U.S. NUCLEAR REGULATORY COMMISSION August 1998

REGULATORY GUIDE 

OFFICE OF NUCLEAR REGULATORY RESEARCH 

REGULATORY GUIDE 1.175 

(Draft was Issued as DG-1 062) 

AN APPROACH FOR PLANT-SPECIFIC, RISK-INFORMED 

DECISIONMAKING: INSERVICE TESTING

A. INTRODUCTON 

Background 

During the last several years both the U.S. Nuclear 

Regulatory Commission (NRC) and the nuclear indus

try have recognized that probabilistic risk assessment 

(PRA) has evolved to be more useful in supplementing 

traditional engineering approaches in reactor regula

tion. After the publication of its policy statement (Ref. 

1) on the use of PRAin nuclear regulatory activities, the 

Commission directed the NRC staff to develop a regu

latory framework that incorporated risk insights. That 

framework was articulated in a November 27,1995, pa

per to the Commission (Ref. 2). This regulatory guide, 

which addresses inservice testing (IST) of pumps and 

valves, and its companion regulatory documents (Refs. 

3-8) implement, in part, the Commission policy state

ment and the staff's framework for incorporating risk 

insights into the regulation of nuclear power plants. 

The NRC's policy statement on probabilistic risk 

analysis encourages greater use of this analysis tech

nique to improve safety decisionmaking and improve 

regulatory efficiency. One activity under way in re

sponse to the policy statement is the use of PRAin sup

port of decisions to modify an individual plant's IST 

program. Licensee-initiated IST program changes that

are consistent with currently approved staff positions 

(e.g., regulatory guides, standard review plans, branch 

technical positions) are normally evaluated by the NRC 

staff using traditional engineering analyses. In such 

cases, the licensee would not be expected to submit risk 

information in support of the proposed change. 

Licensee-initiated IST program change requests that go 

beyond current staff positions may be evaluated by the 

staff using traditional engineering analyses as well as 

the risk-informed approach set forth in this regulatory 

guide. A licensee may be requested to submit supple

mental risk information if such information is not pro

vided in the proposed risk-informed inservice testing 

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

formation on the proposed RI-IST program is not pro

vided to the staff, the staff will review the information 

provided by the licensee to determine whether the ap

plication can be approved based upon the information 

provided using traditional methods, and the staff will 

either approve or reject the application based upon the 

review. For those licensee-initiated RI-IST program 

changes that a licensee chooses to support (or is re

quested by the staff to support) with risk information, 

this regulatory guide describes an acceptable method 

for assessing the nature and impact of proposed RI-IST 

program changes by considering engineering issues 

and applying risk insights. Licensees submitting risk

USNRC REGULATORY GUIDES The guides ae Issued In the following ten broad divisions: 

Regulatory Guides awe Issued to describe and make available to the public auch Wlorma

ton as methods acceptable to he NRC staff for Implementing specific parts of the Com- 1. Power Reactors 6. Products 

mission's regulations, lechniques used by the staff inevaluating specific problemr or pos- 2. Research and Test Reactors 7. Transportation 

tulated accidents, and data needed by the NRC staff in its review of applications for per- 3a Fuels and Materials Facilities & Occupational Health 

mits aid licenss. Regulatory guides are not substitutes for regulations, and comprmence 4. Environmental and Sting 9. Antitrust and Frnancial Review 

with them i not requlred. Methodsnd solutMonsdifferentfrom ho9 setoutIntheguides materials "n Plant Protection 10L General 

wil be acceptable Ithey provide a basis for the findings requisite to the Issuance or con

Unuance of a permit or license by the Commission. Single copies of regulatory goides may be obtained free of charge by writing the Repro

duction and Distribution Services Section, Office of the Chief Information Officer, U.S. Nu

This gilds was Isued after iomments consideration of racaived from the public- Corn- dear Regulatory Commission, Washington, DC 20555-0001; or by fax at (301)415-2289; menits end suggestions lor improvements hn thesae guides areencouraged at all times, on or by a-mail to GRWl@NRC.GOV. 

deswilbe revised, as appropriate, to accommodate comments and to reflect new In

ation or aipennc. Issued guides may also be purchased from the National Technical Information Service on 

Written commerts may be aubmitted lo the Rules Review and Directives Branch, ADM, a standing order basis. Details on this service may be obtained by writing NTIS, 5285 Port 

U.S. Nuclear Regula Commission, Washington, DC 20555-0001. Royal Road, Springfleld, VA 22161.
information should address each of the principles of 

risk-informed regulation discussed in Regulatory 

Guide 1.174, "An Approach for Using Probabilistic 

Risk Assessment in Risk-Informed Decisions on Plant

Specific Changes to the Licensing Basis" (Ref. 3) and 

repeated in this guide. Licensees should identify how 

chosen approaches and methods (whether they are 

quantitative or qualitative, traditional or probabilistic), 

data, and criteria for considering risk are appropriate for 

the decision to be made. 

IST of snubbers was not addressed in this regula

tory guide, however, licensees interested in implement

ing a RI-IST program for snubbers may submit an alter

native to the NRC for consideration. 

Relationship to the Maintenance Rule 

10 CFR 50.65 

The Maintenance Rule, Section 50.65, "Require

ments for Monitoring the Effectiveness of Maintenance 

at Nuclear Power Plants," of 10 CFR Part 50, "Domes

tic licensing of Production and Utilization Facilities," 

requires that licensees monitor the performance or con

dition of structures, systems, or components (SSCs) 

against licensee-established goals in a manner suffi

cient to provide reasonable assurance that such SSCs 

are capable of fulfilling their intended function. Such 

goals are to be established, where practicable, com

mensurate with safety, and they are.to take into account 

industrywide operating experience. When the perfor

mance or condition of a component does not meet es

tablished goals, appropriate corrective actions are to be 

taken. 

Component monitoring that is performed as part of 

the Maintenance Rule implementation can be used to 

satisfy monitoring needs for RI-IST, and for such cases, 

the performance criteria chosen should be compatible 

with both the Maintenance Rule requirements and 

guidance and the RI-IST guidance provided in this 

guide. 

Purpose and Scope 

Current IST programs are performed in com

pliance with the requirements of 10 CFR 50.55a(f) and 

with Section XI of the ASME Boiler and Pressure Ves

sel Code (Ref. 9), which are requirements for all plants. 

This regulatory guide describes an acceptable alterna

tive approach applying risk insights from PRA to make 

changes to a nuclear power plant's IST program. An ac

companying Standard Review Plan (SRP) (Ref. 7) has 

been prepared for use by the NRC staff in reviewing RI

IST applications. Another guidance document, Regula

tory Guide 1.174 (Ref. 3), is referenced throughout this 

report. Regulatory Guide 1.174 provides overall guidance on the technical aspects that are common to devel

oping acceptable risk-informed programs for all ap

plications such as 1ST (this guide), inservice 

inspection, graded quality assurance, and technical 

specifications. 

This regulatory guide provides application

specific details of a method acceptable to the NRC staff 

for developing RI-IST programs and supplements the 

information given in Regulatory Guide 1.174. This 

guide provides guidance on acceptable methods for uti

lizing PRA information with established traditional en

gineering information in the development of RI-IST 

programs that have improved effectiveness regarding 

the utilization of plant resources while still maintaining 

acceptable levels of quality and safety. 

In this regulatory guide, an attempt has been made 

to strike a balance in defining an acceptable process for 

developing RI-IST programs without being overly pre

scriptive. Regulatory Guide 1.174 identifies a list of 

high-level safety principles that must be maintained 

during all risk-informed plant design or operational 

changes. Regulatory Guide 1.174 and this guide iden

tify acceptable approaches for addressing these basic 

high-level safety principles; however, licensees may 

propose other approaches for consideration by the NRC 

staff. It is intended that the approaches presented in this 

guide be regarded as examples of acceptable practice 

and that licensees should have some degree of flexibil

ity in satisfying regulatory needs on the basis of their 

accumulated plant experience and knowledge. 

Organization 

This regulatory guide is structured to follow the ap

proach given in Regulatory Guide 1.174. The discus

sion, Part B, gives a brief overview of a four-element 

process described in Regulatory Guide 1.174 as applied 

to the development of an RI-IST program. This process 

is iterative and generally not sequential. Part C, Regula

tory Position, provides a more detailed discussion of 

the four elements including acceptance guidelines. In 

Part C, Regulatory Position 1 addresses the first ele

ment in the process in which the proposed changes to 

the IST program are described. This description is 

needed to determine what supporting information is 

needed and to define how subsequent reviews will be 

performed. Regulatory Position 2 contains guidance 

for performing the engineering evaluation needed to 

support the proposed changes to the IST program (sec

ond process element). Regulatory Position 3 addresses 

program implementation, performance monitoring, 

and corrective action (third element). Regulatory Posi

tion 4 addresses documentation requirements (fourth 

element) for licensee submittals to the NRC and identi

fies additional information that should be maintained in

1.175-2
the licensee's records in case later review or reference is 

needed. The appendix contains additional guidance for 

dealing with certain IST-related issues such as might 

arise during the deliberations of the licensee in carrying 

out integrated decisionmaking. 

Relationship to Other Guidance Documents 

This regulatory guide provides detailed guidance 

on approaches to implement risk insights in IST pro

grams that are acceptable to the NRC staff. This 

application-specific guide makes extensive reference 

to Regulatory Guide 1.174 (Ref. 3) for general guid

ance. 

Companion regulatory guides (Refs. 4 and 5) ad

dress graded quality assurance and technical specifica

tions, and contain guidance similar to that given in this 

RI-ISTguide. SRP chapters associated with the risk-in

formed regulatory guides are available (Refs. 6-8). The 

SRP chapters are intended for NRC use during the re

view of industry requests for risk-informed program 

changes. SRP Chapter 3.9.7 (Ref. 7) addresses RI-IST 

and is consistent with the guidance given in this regula

tory guide. 

In the 1995-1998 period, the industry developed a 

number of documents addressing the increased use of 

PRAin nuclear plant regulation. The American Society 

of Mechanical Engineers (ASME) developed guide

lines for risk-based IST (Ref. 10) and later initiated 

code cases addressing IST component importance 

ranking and testing of certain plant components using 

risk insights. The Electric Power Research Institute 

(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

plications. The Nuclear Energy Institute (NEI) has also 

been developing guidelines on risk-based IST (Ref. 

12). These documents have provided useful viewpoints 

and proposed approaches for the staff's consideration 

during the development of the NRC regulatory guid

ance documents. 

Abbreviations 

ASME American Society of Mechanical Engi

neers

CCF 

CDF 

EPRI 

FV 

GQA 

HEP 

HSSC

common cause failure 

core damage frequency 

Electric Power Research Institute 

Fussell-Vesely risk importance measure 

graded quality assurance 

human error probability 

high safety-significant component 

I

1.1

ISI 

IST 

LERF 

LSSC 

MCS 

NEI 

NUMARC

inservice inspection 

inservice testing 

containment large early release frequency 

low safety-significant component 

minimal cut set 

Nuclear Energy Institute 

Nuclear Utilities Management Research 

Council

O&M Operations and Maintenance (ASME 

committee) 

PRA probabilistic risk assessment 

PSA probabilistic safety assessment

RAW

RI-IST 

SRP 

SSCs 

THERP 

USAR 

USNRC

risk achievement worth risk importance 

measure 

risk-informed IST (e.g., RI-IST programs) 

standard review plan 

structures, systems, and components 

Technique for Human Error Rate Predic

tion 

Updated Safety Analysis Report 

U.S. Nuclear Regulatory Commission

The information collections contained in this regu

latory guide are covered by the requirements of 10 CFR 

Part 50, which were approved by the Office of Manage

ment and Budget, approval number 3150-0011. The 

NRC may not conduct or sponsor, and a person is not 

required to respond to, a collection of information un

less it displays a currently valid OMB control number. 

B. DISCUSSION 

Key Safety Principles 

Regulatory Guide 1.174 (Ref. 3) identifies five key 

safety principles to be met for all risk-informed applica

tions and to be explicitly addressed in risk-informed 

plant program change applications. As indicated in 

Regulatory Guide 1.174, while these key principles are 

stated in traditional engineering terminology, efforts 

should be made wherever feasible to utilize risk evalua

tion techniques to help ensure and to show that these 

principles are met. These key principles and the loca

tion in this guide where each is addressed for RI-IST 

programs are as follows: 

1. The proposed change meets the current regu

lations unless it Is explicitly related to a requested 

exemption or rule change. (This principle is ad

dressed in Regulatory Positions 1.1 and 2.1 of this 

guide.) 

.75-3
Figure 1 Principles of Risk-Informed Regulation

2. The proposed change is consistent with the 

defense-in-depth philosophy. (Regulatory Position 

2.2.1) 

3. The proposed change maintains sufficient 

safety margins. (Regulatory Position 2.2.2) 

4. When proposed changes result in an increase 

in core damage frequency or risk, the increases 

should be small and consistent with the intent of the 

Commission's Safety Goal Policy Statement. (Regu

latory Positions 2.3, 2.4) 

5. The impact of the proposed change should be 

monitored using performance 

measurement strategies. (Regulatory Position 3.3) 

Regulatory Guide 1.174 gives additional guidance 

on the key safety principles applicable to all risk

informed applications. Figure I of this guide, repeated 

from Regulatory Guide 1.174, illustrates the consider

ation of each of these principles in risk-informed deci

sion making. 

A Four-Element Approach to Risk-Informed 

Decisionmaking for Inservice Testing Programs 

Regulatory Guide 1.174 (Ref. 3) describes a four

element process for developing risk-informed regulato

ry changes. The process is highly iterative. Thus, the fi

nal description of the proposed change to the IST 

program as defined in Element I depends on both the 

analysis performed in Element 2 and the definition of 

the implementation of the IST program performed in 

Element 3. The Regulatory Position of this guide pro

vides guidance on each element. 

While IST is, by its nature, a monitoring program, 

it should be noted that the monitoring referred to in Ele

ment 3 is associated with making sure that the assumptions made about the impact of the changes to the IST 

program are not invalidated. For example, if the test in

tervals are based on an allowable margin to failure, the 

monitoring is performed to make sure that these mar

gins are not eroded. An overview of this process specif

ically related to RI-IST programs is given in this sec

tion. The order in which the elements are performed 

may vary or occur in parallel, depending on the particu

lar application and the preference of the program devel

opers. 

Element 1: Define Proposed Changes to the 

Inservice Testing Program. 

The purpose of this element is to identify (1) the 

particular components that would be affected by the 

proposed changes in testing practices, including those 

currently in the IST program and possibly some that are 

not (if it is determined through new information and in

sights such as the PRA that these additional compo

nents are important in terms of plant risk) and (2) spe

cific revisions to testing schedules and methods for the 

chosen components. Plant systems and functions that 

rely on the affected components should be identified. 

Regulatory Position 1 gives a more detailed description 

of Element 1. 

Element 2: Perform Engineering Analysis 

In this element, both traditional engineering and 

PRA methods are used to help define the scope of the 

changes to the IST program and to evaluate the impact 

of the changes on the overall plant risk. Areas that are to 

be evaluated include the expected effect of the proposed 

RI-IST program on the design basis and severe acci

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

In this evaluation, the results of traditional engineering 

and PRA methods are to be considered together in an 

integrated decision process that will be carried over into

1.175-4
the implementation phase described below in Element 

3. PRA results should be used to provide information 

for the categorization of components into groupings of 

low safety-significant components (LSSC) and high 

safety-significant components (HSSQ. Components 

in the LSSC group would then be candidates for less 

rigorous testing when compared with those in the 

HSSC group. When the revised IST plan has been de

veloped, the plant-specific PRA should be used to eval

uate the effect of the planned program changes on the 

overall plant risk as measured by core damage fre

quency (CDF) and containment large early release fre

quency (LERF). 

During the integration of all the available informa

tion, it is expected that many issues will need to be re

solved through the use of a well-reasoned judgment 

process, often involving a combination of different en

gineering skills. This activity has typically been re

ferred to in industry documents as being performed by 

an "expert panel." As discussed further at the end of this 

section and in the appendix, this important process is 

the licensee's responsibility and may be accomplished 

by means other than a formal panel. In any case, the key 

safety principles discussed in this guide must be ad

dressed and shown to be satisfied regardless of the ap

proach used for RI-IST program decisionmaking. 

Additional application-specific details concerning 

RI-IST programs and Element 2 are contained in Regu

latory Positition 2 of this guide. 

Element 3: Define Implementation and 

Monitoring Program 

In this element, the implementation plan for the 

IST program is developed. This involves determining 

both the methods to be used and the frequency of test

ing. The frequency and method of testing for each com

ponent is commensurate with the component's safety 

significance. To the extent practicable, the testing 

methods should address the relevant failure mecha

nisms that could significantly affect component reli

ability. In addition, a monitoring and corrective action 

program is established to ensure that the assumptions 

upon which the testing strategy has been based contin

ue to be valid, and that no unexpected degradation in 

performance of the HSSCs and LSSCs occurs as a re

sult of the change to the IST program. Specific guid

ance for Element 3 is given in Regulatory Position 3. 

Element 4: Submit Proposed Change 

The final element involves preparing the documen

tation to be included in the submittal and the documen

tation to be maintained by the licensee for later refer

ence, if needed. The submittal will be reviewed by the

NRC according to SRP Chapter 19 and Section 3.9.7 

(Refs. 6 and 7). Guidance on documentation require

ments for RI-IST programs is given in Regulatory Posi

tion 4 of this regulatory guide. 

In carrying out this process, the licensee will make 

a number of decisions based on the best available infor

mation. Some of this information will be derived from 

traditional engineering practice and some will be pro

babilistic in nature resulting from PRA studies. It is the 

licensee's responsibility to ensure that its RI-IST pro

gram is developed using a well-reasoned and integrated 

decision process that considers both forms of input in

formation (traditional engineering and probabilistic) in 

a complementary manner. This important decisionma

king process may at times require the participation of 

special combinations of licensee expertise (licensee 

staff), depending on the technical and other issues in

volved, and may at times also need outside consultants. 

Industry documents have generally referred to the use 

of an expert panel for such decisionmaking. The appen

dix to this guide discusses a number of IST-specific is

sues such as might arise in expert panel deliberations. 

C. REGULATORY POSITION 

1. ELEMENT 1: DEFINE PROPOSED 

CHANGES TO INSERVICE TESTING 

PROGRAM 

In this first element of the process, the proposed 

changes to the IST program are defined. This involves 

describing what IST components (e.g., pumps and 

valves) will be involved and how their testing would be 

changed. Also included in this element is identification 

of supporting information and a proposed plan for the 

licensee's interactions with the NRC throughout the 

implementation of the RI-IST. 

1.1 Description of Proposed Changes 

A full description of the proposed changes in the 

IST program is prepared. This description would in

clude: 

(1) Identification of the aspects of the plant's design, 

operations, and other activities that require NRC 

approval that would be changed by the proposed 

RI-IST program. This will provide a basis from 

which the staff can evaluate the proposed changes. 

(2) Identification of the specific revisions to existing 

testing schedules and methods that would result 

from implementation of the proposed program. 

(3) Identification of the components in the plant that 

are directly and indirectly involved with the pro

posed testing changes. Any components that are 

not presently covered in the plant's IST program

1.175-5
but are determined to be important to safety (e.g., 

through PRA insights) should also be identified. 

In addition, the particular systems that are affected 

by the proposed changes should be identified 

since this information is an aid in planning the 

supporting engineering analyses. 

(4) Identification of the information that will be used 

in support of the changes. This will include perfor

mance data, traditional engineering analyses, and 

PRA information. 

(5) A brief statement describing the way how the pro

posed changes meet the objectives of the Commis

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

1.2 Inservice Testing Program Scope 

IST requirements for certain safety-related pumps 

and valves are specified in 10 CFR 50.55a. These com

ponents are to be tested according to the requirements 

of Section XI of the American Society of Mechanical 

Engineers (ASME) Boiler and Pressure Vessel Code 

(the Code) (Ref. 9) or the applicable ASME Operations 

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

For acceptance guidelines, the licensee's RI-IST 

program would include all components in the current 

Code-prescribed IST program. In addition, the pro

gram should include those non-Code components that 

the licensee's integrated decisionmaking process cate

gorized as HSSC. 

1.3 RI-IST Program Changes After Initial 

Approval 

This section provides guidance on reporting of pro

gram activities. The NRC will formally review the 

changes proposed to RI-IST programs that have al

ready received NRC approval. 

The licensee should implement a process for deter

mining when proposed RI-IST program changes re

quire formal NRC review and approval. Changes made 

to the NRC-approved RI-IST program that could affect 

the process and results that were reviewed and ap

proved by the NRC staff should be evaluated to ensure 

that the basis for the NRC staff's prior approval has not 

been compromised. All changes should be evaluated 

against the change mechanisms described in the regula

tions (e.g., 10 CFR 50.55a, 10 CFR 50.59) to determine 

whether NRC review and approval is required prior to 

implementation. If there is a question regarding this is

sue, the licensee should seek NRC review and approval 

prior to implementation. 

For acceptance guidelines, licensees can change 

their RI-IST programs consistent with the process and 

results that were reviewed and approved by the NRC

staff (i.e., as defined in the approved RI-IST program 

description). Prior to implementation, a process or pro

cedures should be in place to ensure that any such 

changes to the previously approved RI-IST program 

meet the acceptance guidelines of this section. 

The cumulative impact of all RI-IST program 

changes (initial approval plus later changes) should 

comply with the acceptance guidelines given in Regu

latory Position 2.3.3 below. 

Examples of changes to RI-IST programs that 

would require NRC's review and approval include, but 

are not limited to, the following: 

" Changes to the RI-IST program that involve pro

grammatic changes (e.g., changes in the accep

tance guidelines used for the licensee's integrated 

decisionmaking process), 

" Component test method changes that involve devi

ation from the NRC-endorsed Code requirements, 

NRC-endorsed Code Case, or published NRC 

guidance. 

Examples of changes to RI-IST programs that 

would not require NRC's review and approval include, 

but are not limited to, the following: 

" Changes to component groupings, test intervals, 

and test methods that do not involve a change to the 

overall RI-IST approach that was reviewed and ap

proved by the NRC, 

" Component test method changes that involve the 

implementation of an NRC-endorsed ASME Code 

or an NRC-endorsed Code Case, 

" Recategorization of components because of expe

rience, PRA insights, or design changes, but not 

programmatic changes when the process used to 

recategorize the components is consistent with the 

RI-IST process and results that were reviewed and 

approved by the NRC. 

2. ELEMENT 2: PERFORM ENGINEERING 

ANALYSIS 

As part of defining the proposed change to the li

censee's IST program, the licensee should conduct an 

engineering evaluation of the proposed change using a 

combination of traditional engineering methods and 

PRA. The major objective of this evaluation is to con

firm that the proposed program change will not com

promise defense in depth and other key safety prin

ciples described in this guide. Regulatory Guide 1.174 

(Ref. 3) provides general guidance for the performance 

of this evaluation, to be supplemented by the RI-IST

specific guidance in this guide.

1.175-6
2.1 Licensing Considerations 

2.1.1 Evaluating the Proposed Changes 

On a component-specific basis, the licensee should 

determine whether there are instances in which the pro

posed IST program change would affect the design, op

erations, and other activities at the plant, and the li

censee should document the basis for the acceptability 

of the proposed change by addressing the key prin

ciples. In evaluating proposed changes to the plant, the 

licensee should consider other licensing basis docu

ments (e.g., technical specifications, Final Safety Anal

ysis Report (FSAR), responses to NRC generic letters) 

in addition to the IST program documentation. 

The principal focus should be on the use of PRA 

findings and risk insights in support of proposed 

changes to a plant's design, operation, and other activi

ties that require NRC approval. Such changes include 

(but are not limited to) license amendments under 

10 CFR 50.90, requests for use of alternatives under 

10 CFR 50.55a, and exemptions under 10 CFR Part 12. 

However, the reviewer should note that there are certain 

docketed commitments that are not related to regula

tory requirements (e.g., commitments made by the li

censee in response to NRC Generic Letter 89-10 or 

96-05) that may be changed by licensees via processes 

other than as described in NRC regulations (e.g., con

sistent with Reference 14). 

A broad review of the plant's design, operations, 

and other activities may be necessary because proposed 

IST program changes could affect requirements or 

commitments that are not explicitly stated in the licens

ee's FSAR or IST program documentation. Further

more, staff approval of the design, operation, and main

tenance of components at the facility have likely been 

granted in terms other than probability, consequences, 

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

Therefore, it may also be appropriate to evaluate pro

posed IST program changes against other criteria (e.g., 

criteria used in either the licensing process or to deter

mine the acceptability of component design, operation 

and maintenance). 

The Director of the Office of Nuclear Reactor Reg

ulation is allowed by 10 CFR 50.55a to authorize alter

natives to the specific requirements of this regulation 

provided that the proposed alternative will ensure an 

acceptable level of quality and safety. Thus, alterna

tives to the acceptable RI-IST approaches presented in 

this guide may be proposed by licensees so long as sup

porting information is provided that demonstrates that 

the key principles discussed in Chapter 2 of this guide 

are maintained.

For acceptance guidelines, the licensee should re

view applicable documents to identify proposed 

changes to the IST program that would alter the design, 

operations, and other activities of the plant. On a com

ponent-specific basis, the licensee should (1) identify 

instances in which the proposed RI-IST program 

change would affect the design, operations, and other 

activities of the plant, (2) identify the source and nature 

of the requirements (or commitments), and (3) docu

ment the basis for the acceptability of the proposed re

qulrement changes, e.g., by addressing the key prin

ciples. 

The licensee must comply with 10 CFR 50.59, 

50.90, and 50.109 as applicable. The staff recognizes 

that there are certain docketed commitments that are 

not related to regulatory requirements that can be 

changed by licensees via processes other than described 

in NRC regulations (e.g., consistent with Reference 

14). 

2.1.2 Relief Requests and Technical Specification 

Changes 

The licensee should have included in the RI-IST 

program submittal the necessary exemption requests, 

technical specification amendment requests, and relief 

requests necessary to implement their RI-IST program. 

Individual component relief requests are not re

quired for adjusting the test interval of individual com

ponents that are categorized as having low safety sig

nificance (because the licensee's implementation plans 

for extending specific component test intervals should 

have been reviewed and approved by the NRC staff as 

part of the licensee's RI-IST program submittal). Simi

larly, if the proposed alternative includes improved test 

strategies to enhance the test effectiveness of compo

nents, additional relief to implement these improved 

test strategies is not required. 

For acceptance guidelines, the following are to be 

approved by the NRC before implementing the RI-IST 

program: 

" A relief request for any component, or group of 

components, that is not tested in accordance with 

the licensee's ASME Code of record or NRC

approved ASME code case. 

" A technical specification amendment request for 

any component, or group of components, if there 

are changes from technical specification require

ments. 

2.2 Traditional Engineering Evaluation 

This part of the evaluation is based on traditional 

engineering methods (not probabilistic). Areas to be

1.175-7
evaluated from this viewpoint include the potential ef

fect of the proposed RI-IST program on defense-in

depth attributes and safety margins. In addition, de

fense in depth and safety margin should also be 

evaluated, as feasible, using risk techniques (PRA). 

2.2.1 Defense-in-Depth Evaluation 

Because of its importance, both historically during 

the evolution of reactor safety practice and for the con

tinuation of public health and safety, the concept of de

fense in depth has been included in Regulatory Guide 

1.174 (Ref. 3) as one of the five key principles. In refer

ring to a proposed risk-informed program change, Sec

tion 2 of Regulatory Guide 1.174 states that the pro

posed change should be consistent with the 

defense-in-depth philosophy. Furthermore, as stated in 

Section 2.2.1.1, 

Consistency with the defense-in-depth philos

ophy is maintained if: 

" A reasonable balance is preserved among 

prevention of core damage, prevention of 

containment failure, and consequence miti

gation. 

" Over-reliance on programmatic activities 

to compensate for weaknesses in plant de

sign is avoided. 

" System redundancy, independence, and di

versity are preserved commensurate with 

the expected frequency, consequences of 

challenges to the system, and uncertainties 

(e.g., no risk outliers). 

" Defenses against potential common cause 

failures are preserved, and the potential for 

the introduction of new common cause fail

ure mechanisms is assessed. 

"* Independence of barriers is not degraded. 

"* Defenses against human errors are pre

served. 

"* The intent of the General Design Criteria in 

10 CFR Part 50, Appendix A is maintained. 

These defense-in-depth objectives apply to all risk

informed applications, and for some of the issues in

volved (e.g., no over-reliance on programmatic activi

ties and defense against human errors), it is fairly 

straightforward to apply them to the RI-IST program 

evaluation. Some specific examples of how certain oth

er of these objectives may be met for RI-IST applica

tions are as follows. The use of the multiple risk metrics 

of CDF and LERF and controlling their change resulting from the RI-IST program will maintain a balance 

between prevention of core damage, prevention of con

tainment failure, and consequence mitigation. Redun

dancy, diversity, and independence of safety systems 

should be considered after the initial choice is made in 

the categorization of components to ensure that these 

qualities are not degraded by the categorization. Inde

pendence of barriers and defense against common 

cause failures should also be considered in the review 

of the categorization. The improved understanding of 

the relative importance of plant components to risk re

sulting from the development of the RI-IST program 

should promote an improved overall understanding of 

how the components in the IST program contribute to a 

plant's defense in depth, and this should be discussed in 

the application. 

2.2.2 Safety Margin Evaluation 

The maintenance of safety margins is also a very 

important part of ensuring continued reactor safety and 

is included as one of the key safety principles in Section 

2 of Regulatory Guide 1.174 (Ref. 3). This principle 

states that the proposed change maintains sufficient 

safety margins. 

In addition, in Section 2.2.1.2, it is stated that with 

sufficient safety margins: 

"* Codes and standards or alternatives ap

proved for use by the NRC are met. 

" Safety analysis acceptance criteria in the li

censing basis (e.g., FSAR, supporting anal

yses) are met, or proposed revisions pro

vide sufficient margin to account for 

analysis and data uncertainty. 

It is possible that the categorization process will 

identify components that are currently not included in 

the IST program, and their addition as HSSCs will 

clearly improve safety margin in terms of CDF and 

LERF. It is also important that the performance moni

toring program be capable of quickly identifying sig

nificant degradation in performance so that, if neces

sary, corrective measures can be implemented before 

the margin to failure is significantly reduced. The im

proved understanding of the relative importance of 

plant components to risk resulting from the develop

ment of the RI-IST program should promote an im

proved understanding of how the components in the 

IST program contribute to a plant's margin of safety, 

and this should be discussed in the application. 

2.3 Probabilistic Risk Assessment 

Issues specific to the IST risk-informed process are 

discussed in this section. Regulatory Guide 1.174 (Ref.

1.175-8
3) contains much of the general guidance that is apl 

cable for this topic. 

In RI-IST, information obtained from a PI 

should be used in two ways: First, to provide input 

the categorization of SSCs into HSSC and LS' 

groupings; and second, to assess the impact of the pi 

posed change on CDF and LERF. Regulatory Positi 

2.3.1 discusses, in general terms, issues related to I 

quality, scope, and level of detail of a PRA that is us 

for IST applications. More specific considerations 

given in Regulatory Positions 2.3.2, and 2.3.3, whi 

address the use of PRA in categorization and in the 

sessment of the impact on risk metrics respectively 

2.3.1 Scope, Level of Detail, and Quality of 

Probabilistic Risk Assessments for Inservik 

Testing Applications 

For the quantitative results of the PRA to pla3 

major and direct role in decision making, there is a ne 

to ensure that they are derived from "quality" analysi 

and that the extent to which the results apply is well t 

derstood. Section 2.2.3 of Regulatory Guide 1.1 

(Ref. 3) addresses in general terms the issues related 

scope, level of detail, and quality of the PRA applied 

risk-informed applications. 

While a full scope PRA that covers all modes of c 

eration and initiating events is preferred, a lesser sco 

PRA can be used to provide useful risk informatic 

However, it must then be supplemented by additior 

considerations as discussed below. 

For the PRA to be useful in the development ol 

RI-IST program, it is necessary that the PRA model 

developed to the component level for the systems, i 

cluding non-safety systems, considered important I 

prevention of core damage and release of radioactivii 

A PRA used in RI-IST should be performed c( 

rectly and in a manner that is consistent with accept 

practices. The PRA should reflect the actual desiE 

construction, operating practices, and operating expe: 

ence of the plant. The quality required of the PRA 

commensurate with the role it plays in the determin 

tion of test intervals or test methods and with the rc 

the integrated decisionmaking panel plays in compe 

sating for limitations in PRA quality. Regulatory Gui 

1.174 and SRP Chapter 19 (Refs. 3 and 6) further di 

cuss the requirements of PRA quality. 

To be acceptable for application to RI-IST, PR 

models must reflect the as-built, as-operated plant, ai 

they must have been performed in a manner that is co 

sistent with accepted practices. The quality of the PR 

has to be shown to be adequate, commensurate with t] 

role the PRA results play in justifying changes to t]

,li- test intervals or strategies. The PRA model should be 

developed to the component level for the systems im

RA portant to safety. 

to If less than a full-scope PRA is used to support the 

SC proposed RI-IST program, supplemental information 

ro- (deterministic and qualitative) must be considered dur

.on ing the integrated decisionmaking process. 

the Acceptance guidelines for the required PRA quali

ied ty and scope are further defined in Regulatory Guide 

ire 1.174. 

ich 

s 2.3.2 Categorization of Components 

The categorization of components is important in 

the implementation of the RI-IST program since it is an 

efficient and risk-informed way of providing insights in 

,e the areas in which safety margin can be relaxed without 

unacceptable safety consequences. Thus, categoriza

y a dion of components, in addition to the traditional engi

ed neering evaluation described in Regulatory Position 

es, 2.2 and the calculation of change in overall plant risk 

in- described in Regulatory Position 2.3.3, will provide 

74 significant input to the determination of whether the 

to IST program is acceptable or not. 

to The determination of safety significance of com

ponents by the use of PRA-determined importance 

1p. measures is important for several reasons. 

pe • When performed with a series of sensitivity evalu

)n. ations, it can identify potential risk outliers by 

aal identifying IST components that could dominate 

risk for various plant configurations and operation

f a al modes, PRA model assumptions, and data and 

be model uncertainties. 

in- Importance measure evaluations can provide a use

for ful means to identify improvements to current IST 

ty. practices during the risk-informed application pro

cess. )r

ed System- or functional-level importance results can 

PIP provide a high level verification of component-lev

ri- el results and can provide insights into the potential 

is risk significance of IST components that are not 

a- modeled in the PRA. 

le General guidelines for risk categorization of com

n- ponents using importance measures and other informa

de tion are provided in Regulatory Guide 1.174 (Ref. 3). 

is- These general guidelines address acceptable methods 

for carring out categorization and some of the limita

A dions of this process. Guidelines that are specific to the 

Wd IST application are given in this section. As used here, 

)n- risk categorization refers to the process for grouping 

A IST components into LSSC and HSSC categories. 

he Components are initially categorized into HSSC 

he and LSSC groupings based on threshold values for the 

1.175-9
importance measures. Depending on whether the PRA 

is performed using the fault tree linking or event tree 

linking approach, importance measures can most easily 

be provided at the component or train level. In either 

case, the importance measures are applicable to the 

items taken one at a time, and therefore, as discussed in 

Regulatory Guide 1.174, while a licensee is free to 

choose the threshold values of importance measures, it 

will be necessary to demonstrate that the integrated im

pact of the change is such that Principle 4 is met. One 

acceptable approach is discussed in the next section. 

PRA systematically takes credit for non-Code 

components as providing support, acting as alterna

tives, and acting as backups to those components that 

are within the current Code. Accordingly, to ensure that 

the proposed RI-IST program will provide an accept

able level of quality and safety, these additional risk

important components should be included in licensees' 

RI-IST proposals. Specifically, the licensee's RI-IST 

program should include those ASME Code Class 1, 2, 

and 3 and non-Code components that the licensee's in

tegrated decisionmaking process categorized as HSSC 

and thus determined these components to be appropri

ate additional candidates for the RI-IST program. 

Although PRAs model many of the SSCs involved 

in the performance of plant safety functions, other 

SSCs are not modeled for various reasons. However, 

this should not imply that unmodeled components are 

not important in terms of contributions to plant risk. 

For example, some components are not modeled be

cause, certain initiating events may not be modeled 

(e.g., low power and shutdown events, or some external 

events); in other cases, components may not be directly 

modeled because they are grouped together with events 

that are modeled (e.g., initiating events, operator recov

ery events, or within other system or function bound

aries); and in some cases, components are screened out 

from the analysis because of their assumed inherent 

reliability; or failure modes are screened out because of 

their insignificant contribution to risk (e.g., spurious 

closure of a valve). When feasible, adding missing 

components or missing initiators or plant operating 

states to the PRA should be considered by the licensee. 

When this is not feasible, information based on tradi

tional engineering analyses and judgment is used to de

termine whether a component should be treated as an 

LSSC or HSSC. One approach to combining these dif

ferent pieces of information is to use what has been re

ferred to as an expert panel. Appendices B and C of 

Standard Review Plan Chapter 19 (Ref. 6) contain staff 

expectations on the use of expert panels in integrated 

decisionmaking and SSC categorization respectively.

In classifying a component not modeled in the 

PRA as LSSC, the expert panel should have determined 

that: 

"* The component does not perform a safety 

function, or does not perform a support 

function to a safety function, or does not 

complement a safety function. 

"* The component does not support operator 

actions credited in the PRA for either proce

dural or recovery actions. 

"* The failure of the component will not result 

in the eventual occurrence of a PRA initiat

ing event. 

"• The component is not a part of a system that 

acts as a barrier to fission product release 

during severe accidents. 

"* The failure of the component will not result 

in unintentional releases of radioactive ma

terial even in the absence of severe accident 

conditions. 

For acceptance guidelines, when using risk impor

tance measures to identify components that are low risk 

contributors, the potential limitations of these mea

sures have to be addressed. Therefore, information to 

be provided to the licensee's integrated decisionmaking 

process (e.g., expert panel) must include evaluations 

that demonstrate the sensitivity of the risk importance 

results to the important PRA modeling techniques, as

sumptions, and data. Issues that the licensee should 

consider and address when determining low risk con

tributors include truncation limit used, different risk 

metrics (i.e., CDF and LERF), different component 

failure modes, different maintenance states and plant 

configurations, multiple component considerations, 

defense in depth, and analysis of uncertainties (includ

ing sensitivity studies to component data uncertainties, 

common-cause failures, and recovery actions). 

While the categorization process can be used to 

highlight areas in which testing strategy can be im

proved and areas in which sufficient safety margins ex

ist to the point that testing strategy can be relaxed, it is 

the determination of the change in risk from the overall 

changes in the IST program that is of concern in demon

strating that Principle 4 has been met. Therefore, no ge

nerically applicable acceptance guidelines for the 

threshold values of importance measures used to cate

gorize components as HSSC or LSSC are given here. 

Instead, the licensee should demonstrate that the over

all impact of the change on plant risk is small as dis

cussed in Regulatory Position 2.3.3. 

As part of the categorization process, licensees 

must also address the initiating events and plant operat1.175-10
ing modes missing from the PRA evaluation. The li

censee can do this either by providing qualitative argu

ments that the proposed change to the IST program 

does not result in an increase on risk, or by demonstrat

ing that the components significant to risk in these mis

sing contributors are maintained as HSSC. 

2.3.3 Use of a PRA To Evaluate the Risk Increase 

from Changes in the IST Program 

One of the important uses of the PRAis to evaluate 

the impact of the IST change with respect to the accep

tance guidelines on changes in CDF and LERF as dis

cussed in Section 2.2.2 of Regulatory Guide 1.174 

(Ref. 3). In addition, the PRA can provide a baseline 

risk profile of the plant, and the extent of analysis of the 

baseline CDF and LERF depends on the proposed 

change in CDF and LERF. As discussed in Regulatory 

Guide 1.174, if the PRA is not full scope, the impact of 

the change must be considered by supplementing the 

PRA evaluation by qualitative arguments or by bound

ing analyses. 

2.3.3.1 Modeling the Impact of Changes in the 

IST Program. In order for the PRA to support the deci

sion appropriately, there should be a good functional 

mapping between the components associated with IST 

and the PRA basic event probability quantification. 

Part of the basis for the acceptability of the RI-IST pro

gram is a quantitative demonstration by use of a PRA 

that established risk measures are not significantly in

creased by the proposed changes to the IST for selected 

components. To establish this demonstration, the PRA 

includes models that appropriately account for the 

change in reliability of the components as a function of 

the IST program changes. In general, this will include 

not only changes to the test interval but also the effects 

of an enhanced testing method. Enhanced testing might 

be shown to improve or maintain component availabil

ity, even if the interval is extended. That is, a better test 

might compensate for a longer interval between tests. 

Licensees who apply for substantial increases in test in

terval are expected to address this area, i.e., as appropri

ate, consider improvements in testing that would com

pensate for the increased intervals under consideration. 

One model for the relationship between the com

ponent unavailability on demand and the test interval is 

given in NUREG/CR-6141 (Ref. 16), which assumes a 

constant rate (k) of transition to the failed state. Refer

ence 16 also describes how to account for various test 

strategies. 

In addition to transitions to a failed state that occur 

between component demands or tests, there is also a 

demand-related contribution to unavailability, corre

sponding to the probability that a component will fail to

operate when demanded, even though for some pur

poses it would have been considered "good" before be

ing subjected to the stress of the demand itself. This 

would have the effect of adding a constant to the test-in

terval-dependent contribution to the component un

availability on demand. The assumption that the total 

unavailability scales linearly with the test interval (i.e., 

doubles when test interval doubles) is conservative in 

the sense that it scales the test-interval-independent 

contribution along with the test-interval-dependent 

contribution, and in that respect tends to overstate the 

effect of test interval extension. This approximation is 

therefore considered acceptable; however, it should be 

noted that guidance aimed at improving the capability 

of tests to identify loss of performance margin is aimed 

partly at reducing the "demand" contribution as well, so 

that improved modeling in this area would appear to 

have the potential to support further improvements in 

allocation of safety resources. 

This model essentially assumes that failures are 

random occurrences and that the frequency of these oc

currences does not increase as the test interval is in

creased. However, as test intervals are extended, there 

is some concern that the failure rate, X, may increase. 

This failure rate, generally assumed constant, is based 

on data from current IST test intervals and therefore 

does not include effects that may arise from extended 

test intervals. It is possible that insidious effects such as 

corrosion or erosion, intrusion of foreign material into 

working parts, adverse environmental exposure, or 

breakdown of lubrication, which have not been encoun

tered with the current shorter test intervals, could sig

nificantly degrade the component if test intervals be

come excessively long. Therefore, unless it can be 

demonstrated that either degradation is not expected to 

be significant or that the test would identify degrada

tion before failures are likely to occur, use of the 

constant failure rate model could be nonconservative. 

One way to address this uncertainty is to use the 

PRA insights to help design an appropriate imple

mentation and monitoring program, for example, to ap

proach the interval increase in a stepwise fashion rather 

than going to the theoretically allowable maximum in a 

single step, or to stagger the testing of redundant com

ponents (test different trains on alternating schedules) 

so that the population of components is being sampled 

relatively frequently, even though individual members 

of the population are not. By using such approaches, the 

existence of the above effects can be detected and.com

pensatory measures taken to correct the testing of the 

remaining population members. However, it is impor

tant that the monitoring includes enough tests to be 

relevant, and that the tests are capable of detecting the

1.175-11
time-related degradation (performance monitoring is 

discussed in Regulatory Position 3.3). 

A check should also be performed to determine 

whether non-IST manipulation has been credited either 

in IST basic events or in compensating-component ba

sic events. If a component is stroked or challenged be

tween instances of IST, and if these activities are capa

ble of revealing component failure, the effective fault 

exposure time can be less than the RI-IST interval. It 

can be appropriate to take credit for this shortening of 

fault exposure time in the PRA quantification, pro

vided that there is assurance that the important failure 

modes are identified by the stroking or the system chal

lenges. This is not always trivial: If a functional success 

can be achieved by any one of n components in parallel, 

so that the function succeeds even if n-1 of the compo

nents fail, then merely monitoring successful function

al response does not show whether all components are 

operable unless verification of each component's state 

is undertaken. In addition, some instances of revealing 

a component fault through challenge have adverse con

sequences, including functional failure, and if credit is 

taken for shortening fault exposure time through func

tional challenges, it is necessary to account for this 

downside in the quantification of accident frequency. 

2.3.3.2 Evaluating the Change in CDF and 

LERF. Once the impact on the individual basic event 

probabilities has been determined, the change in CDF 

and LERF can be evaluated. There are some issues that 

must be carefully considered, which become more im

portant the larger the change in basic event probabili

ties. When using a fault tree linking approach to PRA, it 

is preferable that the model be re-solved rather than 

simply requantifying the CDF and LERF cutset solu

tions. In addition, it is important to pay attention to the 

parametric uncertainty analysis, especially if the 

change is dominated by cutsets that have multiple 

LSSCs. .The "state of knowledge" correlation effect 

(Ref. 16) could be significant if there are a significant 

number of cutsets with similar SSCs contributing to the 

change in risk. Regulatory Guide 1.174 (Ref. 3) dis

cusses the parametric uncertainty analysis in more 

detail. 

In addition, model and completeness uncertainties 

should be addressed as discussed in Regulatory Guide 

1.174. In particular, initiating events and modes of 

plant operations whose risk impact are not included in 

the PRA need additional analyses or justification that 

the proposed changes do not significantly increase the 

risk from those unmodeled contributors. 

23.3.3 Acceptance Guidelines. The change in 

risk from proposed changes to the IST program should

be consistent with the guidelines provided in Section 

2.2.4 of Regulatory Guide 1.174. In comparing the cal

culated risk to the guidelines, the licensee should ad

dress the model and completeness uncertainty as dis

cussed in Regulatory Guide 1.174 (Ref. 3). In addition, 

the licensee should address parameter uncertainty ei

ther by propagating the uncertainty during sequence 

quantification or by demonstrating that the "state-of

knowledge correlation" effect is not significant, espe

cially in cutsets in which the RI-IST changes affect 

multiple components that are similar. 

In evaluating the change in plant risk from pro

posed changes in the IST program, the licensee should 

perform the following. 

"* Evaluate the risk significance of extending the test 

interval on affected components. This requires that 

the licensee address the change in component 

availability as a function of test interval. The analy

sis should include either a quantitative considera

tion of the degradation of the component failure 

rate as a function of time, supported by appropriate 

data and analysis, or arguments that support the 

conclusion that no significant degradation will oc

cur. 

"* Consider the effects of enhanced testing to the ex

tent needed to substantiate the change. 

Other issues that should be addressed in the quanti

fication of the change in risk include the following. 

"• The impact of the IST change on the frequency of 

event initiators (those already included in the PRA 

and those screened out because of low frequency) 

should be determined. For applications in RI-IST, 

potentially significant initiators include valve fail

ure that could lead to interfacing system loss-of

coolant accidents (LOCAs) or to other sequences 

that fail the containment isolation function. 

"* The effect of common cause failures (CCFs) 

should be addressed either by the use of sensitivity 

studies or by the use of qualitative assessments that 

show that the CCF contribution would not become 

significant under the proposed IST program (e.g., 

by use of phased implementation, staggered test

ing, and monitoring for common cause effects). 

"* Justification of lST relaxations should not be based 

on credit for post-accident recovery of failed com

ponents (repair or ad hoc manual actions, such as 

manually forcing stuck valves to open). However, 

credit may be taken for proceduralized imple

mentation of alternative success strategies. For 

each human action that compensates for a basic 

event probability increasing as a result of IST re1.175-12
laxation, there should be a licensee commitment to 

ensure performance of the function at the level 

credited in the quantification. Excessively low hu

man failure probabilities Qess than 10-3) cannot be 

accepted unless there is adequate justification and 

there are adequate training programs, personnel 

practices, plant policies, etc., to ensure continued 

licensee performance at that level. 

"* The failure rates and probabilities used for compo

nents affected by the proposed change in IST 

should appropriately consider both plant-specific 

and generic data. The licensee should determine 

whether individual components affected by the 

change are performing more poorly than the aver

age associated with their class; the licensee should 

avoid relaxing IST for those components to the 

point that the unavailability of the poor performers 

would be appreciably worse than that assumed in 

the risk analysis. In addition, components that have 

experienced repeated failures should be reviewed 

to see whether the testing scheme (interval and 

methods) would be considered adequate to support 

the performance credited to them in the risk 

analysis. 

" The evaluation should be performed so that the 

truncation of LSSCs is considered. It is preferred 

that solutions be obtained from a re-solution of the 

model, rather than a requantification of CDF and 

LERF cutsets. 

" The cumulative impact of all RI-IST program 

changes (initial approval plus later changes) 

should comply with the acceptance guidelines 

given in this section.

safety principles. Because of the importance of these 

expectations, they will be repeated here. 

* All safety impacts of the proposed change 

are evaluated in an integrated manner as 

part of an overall risk management ap

proach in which the licensee is using risk 

analysis to improve operational and engi

neering decisions broadly by identifying 

and taking advantage of opportunities for 

reducing risk, and not just to eliminate re

quirements the licensee sees as undesirable. 

For those cases when risk increases are pro

posed, the benefits should be described and 

should be commensurate with the proposed 

risk increases. The approach used to iden

tify changes in requirements should be used 

to identify areas where requirements should 

be increased, 1 as well as where they could 

be reduced. 

"* The scope and quality of the engineering 

analyses (including traditional and proba

bilistic analyses) conducted to justify the 

proposed licensing basis change should be 

appropriate for the nature and scope of the 

change, should be based on the as-built and 

as-operated and maintained plant, and 

should reflect operating experience at the 

plant. 

"* The plant-specific PRA supporting li

censee proposals has been subjected to 

quality controls such as an independent 

peer review or certification. 2 

"* Appropriate consideration of uncertainty is 

given in analyses and interpretation of find

ings, including using a program of monitor2.4 Integrated Decisionmaking

This section discusses the integration of all the 

technical considerations involved in reviewing submit

tals from licensees proposing to implement RI-IST pro

grams. General guidance for risk-informed applica

tions is given Regulatory Guide 1.174 (Ref. 3) and in 

the new SRP sections, Chapter 19 (Ref. 6) for general 

guidance, and Section 3.9.7 (Ref. 7) for IST programs. 

These documents discuss a set of regulatory findings 

that form the basis for the staff to prepare an acceptable 

safety evaluation report (SER) for a licensee's risk

informed application. Specifically, Section 2 of Regu

latory Guide 1.174 identifies a set of "expectations" 

that licensees should follow in addressing the key

t

Tbe NRC staff is aware of but does not endorse guide

lines that have been developed (e.g., by NEI/NU

MARC) to assist in identifying potentially beneficial 

changes to requirements. 

2As discussed in Section 2.2.3.3 of Regulatory Guide 

1.174 (Ref. 3) in its discussion of PRA quality, such a peer review or certification is not a replacement for 

NRC review. Certification is defined as a mechanism 

for assuring that a PRA, and the process of developing 

and maintainingthat PRA, meet aset Oftechnicalstan

dards established byadiverse groupofpersonnel expe

rienced in developing PRA models, performing PRAs, 

and performing quality reviews of PRAs. Such a pro cess has been developed and integrated with a peer re viewprocess by, forexample, the BWR Owners Group 

and implemented for the purpose of enhancing quality 

of PRAs at several BWR facilities.

1.175-13
ing, feedback, and corrective action to ad

dress significant uncertainties. 

The use of core damage frequency (CDF) 

and large early release frequency (LERF)3 

as bases for probabilistic risk assessment 

acceptance guidelines is an acceptable ap

proach to addressing Principle 4. Use of the 

Commission's Safety Goal qualitative 

health objectives (QHOs) in lieu of LERF is 

acceptable in principle and licensees may 

propose their use. However, in practice, im

plementing such an approach would require 

an extension to a Level 3 PRA, in which 

case the methods and assumptions used in 

the Level 3 analysis, and associated uncer

tainties, would require additional attention. 

* Increases in estimated CDF and LERF re

sulting from proposed changes will be lim

ited to small increments. The cumulative 

effect of such changes should be tracked 

and considered in the decision process. 

* The acceptability of proposed changes 

should be evaluated by the licensee in an in

tegrated fashion that ensures that all prin

ciples are met.4 

* Data, methods, and assessment criteria 

used to support regulatory decisionmaking 

must be well documented and available for 

public review. 

These expectations apply to both probabilistic and 

traditional engineering considerations, which are ad

dressed in more detail in this chapter and in Regulatory 

Guide 1.174 (Ref. 3). 

Licensees are expected to review commitments re

lated to outage planning and control to verify that they 

3In this context, LERF is being used as a surrogate for 

the early fatality quantitative health objective (QHO). 

It isdefined as the frequency of those accidentsleading 

to significant, unmitigated releases from containment 

in a time frame prior to effective evacuation of the 

close-in population such that there is a potential for 

early health effects. Such accidents generally include 

unscrubbedreleasesassociatedwithearlycontainment 

failure at or shortly after vessel breach, containment 

bypass events, and loss of containment isolation. This 

definition is consistent with accident analyses used in 

the safetygoal screening criteria discussed in the Com

mission's regulatory analysis guidelines. An NRC con

tractor's report (Ref. 15) describes a simple screening 

approach for calculating LERF. 

4One important element of integrated decisionmaking 

can be the use of an'"expert panel." Such a panel is not a 

necessary component of risk-informed decisionmak

ing; butwhen it is used, the key principles and associat

ed decision criteria presented in this regulatory guide 

still apply and must be shown to have been met or tobe 

irrelevant to the issue at hand.

are appropriately reflected in the licensee's component 

grouping. This should include components required to 

maintain adequate defense in depth as well as compo

nents that might be operated as a result of contingency 

plans developed to support the outage. 

Licensees are also expected to review licensing ba

sis documentation to ensure that the traditional engi

neering related factors mentioned above are adequately 

modeled or otherwise addressed in the PRA analysis. 

When making final programmatic decisions, 

choices must be made based on all the available infor

mation. There may be cases when information is in

complete or when conflicts appear to exist between the 

traditional engineering data and the PRA-generated in

formation. It is the responsibility of the licensee in such 

cases to ensure that well-reasoned judgment is used to 

resolve the issues in the best manner possible, includ

ing due consideration to the safety of the plant. This 

process of integrated decisionmaking has been dis

cussed in various industry documents (Refs. 10 

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

The appendix to this regulatory guide includes some 

detailed guidance on certain aspects of integrated deci

sionmaking specific to RI-IST programs. As discussed 

in the appendix, it is not intended that an administrative 

body such as an expert panel must always be formed by 

the licensee to fulfill this function. Some general accep

tance guidelines for this important activity follow, with 

more specific details given in the appendix. 

In summary, acceptability of the proposed change 

should be determined by using an integrated decision

making process that addresses three major areas: (1) an 

evaluation of the proposed change in light of the plant's 

licensing basis, (2) an evaluation of the proposed 

change relative to the key principles and the acceptance 

criteria, and (3) the proposed plans for implementation, 

performance monitoring, and corrective action. As 

stated in the Commission's Policy Statement on the in

creased use of PRA in regulatory matters (Ref. 1), the 

PRA information used to support the RI-IST program 

should be as realistic as possible, with reduced unnec

essary conservatisms, yet include a consideration of 

uncertainties. These factors are very important when 

considering the cumulative plant risk and accounting 

for possible risk increases as well as risk benefits. The 

licensee should carefully document all of these kinds of 

considerations in the RI-IST program description, in

cluding those areas that have been quantified through 

the use of PRA, as well as qualitative arguments for 

those areas that cannot readily be quantified. 

The following are acceptance guidelines.

1.175-14

K
* The licensee's proposed RI-ISTprogram should be 

supported by both a traditional engineering analy

sis and a PRA analysis. 

The licensee's RI-IST program submittal should be 

consistent with the acceptance guidelines con

tained throughout this regulatory guide, specifi

cally with the expectations listed in this section, or 

the submittal should justify why an alternative ap

proach is acceptable. 

If the licensee's proposed RI-IST program is ac

ceptable based on both the deterministic and pro

babilistic analyses, it may be concluded that the 

proposed RI-IST program provides "an acceptable 

level of quality and safety" [see 10 CFR 

50.55a(a)(3)(i)]. 

3. ELEMENT 3: DEFINE IMPLEMENTATION 

AND MONITORING PROGRAM 

Upon approval of an RI-IST program, the licensee 

should have in place an implementation schedule for 

testing all HSSCs and LSSCs identified in their pro

gram. This schedule should include test strategies and 

testing frequencies for HSSCs and LSSCs that are with

in the scope of the licensee's IST program and compo

nents identified as HSSCs that are not currently in the 

IST program. 

3.1 Inservice Testing Program Changes 

This section discusses the test strategy changes 

(i.e., component test frequency and methods changes) 

that licensees should make as part of a RI-IST program. 

For acceptance guidelines, the RI-IST program 

should identify components for which the test strategy 

(i.e., frequency, methods or both) should be more fo

cused as well as components for which the test strategy 

might be relaxed. The information contained in, and de

rived from, the PRA should be used to help construct 

the testing strategy for components. To the extent prac

ticable, components with high safety significance 

should be tested in ways that are effective at detecting 

their risk-important failure modes and causes (e.g., 

ability to detect failure, to detect conditions that are pre

cursors to failure, and predict end of service life). Com

ponents categorized LSSC may be tested less rigor

ously than components categorized as HSSC (e.g., less 

frequent or informative tests). 

In some situations, an acceptable test strategy for 

components categorized HSSC may be to conduct the 

existing approved Code IST test at the Code-prescribed 

frequency. In some situations, an acceptable test strat

egy for components categorized LSSC may be to conduct the existing approved Code IST test at an extended 

interval. 

An acceptable strategy for testing components 

categorized HSSC and LSSC may be defined in NRC

approved ASME risk-informed Code Cases. Licensees 

who choose to pursue RI-IST programs should consid

er adopting test strategies developed by ASME and en

dorsed by the NRC. Deviations from endorsed Code 

Cases must be reviewed and approved by the NRC staff 

as part of the RI-IST program review. 

In establishing the test strategy for components, 

the licensee should consider component design, service 

condition, and performance, as well as risk insights. 

The proposed test strategy should be supported by data 

that are appropriate for the component. The omission of 

either generic or plant-specific data should be justified. 

The proposed test interval should be significantly less 

than the expected time to failure assumed in the PRAof 

the components in question (e.g., an order of magnitude 

less).5 In addition, the licensee should demonstrate that 

adequate component capability (margin) exists, above 

that required during design-basis conditions, such that 

component operating characteristics over time do not 

result in reaching a point of insufficient margin before 

the next scheduled test activity. 

The IST interval should generally not be extended 

beyond once every 6 years or 3 refueling outages 

(whichever is longer) without specific compelling doc

umented justification available on site for review. Ex

tensions beyond 6 years or 3 refueling outages (which

ever is longer) will be considered as component 

performance data at extended intervals is acquired. 

This is not meant to restrict a licensee from fully imple

menting NRC-approved component Code Cases. 

Components categorized HSSc that are not in the 

licensee's current IST program should (where practi

cal) be tested in accordance with the NRC-approved 

ASME risk-informed Code Cases, including com

pliance with all administrative requirements. When 

ASME Section XI or O&M Code testing is not practi

cal, alternative test methods should be developed by the 

licensee to ensure operational readiness and to detect 

component degradation (i.e., degradation associated 

with failure modes identified as being important in the 

licensee's PRA). As a minimum, a summary of these 

components and their proposed testing should be inclu

ded in the RI-IST program. 

For components categorized as HSSC that were the 

subject of a previous NRC-approved relief request (or 

an NRC-authorized alternative test), the licensee 

5

Forexample, the MOVexercise requirement (which is comparable to 

the current stroke time test) should be performed at intervals consid

erably smaller than the expected time to failure.

1.175-15

. I
should discuss the appropriateness of the relief in light 

of the safety significance of the component in their RI

IST submittal. 

If practical, IST components (with the exception of 

certain check valves and relief valves) should, as a 

minimum, be exercised or operated at least once every 

refueling cycle. More frequent exercising should be 

considered for components in any of the following cate

gories, if practical: 

"* Components with high risk significance, 

"* Components in adverse or harsh environmental 

conditions, or 

"• Components with any abnormal characteristics 

(operational, design, or maintenance conditions). 

The testing strategy for each component (or group 

of components) in the licensee's RI-IST program 

should be described in the RI-IST program description. 

The RI-IST program description should summarize all 

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

MOV testing in response to NRC Generic Letter 96-05, 

Ref. 18). The specific testing to be done on each com

ponent (or group of components) should be delineated 

in the licensee's IST program plan and is subject to 

NRC inspection. 

3.2 Program Implementation 

The applicable ASME Code generally requires that 

safety-related components within the program scope as 

defined in the current ASME Code be tested on a quar

terly frequency regardless of safety significance. The 

authorization of a risk-informed inservice testing pro

gram will allow the extension of certain component 

testing intervals and modification of certain component 

testing methods based on the determination of individ

ual component importance. The implementation of an 

authorized program will involve scheduling test inter

vals based on the results of probabilistic analysis and 

deterministic evaluation of each individual component. 

The R1-1ST program should distinguish between 

high and low safety-significant components for testing 

intervals. Components that are being tested using spe

cific ASME Codes, NRC-endorsed Code Cases for RI

IST programs, or other applicable guidance should be 

individually identified in the RI-IST program. The test 

intervals of the HSSC should be included in the R1-IST 

program for verification of compliance with the ASME 

Code requirements and applicable NRC-endorsed 

ASME Code Cases. Any component test interval or 

method that is not in conformance with the above 

should have specific NRC approval. Plant corrective 

action and feedback programs should be appropriately

referenced in the IST program and in the implementing 

and test procedures to ensure that testing failures are re

evaluated for possible adjustment to the component's 

grouping and test strategy. 

It is acceptable to implement RM-IST programs on a 

phased approach. Subsequent to the approval of a RI

IST program, implementation of interval extension for 

LSSC may begin at the discretion of the licensee and 

may take place on a component-, train-, or system

level. However, it is not acceptable to immediately ad-, 

just the test intervals of LSSC to the maximum pro-' 

posed test interval. Normally, test interval increases 

will be done step-wise, with gradual extensions being 

permitted consistent with cumulative performance data 

for operation at the extended intervals. The actual test

ing intervals for each component in the RI-IST program 

should be available at the plant site for inspection. 

It should be noted that the test described in the cur

rent ASME Code may not be particularly effective in 

detecting the important failure modes and causes of a 

component or group of components. A more effective 

test strategy may be to conduct an enhanced test at an 

extended test interval. 

HSSCs that are not in the current IST program 

should be tested, where practical, in accordance with 

the ASME Code, including compliance with all admin

istrative requirements. When ASME Section XI or 

O&M testing is not practical, alternative test methods 

should be developed by the licensee to ensure opera

tional readiness and to detect component degradation 

(i.e., degradation associated with failure modes identi

fied as being important in the licensee's PRA). As a 

minimum, a summary of these components and their 

proposed testing should be provided to the NRC as part 

of this review and prior to implementation of the risk

informed IST program at the plant. 

An acceptable method to extend the test interval for 

LSSC is to group like components and stagger their 

testing equally over the interval identified for a specific 

component based on the probabilistic analysis and de

terministic evaluation of each individual component. 

Initially, it would be desirable to test at least one com

ponent in each group every refueling outage. For exam

ple, component grouping should consider valve actua

tor type for power operated valves and pump driver 

type, as applicable. With this method, generic age

related failures could be identified while allowing im

mediate implementation for some components. For 

component groups that are insufficient in size to test 

one component every refueling outage, the imple

mentation of the interval should be accomplished in a 

more gradual step-wise manner, The selected test fre1.175-16
quency for LSSC that are to be tested on a staggered ba

sis should be justified in the RI-IST program. 

The following implementation activities are ac

ceptable: 

* For components that will be tested in accordance 

with the current NRC-approved Code test frequen

cy and method requirements, no specific imple

mentation schedule is required. The test frequency 

and method should be documented in the licensee's 

RI-IST program. 

* For components that will employ NRC-endorsed 

ASME Codes or Code Case methods, implementa

tion of the revised test strategies (i.e., interval ex

tension plan) should be documented in the licens

ee's RI-IST program. 

* For any alternative test strategies proposed by the 

licensee (i.e., for components within the scope of 

the current ASME code), the licensee should have 

specific NRC approval. 

The licensee should increase the test interval for 

components in a step-wise manner (i.e., equal or suc

cessively smaller steps, not to exceed one refueling 

cycle per step). If no significant time-dependent fail

ures occur, the interval can be gradually extended until 

the component is tested at the maximum proposed ex

tended test interval. An acceptable approach is to group 

similar components and test them on a staggered basis. 

Guidance on grouping components is contained in 

Position 2 of NRC Generic Letter 89-04 (Ref. 19) for 

check valves; Supplement 6 to NRC Generic Letter 

89-10 (Ref. 20), and Section 3.5 of ASME Code Case 

OMN-1 (Ref. 21) for motor-operated valves, or other 

documents endorsed by the NRC. 

3.3 Performance Monitoring 

Performance monitoring in RI-IST programs re

fers to the monitoring of inservice test data for compo

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

cluding both HSSC and LSS). The purpose of 

performance monitoring in a RI-IST program is two

fold. First, performance monitoring should help con

firm that no insidious failure mechanisms that are re

lated to the revised test strategies become important 

enough to alter the failure rates assumed in the justifica

tion of program changes. Second, performance moni

toring should, to the extent practicable, ensure that ade

quate component capability (i.e., margin) exists, above 

that required during design-basis conditions, so that 

component operating characteristics over time do not 

result in reaching a point of insufficient margin before 

the next scheduled test activity. Regulatory Guide 

1.174 (Ref. 3) provides guidance on performance monitoring when testing under design basis conditions is 

impracticable. In most cases, component-level moni

toring will be expected. 

Two important aspects of performance monitoring 

are whether the test frequency is sufficient to provide 

meaningful data and whether the testing methods, pro

cedures, and analysis are adequately developed to en

sure that performance degradation is detected. Compo

nent failure rates cannot be allowed to rise to 

unacceptable levels (e.g., significantly higher than the 

failure rates used to support the change) before detec

tion and corrective action take place. 

The NRC staff expects that licensees will integrate, 

or at least coordinate, their monitoring for RI-IST pro

gram with existing programs for monitoring equipment 

performance and other operating experience on their 

sites and, when appropriate, throughout the industry. In 

particular, monitoring that is performed as part of the 

Maintenance Rule (10 CFR 50.65) implementation can 

be used in the RI-IST program when the monitoring 

performed under the Maintenance Rule is sufficient for 

the SSCs in the RI-IST program. As stated in Regulato

ry Guide 1.174, if an application requires monitoring of 

SSCs not included in the Maintenance Rule, or in

volves SSCs that need a greater resolution of monitor

ing than the Maintenance Rule (e.g., component-level 

vs. train- or plant-level monitoring), it may be advanta

geous for a licensee to adjust the Maintenance Rule 

monitoring program rather than to develop additional 

monitoring programs for RI-IST purposes. Therefore, 

it may be advantageous to adjust the Maintenance Rule 

performance criteria to meet the acceptance guidelines 

below. 

For acceptance guidelines, monitoring programs 

should be proposed that are capable of adequately 

tracking the performance of equipment that, when de

graded, could alter the conclusions that were key to 

supporting the acceptance of the RI-IST program. 

Monitoring programs should be structured such that 

SSCs are monitored commensurate with their safety 

significance. This allows for a reduced level of moni

toring of components categorized as having low safety 

significance provided the guidance below is still met. 

The licensee's performance monitoring process 

should have the following attributes: 

* Enough tests are included to provide meaningful 

data, 

"* The test is devised such that incipient degradation 

can reasonably be expected to be detected, and 

"* The licensee trends appropriate parameters as re

quired by the ASME Code or ASME Code Case 

and as necessary to provide reasonable assurance

1.175-17

f r
that the component will remain operable over the 

test interval. 

Assurance must be established that degradation is 

not significant for components that are placed on an ex

tended test interval, and that failure rate assumptions 

for these components are not compromised by test data. 

It must be clearly established that those test procedures 

and evaluation methods are implemented that reason

ably ensure that degradation will be detected and cor

rective action will be taken. 

3A Feedback and Corrective Action 

The licensee's corrective action program for this 

application should contain a performance-based feed

back mechanism to ensure that if a particular compo

nent's test strategy is adjusted in a way that is ineffec

tive in detecting component degradation and failure, 

particularly potential common cause failure mecha

nisms, the RI-IST program weakness is promptly de

tected and corrected. Performance monitoring should 

be provided for systems, structures, and components 

with feedback to the RI-IST program for appropriate 

adjustments when needed. 

If component failures or degradation occur at a 

higher rate than assumed in the basis for the RI-IST pro

gram, the following basic steps should be followed to 

implement corrective action. 

"* The causes of the failures or degradation should be 

determined and corrective action implemented. 

" The component's test effectiveness should be re

evaluated, and the RI-IST program should be mo

dified accordingly. 

The following are acceptance guidelines. 

The licensee's corrective action program evaluates 

RI-IST components that either fail to meet the test ac

ceptance criteria or are otherwise determined to be in a 

nonconforming condition (e.g., a failure or degraded 

condition discovered during normal plant operation). 

The evaluation: 

(1) Complies with Criterion XVI, "Corrective Ac

tion," of Appendix B to 10 CFR Part 50. 

(2) Promptly determines the impact of the failure or 

nonconforming condition on system/train oper

ability and follows the appropriate technical spec

ification when component capability cannot be 

demonstrated. 

(3) Determines and corrects the apparent or root cause 

of the failure or nonconforming condition (e.g., 

improve testing practices, repair or replace the 

component). The root cause of failure should be

determined for all components categorized as hav

ing high safety significance, as well as for compo

nents categorized as having low safety signifi

cance when the apparent cause of failure may 

contribute to common cause failure. 

(4) Assesses the applicability of the failure ornoncon

forming condition to other components in the RI

IST program (including any test sample expansion 

that may be required for grouped components such 

as relief valves). 

(5) Corrects other susceptible RI-IST components as 

necessary. 

(6) Considers the effectiveness of the component's 

test strategy in detecting the failure or nonconfor

ming condition. Adjust the test interval and/or test 

methods, as appropriate, when the component (or 

group of components) experiences repeated or 

age-related failures or nonconforming conditions. 

The corrective action evaluations should periodi

cally be provided to the licensee's PRA group so that 

any necessary model changes and re-grouping are done 

as might be appropriate. The effect of the failures on 

overall plant risk should be evaluated as well as a con

firmation that the corrective actions taken will restore 

the plant risk to an acceptable level. 

The RI-IST program documents should be revised 

to document any RI-IST program changes resulting 

from corrective actions taken. 

3.5 Periodic Reassessment 

RI-IST programs should contain provisions 

whereby component performance data periodically 

gets fed back into both the component categorization 

and component test strategy determination (i.e., test in

terval and methods) process. These assessments should 

also take into consideration corrective actions that have 

been taken on past IST program components. (This pe

riodic reassessment should not be confused with the 

120-month program updates required by 10 CFR 

50.55a(f)(5)(i), whereby the licensee's IST program 

must comply with later versions of the ASME Code 

that have been endorsed by the NRC.) 

The assessment should: 

" Review and revise as necessary the models and 

data used to categorize components to determine 

whether component groupings have changed. 

" Reevaluate equipment performance to determine 

whether the RI-IST program should be adjusted 

(based on both plant-specific and generic informa

tion).

1.175-18
The licensee should have procedures in place to 

identify the need for more emergent RI-IST program 

updates (e.g., following a major plant modification or 

following a significant equipment performance prob

lem). 

Licensees may wish to coordinate these reviews 

with other related activities such as periodic PRA up

dates, industry operating experience programs, the 

Maintenance Rule program, and other risk-informed 

program initiatives. 

The acceptance guideline is that the test strategy 

for RI-IST components should be periodically assessed 

to reflect changes in plant configuration, component 

performance, test results, and industry experience. 

4. ELEMENT 4: DOCUMENTATION 

The recommended content of an RP-IST submittal 

is presented in this Regulatory Postion. The guidance 

provided below is intended to help ensure the com

pleteness of the information provided and should aid in 

shortening the time needed for the review process. The 

licensee should refer to the appropriate section of this 

regulatory guide to ascertain the level of detail of the 

documentation that should either be submitted to the 

NRC staff for review or retained onsite for inspection. 

To the extent practical the applicable sections of the re

gulatory guide have been identified on each list of 

documents. 

4.1 Documentation That Should Be in The 

Licensee's RI-IST Submittal 

* A request to implement a RI-IST program as an au

thorized alternative to the current NRC-endorsed 

ASME Code pursuant to 10 CFR 50.55a(a)(3)(i). 

0 A description of the change associated with the 

proposed RI-IST program (see Regulatory Posi

tion 1.1 above). 

0 Identification of any changes to the plant's design, 

operations, and other activities associated with the 

proposed RI-IST program and the basis for the ac

ceptability of these changes (see Regulatory Posi

tion 2.1.1). 

* A summary of key technical and administrative as

pects of the overall RI-IST program that includes: 

A description of the process used to identify 

candidates for reduced and enhanced IST re

quirements, including a description of the cate

gorization of components using the PRA and 

the associated sensitivity studies (see Regula

tory Position 2.3.2 above),

* A description of the PRA used for the catego

rization process and for the determination of 

risk impact, in terms of the process to ensure 

quality and the scope of the PRA, and how lim

itations in quality, scope, and level of detail are 

compensated for in the integrated decision

making process (see Regulatory Position 2.3.1 

above), 

* A description of how the impact of the change 

is modeled in the IST components (including a 

quantitative or qualitative treatment of compo

nent degradation) and a description the impact 

of the change on plant risk in terms of CDF and 

LERF and how this impact compares with the 

decision guidelines (see Regulatory Position 

2.3.3), 

• A description of how the key principles were 

(and will continue to be) maintained (see Reg

ulatory Positions 2.2, 2.3, and 2.4), 

* A description ofthe integrated decisionmaking 

process used to help define the RI-IST pro

gram, including any decision criteria used (see 

Regulatory Position 2.4), 

• A general implementation approach or plan 

(see Regulatory Positions 3.1 and 3.2), 

a A description of the testing and monitoring 

proposed for each component group (see Reg

ulatory Position 3.2), 

* A description of the RI-IST corrective action 

plan (see Regulatory Position 3.4), 

0 A description of the RI-IST program periodic 

reassessment plan (see Regulatory Position 3.5 

above). 

" A summary of any previously approved relief re

quests for components categorized as HSSC along 

with any exemption requests, technical specifica

tion changes, and relief requests needed to imple

ment the proposed RI-IST Program (see Regula

tory Position 2.1.2). 

" An assessment of the appropriateness of pre

viously approved relief requests. 

4.2 Documentation That Should Be Available 

Onsite For Inspection 

"* The overall IST Program Plan 

"* Administrative procedures related to RI-IST 

"* Component or system design basis documentation 

"* Piping and instrument diagrams for systems that 

contain components in the RI-IST program

1.175-19

I I I ; I I
" PRA and supporting documentation (see Regula

tory Position 2.3) 

" Categorization results, including the RI-IST pro

cess summary sheet for each component or group 

of components (see Regulatory Position 2.3.2) 

" Integrated decisionmakingprocess procedures, ex

pert panel meeting minutes (if applicable) (see 

Regulatory Position 2.4) 

" Detailed implementation plans and schedules (see 

Regulatory Position 3.2)

" Completed test procedures and any supplemental 

test data related to RI-IST (see Regulatory Position 

3.3) 

" Corrective action procedures (see Regulatory Posi

tion 3.4) 

" Plant-specific performance data (e.g., machinery 

history) for components in the RI-IST program 

(see Regulatory Positions 2.3.3 and 3.1) 

" A description of individual changes made to the 

RI-IST program after implementation (see Regula

tory Position 1.3)

1.175-20
REFERENCES

1. USNRC, "Use of Probabilistic Risk Assessment 

Methods in Nuclear Regulatory Activities: Final 

Policy Statement," Federal Register, Vol. 60, p 

42622, August 16, 1995. 

2. USNRC, "Framework for Applying Probabilistic 

Risk Analysis in Reactor Regulation," 

SECY-95-280, November 27, 1995.1 

3. USNRC, "An Approach for Using Probabilistic 

Risk Assessment in Risk-Informed Decisions on 

Plant-Specific Changes to the Licensing Basis," 

Regulatory Guide 1.174, July 1998.2 

4. USNRC "An Approach for Plant-Specific, Risk

Informed Decisionmaking: Graded Quality As

surance," Regulatory Guide 1.176, August 1998.2 

5. USNRC, "An Approach for Plant-Specific, Risk

Informed Decisionmaking: Technical Specifica

tions," Regulatory Guide 1.177, August 1998.2 

6. USNRC, "Standard Review Plan for Risk

Informed Decision Making," Standard Review 

Plan, NUREG-0800, Chapter 19, July 1998.2 

7. USNRC, "Standard Review Plan for Risk

Informed Decision Making: Inservice Testing," 

Standard Review Plan, NUREG-0800, Chapter 

3.9.7, August 1998.2 

8. USNRC, "Standard Review Plan for Risk

Informed Decision Making: Technical Specifica

tions," Standard Review Plan, NUREG-0800, 

Chapter 16.1, August 1998.2 

9. American Society of Mechanical Engineers 

(ASME) Boiler and Pressure Vessel Code, Section 

XI, ASME.3 

1

Copies are available for inspection or copying for afee from the NRC 

Public Document Room at 2120 L Street NW, Washington, DC; the 

PDR's mailing address is Mail Stop LL-6, Washington, DC 20555; 

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

2

Single copies of regulatory guides, both active and draft, and draft 

NUREG documents may be obtained free of charge by writing the 

Reproduction and Distribution Services Section, OCIO, USNRC, 

Washington, DC 20555-0001, or by fax to (301)415-2289, or by email 

to GRWI@NRC.GOV. Active guides may also be purchased from 

the National Technical Information Serviceonastandingorderbasis. 

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

Royal Road, Springfield, VA22161. Copiesofactive and draftguides are available for inspection or copying for a fee from the NRC Public 

Document Room at 2120 L Street NW, Washington, DC; the PDR's 

mailingaddressisMailStopLL-6,WashingtonDC20555;telephone 

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

3

Copiesmaybe obtained fromASME,345 East 47thStreet, NewYork, 

NY 10017.)

10. American Society of Mechanical Engineers, 

"Risk-Based Inservice Testing-Development of 

Guidelines," Research Report (CRDT-Vol. 40-2, 

Volume 2), 1996.0 

11. Electric Power Research Institute, "PSAApplica

tions Guide," EPRI TR-105396, August 1995.1 

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

dustry Guidelines for Risk-Based Inservice Test

ing," March 19, 1996.1 

13. American Society of Mechanical Engineers 

(ASME) Code for Operations and Maintenance of 

Nuclear Power Plants, OM Code-1995.3 

14. Nuclear Energy Institute, "Guidelines for Manag

ing NRC Commitments," Revision 2, Decem

ber 19, 1995.1 

15. W.T. Pratt et al., "An Approach for Estimating the 

Frequencies of Various Containment Failure 

Modes and Bypass Events," Draft NUREG/ 

CR-6595, December 1997.2 

16. P.K. Samanta et al., "Handbook of Methods for 

Risk-Based Analyses of Technical Specifica

tions," NUREG/CR-6141, December 1994.4 

17. G.E. Apostolakis and S. Kaplan, "Pitfalls in Risk 

Calculations," Reliability Engineering, Vol. 2, 

pages 135-145, 1981. 

18. USNRC, "Periodic Verification of Design-Basis 

Capability of Safety-Related Power-Operated 

Valves," Generic Letter 96-05, September 18, 

1996.1 

19. USNRC, "Guidance on Developing Acceptable 

Inservice Testing Programs," Generic Letter 

89-04, April 3, 1989.1 

20. USNRC, "Safety-Related (1) Motor-Operated 

Valve Testing and Surveillance," Generic Letter 

89-10, June 28, 1989.1 

21. American Society of Mechanical Engineers 

(ASME) Alternative Rules for Preservice and In

service Testing of Certain Electric Motor Oper

ated Valve Assemblies in LWR Power Plants, 

Code Case OMN-1, OM Code-1995; Subsection 

ISTC.3 

4

Copiesare available atcurrent ratesfrom the U.S.GovernmentPrint

ing Office, P.o. Box 37082, Washington, DC 20402-9328 (telephone 

(202)512-2249); or from the National Tbchnical Information Service 

by writing NTIS at 5285 Port Royal Road, Springfield, VA 22161. 

Copies are available forinspection orcopyingforafee from the NRC 

Public Document Room at 2120 L Street NW, Washington, DC; the 

PDR's mailing address is Mail Stop LL-6, Washington, DC 20555; 

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

1.175-21
APPENDIX A 

DETAILED GUIDANCE FOR INTEGRATED DECISIONMAKING

A.1 Introduction 

The increased use of probabilistic risk assessment 

(PRA) in nuclear plant activities such as in risk-in

formed inservice testing (IST) programs will require a 

balanced use of the probabilistic information with the 

more traditional engineering (sometimes referred to as 

"deterministic") information. Some structured process 

for considering both types of information and making 

decisions will be needed that will allow improvements 

to be made in plant effectiveness while maintaining ad

equate safety levels in the plant. This will be particular

ly important during initial program implementation 

and also for the subsequent early phases of the program. 

In some instances, the physical data from the PRA and 

from the deterministic evaluations may be insufficient 

to make a clearcut decision. At times, these two forms 

of information may even seem to conflict. In such 

cases, it is the responsibility of the licensee to assemble 

the appropriate skilled utility staff (and in some cases 

consultants) to consider all the available information in 

its various forms and to supplement this information 

with engineeringjudgment to determine the best course 

of action. The participants involved in this important 

role have generally been referred to in various industry 

documents as an "expert panel." In this appendix, this 

function will be described as being an engineering eval

uation without specifying how the evaluation is to be 

performed administratively. It is not the intention of 

this guidance to indicate that a special administrative 

body needs to be formed within the utility to satisfy this 

role. It is the function that is important and that must be 

performed in some well-organized, repeatable, and 

scrutable manner by the licensee. This function is all

pervasive in the implementation phase of such activi

ties as inservice inspection (ISI) and IST, and accord

ingly, the licensee has the responsibility to see that this 

function is done well. 

A.2 Basic Categories of Information To Be 

Considered 

Risk-importance measures may be used together 

with other available information to determine the rela

tive risk ranking (and thus categorization) of the com

ponents included in the evaluation. Results from all 

these sources are then reviewed prior to making final 

decisions about where to focus IST resources. 

Although the risk ranking of components can be 

used primarily as the basis for prioritizing IST at a 

plant, additional considerations need to be addressed

(e.g., defense in depth, common cause, and the single 

failure criterion), which may be more constraining than 

the risk-based criteria in some cases. Consideration 

must be given to these issues and component perfor

mance experience before the IST requirements for the 

various components are determined. 

IST experience should contribute an understanding 

of the important technical bases underlying the existing 

testing program before it is changed. The critical safety 

aspects of these bases should not be violated inadver

tently in changing over to a RI-IST, and important plant 

experience gained through the traditional IST should be 

considered during the change. 

The plant-specific PRA information should in

clude important perspectives With respect to the limita

tions of PRA modeling and analysis of systems, some 

of which may not be explicitly addressed within the 

PRA analysis. An understanding should also be pro

vided as to how the proposed changes in pump and 

valve testing could affect PRA estimates of plant risk. 

Plant safety experience should provide insights as

sociated with the traditional analyses (Chapter 15 of the 

plant Final Safety Analysis Report) and any effect that 

proposed changes in testing might have on the tradi

tional perspective of overall plant safety. 

Plant operational input should supplement the in

sights of plant safety with additional information re

garding the operational importance of components un

der normal, abnormal, and emergency conditions. 

There should also be input on operating history, system 

interfaces, and industry operating experience to supple

ment information from the IST. 

Maintenance considerations should provide per

spectives on equipment operating history, work prac

tices, and the implementation of the maintenance rule. 

Systems design considerations should include the 

potential effect of different design configurations (e.g., 

piping, valves, and pumps) on planning for a risk

informed IST, particularly if future plant modifications 

are contemplated or if systems are temporarily taken 

out of service for maintenance or replacement or repair. 

A.3 Specific Areas To Be Evaluated 

This section addresses some technical and admin

istrative issues that are currently believed to be particu

larly important for RI-IST applications. Additional is

sues of a more general nature that may arise in expert 

panel deliberations are given in SRP Chapter 19.

1.175-22
It should be confirmed that proper attention has 

been given to component classifications in systems 

identified in emergency operating procedures (and 

other systems) depended upon for operator recov

ery actions, primary fission product barriers ex

cluded from the PRA due to their inherent reliabil

ity (such as the RPV), passive items not modeled in 

the PRA (such as piping, cable, supports, building 

or compartment structures such as the spent fuel 

pool), and systems relied upon to mitigate the ef

fects of external events in cases where the PRA 

considered only internal events. 

Failure modes modeled by the PRA may not be all

inclusive. Consideration should be given to the 

failure modes modeled and the potential for the 

introduction of new failure modes related to the 

IST application. For example, if valve misposi

tioning has been assumed to be a low-probability 

event because of independent verification and 

therefore is not included in the PRA assumptions, 

any changes to such independent verifications 

should be evaluated for potential impact on the 

PRA results. 

Other qualitative or quantitative analyses that shed 

light on the relative safety importance of compo

nents, such as FMEA, shutdown risk, seismic risk, 

and fire protection should be included in the re

source information base.

" Attention should be given to the fact that compo

nent performance can be degraded from the effects 

of aging or harsh environments, and this issue will 

need to be addressed and documented. 

" The engineering evaluation should include the 

choice of new test frequencies, the identification of 

compensatory measures for potentially important 

components, and the choice of test strategies for 

both HSSCs and LSSCs. 

" Until the ASME recommendations for improved 

test methods are available, the existing IST test 

methods should be evaluated prior to choosing the 

test methods tobe used for the HSSCs and LSSCs, 

depending on their expected failure modes, service 

conditions, etc. 

"* Because of the importance of maintaining defense 

in depth, particular attention should be given to 

identifying any containment systems involving 

IST components. 

"* Step-wise program implementation, as discussed 

in Regulatory Position 3.2, should be included as 

part of the licensee's integrated decisionmaking 

process. 

"• The licensee's performance monitoring approach, 

as discussed in Regulatory Position 3.3, should be 

included as part of the licensee's decisionmaking 

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. 

1.175-23

I I
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001

FIRST CLASS MAIL 

POSTAGE AND FEES PAID 

USNRC 

PERMIT NO. G-67

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300