Regulatory Guide 1.178: Difference between revisions

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{{Adams
{{Adams
| number = ML032510128
| number = ML003740181
| issue date = 09/30/2003
| issue date = 09/30/1998
| title = An Approach for Plant-Specific Risk-Informed Decisionmaking for Inservice Inspection of Piping
| title = (Draft Was Issued as DG-1063), an Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping
| author name =  
| author name =  
| author affiliation = NRC/RES
| author affiliation = NRC/RES
Line 10: Line 10:
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = RG-1.178, Rev 1
| document report number = RG-1.178
| document type = Regulatory Guide
| document type = Regulatory Guide
| page count = 33
| page count = 24
}}
}}
{{#Wiki_filter: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 NRCs  regulations, techniques used by the staff in evaluating specific problems or postulated accidents, and data needed by the NRC
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION
staff in its review of applications for permits and licenses.  Regulatory guides are not substitutes for regulations, and compliance with them is not required.
REGULATORY GUIDE
OFFICE OF NUCLEAR REGULATORY RESEARCH
FOR TRIAL USE
REGULATORY GUIDE 1.178 (Draft was Issued as DG-1063)
AN APPROACH FOR PLANT-SPECIFIC RISK-INFORMED
DECISIONMAKING INSERVICE INSPECTION OF PIPING
A. INTRODUCI7ION
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.


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.
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 inspection of piping (ISI),
with its companion Standard Review Plan, Section
3.9.8 of NUREG-0800 (Ref. 3), and other regulatory documents (Refs. 4-10), implement, in part, the Com mission's policy statement and the staff's framework for incorporating risk insights into the regulation of nu clear power plants.


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.
In 1995 and 1996, the industry developed a number of documents addressing the increased use of PRA in nuclear plant regulation. The American Society of Me chanical Engineers (ASME) initiated Code Cases N-560 (Ref. 11), N-577 (Ref. 12), and N-578 (Ref. 13)
that address the importance categorization and inspec- tion of plant piping using risk insights. The Electric Power Research Institute (EPRI) published its "PSA
Applications Guide" (Ref. 14) to provide utilities with guidance on the use of PRA information for both regu latory and nonregulatory applications. The Nuclear En ergy Institute (NEI) has been developing guidelines on risk-based ISI and submitted two methods, one devel oped by EPRI (Ref. 15) and the other developed by the ASME research and the Westinghouse Owners Group (Refs. 16-17), for staff review and approval.


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.
Given the recent initiatives by the ASME in devel oping Code Cases N-560, N-577, and N-578, it is an ticipated that licensees will request changes to their plant's design, operation, or other activities that require NRC approval to incorporate risk insights into their ISI
programs (known as risk-informed inservice inspec tion programs, RI-ISI). Until the RI-ISI is approved for generic use, the 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)(3Xi) in Section
50.55a, "Codes and Standards," of 10 CFR Part 50,  
"Domestic Licensing of Production and Utilization Fa cilities," providing an acceptable level of quality and safety. As always, licensees should identify how the USNRC REGULAIORY GUItES
The guides ma Issued In On following ton broad divisions:
Regulator Guides ae Issued to describe and make available to fte publlc such Informs Ilonesmethodsaoceptabletothe NRCstaffforimplemen ngepedflc partsof*teom-


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 NRCs home page at <WWW.NRC.GOV> through the Electronic Reading Room, Accession Number ML032510128.
===1. Power Reactor ===
6 Products
.tisonrmgulatonr, tachnquesuaedbythesaffinevluating peciflcproblem rpos-
2. Research and Test Reactors
7. lnraon luatedacddentarend data needed by the NRC staff in ts review of licational-or per-
. Fuels and Materials Fadlities S. Ooculxtonal Health mits and cenees. Regulatory Ouides;m not "re-titutes for gtilations, and compllance
4. Envlromentlal and Siting
9. Anttrust and Financal Review with lem Isanot required. Methods and eolutionadtiffent frn mhoeeetoutlnthe des
6. Materials and Plant Protection
10. General will be acceptable If tNhy provide a basisforthefindings requisite to the Isusnc orcon orufnce a permit or lkers by 1he Co--on.


U.S. NUCLEAR REGULATORY COMMISSION              Revision 1 September 2003 REGULATORY GUIDE
Sngte copies of regulatory guides may be obtained free of charge by wilting the Repro
OFFICE OF NUCLEAR REGULATORY RESEARCH
'this guide was Issued ater consieation o i mnwnts received from toe public. COm- duction and Oistbutdon Services Section, Office of the Chief kndon Cfficer, U.S. Nu
REGULATORY GUIDE 1.178 AN APPROACH FOR PLANT-SPECIFIC RISK-INFORMED
"erits stuggestions forImprovements In les guides we encouraged at all tnms, and dear Reglator Commslson, Washingto, DC 20555-00l; o by fax at (301)415-228.
DECISIONMAKING FOR INSERVICE INSPECTION OF PIPING


==A. INTRODUCTION==
vlmds ll be nr4sed, as appropiae to accommodate comnments and ljo reflect new ls,&#xfd;
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
or by *-mail lo GRWI @NRC.GOV
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,
Issued guides may also be purchased fom t*e National Telhnical nfomt..onServionon Witten conunenra may be ubmnltted to the Rules Review and Directves Branch. ADM,  
1.176, and 1.177; SRP Chapters 3.9.7, 16.1, and 19 ), implement, in part, the Commissions Policy Statement and the staffs framework for incorporating risk insights into the regulation of nuclear power plants.
8 8tnding orderbasis. Details on Via service may be obtained bywrlting NTIS, 6285 Po U.S. Nudear Reguatory Commisslork Washington, DC 20555-0001.


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.
Roya Road. Spinged, VA 22181.


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
September 1998
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 guides revisions, the revised guide was not issued in draft form in the manner that has normally been used for regulatory guides.
chosen approach, methods, data, and criteria are ap propriate for the decisions they need to make.


Rather, a more limited stakeholders 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.
In October 1997, the Commission published a draft of this regulatory guide for public comment. This guide's principal focus is on the use of PRA findings and risk insights in support of proposed changes to a plant's design, operations, and other activities that re quire NRC approval. 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 reg ulatory guide describes methods acceptable to the NRC
staff for integrating insights from PRA techniques with traditional engineering analyses into ISI programs for piping.


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)
The draft guide, DG- 1063, was discussed during a public workshop held on November 20-21, 1997, and was peer reviewed. While the public comments and peer review of the document were positive, the staff has not had an opportunity to apply the guidance to indus try's pilot plants. Therefore, this regulatory guide is be ing issued for trial use on the pilot plants. This regula tory guide does not establish any final staff positions, and may be revised in response to experience with its use. As such, this trial regulatory guide does not estab lish a staff position for purposes of the Backfit Rule, 10
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.
CFR 50.109, and any changes to this regulatory guide prior to staff adoption in final form will not be consid ered to be backfits as defined in 10 CFR 50.109(a)(1). 
This will ensure that the lessons learned from regulato ry review of the pilot plants are adequately addressed in this document and that the guidance is sufficient to en hance regulatory stability in the review, approval, and implementation of proposed RI-ISI programs.


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.
In the interest of optimizing limited resources, the appendices that were in DG- 1063 will be incorporated in a future NUREG report. The appendices have been deleted from this guide to focus the NRC staff's limited resources on the review and approval of the pilot plant applications and the topical reports submitted in sup port of the pilot plant analyses. Staff positions on the methodologies will be provided in the staff's safety evaluation of the topical reports and pilot plant submit tals. This process would minimize resources needed to update the RG to address the different methods pro posed by the industry.


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:
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 regula- tory decisionmaking. In August 1995, the NRC  

adopted a policy statement regarding the expanded use of PRA (Ref. 1). 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
t The use of PRA technology should be in creased in all regulatory matters to the ex tent supported by the state-of-the-art in PRA methods and data and in a manner that complements the deterministic approach and supports the NRC's traditional philoso phy of defense-in-depth.


1.178-3 complements the deterministic approach and supports the NRCs traditional philosophy of defense in depth.
* PRA and associated analyses (e.g., sensi tivity studies, uncertainty analyses, and im portance measures) should be used in regu latory matters, where practical within the bounds of the state-of-the-art, to reduce un necessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal of addi tional regulatory requirements in accor dance with 10 CFR 50.109 (Backfit Rule). 
Appropriate procedures for including PRA
in the process for changing regulatory re quirements should be developed and fol lowed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.



"* PRA evaluations in support of regulatory decisions should be as realistic as practica ble and appropriate supporting data should be publicly available for review.
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.



"* The Commission's safety goals for nuclear power plants and subsidiary numerical ob jectives are to be used with appropriate con sideration of uncertainties in making regu latory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.
PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.



In its approval of the policy statement, the Com mission articulated its expectation that implementation of the policy statement will improve the regulatory pro cess in three areas: foremost, through safety decision making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees.
The Commissions 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:
In parallel with the publication of the policy state ment, the staff developed a regulatory framework that incorporates risk insights. That framework was articu-
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.
1.178-2 I'
/i


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 Commissions Policy Statement and the staffs framework for incorporating risk insights into the regulation of nuclear power plants.
lated in a November 27, 1995, paper (SECY-95-280)  
to the Commission. This regulatory guide, which ad dresses ISI programs of piping at nuclear power plants, j
is part of the implementation of the Commission's policy statement and the staff's framework for incorpo rating risk insights into the regulation of nuclear power plants. This document uses the knowledge base docu mented in Revision 1 of NUREG/CR-6181 (Ref. 18),
and it reflects the experience gained from the ASME
initiatives (Code Case development and pilot plant ac tivities). 
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 2000 reactor-years of plant operating experience and extensive research in the areas of material sciences, aging phenomena, and in spection techniques. This information, combined with modem risk assessment techniques and associated data, can be used to develop a more effective approach to ISI programs for piping.


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 compo nents are found in 10 CFR 50.55a and the General De
2 sign Criteria listed in Appendix A to 10 CFR Part 50.


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
These requirements are throughout the General Design Criteria, such as in Criterion I, "Overall Require ments," Criterion II, "Protection by Multiple Fission Product Barriers," Criterion III, "Protection and Reac tivity Control Systems," and Criterion IV, "Fluid Sys tems."
Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) (Ref. 19) is referenced by 10 CFR 50.55a, which addresses the codes and standards for design, fabrication, testing, and inspection of piping systems.


1.178-4 II, "Protection by Multiple Fission Product Barriers"; Criterion III, "Protection and Reactivity Control Systems"; and Criterion IV, "Fluid 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 50.55
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 50.55


====a.   ISI====
====a. 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.
programs are intended to address all piping locations that are subject to degradation. Incorporating risk in sights into the programs can focus inspections on the more important locations and reduce personnel expo sure, while at the same time maintaining or improving public health and safety. The justification for any re duction 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.


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.
As a result of the above insights, more efficient and technically sound means for selecting and scheduling ISIs of piping are under development by the ASME
(Refs. 11-13). 
When categorizing piping segments in terms of their contribution to risk, it is the responsibility of a li censee to ensure that the categorization of piping seg ments and the resulting inspection programs are consis tent 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 (Ref. 4). This regulatory guide augments the guidance presented in Regulatory Guide 1.174 by pro viding guidance specific to incorporating risk insights to inservice inspection programs of piping.


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.
Purpose of the Guide Consistent with Regulatory Guide 1.174 (Ref. 4),  
this regulatory guide focuses on the use of PRA in sup port of a risk-informed ISI program. This guide pro vides guidance on acceptable approaches to meeting the existing Section XI requirements for the scope and frequency of inspection of ISI programs. Its use by li censees 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 com pliance with the requirements of 10 CFR 50.55a and with Section XI of the ASME Boiler and Pressure Ves sel Code, which are part of the plant's licensing basis.


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.
This approach provides an acceptable level of quality and safety (per 10 CFR 50.55a(a)(3)(i)) by incorporat ing insights from probabilistic risk and traditional anal ysis calculations, supplemented with operating reactor data. Licensees who propose to apply risk-informed ISI
programs would amend their final safety analysis re port (FSAR, Sections 5.3.4 and 6.6) accordingly. A
Standard Review Plan (SRP) (Ref. 3) has been prepared for use by the NRC staff in reviewing RI-ISI applica tions.


Purpose of the Guide Consistent with Regulatory Guide 1.174, this regulatory guide focuses on the use of PRA
This document addresses risked-informed meth ods to develop, monitor, and update more efficient ISI
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.
programs for piping at a nuclear power facility. This guidance does not preclude other approaches for incor porating risk insights into the ISI programs. Licensees may propose other approaches for NRC consideration.


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 plants inspection program for pipin
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 regula tions on the basis of their accumulated plant experience and knowledge. This document addresses risk informed approaches that are consistent with the basic
1.178-3


====g. The current ISI====
elements identified in Regulatory Guide 1.174 (Ref. 4).
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 plants licensing basis.
In addition, this document provides guidance on the following for the purposes of RI-ISI.


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.
"
Estimating the probability of a leak, a leak that pre vents the system from performing its function (dis abling 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, re duction of unnecessary radiation exposure to per sonnel,
"* Determining the risk impact of changes to ISI pro grams,  
"* Capturing deterministic considerations in the re vised ISI program, and  
"
Developing an inspection program that monitors the performance of the piping elements for consis tency with the conclusions from the risk assess ment.


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.
Given the recent initiatives by the ASME in devel oping Code Cases N-560, N-577, and N-578 (Refs.


Notable examples of augmented programs for piping inspections include the following topics:
11-13), it is anticipated that licensees will request changes to their plant's design, operation, or other ac tivities that require NRC approval to incorporate risk insights in their ISI programs (RI-ISI). Until the RI-ISI
*
is approved for generic use, the staff anticipates that li censees will request changes to their ISI programs by requesting NRC approval of a proposed inspection pro gram that meets the criteria of 10 CFR 50.55a(a)(3)(i),  
intergranular stress corrosion cracking (IGSCC) of stainless steel piping in boiling water reactors (BWRs) (Generic Letter 88-01),  
providing an acceptable level of quality and safety. The licensee's RI-ISI program will be enforceable under 10
*
CFR 50.55a.
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
Scope of the RI-ISI Program This regulatory guide only addresses changes to the ISI programs for inspection of piping. To adequate ly reflect the risk implications of piping failure, both partial and full-scope RI-ISI programs are acceptable to the NRC staff.
*
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
Partial Scope: A licensee may elect to limit its RI
program provided that the licensee identifies and the staff approves the specific programs and program changes.
ISI program to a subset of piping classes, for example, ASME Class-1 piping only, including piping exempt from the current requirements.


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.
Full Scope: Afull scope RI-ISI program evaluates the piping in a plant as being either high or low safety significant. 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 compo nents that are relied upon to remain functional during and following design basis events to en sure the integrity of the reactor coolant pres sure boundary, the capability to shut down the reactor and maintain it in a safe shutdown con dition, 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.



-
Estimating the probability of a leak, a leak that prevents the system from performing its function (disabling leak), and a rupture for piping segments,  
Non-safety-related structures, systems or com ponents

"* That are relied upon to mitigate accidents or transients or are used in plant emergen cy operating procedures; or
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,
"* 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.
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.
For both the partial and full scope evaluations, the licensee is to demonstrate compliance with the accep tance guidelines and key principles of Regulatory Guide 1.174 (Ref. 4).   
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 regula tory 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 be cause detailed evaluations will often identify welds as the locations most likely to experience degradation.


Scope of the RI-ISI Program This regulatory guide only addresses changes to the ISI programs for inspection of piping.
Welds are most likely to have fabrication defects, welds are often at locations of high stress, and certain de gradation mechanisms (stress corrosion cracking) usu ally occur at welds. Nevertheless, there are other degra dation mechanisms such as flow-assisted-corrosion (e.g., erosion-corrosion) and thermal fatigue that occur independent of welds.


To adequately reflect the risk implications of piping failure, both partial and full-scope RI-ISI
1Generally, ASME Code Class 1 includes all reactor pressure bound.
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.
ary (RCPB) components. ASME Code Class 2 generally includes sys tems or portions of systems important to safety that are designed for post-accident containment and removal of heat and fission products.


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.
ASME Code Cass 3 generally includes those system components or portions of systems important to safety that are designed to provide cooling water and auxiliary feedwater for the front-line systems.


Partial scope applications should include the full population of piping within the selected subset of piping such as ASME Class and/or plant systems.
1.178-4


Full Scope:  A full scope RI-ISI includes:
To ensure that the proposed RI-ISI program would provide an acceptable level of quality and safety, the li censee should use the PRA to identify the appropriate scope of the piping segments to be included in the pro gram. In addition, licensees implementing the risk-in formed process may identify piping segments catego rized as high safety-significant (HSS) that are not currently subject to the traditional Code requirements (e.g., outside the Code boundaries, including Code ex empt piping) or are not being inspected to a level that is commensurate with their risk significance. In this con text, HSS refers to a piping segment that has a relatively high contribution to risk. PRA systematically takes credit for systems with non-Code piping that provide support, act as alternatives, and act as backups to those systems with piping that are within the scope of the cur rent Section XI of the Code.

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.



Organization and Content This regulatory guide is structured to follow the general four-element process for risk-informed ap plications discussed in Regulatory Guide 1.174 (Ref.
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.
4). The Discussion section summarizes the four element process developed by the staff to evaluate pro posed changes related to the development of a RI-ISI
program. Regulatory Position 1 discusses an accept able approach for defining the proposed changes to an ISI program. Regulatory Position 2 addresses, in gen eral, the traditional and probabilistic engineering eval uations performed to support RI-ISI programs and pre sents the risk acceptance goals for determining the acceptability of the proposed change. Regulatory Posi tion 3 presents one acceptable approach for implement ing and monitoring corrective actions for RI-ISI pro grams. The documentation the NRC will need to render its safety decision is discussed in Regulatory Position
4.


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.
Relationship to Other Guidance Documents As stated above, this regulatory guide discusses ac ceptable 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. Regulatory Guide 1.174 de scribes a general approach to risk-informed regulatory decisionmaking and discusses specific topics common to all risk-informed regulatory applications. Topics ad dressed include:
PRA quality-data, assumptions, methods, peer review,
"* PRA scope-internal and external event initiators, at-power and shutdown modes of operation, con sideration of requirements for Level 1, 2, and 32 analyses,  
"* Risk metrics--core damage frequency, large early release frequency and importance measures,  
*
Sensitivity and uncertainty analyses.


2 Level 1 - accident sequence analysis, Level 2 - accident progression and source term analysis, and Level 3 - offsite consequence analysis.
To the extent that a licensee elects to use PRA as an element to enhance or modify its implementation of ac tivities 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.


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.
The information collections contained in this doc ument are covered by the requirements of 10 CFR
Part 50, which were approved by the Office of Manage ment and Budget (OMB),  
approval number
3150-0011. The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of in formation unless it displays a currently valid OMB con trol number.


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.
Abbreviations and Definitions ASME
American Society of Mechanical Engi neers 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 In the context of this regulatory guide, expert elicitation is a process used to esti mate failure rates or probabilities of pip ing when data and computer codes are un available for the intended purpose. It is a process used to estimate the failure proba bility and the associated uncertainties of the material in question under specified degradation mechanisms. For example, if a structural mechanics code is not quali fied to calculate the failure probability of plastic piping and no data are available to estimate its failure probability, experts in plastic piping and their failure may be asked to estimate the failure probabilities.


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.
If applicable industry data are available, an expert elicitation process would not be needed.


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
2Level 1--accident sequence analysis, Level 2-accident progression and source term analysis, and Level 3-offsite consequence analysis.
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
1.178-5

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.
Expert Panel FSAR
HSS
IGSCC
Importance Measures ISI
IST
LERF
LSS
NDE
NEI
NRC
PRA
PSA
RCPB
Normally refers to plant personnel exper ienced in operations, maintenance, PRA,
ISI programs, and other related activities and disciplines that impact the decision under consideration.


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.
Final Safety Analysis Report High safety significance Intergranular stress corrosion cracking Used in PRA to rank systems or compo nents in terms of risk significance Inservice inspection Inservice testing Large early release frequency Low safety significance Nondestructive examination Nuclear Energy Institute Nuclear Regulatory Commission Probabilistic risk assessment Probabilistic safety assessment Reactor coolant pressure boundary RI-ISI
Staff Sensitivity Studies SRP
SRRA
SSCs Tech Spec


Abbreviations and Definitions ASME
==B. DISCUSSION==
American Society of Mechanical Engineers BPVC
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. Figure I il lustrates the five key principles involved in the inte grated decisionmaking process; they are described in detail in Regulatory Guide 1.174 (Ref. 4). In addition, Regulatory Guide 1.174 describes a four-element pro cess for evaluating proposed risk-informed changes as illustrated in Figure 2.
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==
Figure 1 Principles of Risk-Informed Integrated Decisionmaking Figure 2 Principal Elements of Risk-Informed, Plant-Specific Decisionmaking
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.
1.178-6 Risk-informed inservice inspection Refers to NRC employees Varying parameters to assess impact due to uncertainties Standard Review Plan Structural reliability/risk assessment (re fers to fracture mechanics analysis)
Structures, systems and components Technical specifications


The key principles and the section of this guide that addresses each of these principles for RI-ISI programs are as follows.
The key principles and the section of this guide that addresses each of these principles for RI-ISI programs are as follows.


1.
1. The proposed change meets the current regulations unless it is explicitly related to a requested exemp tion 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 core damage frequency or risk, the increases should be small and consistent with the intent of the Com mission's Safety Goal Policy Statement. (Regula tory Position 2.2)
5. The impact of the proposed change should be mon itored by using performance measurement strate gies. (Regulatory Position 3)
The individual principles are discussed in detail in Regulatory Guide 1.174.


The proposed change meets the current regulations unless it is explicitly related to a requested exemption or rule change.  (Regulatory Position 2.1.1)
Section 2 of Regulatory Guide 1.174 describes a four-element process for developing risk-informed reg ulatory changes. An overview of this process is given here and illustrated in Figure 2. 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 pro posed change to the ISI program as defined in Element I depends on both the analysis performed in Element 2 and the definition of the implementation of the ISI pro gram performed in Element 3. While ISI is, by its na ture, 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 moni toring is performed to make sure that these margins are not eroded. Element 4 involves preparing the documen tation to be submitted to the NRC and to be maintained by the licensee for later reference.
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 Commissions 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==
==C. REGULATORY POSITION==
1.
1. ELEMENT 1: DEFINE THE PROPOSED  
 
CHANGES TO ISI PROGRAMS  
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 de- scribing the scope of ISI piping that would be incorpo rated in the overall assessment and how the inspection of this piping would be changed. Also included in this ele ment is identification of supporting information and a proposed plan for the licensee's interactions with the NRC throughout the implementation of the RI-ISI.
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 licensees 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 plants 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 plants 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.



1.1 Description of Proposed Changes A full description of the proposed changes in the ISI
Identification of the elements of the ISI program to be changed.
program is to be prepared. This description should in dude:
"
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 pro posed changes, the licensee should also confirm that the plant's design and operation is in accordance with its current requirements and that engineering infor mation used to develop the proposed RI-ISI program is also consistent with the current requirements.
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 plants 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 elements of the ISI program to be changed.
Identification of the information that will be used to support the changes.  This could include performance data, traditional engineering analyses, and PRA information.



"* Identification of the piping in the plant that is both di rectly and indirectly involved with the proposed changes. Any piping not presently covered in the plant's ISI program but categorized as high safety significant (e.g., through an integrated decisionmak ing process using PRA insights) should be identified and appropriately addressed. In addition, the particu lar systems that are affected by the proposed changes should be identified since this information is an aid in planning the supporting engineering analyses.
A brief statement describing how the proposed changes meet the intent of the Commissions 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.
"* Identification of the information that will be used to support the changes. This could include performance data, traditional engineering analyses, and PRA in formation.


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 staffs approval has not been compromise
"* A brief statement describing how the proposed changes meet the intent of the Commission's PRA
Policy Statement.


====d.   All====
1.2 Changes to Approved RI-ISI Programs This section provides guidance on the need for licen sees to report program activities and guidance on formal NRC review of changes made to RI-ISI programs.


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
The licensee should implement a process for deter mining when RI-ISI program changes require formal NRC review and approval. Changes made to the NRC
review and approval prior to implementation.
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 changes should be evaluated using the change mechanisms
1.178-7


2.
described in the applicable regulations (e.g., 10 CFR
50.55a, 10 CFR 50.59) to determine whether NRC re view and approval are required prior to implementation.


ELEMENT 2:  ENGINEERING ANALYSIS
If there is a question regarding this issue, the licensee should seek NRC review and approval prior to imple mentation.
As part of defining the proposed change to the licensees 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.
2. ELEMENT 2: ENGINEERING ANALYSIS
As part of defining the proposed change to the licens ee's ISI program, the licensee should conduct an engi neering evaluation of the proposed change, using and in tegrating 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 (Ref 4). Regulatory Guide 1.174 provides general guidance for performing this evaluation, which is supplemented by the RI-ISI guidance herein.


The engineering evaluations are to:
Figure 3 Element 2 The regulatory issues and engineering activities that should be considered for a risk-informed ISI pro gram are summarized here. For simplicity, the discus sions are divided into traditional and PRA analyses (see Figure 3). Regulatory Position 2.1 addresses the tradi tional 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 tradi tional and PRA analyses are iterative.

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 Commissions 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.
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 core damage frequency 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.


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
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 Regula tory Guide 1.174.


1.178-12 margins, assessment of failure potential of piping segments, and assessment of primary and secondary effects (failures) that result from piping failures.
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 margins, assessment of failure potential of piping segments, and assessment of pri mary and secondary effects (failures) that result from piping failures.


The engineering analysis for a RI-ISI piping program will achieve the following:
The engineering analysis for a RI-ISI piping pro gram will achieve the following:
1.
1.


Assess compliance with applicable regulations,
2.
2.


Perform defense-in-depth evaluation,
3.
3.


Perform safety margin evaluation,
4.
4.


Define piping segments,
5.
5.


Assess failure potential for the piping segment,
Assess compliance with applicable regulations, Perform defense-in-depth evaluation, Perform safety margin evaluation, Define piping segments, Assess failure potential for the piping segment (from leaks to breaks),
6.
6. Assess the consequences (both direct and indirect)  
 
of piping segment failure,  
Assess the consequences (both direct and indirect) of piping segment failure,
7. Categorize the piping segments in terms of safety (risk) significance,
7.
 
Categorize the piping segments in terms of safety significance,
8.
8.


Develop an inspection program,
9.
9.


Assess the impact of changing the ISI program on CDF and LERF, and
Develop an inspection program, Assess the impact of changing the ISI program on CDF and LERF, and
10.
10. Demonstrate conformance with the key principles (e.g., maintaining sufficient safety margins, de fense 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 com promise compliance with the regulations. The evalua tion should consider the appropriate requirements in the licensing basis and applicable regulatory guidance.
 
Specifically, the evaluation should consider
1.178-8


Demonstrate conformance with the key principles (e.g., maintaining sufficient safety margins, defense in depth consideration, Commissions 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  
10 CFR 50.55a  

*
Appendix A to 10 CFR Part 50
Appendix A to 10 CFR Part 50  

-
Appendix B to 10 CFR Part 50
Criterion I, "Overall Requirements"

-
ASME B&PVC, Section XI (10 CFR Part 50.55a)
Criterion II, "Protection of Multiple Fission Product Barriers"

-
Regulatory Guide 1.84, Design, Fabrication, and Materials Code Case Acceptability, ASME Section III
Criterion III, "Protection and Reactivity Con trol Systems"

-
Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME
Criterion IV, "Fluid Systems," etc
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
ASME Boiler and Pressure Vessel Code, Section XI (10 CFR Part 50.55a)  
a Regulatory Guide 1.84 (Ref. 20)
*
Regulatory Guide 1.85 (Ref. 21)
*
Regulatory Guide 1.147 (Ref. 22)
*
Appendix B to 10 CFR Part 50.


1.178-13 necessary because proposed ISI program changes could affect issues not explicitly stated in the licensees FSAR or ISI program documentation.
In addition, the evaluation should consider wheth er the proposed changes have affected license commit ments. A broad review of the licensing requirements and commitments may be necessary because proposed ISI program changes could affect issues not explicitly stated in the licensee's FSAR or ISI program documen tation.


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
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 accept able RI-ISI approaches presented in this guide may be proposed by licensees so long as supporting informa tion is provided that demonstrates that the key prin ciples discussed in this guide are maintained.
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.
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 pro gram.


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.
NRC-endorsed ASME Code Cases that apply risk informed ISI programs will be consistent with this reg ulatory guide in that they encourage the use of risk in sights in the selection of inspection locations and the use of appropriate and possibly enhanced inspection techniques that are appropriate to the failure mecha nisms 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.
2.1.2 Defense-in-Depth Evaluation As stated in Regulatory Guide 1.174 (Ref. 4), the engineering analysis should evaluate whether the im pact of the proposed change in the ISI program (indi- vidually 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 equip ment and human performance. Where a comprehensive risk analysis can be done, it can be used to help deter mine the appropriate extent of defense-in-depth (e.g.,  
balance among core damage prevention, containment failure, and consequence mitigation) to ensure protec tion of public health and safety. Where a comprehen sive risk analysis is not or cannot be done, traditional defense-in-depth consideration should be used or main tained 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 dam age, 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, Reference 4, for addtional guidance). 
An important element of defense in depth for RI
ISI is maintaining the reliability of independent barri ers to fission product release. Class I piping (primary coolant system) is the second boundary between the ra dioactive fuel and the general public. If a RI-ISI pro gram categorized, for example, all the hot and cold legs of the primary system piping as LSS and calculated that, with no inspections, the frequency of leaks would not increase beyond existing performance history of the ASME Code, the staff would continue to require some level of NDE inspection.


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).
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 accom panying engineering assumptions are intended to ac count 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).   
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
Insufficient safety margins may require additional attention. Prior to a request for relaxation of the existing
1.178-9


1.178-14 increase beyond existing performance history of the ASME Code, the staff would continue to require some level of NDE inspection.
requirements, the licensee must ensure that the uncer tainties are adequately addressed. The quantification of uncertainties would likely require supporting sensitiv ity analyses.


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 appropri;
ate for evaluating whether sufficient safety margins would be maintained if the proposed change were im plemented. An acceptable set of guidelines for making that assessment are summarized below. Other equiva lent decision criteria could also be found acceptable.


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.
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 margin to account for analysis and data uncer tainty.


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.
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; dif ferent criteria or definitions can be applied to each pip ing segment. One acceptable method is to identify seg ments 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 de fined 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 seg ments.


An acceptable set of guidelines for making that assessment are summarized below. Other equivalent decision criteria could also be found acceptable.
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 spe cific definition of a piping segment, but the analysis and the definition of a segment must-be consistent and technically sound.


Sufficient safety margins are maintained when:
2.1.5 Assess Piping Failure Potential The engineering analysis includes evaluating the failure potential of a piping segment. Figure 4 identifies the three means for estimating the failure potential of a piping segment: data, fracture mechanics computer codes, and the expert elicitation process. 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 dy namic loads, flaw size, flaw distribution, inspection pa rameters, 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 defini tion) that will be needed to support the PRA or risk as sessment. The failure potential used in or in support of EsTrMATING FAM- URE P~l

1FRATUE
Codes and standards (see Regulatory Position 2.1.1) or alternatives approved for use by the NRC are met, and
iEXPERT

I
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.
I ELICITATIO
 
CS
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.
PROCESS
 
1 j~CODEIS 1 IFNED)  
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.
(IFN
 
J
2.1.5 Assess Piping Failure Potential The engineering analysis includes evaluating the failure potential of a piping segment.
DE]
 
Figure 4 Estimating Failure Potential of Piping Segments
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).
1.178-10
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
the analysis should be appropriate for the specific envi ronmental conditions, degradation mechanisms, and failure modes for each piping location and break size (e.g., leak, disabling leak, break). When data are ana lyzed to develop a categorization process relating de gradation 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 (Ref. 23, 24). 
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, varia tion in steady-state and transient loads, availability and accuracy of plant operating history, availability of in spection 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 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.


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 clas sification of the risk significance of each segment. Ref erencing applicable generic topical reports approved by the NRC is one acceptable means to standardize the process. This review should be documented and a sum mary discussion of the review should be included in the submittal. When new computer codes are used to de velop quantitative estimates, the techniques should be verified and validated against established industry codes and available data. When data are used to evalu ate the likelihood of piping failures, the data should be submitted to the NRC or ieferenced by an NRC-ap proved topical report. As stated in Regulatory Guide
1.174 (Ref. 4), "data, methods, and assessment criteria used to support regulatory decisionmaking must be scrutable and available for public review." It is the re sponsibility of the licensee to provide the data, meth ods, and justification to support its estimation of the failure potential of piping segments. Since conse quences of and potential for piping failures could differ for leaks, disabling leaks, and breaks, the failure poten tial for all three break types should be addressed.


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 ac complished by performing a detailed walkdown of a nuclear power facility's piping network. Assessment of internal and external events, including resulting pri mary and secondary effects of piping failures (e.g.,
leaks, disabling leaks, and breaks) are important pa rameters to the risk-informed program (see Figure 5).   
Leaks can result in failures of electrical components caused by jet impingement. Disabling leaks and full breaks can lead to a loss of system function, flooding induced damage, and initiating events. Full breaks can lead to damage resulting from pipe whip, as well as flooding and initiating events. Each of these break types has its associated failure potential that is evalu ated in Regulatory Position 2.1.5. A failure modes and consequence assessment is performed to identify the potential failures, from piping leaks to breaks. Internal flooding PRAs can identify the impact of jet impinge ment and flooding to the RI-ISI program. The failures are used as input to the risk analysis. Alternative meth ods for evaluating consequences should be submitted to the NRC for review and approval. These evaluations are expected to provide information for the conse quence analysis. They are not intended to be used in lieu of the plant licensing basis.


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 facilitys 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.
2.1.7 Probabilistic Fracture Mechanics Evaluation When implementing probabilistic fracture me chanics computer programs that estimate structural reliability and are used in risk assessment of piping, or other analytic methods for estimating the failure poten tial of a piping segment, some of the important parame ters that need to be assessed in the analysis include the identification of structural mechanics parameters, deg radation mechanisms, design limit considerations, op erating practices and environment, and the develop ment of a data base or analytic methods for predicting the reliability of piping systems. Design and opera tional stress or strain limits are assessed. This informa tion is available to the licensee in the design informa tion for the plant. The loading and resulting stresses or strains on the piping are needed as input to the calcula tions that predict the failure probability of a piping seg ment. The use of validated computer programs, with appropriate input, is strongly recommended in a quanti tative RI-ISI program because it may facilitate the
1.178-11


2.2 Probabilistic Risk Assessment In accordance with the Commissions 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.
LEAK/BREAK
CONSEQUENCES
Leak Effects from Jet Impingement Disabling Leak or Full Break Loss of System Function Disabling Leak (plant trip) or Initiating Event Full Break Disabling Leak or Full Break Effects from Flooding Full Break Effects from Pipe Whip Figure 5 Mapping of Probabilities and Consequences for RI-ISI Analysis regulatory evaluation of a submittal. The analytic method should be validated with applicable plant and industry piping performance data.


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.
2.2 Probabilistic Risk Assessment In accordance with the Commission's policy on PRA, the risk-informed application process is intended not only to support relaxation (number of inspections, inspection intervals and methods), but also to identify areas where increased resources should be allocated to enhance safety. Therefore, an acceptable RI-ISI pro cess should not focus exclusively on areas in which re duced inspection could be justified. This section ad dresses ISI-specific considerations in the PRA to support relaxation of inspections, enhancement of in spections, and validation of component operability.


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 DG-
The scope of a RI-ISI program, therefore, should in clude a review of Code-exempt piping for partial or full-scope programs and the review of non-Code piping for full-scope RI-ISI programs.
1122). 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
The general methodology for using PRA in regula tory applications is discussed in Regulatory Guide  
Chapter 19 of the Standard Review Plan.
1.174. The PRA can be used to categorize the piping segments into. HSS and LSS classification (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 (Ref. 4).   
The PRA can be used to categorize the piping segments into safety-significant and low- safety-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 licensees 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 licensees 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.
If a licensee elects to use PRA to enhance or modify its activities affecting the safety-related functions of SSCs subject to the provisions of Appendix B to
10 CFR Part 50, the pertinent requirements of Appen dix B will also apply to the PRA. In this context, there fore, a licensee would be expected to control PRA ac tivity in a manner commensurate with its impact on the facility's design and licensing basis and in accordance with all applicable regulations and its QA program de scription. An independent peer review can be an impor- tant element in ensuring this quality. The licensee's submittal should discuss measures used to ensure ade quate 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 report should address any limitations of the analysis that are expected to im pact the conclusion regarding the acceptability of the proposed change. The licensee's resolution of the find ings of the peer review, certification, or cross compari son, when performed, should also be submitted. This response could indicate whether the PRA was modified or could justify why no change to the PRA was neces sary to support decisionmaking for the change under consideration.


2.2.1 Modeling Piping Failures in a PRA  
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.
Input from the traditional engineering analysis ad dressed in Regulatory Position 2.1 includes identifica tion of piping segments from the point of view of the failure potential (degradation mechanisms) and conse quences (resulting failure modes and consequential pri mary and secondary effects). The traditional analysis identifies both the primary and secondary effects that can result from a piping failure, such as a leak, disabling leak, and a break. 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.
Each pipe segment failure may have one of three types of impacts on the plant.


1.
1. Initiating event failures when the failure directly causes a transient and may or may not also fail one or more plant trains or systems.


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 character ized by train or system unavailability that may re quire shutdown because of the technical specifica tions or limiting conditions for operation.


2.
1.178-12 V


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 de mand for a train or system and are usually caused by the transient-induced loads on the segment dur ing system startup.


3.
The impact of the pipe segment failure on risk should be evaluated with the PRA. Evaluation may in volve a quantitative estimate derived from the PRA, a systematic technique to categorize the consequence of the pipe failure on risk, or some combination of quanti fication 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 ini tiating event, for example), the evaluation process should be expanded to assess these events.


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.
PRAs normally do not include events that repre sent failure of individual piping segments nor the struc tural elements within the segments. A quantitative esti mate of the impact of segment failures can be done by modifying the PRA logic to systematically and ex plicitly include the impact of the individual pipe seg ment 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 fail ure (referred to as the surrogate approach). In either case, to assess the impact of a particular segment fail ure, the analyst sets the appropriate events to a failed state in the PRA (by assigning them a frequency or probability of 1.0) and requantifies the PRA or the ap propriate parts of the PRAas needed. The requantifica tion should explicitly address truncation errors, since cut set or truncated sequences may not fully capture the impact of multiple failure events. This yields condi tional CDF (CCDF) and conditional LERF (CLERF)
estimates when the segment failure would trip the plant, and conditional core damage probabilities (CCDP) and conditional large early release probabili ties (CLERP) when the segment failure would not trip the plant.


The impact of the pipe segment failure on risk should be evaluated with the PRA.
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 fail ure, the process should provide for determining which range each segment's failure would lie within. In gen eral, the consequences would range from high, forthose 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 se lected. The use of ranges instead of individual results estimates may require fewer calculations, but the cate gorization process and decision criteria should be justi fied, well defined, and repeatable.


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.
2.2.1.1 Dependencies and Common Cause Fail ures. The effects of dependencies and common cause failures (CCFs) for ISI components need to be consid ered 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 re quired.


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 segments failure on risk can also be estimated without modifying the PRAs 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 segments 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.
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 pro gram (including both operator and equipment qualifi cation), which factors into the probability of detection, should also be addressed.


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.
2.2.2 Use of PRA for Categorizing Piping Segments Once the impact of each segment's failure on plant risk metrics has been determined, the safety signifi cance of the segments is developed. The method of categorizing a piping segment can vary. For example, if the pipe failure event frequency or probability are esti mated by structural mechanics methods as discussed in Regulatory Position 2.1.5 and the events are incorpo rated into the PRA logic model, importance measure calculations and the determination of safety signifi cance, as discussed in Regulatory Guide 1.174 and SRP
Chapter 19 (Refs. 4 and 8), may be performed. Alterna tively, if a CCDF, CLERF, CCDP, or CLERP (depend ing on the impact the segment failure has on the plant)
are estimated for each segment from the PRA, a CDF
and LERF caused only by pipe failures may be devel oped 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 seg ment. The calculations used in such a process should yield well defined estimates of CDF, LERF, and impor tance measures. The licensee should provide a discus sion of and justification for the threshold criteria used.


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.
As discussed in Regulatory Position 2.2.1, the con sequence of segment failures may be represented by categories of consequences instead of quantitative
1.178-13


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.
"estimates for each segment. In this case, the potential for pipe fail'are as discussed in Regulatory Position
2.1.5 would also be developed as categories ranging from high to low depending on the degradation mecha nisms present and the corresponding likelihood that the segment will fail. These consequence and failure likeli hood 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 combina tion.


2.2.2 Use of PRA for Categorizing Piping Segments When the impact of each segments 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
The safety-significance category of the pipe seg ment will help determine the level of inspection effort devoted to the segment. In general, higher safety significant segments will receive more inspections and more demanding inspections than less significant seg ments. In any integrated categorization process, the principles in Regulatory Guide 1.174 need to be ad dressed. Irrespective of the method used in the analysis, the licensee needs to justify the final categorization pro cess as being robust and reasonable with respect to the analysis uncertainties.


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.
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 neutrality. The change is evaluated and compared with the guidelines presented in Regulatory Guide 1.174.


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 staff expects that a RI-ISI program would lead to both risk reduction and reduction in radiation exposure to plant personnel.


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.
2.3 Integrated Decisionmaking Regulatory Positions 2.1 and 2.2 address the ele ments 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 (Ref. 4), as highlighted in Figure 1, are systemat ically addressed. Technical and operations personnel at the plant review the information and render a finding of HSS or LSS categorization for each piping segment un der review. Detailed guidelines for the categorization of piping segments should be developed and discussed with the group responsible for the determination (typi cally performed by the plant's expert panel).
The method for selecting the number of piping ele ments to be inspected should be justified.


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.
3. ELEMENT 3: IMPLEMENTATION,
PERFORMANCE MONITORING, AND
CORRECTIVE ACTION STRATEGIES
Integrating the information obtained from Ele ments 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, perfor mance monitoring, and corrective action strategies.


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 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 program should be self-correct ing as experience dictates. The program should contain performance measures used to confirm the safety in sights gained from the risk analyses.


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 Commissions Safety Goal Policy Statement.
Upon approval of the RI-ISI program, the licensee should have in place a program for inspecting all HSS
and LSS piping identified in its program. (Note that ref erence to HSS piping is broadened when implementing a more detailed graded categorization process, such as low, medium, and high safety significant. For discus sion purposes, a tWo-category process (e.g., HSS and LSS) will be assumed. Requirements for medium and LSS piping will be addressed on a case-by-case basis.)
The number of required inspections should be a product of the systematic application of the risk-informed pro cess.


1.178-20
3.1 Program Implementation A licensee should have in place a schedule for in specting all segments categorized in its RI-ISI program as LSS and HSS. This schedule should include inspec tion strategies and inspection frequencies, inspection methods, the sampling program (the number of ele ments/areas to be inspected, the acceptance criteria, etc.) for the HSS piping that is within the scope of the ISI program, including piping segments identified as HSS that are not currently in the ISI program.
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 plants expert panel).
The analysis for a RI-ISI program will, in most cases, confirm the appropriateness of the inspection in terval 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 inter vals are contained in Section XI of the B&PVC. How ever, should active degradation mechanisms surface, the inspection interval would be modified as appropri ate. Updates to the RI-ISI program should be per formed at least periodically to coincide with the
The method for selecting the number of piping elements to be inspected should be justified.
1.178-14


3.
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.


ELEMENT 3:  IMPLEMENTATION, PERFORMANCE MONITORING, AND
For example, if changes to the PRA impact the deci sions made for the RI-ISI program, if plant design and operations change such that they impact the RI-ISI pro gram, if inspection results identify unexpected flaws, or if replacement activities 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 re quired 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.
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.
Changes to piping performance, the plant procedures that can affect system operating parameters, piping in spection, component and valve lineups, equipment op erating modes, or the ability of the plant personnel to perform actions associated with accident mitigation should be reviewed in any RI-ISI program update.


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
Leakage and flaws identified during scheduled inspec tions should be evaluated as part of the RI-ISI update.
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
Piping segments categorized as HSS that are not in the licensee's current ISI program should (wherever ap j
propriate 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 de veloped 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 imple mentation of the RI-ISI program at the plant.


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.
For piping segments categorized as HSS 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 pip ing segment.


Piping segments categorized as safety significant that are not in the licensees current ISI
3.2 Performance Monitoring
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.
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 per formed more frequently if dictated by any plant proce- dures to update the PRA (which may be more restrictive than a Section XI period type update) or as new de gradation mechanisms are identified.


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.
31.2 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 defini tion and element selection. Some examples of these in puts would include:
"* Operating characteristics (e.g., changes in water chemistry control)
"* Material and configuration changes
"* Welding techniques and procedures
"* Construction and preservice examination results
"* Stress data (operating modes, pressure, and tem perature 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 Performance Monitoring
3.2.3 Changes to Plant Procedures Changes to plant procedures that affect ISI, such as system operating parameters, test intervals, or the abil ity of plant operations personnel to perform actions as sociated with accident mitigation, should be included for review in any RI-ISI program update. Additionally, changes in those procedures that affect component in spection 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.
3.2.1 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
3.2.4 Equ pment Performance Changes Equipment performance changes should be re viewed with system engineers and maintenance per sonnel to ensure that changes in performance parame ters such as valve leakage, increased pump testing, or identification of vibration problems is included in the periodic evaluation of the RI-ISI program update. Spe cific 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.
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
3.2.5 Examination Results When scheduled RI-ISI program NDE examina tions, pressure tests, and cotresponding VT-2 visual examinations for leakage have been completed, and if

1.178-15
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 plants 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.
unacceptable flaws, evidence of service related degra dation, or indications of leakage have been identified, the existence of these conditions should be evaluated.


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
This update of the RI-ISI program should follow the applicable elements of Appendix B to 10 CFR Part 50
program.
to determine the adequacy of the scope of the inspection 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 pro vides failure information and trending information that may have a profound effect on the element locations currently being examined under a RI-ISI program. In dustry failure data is just as important to the overall pro gram as the owner's information. During the periodic update, industry data bases (including available inter national data bases) should be reviewed for applicabil ity to the owner's plant.


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 licensees 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.
3.3 Corrective Action Programs Each licensee of a nuclear power plant is responsi ble for having a corrective action program, consistent with Regulatory Guide 1.174 (Ref. 4). Measures are to be established to ensure that conditions adverse to qual ity, such as failures, malfunctions, deficiencies, devi ations, defective material and equipment, and noncon formances, 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 condi tion adverse to quality, the cause of the condition, and the corrective action are to be documented and reported to appropriate levels of management.


Industry failure data is just as important to the overall program as the owners information.  During the periodic update, industry data bases (including available international data bases) should be reviewed for applicability to the owners plant.
For Code piping categorized as HSS, this correc tive action program should be consistent with applica ble Section XI provisions. For non-Code and Code exempt piping categorized as HSS, appropriate Section XI provisions should also be used, or the licensee should submit an alternative program based on the risk significance of the piping.


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.
3.4 Acceptance Guidelines These acceptance guidelines are for the imple mentation, monitoring, and corrective action programs for the accepted RI-ISI program plan.


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.
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.


3.4 Acceptance Guidelines These acceptance guidelines are for the implementation, monitoring, and corrective action programs for the accepted RI-ISI program plan.
2. The corrective action program should provide rea sonable assurance that a nonconforming compo nent will be brought back into conformance in a timely fashion. The corrective actions required in ASME Section XI should continue to be followed.


1.
3. Evaluations within the corrective action program may also include:
"* Ensuring that the root cause of the condi tion is determined and that corrective ac tions are taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condi tion, and the corrective action are to be documented and reported to appropriate levels of management.


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.
"* Determining the impact of the failure or nonconformance on system or train oper ability since the previous inspection.


2.
"* Assessing the applicability of the failure or nonconforming condition to other components in the RI-ISI program.


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.
"* Correcting other susceptible RI-ISI com ponents as necessary.


3.
"* Incorporating the lessons in the plant data base and computer models, if appropriate.


Evaluations within the corrective action program may also include:
"* 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 com ponent's inspection strategy in detecting the failure or nonconforming condition.
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.



The inspection interval would be reduced or the inspection methods adjusted, as ap propriate, when the component (or group of components) experiences repeated fail ures or nonconforming conditions.
Determining the impact of the failure or nonconformance on system or train operability since the previous inspection.



4. The corrective action evaluation should be pro vided to the licensee's PRA and RI-ISI groups so that any necessary model changes and regrouping are done, as appropriate.
Assessing the applicability of the failure or nonconforming condition to other components in the RI-ISI program.



5. The RI-ISI program documents should be revised to document any RI-ISI program changes resulting from the corrective actions taken.
Correcting other susceptible RI-ISI components as necessary.


1.178-24
6. A program is in place that monitors industry find ings.

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



1.178-16
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 components 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.
7. Piping is subject to examination. The examination requirements include all piping evaluated by the risk-informed process and categorized as high safety significant.


The corrective action evaluation should be provided to the licensees PRA and RI-ISI
8. The inspection pr6gram is to be completed during each ten-year inspection interval with the follow ing exceptions.
groups so that any necessary model changes and regrouping are done, as appropriate.


5.
8.1 If, during the interval, a reevaluation using the RI-ISI process is conducted and scheduled items are no longer required to be examined, these items may be eliminated.


The RI-ISI program documents should be revised to document any RI-ISI program changes resulting from the corrective actions taken.
8.2 If, during the interval, a reevaluation using the RI-ISI process is conducted and items must be added to the examination program, those items will be added.


6.
9. Locations selected for successive and additional inspections should be subjected to successive and additional examinations consistent with Section XI
requirements at appropriate intervals.


A program is in place that monitors industry findings.
10. Examination and Pressure Test Requirements.


7.
Pressure testing and VT-2 visual examinations are to be performed on Class 1, 2, and 3 piping systems in accordance with 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 HSS piping and on non-Code LSS pip ing with high failure potential.


Piping is subject to examination.  The examination requirements include all piping evaluated by the risk-informed process and categorized as safety significant.
Examination qualification and methods and per sonnel 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 re pair or replacement activities are to be performed in accordance with the B&PVC Section XI require ments.


8.
12. Records and reports should be prepared and main tained in accordance with the B&PVC Section XI
Edition and Addenda as specified in the licensee's ISI program.


The inspection program is to be completed during each ten-year inspection interval with the following exceptions.
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 infor mation provided and aid in minimizing the time needed for the review process.


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.
4.1 Documentation that Should Be Included in a Licensee's RI-ISI Submittal Table 1 provides an overall summary of the infor mation needed to support a risk-informed ISI submit- tal. References to NRC-approved generic topical re ports that address the methodology and issues requested in a submittal are acceptable. Since topical reports could cover more issues than applied by a li censee or the licensee may elect to deviate from the full body of issues addressed in the topical report, such dis tinctions should be clearly stated. If a licensee refer ences a topical report that has not been approved by the NRC, the time required to review the submittal may be delayed.


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


9.
"
A request to implement a RI-ISI program as an au thorized 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 commit ments made to the NRC.


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.
"
Detailed discussions on each of the following five key principles of risk-informed regulations (see Section 2 of Regulatory Guide 1.174 (Ref. 4) for more details).   
1. The proposed change meets the current regula tions unless it is explicitly related to an alterna tive requested under 10 CFR 50.55a(a)(3)(i), a requested exemption, or a rule change.


10.
2. The proposed change is consistent with the de fense-in-depth philosophy (see detailed dis cussions 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 core damage frequency and/or risk, the in creases should be small and consistent with the guidance in Regulatory Guide 1.174.


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


1.178-25 Section XI, as specified in the licensees ISI program. The pressure testing and VT-2 examinations are also to be performed on non-Code safety-significant piping.  The non- Code safety-significant piping will be treated as ASME Code Class piping for the purposes of examination and pressure testing.
Identification of the aspects of the plant's current requirements that would be affected by the pro posed RI-ISI program. This identification should include all commitments (for example, the IGSCC
inspections and other commitments arising from generic letters affecting piping integrity) that the li censee intends to change or terminate as part of the RI-ISI program.


Examination methods, equipment qualification, personnel qualification, and procedure qualification are to be in accordance with the edition and addenda endorsed by the NRC
1.178-17
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.
Table 1 Documentation Summary Table PRA Quality Address the adequacy of the PRA model used in the calculations.


12.
Address the acceptance guidelines in Regulatory Position 2 of this document and in Regulatory Guide 1.174 (Ref. 4). 
Failure Probability Calcula- Address the methods used to calculate or categorize the failure probability or tions frequency of a piping element. Any use of expert elicitation should be fully documented.


Records and reports should be prepared and maintained in accordance with the B&PVC
Changes in CDF and LERF
Section XI Edition and Addenda as specified in the licensees ISI program.
Address the change in CDF and LERF resulting from changes to the ISI pro gram ISI Systems Identify all the systems inspected based on the current ISI programs and compare the systems for the RI-ISI programs.


4.
Segmentation Identify methods used to segment piping systems, if applicable.


ELEMENT 4:  DOCUMENTATION
Categorization Identify methods used to categorize piping segments and elements as HSS,
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.
LSS, high failure potential, and low failure potential.


4.1 Documentation that Should Be Included in a Licensees 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.
Identify all the HSS-HFP and HSS-LFP elements (format may differ based on decision matrix employed).   
Sampling Method Identify the method used to calculate the number of elements to be inspected.


The following items should be included in the application to implement a RI-ISI program.
Document the method used to establish elements within a lot. Address how this method provides an acceptable level of quality and safety per 10 CFR
50.55a(a)(3)(i). 
Locations of Inspections Provide a system/piping diagram or table that compares the existing ISI loca tions of inspection with the RI-ISI location of inspection.



"Address the reasons for the changes.
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.



Failure Probabilities Identify the methods used to arrive at the failure probabilities for piping seg ments.
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.
Performance Monitoring Discuss the performance goals and corrective action programs.
 
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).
Periodic Reviews Identify the frequency of performance monitoring and activities in support of the RI-ISI program. Address consistency with other RI programs (e.g.,
3.
Maintenance Rule, IST, Tech Specs).
QA Program Describe the QA program used to ensure proper implementation of RI-ISI
process and categorization and consistency with other RI programs.


The proposed change maintains sufficient safety margins (see detailed discussions in Section 2.2.1.2 in Regulatory Guide 1.174).
Expert Elicitation Identify any use of the expert elicitation process to estimate a failure proba bility for piping. Address the reasons why an expert elicitation was required, provide all supporting information used by the experts, document the conclu sions, and address how the results will be incorporated in an industry data base or computer code, or why it is not necessary to make the findings avail able to the industry.
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.
Each weld to be inspected Identify: 1. The inspection method to be used
2. The applicable degradation mechanism to be inspected, and
3. The frequency of inspection Address each of the key prin- Verify compliance with applicable regulations, defense-in-depth, safety mar ciples and the integrated deci- gins, etc.


5.
sionmaking guidelines (e.g.,
Regulatory Position 2.3)
Implementation and monitor- Address the acceptance guidelines outlined in Regulatory'Position 3 of this ing program regulatory guide.


The impact of the proposed change should be monitored using performance measurement strategies.
1.178-18



SA summary of events involving piping failures that have occurred at the plant or similar plants. Include in the summary any lessons learned from those events and indicate actions taken to prevent or minimize the potential for recurrence of the events.
Identification of the aspects of the plants 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.
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 imple menting a change in the ISI program based on pro babilistic and traditional insights. These should include a description of the integrated decision making process and criteria used for categorizing the safety significance of piping segments, a de scription 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 mak ing the decisions, and the guidelines for making those decisions.
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 r~iewers of the program. These results should include the following information.
The results of the licensees 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 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,
A list of each segment, including the number of welds, weld type and properties of the weld ing material and base metal, the failure poten tial, CDF, CCDF/CCDP, LERF, CLERF, im portance measure results (RAW, F-V, etc.) and justification of the associated threshold val ues, degradation mechanism, test and inspec tion intervals used in or in support of the PRA,  
LERF, CLERF, importance measure results (risk achievement worth (RAW),
etc. Results from other methods used to de velop the consequences and categorization of each segment (or weld) should be documented in a similar level of detail. (NOTE: Table 2 provides an example of a summary of possible methods for obtaining failure probabilities based on specified degradation mechanisms.
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.
The staff recommends that licensees provide such a table with supporting discussions.)
For the selected limiting locations, provide ex amples of the failure mode, failure potential, failure mechanism, weld type, weld location, and properties of the welding material and base metal. Provide a detailed description and justification for the number of elements to be inspected.


It was submitted for staff review in the context of its use in categorizing SSCs with respect to special treatment requirements (i.e.,
-
supporting NRCs 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.
The degradation mechanisms for each seg ment (if segments contain welds exposed to different degradation mechanism, for each weld) used to develop the failure potential of each segment.


1.178-27
-

Equipment assumed to fail as a direct or indi rect consequence of each segment's failure (if segments contain welds with different failure consequences, for each weld).   
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 description of how the impact of the change between the current Section XI and the pro posed RI-ISI programs is evaluated or bounded, and how this impact compares with the risk guidelines in Section 2.2.2.2 of Regu latory Guide 1.174.



The means by which failure probabilities or fre quencies or potential were determined. The data should be provided in the submittal for analyses that rely on operational data for determining failure frequencies or potential. Reliance on fracture me chanics structural reliability and risk analysis codes should be documented and validated. Re liance on the expert elicitation process should be fully documented. (NOTE: Expert elicitation is only used if data are not sufficient to estimate the failure probability and frequency of a piping seg ment. Data assessment is not an expert elicitation process and can normally be performed by plant personnel.)
A detailed description and justification for the number of elements to be inspected.
A description of the PRA used for the categoriza tion process and for the determination of risk im pact, 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 sup porting the ISI submittal. The key assumptions used in the PRA that impact the application (i.e.,
licensee voluntary actions), elements of the moni toring program, and commitments made to support the application should be addressed.



"
Equipment assumed to fail as a direct or indirect consequence of each segments failure (if segments contain welds with different failure consequences, for each weld).
If the submittal includes modified inspection inter vals, the methodology and results of the analysis should be submitted.

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.



"* A description of the implementation, performance monitoring, and corrective action strategies and programs in sufficient detail for the staff to under stand the new ISI program and its implications.
The means by which failure probabilities, frequencies, or potential were determined.



1.178-19
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.
Applicable documentation discussed under the Cumulative Risk documentation for submittal in Section 1.3 of Regulatory Guide 1.174 (Ref. 4). 
"
Reference to NRC-approved topical reports on im plementing a RI-ISI and supporting documents.



Variations from the topical reports and supporting documents should be clearly identified.
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.
Detailed justification for the proposed regulatory action (e.g., how the proposed program meets the requirements set in 10 CFR 50.55a(a)(3)(i)).   
4.2 Documentation That Should Be Available Onsite for Inspection The licensee should maintain at its facility the tech nical 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.



Plant and applicable industry data used in support of the RI-ISI program. All analyses and assump tions used in support of the RI-ISI program and communications with outside organizations sup porting the RI-ISI program (e.g., use of peer and independent reviews, use of expert contractors). 
If the submittal includes modified inspection intervals, the methodology and results of the analysis should be submitted.
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.


1.178-28
Documentation of the plant's baseline PRA used to support the ISI submittal should be of sufficient de tail to allow an independent reviewer to ascertain whether the PRA reflects the current plant configu ration and operational practices commensurate with the role the PRA results play in the integrated decisionmaking process. In addition to documen tation on the PRA itself, analyses performed in support of the IST submittal should be documented in a manner consistent with the baseline documen tation. Such analyses may include:

-
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.
The process used to identify initiating events developed in support of the RI-ISI submittal and the results from the process.



-
Applicable documentation discussed under the cumulative risk documentation for submittal in Section 1.3 of Regulatory Guide 1.174.
Any event and fault trees developed during the RI-ISI submittal preparation.



-
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.
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.


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 plants 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:
-

The techniques used to identify and quantify human actions.
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).
The data used in any uncertainty calculations or sensitivity calculations, consistent with the guidance provided in Regulatory Guide 1.174.



-
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.
How uncertainty was accounted for in the seg ment categorization, and the sensitivity stud ies performed to ensure the robustness of the categorization.



Detailed results of the inspection program corre sponding to the ISI inspection records described in the implementation, performance monitoring, and corrective action program accompanying the RI
Documentation of the plants 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
ISI submittal.
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
For each piping segment, information on weld type, weld location, and properties of welding ma terial and base metal.
submittal and the results from the process.



For each piping segment, information regarding the process and assumptions used to develop fail ure mode and failure potential (frequency/proba bility), in addition to the identification of the fail ure mechanism.
Any event and fault trees developed during the RI-ISI submittal preparation.


1.178-29
K,

1.178-20
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.
K



Table 2 Example of a Summary of Methods Used To Estimate Piping Failure Probabilities for Risk Categorization Failure Mechanism Methods for Estimating Probability Name of Mechanism Contributing Factors Failure Mode Stainless Steel Carbon Steels Other Materials Thermal Striping Crack Code Name Code Name High Cycle Flow Induced Vibration Initiation Failure Fatigue Mechanical Vibration Crack Code Name Code Name Database Growth Thermal Stratification Crack Code Name Code Name Low Cycle Heat-up and Cool-down Initiation Failure Fatigue Thermal Cycling Crack Code Name Code Name Database Growth Coolant Chemistry Crack Code Not Corrosion Crevice Corrosion Initiation Name Applicable Failure Cracking Susceptible Material Database High Stresses Crack Code Not (Residual, Springing)
The techniques used to identify and quantify human actions.
Growth Name Applicable Flow Accelerated. Corrosion Wall Name of Name of Failure Wastage Microbiologically Ind. Corr.



Thinning Code Code Database
The data used in any uncertainty calculations or sensitivity calculations, consistent with the guidance provided in Regulatory Guide 1.174.
_
_Pitting and/or Wear
"Other Creep Damage Miscellaneous Failure Failure Failure Mechanisms Thermal Aging Modes Database Database Database Irra



====d. Embrittlement I ====
How uncertainty was accounted for in the segment categorization, as well as the sensitivity studies performed to ensure the robustness of the categorization.
I
1-1 Co
00
k4



REFERENCES
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.
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,"
For each piping segment, information on weld type, weld location, and properties of welding material and base metal.
SECY-95-280, November 27, 1995.1
3. USNRC, "Standard Review Plan for the Review of Risk-Informed Inservice Inspection of Piping,"
NUREG-0800, Section 3.9.8, September 1998.2
4. USNRC, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," Regulatory Guide 1.174, July 1998.2
5. USNRC, "An Approach for Plant-Specific, Risk Informed Decisionmaking: Inservice Testing,"
Regulatory Guide 1.175, August 1998.2
6. USNRC, "An Approach for Plant-Specific, Risk Informed Decisionmaking: Graded Quality Assur ance," Regulatory Guide 1.176, August 1998.2
7. USNRC, "An Approach for Plant-Specific, Risk Informed Decisionmaking: Technical Specifica tions," Regulatory Guide 1.177, August 1998.2
8. USNRC, "Standard Review Plan for Risk Informed Decision Making," Standard Review Plan, NUREG-0800, Chapter 19, July 1998.3 lCopies 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 mailing address is Mail Stop LL-6, Washington, DC 20555;
telephone (202) 634-3273; fax (202) 634-3343.



2Single copies of regulatoryguides, both active and draft, and standard review plans may be obtained free of charge by writing the Reproduc tion and Distribution Services Section, OCIO, USNRC, Washington, DC 20555-0001, or by fax to (301) 415-2289, or by e-mail to GRWI@NRC.GOV. Active guides may also be purchased from the National Technical Information Service on a standing order basis.
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.
Details on this service may be obtained by writing NTIS, 5285 Port Royal Road, Springfield, VA22161. Copies of active and draft guides 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 mailing address is Mail Stop LL-6, Washington, DC 20555; tele phone (202) 634-3273; fax (202) 634-3343.


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
3Copies are available at current rates from the U.S. Government Printing Office, RO. Box37082, Washington, DC20402-9328 (tele phone (202) 512 - 2249); or from the National Technical Information Service by writing NTIS at 5285 Port Royal Road, Springfield, VA
(301)415-2289; email <DISTRIBUTION@NRC.GOV>. Copies are available for inspection or copying for a fee from the NRC
22161. Copies 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 mailing address is Mail Stop LL-6, Washington, DC  
Public Document Room at 11555 Rockville Pike (first floor), Rockville, MD; the PDRs mailing address is USNRC PDR,
20555; telephone (202) 634-3273; fax (202) 634-3343.
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 NRCs 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 PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone
9. USNRC, "Standard Review Plan for Risk Informed Decision Making: Inservice Testing,"
(301)415-4737 or 1-(800)397-4209; fax (301)415-3548; e-mail <PDR@NRC.GOV>.
Standard Review Plan, NUREG-0800, Chapter
1.178-30
3.9.7, August 1998.3
REFERENCES
10. USNRC, "Standard Review Plan for Risk Informed Decision Making: Technical Specifica tions," Standard Review Plan, NUREG-0800,
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,
Chapter 16.1, August 1998.3
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
11. American Society of Mechanical Engineers, "Case N-560, Alternative Examination Requirements for Class 1, Category B-J Piping Welds Section XI,  
Division 1," August 9, 1996.4
12. American Society of Mechanical Engineers, "Case N-577, Risk-Informed Requirements for Class 1,  
2, and 3 Piping, Method A, Section XI, Divi sion 1," September 2, 1997.4
13. American Society of Mechanical Engineers, "Case N-578, Risk-Informed Requirements for Class 1,  
2, and 3 Piping, Method B, Section XI, Divi sion 1," September 2, 1997.4
14. Electric Power Research Institute, "PSA Applica tions Guide," EPRI TR-105396, August 1995.5
15. Electric Power Research Institute, "Risk-Informed Inservice Inspection Evaluation Procedure," EPRI
TR-106706, June 1996.5
16. Westinghouse Energy Systems, "Westinghouse Owners Group Application of Risk Informed Methods to Piping Inservice Inspection Topical Report," WCAP-14572, Revision 1, October
1997.1
17. Westinghouse Energy Systems, "Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inser vice Inspection," WCAP-14572, Revision 1, Sup plement 1, October 1997.1
18. T.V. Vo et al., "A Pilot Application of Risk-In formed Methods To Establish Inservice Inspection Priorities for Nuclear Components at Surry Unit 1 Nuclear Power Station," USNRC, NUREG/
CR-6181, Revision 1, February 1997.3
4Copies may be obtained from the American Society of Mechanical Engineers, 345 East 47th Street, New York, NY 10017.


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 PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301)415-4737 or 1-
5Copies may be obtained from the EPRI Distribution Center, 207 Coggins Drive, P.O. Box 23205, Pleasant Hill, CA 94523.
(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 PDRs 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.,
1.178-22 K
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
19. American Society of Mechanical Engineers,
Section III," USNRC,  Regulatory Guide 1.84, Revision 32, June 2003.2 Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI,
"Rules for Inservice Inspection of Nuclear Power Plant Components," ASME Boiler and Pressure Vessel Code, Section XI, 1989 Edition, New
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:
York.4
Inservice Testing," USNRC, August 1998.2 Regulatory Guide 1.176, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:
20. USNRC, "Design and Fabrication Code Case Ac ceptability, ASME Section III, Division I," Regu latory Guide 1.84, Revision 30, October 1994.2
Graded Quality Assurance," USNRC, August 1998.2
21. USNRC, "Materials Code Case Acceptability, ASME Section III, Division 1," Regulatory Guide  
1.85, Revision 30, October 1994.2
22. USNRC, "Inservice Inspection Code Case Accept ability, ASME Section XI, Division 1," Re*gulatory Guide 1.147, Revision 11, October 1994.2  
23. M.A. Meyer and J.A. Booker, "Eliciting and Ana lyzing Expert Judgement," NUREG/CR-5424 (Prepared for the NRC by Los Alamos National Laboratory), USNRC, January 1990.3
24. J.P. Kotra et al., "Branch Technical Position on the Use of Expert Elicitation in the High-Level Radio active Waste Program," NUREG-1563, USNRC,  
November 1996.3 REGUILATORY ANALYSIS
A draft regulatory analysis was published with the draft of this guide when it was published for public comment (Task DG-1063, October 1997). No changes were necessary, so a separate regulatory analysis for Regulatory Guide 1.178 has not been prepared. A copy of the draft regulatory analysis is available for inspec tion or copying for a fee in the NRC's Public Document Room at 2120 L Street NW., Washington, DC, under Task DG-1063.


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),
1.178-23
Rockville, MD; the PDRs 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
UNITED STATES
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 PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone
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Latest revision as of 02:06, 17 January 2025

(Draft Was Issued as DG-1063), an Approach for Plant-Specific Risk-Informed Decisionmaking Inservice Inspection of Piping
ML003740181
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Issue date: 09/30/1998
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Office of Nuclear Regulatory Research
To:
References
RG-1.178
Download: ML003740181 (24)
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U.S. NUCLEAR REGULATORY COMMISSION

REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

FOR TRIAL USE

REGULATORY GUIDE 1.178 (Draft was Issued as DG-1063)

AN APPROACH FOR PLANT-SPECIFIC RISK-INFORMED

DECISIONMAKING INSERVICE INSPECTION OF PIPING

A. INTRODUCI7ION

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 inspection of piping (ISI),

with its companion Standard Review Plan, Section

3.9.8 of NUREG-0800 (Ref. 3), and other regulatory documents (Refs. 4-10), implement, in part, the Com mission's policy statement and the staff's framework for incorporating risk insights into the regulation of nu clear power plants.

In 1995 and 1996, the industry developed a number of documents addressing the increased use of PRA in nuclear plant regulation. The American Society of Me chanical Engineers (ASME) initiated Code Cases N-560 (Ref. 11), N-577 (Ref. 12), and N-578 (Ref. 13)

that address the importance categorization and inspec- tion of plant piping using risk insights. The Electric Power Research Institute (EPRI) published its "PSA

Applications Guide" (Ref. 14) to provide utilities with guidance on the use of PRA information for both regu latory and nonregulatory applications. The Nuclear En ergy Institute (NEI) has been developing guidelines on risk-based ISI and submitted two methods, one devel oped by EPRI (Ref. 15) and the other developed by the ASME research and the Westinghouse Owners Group (Refs. 16-17), for staff review and approval.

Given the recent initiatives by the ASME in devel oping Code Cases N-560, N-577, and N-578, it is an ticipated that licensees will request changes to their plant's design, operation, or other activities that require NRC approval to incorporate risk insights into their ISI

programs (known as risk-informed inservice inspec tion programs, RI-ISI). Until the RI-ISI is approved for generic use, the 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)(3Xi) in Section

50.55a, "Codes and Standards," of 10 CFR Part 50,

"Domestic Licensing of Production and Utilization Fa cilities," providing an acceptable level of quality and safety. As always, licensees should identify how the USNRC REGULAIORY GUItES

The guides ma Issued In On following ton broad divisions:

Regulator Guides ae Issued to describe and make available to fte publlc such Informs Ilonesmethodsaoceptabletothe NRCstaffforimplemen ngepedflc partsof*teom-

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10. General will be acceptable If tNhy provide a basisforthefindings requisite to the Isusnc orcon orufnce a permit or lkers by 1he Co--on.

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September 1998

\\

chosen approach, methods, data, and criteria are ap propriate for the decisions they need to make.

In October 1997, the Commission published a draft of this regulatory guide for public comment. This guide's principal focus is on the use of PRA findings and risk insights in support of proposed changes to a plant's design, operations, and other activities that re quire NRC approval. 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 reg ulatory guide describes methods acceptable to the NRC

staff for integrating insights from PRA techniques with traditional engineering analyses into ISI programs for piping.

The draft guide, DG- 1063, was discussed during a public workshop held on November 20-21, 1997, and was peer reviewed. While the public comments and peer review of the document were positive, the staff has not had an opportunity to apply the guidance to indus try's pilot plants. Therefore, this regulatory guide is be ing issued for trial use on the pilot plants. This regula tory guide does not establish any final staff positions, and may be revised in response to experience with its use. As such, this trial regulatory guide does not estab lish a staff position for purposes of the Backfit Rule, 10

CFR 50.109, and any changes to this regulatory guide prior to staff adoption in final form will not be consid ered to be backfits as defined in 10 CFR 50.109(a)(1).

This will ensure that the lessons learned from regulato ry review of the pilot plants are adequately addressed in this document and that the guidance is sufficient to en hance regulatory stability in the review, approval, and implementation of proposed RI-ISI programs.

In the interest of optimizing limited resources, the appendices that were in DG- 1063 will be incorporated in a future NUREG report. The appendices have been deleted from this guide to focus the NRC staff's limited resources on the review and approval of the pilot plant applications and the topical reports submitted in sup port of the pilot plant analyses. Staff positions on the methodologies will be provided in the staff's safety evaluation of the topical reports and pilot plant submit tals. This process would minimize resources needed to update the RG to address the different methods pro posed by the industry.

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 regula- tory decisionmaking. In August 1995, the NRC

adopted a policy statement regarding the expanded use of PRA (Ref. 1). In part, the policy statement states that:

t The use of PRA technology should be in creased in all regulatory matters to the ex tent supported by the state-of-the-art in PRA methods and data and in a manner that complements the deterministic approach and supports the NRC's traditional philoso phy of defense-in-depth.

  • PRA and associated analyses (e.g., sensi tivity studies, uncertainty analyses, and im portance measures) should be used in regu latory matters, where practical within the bounds of the state-of-the-art, to reduce un necessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal of addi tional regulatory requirements in accor dance with 10 CFR 50.109 (Backfit Rule).

Appropriate procedures for including PRA

in the process for changing regulatory re quirements should be developed and fol lowed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.

"* PRA evaluations in support of regulatory decisions should be as realistic as practica ble and appropriate supporting data should be publicly available for review.

"* The Commission's safety goals for nuclear power plants and subsidiary numerical ob jectives are to be used with appropriate con sideration of uncertainties in making regu latory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.

In its approval of the policy statement, the Com mission articulated its expectation that implementation of the policy statement will improve the regulatory pro cess in three areas: foremost, through safety decision making 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 state ment, the staff developed a regulatory framework that incorporates risk insights. That framework was articu-

1.178-2 I'

/i

lated in a November 27, 1995, paper (SECY-95-280)

to the Commission. This regulatory guide, which ad dresses ISI programs of piping at nuclear power plants, j

is part of the implementation of the Commission's policy statement and the staff's framework for incorpo rating risk insights into the regulation of nuclear power plants. This document uses the knowledge base docu mented in Revision 1 of NUREG/CR-6181 (Ref. 18),

and it reflects the experience gained from the ASME

initiatives (Code Case development and pilot plant ac tivities).

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 2000 reactor-years of plant operating experience and extensive research in the areas of material sciences, aging phenomena, and in spection techniques. This information, combined with modem 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 compo nents are found in 10 CFR 50.55a and the General De

2 sign Criteria listed in Appendix A to 10 CFR Part 50.

These requirements are throughout the General Design Criteria, such as in Criterion I, "Overall Require ments," Criterion II, "Protection by Multiple Fission Product Barriers," Criterion III, "Protection and Reac tivity Control Systems," and Criterion IV, "Fluid Sys tems."

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) (Ref. 19) 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 50.55

a. ISI

programs are intended to address all piping locations that are subject to degradation. Incorporating risk in sights into the programs can focus inspections on the more important locations and reduce personnel expo sure, while at the same time maintaining or improving public health and safety. The justification for any re duction 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.

As a result of the above insights, more efficient and technically sound means for selecting and scheduling ISIs of piping are under development by the ASME

(Refs. 11-13).

When categorizing piping segments in terms of their contribution to risk, it is the responsibility of a li censee to ensure that the categorization of piping seg ments and the resulting inspection programs are consis tent 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 (Ref. 4). This regulatory guide augments the guidance presented in Regulatory Guide 1.174 by pro viding guidance specific to incorporating risk insights to inservice inspection programs of piping.

Purpose of the Guide Consistent with Regulatory Guide 1.174 (Ref. 4),

this regulatory guide focuses on the use of PRA in sup port of a risk-informed ISI program. This guide pro vides guidance on acceptable approaches to meeting the existing Section XI requirements for the scope and frequency of inspection of ISI programs. Its use by li censees 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 com pliance with the requirements of 10 CFR 50.55a and with Section XI of the ASME Boiler and Pressure Ves sel 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 incorporat ing insights from probabilistic risk and traditional anal ysis calculations, supplemented with operating reactor data. Licensees who propose to apply risk-informed ISI

programs would amend their final safety analysis re port (FSAR, Sections 5.3.4 and 6.6) accordingly. A

Standard Review Plan (SRP) (Ref. 3) has been prepared for use by the NRC staff in reviewing RI-ISI applica tions.

This document addresses risked-informed meth ods to develop, monitor, and update more efficient ISI

programs for piping at a nuclear power facility. This guidance does not preclude other approaches for incor porating 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 regula tions on the basis of their accumulated plant experience and knowledge. This document addresses risk informed approaches that are consistent with the basic

1.178-3

elements identified in Regulatory Guide 1.174 (Ref. 4).

In addition, this document provides guidance on the following for the purposes of RI-ISI.

"

Estimating the probability of a leak, a leak that pre vents the system from performing its function (dis abling 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, re duction of unnecessary radiation exposure to per sonnel,

"* Determining the risk impact of changes to ISI pro grams,

"* Capturing deterministic considerations in the re vised ISI program, and

"

Developing an inspection program that monitors the performance of the piping elements for consis tency with the conclusions from the risk assess ment.

Given the recent initiatives by the ASME in devel oping Code Cases N-560, N-577, and N-578 (Refs.

11-13), it is anticipated that licensees will request changes to their plant's design, operation, or other ac tivities that require NRC approval to incorporate risk insights in their ISI programs (RI-ISI). Until the RI-ISI

is approved for generic use, the staff anticipates that li censees will request changes to their ISI programs by requesting NRC approval of a proposed inspection pro gram 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 adequate ly reflect the risk implications of piping failure, both partial and full-scope RI-ISI programs are acceptable to the NRC staff.

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.

Full Scope: Afull scope RI-ISI program evaluates the piping in a plant as being either high or low safety significant. 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 compo nents that are relied upon to remain functional during and following design basis events to en sure the integrity of the reactor coolant pres sure boundary, the capability to shut down the reactor and maintain it in a safe shutdown con dition, 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 com ponents

"* That are relied upon to mitigate accidents or transients or are used in plant emergen cy 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 accep tance guidelines and key principles of Regulatory Guide 1.174 (Ref. 4).

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 regula tory 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 be cause 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 de gradation mechanisms (stress corrosion cracking) usu ally occur at welds. Nevertheless, there are other degra dation mechanisms such as flow-assisted-corrosion (e.g., erosion-corrosion) and thermal fatigue that occur independent of welds.

1Generally, ASME Code Class 1 includes all reactor pressure bound.

ary (RCPB) components. ASME Code Class 2 generally includes sys tems or portions of systems important to safety that are designed for post-accident containment and removal of heat and fission products.

ASME Code Cass 3 generally includes those system components or portions of systems important to safety that are designed to provide cooling water and auxiliary feedwater for the front-line systems.

1.178-4

To ensure that the proposed RI-ISI program would provide an acceptable level of quality and safety, the li censee should use the PRA to identify the appropriate scope of the piping segments to be included in the pro gram. In addition, licensees implementing the risk-in formed process may identify piping segments catego rized as high safety-significant (HSS) that are not currently subject to the traditional Code requirements (e.g., outside the Code boundaries, including Code ex empt piping) or are not being inspected to a level that is commensurate with their risk significance. In this con text, HSS refers to a piping segment that has a relatively high contribution to risk. PRA systematically takes credit for systems with non-Code piping that provide support, act as alternatives, and act as backups to those systems with piping that are within the scope of the cur rent Section XI of the Code.

Organization and Content This regulatory guide is structured to follow the general four-element process for risk-informed ap plications discussed in Regulatory Guide 1.174 (Ref.

4). The Discussion section summarizes the four element process developed by the staff to evaluate pro posed changes related to the development of a RI-ISI

program. Regulatory Position 1 discusses an accept able approach for defining the proposed changes to an ISI program. Regulatory Position 2 addresses, in gen eral, the traditional and probabilistic engineering eval uations performed to support RI-ISI programs and pre sents the risk acceptance goals for determining the acceptability of the proposed change. Regulatory Posi tion 3 presents one acceptable approach for implement ing and monitoring corrective actions for RI-ISI pro grams. 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 ac ceptable 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. Regulatory Guide 1.174 de scribes a general approach to risk-informed regulatory decisionmaking and discusses specific topics common to all risk-informed regulatory applications. Topics ad dressed include:

PRA quality-data, assumptions, methods, peer review,

"* PRA scope-internal and external event initiators, at-power and shutdown modes of operation, con sideration of requirements for Level 1, 2, and 32 analyses,

"* Risk metrics--core damage frequency, large early release frequency and importance measures,

Sensitivity and uncertainty analyses.

To the extent that a licensee elects to use PRA as an element to enhance or modify its implementation of ac tivities 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 doc ument are covered by the requirements of 10 CFR

Part 50, which were approved by the Office of Manage ment and Budget (OMB),

approval number

3150-0011. The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of in formation unless it displays a currently valid OMB con trol number.

Abbreviations and Definitions ASME

American Society of Mechanical Engi neers 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 In the context of this regulatory guide, expert elicitation is a process used to esti mate failure rates or probabilities of pip ing when data and computer codes are un available for the intended purpose. It is a process used to estimate the failure proba bility and the associated uncertainties of the material in question under specified degradation mechanisms. For example, if a structural mechanics code is not quali fied to calculate the failure probability of plastic piping and no data are available to estimate its failure probability, experts in plastic piping and their failure may be asked to estimate the failure probabilities.

If applicable industry data are available, an expert elicitation process would not be needed.

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

1.178-5

Expert Panel FSAR

HSS

IGSCC

Importance Measures ISI

IST

LERF

LSS

NDE

NEI

NRC

PRA

PSA

RCPB

Normally refers to plant personnel exper ienced in operations, maintenance, PRA,

ISI programs, and other related activities and disciplines that impact the decision under consideration.

Final Safety Analysis Report High safety significance Intergranular stress corrosion cracking Used in PRA to rank systems or compo nents in terms of risk significance Inservice inspection Inservice testing Large early release frequency Low safety significance Nondestructive examination Nuclear Energy Institute Nuclear Regulatory Commission Probabilistic risk assessment Probabilistic safety assessment Reactor coolant pressure boundary RI-ISI

Staff Sensitivity Studies SRP

SRRA

SSCs Tech Spec

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. Figure I il lustrates the five key principles involved in the inte grated decisionmaking process; they are described in detail in Regulatory Guide 1.174 (Ref. 4). In addition, Regulatory Guide 1.174 describes a four-element pro cess for evaluating proposed risk-informed changes as illustrated in Figure 2.

Figure 1 Principles of Risk-Informed Integrated Decisionmaking Figure 2 Principal Elements of Risk-Informed, Plant-Specific Decisionmaking

1.178-6 Risk-informed inservice inspection Refers to NRC employees Varying parameters to assess impact due to uncertainties Standard Review Plan Structural reliability/risk assessment (re fers to fracture mechanics analysis)

Structures, systems and components Technical specifications

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 exemp tion 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 core damage frequency or risk, the increases should be small and consistent with the intent of the Com mission's Safety Goal Policy Statement. (Regula tory Position 2.2)

5. The impact of the proposed change should be mon itored by using performance measurement strate gies. (Regulatory Position 3)

The individual principles are discussed in detail in Regulatory Guide 1.174.

Section 2 of Regulatory Guide 1.174 describes a four-element process for developing risk-informed reg ulatory changes. An overview of this process is given here and illustrated in Figure 2. 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 pro posed change to the ISI program as defined in Element I depends on both the analysis performed in Element 2 and the definition of the implementation of the ISI pro gram performed in Element 3. While ISI is, by its na ture, 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 moni toring is performed to make sure that these margins are not eroded. Element 4 involves preparing the documen tation to be submitted to the NRC and to be maintained by the licensee for later reference.

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 de- scribing the scope of ISI piping that would be incorpo rated in the overall assessment and how the inspection of this piping would be changed. Also included in this ele ment is identification of supporting information and a proposed plan for the licensee's interactions with the NRC throughout the implementation of the RI-ISI.

1.1 Description of Proposed Changes A full description of the proposed changes in the ISI

program is to be prepared. This description should in dude:

"

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 pro posed changes, the licensee should also confirm that the plant's design and operation is in accordance with its current requirements and that engineering infor mation 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 di rectly and indirectly involved with the proposed changes. Any piping not presently covered in the plant's ISI program but categorized as high safety significant (e.g., through an integrated decisionmak ing process using PRA insights) should be identified and appropriately addressed. In addition, the particu lar 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 in formation.

"* 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 licen sees to report program activities and guidance on formal NRC review of changes made to RI-ISI programs.

The licensee should implement a process for deter mining 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 changes should be evaluated using the change mechanisms

1.178-7

described in the applicable regulations (e.g., 10 CFR

50.55a, 10 CFR 50.59) to determine whether NRC re view 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 imple mentation.

2. ELEMENT 2: ENGINEERING ANALYSIS

As part of defining the proposed change to the licens ee's ISI program, the licensee should conduct an engi neering evaluation of the proposed change, using and in tegrating 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 (Ref 4). Regulatory Guide 1.174 provides general guidance for performing this evaluation, which is supplemented by the RI-ISI guidance herein.

Figure 3 Element 2 The regulatory issues and engineering activities that should be considered for a risk-informed ISI pro gram are summarized here. For simplicity, the discus sions are divided into traditional and PRA analyses (see Figure 3). Regulatory Position 2.1 addresses the tradi tional 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 tradi tional 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 core damage frequency 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 Regula tory 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 margins, assessment of failure potential of piping segments, and assessment of pri mary and secondary effects (failures) that result from piping failures.

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

1.

2.

3.

4.

5.

Assess compliance with applicable regulations, Perform defense-in-depth evaluation, Perform safety margin evaluation, Define piping segments, Assess failure potential for the piping segment (from leaks to breaks),

6. Assess the consequences (both direct and indirect)

of piping segment failure,

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

8.

9.

Develop an inspection program, 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, de fense 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 com promise compliance with the regulations. The evalua tion should consider the appropriate requirements in the licensing basis and applicable regulatory guidance.

Specifically, the evaluation should consider

1.178-8

10 CFR 50.55a

Appendix A to 10 CFR Part 50

-

Criterion I, "Overall Requirements"

-

Criterion II, "Protection of Multiple Fission Product Barriers"

-

Criterion III, "Protection and Reactivity Con trol Systems"

-

Criterion IV, "Fluid Systems," etc

ASME Boiler and Pressure Vessel Code,Section XI (10 CFR Part 50.55a)

a Regulatory Guide 1.84 (Ref. 20)

Regulatory Guide 1.85 (Ref. 21)

Regulatory Guide 1.147 (Ref. 22)

Appendix B to 10 CFR Part 50.

In addition, the evaluation should consider wheth er the proposed changes have affected license commit ments. A broad review of the licensing requirements and commitments may be necessary because proposed ISI program changes could affect issues not explicitly stated in the licensee's FSAR or ISI program documen tation.

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 accept able RI-ISI approaches presented in this guide may be proposed by licensees so long as supporting informa tion is provided that demonstrates that the key prin ciples 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 pro gram.

NRC-endorsed ASME Code Cases that apply risk informed ISI programs will be consistent with this reg ulatory guide in that they encourage the use of risk in sights in the selection of inspection locations and the use of appropriate and possibly enhanced inspection techniques that are appropriate to the failure mecha nisms that contribute most to risk.

2.1.2 Defense-in-Depth Evaluation As stated in Regulatory Guide 1.174 (Ref. 4), the engineering analysis should evaluate whether the im pact of the proposed change in the ISI program (indi- vidually 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 equip ment and human performance. Where a comprehensive risk analysis can be done, it can be used to help deter mine the appropriate extent of defense-in-depth (e.g.,

balance among core damage prevention, containment failure, and consequence mitigation) to ensure protec tion of public health and safety. Where a comprehen sive risk analysis is not or cannot be done, traditional defense-in-depth consideration should be used or main tained 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 dam age, 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, Reference 4, for addtional guidance).

An important element of defense in depth for RI

ISI is maintaining the reliability of independent barri ers to fission product release. Class I piping (primary coolant system) is the second boundary between the ra dioactive fuel and the general public. If a RI-ISI pro gram categorized, for example, all the hot and cold legs of the primary system piping as LSS and calculated that, with no inspections, the frequency of leaks would not 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 accom panying engineering assumptions are intended to ac count 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

1.178-9

requirements, the licensee must ensure that the uncer tainties are adequately addressed. The quantification of uncertainties would likely require supporting sensitiv ity 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 appropri;

ate for evaluating whether sufficient safety margins would be maintained if the proposed change were im plemented. An acceptable set of guidelines for making that assessment are summarized below. Other equiva lent 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 margin to account for analysis and data uncer tainty.

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; dif ferent criteria or definitions can be applied to each pip ing segment. One acceptable method is to identify seg ments 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 de fined 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 seg ments.

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 spe cific 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. Figure 4 identifies the three means for estimating the failure potential of a piping segment: data, fracture mechanics computer codes, and the expert elicitation process. 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 dy namic loads, flaw size, flaw distribution, inspection pa rameters, 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 defini tion) that will be needed to support the PRA or risk as sessment. The failure potential used in or in support of EsTrMATING FAM- URE P~l

1FRATUE

iEXPERT

I

I ELICITATIO

CS

PROCESS

1 j~CODEIS 1 IFNED)

(IFN

J

DE]

Figure 4 Estimating Failure Potential of Piping Segments

1.178-10

the analysis should be appropriate for the specific envi ronmental conditions, degradation mechanisms, and failure modes for each piping location and break size (e.g., leak, disabling leak, break). When data are ana lyzed to develop a categorization process relating de gradation 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 (Ref. 23, 24).

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, varia tion in steady-state and transient loads, availability and accuracy of plant operating history, availability of in spection 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 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 clas sification of the risk significance of each segment. Ref erencing applicable generic topical reports approved by the NRC is one acceptable means to standardize the process. This review should be documented and a sum mary discussion of the review should be included in the submittal. When new computer codes are used to de velop quantitative estimates, the techniques should be verified and validated against established industry codes and available data. When data are used to evalu ate the likelihood of piping failures, the data should be submitted to the NRC or ieferenced by an NRC-ap proved topical report. As stated in Regulatory Guide

1.174 (Ref. 4), "data, methods, and assessment criteria used to support regulatory decisionmaking must be scrutable and available for public review." It is the re sponsibility of the licensee to provide the data, meth ods, and justification to support its estimation of the failure potential of piping segments. Since conse quences of and potential for piping failures could differ for leaks, disabling leaks, and breaks, the failure poten tial for all three break types should be addressed.

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 ac complished by performing a detailed walkdown of a nuclear power facility's piping network. Assessment of internal and external events, including resulting pri mary and secondary effects of piping failures (e.g.,

leaks, disabling leaks, and breaks) are important pa rameters to the risk-informed program (see Figure 5).

Leaks can result in failures of electrical components caused by jet impingement. Disabling leaks and full breaks can lead to a loss of system function, flooding induced damage, and initiating events. Full breaks can lead to damage resulting from pipe whip, as well as flooding and initiating events. Each of these break types has its associated failure potential that is evalu ated in Regulatory Position 2.1.5. A failure modes and consequence assessment is performed to identify the potential failures, from piping leaks to breaks. Internal flooding PRAs can identify the impact of jet impinge ment and flooding to the RI-ISI program. The failures are used as input to the risk analysis. Alternative meth ods for evaluating consequences should be submitted to the NRC for review and approval. These evaluations are expected to provide information for the conse quence analysis. They are not intended to be used in lieu of the plant licensing basis.

2.1.7 Probabilistic Fracture Mechanics Evaluation When implementing probabilistic fracture me chanics computer programs that estimate structural reliability and are used in risk assessment of piping, or other analytic methods for estimating the failure poten tial of a piping segment, some of the important parame ters that need to be assessed in the analysis include the identification of structural mechanics parameters, deg radation mechanisms, design limit considerations, op erating practices and environment, and the develop ment of a data base or analytic methods for predicting the reliability of piping systems. Design and opera tional stress or strain limits are assessed. This informa tion is available to the licensee in the design informa tion for the plant. The loading and resulting stresses or strains on the piping are needed as input to the calcula tions that predict the failure probability of a piping seg ment. The use of validated computer programs, with appropriate input, is strongly recommended in a quanti tative RI-ISI program because it may facilitate the

1.178-11

LEAK/BREAK

CONSEQUENCES

Leak Effects from Jet Impingement Disabling Leak or Full Break Loss of System Function Disabling Leak (plant trip) or Initiating Event Full Break Disabling Leak or Full Break Effects from Flooding Full Break Effects from Pipe Whip Figure 5 Mapping of Probabilities and Consequences for RI-ISI Analysis regulatory evaluation of a submittal. The analytic method should be validated with applicable plant and industry piping performance data.

2.2 Probabilistic Risk Assessment In accordance with the Commission's policy on PRA, the risk-informed application process is intended not only to support relaxation (number of inspections, inspection intervals and methods), but also to identify areas where increased resources should be allocated to enhance safety. Therefore, an acceptable RI-ISI pro cess should not focus exclusively on areas in which re duced inspection could be justified. This section ad dresses ISI-specific considerations in the PRA to support relaxation of inspections, enhancement of in spections, and validation of component operability.

The scope of a RI-ISI program, therefore, should in clude a review of Code-exempt piping for partial or full-scope programs and the review of non-Code piping for full-scope RI-ISI programs.

The general methodology for using PRA in regula tory applications is discussed in Regulatory Guide

1.174. The PRA can be used to categorize the piping segments into. HSS and LSS classification (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 (Ref. 4).

If a licensee elects to use PRA to enhance or modify its activities affecting the safety-related functions of SSCs subject to the provisions of Appendix B to

10 CFR Part 50, the pertinent requirements of Appen dix B will also apply to the PRA. In this context, there fore, a licensee would be expected to control PRA ac tivity in a manner commensurate with its impact on the facility's design and licensing basis and in accordance with all applicable regulations and its QA program de scription. An independent peer review can be an impor- tant element in ensuring this quality. The licensee's submittal should discuss measures used to ensure ade quate 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 report should address any limitations of the analysis that are expected to im pact the conclusion regarding the acceptability of the proposed change. The licensee's resolution of the find ings of the peer review, certification, or cross compari son, when performed, should also be submitted. This response could indicate whether the PRA was modified or could justify why no change to the PRA was neces sary to support decisionmaking for the change under consideration.

2.2.1 Modeling Piping Failures in a PRA

Input from the traditional engineering analysis ad dressed in Regulatory Position 2.1 includes identifica tion of piping segments from the point of view of the failure potential (degradation mechanisms) and conse quences (resulting failure modes and consequential pri mary and secondary effects). The traditional analysis identifies both the primary and secondary effects that can result from a piping failure, such as a leak, disabling leak, and a break. 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 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 character ized by train or system unavailability that may re quire shutdown because of the technical specifica tions or limiting conditions for operation.

1.178-12 V

3. Demand failures are failures accompanying a de mand for a train or system and are usually caused by the transient-induced loads on the segment dur ing system startup.

The impact of the pipe segment failure on risk should be evaluated with the PRA. Evaluation may in volve a quantitative estimate derived from the PRA, a systematic technique to categorize the consequence of the pipe failure on risk, or some combination of quanti fication 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 ini tiating event, for example), the evaluation process should be expanded to assess these events.

PRAs normally do not include events that repre sent failure of individual piping segments nor the struc tural elements within the segments. A quantitative esti mate of the impact of segment failures can be done by modifying the PRA logic to systematically and ex plicitly include the impact of the individual pipe seg ment 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 fail ure (referred to as the surrogate approach). In either case, to assess the impact of a particular segment fail ure, the analyst sets the appropriate events to a failed state in the PRA (by assigning them a frequency or probability of 1.0) and requantifies the PRA or the ap propriate parts of the PRAas needed. The requantifica tion should explicitly address truncation errors, since cut set or truncated sequences may not fully capture the impact of multiple failure events. This yields condi tional CDF (CCDF) and conditional LERF (CLERF)

estimates when the segment failure would trip the plant, and conditional core damage probabilities (CCDP) and conditional large early release probabili ties (CLERP) when the segment failure would not trip the plant.

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 fail ure, the process should provide for determining which range each segment's failure would lie within. In gen eral, the consequences would range from high, forthose 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 se lected. The use of ranges instead of individual results estimates may require fewer calculations, but the cate gorization process and decision criteria should be justi fied, well defined, and repeatable.

2.2.1.1 Dependencies and Common Cause Fail ures. The effects of dependencies and common cause failures (CCFs) for ISI components need to be consid ered 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 re quired.

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 pro gram (including both operator and equipment qualifi cation), which factors into the probability of detection, should also be addressed.

2.2.2 Use of PRA for Categorizing Piping Segments Once the impact of each segment's failure on plant risk metrics has been determined, the safety signifi cance of the segments is developed. The method of categorizing a piping segment can vary. For example, if the pipe failure event frequency or probability are esti mated by structural mechanics methods as discussed in Regulatory Position 2.1.5 and the events are incorpo rated into the PRA logic model, importance measure calculations and the determination of safety signifi cance, as discussed in Regulatory Guide 1.174 and SRP

Chapter 19 (Refs. 4 and 8), may be performed. Alterna tively, if a CCDF, CLERF, CCDP, or CLERP (depend ing on the impact the segment failure has on the plant)

are estimated for each segment from the PRA, a CDF

and LERF caused only by pipe failures may be devel oped 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 seg ment. The calculations used in such a process should yield well defined estimates of CDF, LERF, and impor tance measures. The licensee should provide a discus sion of and justification for the threshold criteria used.

As discussed in Regulatory Position 2.2.1, the con sequence of segment failures may be represented by categories of consequences instead of quantitative

1.178-13

"estimates for each segment. In this case, the potential for pipe fail'are as discussed in Regulatory Position

2.1.5 would also be developed as categories ranging from high to low depending on the degradation mecha nisms present and the corresponding likelihood that the segment will fail. These consequence and failure likeli hood 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 combina tion.

The safety-significance category of the pipe seg ment will help determine the level of inspection effort devoted to the segment. In general, higher safety significant segments will receive more inspections and more demanding inspections than less significant seg ments. In any integrated categorization process, the principles in Regulatory Guide 1.174 need to be ad dressed. Irrespective of the method used in the analysis, the licensee needs to justify the final categorization pro cess 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 neutrality. 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.

2.3 Integrated Decisionmaking Regulatory Positions 2.1 and 2.2 address the ele ments 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 (Ref. 4), as highlighted in Figure 1, are systemat ically addressed. Technical and operations personnel at the plant review the information and render a finding of HSS or LSS categorization for each piping segment un der review. Detailed guidelines for the categorization of piping segments should be developed and discussed with the group responsible for the determination (typi cally performed by the plant's expert panel).

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

3. ELEMENT 3: IMPLEMENTATION,

PERFORMANCE MONITORING, AND

CORRECTIVE ACTION STRATEGIES

Integrating the information obtained from Ele ments 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, perfor mance 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 program should be self-correct ing as experience dictates. The program should contain performance measures used to confirm the safety in sights gained from the risk analyses.

Upon approval of the RI-ISI program, the licensee should have in place a program for inspecting all HSS

and LSS piping identified in its program. (Note that ref erence to HSS piping is broadened when implementing a more detailed graded categorization process, such as low, medium, and high safety significant. For discus sion purposes, a tWo-category process (e.g., HSS and LSS) will be assumed. Requirements for medium and LSS piping will be addressed on a case-by-case basis.)

The number of required inspections should be a product of the systematic application of the risk-informed pro cess.

3.1 Program Implementation A licensee should have in place a schedule for in specting all segments categorized in its RI-ISI program as LSS and HSS. This schedule should include inspec tion strategies and inspection frequencies, inspection methods, the sampling program (the number of ele ments/areas to be inspected, the acceptance criteria, etc.) for the HSS piping that is within the scope of the ISI program, including piping segments identified as HSS 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 in terval 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 inter vals are contained in Section XI of the B&PVC. How ever, should active degradation mechanisms surface, the inspection interval would be modified as appropri ate. Updates to the RI-ISI program should be per formed at least periodically to coincide with the

1.178-14

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 deci sions made for the RI-ISI program, if plant design and operations change such that they impact the RI-ISI pro gram, if inspection results identify unexpected flaws, or if replacement activities 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 re quired 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 in spection, component and valve lineups, equipment op erating 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 inspec tions should be evaluated as part of the RI-ISI update.

Piping segments categorized as HSS that are not in the licensee's current ISI program should (wherever ap j

propriate 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 de veloped 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 imple mentation of the RI-ISI program at the plant.

For piping segments categorized as HSS 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 pip ing 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 per formed more frequently if dictated by any plant proce- dures to update the PRA (which may be more restrictive than a Section XI period type update) or as new de gradation mechanisms are identified.

31.2 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 defini tion and element selection. Some examples of these in puts would include:

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

"* Material and configuration changes

"* Welding techniques and procedures

"* Construction and preservice examination results

"* Stress data (operating modes, pressure, and tem perature 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 abil ity of plant operations personnel to perform actions as sociated with accident mitigation, should be included for review in any RI-ISI program update. Additionally, changes in those procedures that affect component in spection 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 Equ pment Performance Changes Equipment performance changes should be re viewed with system engineers and maintenance per sonnel to ensure that changes in performance parame ters such as valve leakage, increased pump testing, or identification of vibration problems is included in the periodic evaluation of the RI-ISI program update. Spe cific 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 examina tions, pressure tests, and cotresponding VT-2 visual examinations for leakage have been completed, and if

1.178-15

unacceptable flaws, evidence of service related degra dation, 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 pro vides failure information and trending information that may have a profound effect on the element locations currently being examined under a RI-ISI program. In dustry failure data is just as important to the overall pro gram as the owner's information. During the periodic update, industry data bases (including available inter national data bases) should be reviewed for applicabil ity to the owner's plant.

3.3 Corrective Action Programs Each licensee of a nuclear power plant is responsi ble for having a corrective action program, consistent with Regulatory Guide 1.174 (Ref. 4). Measures are to be established to ensure that conditions adverse to qual ity, such as failures, malfunctions, deficiencies, devi ations, defective material and equipment, and noncon formances, 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 condi tion 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 HSS, this correc tive action program should be consistent with applica ble Section XI provisions. For non-Code and Code exempt piping categorized as HSS, 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 imple mentation, 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 rea sonable assurance that a nonconforming compo nent 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 condi tion is determined and that corrective ac tions are taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condi tion, 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 oper ability 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 com ponents as necessary.

"* 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 com ponent's inspection strategy in detecting the failure or nonconforming condition.

The inspection interval would be reduced or the inspection methods adjusted, as ap propriate, when the component (or group of components) experiences repeated fail ures or nonconforming conditions.

4. The corrective action evaluation should be pro vided 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 find ings.

1.178-16

7. Piping is subject to examination. The examination requirements include all piping evaluated by the risk-informed process and categorized as high safety significant.

8. The inspection pr6gram is to be completed during each ten-year inspection interval with the follow ing exceptions.

8.1 If, during the interval, a reevaluation 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 reevaluation using the RI-ISI process is conducted and items must be added to the examination program, those items will be added.

9. Locations selected for successive and additional inspections should be subjected to successive and additional examinations consistent with Section XI

requirements at appropriate intervals.

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 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 HSS piping and on non-Code LSS pip ing with high failure potential.

Examination qualification and methods and per sonnel 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 re pair or replacement activities are to be performed in accordance with the B&PVC Section XI require ments.

12. Records and reports should be prepared and main tained 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 infor mation 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 Table 1 provides an overall summary of the infor mation needed to support a risk-informed ISI submit- tal. References to NRC-approved generic topical re ports that address the methodology and issues requested in a submittal are acceptable. Since topical reports could cover more issues than applied by a li censee or the licensee may elect to deviate from the full body of issues addressed in the topical report, such dis tinctions should be clearly stated. If a licensee refer ences a topical report that has not been approved by the NRC, the time required to review the submittal may be delayed.

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

"

A request to implement a RI-ISI program as an au thorized 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 commit ments made to the NRC.

"

Detailed discussions on each of the following five key principles of risk-informed regulations (see Section 2 of Regulatory Guide 1.174 (Ref. 4) for more details).

1. The proposed change meets the current regula tions unless it is explicitly related to an alterna tive requested under 10 CFR 50.55a(a)(3)(i), a requested exemption, or a rule change.

2. The proposed change is consistent with the de fense-in-depth philosophy (see detailed dis cussions 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 core damage frequency and/or risk, the in creases 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 pro posed RI-ISI program. This identification should include all commitments (for example, the IGSCC

inspections and other commitments arising from generic letters affecting piping integrity) that the li censee intends to change or terminate as part of the RI-ISI program.

1.178-17

Table 1 Documentation Summary Table PRA Quality Address the adequacy of the PRA model used in the calculations.

Address the acceptance guidelines in Regulatory Position 2 of this document and in Regulatory Guide 1.174 (Ref. 4).

Failure Probability Calcula- Address the methods used to calculate or categorize the failure probability or tions frequency of a piping element. Any use of expert elicitation should be fully documented.

Changes in CDF and LERF

Address the change in CDF and LERF resulting from changes to the ISI pro gram ISI Systems Identify all the systems inspected based on the current ISI programs and compare the systems for the RI-ISI programs.

Segmentation Identify methods used to segment piping systems, if applicable.

Categorization Identify methods used to categorize piping segments and elements as HSS,

LSS, high failure potential, and low failure potential.

Identify all the HSS-HFP and HSS-LFP elements (format may differ based on decision matrix employed).

Sampling Method Identify the method used to calculate the number of elements to be inspected.

Document the method used to establish elements within a lot. Address how this method provides an acceptable level of quality and safety per 10 CFR

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

Locations of Inspections Provide a system/piping diagram or table that compares the existing ISI loca tions of inspection with the RI-ISI location of inspection.

"Address the reasons for the changes.

Failure Probabilities Identify the methods used to arrive at the failure probabilities for piping seg ments.

Performance Monitoring Discuss the performance goals and corrective action programs.

Periodic Reviews Identify the frequency of performance monitoring and activities in support of the RI-ISI program. Address consistency with other RI programs (e.g.,

Maintenance Rule, IST, Tech Specs).

QA Program Describe the QA program used to ensure proper implementation of RI-ISI

process and categorization and consistency with other RI programs.

Expert Elicitation Identify any use of the expert elicitation process to estimate a failure proba bility for piping. Address the reasons why an expert elicitation was required, provide all supporting information used by the experts, document the conclu sions, and address how the results will be incorporated in an industry data base or computer code, or why it is not necessary to make the findings avail able to the industry.

Each weld to be inspected Identify: 1. The inspection method to be used

2. The applicable degradation mechanism to be inspected, and

3. The frequency of inspection Address each of the key prin- Verify compliance with applicable regulations, defense-in-depth, safety mar ciples and the integrated deci- gins, etc.

sionmaking guidelines (e.g.,

Regulatory Position 2.3)

Implementation and monitor- Address the acceptance guidelines outlined in Regulatory'Position 3 of this ing program regulatory guide.

1.178-18

SA summary of events involving piping failures that have occurred at the plant or similar plants. Include in the summary any lessons learned from those events and indicate actions taken to prevent or minimize the potential for recurrence of the events.

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 imple menting a change in the ISI program based on pro babilistic and traditional insights. These should include a description of the integrated decision making process and criteria used for categorizing the safety significance of piping segments, a de scription 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 mak ing 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 r~iewers 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 weld ing material and base metal, the failure poten tial, CDF, CCDF/CCDP, LERF, CLERF, im portance measure results (RAW, F-V, etc.) and justification of the associated threshold val ues, degradation mechanism, test and inspec tion intervals used in or in support of the PRA,

etc. Results from other methods used to de velop the consequences and categorization of each segment (or weld) should be documented in a similar level of detail. (NOTE: Table 2 provides an example of a summary of possible methods for obtaining failure probabilities based on specified degradation mechanisms.

The staff recommends that licensees provide such a table with supporting discussions.)

For the selected limiting locations, provide ex amples of the failure mode, failure potential, failure mechanism, weld type, weld location, and properties of the welding material and base metal. Provide a detailed description and justification for the number of elements to be inspected.

-

The degradation mechanisms for each seg ment (if segments contain welds exposed to different degradation mechanism, for each weld) used to develop the failure potential of each segment.

-

Equipment assumed to fail as a direct or indi rect 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 pro posed RI-ISI programs is evaluated or bounded, and how this impact compares with the risk guidelines in Section 2.2.2.2 of Regu latory Guide 1.174.

The means by which failure probabilities or fre quencies or potential were determined. The data should be provided in the submittal for analyses that rely on operational data for determining failure frequencies or potential. Reliance on fracture me chanics structural reliability and risk analysis codes should be documented and validated. Re liance on the expert elicitation process should be fully documented. (NOTE: Expert elicitation is only used if data are not sufficient to estimate the failure probability and frequency of a piping seg ment. Data assessment is not an expert elicitation process and can normally be performed by plant personnel.)

A description of the PRA used for the categoriza tion process and for the determination of risk im pact, 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 sup porting the ISI submittal. The key assumptions used in the PRA that impact the application (i.e.,

licensee voluntary actions), elements of the moni toring program, and commitments made to support the application should be addressed.

"

If the submittal includes modified inspection inter vals, the methodology and results of the analysis should be submitted.

"* A description of the implementation, performance monitoring, and corrective action strategies and programs in sufficient detail for the staff to under stand the new ISI program and its implications.

1.178-19

"

Applicable documentation discussed under the Cumulative Risk documentation for submittal in Section 1.3 of Regulatory Guide 1.174 (Ref. 4).

"

Reference to NRC-approved topical reports on im plementing a RI-ISI and supporting documents.

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

"

Detailed justification for the proposed regulatory action (e.g., how the proposed program meets the requirements set in 10 CFR 50.55a(a)(3)(i)).

4.2 Documentation That Should Be Available Onsite for Inspection The licensee should maintain at its facility the tech nical 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.

Plant and applicable industry data used in support of the RI-ISI program. All analyses and assump tions used in support of the RI-ISI program and communications with outside organizations sup porting 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 de tail to allow an independent reviewer to ascertain whether the PRA reflects the current plant configu ration and operational practices commensurate with the role the PRA results play in the integrated decisionmaking process. In addition to documen tation on the PRA itself, analyses performed in support of the IST submittal should be documented in a manner consistent with the baseline documen tation. 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.

-

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 seg ment categorization, and the sensitivity stud ies performed to ensure the robustness of the categorization.

Detailed results of the inspection program corre sponding 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 ma terial and base metal.

For each piping segment, information regarding the process and assumptions used to develop fail ure mode and failure potential (frequency/proba bility), in addition to the identification of the fail ure mechanism.

K,

1.178-20

K

Table 2 Example of a Summary of Methods Used To Estimate Piping Failure Probabilities for Risk Categorization Failure Mechanism Methods for Estimating Probability Name of Mechanism Contributing Factors Failure Mode Stainless Steel Carbon Steels Other Materials Thermal Striping Crack Code Name Code Name High Cycle Flow Induced Vibration Initiation Failure Fatigue Mechanical Vibration Crack Code Name Code Name Database Growth Thermal Stratification Crack Code Name Code Name Low Cycle Heat-up and Cool-down Initiation Failure Fatigue Thermal Cycling Crack Code Name Code Name Database Growth Coolant Chemistry Crack Code Not Corrosion Crevice Corrosion Initiation Name Applicable Failure Cracking Susceptible Material Database High Stresses Crack Code Not (Residual, Springing)

Growth Name Applicable Flow Accelerated. Corrosion Wall Name of Name of Failure Wastage Microbiologically Ind. Corr.

Thinning Code Code Database

_

_Pitting and/or Wear

"Other Creep Damage Miscellaneous Failure Failure Failure Mechanisms Thermal Aging Modes Database Database Database Irra

d. Embrittlement I

I

1-1 Co

00

k4

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, "Standard Review Plan for the Review of Risk-Informed Inservice Inspection of Piping,"

NUREG-0800, Section 3.9.8, September 1998.2

4. USNRC, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis," Regulatory Guide 1.174, July 1998.2

5. USNRC, "An Approach for Plant-Specific, Risk Informed Decisionmaking: Inservice Testing,"

Regulatory Guide 1.175, August 1998.2

6. USNRC, "An Approach for Plant-Specific, Risk Informed Decisionmaking: Graded Quality Assur ance," Regulatory Guide 1.176, August 1998.2

7. USNRC, "An Approach for Plant-Specific, Risk Informed Decisionmaking: Technical Specifica tions," Regulatory Guide 1.177, August 1998.2

8. USNRC, "Standard Review Plan for Risk Informed Decision Making," Standard Review Plan, NUREG-0800, Chapter 19, July 1998.3 lCopies 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 mailing address is Mail Stop LL-6, Washington, DC 20555;

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

2Single copies of regulatoryguides, both active and draft, and standard review plans may be obtained free of charge by writing the Reproduc tion and Distribution Services Section, OCIO, USNRC, Washington, DC 20555-0001, or by fax to (301) 415-2289, or by e-mail to GRWI@NRC.GOV. Active guides may also be purchased from the National Technical Information Service on a standing order basis.

Details on this service may be obtained by writing NTIS, 5285 Port Royal Road, Springfield, VA22161. Copies of active and draft guides 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 mailing address is Mail Stop LL-6, Washington, DC 20555; tele phone (202) 634-3273; fax (202) 634-3343.

3Copies are available at current rates from the U.S. Government Printing Office, RO. Box37082, Washington, DC20402-9328 (tele phone (202) 512 - 2249); or from the National Technical Information Service by writing NTIS at 5285 Port Royal Road, Springfield, VA

22161. Copies 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 mailing address is Mail Stop LL-6, Washington, DC

20555; telephone (202) 634-3273; fax (202) 634-3343.

9. USNRC, "Standard Review Plan for Risk Informed Decision Making: Inservice Testing,"

Standard Review Plan, NUREG-0800, Chapter

3.9.7, August 1998.3

10. USNRC, "Standard Review Plan for Risk Informed Decision Making: Technical Specifica tions," Standard Review Plan, NUREG-0800,

Chapter 16.1, August 1998.3

11. American Society of Mechanical Engineers, "Case N-560, Alternative Examination Requirements for Class 1, Category B-J Piping WeldsSection XI,

Division 1," August 9, 1996.4

12. American Society of Mechanical Engineers, "Case N-577, Risk-Informed Requirements for Class 1,

2, and 3 Piping, Method A,Section XI, Divi sion 1," September 2, 1997.4

13. American Society of Mechanical Engineers, "Case N-578, Risk-Informed Requirements for Class 1,

2, and 3 Piping, Method B,Section XI, Divi sion 1," September 2, 1997.4

14. Electric Power Research Institute, "PSA Applica tions Guide," EPRI TR-105396, August 1995.5

15. Electric Power Research Institute, "Risk-Informed Inservice Inspection Evaluation Procedure," EPRI

TR-106706, June 1996.5

16. Westinghouse Energy Systems, "Westinghouse Owners Group Application of Risk Informed Methods to Piping Inservice Inspection Topical Report," WCAP-14572, Revision 1, October

1997.1

17. Westinghouse Energy Systems, "Westinghouse Structural Reliability and Risk Assessment (SRRA) Model for Piping Risk-Informed Inser vice Inspection," WCAP-14572, Revision 1, Sup plement 1, October 1997.1

18. T.V. Vo et al., "A Pilot Application of Risk-In formed Methods To Establish Inservice Inspection Priorities for Nuclear Components at Surry Unit 1 Nuclear Power Station," USNRC, NUREG/

CR-6181, Revision 1, February 1997.3

4Copies may be obtained from the American Society of Mechanical Engineers, 345 East 47th Street, New York, NY 10017.

5Copies may be obtained from the EPRI Distribution Center, 207 Coggins Drive, P.O. Box 23205, Pleasant Hill, CA 94523.

1.178-22 K

19. American Society of Mechanical Engineers,

"Rules for Inservice Inspection of Nuclear Power Plant Components," ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition, New

,'

York.4

20. USNRC, "Design and Fabrication Code Case Ac ceptability, ASME Section III, Division I," Regu latory Guide 1.84, Revision 30, October 1994.2

21. USNRC, "Materials Code Case Acceptability, ASME Section III, Division 1," Regulatory Guide

1.85, Revision 30, October 1994.2

22. USNRC, "Inservice Inspection Code Case Accept ability, ASME Section XI, Division 1," Re*gulatory Guide 1.147, Revision 11, October 1994.2

23. M.A. Meyer and J.A. Booker, "Eliciting and Ana lyzing Expert Judgement," NUREG/CR-5424 (Prepared for the NRC by Los Alamos National Laboratory), USNRC, January 1990.3

24. J.P. Kotra et al., "Branch Technical Position on the Use of Expert Elicitation in the High-Level Radio active Waste Program," NUREG-1563, USNRC,

November 1996.3 REGUILATORY ANALYSIS

A draft regulatory analysis was published with the draft of this guide when it was published for public comment (Task DG-1063, October 1997). No changes were necessary, so a separate regulatory analysis for Regulatory Guide 1.178 has not been prepared. A copy of the draft regulatory analysis is available for inspec tion or copying for a fee in the NRC's Public Document Room at 2120 L Street NW., Washington, DC, under Task DG-1063.

1.178-23

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