable for mspection or coPHng fon fedrom the NR C Pubhc Document Room at 2120 L street, or conditica of all SSCs associated with storing, con-NW., Washmgton, DC; the PDR's maihng address is Mais stop Lt.-6, trolling, and maintaining spent fuel in a safe condition, W ashmgton, DC 20555; phone (202%34-3273; f ax (202%34-3343.

in a manner sufficient to provide reasonable assurance chNRC ki GULATORY GetDI S The gJoes are msved in the lotioeng ten broad divisions Regulatory Gmdes are esued to debcnbe and maae evelable to the pubiec sum informa-tion as methods acceptable to the NAC staff for amp 6ementeng specAc pwts at tre Com-mess.on a regulations techrvQues used by the staff movaluatng specshe prob 4 ems or pos' 1 Nwer Reactors 6 Products tulated accidents and data needed t y the NRC staff m #ts review of appucanons for pe'- 2 Rosesch and Test Reactors 7 Transportaterm mets and hcanoes Regulatory gedes are not substitutsis for regulations. and compance 3 Fuens and Materies Facilities 6 Occupatiorial Health wrth them a not reguved Methods and solutions d fterent from those set out en the guices a En virarrtents and Siting 9 Anmrust and Financial Review wit be acceptetxe ll they provice a basis for tre fi%ngs reqwade to tre essaance or cori- 5 Maternats and Plant Protection tO Ge serai tenuance of a permit or hcense by the Commisason This gwde was issued after coro6deralson of Comments received from the pubhc Com- Single copies of regulatory gados may be cetained free of chage by wnting the Othce of ments anu suggestsors frar #mprovements in these gwdes are encouraged at m's times and Administration Attention Distritution and Mai Services Section. V S Nuciew Regmatory gJoes mil be rewmed as appropnate to accommodate comments and to reflect newin- Comm#sseon. Wasnsngton, DC20655- 0001 or by fax at (301)415-2200 v/ Wntten comme 1ts may be submrtled to tne Ru6es Review and rkrectwen Rran& DFIPL Aou u s Naciear Revuory comraswan West.ngian Dc 20%,s- ouu tagued gades may also be pur%ebed from the Natronal fechrmcal IrWormation Lentate on a 3+andmg order bases Detals on1%s service may be obtened by wntrng NTIS $295 Port Rova Road. spnnaheid vA 221e1 1 b Faclosure

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". that such SSCs are capable of fulfilling their intended (i) That are relied upon to mitigate accidents

! , tunctions.2 or transients or are used in plant emergen-Paragraph (a)(2) of 10 CFR 50.65 states that moni- cy perating procedures (EOPs); or toring as specified in Paragraph (a)(1) is not required (ii) Whose failure could prevent safety-where it has been demonstrated that the performance or related structures, systems, and compo-condition of an SSC is being effectively controlled nents from fulfilling their safety-related

. through the performance of appropriate preventive function; or maintenance, such that the SSC remains capable of per-forming its intended function.

(iii) Whose failure could caust a reactor Paragraph (a)(3) of 10 CFR 50.65 reqaires that per. scram or actuation of a safety related formance and condition monitoring activities and asso. system.

ciated goals and preventive maintenance activities be Paragraph (c) of 10 CFR 50.65 states that the rule evaluated at least every refueling cycle provided the in- provisions are to be implemented by licensees no later terval between evaluations does not exceed 24 months. than July 10,1996.

The evaluations must be conducted takinginto account, . .

where practical, industry-wide operating experience. This Regulatory Guide 1.160 is being revised to Adjustments must be made where necessary to ensure endorse Revision 2 of NUMARC 93-01, " Industry that the objective of preventing failures of SSCs Guideline for Monitoring the Effectiveness of Mainte-through maintenance is appropriately balanced against nance at Nuclear Power Plants"d (April 1996), which the objective of minimizing unavailability of SSCs be- has been updated by the Nuclear Energy Institute. The cause of monitoring or preventive maintenance. ' c r- regular ry guidance is intended to provide Dexibility I ' " "'#"***

forming monitoring and preventive maintenance acti .

to stmc'ture iq maintenance pmgram in ities, an assessment of the total plant equipment Nt ac'cordance with the safety sigmficance of those SSCs within the scope of the rule.

out of service should be taken into account to detet . .

the overall effect on performance of safety functiu . 'i ne information collections contained in this regu-Paragraph (b) of 10 CFR 50.65 states that the scope latory guide are covered by the requirements of 10CFR of the monitoring program specified in Paragraph Part 50, which were approved by the Office of Maaage- i ta)(1)is toinclude safety-related and nonsafety-related ment md Budget, approval number 3150-001 .The SSCs as follows.

NRC may not conduct or sponsor, and a personl required to respond to, a collection of information un-(1) Safety-related tiractures, systems, or compo. less it displays a currently valid OMB control number. I nents that are relied upon to remain functional during and following design basis events to B. DISCUSSION ensure the integrity of the reactor coolent pres-sure boundary, the capability to shut down the OBJECTIVE i reactor and maintain it in a safe shutdown con-The objective of 10 CFR 50.65 (referred to hereaf-dition, and the capability to prevent or mitigate ter as the maintenance rule or the rule) is to require the consequences of accidents that could result monitoring of the overall continuing effectiveness ofli-in potential offsite exposure comparable to the censee maintenance programs to ensure that (1) safety-guidelines in 10 CFR 50.34(a)(1) or 100.11 of related and certain nonsafety-related SSCs are capable this chapter, as applicable.3 my M Md We W h e (2) Nonsafety-related structures, systems,orcom- s fety-related equipment, failures will not occur that ponents: prevent the fulfillment of safety-related functions, and failures resulting in scrams and unnecessary actuations 2

Tae specific requirements for decommissionmg plants became effective August 28.1996. Sec 61 i R 39278, July 19.1996,-Decommissioning of Nuclear Power Reactors."

3 'This document is available for inspection or copying for a fee in the NRC This Paragraph (b)(l)of the maintenance rule was changed in the final rulemaking for" Reactor Site Critena including seismic and Earthquake Public Document Room,2:20 L street NW. , Washington. DC; the PDR's Engmeeting Cnteria ror Nuclear Power Plants," December I1,1996 See mailing address is Mail Stop 11-6, Washmgion. DC 20555; phone 61 i R 65157. (202)634-3273; fax (202)634-3343.

1.160 - 2 j

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DEVELOPMENT OF INDUSTRY GUIDELINE, PLANT, SYSTEM, TRAIN, AND COMPONENT NUMARC 93-4)1 MONITORING LEVELS m The nuclear industry developed a document, The extent of monitoring may vary from system to

( ) NUMARC 93-4)l," Industry Guideline for Monitoring system depending on the system's importance to safety.

O the Effectiveness of Maintenance at Nuclear Power Some monitoring at the component level may be neces-Plants"(May 1993),4 that provides guidance to licen- sary; however,it is envisioned that most of the monitor-sees regarding implementation of the maintenance rule. ing could be done at the plant, system, or train level.

This document was prepared by NUM ARC. A verifica- SSCs with high safety significance and standby SSCs tion and validation (V&V) effort was conducted by with low safety significance should be monitored at the NUMARC, with NRC staff observation, to test the system or train level. Except as noted in the Regulatory guidance document on several representative systems. Position of this guide, normally operating SSCs with A number of changes were made to the NUMARC low safety significance may be monitored through guidance dacument based on the results of the V& V ef. plant-level performance criteria, including unplanned fort. The NRC staff reviewed this document and fou:id scrams, safety system actuations, or unplanned capa-that it provided acceptable guidance to hcensees. In bility loss factors. For SSCs monitored in accordance June 1993, the NRC staff issued Regulatory Guide with 10 CFR 50.65(a)(1), additional parameter trend-1.160, " Monitoring the Effectiveness of Maintenance ing may be necessary to ensure that the problem that at Nuclear Power Plants," which endorsed the May caused the SSC to be placed in the Paragraph (a)(1) 1993 version of NUM ARC 934)l. In January 1995, the category is being corrected.

NRC staffissued Revision 1 to Regulatory Guide 1.160 to reflect the amendment to 10 CFR 50.65(a)(3) that USE OF EXISTING LICENSEE PROGRAMS l changed the requirement fo'r performing the periodic The NRC staff encourages licensees to use, to the evaluation from annually to once per refueling cycle, maximum extent practicable, activities currently being not to exceed 24 months between evaluations. conducted, such as technical specification surveillance

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From September 1994 to March 1995, the NRC testing, to satisfy monitoring requirements. Such activ-1 s

staff performed a series of nine pilot site visits to verit.y ities could be integrated with, and provide the basis for, I t j the usabihtv and adequacy of the draft NRC mainte- the requisite level of monitoring. Consistent with the Nv / underlying purposes of the rule, maximum flexibility nance rule inspection procedure and to determme the strengths and weaknesses of the implementation of the should be offered tolicensees in establishing and modi-rule at each site that used the guidance provided in NU- i. .vmg their monitoring activities.

MARC 93-4)l. The findings are described in USE OF RELIABILITY-IIASED PROGRAMS NUREG-1523, " Lessons Learned from Early Imple-Licensees are encouraged to consider the use of mentation of the Maintenance Rule at Nine Nuclear reliability-based methods for developing the preven-Power Plants"S (June 1995). The NRC staff concluded tive maintenance programs covered under 10 CFR that the requirements of the rule could be met more con-50.65(a)(2): however, the use of such methods is not sistently across the industry if some clarifying guid- required.

ance was added to NUMARC 93-4)1 to address the fin-dings noted in NUREG-152',. The NRC staff met with SAFETY SIGNIFICANCE CATEGORIES industry representatives 6 a series of public meetings The mai"'enance rule requires that goals be estab-to discuss proposed revisions to NUMARC 93411 that lished commensurate with safety. In order to imple-would address the findings noted during the site visits. ment this requirement, NUMARC 934)1 established Revision 2 to NUMARC 934)1 (April 1996) resulted two safety significance categories, " risk-significant" from these meetings.

and "non-risk-significant." The process for placing SSCs in either of these two categories is described in section 9.0 of NUMARC 934)l. The statements of consideration for the rule use the terms "more risk-Scopies are asailable at current rates f rom the U s. Government Printing significant" and "less risk-significant." NRC inspec-oMwe, Po Bm 37082, Washmgton, DC 20402-9328 (telephone lion procedure (IP) 62706 4uses the terms "high safety

_' (202)S i2-2249). or trom the National Technical lnformation seruce b>

u riting N'Ils at 5285 Port Royal Road. Springfield. VA 22161. Copies significance" and " low safety significance." After dis-( j are asaii. die for inspcotion or coppng for a ree from the NRc Pubhc cussions with m. dustry representatives, the NRC staff

's D"cument Room at 21201. street NW Washington. DC. the PDP's maihng adJress is Mail slop i L-6. Washington. DC 20555 telephone has determined that the preferred terminology is "high (202 M4-3273, f aq202 )634-3343. . .

safety significance,, and ,, low safety sigmricance.,,

1.160 -3

l Some licensees may elect to define other safety signifi- plant operations. Plant management should be aware of cance categories or may elect to define more than two and have the ability to control these activities.

categories, which would be acceptable if these alterna-tive categories are defined in the licensee's procedures EMERGENCY DIESEL GENERATORS and used in a consistent manner. Industry- and NRC-sponsored probabilistic risk analyses (PRAs) have shown the safety significance of SAFETY SIGNIFICANCE RANK 1NG emergency ac power sources. The station blackout rule METIIODOLOGY (10 CFR 50.63) required plant-specific coping a nalyses The NRC staff endorses the "" ofine SSC safety to ensure that a plant could withstand a totalloss of ac significance ranking methodology described in Revi- power for a specified duration and to determine ap-sion 2 (April 1996) of NUMARC 93-01 as an accept- propriate actions to mitigate the effects of a total loss of able method for meeting the requirements of the main- ac power. During the station blackout reviews, most li-tenance rule.6 Ilowever, because of some unique cerwees: (1) made a com,nitment to implement an aspects of the maintenance rule, including the fact P* emergency diesel generator (EDG) reliability program standby SSCs oflowsafety significance are treated the in accordance with NRC regulatory guidance but re-same as SSCs of high safety significance, this endorse. served the option to later adopt the outcome of Generic ment for purposes of the maintenance rule should not be Issue B-56 resolution, and (2) stated that they had or construed as ari endersement for other applications. will implement an equivalent program. Subsequently, These issues were discussed in SECY 95-265, "Re- utilities docketed commitments to maintain their se-sponse to August 9,1995, Staff Requirements Memo- lected target reiiability values (i.e., maintain the emer-randum Request to Analyze the Generic Applicability gency diese: generator target reliability of 0.95 or of the Risk Determination Process Used in Implement. 0.975). Those values could be used as a goal or as a per-ing the Maintenance Rule."4 formance criterion for emergency diesel generatoi reli-ability under the maintenance rule.

APPLICABILITY OF APPENDIX H TO 10 CFR PART 50 Emergency diesel generator unavailability values were also assumed in plant-specific individual plant ex-With regard to the scope of the maintenance rule, as amination (1PE) analyses. These values should be stated in Paragraph (b) of the rule, it is understood that compared to the plant 4pecific emergency diesel gener-balance of plant (BOP) SSCs may have been designed ator unavailability data regularly monitored and re-and built with normal industrial quality and may not meet the standards in Appendix B to 10 CFR Part 50. It ported as industry-wide plant performance .informa-tion. These values could also be used as the basis for a is not the intent of tha NRC staff to require licensees to goal or performance criterion under the maintenance generate paperworl to documtnt the basis for the de-iuh . In addition,in accordance with Paragraph (a)(3) of sign, fabrication, and construction of BOP equipment the rule, licensees must periodically balance unavail-(i.e., BOP equipment need not meet the requirements of Appendix B to 10 CFR Part 50). ability and reliability of the emergency diesel generators.

Each licensee's maintenance efforts should mini-mize failures in both safety-related and BOP SSCs that C. REGULATORY POSITION affect safe operation of the plant. The effectiveness of 1. NUMARC 93-01 maintenance programs should be maintained for the Revision 2 of NUM ARC 9341," Industry Guide-operational life of the facility.

line for Monitoring the Effectiveness of Maintenance at SWITCIlYARD MAINTENANCE ACTIVITIES Nuclear Power Plants,"4 provides methods that are ac-ceptable to the NRC staff for complying with the provi-As noted in the Regulatory Position of this guide, there may be a need to address maintenance activities sions of 10 CFR 50.65 with the following provisions and clarifications.

that occur in the switchyards that could directly affect 1.1 Scope of the Rule

("The staff is developing guidance that addresses the acceptable entena 1.1.1 "Could Cause" Criterion for the use of PRAs in nsk-informed regulatory matters. The NRC staff anticipales that a future revision to this Regulatory Guide 1.160 would DurinS the nine Pilot site visits, the NRC staff rec-reference the guidance. w ben available. to make the NRC staff's guid-ognized that Some licensees interpreted the words in ance n the use of PRA in the maintenance rule consistent with the NRC staff's guidance in other areas of risk-informed regulation. The industrv seClion 8.2.l.5 of NUMARC 93--011o mean that only will be encouraged to use this guidance at that time.

those SSCs that had actually caused a plant scram or 1.160 - 4

safety system actuation needed to be included within 1,1.3 Function Versus System the reope of the rule. The NRC staff N position is that The rulu provides criteria to determine which SSCs the SSCs to be included under the criterion "could must be included within the scope of the rule. Alterna-cause a reactor scram or actuation of a safety system"

(/n) s should not be limited to SSCs that "did cause" or tively, licensees may use a functional basis te determine which SSCs must be monitored within the scope of the "could likely cause." This position was discussed in rule. That is, the licensee may determine all the func-NUREG-1526, " Lessons Learned from Early imple-tions performed by the SSCs and include within the  !

mentation of the Maintenance Rule at Nine Nuclear scope of the maintenance rule only those functions, and t Power Plants"(June 1995).5 Licensees should consider the associated SSCs that fulfill those functions, that  !

the following SSCs to be within the scope of the rule. meet the scoping criteria of the rule.

1. SSCs whose failure has caused a reactor scram 1.1.4 Systems with Multiple Design Functions i or actuation of a safety-related system at their For systems that have multiple design functions site.

the NRC staff's position is that some design functions may be within the scope of the maintenance rule while

2. SSCs whose failure has caused a reactor scram others may be outside the scope of the rule. Failures of or actuation of a safety-related system at a site l components that affect a design function that is within with a similar configuration. the scope of the maintenance rule would require correc-o ve action and monitoring under the rule. For example,

3. SSCs identified in the licensee's analysis (e.g., the compcnents (piping, pumps, and valves) in the FSAR, IPE) whose failure would cause a reac- high-pressure coolant injection system (llPCI) that are tor scram or actuation of a safety-related needed to perform the design function (injection of system. l high-pressure water into the reactor) would be included  !

,The only exception to items 2 and 3 above would within the scope of the rule because this is a safety-be a licensee who has demonstrated by an analysis (e.g., related function of the system. Ilowever, the compo-FSAR, IPE) and by operational experience that the de- nents that are only used for testing (e.g., test loop, sam-VM vdves) might be excluded from the

( sign or configuration of an SSC is fault-tolerant scope of the rule unless they meet another scoping crite-through redundancy or mstalled standby spares such that a reactor scram or actuation of a safety-related sys- rion (e.g., if they could cause failure of a safety-related fem is implausible. In these cases, the licensee may ex- SSC), because these components are not required for clude the SSC from the scope of the rule. the coolant injection function of the llPCI.

1.2 Definition of Maintenance 1.1.2 SSCs Relied Upon To Mitigate For the purposes of the maintenance rule, mainte-Accidents or Transients or Used in Emergency Operating Procedures nance activities are as described in the " Final Commis-sion Policy Statement on Maintenance of Nuclear Pow-Nonsafety-related SSCs that are relied upon to mit- er Plants."7 This definition is very broad and includes igate accidents or transients or that are used in emergen-all activities associated with the planning, scheduling, cy operating procedures (EOPs) are included in the accomplishment, post-maintenance testing, and return-scope of the rule by 10 CFR 50.65(b)(2)(i). NUM ARC to-service activities for surveillances and preventive 93-01 states that only those SSCs that provide a signifi- and corrective maintenance. These activities are con-cant fraction of the mitigating function need to be in-sidered maintenance regardless of which organization cluded in the scope of the rule. The NRC staff considers performs the activity (e.g., maintenance, operations, this to mean that SSCs that are directly used to address contractors). This definition is referenced in the accident or transient or explicitly used in the EOPs NUMARC 93-01. Some licensees have questioned the are within the scope of the rule, as are SSCs whose use guidance because in section 9.4.5 of NUMARC 93-01 is implied and that provide a significant fraction of the an example of a failure that is not a maintenance-mitigating function. Examples of SSCs that should be preventable functional failure (MPFF) is " failures due considered include communications and emergency to operational errors.. ." The operational errors referred lighting systems, which are necessary to successfully to in that example are those that are not associated with t

mitigate accidents and transients and to use the EOPs, a maintenance activity.

l although they may not directly address the accident or transient, or not be explicitly mentioned in the EOPs. 7 53 rn 9430. March 23.1988-1.160 -5 l

An example of an operator action that would not be 1.5 Monitoring Structures an MPFF would be improper closure of a valve while The maintenance rule does not treat structures dif-filling a tank that results in a me vip followed by a ferently from systems and componeats. Exper ence reactor trip. An example of an operator action that with the rule and NUMARC 93-Ol during the pilot site would be an MPFF could be when an operator failed to visits and the initial period following the effective date reopen a suction valve for a pump following post-of the rule indicated that specific guidance for monitor-maintenance testing and the closed suction valve ing the effectiveness of maintenance for structures was caused pump failure during a subsequent demand-needed, as structures present a different situation than L3 Tu.n d.nm do systems and components. The primary difficulty in implementing the rule for structures using NUMARC NUMARC 93--01 states that activities such as 93-01 was in establishing appropriate criteria for per-cause determinations and moving SSCs from the (a)(2) formance and monitoring structures under Paragraph to the (a)(1) category must be performed in a " timely" (a)(1) instead of Paragraph (a)(2).

manner. Some licensees have requested that the NRC staff provide a specific penod that would be considered The etfectiveness of maimenance can be moni-

" timely."To be consistent with the inient of the mainte- tored by using performance criteria or goals, or by con-nance rule to provide flexibihty to licensees, the NRC dition monitoring. While it is acceptable to use perfor-mance criteria or goals, most licensees have found it staff does not consider it appropriate to provide a spe-cific timeliness criterion. Licensees are to undertake more practical to use condition monitoring for struc-and accomplish activities associated with the mainte- tures. With certain exceptions (e.g., primary contain-nance rule in a manner commensurate with the safety ment), structurer do not have unavailability, and rarely significance of the SSC and the complexity of the issue have demands placed on their safety significant func-being addressed. tions (e.g., maintain integrity under all relevant design basis events), which makes reliability monitoring l A MPFFs as an Indicator of Reliability impr etical.

NUMARC 93-01 states that performance criteria An acceptable structural monitoring program for for SSCs of high safety significance should be estab- the purposes of the maintenance rule should have the lished to assure that reliability and availability assump, foH wing attributes.

tions used in the plant-specific safety analysis are main-tained or adjusted. NUMARC 93-Ol further allows the Consistent with the NUMARC 93-01 ap-use of MPFFs as an indicator of reliability. The mainte- proach for systems and components, most nance rule requires that the performance of SSCs be structures would be monitored in accordance with Paragraph (a)(2), provided there is not monitored commensurate with safety; however, the s gnificant degradation of the structure.

ma;ntenance rule does not require that the assumptions in the safety analysis be validated. Licensees who .

The condition of all structures within the scope choose to use their safety analyses as described in of the rule would be assessed periodically. The NUMARC 93-01 must be able to demonstrate how the appropriate frequency of the assessments number of MPFFs allowed per evaluation period is would be commensurate with the safety signif-consistent with the assumptions in the risk analysis. For icance of the structure and its condition.

standby SSCs, this would require, at a minimum, a rea-sonable estimate of the number of demands during that Licensees would evaluate the results of the time period. assessments to determine the extent and rate of any degradation of the structures. Deficiencies l If a licensee desires to establish a reliability perfor-would be corrected in a timely manner I mance criterion that is not consistent with the assump-commensurate with their safety significance, I tions used in the risk analysis, adequate technical justi-their complexity, and other regulatory fication for the performance criterion must be provided.

requirements. l For some SSCs, an MPFF performance criterion may be too small to be effectively monitored and trended as

• A structure would be monitored in accordance required by the rule. In these cases, the licensee should with Paragraph (a)(1) if either (1) degradation

establish performance or condition monitoring criteria is to the extent that the structure may not meet

) that can be monitored and trended so that the licensee its design basis or (2) the structure has de-can demonstrate that maintenance is effective. graded to the extent that, if the degradation 1.160 - 6 l

. were allowed to continue uncorrected until the 1.7 Normally Operating SSCs of Low Safety next normally scheduled assessment, the Significance structure may not meet its dcsign basis. The g .

1.7.1 Cause Determ.ma tm.ns j structure would continue to be monitored in d accordance with Paragraph (a)(1) until the For all SSCs that are being monitored using plant-degradation and its cause have been corrected. level performance criteria (i.e., normally operating SSCs of low safety significance), the NRC staff's posi-tion is that a cause determination is required whenever

= F,or structures monitored m, accordance with

, any of these performance criteria are exceeded (failed)

Paragraph (a)(1), there would be additional in order to determine which SSC caused w criterion to .

I be exceeded or whether the failure was a repetitive degradation-specific condition monitoring MPFF. As part of the cause determination, it would also and increased frequency of assessments until be necessary to determine whether the SSC was within

, the licensee's corrective actions are comp!ete the scope of the maintenance rule and, if so, whether and the licensee is assured that the structure corrective action and monitoring (tracking, trending, I can fulfill its intended functions and will not goal setting) under 10 CFR 50.65(a)(1) should be degrade to the point that it cannot fulfill its de-performed.

sign basis.

1.7.2 Unplanned Manual Scrams l Consistent with the intent of the rule, licenseu In order to monitor the effectiveness of mainte-should use their existing structural monitoring pro- nance for those SSCs monitored by plant-level criteria, grams (e.g., those required by other regulations or NUMARC 93-01 recommends that only those scrams j codes) to the maximum extent practical. that are automatically initiated be counted. The NRC l staff's position is that all unanticipated scrams be con-sidered, including those scrams that are manually initi-  ;

1.6 Definition of Standby ated in anticipation of an automatic scram. The purpose c of this is not to discourage manual trips but rather to en-(' In NUMARC 9341, standby SSCs of low safety ,

sure that operators do not mask a maintenance perfor- i significance must have SSC-specific performance cri- mance issue. Ifineffective maintenance is forcing plant teria or goals, similar to SSCs of high safety signifi- shutdowns, whether the trip is initiated automatically cance. NUMARC 9341 provides a definition of stand- or manually should not affect how licensees address the by. Some licensees have improperly interpreted this maintenance performance issue under the maintenance definition as meamng that SSCs that are energized are rule.

normally operating. As stated in NUMARC 9341, if the SSC only performs its intended function when initi. 1.7.3 Establishing SSC-Specific Performance ated by either an automatic or manual demand signal, Criteria the SSC is in standby. The maintenas 4a requn a that licensees moni-tor the effectiveness of maintenance for all SSCs within Normally operating SSCs are those whose failure the scope of the rule. NUMARC 93-01 allows licen-would be readily apparent (e.g., a pump failure results sees to monitor SSCs of low safety significance with in loss of flow that causes a trip). Standby SSCs are plant-level criteria. NUMARC 9341 notes that some those whose failure would not become apparent until normally operating SSCs of low safety significance the next demand, actuation, or surveillance. Only those cannot be practically monitored by plant-level criteria.

SSCs of low safety significance, whose failure would Licensees must ensure that the plant-level criteria es-be readily apparent (because they are normally operat. tablished do effectively monitor the maintenance per-ing), should be monitored by plant-level criteria. formance of the normally operating SSCs of low safety significance, or they should establish SSC-specific per-SSCs may have both normally operating and f rmance criteria or goals or use condition monitoring.

standby functions. In order to adequately monitor the For example, a licensee determined that the rod effectiveness of maintenance for the SSCs associated position indication system and the spent fuel pool pit V) g with standby functions, licensees should develop SSC-specific performance criteria or goals, or condition cooling system were within the scope of the mainte-nance rule because they were safety-related at the li-monitoring. censee's site. None of the three plant-level performance 1.160 - 7

criteria described in NUMARC 93-01 (unplanned

. automatic scrams, unplanned capability loss factor, or ing a repetitive MPFF. Therefore, the Paragraph (a)(1) unplanned safety system actuations) would monitor the category could be used as a tool to focus attention on effectiveness of maintenance on these systems. There-those SSCs that need to be monitored more closely. It is possibit that no (or very few) SSCs would be handled fore, additional plant-level performance criteria or system-specific performance criteria must be under the requirements of Paragraph (a)(1). However, established. the rule does not require this approach. Licensees could aho take the approach that all(or most) SSCs would be 1.8 Cisrincation of MPFFs Related to Design handled under Paragraph (a)(1) of the rule and none (or Deficiencies very few) would be considered under Paragraph (a)(2) of the rule. Licensees may take either approach.

The third paragraph of Section 9.4.5 of NUMARC 93-01 provides guidance on the licensee's options fol- Dunng me pd.ot site visits, licensees questioned lowing a failure and on whether, as a result of the licens- e er a l rge nmn er of SSCs monitored under Para-ee's corrective actions, subsequent failures would be graph (a)(1) would be used by the NRC as an mdicator considered MPFFs. In particular, this paragraph ad- '

• " '" "'" P#' ' * # "'# ' """

dresses failures caused by design deficiencies. Ideally, sured the licensees that NRC management would not licensees would make design modifications to elimi- "* *""*#' I

• nit red under Paragraph nate the poorly designed equipment. Ilowever,if the li- (a)(1) s an indicator of mam.tenance performance nor censee determines that such an approach is not cost ef.-

would it be used m determimng the systematic assess-fective (e.g., the cost of modification is prohibitive), ment of licensee performance (SALP) grade m. the the licensee has two options:

ma nieunct area. The number of SSCs monitored un-der Paragraph (a)(1) can vary greatly because of factors (1) Replace or repair the failed equioment and that have nothing to do with the quality of the licensee's make adjustments to the preventive mainte- maintenance activities. For example, two identical nance program as necessary to prevent recur- plants with equally effective maintenance programs rence of the failure. Subsequent failures of the could have different numbers of SSCs monitored under same type that are caused by inadequate cor- Paragraph (a)(1) because of differences in the way sys-rective or preventive maintenance would be tem boundaries were defined (a system with three trains MPFFs, and could be repetitive MPFFs. may be defined as one system at one plant while the same system may be defined as three separate systems (2) Perform an evaluation that demonstrates that at an identical plant) or because of differences in the the equipment can be run to failure (as de- way performance criteria were defined at the two plants scribed in Section 9.3.3 of NUMARC 93-01). licensee who takes a very conservative approach to (a

if the equipment can be run to failure, the li- monitoring against the performance criteria would censee can replace or repair the failed equip- have more SSCs in the (a)(1) category). The NRC staff ment, but adjustments to the preventive main- abo cautioned licensee managers that they should not tenance program are not necessary and view the number of SSCs in the (a)(1) category as an subsequent failures would not be MPFFs. indicator of performance since that attitude might in-hibit the licensees' staff from monitoring an SSC under B.9 SSCs Considered Under 10 CI'R 50.65(a)(1) Paragraph (a)(1) when a performance criterion has been exceeded or a repetitive MPFF has occurred. If there is Paragraph (a)(1) of the maintenance rule requires some doubt about whether a particular SSC should be hat goal setting and monitoring be established for all ESCs within the scope of the rule except for those SSCs monitored under Paragraph (a)(1) or Paragraph (a)(2),

the conservative approach would be to monitor the SSC shose performance or condition is adequately con- under Paragraph (a)(1).

rolled through the performance of appropriate preven-lve maintenance as described in Paragraph (a)(2) of the 4ie. In NUMARC 93-01, all SSCs are initially placed 1.10 Use of Other Methods nder Paragr8ph (a)(2) and are only moved under Para-Licensees may use methods other than those pro-raph (a)(1) if experience indicates that the perfor-sance or condition is not adequately controlled vided in Revision 2 of NUM ARC 93-01 to meet the re-quirements of the maintenance rule, but the NRC will 3 rough preventive maintenance as evidenced by the (

determine the acceptability of other methods on a case-

'titure to meet a performance criterion or by experienc- by-case basis.

i 1.160 -8 l

l l

e

2. OTHER DOCUMENTS REFERENCED IN (i.e., equipment in the switchyard) should be consid-NUMARC 93-01 ered for inclusion as defined in 10 CFR 50.65(b).

Q /

NUMARC 9341 references other documents,1,ut NRC's endorsement of N UMARC 9341 should not be D. IMPLEMENTATION considered an endorsement of the referenced documents. The purpose of this section is to provide informa-tion to applicants and licensees regarding the NRC

3. staff's plans for using this regulatory guide.

INCLUSION OF ELECTRICAL .

DISTRIBUTION EOUIPMENT Except in those cases in which an apph. cant or h. -

censee proposes an acceptable alternative method for The monitoring efforts under the maintenance rule, complying with specified portions of the NRC's regu-as defined in 10 CFR 50.65(b), encompass those SSCs lations, the methods described in this guide will be used that directly and significantly affect plant operations, in the evaluation of the effectiveness of maintenance regardless of what organization actually performs the activities oflicensees who are required to comply with maintenance activities. Maintenance activities that oc- 10 CFR 50.65. The guide will also be used to evaluate cur in the switchyard can directly affect plant opera- the effectiveness ofemergency diesel generator mainte-tions; as a result, electrical distribution equipment out nance activities associated with compliance with 10 to the first inter-tie with the offsite distribution system CFR 50.63.

O 1.160 - 9

RP;UIATORY AND BACKFIT ANALYSES Separate regulatory and backfit anaiyses were not prepared for th*is Revision 20f Regulatory Guide 1.160. the level of protection of public health and safety be-The regulatory analysis and the backfit analysis that yond that currently provided by the Commission's reg-were prepared when this guide was first issued as a ulations, and that the costs ofimplementing the rule are justified in view of this increased protection."* The re-draft, DG-1020, in November 1992, are still applica-ble. The backfit analysis prepared for DG-1020 con- gulatory analysis and backfit analysis for DG-1020 are cluded that no backfit was associated with the regulato- available. in the file for Regulatory Guide 1.160, 'orin-ry guide because it was only providing guidance to spection or copying for a fee in the Commission's Pub-implement the existing requirements of the mainte- lic Document Room,2120 L Street NW., Washington, nance rule. The Commission determined, on the basis DC; the PDR's mailing address is Mail Stop LL-6, of the backfit analysis performed for the maintenance Washington, DC 20555: phone (202)634-3273: fax rule, " that backtatting of the requirements in the (202)634-3343, maintenance rule will provide a substantial increase in 56 rn 3 3:o 3 O

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