ML20205N585
| ML20205N585 | |
| Person / Time | |
|---|---|
| Issue date: | 04/13/1999 |
| From: | Ashley D NRC (Affiliation Not Assigned) |
| To: | Stewart Magruder NRC (Affiliation Not Assigned) |
| References | |
| PROJECT-689, RTR-REGGD-01.082, RTR-REGGD-1.082 NUDOCS 9904190023 | |
| Download: ML20205N585 (21) | |
Text
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E NUCLEAR REGULATORY COMMIS310N "f
W.hSHINGToN, D.C. 20555-0001
/
April 13, 1999 1
MEMORANDUM TO: Stewart L. Magruder, Project Manager Generic issues and Environmental Projects Branch Division o' i% actor Program Management THRU:
Richard P. Correia, Chief y
( h w.,.
Reliability and Maintenance Section Quality Assurance, Vendor Inspection, Maintenance and Allegations Branch Division of Inspection Program Management 7)h(
FROM:
Donnic
',shley, Operations Specialist Reliability and Maintenance Section Quality Assurance, Vendor Inspection,
~
Maintenance and Allegations Branch Division of Inspection Program Management
Subject:
SUMMARY
OF APRIL 6.1999, MEETING BETWEEN THE NUCLEAR REGULATORY COMMISSION (NRC) AND THE NUCLEAR ENERGY INSTITUTE (NEI) REGARDING THE CHANGE TO 10CFR50.65 On April 6,1999, members from the NRC staff met with representatives from Nuclear Energy Institute (NEI), in a public meeting to discuss issns related to the change to 10CFR50.65(a)(3).
NEl made no opening remarks. Discussions at this meeting continued the dialogue and interchange between NRC and NEl from similar meetings held previously (May 21,1998 and July 28,1998) to discuss guidance regarding the conduct of assessments prior to performing g
maintenance. The staff gave a status on the rule change, including recent presentations to the 7g-Advisory Committee on Reactor Safegeard:, (ACRS) and planned presentations to Committee eX for the Review of Generic Require.nents (CPGR). The staff informed NEl that the plans are to have final wording on the rule change to the Commission in May 1999.
The staff made an overview presentation on the basic requirements of the rule change and diso, e sed methods which could be used to reach compliance with the rule (see Attachment 2).
This information was obtained from the draft regulatory guide DG 1082 " Assessing and Managing Risk of Maintenance Activities At Nuclear Power Plants"(see Attachment 3).
Representatives of NEl expressed concern about the scope of the plant structures, systems and components (SSCs) covered by the rule and methods to evaluate licensees' implementation of the rule change. They exprersed their continuing interest in risk informing the Maintenance Rule and were continuing their work on a revision to NUMARC 93-01, 4
" Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants".
However, NUMARC 93-01 would be revised using an NEl versien of the rule. These issues j
were addressed in a letter to the Chairman from NEl dated March 17,1999.
9904190023 990413 PDR REVQP ERONUNRC QQ-OfC ff r
M M EM M aM 1
f S. M;grud:r April 13, 1999 The s'aff provided the audience with copies of the draft regulatory guide DG-1082. Meeting attendees were advised that the (information only) draft regulatory guide and the summary of the meeting would be posted on the Maintenance Rule Web Site as soon as possible. NEl stated that they wanted to have another Maintenance Rule workshop later in the year and invited the NRC staff to participate. NEl would advise the NRC of the date when finalized.
l Project No. 689 cc: See next page Attachments: 1.
Attendance List 2.
NRC handouts used during the meeting 3.
Draft Regula".ory Guide DG-1082 DISTRIBJTION Hard Cooies Central Files OGC ACRS lOMB R/F PUBLIC EMAIL*
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DATE 4 /If/99 i 4 /IN90
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OfflCIAL RECORD COPY I
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a 3-Nuclear Energy Institute Project No. 689 4
cc:
Mr. Ralph Beedle Ms. Lynnette Hendricks, Director f
Senior Vice President Plant Support and Chief Nuclear Officer Nuclear Energy Institute Nuclear Energy Institute Suite 400
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Suite 400 1776 i Street N.W.
j 1776 i Street N.W.
Washington, D.C. 20006-3708 Washington, D.C. 20006-3708 Mr. Nicholas J. Liparulo, Manager Mr. Alex Marion, Director Programs Nuclear Safety and Regulatory Activities Nuclear Energy Institute Nuclaar and Advanced Technology Division Suite 400 Wess :.ghouse Electric Corporation 1776 i Street N.W.
P.O. Box 355 Washington, D.C. 20006-3708 Pittsburgh, PA 15230 Mr. David Modeen, Director Engineering Nuclear Energy institute Suite 400 1776 l Street N.W.
Washington, D.C. 20006-3703 Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy Institute Suite 400 1776 i Street N.W.
Washington, D.C. 20006-3708 Mr. Jim Davis, Director Operations Nuclear Energy Institute Suite 400 1776 i Street N.W.
Washington, D.C. 20006-3708
I 1
Attendance Li.st NAMEITITLE i,
~ '$., ' p TELEPHONE $,* ORGANIZATION AFFlUATION
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Donnie Ashley 30',-415-3191 NPR/DIPM/lOMB Rich Correia, Section Chief 301 d.15-1009 NRPJDIPM/lOMB Ted Quay, Branch Chief 301-415-2402 NRR/DIPM/lOMB Claudia Craig 301-415-1832 Comm. Dieus' Office Marc Miller 301-415-1972 OCM/SJ Nanette Gilles 301-415-1180 NRR/ADPR/TSB Mike T. Schiltz 301-415-1733 OEDO Wayne Scott 301-415-1020 NRR/DIPM/lOMB Amarjit Singh 301-415-6899 ACRS 2-R. M. Latta 301-415-1023 NRR/DIPM/lOMB Dan Steneger 202-371-5742 Winston and Strawn Michael Knapik 202-383-2167 McGraw-Hill Kim Green 301-255-2289 NUSIS Dean Raleigh 301-417-4868 Bschtel Power Corp Bi'f Bradley, Sr. Project Manager 202-739-8083 NEl Tony Pietrangelo, Director Licensing 202-739-8081 NEl Francis X. Talbot 301-415-3146 NRR/DIPM/iOMB i
See-Meng Wong 301-415-1125 NRR/DSSA/SPSB
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U.S. NUCLEAR REGULATORY COMMISSION April 1999
[
g OFFICE OF NUCLEAR REGULATORY RESEARCH Division 1
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Draft DG-1082 5
+..../
DRAFT REGULATORY GUIDE
Contact:
R. P. Correia (301)415-1009 l
l DRAFT REGULATORY GUIDE DG-1082:
ASSESSING AND MANAGING RISK BEFORE MAINTENANCE ACTIVITIES"AT; NUCLEAR POWER PLANTS i
A. INTRODUCTION The NRC staff has amende) the Maintenan'ce R fe[10 CFR 50.65, by adding a
%'f new paragraph (a)(4)-
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- Before performing maintenance actidities (inclUdiNd[tskhotYmited to, surveillances, post-maintenance testing, and corrective maintenanbe'end preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities." (
.c As of July 1998, Maintenance Rule Baseline inspections at all U.S. nuclear power plant sites were comf'letedq NRC staff experience during the baseline inspections indicated that all licenseespave jieveloped programs to implement the pre-maintenance assessment provision of the original paragral:h (a)(3). Howeverl the baseline inspections icientified a number of instances in which t%se assessmeilts were not performed (including some that caused a signifm..t inerease in risk) and identified weaknesses lin licensees' programs that could result in failures to perform ac' equate assessments priot to maintenance activities. Partly because of Mese inspection findings, the Commission approved the amendment to ensure that licensees assess and manage increases in risk associated with ma ntenance activities.
f iW
/d if a series of pubhc meetings, the NRC staff has met with indust y representatives to discuspthe change in the rule in relation to proposed revis, ions to NIA4ABC 93-01, " Indust $ Guide'ine for Monitoring the Effectiveness of Maintenance at Nuclear Poipgants" (MaygG3). NUMARC 93-01 was prepared ty the Nuclear Energy Institute (NEI) and itjedndorsedby Revision 2 of Regulatory Guide 1.160, *Monitonng the Effectiveness of Maintensnos et. Nuclear Power Plante." Partially based on these dis:ussions with industry I
representativeh, this Draft Regulatory Guide DG-1082 is being developed to propose guidance on implementing the provisions of 10 CFR 50.65(c.)(4). The finaf version of DG-1082 is meant to be used with Regulatory Guide t.160 as guidance on,mothods acceptable toithe NRC staff for assessment processes befoto maintenance activities to manage the risk from maintenance activities.
1 s
The information collections contained in this draft regulatory guide are covered by the requirements of 10 CFR Part 50, which were approved by the Office of Management and Budget, approval number 3150-0011. The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.
B. DISCUSSION OBJECTIVE The objective of the Maintenance Rbe,10 CFR'5b.65, is to require monitoring of the overall continuing effectiveness of licensee maintenance programs to ensure (1) that safety-related and certain nonsafety-rela.ed structures,,, systems, and components (SSCs) are capable of performing their intended functions, (2) fdr hensafety-related equipment,'that
^
failures will not occur that prevent the fulfillment of safety related functions, and (3) that failures resuiting in scrams and unnecessary actuations of safety 7related syitems are minimized. The intent of 10 CFR 50.65(a)(4) is to require that licensees perforrn assessments before mainter ance activities are performed on SSCs covered by the Maintenance Rule and to manage risk that may result from the proposed activities.
l, g
- swA L:c V
SAFETY ASSESSMENTS FOR MAINTENANCE ACTIVITIESi y
ng Paragraph (a)(4) of the Rule requires that licensees assess and manage ritk that may resuh from maintenance activities durMg all modes of plant operation (i.e., including all normal power and shutdown conditions). An appropriate assessment would include a review of the current configuration of the plant and the plant configu' ration expected during the planned mantenance activity. Assessing the current plaqt:config0 ration as well as changes to plant conJguratior) expected from the planned maintenance activities is intended to ensure that the plant is nct inadvertently p? aced.in riskfsignificant config'Urations. These assessments do not necessarily reque e that alqusntitative assessme.nt of probabilistic risk be performed. The level of sophhtication%lth which such assessments are performed is expected to vary, ba. sed on circunstances iMolvMMt should be"isiderstood, however, that the contribution to risk cf a specific plant configurationA(Js'nds'on both t'ie degree to which the safety functions are degraded and the ouratio, for which the plant.isjn that configuration. Furthermore, assessing the degree of safety functior giogradation nec'esaltstes:an understanding of the impact of maintenance activities on the I
capabiliU of the plant to prevent or mitigste accidents and transients, as well as an understanding of the p; wtial impact of extemal conditions (e.g., incle" Mt weather, electrical grid instabllity, floodirt or seismic evenis) on plant meintenance conf!pations. The assessments may range froth {orrned determinisic judgments to the use of an on-line, probabilistic risk assessment (PRA) i tod 4g e. x 9
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These assessments shoJd consider the planned maintenance activities h
y I
fcr at is i ciuded in the scope of the rule specified by 10 CFR 50.65(b). Emphasis should be placed on assessments of the SSCs modeled in the hcensee's PRA in addition to au SSCs considerec to be risk significant by the licensee's maintenance rule expert panel. The I
assessments should also consider combinations of low safety (rish) significar,t SGCs for planned maintenance that could renuit in hips ssfety \\(nsk) sig ficant situations. To reduce burden, licensees may perform a *cne-time assessment to identify low nsk-significant SECS from maintenanw activh es or combinations of maintenance activities with low safety impact.
2-
i An assessment should be reevaluated following the discovery of emergent failures or changes in plant conditions to determine tne safety impact of the failure or change in plant conditions. However, the reevaluation of prior assessments should not interfere with, or delay, the operator and maintenance crew from taking timely actions to restore the apprcpriate SSC to service or taking compensatory actions necessary to ensure plant safety is maintained, if the SSC is restored to service before the assessment is performed the evaluation need not be conducted.
The process for performing these assessments should be scrutable and repeatable. Known limitations in the assessment process should be described in the licensee's assessment process in accordance with 10 CFR 50.65(a)(4) that is part of the maintenance rule program documentation. The licensee's maintenance rule expert panel should review this process to ensure that the process is sufficiently robust and compreshensive.
4
- c. REGULATORY POS[DQN ASSESSMENTS FOR MAINTENANCE ACTIVITIES DURING POWER OPERATING CONDITIONS -
j 4m gy y Licensees should perform k to be implemented during power operating conditionQ{segaments of myint$ n Power operating conditions are defined a s plant modes other than a hot shutdown, cold' shutdown, ' refueling',)dt defueled condition.
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%w This guidance should be applicable to the two methods commonly used to evaluate the risk impact of plant mainterance configurations: (1) using a plant " risk monitor" b computer-based risk analysis tool);and (2) using a matrgof pre-analyzed plant configurations Most plant risk monitors are customized to evaluate the Wsk impact of maintenance activities on SSCs used tofmitigate events,~as weil as SSCsjthat may initiate events (e.g., switchyard maintenapop)dThe adequacyfsp(quality of this assessment tool depends on the fidelity of the PRA modd3ndCie' accuracy of the input assumptions. It ts expected that the scope of the PRA mode! in a plant r11sk monitor should reflect the "as-built, as operated" plant configuration to limit the underestimation bf dak Associated with maintenance config trations. Additionally, full requantification (rpther tharicutset Editing) of the PRA modC for the assessment of each maintenance co~nfiguration isl desirable to ensure a greater fk elity of results when multiple components are involved?,Qk
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,f If a matrix of pre-analyzed plant confi urations is used for the t
asr.essment, the limitations of the risk matrix should be clearly i 3entified and the users of this tool should have sufficient kjj6wledge and familiarity with the tool's li, litations. The adequacy of the s
' ' sessment tools should be etaluated by the licensee's e; cert panel to determine the p
7The known limitations of the assessment toc should be described in the lic nance Rule program documentation, and training on the limitations should be provided.% #
The sophistication of M assessment for eatuating the risk of a maintenance configuration should be commensurate with the compk (ity of the configuration. The level of sophistication and attiibutes of the assessment for evaluating 'ncreases in risk arising from maintanance during power operating conditions are as follows.
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1.1 Maintenance Activity on a Single SSC j
(1)
If the planned maintenance configuration consists of one SSC or one SSC traia only, qualitative assessment based on the j
informed judgment of a trained licensed operator is sufficient to evaluate the safety impact of the maintenance activity on the SSC or SSC train.
(2)
The operators making the informed judgment should be knowledgeable of the SSC's contribution to plant risk, or which key safety functions are degraded by the maintenance on the SSC.
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Tha licensee's assessrnent process should containprovisions to evaluate the safety; impact'when emergent failures, changes in the plant,20r chhnges in external conditions occur, j
(4)
The operators shoudbe'awareM the potential impacts of external conditions (e.g., inclement weather) and other activities (e 9.iswitchyard maintenance) that may increase the likelihood.of,an iptiating event..
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a 1.2 Maintenance Activities on Twd $$Cshj D.i "
(1)
If.the planned maintenance configuration consists of any two gSSCs, eit@r a qualitative or quantitative assessment should
,,be used to managq chariges in risk to ensure that the plant is pf not inadvertently placed in a risk-significant configuration or in 4
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TgJTQelicensee's assessment process should establishYg i
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- slNdify the levels of risk that would require certain actions Tj (e'.g., management approvac, or contingency planning).
(3) $ ' "
The assessments may be based on evaluations of precetermined configurations (e.g., various combinations of M
L two SSCs), or the assessmen*s could be perforraed on an
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"as-naeded" basis.
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.kg 1 The liter;see's assessment promss should contain provisions 1,y A ;
9" to evallate the safety impact when emergent failures,
^'.A changes in the plant, or changes in external conditions occur.
(5)
The licensee's process should contain provis;ons to ensure tnat plant configurations not recognized by the assessment tool are e valuated by appropriate personnel such as a risk analyst or the expert panel. This expectation does not apply to those combinations of two SSCs that were determined through a one-time assessment to have little or no contributior-to plant risk.
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1.3 Maintenance Activities on N.ne Than Two SSCs (1)
If the planned maintenance corfiguration consists of more than two SSCs, the assessment should be strongly focused on risk insights derived from PRA analyses.
(2)
The assessments may be based on evaluations of predetermined configurations (e.g., various combinations of multiple SSCs), or the assessments could be performed on an "as-needed" basis.
(3)
The licensee's process should contain provisions to ensure that plant configurations not recognized by the; assessment tool are evaluated by; appropriate personnel such as a risk analyst or the expert pane:,' This expectation does not apply to those combinations'of multiple SSCs that were determined through a one-timgassessment to have little or no contribution to plant risk.
(4)
The licensee's safety assessment pr ess should establish guidelines for managing risk, i.e.,-the process should specify the levels of riskincrease that would require certain actions, e.g,Lmanag sent approval or" contingency planning.
(5)
The licensee's assessment process should contain provisions
- 7. Sto evaluate.the safety impact following emergent failures or
!E changes in' plant conditions after the assessment of the planned configuration.
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2.
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- x. ASS $SSMENTS FOR MAINTENANCE ACTIVITIES DURING E 9 (SHUTDOWN CONDITIONS L7, f,f
' 2The_ performance of safety assessments for maintenance activities during shutdown conditionsinvolve the same general guidance as described above. However, there are some considerations that are different than for power operation conditions.
Sophisticated quantitativej.isssssment tools are not gen.erally available, as PRA models for shutdown plant conditions are not widely used. If a PRA modelis not available, an assessment of degradation of the keisafety functions for shutdown conditions should be made for any malmenance activity on the in-scope SSCs.
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h Key safety functions for shutdown conditions are decay heat removal capability, reactor coolant inventory control, electrical power availability, reactivity control, and containment closure (primary and secondary). Assessments for shutdown maintenance activities must take into account outage conditions and plant configurations that impact key safety functions. For example, assessment of maintenance activities that impact the decay heat removal capability should consider these key aspects:
Initial magnitude of decay heat Time *o reactor coolant boiling r
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l Time to uncover the core l
Time to achieve containment c;osure Initial reactor coolant systern inventory condition (e.g., filled, reduced, midloop, refueling canal filled, reactor cavity flooded)
,.. m Reactor coolant system co.nfiguration'(e.g., open or closed, l
nozzle dams installed or loop isolation valves closed, steam generator manways on 'or off, vent paths"available, temporary covers or thimble tube plugs installod main i
steam line plugs installed)
Natural circulation capahlity with heat transfer to steam generator shell sideg qMe w
m Instrumentation needed to mbnitof the conditions 8&
&W Additional aspects of the fuel handling and core alteration activities should be considered. These considerationsph'ould address systems needed to mitigate fuel l
handling accidents such as radiation monitdring and. ventilation and filtration systems. In l
addition, spent fuel cooling capability should be copsidereo'.g,
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3.
RISK-SIGNIFICANT CONFIGURATIONS
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equipment outages for which~,k-significant configurab$is (1) a configuration of co A ris the incremental contribution to the annuai risk is substantial or (2) l a configuration that would'signdicantly affect tiie"peirformance of safety functions. Because the i
risk of equipment-outage configurations depends on the maintenance configuration and duration of the configuration; the, risk metnce for evaluating the risk significance of a plant configuration should be estimated increases in oore' damage probability (CDP) or large early release
$1he' CDP (or LERP)is the integration of expected core damage frequency probability (LERP)E5d frequency) over"a specified time interval. Regulatory Guide 1.1 l
s (or large early refe l
Approach for Using ProbabillMic. Risk Assessment in Risk-Informed Decisions on Plant-Specific Changeslo the Licensing % asis;"2 provides risk acceptance guidelines for allowable small heresses above the plant baseline CDF (or LERF) for a permanent plant change, whereby the allowed increases must be" consistent with the intent of NRC's Safety Goal Policy Statement.
The NRC staff's position;ls that an equipment-outage configuration becomes risk significant when the increase in CDP (or LERP) exceeds a pre-determined level.
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yNQQ7 Because of variations in riuk profiles between nuclear plants, quar 1tilme screening criteria for risk-significant configurations should be established on a ptant-specific basis to reflect the p; ant risk profile. For temporsry risk changes associated with equipment-outage configurations, relative criteria could be used on a sliding scale to determine risk significance for base CDF values less than 1E-4/yr and base LERF values less than 1E-5/yr.
The temporary risk-acceptance criteria should be based on increases in CDP (or LERP) that account for risk contribution that is due to the duration of the configuration. If the base CDF l
value of a plant is in the range between 1E-5 and 1E-4/yr, a relative change not exceeding a 10 percent increase in the CDP (or LERP) estimate could be selected to establish the non-risk significant demarcation for a given configuration. If the base CDF value of the plant is below 1E-i i
1E-5/yr, a relative change not exceeding a 50 percent increase in the CDP (or LERP) estimate could be used. This approach would limit the cumulative risk impact of all equipment-outage configurations to low CDP values, i.e., less than 1E-6. Thus, a threshold probability of 1E-6 could be used as the criterion for determining the risk significance of a plant maintenance configuration.
4.
QUALITY AND SOPHISTICATION OF SAFETY ASSESSMENTS The quality and sophistication of the safety assessment tools are important to assure appropriate analyses of plant configurations for risk management purposes.
The assessment tools should be of sufficient quality to identify risk-significant configurations. If the assessment tool is a plant-specific PRA, the PRA model used for evaluating" plant-configuration risks should reflect the "as-built and as-operated" plant.' In addition,' the quality of PRA models and assumptions used for this analysis should reflect the current industry practices of achieving meaningful PRA results. Thus, the PRA model and its database should be updated when necessary to account for plant design modifications and changes in operational practices and equipment reliability. This ensures that the quantitative assessment results adequately reflect the modeled contributors to risk.
5.
MANAGING RISK
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The assessments provide insights to identify and limit risk-significant mairnenance activities and their duration. Licensees should avoid entering risk-significant plant configurations or places where a key safety function has been 'significantly degraded when conducting maintenance activities. If a plant configuration is planned that exceeds the risk-acceptance guidelines and the maintenance activity needs to be conducted, the licensee should implement the following:
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,3 Duratjokof the niaintenance"hetivity should be minimized through preplanning and prestaging necessary equipment.
Plant management approval should be obtained before s entering the configuration, Compensatory actions and contingency plans should be
', ; q' implemented if possible, and 7y 7i Site personnel should be at a heightened state of risk
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awareness while the plant is in the risk-significant I
configuration or in a condition in which a key safety function is significantly degraded.
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Implementint these practices is a prudent approach to ensure that the risk ofrbS$tenance activities involving potentially risk-significant configurations is effectively managed.
D. IMPLEMENTATION The purpose of this section is to provide irformation to licensees and applicants regarding the NRC staff's plans for using this regulatory guide.
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];o This draft regulatory guide has'I'fbeen relesed}\\-r,4to encourage public ggb) ;,U participation in its development. Except in those casesin which a lisenseoloi applicant proposes an acceptable alternative method for complying wit (MMdlIled_ portions of the NRC's regulations, the method to be described in the active guide ~ reflecting,pubBc comments will be used with Regulatory Guide 1.160 in the evaluation of a'ssessment processes before maintenance activities to manage the risk from maintenance activities.
1 REGULATORY ANALYSIS
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if kQwaye. lll The Regulatory Analysis 'hefwas published with the rulemaking for the A
t amendment to the Ma;ntenance Rule,10 CFR.50.65(a)(4), is also applicable for this Draft Regulatorf Guide DG-1082.* A copy of the regulatory analysis is available for inspection or copying for a fee _in the Commission's Public Document Room at 2120 L Street NW.,
Washington, DC,#under task DG-1082M p
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