ML20151U499

Reg Guide 1.175, Approach for Plant-Specific,Risk-Informed Decisionmaking Inservice Testing

ML20151U499
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Issue date:	08/31/1998
From:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
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TASK-*****, TASK-RE REGGD-01.175, REGGD-1.175, NUDOCS 9809110024
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U.S. NUCLEAR REGULATORY COMMISSION Augu;t 1998 O Q[ c o g \\

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REGULATORY GUIDE OFFICE OF NUCLEAR REGULATORY RESEARCH REGULATORY GUIDE 1.175 (Draft was issued as DG 1062)

AN APPROACH FOR PLANT-SPECIFIC, RISK-INFORMED DECISIONMAKING: INSERVICE TESTING A. INTRODUCTION are consistent with currently approved staff positions (e.g., regulatory guides, standard review plans, branch Hackground technical positions) are normally evaluated by the NRC staff using traditional engineerine analyses. In such During the last several years both the U.S. Nuclear Regulatory Commission (NRC) and the nuclear indus-cases, the licensee would not be expected to submit risk mformation in support of the proposed change.

try have recognized that probabilistic risk assessment (PRA) has evolved to be more useful in supplementing Licensee-initiated IST program change requests that go beyond current staff positions may be evaluated by the traditional engineering approaches in reactor regula-staff using traditional engineering analyses as well as tion. After the publication ofits policy statement (Ref.

the risk-m. formed approach set forth.m this regulatory

1) on the use of PRA in nuclear regulatory activities, the guide. A licensee may be requested to subm.t supple-i Commission directed the NRC staff to develop a regu-mental risk m. formation if such information is not pro-

/_s latory framework that incorporated risk insights. That i

vided in the proposed risk-informed inservice testing framework was articulated m.a November 27,1995,pa-

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(RI-IST) program submitted by the licensee. If risk m.-

per to the Commission (Ref. 2). This regulatory guide, formation on the proposed RI-IST program is not pro-which addresses inservice testing (IST) of pumps and vided to the staff, the staff will review the information valves, and its companion regulatory documents (Refs.

provided by the licensee to determme whether the ap-3-8) implement,in part, the Commission pohey state-pl cation can be approved based upon the information ment and the staff's framework for incorporating risk provided using traditional methods, and the staff will insights into the regulation of nuclear power plants.

e ther approve or reject the application based upon the The NRC's policy statement on probabilistic risk review. For those licensee-initiated RI-IST program analysis encourages greater use of this analysis tech-changes that a licensee chooses to support (or is re-nique to improve safety decisionmaking and improve quested by the staff to support) with risk information, regulatory efficiency. One activity under way in re-this regulatory guide describes an acceptable method sponse to the policy statement is the use of PRA in sup-for assessing the nature and impact of proposed RI-IST port of decisions to modify an individual plant's IST program changes by considering engineering issues program. Licensee-initiated IST program changes that and applying risk insights. Licensees submitting risk USNRC R LGULATORY GUIDES The gudes are issued in the following ten broad divmeore Aegulatory Gwdes are issued to desenbe and mah e avalable to the pubitc such informa-tion as methods acceptable to the NAC staM for implementmg specac parts of the Com.

1 Power Reactors 6 Products

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masion s regulations tectvvques used by the staff m avalvating specac problems or pos.

2 Aesearch and Test Aeactors 7 Transportation lulated accidents, and data needed by the NAC staM m its review of appbcations for per.

3 Fueis and Materials Fachties 8 Occupational Heattn mits and hcenses Regulatory guedes are not substitutes for regulations. and compliance 4 Environmental and Scting 9 Antrtrust and Fmancai Review with them e not requered Methods and solutions d#erent kom those set out in the gwdes 5 Materias and Plant Protectm 10 General weil be acceptable if they prowse a basis for the findings req'.ns:te to the essuance or cork Q

tinuance of a permet or kcense by the Commission Sengie copies of regulatory gudes may be ottened kee of charge by woting tne Repro.

TNs gude was issued after consedersoon of comments received from the pubhc Com-

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ss 5-or ments and suggestions for emprovements in these gwdes are encouraged at all times and 9

guides will be revised. as appropnate to accommodate comments and to reflect riew uv or Dy e mail to GAW1 gNAC GOV lasved gudes may also te purcriased kom the Nat onal Tecnnecal information Serwce on Wntten comments may be submmed to the Aules Review and Directives Branch. ADM.

a stanoing order bases Detals on this senrice may be obtained by woting NT!S 5285 Port U S Nuclear Regulatory Commission. Wasnington. DC 20555 0001 Royal Road Spnngheid VA 22161

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PDR REGCD 01.175 R PDR

l information should address each of the principles of ance on the technical aspects that are common to devel-risk-informed regulation discussed in Regulatory oping acceptable risk-informed programs for all ap-Guide 1.174, "An Approach for Using Probabilistic plications such as IST (this guide), inservice Risk Assessment in Risk-Informed Decisions on Plant-inspection, graded quality assurance, and technical Specific Changes to the Licensing Basis" (Ref. 3) and specifications.

l repeated in this guide. Licensees should identify how This regulatory guide provides application-chosen approaches and methods (whether they are specific details of a method acceptable to the NRC staff quantitative or qualitative, traditional or probabilistic),

for developing RI-IST programs and supplements the i

data, and criteria for considering risk are appropriate for information given in Regulatory Guide 1.174. This the decision to be made-guide provides guidance on acceptable methods for uti-IST of snubbers was not addressed in this regula.

lizing PRA information with established traditional en-gineering information in the development of RI IST tory guide, however, licensees inte rested in implement-l ing a RI-IST program for snubbers may submit an alter.

programs that have improved effectiveness regarding the utilization of plant resources while still maintaining native to the NRC for consideration.

acceptable levels of quality and safety.

Relationship to the Maintenance Rule In this regulatory guide, an attempt has been made l

10 CFR 50.65 to strike a balance in defining an acceptable process for l

The Maintenance Rule, Section 50.65," Require-developing RI IST programs withombeing overly pre-ments for Monitoring the Effectiveness of Maintenance scriptive. Regulatory Guide 1.174 identifies a list of at Nuclear Power Plants," of 10 CFR Part 50," Domes.

high-level safety principles that must be maintained tic Licensing of Production and Utilization Facilities,"

during all risk-informed plant design or operational requires that licensees monitor the performance or con.

changes. Regulatory Guide 1.174 and this guide iden-dition of structures, systems, or components (SSCs) tify acceptable approaches for addressing these basic against licensee-established goals in a manner suffi-high-level safety principles; however, licensees may cient to provide reasonable assurance that such SSCs propose other approaches for conside ration by the NRC are capable of fulfilling their intended function. Such staff. It is intended that the approaches presented in this goals are to be established, where practicable, com-guide be regarded as examples of acceptable practice l

mensurate with safety, and they are to take into account and that licensees should have some degree of flexibil-industrywide operating experience. When the perfor-ity in satisfying regulatory needs on the basis of their mance or condition of a component does not meet es-accumulated plant experience and knowledge.

tablished goals, appropriate corrective actions are to be Organization taken' This regulatory guide is structured to follow the ap-Component monitoring that is performed as part of proach given in Regulatory Guide 1.174. The discus-the Maintenance Rule implementation can be used to sion, Part B, gives a brief overview of a four-element satisfy monitoring needs for RI-IST, and for such cases, process described in Regulatory Guide 1.174 as applied the performance criteria chosen should be compatible to the development of an RI-IST program.This process with both the Maintenance Rule requirements and is iterative and generally not sequential. Part C, Regula-guidance and the RI-IST guidance provided in this tory Position, provides a more detailed discussion of guide.

the four elements including acceptance guidelines. In Part C, Regulatory Position 1 addresses the first ele-Purpose and Scope ment in the process in which the proposed changes to Current IST programs are performed in com-the IST program are described. This description is pliance with the requirements of 10 CFR 50.55a(f) and needed to determine what supporting information is with Section XI of the ASME Boiler and Pressure Ves-needed and to define how subsequent reviews will be sel Code (Ref. 9), which are requirements for all plants.

performed. Regulatory Position 2 contains guidance This regulatory guide describes an acceptable alterna-for performing the engineering evaluation needed to tive approach applying risk insights from PRA to make support the proposed changes to the IST program (sec-changes to a nuclear power plant's IST program. An ac-ond process element). Regulatory Position 3 addresses companying Standard Review Plan (SRP)(Ref. 7) has program implementation, performance monitoring, been prepared for use by the NRCstaffin reviewing RI-and corrective action (third element). Regulatory Posi-ISTapplications. Anotherguidance document,Regula-tion 4 addresses documentation requirements (fourth tory Guide 1.174 (Ref. 3),is referenced throughout this element) for licensee submittals to the NRC and identi-report. Regulatory Guide 1.174 provides overall guid-fies additional information that should be maintained in 1.175 - 2

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the licensee 's records in case later review or re ference is ISI inservice inspection needed. The appendix contains additional guidance for IST inservice testing dealing with cortain IST-related issues such as might LERF containment large early release frequency arise during the deliberations of the licensee in carrying out integrated decisionmaking.

LSSC low safety-significant component s

MCS minimal cut set Relationship to Other Guidance Documents NEl Nuclear Energy Institute This regulatory guide provides detailed guidance NUMARC Nuclear Utilities Management Research on approaches to implement risk insights in IST pro-Council grams that are acceptable to the NRC staff. This application-specific guide makes extensive reference Operations and Maintenance (ASME to Regulatory Guide 1.174 (Ref. 3) for general guid-

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ance.

PRA probabilistic risk assessment PSA probabilistic safety assessment Companion regulatory guides (Refs. 4 and 5) ad-dress graded quality assurance and technical specifica.

F'N risk achievement worth risk importance tions, and contain guidance similar to that given in this measure RI ISTguide. SRP chapters associated with the risk-in-RI-IST risk-informed IST (e.g., RI-IST programs) l formed regulatory guides are available (Refs. 6-8). The SRP standard review plan SRP chapters are intended for NRC use during the re-SSCs view of industry requests for nsk-informed program structures, systems, and components changes. SRP Chapter 3.9.7 (Ref. 7) addresses RI-IST THERP Technique for Human Error Rate Predic-and is consistent with the guidance given in this regula-tion tory guide.

USAR Updated Safety Analysis Report In the 1995-1998 period, the industry developed a USNRC U.S. Nuclear Regulatory Commission number of documents addressing the increased use of PRAin nuclear plant regulation.The American Society The information collections contained in this regu.

D of Mechanical Engineers (ASME) developed guide-latory guide are covered by the requirements of 10 CFR lines for risk-based IST (Ref.10) and later initiated Part 50,which were approved by the Office of Manage-ment and Budget, approval number 3150-0011. The code cases addressing IST component importance ranking and testing of certain plant components using NRC may not conduct or sponsor, and a person is not risk insights. The Electric Power Research Institute required to respond to, a collection ofinformation un-less it displays a currently valid OMB control number.

(EPRI) published its "PSA Applications Guide"(Ref.

11) to provide utilities with guidance on the use of PRA information for both regulatory and nonregulatory ap-B. DISCUSSION plications. The Nuclear Energy Institute (NEI) has also been developing guidelines on risk based IST (Ref.

Key Safety Principles 12). These documents have provided useful viewpoints and proposed approaches for the staff s consideration Regulatory Guide 1.174 (Ref. 3) identifies five key during the development of the NRC regulatory guid-safety principles to be met for all risk-informed applica-ance documents, tions and to be explicitly addressed in risk-informed plant program change applications. As indicated in l

Abbreviations Regulatory Guide 1.174, while these key principles are stated in traditional engineering terminology, efforts ASME American Society of Mechanical Engi-should be made wherever feasible to utilize risk evalua-l tion techniques to help ensure and to show that these

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CCF common cause failure principles are met. These key principles and the loca-CDF core damage frequency tion in this guide where each is addressed for RI-IST EPRI Electric Power Research Institute Programs are as follows:

FV Fussell-Vesely risk importance measure

1. The pmposed change meets the current regu.

GOA graded quality assurance lations unless it is explicitly related to a requested

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exemption or rule change. (This principle is ad-HEP human error probability dressed in Regulatory Positions 1.1 and 2.1 of this HSSC high safety-significant component guide.)

1.175 - 3

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Figure 1 Principles of Risk Informed Regulation l

2. The proposed change is consistt9t with the tions made about the impact of the changes to the IST dIfense in depth philosophy. (Regulatory Position program are not invalidated. For example,if the test in-tervals are based on an allowable margin to failure, the 2.2.1) monitoring is performed to make sure that these mar-

3. The proposed change mainta.ms sufficient gins are not eroded. An overview of this process specif-safety margins. (Regulatory Positmn 2.2.2) ically related to RI IST programs is g;ven in this dec-

4. When proposed changes result in an increase tion. The order in which the elements are performed In core damage frequency or riske the increases may vary or occur in parallel, depending on the particu-should be small and consistent with the intent of the lar application and the preference of the program devel-Commission's Safety Goal Policy Statement. (Regu-opers.

latory Positions 2.3,2.4)

Element 1: Define Proposed Changes to the l

S. The impact of the proposed change should be Inservice Testing Program.

monitored using performance The purpose of this element is to identify (1) the l

measurement strategies. (Regulatory Position 3.3) particular components that would be affected by the Regulatory Guide 1.174 gives additional guidance proposed changes in testing practices, including those on the key safety principles applicable to all risk-currently in the IST program and possibly some that are infctmed applications. Figere 1 of this guide, repeated not (if it is determined through new in formation and in-from Regulatory Guide 1.174, illustrates the consider, sights such as the PRA that these additional compo-ation of each of these principles in risk-informed deci-nents are important in terms of plant risk) and (2) spe-cific revisions to testing schedules and methods for the sion making.

chosen components. Plant systems and functions that l

A Four Element Approach to Risk Informed rely on the affected components should be identified.

Decisionmaking for Inservice Testing Programs Regulatory Position 1 gives a more detailed description f Element 1.

Regulatory Guide 1.17-lef. 3) describes a four-ele me nt process for developiog risk-informed re gulato-Element 2: Perform Engineering Analysis ry changes. The process is highly iterative.Thus, the fi-In this element, both traditional engineering and nal description of the proposed change to the IST PRA methods are used to help define the scope of the progra n as defined in Element 1 depends on both the changes to the IST program and to evaluate the impact analysis performed in Element 2 and the definition of f the changes on the overall plant risk. Areas that are to the implementation of the IST program performed in be evaluated include the expected effect of the proposed Element 3.The Regulatory Position of this guide pro-T pmgram on the design basis and severe acci-vides guidance on each element.

dents, defense-in-depth attributes, and safety margins.

While IST is, by its nature, a monitoring program, In this evaluation, the results o' -aditional engineering it should be noted that the monitoring referred to in Ele-and PRA methods are to be considered together in an ment 3 is associated with making sure that the assump-integrated decision process that will be carried over into 1.175 - 4

l the implementation phase described below in Element NRC according to SRP Chapter 19 and Section 3.9.7

3. PRA results should be used to provide information (Refs. 6 and 7). Guidanu on documentation require-p for the categorization of components into groupings of ments for RI-IST programs is given in Regulatory Posi-low safety-significant components (LSSC) and high tion 4 of this regulatory guide.

safety-significant components (llSSC). Components m the LSSC group would then be candidates for less in carrying out this process, the licensee will make a number of decisions based on the best available infor-rigorous testing when compared with those m the mation. Some of this information will be derived from IISSC group. When the revised IST plan has been de-veloped, the plant-specific PRA should be used to eval-traditional engineering practice and some will be pro-babilistic in nature resulting from PRA studies. It is the uate the effect of the planned program changes on the overall plant risk as measured by core damage fre-licensee's responsibility to ensure that its RI IST pro-quency (CDF) and contamment large early release fre-gram is developed using a well-reasoned and integrated quency (LERF).

decision process that considers both forms of input in-formation (traditional engineering and probabilistic)in l

During the integration of all the available informa-a complementary manner.This important decisionma-tion,it is expected that many issues will need to be re.

king process may at times require the participation of solved through the use of a well-reasoned judgment special combinations of licensee expertise (licensee process, often involving a combination of different en-staft), depending on the technical and other issues in-gineering skills. This activity has typically been re-volved, and may at times also need outside consultants.

ferred to in industry documents as being performed by Industry documents have generally referred to the use an" expert panel." As discussed further at the end of this of an expert panel for such decisionmaking.The appen-section and in the appendix, this important process is dix to this guide discusses a number of IST-specific is-the licensee's responsibility and may be accomplished sues such as might arise in expert panel deliberations.

by means other than a formal panel. In any case, the key C. REGULATORY POSITION safety principles discussed in this guide must be ad-dressed and shown to be satisfied regardless of the ap-1.

ELEMENT 1: DEFINE PROPOSED proach used for RI-IST program decismnmakmg.

CHANGES TO INSERVICE TESTING Additional application-specific details concerning PROGRAM RI IST programs and Element 2 are contained in Regu-In this first element of the process, the proposed latory Positition 2 of this guide-changes to the IST program are defined. This involves E*

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• P""* "I" (" O' Element 3: Define Implementation and valves) will be mvolved and how their tes mg would be Monitorm.g Pmgram changed. Alsoincludedin thiselementisidentification In this element, the implementation plan for the of supporting information and a proposed plan for the IST pmgram is developed. This involves determining licensee's interactmns with the NRC throughout the both the methods to be used and the frequency of test-implementation of the RI-IST.

ing. The frequency and method of testing for each com-ponent is commensurate with the component's safety 1.1 Description of Proposed Changes significance. To the extent practicable, the testing A full description of the proposed changes in the methods should address the relevant failure mecha-IST program is prepared. This description would in-nisms that could significantly affect component reli-clude:

ability. In addition, a monitoring and corrective action program is established to ensure that the assumptions (1) Identificatmn of the aspects of. he plant,s design, t

upon which the testing strategy has been based contin-per ti ns, and other activities that require NRL appmval t wouM be changd h h proposed ue to be valid, and that no unexpected degradation in T pmgram. His wm pmee a basis fmm performance of the HSSCs and LSSCs occurs as a re-which the staff can evaluate the proposed changes.

sult of the change to the IST program. Specif.ic guid-ance for Element 3 is given in Regulatory Position 3.

(2) Identification of the specific revisions to existing testing schedules and methods that would result Element 4: Submit Proposed Change from implementation of the proposed program.

The finat element involves preparing the documen-(3) Identification of the components in the plant that tation to be included in the submittal and the documen-are directly and indirectly involved with the pro-l tation to be maintained by the licensee for later refer-posed testing changes. Any components that are j

ence, if needed. The submittal will be reviewed by the not presently covered in the plant's IST program l

1.175 - 5 l

1 l

but are determined to be important to safety (e.g.,

staff (i.e., as defined in the approved RI IST program through PRA insights) should also be identified.

description). Prior to implementation, a process or pro-In addition, the particular systems that are affected cedures should be in place to ensure that any such by the proposed changes should be identified changes to the previously approved RI-IST program since this information is an aid in planning the meet the acceptance guidelines of this section.

1 supporting engineering analyses.

The cumulative impact of all RI-IST program (4) Identification of the information that will be used changes (initial approval plus later changes) should in support of the changes. This willinclude perfor-comply with the acceptance guidelines given in Regu-mance data, traditional engineering analyses, and latory Position 2.3.3 below.

PRA information.

Examples of changes to RI-IST programs that (5) A brief statement describing the way how the pro-would require NRC's review and approval include, but posed changes meet the objectives of the Commis-are not limited to, the following:

sion's PRA Policy Statement (Ref.1).

Changes to the RI-IST program that involve pro-Ic2 Inservice Testing Pavgram Scope grammatic changes (e.g., changes in the accep-tance guidelines used for the licensee's integrated IST requirements for certain safety-related pumps decisionmaking process),

and valves are specified in 10 CFR 50.55a. These com-Component test method changes that involve devi-ponents are to be tested according to the requirements of Section XI of the American Society of Mechanical ation from the NRC-endorsed Code requirements, Engineers (ASME) Boiler and Pressure Vessel Code NRC-endorsed Code Case, or published NRC (the Code)(Ref. 9)or the applicable ASME Operations guidance.

and Maintenance (O&M) Code (Ref.13).

Examples of changes to RI-IST programs that For acceptance guidelines, the licensee's RI-IST would not require NRC's review and approvalinclude, program would include all components in the current but are not limited to, the following:

Code-prescribed IST program. In addition, the pro-gram should include those non-Code components that Changes to component groupings, test intervals, and test methods that do not involve a change to the the licensee's integrated decisionmaking process cate, overall RI-IST approach that was reviewed and ap-gorized as HSSC.

proved by the NRC, 13 RI IST Program Changes After Initial Component test method changes that involve the Approval implementation of an NRC-endorsed ASME Code This section provides guidance on reporting of pro-or an NRC-endorsed Code Case, gram activities. The NRC will formally review the Recategorization of components because of expe-changes proposed to RI-IST programs that have al-rience, PRA insights, or design changes, but not ready received NRC approval.

programmatic changes when the process used to The licensee should implement a process for deter-recategorize the components is consistent with the mining when proposed RI-IST program changes re-RI-IST process and results that were reviewed and quire formal NRC review and approval. Changes made approved by the NRC.

to the NRC-approved RI-IST program that could affect the process and results that were reviewed and ap-2.

ELEMENT 2: PERFORM ENGINEERING proved by the NRC staff should be evaluated to ensure ANALYSIS that the basis for the NRC staff's prior approval has not As part of defining the proposed change to the li-been compromited. All changes should be evaluated censee's IST program, the licensee should conduct an against the change mechanisms described in the regula-engineering evaluation of the proposed change using a tions (e.g.,10 CFR 50.55a,10 CFR 50.59) to determine combination of traditional engineering methods and whether NRC review and approval is required prior t PRA. The major objective of this evaluation is to con-implementation. If there is a question regarding this is-firm that the proposed program change will not com-sue, the ticensee should seek NRC review and approval promise defense in depth and other key safety prin-prior to implementation.

ciples describeo in this guide. Regulatory Guide 1.174 For accepance guidelines, licensees can change (Ref. 3) provides general guidance for the performance their RI IST p ograms consistent with the process and of this evaluation, to be supplemented by the RI-IST-results that we e reviewed and approved by the NRC specific guidance in this guide.

1.175 - 6

2.1 Licensing Considerations For acceptance guidelines, the licensee should re-view applicable documents to identify proposed 2.1.1 Evaluating the Proposed Changes changes to the IST program that would alter the design, On a component-specific basis, t he lice nsee should perations, and other activities of the phnt. On a com-determine whether there are instances in which the pro-p nent-specific basis, the licensee should (1) identify l

posed IST program change would affect the design, op-instances in which the proposed RI-IST program l

erations, and other activities at the plant, and the li-change would affect the design, operations, and other censee should document the basis for the acceptability activities of the plant,(2) identify the source and nature of the proposed change by addressing the key prin-f the requirements (or commitments), and (3) docu-ciples,in evaluating proposed changes to the plant, the ment the basis for the acceptability of the proposed re-licensee should consider other licensing basis docu.

9.uirement changes, e.g., by addressing the key prin-l ments (e.g., technical specifications, Final Safety Anal.

CIPes-l ysis Report (FS AR), responses to NRC generic letters)

The licensee must comply with 10 CFR 50.59, in addition to the IST program documentation.

50.90, and 50.109 as applicable. The staff recognizes The principal focus should be on the use of PRA s

not related to regulatory requirements that can be findings and n. k insights m support of proposed changed by licensees via processes other than described changes to a plant's design, operation, and other activi-in NRC regulations (e.g., consistent with Reference ties that require NRC approval. Such changes include 34)'

(but are not limited to) license amendments under 10 CFR 50.90, requests for use of alternatives under 2.1.2 Relief Requests and Technical Specification 10 CFR 50.55a, and exemptions under 10 CFR Part 12.

Changes However,the reviewer should note that there are certain The licensee should have included in the RI IST docketed commitments that are not related to regula-program submittal the necessary exemption requests, tory requirements (e.g., con:mitments made by the h-technical specification amendment requests, and relief censee in response to NRC Generic Letter 8910 or requests necessary to implement their RI-IST program.

,a 96-05) that may be changed by licensees via processes other than as described in NRC regulations (e.g., con-Individual component relief requests are not re-sistent with Reference 14).

quired for adjusting the test interval ofindividual com-ponents that are categorized as having low safety sig-A broad review of the plant's design, operations, nificance (because the licensee's implementation plans and other activities may be necessary because proposed for extending specific component test intervals should IST program changes could affect requirements or have been reviewed and approved by the NRC staff as commitments that are not explicitly stated in the licens-part of the licensee's RI-IST program submittal). Simi-ee's FSAR or IST program documentation. Further-larly,if the proposed alternative includes improved test more, staff approval of the design, operation, and main-strategies to enhance the test effectiveness of compo.

tenance of components at the facility have likely been nents, additional relief to implement these improved granted in terms other than probability, consequences, test strategies is not required.

or margin of safety (i.e., the 10 CFR 50.59 criteria).

I. r cceptance guidelines, the following are to be l

Therefore,it may also be appropriate to evaluate pro-posed IST program changes against other criteria (e.g.,

approved by the NRC before implementing the RI-IST criteria used in either the licensing process or to deter.

Program:

mine the acceptability of component design, operation A relief request for any component, or group of l

and maintenance).

components, that is not tested in accordance with the licensee's ASME Code of record or NRC-The Director of the Office of Nuclear Reactor Reg" approved ASME code case.

ulation is allowed by 10 CFR 50.55a to authorize alter-A technical specification amendment request for natives to the specific requirements of this regulation I

provided that the proposed alternative will ensure an any component, oi group of components, if there acceptable level of quality and safety. Thus, alterna-are changes from technical specification require-tives to the acceptable RI-IST approaches presented in ments.

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this guide may be proposed by licensees so long as sup.

2.2 Traditional Engineering Evaluation

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porting information is provided that demonstrates that the key principles discussed in Chapter 2 of this guide This part of the evaluation is based on traditional are maintained.

engineering methods (not probabilistic). Areas to be 1.175 - 7

evaluated from this viewpoint include the potential ef-ing from the RI IST program will maintain a balance feet of the proposed RI-IST program on defense-in-between prevention of core damage, prevention of con-depth attributes and safety margins. In addition, de-tainment failure, and consequence mitigation. Redun-fense in depth and safety margin should also be dancy, diversity, and independence of safety systems evaluated, as feasible, using risk techniques (PRA).

should be considered after the initial choice is made in the categorization of components to ensure that these 2,2.1 Defense in Depth Evaluation qualities are not degraded by the categorization. Inde-Because ofits importance, both historically during pendence of barriers and defense against common the evolution of reactor safety practice and for the con.

cause failures should also be considered in the review tinuation of public health and safety, the concept of de-of the categorization. The improved understanding of fense in depth has been included in Regulatory Guide the relative importance of plant components to risk re-1.174 (Ref. 3) as one of the five key principles. In refer-sulting from the development of the RI-IST program ring to a proposed risk-informed program change, Sec-should promote an improved overall understanding of tion 2 of Regulatory Guide 1.174 states that the pro-how the components in the IST program contribute to a posed change should be consistent with the plant's de fense in depth, and this should be discussed in defeuse-in depth philosophy. Furthermore, as stated in the application.

Section 2.2.1.1, 2.2.2 Safety Margin Evaluation Consiste ncy wit h the de fe nse-i n-dept h philos-The maintenance of safety margins is also a very ophy is maintained if:

important part of ensuring continued reactor safety and isincluded as one of the key safety principlet m Section A reasonable balance is preserved among 2 of Regulatory Guide 1.174 (Ref. 3). This principle prevention of core damage, prevention of states that the proposed change maintains sufficient containment failure, and consequence miti-s fety margins.

gation.

In addition, in Section 2.2.1.2, it is stated that with Over-reliance on programmatic activities sufficient salety margins:

to compensate for weaknesses in plant de-Codes and standards or alternatives ap-sign is avoided.

proved for use by the NRC are met.

System redundancy, m. dependence, and di-Safety analysis acceptance criteria in the li-versity are preserved commensurate with the expected frequency, consequences of censing basis (e.g., FS AR, supporting anal-challenges to the system, and uncertainties yses) are met, or proposed revisions pro-vide sufficient margin to account for (e.g., no risk outliers).

nalysis and data uncertainty.

Defenses against potential common cause failures are preserved, and the potential for It is possible that the categorization process will the introduction of new common cause fail-identify components that are currently not included in l

ure mechanisms is assessed.

the IST program, and their addition as llSSCs will Independence of barriers is not degraded.

clearly improve safety margin in terms of CDF and l

LERF. It is also important that the performance moni-Defenses against human errors are pre-toring program be capable of quickly identifying sig-

served, nificant degradation in performance so that,if neces-The intent of the General Design Criteria in sary, corrective measures can be implemented before 10 CFR Part 50, Appendix A is maintained, the margin to failure is significantly reduced. The im-proved understanding of the relative importance of These defense-in-depth objectives apply to all risk-plant components to risk resulting from the develop-informed applications, and for some of the issues in-ment of the RI-IST program should promote an im-volved (e.g., no over-reliance on programmatic activi-proved understanding of how the components in the l

ties and defense against human errors), it is fairly IST program contribute to a plant's margin of safety, straightforward to apply them to the RI-IST program and this should be discussed in the application.

evaluation. Some specific examples of how certain oth-2.3 Probabilistic Risk Assessment er of these objectives may be met for RI-IST applica.

tions are as follows.The use of the multiple risk metrics issues specific to the IST risk-informed process are of CDF and LERF and controlling their change result-discussed in this section. Regulatory Guide 1.174 (Ref.

l 1.175 - 8 l

t

3) contains much of the general guidance that is appli-test intervals or strategies. The PRA model should be l

l cable for this topic.

developed to the component level for the systems im-p in RI-IST, information obtained from a PRA portant to safety.

Q should be used in two ways: First, to provide input to if less than a full-scope PRA is used to support the the categorization of SSCs into llSSC and LSSC proposed RI IST program, supplemental information groupings; and second, to assess the impact of the pro.

(deterministic and qualitative) must be considered dur-posed change on CDF and LERF. Regulatory Position ing the integrated decisionmaking process.

2.3.1 discusses, in general terms, issues related to the Acceptance guidelines for the required PRA quali-quality, scope, and level of detail of a PRA that is used ty and scope are further defined in Regulatory Guide for IST applications. More specific considerations are 1.174.

given in Regulatory Positions 2.3.2, and 2.3.3, which address the use of PRA in categorization and in the as-2.3.2 Categorization of Components sessment of the impact on risk metrics respectively.

The categorization of components is important in the implementation of the RI-IST program since it is an 2.3.1 Scope, Level of Detail, and Quality of efficient and risk-informed way of providing insights in Probabilistic Risk Assessments for Inservice the areas in which safety margin can be relaxed without Testing Applications unacceptable safety consequences. Thus, categoriza-For the quantitative results of the PRA to play a tion of components,in addition to the traditional engi-major and direct role in decision making, there is a need neering evaluation described in Regulatory Position to ensure that they are derived from " quality" analyses, 2.2 and the calculation of change in overall plant risk and that the extent to which the results apply is well un.

described in Regulatory Position 2.3.3, will provide derstood. Section 2.2.3 of Regulatory Guide 1.174 Significant input to the determination of whether the (Ref. 3) addresses in general terms the issues related to IST program is acceptable or not.

scope, level of detail, and quality of the PRA applied to The determination of safety significance of com-risk-informed applications.

ponents by the use of PRA-determined importance i

W While a full scope PRA that covers all modes of op, measures is important for several reasons.

)

eration and initiating events is preferred, a lesser scope When performed with a series of sensitivity evalu-V PRA can be used to provide useful risk information.

ations, it can identify potential risk outliers by flowever, it must then be supplemented by additional identifying IST components that could dominate considerations as discussed below.

risk for various plant configurations and operation-1 m des, PRA model assumptions, and data and For the PRA to be usefulin the development of a model uncertainties.

RI-IST program,it is necessary that the PRA model be Importance measure evaluations can provide a use-developed to the component level for the systems, in-ciuding non-safety systems, considered important for ful means to identify improvements to current IST prevention of core damage and release of radioactivity, practices during the risk-informed application pro-ass.

A PRA used in RI-IST should be performed cor-rectly and in a manner that is consistent with accepted System-or functional-level importance results can practices. The PRA should reflect the actual design, provide a high level verification of component-lev-construction, ope rating p ractices, and ope rating e x pe ri.

el results and can provide insights into the potential ence of the plant. The quality required of the PRA is risk significance of IST components that are not commensurate with the role it plays in the determina-modeled in the PRA.

l tion of test intervals or test methods and with the role General guidelines for risk categorization of com-the integrated decisionmaking panel plays in compen-ponents using importance measures and other informa-sating for limitations in PRA quality. Regulatory Guide tion are provided in Regulatory Guide 1.174 (Ref. 3).

1.174 and SRP Chapter 19 (Refs. 3 and 6) further dis-These general guidelines address acceptable methods cuss the requirements of PRA quality.

for carrying out categorization and some of the limita-To be acceptable for application to RI IST, PRA tions of this process. Guidelines that are specific to the models must reflect the as-built, as-operated plant, and IST application are given in this section. As used here, G

they must have been performed in a manner that is con.

risk categorization refers to the process for grouping sistent with accepted practices.The quality of the PRA IST components into LSSC and HSSC categories.

has to be shown to be adequate, commensurate with the -

Components are initially categorized into HSSC role the PRA results play in justifying changes to the and LSSC groupings based on threshold valuee %r the l

1.175 - 9

importance measures. Depending on whether the PIU In classifying a component not modeled in the is performed using the fault tree linking or event tree PRA as LSSC, the expert panel should have determined linking approach,importance measures can most easily that:

be provided at the component or train level. In eithet The component does not perform a safety case, the importance measures are applicable to th a function, or does not perform a support items taken one at a time, and therefore, as discussed il function to a safety function, or does not Regulatory Guide 1.174, while a licensee is free ta complement a safety function.

choose the threshold values ofimportance measures,it The component does not support operator will be necessary to demonstrate that the integrated im-etions credited in the PRA for either proce-pact of the change is such that Principle 4 is met. One dur 1 r rec very acti ns.

acceptable approach is discussed in the next section.

The failure of the component will not result in the eventual occurrence of a PRA initiat-PRA systematically takes credit for non-Code components as providing support, acting as alterna.

ing event, The component is not a part of a system that tives, and acting as backups to those components that are within the current Code. Accordingly, to ensure that acts as a barrier to fission product release the proposed RI-IST program will provide an accept-during severe accidents.

able level of quality and safety, these additional risk-The failure of the component will not result important components should be included in licensees' in unintentional releases of radioactive ma-RI-IST proposals. Specifically, the licensee's RI-IST terial even in the absence of severe accident program should include those ASME Code Class 1,2, conditions.

and 3 and non-Code components that the licensee's in-For acceptance guidelines, when using risk impol-tegrated decisionmaking process categorized as HSSC tance meames to identify components that are low risk and thus determined these components to be appropri-contributors, the potential limitations of these mea-ate additional candidates for the RI-IST program.

sures have to be addressed. Therefore, information to be provided to the licensee's integrated decisionm aking Although PRAs model many of the SSCs involved pr cess @g., expert panel) must include evaluations in the performance of plant safety functions, other that demonstrate the sensitivity of the risk importance SSCs are not modeled for various reasons. However, results to the important PRA modeling techniques, as-this should not imply that unmodeled components are sumptions, and data. Issues that the licensee should not important in terms of contributions to plant risk.

c nsider and address when determining low risk con-For example, some components are not modeled be-tributors include truncation limit used, different risk cause certain initiating events may not be modeled metrics (i.e., CDF and LERF), different component (e.g., low power and shutdown events, or some external failur m des, different maintenance states and plant events);in other cases, components may not be directly

' "II "'.ations, multiple component considerations, E

modeled because they are grouped together with events defense in depth, and analysis of uncertainties (includ-that are modeled (e.g., initiating events, operator recov-ing sensitivity studies to component data uncertainties, ery events, or within other system or function bound-common-cause failures, and recovery actions).

aries); and in some cases, components are screened out from the analysis because of their assumed inherent While the categorization process can be used to reliability; or failure modes are screened out because of highlight areas in which testing strategy can be im-their insignificant contribution to risk (e.g., spurious proved and areas in w hich sufficient safety margins ex-closure of a valve). When feasible, adding missing ist to the point that testing strategy can be relaxed,it is components or missing initiators or plant operating the determination of the change in risk from the overall states to the PRA should be considered by the licensee.

changes in the IST program that is ofconcern in demon-When this is not feasible, information based on tradi, strating that Principle 4 has been met.Therefore, no ge-tional engineering analyses andjudgment is used to de.

nerically applicable acceptance guidelines for the termine whether a component should be treated as an threshold values ofimportance measures used to cate-LSSC or HSSC. One approach to combining these dif.

gorize components as HSSC or LSSC are given here.

ferent pieces ofinformation is to use what has been re.

Instead, the licensee should demonstrate that the over-ferred to as an expert panel. Appendices B and C of allimpact of the change on plant risk is small as dis-Standard Review Plan Chapter 19 (Ref. 6) contain staff cussed in Regulatory Position 2.3.3.

expectations on the use of expert panels in integrated As part of the categorization process, licensees decisionmaking and SSC categorization respectively.

must also address the initiating events and plant operat-1.175 - 10

ing modes missing from the PRA evaluation. The li-operate when demanded, even though for some pur-censee can do this either by providing qualitative argu-poses it would have been considered " good" before be-n ments that the proposed change to the IST program ing subjected to the stress of the demand itself. This

)

does not result in an increase on risk, or by demonstrat-would have the effect of adding a constant to the test in-V ing that the components significant to risk in these mis-terval-dependent contribution to the component un-sing contributors are maintained as llSSC.

availability on demand. The assumption that the total unavailability scales linearly with the test interval (i.e.,

2.3.3 Use of a PRA To Evaluate the Risk Increase doubles when test interval doubles)is conservative in from Changes in the IST Program the sense that it scales the test-interval-independent One of the important uses of the PRA is to evaluate contribution along with the test-interval-dependent the impact of the IST change with respect to the accep-contribution, and in that respect tends to overstate the tance guidelines on changes in CDF a:.d LERF as dis-effect of test interval extension. This approximation is cussed in Section 2.2.2 of Regulatory Guide 1.174 therefore considered acceptable; however, it should be (Ref. 3). In addition, the PRA can provide a baseline noted that guidance aimed at improving the capability j

risk profile of the plant, and the extent of analysis of the of tests to identify loss of performance margin is aimed j

baseline CDF and LERF depends on the proposed partly at reducing the " demand" cont ribution as wd !so change in CDF and LERF. As discussed in Regulatory that improved modeling in this area would appear to Guide 1.174,if the PRA is not full scope, the impact of have the potential to support further improvements m the change must be considered by supplementing the allocation of safety resources.

PRA evaluation by qualitative arguments or by bound-This model essentially assumes that failures are ing analyses.

random occurrences and that the frequency of these oc-2.3.3.1 Modeling the Impact of Changes in the currences does not increase as the test interval is in-IST Program. In order for the PRA to support the deci-creased. However, as test intervals are extended, there sion appropriately, there should be a good functional is some concern that the failure rate, A, may increase.

mapping between the components associated with IST This failure rate, generally assumed constant,is based and the PRA basic event probability quantification.

on data from current IST test intervals and therefore

)

Part of the basis for the acceptability of the RI-IST pro-does not include effects that may arise from extended

,/

gram is a quantitative demonstration by use of a PRA test intervals. it is possible that insidious effects such as that established risk measures are not significantly in-corrosion or erosion, intrusion of foreign materialinto creased by the proposed changes to the IST for selected working parts, adverse environmental exposure, or components. To establish this demonstration, the PRA breakdown oflubrication,which have not been encoun-includes models that appropriately account for the tered with the current shorter test intervals, could sig-change in reliability of the components as a function of nificantly degrade the component if test intervals be-the IST program changes. In general, this will include come excessively long. Therefore, unless it can be not only changes to the test interval but also the effects demonstrated that either degradation is not expected to of an enhanced testing method. Enhanced testing might be significant or that the test would identify degrada-be shown to improve or maintain component availabil-tion before failures are likely to occur, use of the ity, even if the interval is extended. That is, a better test constant failure rate model could be nonconservative.

might compensate for a longer interval between tests.

Licensees who apply for substantiat increases in test in.

One way to address this uncertainty is to use the terval arc expected to address this area,i.e., as appropri.

PRA insights to help design an appropriate imple-ate, consider improvements in testing that would com.

m"ntation and monitoring program, for example, to ap-pensate for the increased intervals under consideration.

pi cach the interval increase in a stepwise fashion rather th in going to the theoretically allowable maximum in a One rnodel for the relationship between the com-sinw step, or to stagger the testing of redundant com-ponent unavailability on demand and the test interval is ponents (test differera %s on alternating schedules) given in NUREG/CR-6141 (Ref.16), which assumes a so that the population ca components is being sampled constant rate (A) of transition to the failed state. Refer-relatively frequently, even though individual members ence 16 also describes how to account for various test of the population are not. By using such approaches, the strategies.

ex stence of the above effects can be detected and com-In addition to transitions to a failed state that occur pensatory measures taken to correct the testing of the 9

between component demands or tests, there is also a remaining population members. Ilowever, it is impor-demand related contribution to unavailability, corre-tant that the monitoring includes enough tests to be spending to the probability that a component will fail to relevant, and that the tests are capable of detecting the 1.175 - 11

time-related degradation (performance monitoring is be consistent with the guidelines provided in Section discussed in Regulatory Position 3.3).

2.2.4 of Regulatory Guide 1.174. In comparing the cal-culated risk to the guidelines, the licensee should ad-A check should also be performed to determm.e dress the model and completeness uncertainty as dis-whether non-IST manipulation has been credited either cussed in Regulatory Guide 1.174 (Ref. 3). In addition, in IST basic events or in compensatmg-component ba-the licensee should address parameter uncertainty ei-sie events. If a component is stroked or challenged be-ther by propagating the uncertainty during sequence tween instances ofIST, and if these activities are capa-quantification or by demonstrating that the " state-of-ble of revealing component failure, the effective fault knowledge correlation" effect is not significant, espe-exposure time can be less than the RI-IST interval. It cially in cutsets in which the RI-IST changes affect can be appropriate to take credit for this shortemng of multiple components that are similar.

fault e.xposure time in the PRA quantification, pro-vided that there is assurance that the important failure in evaluating the change in plant risk from pro-modes are identified by the stroking or the system chal.

posed changes in the IST program, the licensee should lenges. This is not always trivial: If a functional success perform the following.

can be achieved by any one of n components in parallel, Evaluate the risk significance of extending the test so that the function succeeds even if n-1 of the compo-interval on affected components. This requires that nents fail, then merely monitoring successful function-the licensee address the change in component al response does not show whether all components are availability as a function of test interval.The analy-operable unless verification of each component's state sis should include either a quantitative considera-is undertaken. In addition, some instances of revealing tion of the degradation of the component failure a component fault through challenge have adverse con-rate as a function of time, supported by appropriate sequences, including functional failure, and if credit is data and analysis, or arguments that support the taken for shortening fault exposure time through func-conclusion that no significant degradation will oc-tional challenges, it is necessary to account for this cur.

downside in the quantification of accident frequency.

Consider the effects of enhanced testing to the ex-2.3.3.2 Evaluating the Change in CDF and tent needed to substantiate the change.

LERF. Once the impact on the individual basic event probabilities has been determined, the change in CDF Other issues that should be addressed in the quanti-and LERF can be evaluated.There are some issues that fication of the change in risk include the following.

must be carefully considered, which become more im-The impact of the IST change on the frequency of portant the larger the change in basic event probabili-event initiators (those already included in the PRA ties. When using a fault tree linking approach to PRA,it and those screened out because of low frequency) is preferable that the model be re-solved rather than should be determined. For applications in RI-IST, simply requantifying the CDF and LERF cutset solu-potentially significant initiators include valve fail-tions. In addition,it is important to pay attention to the ure that could lead to interfacing system loss-of-parametric uncertainty analysis, especially if the coolant accidents (LOCAs) or to other sequences change is domiriated by cutsets that have multiple that fail the containment iso 3ation function.

LSSCs. The " state of knowledge" correlation effect The effect of common cause failures (CCFs)

+

(Ref.16) could be significant if there are a significant should be addressed either by the use of sensitivity number ofcutsets with similar SSCs contributing to the change in risk. Regulatory Guide 1.174 (Ref. 3) dis-studies or by the use of qualitative assessments that show that the CCF contribution would not become cusses the parametite uncertainty analysis in more g;;'

sigmficant under the prc, posed IST program (e.g.,

by use of phased implementation, staggered test-In addition, model and completeness uncertainties ing, and monitoring for common cause effects).

should be addressed as discussed in Regulatory Guide Justification ofIST relaxations should not be based 1.174. In particular, initiating events and modes of on credit for post-accident recovery of failed com-plant operations whose risk impact are not included in ponents (repair or ad hoc manual actions, such as the PRA need additional analyses or justification that the proposed changes do not sigmficantly increase the manually forcing stuck valves to open). Ilowever, risk from those unmodeled contributors.

credit may be taken for proceduralized imple-mentation of alternative success strategies. For 2.3.3.3 Acceptance Guidelines. The change in each human ac' ion that compensates for a basic risk from proposed changes to the IST program should event probability increasing as a result of IST re-1.175 - 12

laxation, there should be a licensee commitment to safety principles. Because of the importance of these ensure performance of the function at the level expectations, they will be repeated here, n

credited in the quantification. Excessively low hu-All safety impacts of the proposed change man failure probabilities (less than 10-3) cannot be are evaluated in an integrated manner as accepted unless there is adequate justification and part of an overall risk management ap-

{

there are adequate traimng programs, personnel proach in which the licensee is using risk l

practices, plant policies, etc., to ensure continued analysis to improve operational and engi-licensee performance at that level.

neering decisions broadly by identifying and taking advantage of opportunities for l

The failure rates and probabilities used for compo-reducing risk, and not just to eliminate re-l nents affected by the proposed change in IST quirements the ticensee sees as undesirable.

l should appropriately consider both plant-specific For those cases when risk increases are pro-l and generic data. The licensee should determine posed, the benefits should be described and whether individual components affected by the should be commensurate with the proposed l

change are performing more poorly than the aver-risk increases. The approach used to iden-l age associated with their class; the licensee should tify changes in requirements should be used avoid relaxing IST for those components to the to identify areas where requirements should point that the unavailability of the poor performers be increased,1 as well as where they could would be appreciably worse than that assumed in be reduced.

l the risk analysis. In addition, components that have experienced repeated failures should be reviewed The scope and quality of the engineering analyses (including traditional and proba-to see whether the testing scheme (interval and bilistic analyses) conducted to justify the methods) would be considered adequate to support proposed licensing basis change should be the performance credited to them m the risk appropriate for the nature and scope of the analysis.

change, should be based on the as-built and m

as-operated and maintained plant, and The evaluation should be performed so that the should reflect operating experience at the i

a x_)

truncation of LSSCs is considered. It is preferred plant.

that solutions be obtained from a re-solution of the The plant-specific PRA supporting li-model, rather than a requantification of CDF and LERF eutsets.

censee pmposab has been subjected to quality controls such as an independent peer review or certification.2

.The cumulative impact of all RI-IST program Appropriate consideration of uncertainty is changes (initial approval plus later changes) should comply with the acceptance guidelines given in analyses and interpretation of find-given in this section.

ing.s, including using a program of monitor-2.4 Integrated Decisionmaking This section discusses the integration of all the 1

technical considerations involved in reviewing submit.

The NRC staffis aware of but does not endorse guide-Imes that have been developed (e g., by NEl/NU-tais from licensees proposing to implement RI-IST pro-AlARC) to assist in identifying potentially beneficial grams. General guidance for risk-informed applica.

changes to requirements.

2As discussed in Section 2.2.3.3 of Regulatory Guide tions is given Regulatory Guide 1.174 (Ref. 3) and in 3MR'E 3)'"id'scuni n f PRA quainy, such a the new SRP sections, Chapter 19 (Ref. 6) for Seneral peer renew or certification is not a replacement for guidance, and Section 3.9.7 (Ref. 7) for IST programs.

NRC review. Certification is defined as a mechanism These documents discuss a set of reb'ulatorY findinbvs f r assurin8 that a PRA, and the process of deyeloping and mamtaming that PRA. meet a set of techmcal stan-that form the basis for the staff to prepare an acceptable dard sestablished by a dive rse grou pof pe rsonnel expe.

'iencedindevel Ping PRA models, performing PRAs, O

safety evaluation reEort (SER) for a licensee's risk-and performing quality reviews of PRAs. Such a pro-

{~s informed application. Specifically, Section 2 0f Regu-cess has bee n developed and integrated with a peer re-f latory Guide 1.174 identifies a set of " expectations" v'ew processby,for example.the BWR Owners Group and implemented for the purpose of enhancmg quahty that licensees should follow in addressing the key of PRAs at several BWR facilities.

1.175 - 13

ing, feedback, and corrective action to ad-are appropriately renected in the licensee's component dress significant uncertainties.

grouping. This should include components required to m intain adequate defense in depth as well as compo-The use of core damage frequency (CDF) nents that might be operated as a result of contingency and large early release frequency (LERF)3 pl ns developed to support the outage.

as bases for probabilistic risk assessment acceptance guidelines is an acceptable ap-L censees are also expected to review licensing ba-proach to addressing Principle 4. Use of the sis documentation to ensure that the traditional engi-l Commission's Safety Goal qualitative neering related factors mentioned above are adequately health objectives (OHOs)in lieu of LERF is modeled or otherwise addressed in the PRA analysis.

acceptable in principle and licensees may propose their use. However,in practice,im-When making final programmatic decisions, plementing such an approach would require choices must be made based on all the available infor-an extension to a Level 3 PRA, in which mation. There may be cases when information is in-case the methods and assumptions used in complete or when conflicts appear to exist between the the Level 3 analysis, and associated uncer-traditional engineering data and the PRA-generated in-tainties,would require additional attention.

formation. It is the responsibility of the licensee in such cases to ensure that well-reasoned judgment is used to Increases in estimated CDF and LERF re-

=

sulting from proposed changes will be lim-

[esolve the issues in the best manner possible,includ-ing due consideration to the safety of the plant. This ited to small increments. The cumulative pr cess f integrated decisionmaking has been dis-effect of such changes should be tracked cussed in vanous industry documents (Refs. 10 and considered in the decision process.

through 12) with reference to the use of an expert panel.

The acceptability of proposed changes The appendix to this regulatory guide includes some should be evaluated by the licensee in an in-detailed guidance on certain aspects of integrated deci-tegrated fashion that ensures that all prin-sionmaking specific to RI-IST programs. As discussed ciples are met.4 in the appendix,it is not intended that an administrative Data, methods, and assessment criteria body such as an expert panel must always be formed by the licensee to fulfill this function. Some general accep-used to support regulatory decisionmaking tance guidelines for thisimportant activity follow,with must be well documented and available for more specific details given in the appendix.

public review.

These expectations apply to both probabilistic and In summary, acceptability of the proposed change traditional engineering considerations, which are ad-should be determined by using an integrated decision-dressed in more detail in this chapter and in Regulatory making process that addresses three major areas: (1) an l

Guide 1.174 (Ref. 3).

evaluation of the proposed change in light of the plant's licensing basis, (2) an evaluation of the proposed l

Licensees are expected to review commitments re-chang;e relative to the key pnnciples and the acceptance lated to outage planning and control to verify that they entena, and (3) the proposed plans for implementation, 3tn this context, LERF is being used as a surrogate for performance momu ring, and corrective action. As the early fatahty quantitative health objective (QHO).

stated in the Commission's Policy Statement on the in-lt is defined as the frequency of those accidentsleading to significant, unmitigated releases from containment creased use of I'F A in regulatory matters (Ref.1), the i"

• ti.me frame prior to effective esacuation of the PRA information used to support the RI-IST program close-in population such that there is a potential for early health effects. Such accidents generally include should be as realistic as possible, with reduced unnec-unscrubbed releasesassociated withcarly containme nt essary conservatisms, yet include a consideration of failure at or shortly after vessel breach, containment bypass events. and loss of containme nt isolation. This tincertainties. These factors are very important when definition is consistent with accident analyses used in considering the cumulative plant risk and accounting the safety goal scree mng erite ria discussed in the Com-mission'a regulatory analysis guidelines. An N RCcon.

for possible risk increases as well as risk benefits. The tractor's report (Ref.15) describes a simple screening licensee should carefully document all of these kinds of appr ach for calculating LERE considerations in the RI-IST program description, in-done important element ofintegrated decisionmaking cluding those areas that have been quantified throuSh canbetheuseofan expertpanel. Suchapanelisnota necessary componeni of nsk. informed decisionmak.

the use of PRA, as well as qualitMve arguments for ing; but w hen it is used, the key principles and associat.

those areas that cannot readily be (.antified.

ed decision criteria presented in thm regulatory guide still apply and must be shown to have bee n met or to be irrelevant to the issue at hand.

The following are acceptance guidelines.

1.175 - 14

The licensee's proposed RI-ISTprogram should be duct the existing approved Code IST test at an extended e

supported by both a traditional engineering analy-interval.

sis and a PRA analysis.

An acceptable strategy for testing components

)

The licensee's RI-IST program submittal should be c tegorized IISSC and LSSC may be defined in NRC.

x" consistent with the acceptance guidelines con-approved ASME risk-informed Code Cases. Licensees tained throughout this regulatory guide, specifi-who choose to pursue RI-IST programs should consid-cally with the expectations listed in this section, or er adopting test strategies developed by ASME and en-the submittal shouldjustify why an alternative ap-dorsed by the NRC. Deviations from endorsed Code proach is acceptable.

Cases must be reviewed and approved by the NRC staff as part of the RI-IST program review.

If the licensee's proposed RI IST program is ac-In establishing the test strategy for components, ceptable based on both the deterministic and pro-the licensee should consider component design, service babilistic analyses, it may be concluded that the condition, and performance, as well as risk insights.

proposed RI-IST program provides "an acceptable The proposed test strategy should be supported by data level of quality and safety" [see 10 CFR that are appropriate for the component.The omission of 50.55a(a)(3)(i)].

either generic or plant-specific data should bejustified.

The proposed test interval should be significantly less 3.

ELEMENT 3: DEFINE IMPLEMENTATION than the expected time to failure assumed in the PRA of AND MONITORING PROGRAM the components in question (e.g., an order of magnitude Upon approval of an RI IST program, the licensee less).5In addition, the licensee should demonstrate that should have in place an im;)lementation schedule for adequate component capability (margin) exists, above testing all liSSCs and LSSCs identified in their pro.

that required during design basis conditions, such that gram. This schedule should include test strategies and component operating characteristics over time do not testing frequencies for flSSCs and LSSCs that are with.

result in reaching a point of insufficient margin before in the scope of the licensee's IST program and compo.

the next scheduled test activity.

nents identified as llSSCs that are not currently in the The IST interval should generally not be extended j

IST program, beyond once every 6 years or 3 refueling outages j

(whichever is longe r) without specific compelling doc-3.1 Inservice Testing Program Changes umented justification available on site for review. Ex-This section discusses the test strategy changes tensions beyond 6 years or 3 refueling outages (which-(i.e., component test frequency and methods changes) ever is longer) will be considered as component that licensees should make as part of a RI-ISTprogram.

Performance data at extended intervals is acquired.

This is not meant to restrict a licensee from fully imple.

For acceptance guidelines, the RI-IST program menting NRC-approved component Code Cases, should identify components for which the test strategy Components categorized IISSC that are not in the (i.e., frequency, methods or both) should be more fo-licensee's current IST program should (where practi-cused as well as components for which the test strategy cal) be tested in accordance with the NRC-approved might be relaxed. The information contained in, and de-ASME risk-informed Code Cases, including com-rived from, the PRA should be used to help construct pliance with all administrative requirements. When the testing strategy for components. To the extent prac-ASME Section XI or O&M Code testing is not practi-ticable, components with high safety significance cal, alternative test methods should be developed by the should be tested in ways that are effective at detecting licensee to ensure operational readiness and to detect their risk-important failure modes and causes (e.g.,

component degrsdation (i.e., degradation associated ability to detect failure, to detect conditions that are pre-with failure modes identified as being important in the cursors to failure, and predict end of service life). Com-licensee's PRA). As a minimum, a summary of these ponents categorized LSSC may be tested less rigor-components and their proposed testing should be inclu-ously than components categorized as llSSC (e.g., less ded in the RI-IST program.

l frequent or informative tests).

For components categorized as liSSCthat were the In some situations, an acceptable test strategy for subject of a previous NRC-approved relief request (or

]

components categorized IISSC may be to conduct the an NRC-authorized alternative test), the licensee 1

existing approved Code IST test at the Code prescribed D'j^*n$ejime)cstIs$o'uiNp*erf aIiNe~aNrfsiI P

frequency. In some situations, an acceptable test strat-egy for components categorized LSSC may be to con-etably smaHer than the expected time to failure.

l 1.175 - 15 l

l

should discuss the appropriateness of the relief in light referenced in the IST program and in the implementing of the safety significance of the component in their RI.

and test procedures to ensure that testing failures are re-evaluated for possible adjustment to the component's IST submittal, grouping and test strategy.

If practical,lST components (with the exception of certain check valves and relief valves) should, as a lt is acceptable to implement RI-IST programs on a minimum, be exercised or operated at least once every phased approach. Subsequent to the approval of a RI-refueling cycle. More frequent exercising should be IST program, implementation of interval extension for considered for components in any of the following cate-LSSC may begin at the discretion of the licensee and gories,if practical:

may take place on a component, train, or system-s level, llowever,it is not acceptable to immediately ad-Components with high n. k s.igmficance, just the test intervals of LSSC to the maximum pro-Components in adverse or harsh environmental posed test interval. Normally, test interval increases conditions, or will be done step-wise, with gradual extensions being permitted consistent with eumulative performance data Components with any abnormal characteristics g

e (operat,onal, des,gn, or maintenance conditions).

i o

i M ST g m i

i The testing strategy for each component (or group should be available at the plant site for inspection.

of components) in the licensee's RI IST program It should be noted that the test described in the cur-should be described in the RI IST program description.

rent ASME Code may not be particularly effective m The RI-IST program description should summarize all detecting the,mportant failure modes and causes of a i

testing to be performed on a group of components (e.g.,

e mp nent or group of components. A more effective MOV testing in response to NRC Generic Letter 96-05, test strategy may be to conduct an enhanced test at an Ref.18). The specific testing to be done on each com-extended test interval.

ponent (or group of components) should be delineated in the licensee's IST program plan and is subject to IISSCs that are not in the current IST program NRC inspection.

should be tested, where practical, in accordance with the ASME Code, including compliance with all admin-3.2 Program Implementation istrative requirements. When ASME Section XI or i

The applicable ASME Code generally requiresthat O&M testing is not practical, alternative test methods safety-related components within the program scope as should be developed by the licensee to ensure opera-l defined in the current ASME Code be tested on a quar-tional readiness and to detect component degradation terly frequency regardless of safety significance. The (i.e., degradation associated with failure modes identi-authorization of a risk-informed inservice testing pro-fied as being important in the licensee's PRA). As a gram will allow the extension of certain component minimum, a summary of these components and their testing intervals and modification of certain component proposed testing should be provided to the NRC as part testing methods based on the determination of individ-of this review and prior to implementation of the risk-ual component importance. The implementation of an informed IST program at the plant.

authorized program d involve scheduling test inter-An acceptable method to extend the test interval for vals based on the res..lts of probabilistic analysis and LSSC is to group like components and stagger their deterministic evaluation of each individual compone nt.

testing equally over the interval identified for a specific The RI IST program should distinguish between component based on the probabilistic analysis and de-high and low safety-significant components for testing terministic evaluation of each individual component.

intervals. Components that are being tested using spe-Initially, it would be desirable to test at least one com-cific ASME Codes, NRC-endorsed Code Cases for RI-ponent in each group every refueling outage. For exam-IST programs, or other applicable guidance should be ple, component grouping should consider valve actua-individually identified in the RI-IST program. The test tor type for power operated valves and pump driver intervals of the llSSC should be included in the RI-IST type, as applicable. With this method, generic age-program fo verificationof compliancewiththe ASME related failures could be identified while allowing im-Code requirements and applicable NRC-endorsed mediate implementation for some components. For ASME Code Cases. Any component test interval or component groups that are insufficient in size to test method that is not in conformance with the above one component every refueling outage, the imple-should have specific NRC approval. Plant corrective mentation of the interval should be accomplished in a action and feedback programs should be appropriately more gradual step wise manner. The selected test fre-1.175 - 16

quency for LSSC that are to be tested on a staggered ba-itoring when testing under design basis conditions is sis should be justified in the RI-IST program.

impracticable. In most cases, component-level moni-The following implementation activities are ac-toring will be expected.

s ceptable:

Two important aspects of performance monitoring are whether the test frequency is sufficient to provide For components that will be tested in accordance meaningful data and whether the testing methods, pro-with the current NRC-approved Code test frequen-cedures, and analysis are adequately developed to en-cy and method requirements, no specific imple-sur that performance degradation is detected.Compo-mentation schedule is required.The test frequency and method should be documented in thelicensee's nent faHure rates cannoj be aHowed to nse to unacceptable levels (e.g., sigmficantly higher than the P' 8'"**

failure rates used to support the change) before detec-For components that will employ NRC-endorsed tion and corrective action take place.

ASME Codes or Code Case methods,implementa-The N RC staff expects that licensees will integrate, tmn of the revised test strategies (i.e., interval ex-or at least coordinate, their monitoring for RI-IST pro-tension plan) should be documented in the licens-gram with existing programs for monito ing equipment ee's RI-IST program, performance and other operating experience on their For any alternative test strategies proposed by the sites and,when appropriate,throughout the industry. In licensee (i.e., for components within the scope of particular, monitoring that is performed as part of the the current ASME code), the licensee should have Maintenance Rule (10 CFR 50.65) implementation can l

specific NRC approval.

be used in the RI-IST program when the monitoring l

performed under the Maintenance Rule is sufficient for The licensee should increase the test interval for the SSCs in the RI-IST program. As stated in Regulato-components m a step-wise manner (i.e., equal or suc-ry Guide 1.174,if an application requires monitoring of cessively smaller steps, not to exceed one refueling SSCs not included in the Maintenance Rule, or in-cycle per step). If no sigmficant time-dependent fail-volves SSCs that need a greater resolution of monitor-ures occur,the mterval can be gradually extended until ing than the Maintenance Rule (e.g., component level

]

the component is tested at the maximum proposed ex-vs. train-or plant-level monitoring), it may be advanta-tended test interval. An acceptable approach is to group geous for a licensee to adjust the Maintenance Rule l

similar components and test them on a staggered basis.

monitoring program rather than to develop additional Guidance on grouping components is contained in monitoring programs for RI-IST purposes. Therefore, Position 2 of NRC Genenc Letter 89-04 (Ref.19) for it may be advantageous to adjust the Maintenance Rule check valves; Supplement 6 to NRC Genenc Letter performance criteria to meet the acceptance guidelines j

8910 (Ref. 20), and Section 3.5 of ASME Code Case below.

OMN-1 (Ref. 21) for motor-operated valves, or other documents endorsed by the NRC.

For acceptance guideline, monitoring programs s

should be proposed that are capable of adequately 3.3 Performance Monitoring tracking the performance of equipment that, when de-Performance monitoring in RI-IST programs re, graded, could alter the conclusions that were key to fers to the monitoring ofinservice test data for compo.

supporting the acceptance of the RI IST program.

nents within the scope of the RI-IST program (i.e.,in.

Monitoring programs should be structured such that cluding both HSSC and LSSC). The purpose of SSCs are monitored commensurate with their safety performance monitoring in a RI-IST program is two.

significance. This allows for a reduced level of moni-fold. First, performance monitoring should help con.

toring of components categorized as having low safety l

firm that no insidious failure mechanisms that are re.

significance provided the guidance below is still met.

l lated to the revised test strategies become important The licensee's performance monitoring process enough to alter the failure rates assumed in the justifica-should have the following attributes:

tion of program changes. Second, performance moni-Enough tests are included to provide meaningful l

toring should, to the extent practicable, ensure that ade-

data, quate component capability (i.e., margin) exists, above The test is devised such that menpient degradation that required during design-basis conditions, so that can reasonably be expected to be detected, and h

component operating characteristics over time do not The licensee trends appropriate parameters as re-Q result in reaching a point ofinsufficient margin before the next scheduled test activity. Regulatory Guide quired by the ASME Code or ASME Code Case 1.174 (Ref. 3) provides guidance on performance mon-and as necessary to provide reasonable assurance 1.175 - 17 i

that the component will remain operable over the determined for all components categorized as hav-test interval.

ing high safety significance, as well as for compo-nents categorized as having low safety signifi-Assurance must be established that degradation is c nee when the apparent cause of failure may not significant for components that are placd en an ex-e ntribute to common cause failure.

tended test interval, and that failure rate assumptions for these components are not compromised by test data.

(4) Assesses the applicability of the failure or noncon.

It must be clearly established that those test procedures forming condition to other components in the RI-and evaluation methods are implemented that reason-IST program (including any test sample expansion ably ensure that degradation will be detected and cor-that may be required for grouped components such rective action will be taken.

as relief valves).

3.4 Feedback and Corrective Action (5) Corrects other susceptible RI-IST components as necessary.

The licensee's corrective action program for this application should contain a performance-based feed-(6) Considers the effectiveness of the component's back mechanism to ensure that if a particular compo-test strategy in detecting the failure or nonconfor-nent's test strategy is adjusted in a way that is ineffec-ming condition. Adjust the test interval and/or test tive in detecting component degradation and failure, ethods, as appropriate, when the component (or particularly potential common cause failure mecha-gmup f components) experiences repeated or nisms, the RI-IST program weakness is promptly de-age-rei ted failuresornonconformmgconditions.

tected and corrected. Performance monitoring should The corrective action evaluations should periodi-be provided for systems, structures, and components cally be provided to the licensee's PRA group so that with feedback to the RI-IST program for appropriate any necessary model changes and re-grouping are done adjustments when needed.

as might be appropriate. The effect of the failures on If component failures or degradation occur at a ver 11 pl nt risk should be evaluated as well as a con-firm ti n that the corrective actions taken will restore higher rate than assumed in the basis for the RI-IST pro-gram, the following basic steps should be followed to the plant risk to an acceptable level, implement corrective action.

The RI-IST program documents should be revised The causes of the failures or degradation should be to document any RI-IST program changes resulting from corrective actions taken.

determined and corrective action implemented.

The component's test effectiveness should be re-e eansment evaluated, and the RI-IST program should be mo-dified accordingly.

RI IST programs should contain provisions whereby com1.onent performance data periodically The following are acceptance guidelines.

gets fed back into both the component categorization The licensee's corrective action program evaluates and component test strategy determination (i.e., test in-l RI-IST components that either fail to meet the test ac-terval and methods) process.These assessments should ceptance criteria or are otherwise determined to be in a also take into consideration corrective actions that have nonconforming condition (e.g., a failure or degraded been taken on past IST program components. (This pe-condition discovered during normal plant operation).

riodic reassessment should not be confused with the The evaluation:

120-month program updates required by 10 CFR 50.55a(f)(5)(i), whereby the licensee's IST program (1) Complies with Criterion XVI, " Corrective Ac-must comply with later versions of the ASME Code tion," of Appendix B to 10 CFR Part 50.

that have been endorsed by the NRC.)

(2) Promptly determines the impact of the failure or The assessment should:

nonconforming condition on system / train oper-ability and follows the appropriate technical spec-Review and revise as necessary the models and ification when component capability cannot be data used to categorize components to determine demonstrated.

whether component groupings have changed.

(3) Determines and corrects the apparent or root cause Reevaluate equipment performance to determine of the failure or nonconforming condition (e.g.,

whether the RI-IST progcam should be adjusted improve testing practices, repair or replace the (based on both plant-specific and generic informa-component). The root cause of failure should be tion).

1.175 - 18

i l

The licensee should have procedures in place to A description of the PRA used for the catego-identify the need for more emergent RI-IST program rization process and for the determination of

,m updates (e.g., following a major plant modification or risk impact,in terms of the process to ensure I(

following a significant equipment performance prob-quality and the scope of the PRA, and how lim-lem).

itations in quality, scope, and level of detail are Licensees may wish to coordinate these reviews e mpensated for in the integrated decision-making process (see Regulatory Position 2.3.1 with other related activities such as periodic PRA up-dates, industry operating experience programs, the above),

A description of how the impa:t of the change Maintenance Rule program, and other risk-informed program initiatives.

is modeled in the IST components (including a quantitative or qualitative treatment of compo-l The acceptance guideline is that the test strategy nent degradation) and a description the impact for RI-ISTcomponents should be periodically assessed of the change on plant risk in terms of CDF and i

to reflect changes in plant configuration, component LERF and how this impact compares with the performance, test results, and industry experience.

decision guidelines (see Regulatory Position

4. ELEMENT 4: DOCUMENTATION

)'

A description of how the key principles were The recommended content of an RI-IST submittal (and will continue to be) maintained (see Reg-is presented in this Regulatory Postion. The guidance ulatory Positions 2.2,2.3, and.2.4),

provided below is intended to help ensure the com-i pleteness of the information provided and should aid in A description of the integrated decisionmaking shortening the time needed for the review process. The Process used to help define the RI-IST pro-j licensee should refer to the appropriate section of this gram, including any decision criteria used (see regulatory guide to ascertain the level of detail of the Regulatory Position 2.4),

i l

documentation that should either be submitted to the A general implementation approach or plan NRC staff for review or retained onsite for inspection.

(see Regulatory Positions 3.1 and 3.2),

To the extent practical the applicable sections of the re-gulatory guide have been identified on each list of A description of the testing and monitoring documents.

pmp sed for each component group (see Reg-ulatory Position 3.2),

l 4.1 Documentation That Should Be in The A description of the RI-IST corrective action Licensee's RI IST Submittal plan (see Regulatory Position 3.4),

A description of the RI-IST program periodic A request to implement a RI-IST program as an au-thorized alternative to the current NRC-endorsed reassessment plan (see Regulatory Position 3.5 ASME Code pursuant to 10 CFR 50.55a(a)(3)(i).

above).

A description of the change associated with the A summary of any previously approved relief re-proposed RI IST program (see Regulatory Posi-quests for components categorized as IISSC along tion 1.1 above).

with any exemption requests, technical specifica-tion changes, and relief requests needed to imple-Identification of any changes to the plant's design, ment the proposed RI-IST Program (see Regula-operations, and other activities associated with the tory Position 2.1.2).

proposed RI-IST program and the basis for the ac-ceptability of these changes (see Regulatory Posi.

An assessment of the appropriateness of pre-tion 2.1.1).

vi usly approved relief requests.

A summary of key technical and administrative as-4.2 Documentation That Should Be Available i

pects of the overall RI IST program that includes:

Onsite For Inspection The overall IST Program Plan A description of the process used to identify candidates for reduced and enhanced IST re-Administrative procedures related to RI-IST 9

quirements, including a description of the cate-gorization of components using the PRA and Component or system design basis documentation Piping and instrunent diagrams for systems that the associated sensitivity studies (see Regula-I tory Position 2.3.2 above),

contain components in the RI IST program 1.175 - 19 i

Completed test procedures and any supplemental PRA and supporting documentation (see Regula-tory Position 2.3) test data related to RI-IST(see Regulatory Position 3.3)

Categorization results, including the RI-IST pro-cess summary sheet for each component or group Corrective action procedures (see Regulatory Posi-a of components (see Regulatory Position 2.3.2) tion 3.4)

Plant-specific performance data (e.g., machinery I ntegrated decisionmaking process procedures, ex-history) for components in the RI-IST program pert panel meeting minutes (if applicable) (see (see Regulatory Positions 2.3.3 and 3.1)

Regulatory Position 2.4)

A description of individual changes made to the Detailed implementation plans and schedules (see RI-IST program a fte r implementation (see Regula-l Regulatory Position 3.2) tory Position 1.3)

O O

1.175 - 20

l REFERENCES 9

1.

USNRC,"Use of Probabilistic Risk Assessment

10. American Society of Mechanical Engineers, Methods in Nuclear Regulatory Activities: Final

" Risk-Based Inservice Testing-Development of Policy Statement," Federal Register, Vol. 60, p Guidelines," Research Report (CRDT-Vol. 40-2, 42622, August 16,1995.

Volume 2),1996.3 2.

USNRC," Framework for Applying Probabilistic

11. Electric Power Research Institute,"PSA Applica-Risk Analysis in Reactor Regulation,"

tions Guide," EPRI TR-105396, August 1995.1 SECY-95-280, November 27,1995.1

12. Nuclear Energy Institute Draft (Revision B),"In-3.

USNRC,"An Approach for Using Probabilistic dustry Guidelines for Risk-Based Inservice Test-Risk Assessment in Risk-Informed Decisions on ing," March 19,1996.1 Plant-Specific Changes to the Licensing Basis,"

13. American Society of Mechanical Engineers Regulatory Guide 1.174, July 1998.2 (ASM E) Code for Operations and Maintenance of 4.

USNRC "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Graded Quality As-

14. Nuclear Energy Institute," Guidelines for Manag-surance," Regulatory Guide l.176, August 1998.2 ing NRC Commitments," Revision 2, Decem-ber 19,1995.1 5.

USNRC,"An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifica-

15. W.T. Pratt et al.,"An Approach for Estimating the tions," Regulatory Guide 1.177, August 1998.2 Frequencies of Various Containment Failure Modes and Bypass Events," Draft NUREG/

6.

USNRC, " Standard Review Plan for Risk-CR-6595, December 1997.2 Informed Decision Making," Standard Review

16. P.K. Samanta et al., "IIandbook of Methods for Plan, NUREG-0800, Chapter 19, July 1998.2 Risk-Based Analyses of Technical Specifica-7.

USNRC, " Standard Review Plan for Risk-tions," NUREG/CR-6141, December 1994.4 g

Informed Decision Making: Inservice Testing,"

17. G.E. Apostolakis and S. Kaplan," Pitfalls in Risk Standard Review Plan, NUREG-0800, Chapter Calculations," Reliability Engineering, Vol. 2, 3.9.7, August 1998.2 pages 135-145,1981.

8.

USNRC, ' Standard Review Plan for Risk-

18. USNRC," Periodic Verification of Design-Basis Informed Decision Making: Technical Specifica-Capability of Safety-Related Power-Operated tions," Standard Review Plan, NUREG-0800, Valves," Generic Letter 96-05, September 18, Chapter 16.1, August 1998.2 1996.1 9.

American Society of Mechanical Engineers

19. USNRC, " Guidance on Developing Acceptable (ASM E) Boiler and Pressure Vessel Code, Section Inservice Testing Programs," Generic Letter XI, ASME.3 89-04, April 3,1989.3

20. USNRC, " Safety-Related (1) Motor-Operated Valve Testing and Surveillance," Generic Letter ICopies are available for inspection or copying for a fee from the N RC 89-10, June 28,1989.1 Pubhc Document Room at el20 L Street NW, Washington, DC; the

21. American Society of Mechanical Engineers PDR's maihng address is Mail Stop LL-6, Washington, DC 20555; telephone (202)634 3273; fax (202)634-3343.

(ASME) Alternative Rules for Preservice andIn-esting M Certain Bedric hr Opm se 2Single copies of regulatory guides, both active and draft, and draft NUREG documents may be obtained free of charge by writing the ated Valve Assemblies in LWR Power Plants, Reproduction and Distribution Services Section, OCIO, USNRC, Code Case OMN-1, OM Code-1995; Subsection Washington, DC 20$55-0001, or by fax to (30 !)415-2289, or by email to GRWl(d NRC. GOV. Active guides may also be purchased from IbI,C 3 the National Technical!nformation Service on a standing order basis.

Details on this service may be obtained by writing NTIS,5285 Port 4

Royal Road, Suringfield, VA 22161. Copies of active and draft guides Copies arc available at eurrent i ates from thc U.S. Governme n t Print.

are available for inspection or copying for a fee from the NRC Public ing Office, P.O. Box 37082, Washington. DC 20402-9328 (tele pho,ne (202)512-2249);or from the National Technicallnformation Service D

(Document Room at 2120 L Street NW, Washington DC;the PDR's maih ng add re ss is M ail Stop LL-6, Washington, DC 20555; te le phone by wnting NTIS at 5235 Port Royal Road, Springfield, VA 22161.

202)634-3273; fax (202)634-3343.

Copies are available for inspection or copying for a fee from the N RC Public Document Room at 2120 L Street NW., Washington, DC; the 3

(

Copics may be obtained from ASME,345 Ent 47th Street,NewYork, PDR's mailmg address is Mail Stop LL 6, Washington. DC 20555; NY 10017.)

telephone (202)634-3273; fax (202)634-334 L 1.175 - 21

l

[

APPENDIX A DETAILED GUIDANCE FOR INTEGRATED DECISIONNIAKING A.1 Introduction (e.g., defense in depth, common cause, and the single failure criterion),which may be more constraining than The increased use of probabilistic risk assessment the nsk-based criteria in some cases. Consideration i

(PRA) in nuclear plant activities such as in risk-in-must be given to these issues and component perfor-formed inservice testing (IST) programs will require a m nce expenence before the IST requirements for the balanced use of the probabilistic information with the v n us mmponents are determined.

i more traditional engineering (sometimes referred to as

" deterministic") information. Some structured process IST experience should contribute an unde rstanding for considering both types of information and making of the important technical bases underlying the existing decisions will be needed that will allow improvements testing program before it is changed. The critical safety to be made in plant effectiveness while maintaining ad-aspects of these bases should not be violated inadver-equate safety levels in the plant. This will be particular-tently in changing over to a RI-IST, and important plant ly important during initial program implementation experience gained through the traditional ISTshould be and also for the subsequent early phases of the program.

considered during the change.

In some instances, the physical data from the PRA and The plant-specific PRA information should in-from the determimstic evaluations may be insufficient clude important perspectives with respect to the limita-to make a clearcut decision. At times, these two forms uns of PRA modeling and analysis of systems, some of information may even seem to conflict. In 4 which may not be explicitly addressed within the cases,it is the responsibility of the licensee to asse m m PRA analysis. An understanding should also be pro-the appropriate skilled utility staff (and in some cases vided as to how the proposed changes in pump and consultants) to consider all the available information in valve testing could affect PRA estimates of plant risk.

its various forms and to supplement this mformation with engineeringjudgment to determine the best course Plant safety experience should provide insights as-l of action. The participants involved in this important sociated with the traditional analyses (Chapter 15 of the j

role have generally been referred to in various industry plant Final Safety Analysis Report) and any effect that I

documents as an " expert panel." In this appendix, this proposed changes in testing might have on the tradi-function will be described as being an engineering eval.

tional perspective of overall plant safety.

nation without specifying how the evaluation is to be Plant operational input should supplement the in-performed administratively. It is not the intention of sights of plant safety with additional information re-j this guidance to indicate that a special administrative garding the operational importance of components un-l body needs to be formed within the utility to satisfy this der normal, abnormal, and emergency conditions.

l role. It is the function that is important and that must be There should also be input on operating history, system performed in some well-organized, repeatable, and interfaces, and industry operating experience to supple-scrutable manner by the licensee. This function is all-ment information from the IST.

pervasive in the implementation phase of such activi-Maintenance considerations should provide per-ties as inservice inspection (ISI) and IST, and accord-ingly, the licensee has the responsibility to see that this spectives on equipment operating history, work prac-function is done well.

tices, and the implementation of the maintenance rule.

Systems design considerations should include the A.2 Basic Categories ofInformation To Be potential effect of different design configurations (e.g.,

Considered piping, valves, and pumps) on planning for a risk-inf rmed IST,particularly if future plant modifications Risk-importance measures may be used together are mnten1 plated or if systems are temporarily taken with other available information to determine the rela-ut ofservice for maintenance or replacement or repair.

tive risk ranking (and 'ho categorization) of the com-ponece included in the evaluation. Results from all A.3 Specific Arras To Be Evaluated thes.. aourco are then reviewed prior to making final This section addresses some technical and adm.m-decisions about where to focus IST resources.

istrative issues that are currently believed to be particu-Although the risk ranking of components can be larly important for RI-IST applications. Additionalis-used primarily as the basis for prioritizing IST at a sues of a more general nature that may arise in expert plant, additional considerations need to be addressed panel deliberations are given in SRP Chapter 19.

1.175 - 22

It should be confirmed that proper attention has Attention should be given to the fact that compo-e been given to component classifications in systems nent performance can be degraded from the effects identified in emergency operating procedures (and of aging or harsh environments, and this issue will i

other systems) depended upon for operator recov-need to be addressed and documented.

()

ery actions, primary fission product barriers ex-The engineering evaluation should include the cluded from the PRA due to their inherent reliabil-choice of new test frequencies, the identification of sty (such as the RPV), passive items not modeled in compensatory measures for potentially important the PRA (such as piping, cable, supports, building components, and the choice of test strategies for or compartment structures such as the spent fuel both HSSCs and LSSCs.

pool), and systems relied upon to mitigate the ef-Until the ASME recommendations for improved fects of external events in cases where the PRA considered only internal events.

test methods are available, the existing IST test methods should be evaluated prior to choosing the Failure modes modeled by the PRA may not be all-test methods to be used for the HSSCs and LSSCs, a

inclusive. Consideration should be given to the depending on their expected failure modes, service failure modes modeled and the potential for the conditions, etc.

introduction of new failure modes related to the IST application. For example, if valve mispos.

Because of the importance of maintaining defense tioning has been assumed to be a low-probability in depth, particular attention should be given to event because of independent verification and identifying any containment systems involving therefore is not included in the PRA assumptions, IST components.

any changes to such independent verifications Step-wise program implementation, as discussed should be evaluated for potential impact on the in Regulatory Position 3.2, should be included as PRA results-part of the licensce's integrated decisionmaking Other qualitative or quantitative analyses that shed pr cess.

=

light on the relative safety importance of compo-The licensee's performance monitoring approach, nents, such as FMEA, shutdown risk, seismic risk, as discussed in Regulatory Position 3.3, should be I

and fire protection should be included in the re-included as part of the licensee's decisionmaking source information base.

process.

Value/ Impact Statement A draft value/ impact statement was published with the draft of this guide (DG-1062) when it was issued for public comment in June 1997. No significant changes were necessary from the original draft, so a separate value/ impact statement for this final guide has not been prepared. A copy of the draft value/ impact statement is available for inspection or copying for a fee in the Commission's Public Document Room at 2120 L Street NW, Washington, DC.

I 1.175 - 23

i i

UNITED STATES FIRST CLASS MAIL NUCLEAR REGULATORY COMMISSION POSTAGE AND FEES PAID WASHINGTON, DC 20555-0001 PERM M7 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300 l

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