Regulatory guides are issued to describe and make available to the public such information as methods acceptable to the NRC staff for implementing

specific parts of the NRC’s regulations, techniques used by the staff in evaluating specific problems or postulated accidents, and data needed by the NRC

staff in its review of applications for permits and licenses. Regulatory guides are not substitutes for regulations, and compliance with them is not required. 

Methods and solutions different from those set out in the guides will be acceptable if they provide a basis for the findings requisite to the issuance or

continuance of a permit or license by the Commission.

This guide was issued after consideration of comments received from the public. Comments and suggestions for improvements in these guides are

encouraged at all times, and guides will be revised, as appropriate, to accommodate comments and to reflect new information or experience. Written

comments may be submitted to the Rules and Directives Branch, ADM, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. 

Regulatory guides are issued in ten broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities; 4, Environmental

and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review; and 10, General. 

Single copies of regulatory guides (which may be reproduced) may be obtained free of charge by writing the Distribution Services Section, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, or by fax to (301)415-2289, or by email to DISTRIBUTION@NRC.GOV. Electronic copies of this

guide and other recently issued guides are available at NRC’s home page at <WWW.NRC.GOV> through the Electronic Reading Room, Accession Number

ML032510128.

U.S. NUCLEAR REGULATORY COMMISSION Revision 1

September 2003

REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 1.178

AN APPROACH FOR PLANT-SPECIFIC RISK-INFORMED 

DECISIONMAKING FOR INSERVICE INSPECTION OF PIPING

A. INTRODUCTION

During the last several years, both the U.S. Nuclear Regulatory Commission (NRC) and the

nuclear industry have recognized that probabilistic risk assessment (PRA) has evolved to be more useful

in supplementing traditional engineering approaches in reactor regulation. After the publication of its

Policy Statement on the use of PRA in nuclear regulatory activities, the Commission directed the NRC

staff to develop a regulatory framework that incorporated risk insights. That framework was articulated

in a November 27, 1995, paper to the Commission (SECY-95-280). This regulatory guide, which

addresses inservice inspection (ISI) of piping, with its companion Standard Review Plan, Section 3.9.8

of NUREG-0800 (SRP Chapter 3.9.8), and other regulatory documents (Regulatory Guides 1.174, 1.175,

1.176, and 1.177; SRP Chapters 3.9.7, 16.1, and 19 ), implement, in part, the Commission’s Policy

Statement and the staff’s framework for incorporating risk insights into the regulation of nuclear power

plants.

In September 1998, the Commission published a version of this regulatory guide for trial use. As

stated therein, that regulatory guide issued for trial use did not establish any final staff positions for

purposes of the Backfit Rule, 10 CFR 50.109, and any changes to the regulatory guide prior to staff

adoption in final form would not be considered to be backfits as defined in 10 CFR 50.109(a)(1). This

was intended to ensure that the lessons learned from the subsequent regulatory review of industry

methodologies and the pilot plant applications could be adequately addressed in this document and that

the guidance is sufficient to enhance regulatory stability in the review, approval, and implementation of

proposed RI-ISI programs. 
1.178-2

The NRC staff has approved two methods describing how risk-informed ISI programs can

be developed and implemented. One methodology (EPRI TR-112657) was developed by the

Electric Power Research Institute (EPRI). The other methodology (WCAP-14572) was developed

by the Westinghouse Owners Group (WOG). Regulatory Guide 1.178 (for trial use) was used to

support the review and approval of the two industry-developed methodologies. Based on the

experience during the review and approval of the industry methodologies and the review and

approval of numerous plant-specific relief requests for inservice inspection programs, the NRC

staff is now issuing this updated version of this regulatory guide. 

While there has been a substantial effort to provide an opportunity for stakeholders in this

methodology to intereact with the staff in the development of this guide’s revisions, the revised

guide was not issued in draft form in the manner that has normally been used for regulatory guides. 

Rather, a more limited stakeholder’s review was utilized, including a public meeting noticed on

the NRC web site. Prior to the public meeting, the draft of this guide was placed in ADAMS with

a public notice identifying the ADAMS accession number. This approach was considered

adequate since the revisions are generally editorial in nature and intended to either update certain

information or to clarify language without substantial changes to the methodology itself. In

addition, since this guide was first issued in 1998, it has been successfully used by the industry and

staff in processing numerous requests by licensees to make risk-informed changes to inservice

inspection programs. 

Until the risk-informed inservice inspection (RI-ISI) process is approved for generic use,

the NRC staff anticipates that licensees will request changes to their ISI programs by requesting

NRC approval of alternative inspection programs that meet the criteria of 10 CFR 50.55a(a)(3)(i)

in Section 50.55a, "Codes and Standards," of 10 CFR Part 50, "Domestic Licensing of Production

and Utilization Facilities," providing an acceptable level of quality and safety. As always,

licensees should identify how the chosen approach, methods, data, and criteria are appropriate for

the decisions they need to make.

This guide's principal focus is on the use of PRA findings and risk insights for decisions on

changes proposed to plants’ inservice inspection programs for piping. Such changes include (but

are not limited to) license amendments under 10 CFR 50.90, requests for the use of alternatives

under 10 CFR 50.55a, and exemptions under 10 CFR 50.12. This regulatory guide describes

methods acceptable to the NRC staff for integrating insights from PRA techniques with traditional

engineering analyses into ISI programs for piping. 

Background

During recent years, both the NRC and the nuclear industry have recognized that PRA has

evolved to the point that it can be used increasingly as a tool in regulatory decisionmaking. In

August 1995, the NRC adopted a Policy Statement regarding the expanded use of PRA. In part,

the Policy Statement states that:

 The use of PRA technology should be increased in all regulatory matters to the extent

supported by the state of the art in PRA methods and data and in a manner that
1.178-3

complements the deterministic approach and supports the NRC’s traditional philosophy of

defense in depth.

 PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance

measures) should be used in regulatory matters, where practical within the bounds of the

state of-the-art, to reduce unnecessary conservatism associated with current regulatory

requirements, regulatory guides, license commitments, and staff practices. Where

appropriate, PRA should be used to support the proposal of additional regulatory

requirements in accordance with 10 CFR 50.109 (Backfit Rule). Appropriate procedures

for including PRA in the process for changing regulatory requirements should be

developed and followed. It is, of course, understood that the intent of this policy is that

existing rules and regulations be complied with unless these rules and regulations are

revised.

 PRA evaluations in support of regulatory decisions should be as realistic as practicable and

appropriate supporting data should be publicly available for review.

 The Commission’s safety goals for nuclear power plants and subsidiary numerical

objectives are to be used with appropriate consideration of uncertainties in making

regulatory judgments on the need for proposing and backfitting new generic requirements

on nuclear power plant licensees.

In its approval of the Policy Statement, the Commission articulated its expectation that

implementation of the Policy Statement will improve the regulatory process in three areas:

foremost, through safety decisionmaking enhanced by the use of PRA insights, through more

efficient use of agency resources, and through a reduction in unnecessary burdens on licensees.

In parallel with the publication of the Policy Statement, the staff developed a regulatory

framework that incorporates risk insights. That framework was articulated in a paper

(SECY-95-280) to the Commission. This regulatory guide, which addresses ISI programs of

piping at nuclear power plants, is part of the implementation of the Commission’s Policy Statement

and the staff’s framework for incorporating risk insights into the regulation of nuclear power

plants.

While the conventional regulatory framework, based on traditional engineering criteria,

continues to serve its purpose in ensuring the protection of public health and safety, the current

information base contains insights gained from over 2500 reactor-years of plant operating

experience and extensive research in the areas of material sciences, aging phenomena, and

inspection techniques. This information, combined with modern risk assessment techniques and

associated data, can be used to develop a more effective approach to ISI programs for piping. 

The current ISI requirements for piping components are found in 10 CFR 50.55a and the

General Design Criteria listed in Appendix A to 10 CFR Part 50. These requirements are

throughout the General Design Criteria, such as in Criterion I, "Overall Requirements"; Criterion
1.178-4

II, "Protection by Multiple Fission Product Barriers"; Criterion III, "Protection and Reactivity

Control Systems"; and Criterion IV, "Fluid Systems."

Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel

Code (ASME B&PVC) is referenced by 10 CFR 50.55a, which addresses the codes and standards

for design, fabrication, testing, and inspection of piping systems. The objective of the ISI program

is to identify service-induced degradation that might lead to pipe leaks and ruptures, thereby

meeting, in part, the requirements set in the General Design Criteria and 10 CFR 50.55a. ISI

programs are intended to address all piping locations that are subject to degradation. Incorporating

risk insights into the programs can focus inspections on the more important locations and reduce

personnel exposure, while at the same time maintaining or improving public health and safety. 

The justification for any reduction in the number of inspections should address the issue that an

increase in leakage frequency or a loss of defense in depth should not result from decreases in the

numbers of inspections. 

When categorizing piping segments in terms of their contribution to risk, it is the

responsibility of a licensee to ensure that the categorization of piping segments and the resulting

inspection programs are consistent with the key principles and risk guidelines (e.g., core damage

frequency (CDF) and large early release frequency (LERF)) addressed in Regulatory Guide 1.174. 

This regulatory guide augments the guidance presented in Regulatory Guide 1.174 by providing

guidance specific to incorporating risk insights to inservice inspection programs of piping. 

Purpose of the Guide

Consistent with Regulatory Guide 1.174, this regulatory guide focuses on the use of PRA

in support of a RI-ISI program. This guide provides guidance on acceptable approaches to meeting

the existing Section XI requirements for the scope and frequency of inspection of ISI programs. 

Its use by licensees is voluntary. Its principal focus is the use of PRA findings and risk insights for

decisions on changes proposed to a plant’s inspection program for piping. The current ISI

programs are performed in compliance with the requirements of 10 CFR 50.55a and with Section

XI of the ASME Boiler and Pressure Vessel Code, which are part of the plant’s licensing basis. 

This approach provides an acceptable level of quality and safety (per 10 CFR 50.55a(a)(3)(i)) by

incorporating insights from probabilistic risk and traditional analysis calculations, supplemented

with operating reactor data. Licensees who propose to apply RI-ISI programs would amend their

final safety analysis report (FSAR, Sections 5.3.4 and 6.6) accordingly. A Standard Review Plan

(SRP Chapter 3.9.8) has been prepared for use by the NRC staff in reviewing RI-ISI applications.

Additional augmented inspection programs to address generic piping degradation problems

have been recommended by the NRC to preclude piping failure and implemented by the industry. 

Notable examples of augmented programs for piping inspections include the following topics:

• intergranular stress corrosion cracking (IGSCC) of stainless steel piping in boiling

water reactors (BWRs) (Generic Letter 88-01), 

• thermal fatigue (NRC Bulletins 88-08 and 88-11, NRC Information Notice 93-20),

• stress corrosion cracking in pressurized water reactors (PWRs) (IE Bulletin 79-17),
1.178-5

• Service Water Integrity Program (NRC Generic Letter 89-13), 

• flow accelerated corrosion (FAC) in the balance of plant for both PWRs and BWRs

(NRC Generic Letter 89-08). 

Augmented programs have generally been developed to address observed degradation and

the inspections tend to be targeted at locations where the most severe effects are expected. 

Selected augmented programs, or parts of the programs, may be incorporated into a RI-ISI

program provided that the licensee identifies and the staff approves the specific programs and

program changes.

This document addresses risked-informed methods to develop, monitor, and update 

more efficient ISI programs for piping at a nuclear power facility. This guidance does not preclude

other approaches for incorporating risk insights into the ISI programs. Licensees may propose

other approaches for NRC consideration. It is intended that the methods presented in this guide be

regarded as examples of acceptable practices; licensees should have some flexibility in satisfying

the regulations on the basis of their accumulated plant experience and knowledge. This document

addresses risk-informed approaches that are consistent with the basic elements identified in

Regulatory Guide 1.174. In addition, this document provides guidance on the following for the

purposes of RI-ISI. 

 Estimating the probability of a leak, a leak that prevents the system from

performing its function (disabling leak), and a rupture for piping segments, 

 Identifying the structural elements for which ISI can be modified (reduced or

increased), based on factors such as risk insights, defense in depth, reduction of

unnecessary radiation exposure to personnel,

 Determining the risk impact of changes to ISI programs,

 Capturing deterministic considerations in the revised ISI program, and

 Developing an inspection program that monitors the performance of the piping

elements for consistency with the conclusions from the risk assessment. 

Until the RI-ISI process is approved for generic use, the staff anticipates that licensees will

request changes to their ISI programs by requesting NRC approval of a proposed inspection

program that meets the criteria of 10 CFR 50.55a(a)(3)(i), providing an acceptable level of quality

and safety. The licensee's RI-ISI program will be enforceable under 10 CFR 50.55a. 

Scope of the RI-ISI Program

This regulatory guide only addresses changes to the ISI programs for inspection of piping. 

To adequately reflect the risk implications of piping failure, both partial and full-scope RI-ISI

programs are acceptable to the NRC staff. 
1

 Generally, ASME Code Class 1 includes all reactor pressure boundary components. ASME Code Class 2 generally includes

systems or portions of systems important to safety that are designed for post-accident containment and removal of heat and

fission products. ASME Code Class 3 generally includes the system components or portions of systems important to safety that

are designed to provide cooling water and auxiliary feedwater for the front-line systems. 

1.178-6

Partial Scope: A licensee may elect to limit its RI-ISI program to a subset of piping classes, for

example, ASME Class-1 piping only, including piping exempt from the current requirements. 

Partial scope applications should include the full population of piping within the selected subset of

piping such as ASME Class and/or plant systems. 

Full Scope: A full scope RI-ISI includes:

 All Class 1, 2, and 31

 piping within the current ASME Section XI programs, and

 All piping whose failure would compromise:

 Safety-related structures, systems, or components that are relied upon to remain

functional during and following design basis events to ensure the integrity of the

reactor coolant pressure boundary, the capability to shut down the reactor and

maintain it in a safe shutdown condition, or the capability to prevent or mitigate the

consequences of accidents that could result in potential offsite exposure comparable

to 10 CFR Part 100 guidelines.

 Non-safety-related structures, systems, or components:

 That are relied upon to mitigate accidents or transients or are used in plant

emergency operating procedures; or

 Whose failure could prevent safety-related structures, systems, or

components from fulfilling their safety-related function; or

 Whose failure could cause a reactor scram or actuation of a safety-related

system.

For both the partial and full scope evaluations, the licensee is to demonstrate compliance

with the acceptance guidelines and key principles of Regulatory Guide 1.174.

The inspection locations of concern include all weld and base metal locations at which

degradation may occur, although pipe welds are the usual point of interest in the inspection

program. Within this regulatory guide, references to "welds" are intended in a broad sense to

address inspections of critical structural locations in general, including the base metal as well as

weld metal. Inspections will often focus on welds because detailed evaluations will often identify

welds as the locations most likely to experience degradation. Welds are most likely to have

fabrication defects, welds are often at locations of high stress, and certain degradation mechanisms

(stress corrosion cracking) usually occur at welds. Nevertheless, there are other degradation

mechanisms such as flow-assisted-corrosion (e.g., erosion-corrosion) and thermal fatigue that

occur independent of welds. 
2

 Level 1 - accident sequence analysis, Level 2 - accident progression and source term analysis, and Level 3 - offsite consequence

analysis.

1.178-7

Licensees implementing the risk-informed process may identify piping segments

categorized as safety significant that are not currently subject to the traditional Code requirements

(e.g., outside the Code boundaries, including Code exempt piping) or are not being inspected to a

level that is commensurate with their risk significance. In this context, safety significant refers to a

piping segment that has a relatively high contribution to risk. PRA systematically takes credit for

systems with non-Code or exempt piping that provide support, act as alternatives, and act as

backups to those systems with piping that are within the scope of the current Section XI of the

Code. The RI-ISI program should result in inspections of safety significant piping. 

Organization and Content

This regulatory guide is structured to follow the general four-element process for

risk-informed applications discussed in Regulatory Guide 1.174. The Discussion section

summarizes the four-element process developed by the staff to evaluate proposed changes related

to the development of a RI-ISI program. Regulatory Position 1 discusses an acceptable approach

for defining the proposed changes to an ISI program. Regulatory Position 2 addresses, in general,

the traditional and probabilistic engineering evaluations performed to support RI-ISI programs and

presents the risk acceptance goals for determining the acceptability of the proposed change. 

Regulatory Position 3 presents one acceptable approach for implementing and monitoring

corrective actions for RI-ISI programs. The documentation the NRC will need to render its safety

decision is discussed in Regulatory Position 4. 

Relationship to Other Guidance Documents

As stated above, this regulatory guide discusses acceptable approaches to incorporate risk

insights into an ISI program and directs the reader to Regulatory Guide 1.174 and SRP Chapters

19 and 3.9.8 for additional guidance, as appropriate. Further guidance is being developed in Draft

Regulatory Guide DG-1122 and Draft SRP Chapter 19.1. Regulatory Guide 1.174 describes a

general approach to risk-informed regulatory decisionmaking and discusses specific topics

common to all risk-informed regulatory applications. Draft Regulatory Guide DG-1122, when

finalized, will provide guidance on determining the quality of the PRA, in toto or for those parts

that are used to support an application and are sufficient to provide confidence in the results such

that they can be used in regulatory decisionmaking for light-water reactors. Topics addressed in

these documents include:

 PRA quality - characteristics and attributes for technical elements of a PRA,

 PRA scope - internal and external event initiators, at-power and shutdown modes of

operation, consideration of requirements for Level 1, 2, and 32

 analyses,

 PRA peer review - approach, process, and documentation,

 Risk metrics - CDF, LERF, importance measures,
1.178-8

 Sensitivity and uncertainty analyses.

To the extent that a licensee elects to use PRA as an element to enhance or modify its

implementation of activities affecting the safety-related functions of SSCs subject to the provisions

of Appendix B to 10 CFR Part 50, the pertinent requirements of Appendix B are applicable. 

The information collections contained in this regulatory guide are covered by the

requirements of 10 CFR Part 50, which were approved by the Office of Management and Budget

(OMB), approval number 3150-0011. The NRC may not conduct or sponsor, and a person is not

required to respond to, a request for information or an information collection requirement unless

the requesting document displays a currently valid OMB control number.

Abbreviations and Definitions

ASME American Society of Mechanical Engineers

BPVC Boiler and Pressure Vessel Code

CCDF Conditional core damage frequency

CCF Common cause failure

CDF Core damage frequency

CLERF Conditional large early release frequency

Expert Elicitation A process used to estimate failure rates or probabilities of piping when

data and computer codes are unavailable for the intended purpose.

Expert Panel Normally refers to plant personnel experienced in operations,

maintenance, PRA, ISI programs, and other related activities and

disciplines that impact the decision under consideration.

FSAR Final Safety Analysis Report

IGSCC Intergranular stress corrosion cracking 

Importance Measures Used in PRA to rank systems or components in terms of risk significance

ISI Inservice inspection

IST Inservice testing

LERF Large early release frequency

NDE Nondestructive examination

NEI Nuclear Energy Institute

NRC Nuclear Regulatory Commission

PRA Probabilistic risk assessment

PSA Probabilistic safety assessment

RCPB Reactor coolant pressure boundary

RI-ISI Risk-informed inservice inspection

Staff Refers to NRC employees

Sensitivity Studies Varying parameters to assess impact due to uncertainties

SRP Standard Review Plan

SRRA Structural reliability/risk assessment (refers to fracture mechanics

analysis)

SSCs Structures, systems and components

Tech Specs Technical specifications
1.178-9

B. DISCUSSION

When a licensee elects to incorporate risk insights into its ISI programs, it is anticipated

that the licensee will build upon its existing PRA activities. The five key principles involved in

the integrated decisionmaking process are described in detail in Regulatory Guide 1.174. In

addition, Regulatory Guide 1.174 describes a four-element process for evaluating proposed

risk-informed changes. 

The key principles and the section of this guide that addresses each of these principles for

RI-ISI programs are as follows. 

1. The proposed change meets the current regulations unless it is explicitly related to a

requested exemption or rule change. (Regulatory Position 2.1.1)

2. The proposed change is consistent with the defense-in-depth philosophy. (Regulatory

Position 2.1.2)

3. The proposed change maintains sufficient safety margins. (Regulatory Position 2.1.3)

4. When proposed changes result in an increase in CDF or risk, the increases should be small

and consistent with the intent of the Commission’s Safety Goal Policy Statement. 

(Regulatory Position 2.2)

5. The impact of the proposed change should be monitored by using performance

measurement strategies. (Regulatory Position 3)

Section 2 of Regulatory Guide 1.174 describes a four-element process for developing

risk-informed regulatory changes. These are: define the change, perform an engineering analysis,

define the implementation and monitoring program, and submit the proposed change. The order in

which the elements are performed may vary or they may occur in parallel, depending on the

particular application and the preference of the program developers. The process is highly

iterative. Thus, the final description of the proposed change to the ISI program as defined in

Element 1 depends on both the analysis performed in Element 2 and the definition of the

implementation of the ISI program performed in Element 3. While ISI is, by its nature, an

inspection and monitoring program, it should be noted that the monitoring referred to in Element 3

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

ISI program are not invalidated. For example, if the inspection intervals are based on an allowable

margin to failure, the monitoring is performed to make sure that these margins are not eroded. 

Element 4 involves preparing the documentation to be submitted to the NRC and to be maintained

by the licensee for later reference.
1.178-10

C. REGULATORY POSITION

1. ELEMENT 1: DEFINE THE PROPOSED CHANGES TO ISI PROGRAMS

In this first element of the process, the proposed changes to the ISI program are defined. 

This involves describing the scope of piping that would be incorporated in the overall assessment

and how the inspection of this piping 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-ISI program.

1.1 Description of Proposed Changes

A full description of the proposed changes in the ISI program is to be prepared. This

description should include:

 Identification of the plant’s current requirements that would be affected by the proposed

RI-ISI program. To provide a basis from which to evaluate the proposed changes, the

licensee should also confirm that the plant’s design and operation is in accordance with its

current requirements and that engineering information used to develop the proposed RI-ISI

program is also consistent with the current requirements.

 Identification of the elements of the ISI program to be changed.

 Identification of the piping in the plant that is both directly and indirectly involved with the

proposed changes. Any piping not presently covered in the plant’s ISI program but

categorized as safety significant (e.g., through an integrated decisionmaking process using

PRA insights) should be identified and appropriately addressed. 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.

 Identification of the information that will be used to support the changes. This could

include performance data, traditional engineering analyses, and PRA information.

 A brief statement describing how the proposed changes meet the intent of the

Commission’s PRA Policy Statement.

1.2 Changes to Approved RI-ISI Programs

This section provides guidance on the need for licensees to report program activities and

guidance on formal NRC review of changes made to RI-ISI programs.

The licensee should implement a process for determining when RI-ISI program changes

require formal NRC review and approval. Changes made to the NRC-approved RI-ISI program

that could affect the process and results that were reviewed and approved by the NRC staff should

be evaluated to ensure that the basis for the staff’s approval has not been compromised. All
1.178-11

changes should be evaluated using the change mechanisms described in the applicable regulations

(e.g., 10 CFR 50.55a, 10 CFR 50.59) to determine whether NRC review and approval are required

prior to implementation. If there is a question regarding this issue, the licensee should seek NRC

review and approval prior to implementation.

2. ELEMENT 2: ENGINEERING ANALYSIS

As part of defining the proposed change to the licensee’s ISI program, the licensee should

conduct an engineering evaluation of the proposed change, using and integrating a combination of

traditional engineering methods and PRA. The major objective of this evaluation is to confirm

that the proposed program change will not compromise defense in depth, safety margins, and other

key principles described in this guide and in Regulatory Guide 1.174. Regulatory Guide 1.174

provides general guidance for performing this evaluation, which is supplemented by the RI-ISI

guidance herein.

The regulatory issues and engineering activities that should be considered for a

risk-informed ISI program are summarized here. For simplicity, the discussions are divided into

traditional and PRA analyses. Regulatory Position 2.1 addresses the traditional engineering

analysis, Regulatory Position 2.2 addresses the PRA-related analysis, and Regulatory Position 2.3

describes the integration of the traditional and PRA analyses. In reality, many facets of the

traditional and PRA analyses are iterative.

The engineering evaluations are to:

 Demonstrate that the change is consistent with the defense-in-depth philosophy;

 Demonstrate that the proposed change maintains sufficient safety margins;

 Demonstrate that when proposed changes result in an increase in CDF or risk, the increase

is small and consistent with the intent of the Commission’s Safety Goal Policy Statement;

and

 Support the integrated decisionmaking process.

The scope and quality of the engineering analyses performed to justify the changes

proposed to the ISI programs should be appropriate for the nature and scope of the change. The

decision criteria associated with each key principle identified above are presented in the following

subsections. Equivalent criteria can be proposed by the licensee if such criteria can be shown to

meet the key principles set forth in Section 2 of Regulatory Guide 1.174. 

2.1 Traditional Engineering Analysis 

This part of the evaluation is based on traditional engineering methods. Areas to be

evaluated from this viewpoint include meeting the regulations, defense-in-depth attributes, safety
1.178-12

margins, assessment of failure potential of piping segments, and assessment of primary and

secondary effects (failures) that result from piping failures. 

The engineering analysis for a RI-ISI piping program will achieve the following:

1. Assess compliance with applicable regulations,

2. Perform defense-in-depth evaluation,

3. Perform safety margin evaluation,

4. Define piping segments, 

5. Assess failure potential for the piping segment,

6. Assess the consequences (both direct and indirect) of piping segment failure,

7. Categorize the piping segments in terms of safety significance,

8. Develop an inspection program,

9. Assess the impact of changing the ISI program on CDF and LERF, and

10. Demonstrate conformance with the key principles (e.g., maintaining sufficient safety

margins, defense in depth consideration, Commission’s Safety Goal Policy, etc.).

2.1.1 Assess Compliance with Applicable Regulations

The engineering evaluation should assess whether the proposed changes to the ISI

programs would compromise compliance with the regulations. The evaluation should consider the

appropriate requirements in the licensing basis and applicable regulatory guidance. Specifically,

the evaluation should consider:

 10 CFR 50.55a 

 Appendix A to 10 CFR Part 50

 Appendix B to 10 CFR Part 50

 ASME B&PVC, Section XI (10 CFR Part 50.55a)

 Regulatory Guide 1.84, “Design, Fabrication, and Materials Code Case Acceptability,

ASME Section III”

 Regulatory Guide 1.147, “Inservice Inspection Code Case Acceptability, ASME

Section XI, Division 1"

In addition, the evaluation should consider whether the proposed changes have affected

license commitments. A broad review of the licensing requirements and commitments may be
1.178-13

necessary because proposed ISI program changes could affect issues not explicitly stated in the

licensee’s FSAR or ISI program documentation. 

The Director of the Office of Nuclear Regulation is allowed by 10 CFR 50.55a to authorize

alternatives to the specific requirements of this regulation provided the proposed alternative will

ensure an acceptable level of quality and safety. Thus, alternatives to the acceptable RI-ISI

approaches presented in this guide may be proposed by licensees so long as supporting information

is provided that demonstrates that the key principles discussed in this guide are maintained. 

The licensee should include in its RI-ISI program submittal the necessary exemption

requests, technical specification amendment requests (if applicable), and relief requests necessary

to implement its RI-ISI program. 

NRC-endorsed ASME Code Cases that apply risk-informed ISI programs are consistent

with this regulatory guide in that they encourage the use of risk insights in the selection of

inspection locations and the use of appropriate and possibly enhanced inspection techniques that

are appropriate to the failure mechanisms that contribute most to risk.

2.1.2 Defense-in-Depth Evaluation

As stated in Regulatory Guide 1.174, the engineering analysis should evaluate whether the

impact of the proposed change in the ISI program (individually and cumulatively) is consistent

with the defense-in-depth philosophy. In this regard, the intent of this key principle is to ensure

that the philosophy of defense in depth is maintained, not to prevent changes in the way defense in

depth is achieved. The defense-in-depth philosophy has traditionally been applied in reactor

design and operation to provide multiple means to accomplish safety functions and prevent the

release of radioactive material. It has been and continues to be an effective way to account for

uncertainties in equipment and human performance. Where a comprehensive risk analysis can be

done, it can be used to help determine the appropriate extent of defense in depth (e.g., balance

among core damage prevention, containment failure, and consequence mitigation) to ensure

protection of public health and safety. Where a comprehensive risk analysis is not or cannot be

done, traditional defense-in-depth consideration should be used or maintained to account for

uncertainties. The evaluation should consider the intent of the general design criteria, national

standards, and engineering principles such as the single failure criterion. Further, the evaluation

should consider the impact of the proposed change on barriers (both preventive and mitigative) to

core damage, containment failure or bypass, and the balance among defense-in-depth attributes. 

The licensee should select the engineering analysis techniques, whether quantitative or qualitative,

appropriate to the proposed change (see Regulatory Guide 1.174 for additional guidance).

An important element of defense in depth for RI-ISI is maintaining the reliability of

independent barriers to fission product release. Class 1 piping (primary coolant system) is the

second boundary between the radioactive fuel and the general public. If a RI-ISI program

categorized, for example, all segments in the hot and cold legs of the primary system piping as low

safety significant and calculated that, with no inspections, the frequency of leaks would not
1.178-14

increase beyond existing performance history of the ASME Code, the staff would continue to

require some level of NDE inspection. 

2.1.3 Safety Margins

In engineering programs that affect public health and safety, safety margins are applied to

the design and operation of a system. These safety margins and accompanying engineering

assumptions are intended to account for uncertainties, but in some cases can lead to operational

and design constraints that are excessive and costly, or that could detract from safety (e.g., result in

unnecessary radiation exposure to plant personnel). Insufficient safety margins may require

additional attention. Prior to a request for relaxation of the existing requirements, the licensee

must ensure that the uncertainties are adequately addressed. The quantification of uncertainties

would likely require supporting sensitivity analyses. 

The engineering analyses should address whether the impacts of the changes proposed to

the ISI program are consistent with the key principle that adequate safety margins are maintained. 

The licensee is expected to select the method of engineering analysis appropriate for evaluating

whether sufficient safety margins would be maintained if the proposed change were implemented. 

An acceptable set of guidelines for making that assessment are summarized below. Other

equivalent decision criteria could also be found acceptable.

Sufficient safety margins are maintained when:

 Codes and standards (see Regulatory Position 2.1.1) or alternatives approved for use by the

NRC are met, and

 Safety analysis acceptance criteria in the licensing basis (e.g., updated FSAR, supporting

analyses) are met, or proposed revisions provide sufficient margins to account for analysis

and data uncertainty.

2.1.4 Piping Segments

A systematic approach should be applied when analyzing piping systems. One acceptable

approach is to divide or separate a piping system into segments; different criteria or definitions can

be applied to each piping segment. One acceptable method is to identify segments of piping

within the piping systems that have the same consequences of failure. Other methods could

subdivide a segment that exhibits a given consequence into segments with similar degradation

mechanisms or similar failure potential. The definition of a segment could encompass multiple

criteria, as long as a sound engineering and accounting record is maintained and can be applied to

an engineering analysis in a consistent and sound process. Consequences of failure may be

defined in terms of an initiating event, loss of a particular train, loss of a system, or combinations

thereof. The location of the piping in the plant, and whether inside or outside the containment or

compartment, should be taken into consideration when defining piping segments. 
1.178-15

The definition of a piping segment can vary with the methodology. Defining piping

segments can be an iterative process. In general, an analyst may need to modify the description of

the piping segments before they are finalized. This guide does not impose any specific definition

of a piping segment, but the analysis and the definition of a segment must be consistent and

technically sound. 

2.1.5 Assess Piping Failure Potential

The engineering analysis includes evaluating the failure potential of a piping segment. 

Determining the failure potential of piping segments, either with a quantitative estimate or by

categorization into groups, should be based on an understanding of degradation mechanisms,

operational characteristics, potential dynamic loads, flaw size, flaw distribution, inspection

parameters, experience data base, etc. The evaluation should state the appropriate definition of the

failure potential (e.g., failure on demand or operating failures associated with the piping, with the

basis for the definition) that will be needed to support the PRA or risk assessment. The failure

potential used in or in support of the analysis should be appropriate for the specific environmental

conditions, degradation mechanisms, and failure modes for each piping location. When data are

analyzed to develop a categorization process relating degradation mechanisms to failure potential,

the data should be appropriate and publicly available. When an elicitation of expert opinion is

used in conjunction with, or in lieu of, probabilistic fracture mechanics analysis or operating data,

a systematic process should be developed for conducting such an elicitation. In such cases, a

suitable team of experts should be selected and trained (NUREG/CR-5424 and NUREG-1563).

When implementing probabilistic fracture mechanics computer programs that estimate

structural reliability and are used in risk assessment of piping, or other analytic methods for

estimating the failure potential of a piping segment, some of the important parameters that need to

be assessed in the analysis include the identification of structural mechanics parameters,

degradation mechanisms, design limit considerations, operating practices and environment, and

the development of a data base or analytic methods for predicting the reliability of piping systems. 

Design and operational stress or strain limits are assessed. This information is available to the

licensee in the design information for the plant. The loading and resulting stresses or strains on the

piping are needed as input to the calculations that predict the failure probability of a piping

segment. The use of validated computer programs, with appropriate input, is strongly

recommended in a quantitative RI-ISI program because it may facilitate the regulatory evaluation

of a submittal. The analytic method should be validated with applicable plant and industry piping

performance data. 

To understand the impact of specific assumptions or models used to characterize the

potential for piping failure, appropriate sensitivity or uncertainty studies should be performed. 

These uncertainties include, but are not limited to, design versus fabrication differences, variations

in material properties and strengths, effects of various degradation and aging mechanisms,

variation in steady-state and transient loads, availability and accuracy of plant operating history,

availability of inspection and maintenance program data, applicability and size of the data base to

the specific degradation and piping, and the capabilities of analytic methods and models to predict
1.178-16

realistic results. Evaluation of these uncertainties provides insights to the input parameters that

affect the failure potential and, therefore, require careful consideration in the analysis. 

The methodology, process, and rationale used to determine the likelihood of failure of

piping segments should be independently reviewed during the final classification of the risk

significance of each segment. Referencing applicable generic topical reports approved by the

NRC is one acceptable means to standardize the process. When new computer codes are used to

develop quantitative estimates, the techniques should be verified and validated against established

industry codes and available data. When data are used to evaluate the likelihood of piping failures,

the data should be submitted to the NRC or referenced by an NRC-approved topical report. As

stated in Regulatory Guide 1.174, "data, methods, and assessment criteria used to support

regulatory decisionmaking must be scrutable and available for public review." It is the

responsibility of the licensee to provide the data, methods, and justification to support its

estimation of the failure potential of piping segments. 

2.1.6 Assess Consequences of Piping Segment Failures

When evaluating the risk from piping failures, the analyst needs to evaluate the potential

consequences, or failures, that a piping failure can initiate. This can be accomplished by

performing a detailed walkdown of a nuclear power facility’s piping network. The consequences

of the postulated pipe segment failure include direct and indirect effects of the failure. Direct

effects include the loss of a train or system and associated possible diversion of flow or an

initiating event such as a loss of coolant accident (LOCA) or both. Indirect effects include the

spatial effects of flood, spray, pipe whip, or jet impingement that may affect adjacent SSCs or

depletion of water sources and loss of associated systems. 

2.2 Probabilistic Risk Assessment

In accordance with the Commission’s Policy Statement on PRA, the risk-informed

application process is intended not only to support reduction in the number of inspections, but also

to identify areas where increased resources should be allocated to enhance safety. Therefore, an

acceptable RI-ISI process should not focus exclusively on areas in which reduced inspection could

be justified. This section addresses ISI-specific considerations in the PRA to support relaxation of

inspections, enhancement of inspections, and validation of component operability. 

ASME has published a PRA standard that addresses a Level 1 and limited Level 2 PRA for

full-power operation for internal events (excluding internal fire) (ASME RA-S-2002). Other

standards for external events (i.e., seismic, wind and flood), low power and shutdown conditions

and internal fires are under development by ANS.

The NRC staff is currently developing a regulatory guide to provide guidance to licensees

on determining the technical adequacy of a PRA used in a risk-informed integrated decision

making process, and to endorse standards and industry guidance (see Draft Regulatory Guide DG1122). The NRC staff is continuing to evaluate PRAs submitted in support of specific applications
1.178-17

using the guidelines given in Section 2.2.3 and Section 2.5 of Regulatory Guide 1.174 and in SRP

Chapter 19 of the Standard Review Plan.

The PRA can be used to categorize the piping segments into safety-significant and lowsafety-significant classifications (or more classifications, if a finer graded approach is desired) and

to confirm that the change in risk caused by the change in the ISI program is in accordance with

the guidance of Regulatory Guide 1.174. The licensee’s submittal should discuss measures used to

ensure adequate quality, such as a report of a peer review, when performed, that addresses the

appropriateness of the PRA model for supporting a risk assessment of the change under

consideration. The submittal should address any limitations of the analysis that are expected to

impact the conclusion regarding the acceptability of the proposed change. The licensee’s

resolution of the findings of the peer review, certification, or cross comparison, when performed,

should also be submitted. This response could indicate whether the PRA was modified or justify

why no change to the PRA was necessary to support decisionmaking for the change under

consideration. 

2.2.1 Modeling Piping Failures in a PRA 

Input from the traditional engineering analysis addressed in Regulatory Position 2.1

includes identification of piping segments from the point of view of the failure potential

(degradation mechanisms) and consequences (resulting failure modes and consequential primary

and secondary effects). The traditional analysis identifies both the primary and secondary effects

that can result from a piping failure. The assessment of the primary and secondary failures

identifies the portions of the PRA that are affected by the piping failure. 

Each pipe segment failure may have one of three types of impacts on the plant. 

1. Initiating event failures are when the failure directly causes a transient and may or may not

also fail one or more plant trains or systems. 

2. Standby failures are those failures that cause the loss of a train or system but which do not

directly cause a transient. Standby failures are characterized by train or system

unavailability that may require shutdown because of the technical specifications or limiting

conditions for operation. 

3. Demand failures are failures accompanying a demand for a train or system and are usually

caused by the transient-induced loads on the segment during system startup. 

The impact of the pipe segment failure on risk should be evaluated with the PRA. 

Evaluation may involve a quantitative estimate derived from the PRA, a systematic technique to

categorize the consequence of the pipe failure on risk, or some combination of quantification and

categorization. If a segment failure were to lead to plant transients and equipment failures that are

not at all represented in the PRA (a new and specific initiating event, for example), the evaluation

process should be expanded to assess these events. 
1.178-18

PRAs normally do not include events that represent failure of individual piping segments

nor the structural elements within the segments. A quantitative estimate of the impact of segment

failures can be done by modifying the PRA logic to systematically and explicitly include the

impact of the individual pipe segment failures. The impact of each segment’s failure on risk can

also be estimated without modifying the PRA’s logic by identifying an initiating event, basic

event, or group of events, already modeled in the PRA, whose failures capture the effects of the

piping segment’s failure (referred to as the surrogate approach). In either case, to assess the

impact of a particular segment failure, the analyst sets the appropriate events to a failed state in the

PRA and requantifies the PRA or the appropriate parts of the PRA as needed. The analysis should

appropriately incorporate segment failures that only cause an initiating event, that only degrade or

fail a mitigating system required to respond to an independent initiating event, and that

simultaneously cause an initiating event and degrade or fail a mitigating system responding to the

initiating event. The requantification should explicitly address truncation errors, since cut set or

truncated sequences may not fully capture the impact of multiple failure events.

If a systematic technique is used to categorize the consequence of pipe failures, it should

also be based on PRA results. In this case, however, the categories may be represented by ranges

of conditional results, and instead of quantifying the impact of each segment failure, the process

should provide for determining the range within which each segment's failure would lie. In

general, the consequences would range from high, for those segments whose failure would have a

high likelihood of leading to core damage or large early release, to low for those segments whose

failure would likely not lead to core damage or large early release. The licensee should provide a

discussion and justification of the ranges selected. The use of ranges instead of individual results

estimates may require fewer calculations, but the categorization process and decision criteria

should be justified, well defined, and repeatable. 

2.2.1.1 Dependencies and Common Cause Failures

The effects of dependencies and common cause failures (CCFs) for ISI components need to

be considered carefully because of the significance they can have on CDF. Generally, data are

insufficient to produce plant-specific estimates based solely on plant-specific data. For CCFs, data

from generic sources may be required. 

2.2.1.2 Human Reliability Analyses To Isolate Piping Breaks

For ISI-specific analyses, the human reliability analysis methodology used in the PRA must

account for the impact that the piping segment break would have on the operator's ability to

respond to the event. In addition, the reliability of the inspection program (including both operator

and equipment qualification), which factors into the probability of detection, should also be

addressed. 

2.2.2 Use of PRA for Categorizing Piping Segments

When the impact of each segment’s failure on plant risk metrics has been determined, the

safety significance of the segments is developed. The method of categorizing a piping segment

can vary. For example, if the pipe failure event frequency or probability is estimated and the

events are incorporated into the PRA logic model, importance measure calculations and the
1.178-19

determination of safety significance, as discussed in Regulatory Guide 1.174 and SRP Chapter 19,

may be performed. Alternatively, if a conditional core damage frequency (CCDF), conditional

large early release frequency (CLERF), conditional core damage probability (CCDP), or

conditional large early release probability (CLERP) (depending on the impact the segment failure

has on the plant) is estimated for each segment from the PRA, a CDF and LERF caused only by

pipe failures may be developed by combining the conditional consequences and segment failure

probabilities or frequencies external to the PRA logic model. Importance measures can also be

developed using these results and these measures compared to appropriate threshold criteria to

support the determination of the safety significance of each segment. The calculations used in

such a process should yield well-defined estimates of CDF, LERF, and importance measures. The

licensee should provide a discussion of and justification for the threshold criteria used. 

As discussed in Regulatory Position 2.2.1, the consequence of segment failures may be

represented by categories of consequences instead of quantitative estimates for each segment. In

this case, the potential for pipe failure as discussed in Regulatory Position 2.1.5 would also be

developed as categories ranging from high to low depending on the degradation mechanisms

present and the corresponding likelihood that the segment will fail. These consequence and failure

likelihood categories should be systematically combined to develop categories of safety

significance. The licensee should provide a discussion and justification relating the consequence

and failure likelihood categories to the safety-significant category assigned to each combination. 

The safety-significance category of the pipe segment will help determine the level of

inspection effort devoted to the segment. In general, safety-significant segments will receive more

inspections and more demanding inspections than low safety-significant segments. In any

integrated categorization process, the principles in Regulatory Guide 1.174 need to be addressed. 

Irrespective of the method used in the analysis, the licensee needs to justify the final categorization

process as being robust and reasonable with respect to the analysis uncertainties.

2.2.3 Demonstrate Change in Risk Resulting from Change in ISI Program

Any change in the ISI program has an associated risk impact. Evaluation of the change in

risk may be a detailed calculation or it may be a bounding estimate supported by sensitivity studies

as appropriate. The change may be a risk increase, a risk decrease, or risk neutral. The change is

evaluated and compared with the guidelines presented in Regulatory Guide 1.174. The staff

expects that a RI-ISI program would lead to both risk reduction and reduction in radiation

exposure to plant personnel. 

The change in risk estimate should appropriately account for the change in the number of

elements inspected and the effects of enhanced inspection. The methods used to determine the

piping failure potential, the piping failure consequence, and the impact of the change in the

number of inspections should together provide confidence that any increase in CDF or risk is small

and acceptable in accordance with Regulatory Guide 1.174 guidelines and consistent with the

intent of the Commission’s Safety Goal Policy Statement.
1.178-20

2.3 Integrated Decisionmaking

Regulatory Positions 2.1 and 2.2 address the elements of traditional analysis and PRA

analysis of a RI-ISI program. These elements are part of an integrated decisionmaking process that

assesses the acceptability of the program. The key principles of Regulatory Guide 1.174 are

systematically addressed. Technical and operations personnel at the plant review the information

and render a finding of the safety-significance category for each piping segment under review. 

Detailed guidelines for the categorization of piping segments should be developed and discussed

with the group responsible for the determination (typically performed by the plant’s expert panel).

The method for selecting the number of piping elements to be inspected should be justified. 

3. ELEMENT 3: IMPLEMENTATION, PERFORMANCE MONITORING, AND

CORRECTIVE ACTION STRATEGIES

Integrating the information obtained from Elements 1 and 2 of the RI-ISI process (as

described in Regulatory Positions 1 and 2 of this guide), the licensee develops proposed RI-ISI

implementation, performance monitoring, and corrective action strategies. The RI-ISI program

should identify piping segments whose inspection strategy (i.e., frequency, number of inspections,

methods, or all three) should be increased as well as piping segments whose inspection strategies

might be relaxed. The number of required inspections should be a product of the systematic

application of the risk-informed process. The program should be self-correcting as experience

dictates. The program should contain performance measures used to confirm the safety insights

gained from the risk analyses.

3.1 Program Implementation

A licensee should have in place a schedule for inspecting all segments categorized as safety

significant in its RI-ISI program. This schedule should include inspection strategies and inspection

frequencies, inspection methods, the sampling program (the number of elements/areas to be

inspected, the acceptance criteria, etc.) for the safety-significant piping that is within the scope of

the ISI program, including piping segments identified as safety significant that are not currently in

the ISI program.

The analysis for a RI-ISI program will, in most cases, confirm the appropriateness of the

inspection interval and scope requirements of the ASME Boiler and Pressure Vessel Code

(B&PVC) Section XI Edition and Addenda committed to by a licensee in accordance with 10 CFR

50.55a. The requirements for these intervals are contained in Section XI of the B&PVC. 

However, should active degradation mechanisms surface, the inspection interval would be

modified as appropriate. Updates to the RI-ISI program should be performed at least periodically

to coincide with the inspection program requirements contained in Section XI under Inspection

Program B. The RI-ISI program should be evaluated periodically as new information becomes

available that could impact the ISI program. For example, if changes to the PRA impact the

decisions made for the RI-ISI program, if plant design and operations change such that they impact

the RI-ISI program, if inspection results identify unexpected flaws, or if replacement activities
1.178-21

impact the failure potential of piping, the effects of the new information should be assessed. The

periodic evaluation may result in updates to the RI-ISI program that are more restrictive than

required by Section XI. As plant design feature changes are implemented, changes to the input

associated with the RI-ISI program segment definition and element selections should be reviewed

and modified as needed. Changes to piping performance, the plant procedures that can affect

system operating parameters, piping inspection, component and valve lineups, equipment

operating modes, or the ability of the plant personnel to perform actions associated with accident

mitigation should be reviewed in any RI-ISI program update. Leakage and flaws identified during

scheduled inspections should be evaluated as part of the RI-ISI update. 

Piping segments categorized as safety significant that are not in the licensee’s current ISI

program should (wherever appropriate and practical) be inspected in accordance with applicable

ASME Code Cases (or revised ASME Code), including compliance with all administrative

requirements. Where ASME Section XI inspection is not practical or appropriate, or does not

conform to the key principles identified in this document, alternative inspection intervals, scope,

and methods should be developed by the licensee to ensure piping integrity and to detect piping

degradation. A summary of the piping segments and their proposed inspection intervals and scope

should be provided to the NRC prior to implementation of the RI-ISI program at the plant. 

For piping segments categorized as safety significant that were the subject of a previous

NRC-approved relief request or were exempt under existing Section XI criteria, the licensee

should assess the appropriateness of the relief or exemption in light of the risk significance of the

piping segment. 

3.2 Performance Monitoring

3.2.1 Periodic Updates

The RI-ISI program should be updated at least on the basis of periods that coincide with the

inspection program requirements contained in Section XI under Inspection Program B. These

updates should be performed more frequently if dictated by any plant procedures to update the

PRA (which may be more restrictive than a Section XI period type update) or as new degradation

mechanisms are identified. 

3.2.2 Changes to Plant Design Features

As changes to plant design are implemented, changes to the inputs associated with RI-ISI

program segment definition and element selections may occur. It is important to address these

changes to the inputs used in any assessment that may affect resultant pipe failure potentials used

to support the RI-ISI segment definition and element selection. Some examples of these inputs

would include:

 Operating characteristics (e.g., changes in water chemistry control)

 Material and configuration changes

 Welding techniques and procedures
1.178-22

 Construction and preservice examination results

 Stress data (operating modes, pressure, and temperature changes)

In addition, plant design changes could result in significant changes to a plant’s CDF or

LERF, which in turn could result in a change in consequence of failure for system piping

segments.

3.2.3 Changes to Plant Procedures

Changes to plant procedures that affect ISI, such as system operating parameters, test

intervals, or the ability of plant operations personnel to perform actions associated with accident

mitigation, should be included for review in any RI-ISI program update. Additionally, changes in

those procedures that affect component inspection intervals, valve lineups, or operational modes of

equipment should also be assessed for their impact on changes in postulated failure mechanism

initiation or CDF/LERF contribution. 

3.2.4 Equipment Performance Changes

Equipment performance changes should be reviewed with system engineers and

maintenance personnel to ensure that changes in performance parameters such as valve leakage,

increased pump testing, or identification of vibration problems is included in the periodic

evaluation of the RI-ISI program update. Specific attention should be paid to these conditions if

they were not previously assessed in the qualitative inputs to the element selections of the RI-ISI

program.

3.2.5 Examination Results

When scheduled RI-ISI program NDE examinations, pressure tests, and corresponding

VT-2 visual examinations for leakage have been completed, and if unacceptable flaws, evidence of

service related degradation, or indications of leakage have been identified, the existence of these

conditions should be evaluated. This update of the RI-ISI program should follow the applicable

elements of Appendix B to 10 CFR Part 50 to determine the adequacy of the scope of the

inspection program.

3.2.6 Information on Individual Plant and Industry Failures

Review of individual plant maintenance activities associated with repairs or replacements,

including identified flaw evaluations, is an important part of any periodic update, regardless of

whether the activity is the result of a RI-ISI program examination. Evaluating this information as

it relates to a licensee’s plant provides failure information and trending information that may have

a profound effect on the element locations currently being examined under a RI-ISI program. 

Industry failure data is just as important to the overall program as the owner’s information. During

the periodic update, industry data bases (including available international data bases) should be

reviewed for applicability to the owner’s plant. 
1.178-23

3.3 Corrective Action Programs

Each licensee of a nuclear power plant is responsible for having a corrective action

program, consistent with Regulatory Guide 1.174. Measures are to be established to ensure that

conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective

material and equipment, and nonconformances, are promptly identified and corrected. In the case

of significant conditions adverse to quality, the measures must ensure that the cause of the

condition is determined and corrective action is taken to preclude repetition. The identification of

the significant condition adverse to quality, the cause of the condition, and the corrective action are

to be documented and reported to appropriate levels of management.

For Code piping categorized as safety significant, this corrective action program should be

consistent with applicable Section XI provisions. For non-Code and Code-exempt piping

categorized as safety significant, appropriate Section XI provisions should also be used, or the

licensee should submit an alternative program based on the risk significance of the piping.

3.4 Acceptance Guidelines

These acceptance guidelines are for the implementation, monitoring, and corrective action

programs for the accepted RI-ISI program plan. 

1. The evaluation of the implementation program will be based on the attributes presented in

Regulatory Positions 3.1 through 3.3 of this Regulatory Guide 1.178.

2. The corrective action program should provide reasonable assurance that a nonconforming

component will be brought back into conformance in a timely fashion. The corrective

actions required in ASME Section XI should continue to be followed.

3. Evaluations within the corrective action program may also include:

 Ensuring that the root cause of the condition is determined and that

corrective actions are taken to preclude repetition. The identification of

the significant condition adverse to quality, the cause of the condition, and

the corrective action are to be documented and reported to appropriate

levels of management.

 Determining the impact of the failure or nonconformance on system or

train operability since the previous inspection.

 Assessing the applicability of the failure or nonconforming condition to

other components in the RI-ISI program.

 Correcting other susceptible RI-ISI components as necessary.
1.178-24

 Incorporating the lessons in the plant data base and computer models, if

appropriate.

 Assessing the validity of the failure rate and unavailability assumptions

that can result from piping failures used in the PRA or in support of the

PRA, and

 Considering the effectiveness of the component’s inspection strategy in

detecting the failure or nonconforming condition. The inspection interval

would be reduced or the inspection methods adjusted, as appropriate,

when the component (or group of components) experiences repeated

failures or nonconforming conditions.

4. The corrective action evaluation should be provided to the licensee’s PRA and RI-ISI

groups so that any necessary model changes and regrouping are done, as appropriate.

5. The RI-ISI program documents should be revised to document any RI-ISI program changes

resulting from the corrective actions taken.

6. A program is in place that monitors industry findings.

7. Piping is subject to examination. The examination requirements include all piping

evaluated by the risk-informed process and categorized as safety significant.

8. The inspection program is to be completed during each ten-year inspection interval with

the following exceptions.

8.1 If, during the interval, a re-evaluation using the RI-ISI process is

conducted and scheduled items are no longer required to be examined,

these items may be eliminated.

8.2 If, during the interval, a re-evaluation using the RI-ISI process is

conducted and items must be added to the examination program, those

items will be added.

9. If additional examinations are needed following the identification of unacceptable flaws,

additional examinations will be performed on the elements with the same root cause or

degradation mechanisms as the identified flaw or relevant condition. The number of

additional examinations should be equivalent to the number of elements required to be

inspected during the current outage. If unacceptable flaws or relevant conditions are again

found similar to the initial problem, the remaining elements identified as susceptible will

be examined. All additional examinations should be performed during the current outage.

10. Examination and Pressure Test Requirements. Pressure testing and VT-2 visual

examinations are to be performed on Class 1, 2, and 3 piping systems in accordance with
1.178-25

Section XI, as specified in the licensee’s ISI program. The pressure testing and VT-2

examinations are also to be performed on non-Code safety-significant piping. The nonCode safety-significant piping will be treated as ASME Code Class piping for the purposes

of examination and pressure testing.

Examination methods, equipment qualification, personnel qualification, and procedure

qualification are to be in accordance with the edition and addenda endorsed by the NRC

through 10 CFR 50.55a, "Codes and Standards."

11. Acceptance standards for identified flaws and repair or replacement activities are to be

performed in accordance with the B&PVC Section XI requirements.

12. Records and reports should be prepared and maintained in accordance with the B&PVC

Section XI Edition and Addenda as specified in the licensee’s ISI program.

4. ELEMENT 4: DOCUMENTATION

The recommended contents for a plant-specific risk-informed ISI submittal are presented

here. This guidance will help ensure the completeness of the information provided and aid in

minimizing the time needed for the review process. 

4.1 Documentation that Should Be Included in a Licensee’s RI-ISI Submittal

References to NRC-approved generic topical reports that address the methodology and

issues requested in a submittal are acceptable. Documentation guidelines specified in approved

topical reports may be used instead of the following guidelines when the methodology from an

approved topical report is used. Since topical reports could cover more issues than applied by a

licensee or the licensee may elect to deviate from the full body of issues addressed in the topical

report, such distinctions should be clearly stated.

The following items should be included in the application to implement a RI-ISI program.

 A request to implement a RI-ISI program as an authorized alternative to the current NRC

endorsed ASME Code pursuant to 10 CFR 50.55a(a)(3)(i). The licensee should also

provide a description of how the proposed change impacts any commitments made to the

NRC. 

 Discussions on each of the following five key principles of risk-informed regulations (see

Section 2 of Regulatory Guide 1.174 for more details).

1. The proposed change meets the current regulations unless it is explicitly related to

an alternative requested under 10 CFR 50.55a(a)(3)(i), a requested exemption, or a

rule change.
1.178-26

2. The proposed change is consistent with the defense-in-depth philosophy (see

detailed discussions in Section 2.2.1.1 of Regulatory Guide 1.174). 

3. The proposed change maintains sufficient safety margins (see detailed discussions

in Section 2.2.1.2 in Regulatory Guide 1.174). 

4. When proposed changes result in an increase in CDF and/or risk, the increases

should be small and consistent with the guidance in Regulatory Guide 1.174.

5. The impact of the proposed change should be monitored using performance

measurement strategies.

 Identification of the aspects of the plant’s current requirements that would be affected by

the proposed RI-ISI program. This identification should include all commitments and

augmented programs (for example, the IGSCC inspections and other commitments arising

from generic letters affecting piping integrity) that the licensee intends to change or

terminate as part of the RI-ISI program. The application of the RI-ISI methodology to

incorporate and change the augmented program should be justified. 

 Identification of the specific revisions to existing inspection schedules, locations, and

methods that would result from implementation of the proposed program.

 Plant procedures or documentation containing the guidelines for all phases of evaluating

and implementing a change in the ISI program based on probabilistic and traditional

insights. These should include a description of the integrated decisionmaking process and

criteria used for categorizing the safety significance of piping segments, a description of

how the integrated decisionmaking was performed, a description and justification of the

number of elements to be inspected in a piping segment, the qualifications of the

individuals making the decisions, and the guidelines for making those decisions.

 The results of the licensee’s ISI-specific analyses used to support the program change with

enough detail to be clearly understandable to the reviewers of the program. These results

should include the following information. 

 A list of the piping systems reviewed.

 A list of each segment, including the number of welds, weld type, and properties of

the welding material and base metal, the failure potential, CDF, CCDF/CCDP,

LERF, CLERF, importance measure results (risk achievement worth (RAW),

Fussel-Vesely (F-V), etc.) and justification of the associated threshold values,

degradation mechanism, test and inspection intervals used in or in support of the

PRA, etc. Results from other methods used to develop the consequences and

categorization of each segment (or weld) should be documented in a similar level of

detail.
3

 In April 2000, the Nuclear Energy Institute submitted a process (Letter to S.J. Collins, NRC) for peer review of licensee PRAs. 

It was submitted for staff review in the context of its use in categorizing SSCs with respect to special treatment requirements (i.e.,

supporting NRC’s risk-informed proposed rulemaking to add new section 10 CFR 50.69, "Risk-Informed Categorization and

Treatment of Structures, Systems, and Components""Option 2" work (SECY-02-0176)). This process, when endorsed by the

NRC, may also be of use in making licensing basis changes (as well as other regulatory activities not addressed here); if so,

future revisions of this regulatory guide may endorse this certification process for this purpose.

1.178-27

 The degradation mechanisms for each segment (if segments contain welds exposed

to different degradation mechanism, for each weld) used to develop the failure

potential of each segment. For the selected limiting locations, provide examples of

the failure mode, failure potential, failure mechanism, weld type, weld location, and

properties of the welding material and base metal. 

 A detailed description and justification for the number of elements to be inspected.

 Equipment assumed to fail as a direct or indirect consequence of each segment’s

failure (if segments contain welds with different failure consequences, for each

weld).

 A description of how the impact of the change between the current Section XI and

the proposed RI-ISI programs is evaluated or bounded, and how this impact

compares with the risk guidelines in Section 2.2.2.2 of Regulatory Guide 1.174.

 The means by which failure probabilities, frequencies, or potential were determined. 

 A description of the PRA used for the categorization process and for the determination of

risk impact, in terms of the process to ensure quality, scope, and level of detail, and how

limitations in quality, scope, and level of detail are compensated for in the integrated

decisionmaking process supporting the ISI submittal. At a minimum, the submittal should

include the following information.

 The CDF and LERF estimates and the version, calculation, or other reference

number that identifies which version of the PRA was used.

 A description of the process used to up-date the PRA to ensure that the PRA

analyses adequately represent the current design, construction, operational

practices, and operational experience of the plant and its operator.

 A description of the staff and industry reviews performed on the PRA.3

Limitations, weakness, or improvements identified by the reviewers that could

change the results of the PRA should be discussed. The resolution of the reviewer

comments, or an explanation of the insensitivity of the analysis used to support the

submittal to the comment, should be provided.

 If the submittal includes modified inspection intervals, the methodology and results of the

analysis should be submitted.
1.178-28

 A description of the implementation, performance monitoring, and corrective action

strategies and programs in sufficient detail for the staff to understand the new ISI program

and its implications.

 Applicable documentation discussed under the cumulative risk documentation for

submittal in Section 1.3 of Regulatory Guide 1.174.

 Reference to NRC-approved topical reports on implementing a RI-ISI and supporting

documents. Variations from the topical reports and supporting documents should be

clearly identified.

4.2 Documentation that Should Be Available Onsite for Inspection

The licensee should maintain at its facility the technical and administrative records used in

support of its submittal or should be able to generate the information on request. This information

should be available for NRC review and audit. If changes are planned to the ISI program based on

internal procedures and without prior NRC approval, the following information should also be

placed in the plant’s document control system so that the analyses for any given change can be

identified and reviewed. The record should include, but not be limited to, the following

information:

 All the documentation discussed in 4.1. Although the documentation requirements in a

submittal may be reduced when referring to NRC-approved topical reports, all the

documentation included under 4.1 should be available for onsite inspection.

 Plant and applicable industry data used in support of the RI-ISI program. All analyses and

assumptions used in support of the RI-ISI program and communications with outside

organizations supporting the RI-ISI program (e.g., use of peer and independent reviews,

use of expert contractors).

 Detailed procedures and analyses performed by an expert panel, or other technical groups,

if relied upon for the RI-ISI program, including a record of deliberations,

recommendations, and findings.

 Documentation of the plant’s baseline PRA used to support the ISI submittal should be of

sufficient detail to allow an independent reviewer to ascertain whether the PRA reflects the

current plant configuration and operational practices commensurate with the role the PRA

results play in the integrated decisionmaking process. In addition to documentation on the

PRA itself, analyses performed in support of the ISI submittal should be documented in a

manner consistent with the baseline documentation. Such analyses may include:

 The process used to identify initiating events developed in support of the RI-ISI

submittal and the results from the process.

 Any event and fault trees developed during the RI-ISI submittal preparation.
1.178-29

 Documentation of the methods and techniques used to identify and quantify the

impact of pipe failures using the PRA, or in support of the PRA, if different from

those used during the development of the baseline PRA.

 The techniques used to identify and quantify human actions. 

 The data used in any uncertainty calculations or sensitivity calculations, consistent

with the guidance provided in Regulatory Guide 1.174. 

 How uncertainty was accounted for in the segment categorization, as well as the

sensitivity studies performed to ensure the robustness of the categorization.

 Detailed results of the inspection program corresponding to the ISI inspection records

described in the implementation, performance monitoring, and corrective action program

accompanying the RI-ISI submittal.

 For each piping segment, information on weld type, weld location, and properties of

welding material and base metal.

 For each piping segment, information regarding the process and assumptions used to

develop failure mode and failure potential (frequency/probability), in addition to the

identification of the failure mechanism.
1

 Copies may be obtained from the American Society of Mechanical Engineers, Three Park Avenue, New York, NY 10016.

2

 Requests for single copies of draft or active regulatory guides (which may be reproduced) or for placement on an automatic

distribution list for single copies of future draft guides in specific divisions should be made in writing to the U.S. Nuclear

Regulatory Commission, Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to

(301)415-2289; email <DISTRIBUTION@NRC.GOV>. Copies are available for inspection or copying for a fee from the NRC

Public Document Room at 11555 Rockville Pike (first floor), Rockville, MD; the PDR’s mailing address is USNRC PDR,

Washington, DC 20555; telephone (301)415-4737 or 1-(800)397-4209; fax (301)415-3548; e-mail <PDR@NRC.GOV>. Copies

of many regulatory guides are available on NRC’s web site, <WWW.NRC.GOV>.

3

 Copies may be obtained from the EPRI Distribution Center, 207 Coggins Drive, P.O. Box 23205, Pleasant Hill, CA 94523.

4

 Electronic copies are available on the NRC web site at <WWW.NRC.GOV> in the Document Collections under Generic

Communications. Copies are available for inspection or copying for a fee from the NRC Public Document Room at 11555

Rockville Pike (first floor), Rockville, MD; the PDR’s mailing address is USNRC PDR, Washington, DC 20555; telephone

(301)415-4737 or 1-(800)397-4209; fax (301)415-3548; e-mail <PDR@NRC.GOV>.

1.178-30

REFERENCES

ASME B&PVC, "Rules for Inservice Inspection of Nuclear Power Plant Components," American

Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section XI, 1989 Edition.1

ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant

Applications," American Society of Mechanical Engineers, April 5, 2002.1

Draft Regulatory Guide DG-1122, “An Approach for Determining the Technical Adequacy of

Probabilistic Risk Assessment Results for Risk-Informed Activities,” USNRC, November 2002.2

Draft SRP Chapter 19.1, “An Approach for Determining the Technical Adequacy of Probabilistic

Risk Assessment Results for Risk-Informed Activities,” USNRC, November 2002.2

EPRI TR-112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," Electric

Power Research Institute, Revision B-A, December 1999.3

Generic Letter 89-13, “Service Water System Problems Affecting Safety-Related Equipment,”

USNRC, July 18, 1989.4

Generic Letter 89-08, “Erosion/Corrosion-Induced Pipe Wall Thinning,” USNRC, May 2, 1989.4

Generic Letter 88-01, “NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping,”

USNRC, January 25, 1988.4

IE Bulletin 79-17, “Pipe Cracks in Stagnant Borated Water Systems at PWR Plants,” USNRC,

October 29, 1979.4

Letter to S.J. Collins, NRC, from Ralph E. Beedle, NEI, with attached "Probabilistic Risk Analysis

(PRA) Peer Review Guidance," Rev. A3, NEI 00-02, Prepared for NEI Risk-Based Applications
5

 Copies are available for inspection or copying for a fee from the NRC Public Document Room at 11555 Rockville Pike (first

floor), Rockville, MD; the PDR’s mailing address is USNRC PDR, Washington, DC 20555; telephone (301)415-4737 or 1-

(800)397-4209; fax (301)415-3548; e-mail <PDR@NRC.GOV>.

6

 Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328

(telephone (202)512-1800); or from the National Technical Information Service by writing NTIS at 5285 Port Royal Road,

Springfield, VA 22161; <http://www.ntis.gov/ordernow>; telephone (703)487-4650. Copies are available for inspection or

copying for a fee from the NRC Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR’s mailing address is

USNRC PDR, Washington, DC 20555; telephone (301)415-4737 or (800)397-4209; fax (301)415-3548; email is

PDR@NRC.GOV. Certain NUREG documents are available electronically at <WWW.NRC.GOV> under Document

Collections.

1.178-31

Task Force by WOG/Westinghouse Electric Co., and B&WOG/Framatome Technologies, Inc.,

April 24, 2000.5

NRC Bulletin No. 88-11, “Pressurizer Surge Line Thermal Stratification,” USNRC, December 20,

1988.4

NRC Bulletin No. 88-08, “Thermal Stresses in Piping Connected to Reactor Coolant Systems,”

USNRC, June 22, 1988.4

NRC Information Notice 93-20, “Thermal Fatigue Cracking of Feedwater Piping to Steam

Generators,” USNRC, March 24, 1993.4

NUREG-1563, J.P. Kotra et al., “Branch Technical Position on the Use of Expert Elicitation in the

High-Level Radioactive Waste Program,” USNRC, November 1996.6

NUREG/CR-5424, M.A. Meyer and J.A. Booker, "Eliciting and Analyzing Expert Judgment,"

(Prepared for the NRC by Los Alamos National Laboratory), USNRC, January 1990.6

Policy Statement, "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory

Activities; Final Policy Statement," Federal Register, Vol. 60, p 42622, USNRC, August 16,

1995.

Regulatory Guide 1.84, "Design, Fabrication, and Materials Code Case Acceptability, ASME

Section III," USNRC, Regulatory Guide 1.84, Revision 32, June 2003.2

Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI,

Division 1," USNRC, Regulatory Guide 1.147, Revision 13, June 2003.2

Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed

Decisions on Plant-Specific Changes to the Current Licensing Basis," USNRC, Regulatory Guide

1.174, Revision 1, November 2002.2

Regulatory Guide 1.175, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Inservice Testing," USNRC, August 1998.2

Regulatory Guide 1.176, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Graded Quality Assurance," USNRC, August 1998.2
7

 Electronic copies are available at <WWW.NRC.GOV> in the Document Collections under Commission papers. Copies are

available for inspection or copying for a fee from the NRC Public Document Room at 11555 Rockville Pike (first floor),

Rockville, MD; the PDR’s mailing address is USNRC PDR, Washington, DC 20555; telephone (301)415-4737 or 1-(800)397-

4209; fax (301)415-3548; e-mail <PDR@NRC.GOV>.

1.178-32

Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," USNRC, August 1998.2

Safety Goal Policy Statement, “Safety Goals for the Operations of Nuclear Power Plants; Policy

Statement,” USNRC, Federal Register, Vol. 51, p. 30028 (51 FR 30028), August 4, 1986.

SECY-02-0176, Proposed Rulemaking To Add New Section 10 CFR 50.69, "Risk-Informed

Categorization and Treatment of Structures, Systems, and Components," September 30, 2002.7

SECY-00-0162, “Addressing PRA Quality in Risk-Informed Activities,” USNRC, July 28, 2000.7 

SECY-95-280, "Framework for Applying Probabilistic Risk Analysis in Reactor Regulation,"

USNRC, November 27, 1995.7

SRP Chapter 3.9.7, "Standard Review Plan for Risk-Informed Decision Making: Inservice

Testing," USNRC, NUREG-0800, Chapter 3.9.7, Revision 1, April 2003.2

SRP Chapter 3.9.8, "Standard Review Plan for the Review of Risk-Informed Inservice Inspection

of Piping," USNRC, NUREG-0800, Chapter 3.9.8, September 2003.2

SRP Chapter 16.1, "Standard Review Plan for Risk-Informed Decision Making: Technical

Specifications," USNRC, NUREG-0800, Chapter 16.1, August 1998.2

SRP Chapter 19, "Standard Review Plan for Risk-Informed Decision Making," USNRC, 

NUREG-0800, Chapter 19, Revision 1, November 2002.2

WCAP-14572, "Westinghouse Owners Group Application of Risk Informed Methods to Piping

Inservice Inspection Topical Report," WCAP-14572, Revision 1, NP-A, Westinghouse Energy

Systems, February 1999.

 
1.178-33

REGULATORY ANALYSIS

A separate regulatory analysis was not prepared for this regulatory guide. The regulatory

analysis prepared for Draft Regulatory Guide DG-1063, October 1997, provides the regulatory

basis for this regulatory guide as well. DG-1063 was issued for public comment as the draft of this

regulatory guide. A copy of the regulatory analysis is available for inspection and copying for a

fee at the U.S. Nuclear Regulatory Commission Public Document Room, 11555 Rockville Pike,

Rockville, MD; the PDR’s mailing address is USNRC PDR, Washington, DC 20555; telephone

(301)415-4737 or 1-(800)397-4209; fax (301)415-3548; e-mail <PDR@NRC.GOV>.