ML20134K718
| ML20134K718 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1990 |
| From: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| TASK-DG-1009, TASK-RE REGGD-01.XXX, REGGD-1.XXX, NUDOCS 9611200003 | |
| Download: ML20134K718 (83) | |
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,t U.S. NUCLEAR REGULATORY COMMISSION i
December 1990 '
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.... o DRAFT REGULATORY GUIDE l
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DRAFT REGULATORY GUIDE DG-1009 - i fp% !
R$ f STANDARD FORMAT AND CONTENT OFf TECHNICAL INFORMATION ;
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This regulatory guide is being issued in draft form to involve the public in the early stages of the develop-ment of a regulatory position in this area. It has not received complete staff review and does not represent an official NRC staff position.
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Public comments are being solicited on the draft guide (including any implementation schedule) and its associ-ated regulatory analysis or value/ impact statement. Comments should be accompanied by appropriate supporting i data. Written comments may be submitted to the Regulatory Publications Branch, DFIPS, Office of Administra- l tion U.S. Nuclear Regulatory Comnission, Washington, DC 20555. Copies of comments received may be examined -
at the NRC Public Document Room, 2120 L Street NW., Washington, DC. Comments will be most helpful if received i
by March 8 1991.
Requests for single copies of draft guides (which may be reproduced) or for placement on an automatic distri-bution 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: Director Division of Information
. Support Services.
9611200003 901231 PDR REGOD ;
C1.XXX R PDR
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CONTENTS ,
A. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Purpose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Scope . . . . . . . . . .................... 2 '
l l Applicability . . . . . . . . . . . . . . . . . . . . . . . . . . 2 I
l B. DISCUSSION ............................ 3 l Use of Standard Format ......'............... 3 l
C. REGULATORY POSITION . . . . . . . . . . . . . . . . . . . . . . . 4
- 1. FORMAT FOR TECHNICAL INFORMATION . . . . . . . . . . . . . . 4 1.1 Fo rmal Appl i cati o n . . . . . . . . . . . . . . . . . . . . . 5 l
1.2 FSAR Supplement. . . . . . . . . . . . . . . . . . . . . . . 7 1.2.1 Review Objectives . . . . . . . . . . . . . . . . . . 8 l t l 1.2.2 Supporting Documentation. . . . . . . . . . . . . . . 9 ,
I 2. TECHNICAL INFORMATION CONTENT ............... 14 2.1 Types of SSCs for Which Aging Should Be Considered for License Renewal. . . . . . . . . . . . . . . . . . . . . . . 14 2.2 Integrated Plant Assessment - Programs for Addressin l
Age-Related Degradation. . . . . . . . . . . . . . .g .... 15 l
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2.2.1 Established Effective Programs. . . . . . . . . . . . 17 2.2.2 Actions To Be Taken To Manage Age-Related Degradation . . . . . . . . . . . . . . . . . . . . . 17 l 2.2.3 Structures and Components Not Subject to Significant Age-Related Degradation . . . . . . . . . . . . . . . 17 2.2.4 Summary of Acceptable Programs and Practices for Understanding Aging . . . . . . . . . . . . . . . . . 18 .
2.2.5 Summary of Acceptable Programs and Replacement /
Refurbishment Practices for Managing Aging. . . . . . 19 D. IMPLEMENTATION ......................... 21 1
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CONTENTS (Continued)
APPENDIX A -
SUMMARY
OF AGE-RELATED DEGRADATION PROCESSES AND THEIR MANAGEMENT IN OPERATING NUCLEAR POWER PLANTS ...... A-1 APPENDIX B - REPRESENTATIVE SYSTEMS, STRUCTURES, AND COMPONENTS POTENTIALLY IMFCRTf.NT TO LICENSE RENEWAL ........ B-1 REGULATORY ANALYSIS. . . . . . . . . . . . . . . . . . . . . . . . . . RA-1 BACKFIT ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . . . RA-1 FIGURES
- 1. Process for Selecting Systems, Structures, and Components Important to License Renewal and for Understanding and Managing Age-Related Degradation . . . . . . . . . . . . . . . . . . . . . 22 1A. Integrated Plant Assessment -- Identification of Important-To-License-Renewal SSCs and SCs Requiring Evaluation of Age-Related Degradation . . . . . . . . . . . . . . . . . . . . . 24 ,
- 18. Integrated Plant Assessment -- Evaluation of Age-Related Degradation . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 TABLES I. Information To Be Included in the FSAR Supplement - General and SSC Specific. . . . . . . . . . . . . . . . . . . . . . . . . 27 II. Technical Information Needed for License Renewal. . . . . . . . . 32 III. Generic Functional Nuclear Power Plant SSCs Important to License Renewal . . . . . . . . . . . .. . . . . . . . .. . . . . 36 F
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1 A. INTRODUCTION 2 The Nuclear Regulatory Commission (NRC) is proposing to. supplement its 3' regulations in Title 10 of the Code of Federal Regulations by adding Part 54, !
4 " Requirements for Renewal of Operating Licenses for Nuclear Power Plants."* !
5 l Section 54.21 of the proposed rule specifies the technical information to be '
6 included as part of an application for license renewal. This information is to !
7 be included in a supplement to the current, updated Final Safety Analysis Report 8 (FSAR). The supplement is to be included in the application submitted by a
-9 nuclear power plant licensee for a renewed operating license. The FSAR supple- l 10 ment will include an evaluation of the aging mechanisms that result in degrada-11 f
tion of the plant's systems, structures, and components (SSCs) important to j 12 license renewal, as defined in 10 CFR 54.3(a). The FSAR supplement will provide i 13 information to show that the effects of such degradation will be effectively 14
.l managed so that the current licensing basis for the plant, as defined in 10 CFR !
15 54.3(a), will be maintained throughout the renewal term. Each FSAR supplement !
16 is to contain the information required by 10 CFR 54.21.
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i 17 Purpose I
18 The purpose of this regulatory guide is to establish a uniform format and 19 content acceptable to the NRC staff for structuring and presenting the technical !
20 information to be compiled by an applicant for a renewed nuclear power plant I 21 operating license. The guide also establishes a uniform format for structuring l 22 and presenting the technical information to be submitted by the applicant as 23 part of an application for a renewed license. This regulatory guide identifies i 24 the content of and provides technical criteria for the compiled technical infor- f 25 mation. Use of this format will help to ensure the completeness of the informa-l 26 tion compiled or provided, will assist the NRC staff and others in locating the I 27 information, and will aid in shortening'the time needed for the process of 28 reviewing the license renewal application.
29 30 *This draft regulatory guide is based on the proposed license renewal rule, !
31 " Requirements for Renewal of Operating Licenses for Nuclear Power Plants," 10 32 CFR Part 54 (Federal Register, Vol. 55, No. 137, July 17, 1990). Future mod-33 ifications to the proposed rule will be reflected in commensurate changes in 1
34 the draft regulatory guide. [
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l l 1 Scope l 2 This regulatory guide provides a standard forsaat and content for the l 3 technical information to be compiled or submitted in support of an application 4 for a renewed operating license, as will be required by 10 CFR Part 54. Detailed 5 technical-information needs and a description of a standard format that is l 6 acceptable to the NRC staff are presented in the Regulatory Position of this l
7 regulatory guide.
8 This regulatory guide also provides for meeting the technical information
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9 requirements of 10 CFR Part 54, including (1) the content of the technical infor-l 10 mation to be included in license renewal applications,~ (2) criteria for selec--
11 tion of SSCs important to license renewal and their constituent structures and 12 components for which age-releted degradation should be assessed and accounted 13 for, (3) evaluation of design, operational, and environmental factors that con-14 tribute to age-related degradation, (4) identification of the aging mechanisms 15 and specific sites involved in degradation processes, and (5) attributes of 16 established effective programs and of acceptable actions taken or to be taken +
17 to understand and manage age-related degradation. Detailed information on under-
~ 18 - standing and managing aging that will be useful to a license renewal applicant 19 in implementing these methods is contained in Appendix A to this regulatory 20 guide.
21 The guidance provided herein is expected to ensure that actions have been 22 identified and have been taken or will be taken with respect to age-related l
23 degradation of those SSCs important to license renewal, so that there is reason-24 able assurance that the activities authorized by the renewed license can be 25 conducted in accordance with the current licensing basis.
26 Applicability l 27 This regulatory guide applies to applications for renewal of operating L
28 licenses for commercial nuclear power plants and to the constituent SSCs of. .,
29 these facilities that are designated important to license renewal as defined in !)
30 10 CFR 54.3(a). l l 31 Any information collection activities mentioned in this draft regulatory
[ 32 guide are contained as requirements in those sections of 10 CFR Part 54 that 8
33 provide the regulatory basis for this guide. The proposed additions to 10 CFR i
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1 Part 54 have been submitted to the Office of Management and Budget for. clearance, 2 as appropriate, under the Paperwork Reduction Act. Such clearance, if obtained, 3 would also apply to any information collection activities identified in this 4 guide.
5 B. DISCUSSION 6 The. technical information developed and submitted or retained by an ,
7 applicant for license renewal should allow the NRC staff to make the determina-8 tion that the requirements of 10 CFR Part 54 have been met. The format in which 9 this information is presented should satisfy the requirements of 10 CFR Part j 10 54, should provide for optimum utilization of the applicant's resources, and !
11 should facilitate the NRC staff review of a license renewal application. {
' 12 Technical information submitted by an applicant should focus on aging mechanisms '
13 and the resultant age-related degradation that can lead, in the context of the i 14 renewed license term, to unacceptable deterioration of SSCs important to license l
15 renewal. The technical information content of a license renewal application v I 16 should be sufficient to provide an NRC reviewer of the application with a sound {
17 understanding of the aging processes that contribute to degradation of SSCs l 18 important to license renewal and how these processes are or will be managed.
19 The information needed to impart this understanding should address the integrated l,
-20 effects of materials, design, environment, stressors, and plant operating history 21 on SSC degradation attributable to specific aging mechanisms. These effects ;
22 are discussed in Appendix A to this regulatory guide.
23 The technical information will be reviewed by the NRC staff to assess the 24 effectiveness of an applicant's established or proposed programs for understand-25 ing and managing age-related degradation of SSCs important to license renewal 26 during a renewed license term and to evaluate acceptability of the application ,
27 for a renewed license. I 28 Use of Standard Format 29 Conformance with the standard format described in the Regulatory Position 30 is encouraged but not required. License renewal applications with different 31 formats will be acceptable to the staff if they provide an adequate basis for :
32 the findings requisite to the issuance of a renewed license. However, because 33 it may be more difficult to locate needed information, the staff review time 3
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1 1 for such applications may be longer, and there is a greater likelihood that 2 the staff may regard the application as incomplete.
3 Upon receipt of an application for license renewal, the NRC staff will 4 perform a preliminary review to determine if the application provides a reason-5 ably complete presentation of the information needed for issuance of a license 6 in accordance with 10 CFR 54.29. The purpose of this review will be to deter-7~ mine if the submittal is sufficient according to the provisions of 10 CFR
-8 2.109(b). The standard format will be used by the staff as a guideline to iden-9 tify the type of information needed unless there is good reason for not doing 10 so. If the application does not provide a reasonably complete presentation of 11 the necessary information, further review of the application will not be ini-12 tiated and the provisions of 10 CFR 2.109(b) will not apply until a reasonably 13 complete presentation is provided. The information provided in the application 14 should be current with respect to the status of the plant and the state of tech-15 nology concerning age related degradation in operating nuclear power plants.
16 Furthermore, this.information should take into account, as appropriate, recent 17 changes in regulations and in industry codes and standards; results of recent w 18 developments in nuclear reactor safety; and experience in plant-specific design, 19 construction, and operation.
20 C. REGULATORY POSITION l 1
21 The methods described in this section are acceptable to the NRC staff for 22 satisfying the requirements proposed in 10 CFR 54 and 10 CFR 2.109 pertaining ;
23 to the technical information to be compiled or submitted in support of an l 24 application for a nuclear power plant operating license renewal. '!
25 1. FORMAT FOR TECHNICAL INFORMATION i
26 An application for license renewal must meet the applicable provisions of 27' 10 CFR Part 54. Provisions contained in 10 CFR 54.23 deal with environmental 28 -information to be submitted. Environmental issues are addressed in Regulatory 29 Guide 4.2, " Preparation of Environmental Reports for Nuclear Power Stations."
30 The license renewal application is to contain~two separate parts: a formal 31 application and a supplement to the current FSAR. Regulatory Position 1.1 32 describes the basic'information that should be included in the formal applica-33 tion. Regulatory Position 1.2 describes the information that should be included 4
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1 in the FSAR supplement, which will be an attachment to the formal application. '
2 The FSAR supplement will consist of a new chapter added to the current FSAR for 3 the sole purpose of license renewal. This new FSAR chapter will contain the 4 detailed technical information to be included as part of the application. As 5 described in 10 CFR 54.17(e), the application may incorporate, by reference, ,
6 information contained in previous submittals provided such references are clear '
7 and specific.
8 1.1 Formal Application 9 The formal application is to contain the following elements:
10 1. A table of contents.
11 2. An introduction providing general information concerning the application.
12 This should include:
13 a. The information specified in 10 CFR 50.33(a) through'(e), 50.33(h),
14 and 50.33(i),
15 b. The earliest effective date and length of the renewal term, 1 16 c. A statement summarizing how and the extent to which the application !
17 meets the regulatory requirements for license renewal (10 CFR 54).
18 Exceptions should be listed and justified.
19 d. A description of the scope and organization of the remaining sections ;
20 of the application, 21 e. The information specified in 10 CFR 54.17(g), and
- 22 f. An acknowledgment that the commitments and requirements contained 23 in the licensee's current Part 50 license not affected or superseded 24 by the license renewal application will remain in effect when the 25 Part 54 license is issued.
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I 1 3.
A characterization and summation af the licensee's findings.
2 This section should provide the justification, in summary form, to support 3
the conclusion that appropriate actions have been or will be taken to manage 4
the effects of age-related degradation of the facility SSCs important to 5 license renewal. Details supporting these findings are to be contained in 6 thet FSAR supplement.
7 4. An implementation plan that includes the following elements:
8 a. Summary of Commitments 9
List the commitments described in the license renewal application.
10 b. Description of Administrative Controls 11 l Describe the administrative control program used by the licensee to 12 establish and maintain the commitments listed above. Such a program 13 should ensure that changes to the commitments are evaluated for aging l l 14 considerations prior to revision and conform to the requirements for i 15 an established effective program contained in 10 CFR 54.3(a).
15 c. Tasks and Schedule 17 Detail the commitments pertaining to age-related degradation that 18 will be completed following renewal of the operating license. These 19 commitments may include system design changes, one-time replacements, 20 program enhancements, and new programs. )
21 A schedule should be established and provided for the specific actions ,
( 22 committed to in this section, and this schedule should be consistent 1 23 with the evaluations made in the license renewal application.
24 5. Submittals of aay Technical Specification and program changes and additions 25 identified as necessary to manage age-related degradation during the renewal 26 term.
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1 6. A list of changes in calculations or analytical approaches resulting from i l 2 licensing renewal activities, including justification for the changes.
3 1.2 FSAR Supplement 4 The detailed technical information for a license renewal application is to 5 be contained in a supplement to the current FSAR in accordance with 10 CFR 54.21.
6 This supplement will consist of a new FSAR chapter that conforms to the adminis-7 trative requirements for FSAR chapters. This new chapter should be cross-l 8 referenced, as necessary, to other chapters in the FSAR. ;
9 The supplemental FSAR chapter should contain sections that describe the 10 licensee's methodology and results for satisfying each element of 10 CFR 54.21.
11 The NRC staff will review the licensee's application on a systems basis accord-12 ing to review procedures set forth in a new Standard Review Plan chapter that 13 deals with the review of applications for renewed licenses, Draft NUREG-1299, !
14 " Standard Review Plan for License Renewal" (SRP-LR).* ;
l 15 The technical information compiled or submitted by a licensee in compliance "r 16 with 10 CFR 54.21 should conform in format and content to this regulatory guide. !
- 17 This expectation is based on the requirements in the proposed 10 CFR 54.21 to 18 ensure that a facility's licensing basis will be maintained throughout the term 19 of the renewed license and on the definition of current licensing basis as l
20 stated in 10 CFR 54.3(a). Maintenance of the current licensing basis, as defined 21 in 10 CFR 54.3(a), requires a licensee to comply with 10 CFR 50 among other i 22 things. This includes the requirement cited in 10 CFR 50.34(g) that nuclear 23 power plant operating licenses docketed after May 17, 1982, include an evaluation 24 of the facility against the SRP revision in effect six months prior to the
- 25 docket date of the application.
26 Table I outlines the information that should be included in the FSAR supple-27 ment. This information is structured to conform to the process that will be 28 followed by the NRC staff in reviewing applications for license renewal. As 29 indicated in Table I, the FSAR supplement should contain three categories of 30 technical information:
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- Single copies of Draft NUREG-1299 are available from the U.S. Nuclear 33 Regulatory Commission, Washington, DC 20555, Attention: Director, Division 34 of Information Support Services, while supplies last.
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l 1 Par 1 A: General Information and Discussion 2 information of an introductory or general nature such as purpose, scope, 3 definitions, organization, and relationship to 10 CFR 54; conformance to 4 regulatory guides; citations for referenced information; general technical 5 information required by 10 CFR 54 and described in Regulatory Position 6 1.2.2; and a description of any deviations from the acceptance criteria 7 contained in the SRP-LR.
l 8 Part B: Nuclear Power Plant Systems 9 Information specific to the principal systems and subsystems of the plant !
10 that contain structures or components important to license renewal.
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11 Part C: Generic Components l 12 Information related to structures and components important to license j 13 renewal for which age-related degradation may be generically addressed. ;
14 Not all the items included in the lists of SSCs in Table I are germane to l 15 all plants, and the lists may not include all SSCs important to license renewal ,
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16 for any particular plant. l l 17 1.2.1 Review Objectives 18 The FSAR supplement should allow the NRC staff to reach the following l
19 conclusions:
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20 1. Sufficient technical information has been submitted as part of the 21 application to begin the review.
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. 22 2. The licensee's screening methodologies and resulting list of SSCs important 23 to license renewal and the lists af their constituent structures and 24 components are acceptable.
25 3. The licensee's methodologies for identifying established effective programs
, 26 [ defined in 10 CFR 54.3(a)] for structures and components requiring 27 evaluation of age-related degradation are acceptable.
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l 1 4. -The licensee's established programs effectively manage age-related ,
! '2 degradation of structures and components important to license renewal. .
3 5. The licensee actions are adequate to manage age related degradation in 4 structures and components important to license renewal that are not 5 currently subject to established effective programs.
6 1.2.2 Supporting Documentation 7 The FSAR supplement should provide the facility-specific technical l 8 information needed by the NRC staff to make the judgments cited above. The 9 licensee should submit this information to meet the requirements stated in 10 10 CFR 54.21. Specifically, the FSAR supplement should contain:
l 11 1. A detailed description of the licensee's integrated plant assessment, 12 including the screening methodology for identifying SSCs important to 13 license renewal and the methodology for determining if an established 14 program is effective in managing age-related degradation.
15 2. Information specific to the total facility as required by 10 CFR 54.21, 16 specifically:
17 a. A list of SSCs important to license renewal as defined in 10 CFR 18 54.3(a).
19 b. A list of all structures and components that are constituent elements 20 of the SSCs listed in 2.a.
21 c. A list of all structures and components from 2.b that require 22 evaluation of age-related degradation.
23 d. Justification for conclusions that any structures and components
- 24 listed in 2.b but not listed in 2.c do not contribute to the perform-25 ance of a safety function of an SSC important to license renewal or i 26 that their failure would not prevent an SSC important to license 27 renewal from performing its intended safety function.
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A list of structures and components for which age related degradation 2-is not significant with respect to the current licensing basis [as 3 ,
defined in 10 CFR 54.3(a)] through the renewed license period and 4
documentation of the evaluations that support these findings.
5- f.
A list of all structures and components that are subject to an 6
established effective program as defined in 10 CFR 54.3 (a), the I 7
associated effective programs, and the basis for continuing these 8 programs through the renewed license period.
9 g. A description of and the basis for actions taken or to be taken to 10 manage age-related degradation, including, where appropriate, a 11 description of revisions to replacement / refurbishment programs to 12 demonstrate their adequacy.
13 h. A list of all plant-specific exemptions granted pursuant to 10 CFR '
14 50.12 and reliefs granted pursuant to 10 CFR 50.55a. This list meet
15 include an identification of those reliefs and exemptions granted on 16 the basis of an assumed service life or period of operation bound by 17 the original license term of the facility or otherwise related to SSCs '
18 subject to age related degradation. Also, for reliefs and exemptions 19 granted on the basis of an assumed service life cr period of operation t 20 bounded by the original license term of the facility or otherwise 21 related to SSCs subject to age-related degradation, justification for 22 their continuation should be provided in either this section or the 23 system-specific section below. .
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l 24 1. A list and description of any proposed modifications related to age-25 related degradation that have been or will be made to the facility or .
26 its administrative control procedures resulting from the integrated 27 plant assessment [10 CFR 54.21(a)] or exemptions and reliefs described 28 in 2.h [10 CFR 54.21(b)).
29 3. Information specific to the systems listed in Part B of Table I and the 30 generic structures and components listed in Part C of Table I. Information 31 related to components that can be grouped in terms of component type and 32 expected age-related degradation may be cited once in generic form, and 10 k
l l 1 that citation may be referenced for subsequent components that fit the 2 appropriate grouping. For each system (Part B of Table I) or generic !
!- 3 structure or component (Part C of Table I) applicable to the facility, the 4 following information should be presented:
5 a. An SSC-specific list of constituent structures and components that [
6 are important to license renewal. A reference to the lists provided 7 pursuant to 2.a and 2.b will suffice, provided these lists clearly 8 identify all structures and components associated with that particular 9 SSC.
10 b. Identification of age-related degradation sites, site-specific aging !
11 mechanisms, and root causes, when practicable, for structures and 12 components listed pursuant to 2.c. Appendix A summarizes the age-13 related degradation processes that should be discussed.
14 c. For structures and components important to license renewal, a summary 15 discussion of the evaluation of key properties and parameters that 16 may change with time and that are affected by plant operational and 17 service conditions. The initial values at the start of operating 18 life of these properties and parameters (such as fatigue cycle life, 19 cable insulation dielectric strength, fracture toughness, tensile 20 strength, and pressure boundary wall thickness) as established by l 21 measurement, analyses, or qualifications should be included, along 22 with results of evaluations of past operating environments and service 23 conditions to determine the rates of change experienced and residual l
24 values for these properties and parameters. This summary should also 25 include a discussion of changes to analyses resulting from age-related 26 degradation evaluations. These values should be used in trending ,
l 27 and analyses to establish predicted extended operating lives and to i 28 identify actions needed to maintain key properties and parameters 29 within acceptable limits during the renewal term. (See Appendix A of l 30 this regulatory guide for further details.)
! 31 d. An identification of the structures and components listed pursuant to 32 3.a that are subject to established effective programs, including i
33 a technical justification of how these programs effectively manage 11
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'I age-related degradation. Reference may be made to'the list provided 2
pursuant to 2.f provided this list clearly identifies all components 3
and structures associated _with a particular system that are subject 4 to established effective programs. Descriptions of established effec-5 tive programs either should be provided as part of the license renewal 6
FSAR or incorporated by reference from the'most recent update of the 7
facility FSAR or any other material referenced in the facility's docket 8 (such as the FSAR, Quality Assurance' Manual, Inservice Inspection and 9 Testing Programs, and training programs). The program description 10 should not be a general description of the overall program but sho'ld u
11 be specific and justify why the program is effective in managing the 12 age related degradation identified pursuant to 3.b and 3.c and 13 described in Appendix A.
14 e. A description of actions taken or to be taken to manage age-related 15- degradation in SSCs important to license renewal not currently subject 16 to established effective programs and how programs resulting from ~
17 these actions will be implemented and maintained effectively from an 18 l age related degradation perspective. For' structures and components 19 applicable to the system under consideration and included in the list 20 in 2.c but not included in the list in 2.f, this description should 21 contain proposed revisions to maintenance or other program elements, 22 including administrative control , that will be implemented and con-23 trolled throughout the renewal period to manage age-related degrada-24 tion in these components. Alternatively, technical evaluations in 25 the facility docket that provide adequate assurance that the SSCs 26 will not degrade below acceptable levels of safety during the renewal 27 term may be provided or referenced.
28 f. A summary of all current exemptions granted pursuant to 10 CFR 50.12 29 and reliefs granted pursuant to 10 CFR'50.55a(a)(3). For exemptions 30 or reliefs that were granted based on an assumed service life or period
- 31 of operation bounded by the original license term of the facility, a 32 justification for continuing these exemptions and reliefs must be 33 provided. A reference to the list provided pursuant to 2.h will 34 suffice, provided this list clearly identifies all current exemptions 12 9
i 1 and reliefs associated with each system. Justification for i 2 continuation may be supplied either in each system section or with 3 the list in 2.h.
4 .g. A description of actions to be taken with respect to any proposed 5 modifications to the facility or its administrative control procedures 6 resulting from the integrated plant assessment [10 CFR 54.21(a)] or 7 exemptions and reliefs described in 3.f. A reference to the list 8 provided pursuant to 2.1 will suffice, provided this list clearly :
9 identifies all proposed modifications associated with each system.
10 h. For existing and new programs identified as necessary for managing l
11 age-related degradation, a description of how these programs are or 12 will be implemented and controlled to ensure that their effectiveness '
13 is maintained throughout the renewal term.
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l 14 1. A description of the methods to be employed in obtaining and 15 maintaining records of the documentation described in this section or 16 to be generated in the course of performing activities described in ]
17 this section. The description should identify which records are to l
18 be kept, in what form, and over what period of time. Records that 19 permit verification that all SSCs that are important to license renewal l 20 meet their specific performance requirements should be retained in an 21 auditable and retrievable form for the renewal term plus whatever l 22 additional period is required in accordance with the licensing basis.
I 23 Additions or other changes to the Technical Specifications pertaining to i l 24 age-related degradation may be necessary to account for modifications in the 25 plant design or limitations on plant operations during the renewal term. If '
26 such changes are deemed necessary, the license renewal application (described j 27 in 1.1 Formal Application) should specifically request such changes and should -
l L 28 provide the technical justification for the changes. Such changes should be 29 limited to those necessary to address effects of age-related degradation.
30 Items 3.a through 3.1, above, specify information that should be included i 31 in the FSAR supplement for each SSC important to license renewal. Additional 32 information that should be submitted as part of a license renewal application j l.
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l 1 or compiled and retained by the licensee is described in Table II. Some of the 2 information items listed in Table II should be included in the FSAR supplement ,
! 3 (information described in Regulatory Position 1.2.2). The remainder of.the l 4 information specified in Table II should be submitted as part of the application 5 but independent of the FSAR supplement or should be retained by the licensee as 6 appropriate. !
7 2. TECHNICAL INFORMATION CONTENT 8 As required by 10 CFR 54.21, "The FSAR supplement (that presents the ,
9 technical information requirement for license renewal) must include an evalua-10 tion of the aging mechanisms that are present and that result in degradation of '
11 the plant's systems, structures, and components, and a demonstration that the i 12 effects of such degradation will be effectively managed throughout the renewal 13 ters." To meet these requirements, the evaluation must cover those SSCs that 14 are important to license renewal and must be based on principles of understand-15 ing and managing age-related degradation. This Regulatory Position 2 provides l
16 guidance on selection criteria for SSCs important to license renewal and on 1 17 programs and practices for understanding and managing aging that are acceptable 18 to the NRC staff as the bases for the " evaluation" and " demonstration" required 19 by 10 CFR 54.21 as noted above.
20 2.1 Types of SSCs for Which Aging Should Be Considered for License Renewal 21 The process for identifying SSCs important to license renewal and their 22 constituent structures and components requiring evaluation of age-related de-23 gradation is illustrated in Figure 1A. Figures 1A and IB also summarize some of 24 the information to be collected by the licensee and either submitted as part of 25 a license renewal application or retained in auditable, retrievable form for the 26 ters of the renewed operating license. Figure IB delineates approaches to infor-l 27 mation collection and evaluation for a licensee to demonstrate an acceptable 28 level of understanding of age-related degradation in those structures and compon-29 ents identified through the selection process illustrated in Figure 1A. Accept-l 30 able management of age-related degradation for structures and components that are
, 31 important to license renewal requires evaluations of the effectiveness of exist-l 32 ing programs and practices for timely mitigation of the age-related degradation.
I 14 3 I
l l
-1 Figure 18 also delineates a process for identifying deficiencies in existing 2 programs for addressing age-related degradation in specific structures and com- !
3 ponents and for specifying-actions to be pursued in addressing these deficiencies !
4 in support of a license renewal application. '
5 Generic functional plant SSCs that are composed, wholly or partially, of 6 SSCs important to license renewal are listed in Table III. The information 7
l in Table III is included in this~ regulatory guide as a supplement to Table I. '
8- This.information provides guidance to the licensee for compiling the plant- ;
9 specific list of SSCs important to license renewal and is not intended to be ;
10 all-inclusive. Appendix B provides a detailed hierarchy of systems and compo-11 nents for safety categories representative of a typical pressurized water reac-12 tor-(PWR) and a typical boiling water reactor (BWR). The representative systems, l
13 structures, and components listed in Appendix B are referenced to Table III, l 14 " Generic Functional Nuclear Power Plant SSCs Important to License Renewal," to l 15 corresponding sections of the." Standard Review Plan" (NUREG 0800), and to the 16 Standard Technical Specifications. For plants to which these documents apply, o i 17 they provide more detailed information on system functions, configurations, 18 limitations, testing needs, habitability limits'for personnel, and safe environ-19- mental limits for vital equipment.
20 2.2 Integrated Plant Assessment - Programs for Addressing Age-Related 21 Degradation
- l l 22 Elements of the integrated plant assessment for the evaluation of age-
< 23 -related degradation should be based on sound principles and practices for under-24 standing and managing aging. SSC-specific understanding of aging mechanisms and I 25 the degradation sites at which they operate is useful to evaluate the effective-l 26 ness of existing programs and replacement / refurbishment practices for managing i 27 aging or to develop acceptable new programs or practices. An established effec-28 tive program should include, but is not limited to, inspection, surveillance, 29 maintenance, trending, recordkeeping, replacement / refurbishment, and the assess-30 ment of operational life for the purpose of timely mitigation of the effects of i 31 age-related degradation.
i
- 32 The prime criteria for such a program are that it be documented and that j 33 it ensures that SSCs important to license renewal will continue to perform 34 35 *For expanded discussions, see Appendix A to this regulatory guide.
15 ,
1 adequately, thus ensuring that the current licensing basis will be maintained 2 during the renewal period. In addition, an established effective program will 3 (1) be clearly defined and documented in the FSAR, (2) be approved by onsite 4 review committees, (3) be implemented by the facility operating procedures, (4) 5 establish documented acceptance criteria against which the need for corrective 6 action is evaluated to ensure that age-related degradation will not directly or
.7 indirectly prevent SSCs from performing their intended functions, and (5) ensure 8 that corrective action is taken when applicable acceptance criteria are not 9 met. Programs for understanding and managing aging should be implemented and 10- maintained through a system of specific administrative controls that ensures 11 continuing program effectiveness throughout the ters of a renewed license.
12 Requirements for a licensee to demonstrate the adequacy of the plant program 33 for addressing age-related degradation in structures and components important to 14 license renewal are specified in 10 CFR 54.21. These are (1) by substantiating 15 that established programs (ongoing programs that are currently in place) are 16 effective in ensuring the capability of SSCs important to license renewal to per-17 form their safety functicns throughout the renewal period, (2) by taking actions 18 or committing to actions to manage age-related degradation not adequately ad-19 dressed by established programs [10 CFR 54.21(a)(4)], or (3) by demonstrating 20 that age-related degradation is not significant with respect to the current li-21 censing basis [10 CFR 54.21(a)(4)]. Adequacy of the program for addressing 22- age-related degradation must be demonstrated for each structure, component, or 23 group of similar components important to license renewal using one of the afore-
- 24. mentioned methods. Both existing program'. and new actions taken or to be taken 25 to manage age-related degradation are subject to the same effectiveness criteria.
26 Criteria that relate to program structure and administration are cited in the 27 preceding paragraph. Criteria that relate to established effective programs and 28 to new actions taken or to be taken to manage aging are discussed in Regulatory 29 Positions 2.2.1 and 2.2.2, iespectively. Regulatory Position 2.2.3 deals with 30 the exclusion of structures and components not subject to significant age-related 31 degradation during the renewed licensing term. Technical criteria for practices 32 employed in understanding and managing aging are described in Regulatory Posi-33 tions 2.2.4 and 2.2.5, respectively.
16
L 1~ 2.2.1 Established Effective Programs f
( 2 Established effective programs for managing age-related degradation are
! 3 defined in 10 CFR 54.3(a). The structural'and administrative criteria provided l 4 in that definition plus the technical criteria in Regulatory Positions 2.2.4 5 ahd 2.2.5 should be applied by a licensee in evaluating the current plant program l l 6 for managing aging in structures and consponents important to license renewal. ;
l-7 The methodology employed in performing these evaluations should be described, j 8 and results of the evaluations should be provided for all important-to-license-l 9 renewal structures and components that are subject to established programs.
- 10 Two essential products of these evaluations will be identification of (1) those
! l 11 important-to-license-renewal structures and components that are regularly l 12 inspected and routinely replaced / refurbished at defined intervals and (2) struc-13 tures and components for which established programs do not effectively address l
- 14. age-related degradation. l 15 2.2.2 Actions To Be Taken To Manage Age-Related Degradation
.i l
16 It is expected that some important-to-license-renewal structures and 17 components with potentially significant age-related degradation will not be l
l 18 subject to programs that address that degradation and that others will be subject 19 to programs that are not fully effective. Such structures and components should 20 be identified, and the bases for actions that have been taken or will be taken 21 to manage age-related degradation in these structures and components should be 22 described by the licensee. These actions and~ judgments concerning adequacy 23 should be based on the same criteria cited in Regulatory Position 2.2.1 for 24 established effective programs. _
25 2.2.3 Structures and Components Not Subject to Significant Age-Related 26 Degradation 27 For some structures and components important to license renewal, age-related 28 degradation may not lead to significant reduction in the capacity of the struc-
{
29 ture or component to perform its safety functions, i.e., functions defined as j 30 important to license renewal. If a licensee explicitly demonstrates that age- :
j 31 related degradation of an important-to-license-renewal structure or component
! 17 r
I I will not compromise the current licensing basis during the license renewal 2 ters, that structure or component may be excluded from further consideration of :
3 age-related degradation.
4 2.2.4 Summary of Acceptable Programs and Practices for Understanding 5 Aging 6 Programs to assess age-related degradation processes in structures and 7 components important to license renewal should be implemented on a plant-specific 8 basis by qualified licensee staff using state-of-the-art knowledge of age-related 9 degradation in NPPs. Efforts to understand aging mechanisms and degradation 10 should be systematically structured, as illustrated in Figure 1B. For some 11 structures and components, e.g., reactor pressure vessel shells, it will be 12 necessary to evaluate design, materials, and environmental and operational 13 stressors and their interactions. Analysis of these factors will lead to iden-14 tification of potential aging mechanisms, degradation sites, and when practicable, 15 root causes. This information is the basis for developing and implementing
~
16 programs for monitoring and timely mitigation of age-related degradation. Not 17 all important-to-license-renewal structures and components require in-depth 18 evaluation of the basic factors that contribute to age-related degradation. i 19 For some structures and components, reference to empirical information such as 20 operational records, manufacturers' information, test data, and onging regula-21 tory requirements will be sufficient. To the extent practicable, quipment 22 designers and manufacturers should be requested to identify aging mechanisms, 23 rates, and any other pertinent information that they may possess. For other 24 structures and components, descriptions of surveillance or. replacement programs 25 will suffice.
26 The process of assessing age-related degradation involves integrating the 27 relevant materials science concepts that describe degradation processes with 28 plant , SSC , and structure-or-component-specific design and operational informa-29 tion to understand aging mechanisms, degradation sites and rates, and the conse-30 quences of degradation with respect to plant safety. The individual and inter- l 31 active influences of structure and component design, constituent materials, and l 32 both normal and abnormal stressors and environments establish the feasibility of 33 the mechanisms that can degrade SSCs important to license renewal and determine 34 the rates at which degradation progresses. An effective program to understand 18 i
( ,
i 1 aging will selectively integrate a sound understanding of these basic principles 2 with plant-specific design, operational experience, manufacturers' and design 3 information, research and test data, applicable regulatory instruments and 4 requirements, and qualified technical judgment to characterize age-related de-l 5 gradation processes that are operative in important-to-license-renewal structures ;
i 6 and components. These characterizations should be expressed in terms of specific l l 7 degradation sites, site-specific aging mechanisms, root causes for degradation l
8 when practicable, and projected effects of degradation on SSC functions. :
l ,
9 2.2.5 Summary of Acceptable Programs and Replacement / Refurbishment :
10 Practices for Managing Aging !
11 Acceptable practices for managing aging in structures and components for 12 which no defined replacement / refurbishment programs exist may employ combined 13 mechanistic and empirical approaches for understanding aging mechanisms and 14 identifying degradation sites, and, where practicable, root causes. This infor-15 mation may form the basis-for inspection, surveillance, condition monitoring, 16 test methods and frequencies, and residual lifetime evaluations that will dic-l 17 tate timely and offective preventive and corrective maintenance methods and .
j 18 frequencies, as well as associated recordkeeping needs during the renewal term.
19 Monitoring methods (e.g., inspection, surveillance, testing, condition 20 monitoring) should reflect mechanistic and empirical assessments performed by 21 .the licensee staff in their efforts to assess and mitigate age-related degrada-22 tion. These methods should employ state of-the-art nondestructive examination 23 (e.g., ultrasonic testing, signature analysis, vibration analysic, dielectric 24 performance measurements) and other measuring techniques, performed by qualified 25 staff. Measurement results should be trended and analyzed with respect to 26 implications for the residual lifetime of SSCs important to license renewal.
27 Practices for managing aging in structures and components that are regularly inspected and routinely replaced or refurbished at defined intervals should be 28 29 evaluated for the adequacy of the inspection and replacement or refurbishment 30 programs to ensure timely mitigation of age-related degradation during the 31 renewal term. The evaluation process should include reviews of the operational 32 experience and, as appropriate, design and manufacturers' information, known i 33 aging mechanisms, and other available information.
}
i 19 i
= = - - - - - - - - . - - . , , w w -..w & n - -- -.- .,
. - . - - - _ - - . _ - . _. - . - - - ~ . . - _ . - - _ . . -
1 The objective of aging management in support of license renewal should be 2 to ensure that SSCs important to license renewal are subject to surveillance 3
and maintenance that control, at intervals commensurate with expected component !
- 4. _ lifetimes, processes that could degrade their operability and reliability. f 5
The maintenance program is the principal vehicle through which age-related !
6 degradation is managed. Operational and maintenance records and input from >
7 '
monitoring programs should be employed in the maintenance program for scoping 8 and scheduling both preventive and corrective maintenance activities intended i 9 to manage age-related degradation. These activities should be carried out by 10 experienced, qualified maintenance personnel and should lead to needed servic-11 ing, repair, refurbishment, or replacement with a frequency sufficient to ensure 12 the capability of SSCs important to license renewal to perform their safety >
13 functions during the renewal term. In evaluating the effectiveness of mainte-14 nance in managing aging, the maintenance and surveillance intervals should be 15 considered along with (1) the probability of defect detection and diagnosis and 16 (2) the probability of effective defect correction given defect detection and I
-17 diagnosis.
18 A critical first step in implementing an aging management program is to 19 define those surveillance, maintenance, or other mitigative program elements to 20 be implemented to deal with the degradation processes revealed by programmatic 21 efforts to understand aging (Regulatory Position 2.2.4). These program elements.
22 may include inspections, tests using a wide variety of nondestructive examination 23 . (NDE) and other methods, general surveillances, condition monitoring, diagnostic 24 and root-cause analysis, preventive maintenance, corrective maintenance, predic-25 tive maintenance, and reliability-centered maintenance. The primary goal should 26 be to develop effective aging management practices implemented through replace-27 ment, refurbishment, and repair programs that accurately reflect residual life-28 times for specific structures and components and the safety significance of 29 anticipated degradation. The effectiveness of aging management programs should
. 30 be evaluated for specific structures and components using guidance such as that 31 contained in relevant codes and standards and approved industry technical reports. l 32 The process for evaluating effectiveness should reveal those deficiencies that 33 require correction through improved or new programs for managing aging. This 34 process is outlined schematically in Figure 1B. An accurate assessment of !
35 current program effectiveness, based on the principles of understanding aging l 36 summarized in Regulatory Position 2.2.4 and in Appendix A, and implementation 1
20
- p. y , . - - -
--.v, -
.m -
1 of program enhancements to address revealed deficiencies in aging management 2 practices are prerequisites to operating license renewal in nuclear power plants.
3 An important aspect of plant surveillance and maintenance programs is 4 retention of essential data in complete, auditable, easily retrievable records.
5 The records system and its contents should conform to good maintenance practices 6 as well as the requirements of Appendix B to 10 CFR Part 50 and the plant quality 7 assurance r*ogram to the extent that these documents apply to SSCs important to 8 license renewal. A record of the documentation required by'or necessary to '
9 document compliance with the provisions of 10 CFR Part 54 and a record of the 10 administrative processes for controlling changes to such documents should be 11 retained by the licensee in an auditable and retrievable form for the renewal 12 term. This record should include a listing of and the justification for those 13 structures, systems, and components important to license renewal and included 14 in established effective programs as defined in 10 CFR 54.3(a) or subject to 15 actions taken or to be taken. Records and related data should be employed to 16 the extent practicable in trending degradation processes, thereby providing 17 assurance of controlled, timely maintenance. Trends that are based on data 18 contained in the maintenance records and that account for both normal and off-19 normal operating conditions should be used to monitor the effectiveness of aging 20 management for selected structures and components. I l
21 D. IMPLEMENTATION i 22 The purpose of this section is to provide information to applicants 23 regarding the NRC staff's plans for using this regulatory guide.
24 This draft guide has been released to encourage public participation in its !
25 development. Except in those cases in which an applicant proposes an acceptable 26 alternative method for complying with specified portions of the Commission's 27 regulations, the method to be described in the active guide reflecting public l 28 comments will be used by the NRC staff in evaluating applications for renewal l 29 of operating licenses for commercial nuclear power plants.
l 21 l $
I
1 FIGURE 1 2 PROCESS FOR SELECTING SYSTEMS, STRUCTURES, AND COMPONENTS INPORTANT ,
3 TO LICENSE RENEWAL AND FOR UNDERSTANDING AND MANAGING AGE-RELATED 4 DEGRADATION 5 Figures 1A and IB constitute a flow chart.that outlines a process, l 6 acceptable to the NRC staff for license renewal purposes, for selecting indivi- -!
7 dual structures and components. that are constituent elements of systems, struc- i tures, and components (SSCs) important to license renewal. This process is also 8
9 acceptable to the NRC staff for developing and implementing programs for under-10 standing and managing age-related degradation in these structures and components :
j 11 during license renewal term. I 12 Figure 1A portrays the process for selecting SSCs important to license I 13 renewal and their constituent structures and components. , Input to this process
.14 is defined by the four types of SSCs included in the definition of " systems, 15 structures, and components important to license renewal" [10 CFR 54.3(a)]. Each 16 of the input SSCs is subdivided into individual structures and components that I 17 are then screened [using the criteria contained in 10 CFR 54.21(a)(2)] to yield )
18 those structures and components that require evaluation of age-related degrada-tion (block 11, Figure 1A). This collection of structures and components con- l 19 20 stitutes the input to the continuation of the process of evaluating age related 21 degradation as part of the Integrated Plant Assessment as shown in Figure 18.
'22
~
Two methods are presented as guidance for determining-the scope and depth !
of analysis necessary to define age-related degradation mechanisms and to l
23 24 evaluate the adequacy of-aging management programs. The first method may be l 25 applicable when evaluating structures and components that are regularly i i
26 inspected and replaced or refurbished routinely at defined intervals. These 27 structures and components include items such as batteries, relays, and selected l snubbers. Programs that may be considered to provide for management of aging .i 28 ;
29 in these structures and components include inspection, surveillance, testing, 30 condition monitoring, residual lifetime evaluations, predictive maintenance, !
31 preventive maintenance, reliability centered maintenance, and other maintenance 32 or similar programs that provide timely mitigation of age-related degradation _ {
33~ during the license' renewal ters. The second method may be applicable when eval- l 34 uating structures and components that are not regularly inspected and routinely- l replaced or refurbished. Such structures and components typically were designed 35 l 36 for and expected to be in place.for the original 40 years of plant operation.
37 These "long-lived" structures and components include items such as the reactor ,
coolant system, large-diameter piping, the reactor pressure vessel, steam gen- (
38 !
39 erators, and cables.
40 Structures and components that are regularly inspected and routinely ;
41 replaced or refurbished are not automatically considered as subject to "estab- i 42 lished effective programs." The effectiveness of aging management programs for !
these structures and components should be evaluated based on a review of the :
43 i
~44 adequacy of the inspection and replacement or refurbishment programs for tindy i 45 mitigation of age-related degradation during the license renewal term. An :
46 acceptable approach would be to evaluate operational experience, replacement or 47 refurbishment intervals, and, as appropriate, relevant design and manufacturers' ,
48 information, known aging mechanisms, and other available information. Based on l 49 this review and a conclusion that the structure or component will remain func- j 50 tional during the defined interval between replacement or refurbishment, an 51 applicant may establish that the current program is adequate. When the ongoing ;
i 22
i 1 replacement program is demonstrated to be adequate for timely mitigation of age-2 related degradation and if it is in the FSAR, approved by the onsite review 3 committee, and implemented by the facility operating procedure, it is considered 4 to be an " established effective program."
5 When the ongoing replacement or refurbishment programs are not adequate for 6 timely mitigation of age-related degradation, the licensee should describe 7 revisions to the replacement or refurbishment program and demonstrate its ade-8 quacy or perform detailed mechanistic analyses. The bases should be provided 9 for actions taken or to be taken to manage age-related degradation during the 10 license renewal term.
11 Structures and components that are not regularly inspected and routinely 12 replaced or refurbished may be evaluated based on a detailed mechanistic anal-13 ysis of age-related degradation mechanisms. When the analysis indicates that 14 age-related degradation is not significant with respect to the current licensing 15 basis throughout the license renewal term, the result of the analysis should be 16 documented. For those structures and components susceptible to significant age-17 related degradation mechanisms, evaluations should be made to determine if they 18 are subject to an established effective program. A list of structures and com-19 ponents identified as being subject to an established effective program should 20 be provided as well as descriptions of the programs and the basis for continuing 21 them through the license renewal term. For those structures and components that 22 are not subject to an established effective program, the basis for actions taken
- 23 or to be taken to manage age-related degradation during the license renewal term i 24 should be described and provided.
Throughout Figures 1A and 18, individual activities and decision points 25 26 have been referenced as appropriate to the specific parts and subparts of 10 CFR
- 27 Part 54 that provide their justification. These references to 10 CFR Part 54 28 are denoted within appropriate nodes in the figures. ;
I l
l l
23 l
i
.- - - . - - . . ._. _.~ - , - - . . ..
l 54.3 l 54.21(a)l tjst documents I
m idenufying portions of the CLB
, "8 relevant to the integrated plant CW F assessment [2 9
t t , t t
$4.3(a) l 54.3(a) l 54.3(a) l 54.3(a) l i identify any, including idenHfy safety-related identify SSCs used in a nonsafety-related, SSCs idonefy post-accident systems structures' safety analysis or plant whose failure could prevent monitodng equipment 4
and components evaluation for the settsfactory accompNehment as defined in 10 CFR IIC*nsing beels of required safety functions 50.49(b)(3)
(SSCs) 7 y of safety-related SSCs g y I I I I
- N l 54.21(a)(1)l List SSCs important to license renewal (ILR)
+
54.21(a)(2)l List structures and components (SCs) that are consutuent elements of the SSCs ILR Q 1 r 54.21(a)(2)
Doesthe 90 contribute to the perform-ance of a safety funcuan* of F No Provide an SSCILR or could failure of the SC prevent an SSC ILR L jusufication kJ from performing hs intended safety function?
9
- Safety funcdon: Any function Y" that causes an SSC to be Identiflod as important to license 54.21(a)(2)l renewel;this le not limited to the List SCe requiring narTow definition of a safety evaluadon of w elo w function associated with safety.
degrade #en I11 reisted equipment. -
t Figure 1B 1 FIGURE 1A Integrated Plant Assessment--Identification of Important-To-2 License-Renewal SSCs and SCs Requiring Evaluation of Age-Related 3 Degradation 24 i
l l
Figure 1 A 1P 1
54.21(a)(2)l l List SCs requiring evaluaSon r
of :;;: ; ' f 4 f:f:r, -
W 1 l
l l
le the SC sut4act to regular
~
inspection and routinely Yee replaced / refurbished at donned )
1P intervals? 1 P Perform detailed mechanistic analysis that Evaluate operational experience and the considers design,insterials of construction, inspection and replacement / refurbishment operational and environmental stressors, peograms and, as appropriate, design and l and their interactions that may cause or manufacturers
- Information, known aging l contribule to age-related degradation mechanisms, and other available i Q informah y '
I 1r Can it be I"N#O demonstrated that 7 inapoction and replacemen is age <eleted F54.21(a)(4)(1)l No refurbishment programs for the y degradah significant with List such idendfied SCe SC are adequate for timely mitigation of
'**Pect to the CLB through and demonstrate by related degredation to ensure tha the renewedliconee g walus h EJ the CLB will be maintained tenn? through the renewed N5/Yes Ncense term?
7f Yes 1P 1r 1 r II Describe the revleions to the inspection and r ;' x t/ refurbishment programs to 54.21(a)(3) demonstrate adequacy or perform detailed is the SC mechanis6c analyses. See node 13, above, for Yes @ 2 en 6 No guidance on performing analyses Q Nehed offeceve program k manage age-related degradation during the .
1y renewal term? ,
3 3 7 54.21(a)(3) l [ 54.21(a)(4)(1)l Ust auch idondfied SCs, the associated Describe and provide the basis for established offective program (s), and the beels for continuing them through actions taken or to be taken to manage age related degradation the renewed license term 5 g 1 FIGURE IB Integrated Plant Assessment--Evaluation of Age-Related Degradation 25
i Sources of Information and instructions for Implementing ProCeSeeS Described in Figures 1 A and 1B. ;
1 Corryte and memtern all documeras descritmg the current liconemg t>ases (CLB) rt an au6tatWe and removetWe form as per 10 CFR $421(a). !
2 Sulwntt a let of documente identi'ylng portene of the CLB relevant to the integrmed plant aseosoment (IPA) as por 10 CFR 5421(a).
3, 4, 5, 8 These four categories are doened by to CFR 54 3(a) as being important to teense renewal ? , ..a of systems, struchwee, and componoats (SSCe) to me -
approprises categories should be accor'pl6ehed by stureng wem en plant SSCe and doenbueng mese among the four canegones by applymg guideenes such as thoes I contained in me sources referenced in me fotoedng tour notes. SSCe that 6t none of the leur categorlos are not important to hoones renewal and require no furmer considersson for teense renewal purposes.
3 Q List, Final Sepoly Analysis Report (FSAR), Quality Assurance Program and al omer elemente at the Cunent Liconome Beels as denned h 10 CFR 54. SSCe used in I a selety analysis or plant evalusten for me Liconoing Bene include, but are not Ilmited to SSCe identMled h me FSAR, the Teeniced Speerncetone, Plant Operseons Manuale, ciping and hetrumentoson Degrams (PalDo),TatNo 111 of mis regulatory guide (Genene Funceonel SSCe important to License Renewer), and evaluetone outwnitted to show comphones wie me Commrosion*e reguissons such as Anticipeopd Transente iMthout Scram (ATWS), Sestion Bloomut, Fire Protection, Pressured i
Thermal Shock (PTS), and Environmental Oumhficahon.
4 Current Liconome Basis.
5 FSAR, P&lDs, as-bbilt current dromings, plant modittation records, mester equipment lot, plant anfiguranon control system date.
6 10 CFR 80.40 and mesociated reguisery guides (e.g. RG.1 Ag, RG.1.g7), FSAR.
7 Describe m6thodology for identfying SSCe important a teense renewel, as defined h 10CFR54.3(a), and submit a let of the identted SSCe se part of the inangrened Plant Amoseement (IPA).
i 4,g,10,11 Desante methodology for identfring moes structuos and componente (SCs) met are conettuent elements of the SSce important to liconee renewal and met roeptre evolusson of agwelened degradefon, y
11 Prohmbinatic rielt- . ,t (PRA) techrnques may eleo be used to supplement me determinienc approach shoem in Figure 1 A by addng edetonal SCe to be tot of i those regJiring evalueton of ageralmted ,; . _ ' ' ~ j 12 Plant records and CLB. l 13,14 Iraput ham me Nucieer Plant Aging Reseenh (NPAR) Program: codes and standerde; nondestructive exammaton (NDE): Industry studies, and other programe related i to effects of meterial veistdes, aa -,;. and airesecre on agueisted degradeton.
M earectortraton includes inHind design, cone *ruchon, installation plus changes introduced by modilcaton, marntenance,or repiscoment.
samments derin thwn both emdronmental and service condtone
. Err:6cnmonimi conditone include rede6on, temperstav, asnosphere, humidty, and chemical environment for alt design basis events ;
- Service corwttiene include steady-staan, cychc, or other tenaient loedings imposed during normal opere6cn, lostng, or oftnormal events. Moeanical and elecmcel j loadings predommeio 15,16 For moes structures and componente (SCs) sor which it can be demonstrated that age-related degradeton is not significent with roeped to the CLB through the renewed t license period, no further actone need be taken to manage their v . . _l degradeson. !
I 17,is Decisions to be bened upon criterie developed by the beenees and P ma reviser by me NRC. Need for rHlopth evalueton wig reRect such conesserotone as setety i
esgniflamnce, fmpure tzmesquence, erstem and degradaten procoes compiedty,inteneiy er caneervotom of cunent mitgemon program, and other todore deemed
?
Irnportant by the licensee and NRC for spec 6Ac cases. Supportng informaton for those dociosons wm derin kom various spent-opecte sourme inclueng reconis of operatone and maintenance, tecnical specinessons, mornenenon/survoisance procedures, inservice inspecten program (ash 8E Socson XI), PTS anefyese.
- preventive-prodcove maintenance prograrre.
1g,20,21 Programs for managmg T .-- degradation include inspeden, testing, surveillance, condson monnor.ng, foot cause analyse. .JL% .. W.--, -- a t t
riait-basseretietetty contored maintonence, record hoopwql and troneng, omer acevitse, remdud ble aseseement, and reopenses e changes in operseng and design paramotore and envvenmente Relevant informaten een be found in records of operatione and mensenance, toenical spoeshcatkme, meineenenosteuneelance pecedures, inservice inspecton program (ASME Secton XI), pump and valve testing programe, environmental C programe, preventve-presc9ve mer'tenance program records, and the current body of regulosory -G. .: ,
Note A: Roeuhe of acevites 2,7,8,10,11,16,1g,20, and 21 should be submeed with the license renewal application. These roeuHe plus results of activites 1.3.4,5,6,13 f 14, and 17 should be documented and retained by me liconese in audrtable, removable form.
i I
l l
1 TABLE I l
2 INFORMATION TO BE INCLUDED IN THE FSAR SUPPLEMEN( -- -
3 GENERAL AND SSC SPECIFIC l
l 4 Part A: General Information and Discussion
- l
! 5 (Part A should include, but is not limited to, 6 the following types of information.) '
7 1. Introduction I i 8 1.1 Purpose of the FSAR Supplement 9 1.2 Scope of the FSAR Supplement 10 1.3 Definitions of Terms 11 1.4 Organization of the FSAR Supplement 12 1. 5 Relationship of the FSAR Supplement to 10 CFR 54 l 13 2. Conformance to Applicable Regulatory Guides i 14 3. Listing and Summary of Material Incorporated by Reference 15 4. Description of Integrated Plant Assessment (see Regulatory Position 1.2.2 16 of this regulatory guide) 17 5. Information Required by 10 CFR 54.21 (see Regulatory Position 1.2.2 of 18 this regulatory guide) i 19 6. Deviations from SRP-LR Acceptance Criteria * ]
20 Part B: NPP Systems .
21 (Numbers in brackets refer to corresponding chapters in the 22 FSAR and in Regulatory Guide 1.70.)
23 1. Nuclear Systems - the reactor core and those systems and subsystems that l 24 monitor and control the core's reactivity, remove heat from the core, and 25 otherwise directly support the safe operation of the reactor.
26 1.1 Reactor Pressure Vessel (includes reactor core and internals) [5,3]
27 28 *An applicant for a renewed license will find it useful to consult NUREG-1299, 29 " Standard Review Plan for License Renewal," the SRP-LR, for descriptions of :
j 30 criteria and review procedures to be applied by the NRC to applications for '
31 license renewal.
27
1 TABLE I (contd) 2 1.2 Reactor Coolant System (includes piping, reactor coolant pumps, and 3 steam generators) [5,3] -
4 1. 3 Reactor Control System [3]
5 1.4 Control Rod Drive System [3]
6 1.5 Reactor Protection System [7,3]
7 1.6 Nuclear Monitoring / Nuclear Instrumentation System [7,3] i i
8 1.7 Reactor Water Cleanup System (BWR) [5,6]
9 1.8 Standby Liquid Control System (BWR) [9,3,6]
10 1.9 Chemical and Volume Control System and Emergency Boration (PWR) :
11 [3,5,6,9]
12 2. Engineered Safety Features - systems, other than containment systems, that 13 are used to mitigate the effects of a reactor uccident such as a LOCA.
14 2.1 Engineered Safety Features Actuation System (PWR) [7,3] >
15 2.2 Safety Injection Systems [3,5,6]
16 2.2.1 Reactor Core Isolation Cooling (BWR) 17 2.2.2 High Pressure and Intermediate Pressure Safety Injection !
18 System (PWR) 19 2.2.3 Core Flood System (PWR) 20 2.2.4 RHR/ Low Pressure Safety (Core) Injection (includes shutdown 21 cooling function) i 22 2.2.5 Core Spray Systems (BWR) 23 2.2.6 High Pressure Coolant Injection (HPCI) System (BWR) 24 2.3 Auxiliary Feedwater System (PWR) [3,6,10]
25 2.4 Automatic Depressurization System (BWR) [3,6] ,
26 2.5 Remote Shutdown System / Safe Shutdown Systems [7]
27 3. Containment Systems - the containment (primary and secondary, as applicable) !
28 and those systems needed to prevent containment over pressure, to prevent 29 excessive leakage from the containment to the environment, and to provide 30 a habitable atmosphere inside containment.
31 3.1 Primary Containment Structure [3,6]
2.8 I
f
! i l
1 1 - TABLE I (contd) i 2 3.2 Secondary Containment [3,6] l L
i 3 3.3 Containment Heat Removal. System.[6] I I
4 3.4 Containment Isolation System [6]'
l 5 3.5 Containment Purge System [6]
l 6 3.6 Standby Gas Treatment System (8WR) [6]'
l' I
7 3. 7 Containment Combustible Gas Centrol System [6]
8 3.8 Containment Spray System [6]
l 1
l 9 3.9 Containment Ventilation System [9] i 1
10 4 .' Electrical Systems - systems that supply electric power to the utility grid j 11 or other plant systems, or that are purely electric in nature. i 12 4.1 Main Power [8]
13 4.1.1 Protective Relaying and Controls -
14 4.2 P,lant AC Distribution System [8]
l 15 4.2.1 Essential Power System i l 16 4.2.2 Nonessential Power System ;
l l l
17 4.2.3 HPCS Power System (8WR) 18 4.3 Instrument and Control Power Systems [8]
19 4.3.1 DC Power System l
20 4.3.2 Instrument AC Power System i
L 21 4.4 Emergency Diesel Generators (EDG) [8,9]
, 22 4.4.1 EDG Instrumentation and Control Subsystem [8]
l 23 4.4.2 EDG Starting Subsystem [9]
24 4.4.3 EDG Cooling Subsystem [9]
[
25 4.4.4 EDG Fuel Oil Subsystem [9]
! 26 4.4.5 EDG. Lubricating Oil Subsystem [9]
i 27 4.5 Plant Essential Lighting System [8,9]
i 29 , j i
i
-.g- 4-,--. - - am - -m a r- + - '7t'
__ _ .. _ _ _ _ _ _ _ . _ _ . . . ~
l i .
! 1 TABLE I (contd) l 2 4.6 Plant Computer [7]
3 4.7 Switchyard [8]
4 4.7.1- DC Control Power System .
l 5 4.0 Information Systems Important to Safety [7]
l l 6 5. Process Auxiliary Systems - system and subsystems that support the plant
! 7 systems directly involved in the process of safely producing electrical l 8 power.
9 5.1 Offgas System (BWR) [11]
10 5.2 Radiation Monitoring System [12]
11 5.3 Component Cooling Water System [9] i l 12 5.4 Service Water System [9] l 13 5.5 Ultimate Heat Sink [9]
l l
14 5.6 Refueling System [9]
15 5.7 Spent Fuel Storage [9]
i 16 5.8 Compressed Air System [9]
17 6. Plant Auxiliary Systems - systems provided to suppart plant activities and 18 personnel. They are typically nonsafety systems. Design of these systems i 19 varies greatly from plant to plant.
20- 6.1 Fire Protection System [9]
21 6.2 Communications [9]
22 6.3 Control Room Habitability System [6]
23 6.4 Auxiliary HVAC Syr,tems [6]
24 Part C: Generic Components l 25 (These relate to various elements of the preceding 7 26 subsection dealing with NPP Systems) >
27 1. Mechanical 1 4
28 1.1 Piping 30 l l
I 1 TABLE I (contd) -
t 2 - 1. 2 Valves l 3 1.3. Pumps i
i 4 1.4 Heat Exchangers ,
! 5 1.5 Tanks and Vessels j 6 1.6 Equipment and Component Supports 7 2. Electrical l- 8 2.1 Cable and Wiring 9 2.2 Junctions l
10 2.3 Electrical Penetrations l
l 11 2.4 Relays, Circuit Breakers, and Switchgear i
l 12 2.5 Transformers .
13 2.6 Solenoid Operated Valves 14 2.7 Electric Motors 15 3. Instrumentation 16 3.1 Sensors l 17 3.2 Electronic Components 118 3.3 Electronic Devices 19 4. Civil Structures l
4 31 i
i 1
1 TABLE II I 2 TECHNICAL INFORMATION NEEDED FOR LICENSE RENEWAL (LR)'
3 _( Should demonstrate the current licensing basis for SSCs that are important 4 to license renewal and that should be subject to established effective 5 ' programs or subject to actions taken or to be taken to manage age-related 6 degradation during the license renewal.ters.) '
7 SUBMIT WITH I 8- TECHNICAL INFORMATION TO BE GENERATED AND DOCUMENTED IN LR APPLICATION?
9 THE FORM OF AUDITABLE, RETRIEVABLE RECORDS _ Y/N (Yes/No) 10 The principal vehicle for providing technical information 11 in' support of a license renewal application will be the 12 FSAR supplement described in detail in Regulatory Position 13 1.2 of this regulatory guide. The FSAR supplement, which 14 is to be submitted along with the Formal Application for 15 License Renewal described in Regulatory Position 1.1, will ,
16 contain or reference various compilations of technical ,
17 information including, but not limited to, the following: i
~18 1. The most recent update of the facility FSAR and any N 19 other manuals or program documents referenced in the 20 FSAR, reports such as the Quality Assurance Manual, -
21 Emergency Response Plans, Inservice Inspection and' 22 Testing Programs, and training programs. These l
-23 should be incorporated by- reference in the license 24 renewal application.
25 -2. A list of all current exemptions granted pursuant to Y )
- 26. 10 CFR 50.12 and reliefs granted pursuant to 10 CFR '
27 50.55(a)(3). For exemptions or reliefs that were 28 granted based on an assumed service life or period of
~29 operation bounded by the. original license term of the 30 facility, a justification for continuing these exemp-31 tions and reliefs shall be provided.
32 3. A description of any proposed modifications related Y 33 to age-related degradation that have been or will be 34 made to the facility or its administrative control 35 procedures resulting from the evaluation or analysis 36 required by number 2 above. .
37 4. A description of additions or other changes.to the Y 38 Technical Specifications as appropriate, including 39 technical bases for these changes, that will be 40 needed to account for the modifications to the plant L41 design, age-related degradation, or limitations on 42 plant operations during the' renewal ters. Technical 43 Specification changes should not be contained in the 44 FSAR supplement but should be contained and justified 45 .in the formal application.
32
i l
1 TABLE II. (contd) i
! 2 SUBMIT WITH .
3 TECHNICAL INFORMATION TO BE GENERATED AND DOCUMENTED IN LR APPLICATION? !
4 THE FORM 0F AUDITABLE, RETRIEVABLE RECORDS Y/N (Yes/No) 5 5. A facility-specific list of SSCs that are important Y 6 to license renewal as defined in 10 CFR 54.3(a) as l 7 required in 10 CFR 54.21(a)(1). Included with this 1 8 list should be a description of the process used to 9 identify SSCs important to license renewal (see 10 Figure 1A for a schematic of such a process). ;
i 11 6. A facility-specific list of structures and components Y 12 that are constituent elements of the SSCs important i 13 to license renewal, listed in number 5 above.
14 Included with this list should be a description of 15 the process used to identify the SCs.
16 7. Justification for conclusions that any selected Y .
17 structures and components do not contribute to the !
18 performance of a safety function of an SSC important i 19 to license renewal or that their failure would not 20 prevent an SSC important to license renewal from per-21 forming its intended safety function.
22 8. A list of structures and components requiring Y '
23 evaluation of age-related degradation as required in i 24 10 CFR 54.21(a)(2).
25 9. A list of structures and components whose age-related Y 26 degradation is not significant with respect to the '
l 27 CLB through the renewed license period and documenta-28 tion of the evaluations that support these findings 29 as required in 10 CFR 54.21(a)(4)(i).
30 10. A list of the structures and components subject to an Y 31 established effective program, the associated estab- l 32 lished effective programs, and the basis for con-33 tinuing them through the renewed license period as 34 required in 10 CFR 54.21(a)(3).
35 11. A description of and the basis for actions taken or Y 36 to be taken to manage age-related degradation as i 37 required in 10 CFR 54.21(a)(4)(1), including changes l 38 in the refurbishment / replacement program to .
. 39 demonstration adequacy.
40 12. For the structures and components or similar Y t 41 structure and component groups cited in the facility-42 specific list specified in number 6 above, identi-43 fication of degradation sites, site-specific l 44 mechanisms, and when practicable, root causes.
33
I r
l 1 TABLE II. (contd) 2 SUBMIT WITH 3 TECHNICAL INFORMATION TO BE GENERATED AND DOCUMENTED IN LR APPLICATION?
4 THE FORM OF AUDITABLE, RETRIEVABLE RECORDS Y/N (Yes/No) 5 13. For structures and components important to license Y 6 renewal, a summary discussion of the evaluation of 7 key properties and parameters that may change with 8 time and that are affected by NPP operational and 9 service conditions. The initial values at the start 10 of operating life of these properties and parcmeters 11 (such as fatigue cycle life, cable insulation dielec-12 tric strength, fracture toughness, tensile strength, 13 and pressure boundary wall thickness) as established l 14 by measurement, analyses, or qualifications should be 15 included, along with results of evaluations of past 16 operating environments and service conditions to 17 determine the rates of change experienced and resid-ual values for these properties and parameters. This 18 19 summary should also include a discussion of changes 20 to analyses resulting from age-related degradation 21 evaluations. These values should be used in trending 22 and analyses to establish predicted, extended operat- ,
, 23 ing lives and to identify actions needed to maintain l 24- key properties and parameters within acceptable limits l 25 during the renewal term. (See Appendix A of this l 26 regulatory guide for further details.)
27 14. A description, including the technical bases, for all Y 28 completed actions to incorporate the structures and 29 components listed in number 6 above into existing 30 maintenance, surveillance, and inspection programs.
31 These may include tests or inspections, mainter.ance 32 and surveillance, and references to generic technical 33 evaluations that provide assurance that the SSCs will 34 not degrade below acceptable levels of safety during 35 the renewal term.
36 15. A specific description of maintenance or other Y 37 program elements, including administrative controls, 38 that will be implemented to provide for needed addi-39 tional understanding and management of aging in 40 structures and components listed in number 6, above.
41 16. A description of the methods to be employed in Y 42 maintaining records of the documentation de w ribed in
! 43 this section or to be generated in the course of per-
, 44 forming activities prescribed by this section. This 45 should include identification of records to be kept, 46 in what form, and over what period of time. Records 47 that permit verification that all SSCs that are l 34 s
- 1
l l
i- ,
1 TABLE II. (contd) i l 2 SUBMIT WITH l 3 TECHNICAL INFORMATION TO BE GENERATED AND DOCUMENTED IN LR APPLICATION?
l 4 THE FORM OF AUDITABLE, RETRIEVABLE RECORDS Y/N (Yes/No) 5 important to license renewal meet their specific l 6 performance requirements should be retained in an
! 7 auditable and retrievable form for the renewal term 8 plus whatever additional period is required in accor-j 9 dance with the current licensing basis.
10 17. A compilation of the facility's CLB. To ensure N 11 auditability and retrievability to the maximum extent .
12 possible, information composing the CLB should be 13 structured as or easily relatable to the FSAR format.
14 18. A list of documents identifying portions of the CLB Y 15 that are relevar.t to the integrated plant assessment. l 1
i l
l 1
.l l i l
I 35 k
l l
1 TABLE III !
, i 2 GENERIC FUNCTIONAL NUCLEAR POWER PLANT SSCs IMPORTANT TO LICENSE RENEWAL
- 3 The following provides o generic basis for identifying SSCs important to l 4 license renewal for both PWR and BWR nuclear power plants:
l 5 A. All components that constitute the reactor coolant pressure boundary. i 6 B. The reactor core and reactor vessel internals. ;
7- C. Systems or portions.of systems that are required for (1) emergency core 8 _ cooling, (2).postaccident containment heat removal, or (3) postaccident 9 containment' atmosphere cleanup (e.g., hydrogen removal system). !
10 D. Systems or portions of systems that are required for (1) reactor shutdown, i 11 (2) residual heat removal, or (3) cooling the spent fuel storage pool.
12 E. Those portions of the steam systems of BWRs extending from the outermost !
13 containment isolation valve to the turbine stop valve, and connected piping !
14' of 2-1/2 inches or larger nominal pipe size to and including the first !
15 valve that is either normally closed or capable of automatic closure
- 16 during all modes of normal reactor operation. j i
17 F. Those portions of the steam and feedwater systems of PWRs extending fros' ~ " ~
18 and including the secondary side of steam generators to and including the .
19- outermost containment isolation valves, and connected piping of 2-1/2 inches 20 or larger nominal pipe size to and including the first valve (including a i 21 safety or relief valve) that is either normally closed or capable of auto- !
22 matic closure during all modes of normal reactor operation. ,
t 23 G. Cooling water, component cooling, and auxiliary feedwater systems or :
24 portions of these systems, including the intake structures, that are 25 required for (1) emergency core cooling, (2) postaccident containment heat l 26 removal, (3) postaccident containment atmosphere cleanup, (4) residual heat ;
27 removal from the reactor, or (5) cooling the spent fuel storage pool. l t
28 H. Cooling water and seal water systems or portions of these systems that are l 29 required for functioning of reactor coolant system components important to l 30 safety, such as reactor coolant pumps. :
i 31 I. Systems or portions of systems that are required to supp.ly fuel for !
32 emergency equipment. l 33 J. All electric and mechanical devices and circuitry between the process and- l 34 the input terminals of the actuator systems involved in generating signals s}
35 that initiate protective action. L I
36 37 *This table provides supplemental guidance for developing plant-specific lists 38 -of SSCs important to license renewal. This guidance is derived from Regulatory 39 Guide 1.29, " Seismic Design Classification."
i l
36 :
1 TABLE III (contd) i 2 K. Systems or portions of systems that are required for (1) monitoring of 3 systems important to safety a>.d (2) actuation of systems important to 4 safety.
5 L. The spent fuel storage pool structure, including the fuel racks.
l 6 M. The reactivity control systems, e.g. , control rods, control rod drives, ~
7 and boron injection system. ,
8 N. The control room, including its associated equipment and all equipment 9 needed to maintain the control room within safe habitability limits for i 10 personnel and safe environmental limits for vital equipment.
11 0. Primary and secondary reactor containment. !
12 P. Systems, other than radioactive waste management systems, not covered by
~13 items (A) through (0) above that contain or may contain radioactive mate-14 rial and whose postulated failure would result in conservatively calculated j 15 potential offsite doses (using meteorology as recommended in Regulatory l 16 Guide 1.3, " Assumptions Used for Evaluating the Potential Radiological Con- !
17 sequences of a Loss of Coolant Accident for Boiling Water Reactors," and 18 Regulatory Guide 1.4, " Assumptions Used for Evaluating the Potential -
19 Radiological Consequences of a Loss of Coolant Accident for Pressurized 20 Water Reactors") that are more than 0.5 rem to the whole body or its 21 equivalent to any part of the body.
22 Q. Class 1E electric systems, including the auxiliary systems for the onsite 1 23 electric power supplies, that provide the emergency electric power needed 24 for functioning of plant features included in items (A) through (P) above.
25 R. Those portions of SSCs whose continued function is not required but whose 26 failure could reduce the functioning of any plant feature included in items 27 (A) through (Q) above to an unacceptable safety level or could result in 28 incapacitating injury to occupants of the control room.
l 29 S. The first seismic restraint beyond the boundaries defined in items (A) 30 through (R) above and those portions of SSCs that form interfaces between 31 Seismic Category I and non-Seismic Category I features.
! I l
1
\
37 i
i 4
l o !
l !
1 APPENDIX A !
I I
I 2 SUMARY OF AGE-RELATED DEGRADATION PROCESSES i 3 AND THEIR' MANAGEMENT IN OPERATING NUCLEAR POWER PLANTS l l !
l i
! 4 '
This appendix discusses the significant mechanisms that cause age-related 5 degradation in nuclear power plants and the principles involved in understand- !
6 -ing and mitigating this degradation. Methods for selecting systems, structures, 1 7- and components (SSCs) in which aging is a license renewal concern are also 8 described. The information that follows is of a summary nature and is not j 9 intended to characterize in detail the age-related degradation in nuclear power f 10 plants. As research continues, additional information concerning ago-related 11 degradation will be forthcoming.
12 As a plant ages, various degradation mechanisms with the potential for y, j 13 reducing SSC reliability are operative. Unmitigated, some of these processes l
14 could lead to reductions in safety levels below those implied in the plant's 15 current licensing basis.- Known aging mechanisms and criteria for understanding 16 and mitigating them are described in the following sections. Many aging mecha-17 nisms and means for mitigating age-related degradation are addressed in ongoing 18 regulatory and industry programs. For nuclear power plant license renewal, how-l 19- ever, some aspects of age-related degradation require additional attention. This 20 regulatory guide, together. with requirements stated in 10 CFR Part 54, provides 21' the guidance needed to ensure that the technical information content of a license
~
l 22 renewal application is adequate to evaluate the effectiveness of the technical 23 oversight and control applied to age-related degradation in SSCs that are impor-
, 24 tant to license renewal. This guidance relates specifically to age-related l-25 degradation concerns that should be addressed by programs for understanding and i 26 managing aging during a license' renewal ters. Because these concerns center on 27 aging mechanisms, many.of which are operative over a number of years, oversight 28 of these mechanisms must be in place before initiating a license renewal request.
l 29' This will provide the auditable and retrievable documentation of SSC performance l
30 and maintenance needed to support'a license renewal application.
y 31 The following sections provide information that relates to (1) selecting ;
32 -SSCs important to license renewal, (2)' understanding age-related degradation in A-1
. - - .-. - . ~ - - - - . - - . - . - . - . - . . . . . - - -
I i
1 structures a M components important to license renewal, and (3) managing aging l
2 in structures and components important to license renewal.
i 3 A.1 SELECTION OF SSCs IMPORTANT TO LICENSE RENEWAL l
I 4 The process for selecting SSCs important to license renewal and for acquir-l 5 ing information that needs to be included in the license renewal application is j 6' outlined in Figures 1A and 18. This process provides for selecting the SSCs for l 7 which age-related degradation should be addressed and for ensuring adequate [
8 understanding and management of age-related degradation in support of a license l 9 renewal. application. 'As described in 'the Regulatory Position, products of this 10 process represent a major part of the technical information to be. compiled in i 11 support of, or included with, an application.
12 As required by 10 CFR 54.21, acceptable implementation of the process shown l 13 in Figures 1A and 18'should' demonstrate that degradation of SSCs important to
'14 license renewal has been identified, evaluated, and accounted for in ensuring 15 that the current licensing basis, as defined in 10 CFR 54.3(a), will be main- e, 16 tained throughout the license renewal term. Consistent with requirements for i 17 continued compliance with the current licensing basis, the selection process to l 18 be'ap' plied to SSCs with known safety functions emphasizes deterministically 19 based evaluation of aging mechanisms and their effects. The license renewal 20 applicant may also use probabilistic risk assessment (PRA) techniques as a sup-21 plement to the primarily deterministic methods to add additional components to 22 the list of SSCs designated as important to license renewal.
23 The process shown in Figures 1A and IB utilizes the knowledge gained from 24 engineering design information, tests, and operating experience. Also, data 25 from in situ assessments, condition monitoring, maintenance and other records, 26 and post-service examination and tests are recommended inputs to this process.
27 A.2 ELEMENTS OF AN EFFECTIVE PROGRAM TO ADDRESS AGING DEGRADATION 28 An effective program for addressing age-related degradation will provide 29 for both understanding and managing the aging that occurs in nuclear power 30 plants. Aging mechanisms and their effects should be understood with sufficient 31 accuracy and detail to provide the basis for developing and implementing aging 32 management strategies that address, in a prioritized and timely fashion, actual 33 or potential root causes of SSC failure.
A-2 i l
l 1 A.2.1 Understanding Age-Related Degradation j 2 The aging mechanisms that occur in nuclear power plant SSCs should be ;
3
-understood if age-related degradation is to be effectively managed. The requi-4 site understanding may be either empirical or mechanistic, depending on the '
5 nature and potential consequences of a particular degradation mechanism. An
! 6 understanding of age-related degradation requires a detailed awareness of SSC 7
design, fabrication, installation, testing, inservice operation, and maintenance 1
8 cycles. All of these elements in the life cycle of SSCs involve their inter-9 i
action with stressors associated with service environments.
10 Age related degradations of SSCs are time-dependent phenomena that depend 11 on the interactions of materials and environmental and operational stressors. ,
12 Assessments of age related degradation should consider the integrated effects l 13 of these interactions, and all SSCs that are important to license renewal should !
14 be evaluated in this context. I 15 A.2.1.1 Materials -
16 Most materials used in the fabrication of SSCs are subject to some level of 17 age-related degradation. Whether this degradation can affect the operability or.
18 reliability of SSCs such that operation of the plant is reduced below acceptable 19 safety levels is an important concern. It is important to under tand how and 20 at what rate the metallic, nonmetallic, and composite materials used in plant 21 components degrade with time and how this degradation can be managed to ensure 22 the operability or reliability of SSCs. This knowledge of material behavior is 23 important in design and operations and in developing quality assurance, plant 24 inspections, condition monitoring, and maintenance programs. As more is learned 25 about the age-related behavior of materials and how to use this knowledge in the 26 design and operation of SSCs constructed from these materials, confidence will 27 grow in predictions of SSC lifetime behavior and plant operational safety.
28 A.2.1.2 Aging Stressors 29 Of the factors that can affect the age-related degradation of nuclear power 30 plant SSCs, the stressors associated with environmental and service conditions
- 31 are generally the most difficult to understand. Stressors caused by service A-3 k
i 1 conditions assume various forms (e.g., mechanical, electrical) and can originate 2 or are intensified during component fabrication, assembly, transportation, j 3 installation, operation, testing, and maintenance. Those who design, fabricate, ;
4 operate, and maintain structures and components should understand how stressors 5 can degrade their operational capabilities.
6 A.2.1.2.1 Environmental Conditions. Environmental conditions under which !
7 SSCs are designed'to function contribute individually and in concert with other l
8 . stressors.to age-related degradation. Environmental elements include ambient - !
9 operating conditions (humidity and temperature within the plant or within a l i
10 storage facility), chemicals that contact the material (pollutants, acids, 11 lubricants, etc.), and radiation. Environmental effects can individually cause i 12 degradation or influence the rate at which degradation progresses or may act in 13- combination with other factors (e.g., material type and condition, heat, and 14 stress).
15 A.2.1.2.2 Service Conditions. Service conditions consist of steady-state, "'
16 cyclic, or other transient loadings imposed on SSCs during normal operation, !
17 testing, or abnormal events. The principal loadings are mechanical in nature.
18 Significant age-related degradation can also occur because of electrical 19 loadings.
I 20 1. Mechanical loads are generally associated with physical movements, pressure ;
l l
21 differentials, and dimensional changes. The operation of SSCs either dur- ;
. 22 ing normal operation (including testing) or under accident conditions j 23 usually induces time-dependent mechanical stresses. These stresses are l 4 ;
24 caused by dynamic loads, internal or external pressure changes, impact, l 25 vibration loads, temperature changes, component test loads, and seismically {
- - 26 induced motions. The operational motions of active SSCs (e.g., valve oper-- i ation and pump rotations) produce time-dependent distortions and inertial 27
)
28 stresses as well as wear. The effects of these loads in degrading SSCs are l 29 generally understood,.but degradation rates are usually only estimates
- 30 obtained from the analysis of inservice monitoring data, inspection reports, l 31 and maintenance information. Proper maintenance can mitigate much or all 32 of the degradation caused by mechanical loads.
4 A-4 i
1 Internal and external pressure loads approaching operational or i 2 accident design limits also can produce high stresses that can cause dis-l 3 tortions and, after sufficient cycles, can result in strain hardening and 4 fatigue damage to SSC materials. If these stresses are combined with j 5 vibration and thermal stresses, measurable degradation can occur in a i 6 period that is short relative to the anticipated operational life of the
! 7 SSCs.
8 Seismic events or similar but more localized events, e.g. , water 9 hammer, can inflict immediate damage to SSCs at any point during their l
10 operational life. Even though the SSCs may not fail during the impact, i
11
~
their functional capability may be degraded such that the operational life
{
1 12 is shortened. The extent of the damage to SSCs resulting from external l 13 sources must be understood to anticipate any associated reduction in 14 lifetime.
15 Vibrational loads can cause fatigue damage. Methods of analyzing 16 vibrational fatigue damage are available; however, the results often 17 include large uncertainties. These uncertainties are associated with mate-18 rial fatigue properties and the distribution and magnitude of the induced l 19 dynamic stresses. Vibrational stresses may be induced by plant operational 20 modes, during transportation if a component is not properly isolated, and l
i 21 by ground or seismic vibrations. The source of vibrational loads that i 22 develop during the operational life of SSCs, the distribution of the asso-23 ciated stresses, and the endurance limits of the materials must be known l l 24 for lifetime prediction. I 25 Thermal stresses develop in SSCs because of temperature gradient-26 induced differences in thermal expansion and the fact that different mate- ,
27 rials expand at different rates when heated. Differential expansions may I
28 be resisted internally or by interference with adjacent component surfaces. '
29 This resistance results in time-dependent. thermal stresses that can cause 30 age-related degradation, either separately or when combined with the effects 31 of other stressors. Typical of such degradation are the thermal fatigue 32 cracks that have appeared in high-temperature coolant water piping and 33 nozzles and embrittlement of insulating materials.
34 2. Electrical stresses are induced in the insulating materials used in the 35 fabrication of electrical and electromechanical parts and components. Both
- A-5 i
1 passive SSCs (cables, connectors, electrical penetrations, transformers, 2 terminal boards, etc.) and active SSCs (motors, circuit-breakers, relays, 3 voltage and current activated devices, etc.) experience voltage gradients 4 during normal operation and testing. Of primary concern are the higher 5- levels of electrical stresses that are generated during switching operations-6 and during accident and post-accident situations. The nature of electrical ,
7 voltage loads varies depending on the design and functional application of l 8 the device. Voltage gradients can be very high and may be imposed by dc, j 9 ac, or nonperiodic fast or slow transients. The most' severe voltage j
'10 gradients are experienced when a device is subjected to various combina-l
- 11. tions of these voltages superimposed at the same. time. The magnitude and 12 duration of voltage- and current-related stresses in plant electrical -
13 structures and components should be accurately assessed during normal oper-14 ating conditions, test sequences, and accident and post-accident situations.
15 A.2.1.3 . Aging Mechanisms k
16- Stressors and environments act in concert on SSC constituent materials to .l 17 cause age-related degradation. Many mechanisms potentially can contribute to ;
18 degradation processes. Extensive analytical and experimental efforts by both l 19 government and industry have identified numerous aging mechanisms that are oper-20 ative in nuclear power plants. These mechanisms vary widely in terms of their 21 potential effects. Some mechanisms affect numerous types of SSCs over wide 22 variations in environment and stressor level; others are limited in their 23 effects to specific components or. materials over narrow ranges of conditions.
24 Aging mechanisms of concern in nuclear. power plants include the following.
25 1. Corrosion 26 Corrosion is a common form of degradation in nuclear power plants,
~27 resulting in wall thinning in steam and condensate systems, pitting in '
28 service water systems, and transport of activated corrosion products. Many 29 localized corrosion processes are operative in nuclear power plants, e.g.,
30 crevice corrosion, pitting corrosion, galvanic corrosion, various types of 31 stress-enhanced or irradiation-enhanced corrosion, and microbiological 1y 32 influenced corrosion. These processes can result in local wall thinning 33 that may lead to failure. .
I A-6 l
l
l
)
a
{ 1 0xidation to produce a surface oxide scale takes place in metals by 1
2 direct reaction with an oxidizing atmosphere. If the scale is nonporous ;
! 3 and completely covers the surface, the reaction rate will decrease as the 1
I 4 oxide thickens because the transport of reactive species through the scale j 5 becomes rate controlling. Factors such as electrical potential, concentra-i 6 tion gradients, or preferential migration paths through the film may con-l 7 trol the overall corrosion rate. The breakdown of surface scales, typi- I 8 cally through mechanical or chemical processes, often leads to a loss in N 9 protective quality of the scale.
g 10 Pitting is a localized form of corrosion'that results in small craters j- 11 or holes in the metal. Pitting is potentially one of the most insidious 12 forms of corrosion because it can lead to component failure by perforation i 13- while producing only a small loss of metal. Because of their small size f 14 and because the pits are often covered with corrosion products, they can be 15 difficult to detect. Pitting occurs when one area of a metal surface i i
16 becomes anodic with respect to the rest of the surface or when highly 1 17 localized changes in the environment in contact with the surface cause ^' '
- 18 accelerated attack. Causes of pitting include local inhomogeneities on or 19 f
[
i beneath the metal surface, local loss of passivity, mechanical or chemical j 20 rupture of the protective oxide surface film, galvanic corrosion from a j 21 relatively distant cathode, and the formation of a metal ion or oxygen con- l 22 centration cell under a solid deposit (crevice corrosion). The rate of ;
23 penetration into the metal by pitting may be 10 to 100 times greater than i j 24 for general corrosion. The most common causes of pitting in steels are l 25 surface deposits that set up local concentration cells and dissolved halides l 26 that produce local anodes by rupture of the protective surface scale. With l 27 corrosion-resistant alloys such as stainless steels, the most common cause 28 of pitting is the highly localized destruction of passivity through contact
. '29 with a halide-containing environment.
30 Uniform attack is normally characterized by a chemical or 31 electrochemical reaction that proceeds uniformly over the entire exposed 32 surface or over a large area. The metal becomes thinner and eventually 33 fails. Wall thinning of stoam generator tubes has occurred because of uni-
- 34 ' form attack by acid phosphate residues concentrated in low-flow areas.
35 Uniform attack of carbon or low-alloy steel by concentrated boric acid has 36 also been observed.
A-7 l
i 1 Intercranular attack is preferential dissolution of the grain boundary .
t 2- _ regions of a metal with only slight or negligible attack of the grain ;
3 matrix. This preferential attack can be enhanced by segregation of speci- l 4 fic elements or impurities, by enrichment of one of the alloying' elements :
5 in the grain boundaries, or by the depletion of an element that' imparts 6 -corrosion resistance to the grain boundary areas. Susceptibility to inter-7' granular attack usually develops during thermal processing such as welding ;
8 or heat treatments. The susceptibility to intergranular attack can often 9 be corrected by redistributing alloying elements more uniformly through 10 solution heat treatment, by modifying the alloy to increase resistance to 11 segregation, or by using a completely different alloy.
12- Stress corrosion crackina (SCC) is an aging mechanism that occurs in 13 engineering materials by the combined and synergistic interaction of a' q 14 chemically aggressive environment, a susceptible material, and a tensile t
15 stress or radiation field. The material fails by slow, environmentally l 16 induced crack growth that occurs with little or no attendant macroscopic !
17 plastic deformation. Although a tensile stress is not necessary for ,. . j
'18 irradiation-assisted SCC, it can aggravate the' phenomenon. The stresses !
19 required to cause SCC are usually below the yield strength and are tensile 20 in nature. These stresses can be either applied or residual and may result 21 from the fabrication process or inservice loading of the component or 22 structure. Common sources of stress include thermal processing and stress 23 risers created during surface finishing, fabrication, or assembly. The 24 length of time required to produce SCC decreases for increasing stress 25 level. The minimum stress at which cracking will occur depends on the 26 temperature, the composition and microstructure of the alloy, and the 27 ,
environnerft. SCC may initiate at pre-existing mechanical cracks or other 28 surface discontinuities such as pits produced by chemical attack.
29 Microbiological 1y influenced corrosion (MIC) occurs when biological 30 organisms affect corrosion processes on metals by directly influencing the 31 anodic and cathodic reactions, by affecting the protective surface scales 32 on metals, by producing corrosive substances, or by creating solid deposits.
33 These organisms include microscopic forms such as bacteria and macroscopic 34 types such as algae and barnacles. Microscopic and macroscopic organisms 35 have been observed to live and reproduce under broad ranges of pressure, 36 temperature, humidity, and pH; thus, biological organisms may influence l A-8 1
r 1 corrosion in a variety of environments. MIC effects on carbon steel may
.2 result in random pitting, general corrosion, or severe hydraulic effects i 3 caused by formation of tubercles and massive corrosion product deposits. !
4 MIC attack on stainless steel is characterized by pitting, most commonly I 5' at weldments.
6 Saline water attack has resulted in the degradation of reinforced i
' i
'7. concrete structures. The degradation mechanism involves water seepage into 8 the concrete thereby providing a high chloride environment to the reinforc-9 ing bars. The reinforcing bars corrode, resulting in expansion that leads 10 to cracking and spalling of the concrete. This aging mechanism is of par-11 ticular concern for Category I structures, or parts thereof, that cannot be 12 routinely inspected or examined because of submergence in water or physical 13 inaccessibility because of intense radioactivity.
14 2. Erosion
- 15 Erosion caused by high-velocity steam, water, or two phase mixtures e 16 (which may include silt or other particulates) has contributed to failures 17 of power plant equipment. Degradation. processes of importance include 18 cavitation and particulate wear. Erosion caused by cavitation involves 19 the creation of a two phase gas-liquid zone in-the vicinity of high-speed 20 rotating parts (e.g. , pump impe11ers) or in components in which steep 21 pressure gradients occur (such as throttling valves and orifices).
22 Erosion-corrosion is an accelerated form of corrosion caused by the I 23 relative motion of a corrosive fluid with respect to a metal component.
24 The corrosion process is accelerated because of erosive destruction of the .
protective oxide film, resulting in chemical attack or dissolution of the 1
25 26 underlying metal. The carbon steel secondary piping systems in nuclear 27 power plants are susceptible to erosion-corrosion. The damage morphology .
28 is usually characterized by grooves, waves, and valleys oriented in a con-29 sistent direction. Highest erosion rates tend to occur in regions where t.y ;
30 the metal is in contact with wet steam. Alloy additions to carbon steel 31 can reduce or eliminate erosion-corrosion in'most cases. Chromium is the t
, 32 most effective alloying element for improving resistance. Other elements i 33 such as copper and molybdenum also have a beneficial effect.
5 A-9 i
$ I
1 3. Embrittlement 2 Structural or chemical changes induced by radiation, elevated
.3 temperature, or atmospheric contaminants cause embrittlement of metals and 4 polymers used as electrically insulating barriers that can lead to fragility 5 and failure under dynamic loading. Metallic components are most susceptible 6 to embrittlement from neutron radiation; thus, components in proximity to 7 the reactor core are most affected. Embrittlement with loss in toughness 8 for critical components such as pressure vessels and supports represents 9 the most significant contribution of radiation to aging. Organic and elec-10 tronic materials are particularly susceptible to radiation damage from gamma 11 rays. Thermal embrittlement is associated with chemical or metallurgical 12 changes and results from such processes as thermal aging leading to reduced 13 toughness of ferrous alloys, high temperature sensitization to intergranular 14 stress corrosion cracking in austenitic stainless steels, and oxidation or 15 cross linking of polymers with a resultant loss in toughness and dielectric 16 strength. Hydrogen absorption by metallic alloys can also lead to loss of 17 toughness and brittle fracture. i 18 Neutron irradiation of metal components can result in a significant !
19 increase in yield strength with accompanying decreases in ductility and j 20 fracture toughness. Irradiation embrittlement is primarily caused by j 21 irradiation-induced precipitation of fine-scale copper precipitates and 22 formation of radiation-induced point defect clusters. These mechanisms 23 produce barriers to dislocation movement, thereby causing an increase in 24 the yield stress of the steel, a shift in the ductile-to-brittle transition 25 temperature, and a decrease in fracture energy. The major variables con-26 trolling irradiation embrittlement in reactor steels are the copper and 27 nickel content of the steel and the neutron fluence. Other factors that 28 contribute include irradiation temperature, neutron spectrum and flux, 29 phosphorus content, thermomechanical history, and concentrations of other 30 impurities and minor alloying elements.
31 Thermal embrittlement can occur in cast austenitic-ferritic (duplex) 32 stainless steel piping. The embrittlement is associated with the formation 33 of precipitates in the ferritic phase, leading to cleavage of the ferrite 34 or separation of the ferrite /austenite phase boundaries. The degree of 35 aging is related to the volume fraction of ferrite in the material. In A-10
I 1 addition, the precipitation and growth of phase-boundary carbides or 2 nitrides can lead to brittle fracture. In general, low carbon grades of-3 cast stainless steel are the most resistant, and molybdenum-containing high 4 carbon grades are the most susceptible to thermal embrittlement.
5 Hydrogen damage is an environmentally assisted degradation proces's 6
that.usually results from the combined action of hydrogen and residual or l 7 applied tensile stresses. Hydrogen damage occurs in several ways, such as 8
hydrogen embrittlement, blistering, and cracking from hydride formation.
9 Hydrogen embrittlement is usually associated with loss of tensile ductility
'10 in carbon steels and high-strength alloys and is a function of the stress 11 level and time. Steel can be embrittied by only a few parts per million 12 hydrogen, which can originate from the fabrication process or inservice 13 corrosion reactions. A similar effect may occur in austenitic stainless 14 steels, but required hydrogen levels are many times the levels for carbon 15 steels. Above about 400'F, hydrogen diffuses rapidly in steel and is 16 eliminated by offgasing. Unless the steel has an impermeable coating or 17 is used in a high hydrogen pressure environment, embrittlement should not '
18 pose a problem at these temperatures.
At low temperatures, hydrogen
! 19 embrittlement can occur in high strength, low alloy steel.
1 20 4. Mechanical Degradation Mechanisms 21 Fatigue is a common degradation process that' occurs in rotating or 22 reciprocating equipment or under other service conditions that place 23 periodic or cyclic loads on SSCs. Fatigue damage results in progressive, 24 localized structural change in materials subjected to fluctuating stresses 25 and strains. Associated failures may occur at either high or low cycles 26 in response to various kinds of loads, e.g. , mechanical or vibrational 27 loads, thermal cycles, or pressure cycles. The process of fatigue consists
'28 of three stages: (1) initial fatigue damage leading to crack initiation, 29 (2) crack propagation, and (3) sudden fracture of the remaining ligament.
30 Fatigue cracks initiate and propagate in regions of stress concentration 31 that intensify strain, e.g. , structural defects. The fatigue life of any 32 ' SSC is the number of stress or strain cycles required to cause failure.
33 This number is a function of several variables such as stress level, stress j 34 state, cydic waveform, fatigue environment, and the metallurgical condition A-11 I
i i
l 1 of the material. Stress cycles can be generated by the direct application 2- of mechanical loads, differential thermal expansion of mechanically con- ,
3 strained components, or. temperature fluctuations. Although the loading 4 conditions are different, the resultant fatigue is considered to be addi- ;
5 tive. Fatigue cracks form at the point of maximum local stress and minimum 6 local strength. The local stress pattern is governed by the geometry of
- 7. the SSC, including local features such as surface and metallurgical imper-8 fections that concentrate stress, and by the type and amplitude of the ,
loading. Surface imperfections such as scratches, mars, burrs, and other 9
-10 fabrication flaws are locations where fatigue cracks may start. Inclusions, l
'll hard precipitates, and crystal discontinuities such as grain boundaries are 12 examples of microscopic stress concentrators. Pitting corrosion, stress .
13 corrosion cracking, and other effects of a hostile environment may also be +
14 important. For example, many. fatigue failures originate in fretted areas. I 15 In many large structural components, the existence of a crack does not i i
16 necessarily imply imminent failure of the component. Significant struc-w I 17 tural life may remain before.the crack grows to a size at which failure 18 occurs. The growth of a fatigue crack under cyclic loading is principally 19 controlled by the maximum load and the ratio of maximum to minimum load.
20 Wear is a general concern for rotating or other sliding surfaces where 21 tolerances can affect performance. Lubricant loss or degradation, e.g.,
22 because of. contaminants or chemical breakdown, can greatly accelerate wear.
23 Fretting is a wear phenomenon that occurs between tight-fitting surfaces 24 that'are subjected to a cyclic, relative motion of extremely small ampli-25 tude. Fretting is frequently accompanied by corrosion. Common sites for 26 fretting are in joints that are bolted, keyed, pinned, press fit, or 27 riveted; in oscillating bearings, couplings, spindles, and seals; in press 28 fits on shafts; and in universal joints. Under fretting conditions, 29 fatigue cracks may be initiated at stresses well below the endurance limit 30 of nonfretted specimens. The initiation of fatigue cracks depends mainly 31- on surface residual stresses superimposed on applied cyclic stresses.
32 Shrinkage or creep can occur in most materials and are common phenomena 33 in plastics and in metals at high temperatures. Polymers and composites 34 used as electrical insulators, supports, and protective coatings may exhibit 35 dimensional changes caused by exposure to high temperatures, moisture, 36 mechanical stresses, or radiation. These effects can lead to deterioration )
A-12
s ,
3 !
in insulating and structural properties.
~
1 Shrinkage of concrete in nuclear !
) 2 power plants is caused mainly by long-term dehydration. Dimensional changes f
3 in concrete as it ages do not degrade the properties of concrete; however, l
) 4 when these dimensional changes cause interference (e.g., with other compo- i
- 5 nents in prestressed reinforced concrete structures), degradation can i 1 i j 6 occur. Shrinkage is the main contributor to the loss of prestressing forces l j 7 in prestressed concrete containments. ;
d I A.2.1.4 Degradation Resulting from Operational Environment f 8
9 The operational environment of a nuclear power plant has age-related j' 10 degradation implications over the plant operating history that should be properly 11 accounted for. Some SSCs were initially designed or qualified for a finite
]
4 i
! 12 lifetime (usually 40 years or less) with an associated design margin or safety i 13 factor that may, in practice, change during service. Primary piping and reactor l 14 pressure vessels are examples of components that were designed, using corrosion i 15 and fatigue allowances, to last nominally 40 years. In effect, the original ,
l 16 design or qualification provided initial values and minimum acceptable values l
} 17 for key design properties and parameters such as minimum values of wall thick- i
! 18 ness, fatigue cyclic life, dielectric strength, fracture toughness, and tensile
- 9 strength. These properties and parameters may change with time as SSCs are [
20 subjected to loadings and environmental stressors from desigr. hasis events and
{
21 also from events not included in the original design. In tre license renewal i 22 process, each licensee should return to the initial design or qualification l 23 analyses as supplied by the original equipment manufacturer (including all !
24 modifications and revisions thereto), evaluate the past service experience to l 25 determine residual values, and determine actual rates cf change for key design i 26 properties and parameters. Actual rates of change together with minimum accept-27 able values of key properties and parameters will be useful in establishing an 28- acceptable extended operating license period. An example of an event that may ;
29 not have been included in the original design but should be considered is leak- !
30 age of hot primary cooling water into low temperature piping. While such leak- l' 31' age would have been evaluated as an isolated event at the time of occurrence, 32 other related aspects of the plant operation should be evaluated to ensure that !
33 a fatigue effect does not go unevaluated. Each event with aging consequences ,
34 should be evaluated and reconciled with the original design or original quali- !
15 fication to both ensure that the design conditions were not exceeded and that I
A-13
_ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . . _ _ ~ _ _ . . _ _ _ _ _ . -
-_ _ - , _ . - I
1 transients did not contribute to limiting the expected lifetime of the as fected -
2 SSC. Normal operating, testing, and environmental stressors (including those 3' ' caused by electrical, mechanical, and thermal loadings), also contribute to
-l 4 age-related degradation that should be evaluated prior to extended life. !
5 Original equipment designers and manufacturers should be consulted for identi- !
6 fication of aging mechanisms specific to particular SSCs.
7 -A.2.1.5 Degradation Sites ,
8 Most SSCs are not uniformly susceptible to degradation. Certain sites '
9 (physical locations in or on structures or components) exhibit more deteriora- !
10 tion than others; and for many SSCs, degradation is limited to only a specific 11 . location. Factors that affect vulnerability to degradation include localized i 12 chemical or metallurgical . variations, geometry with respect to fluid flow or 13 chemical potential gradients, proximity to mechanically or chemically incompati- {
14- ble materials, and localized high stresses. Examples of. site-specific degrada-15 tion include (1) localized erosion corrosion in ferritic steel piping because of -
1 16 local high fluid velocities, (2) enhanced intergranular stress corrosion crack-17 ing in heat-affected zones near welds in austenitic stainless steels, (3) exces-18 sive hinge pin wear in check valves subject tn flutter, (4) rapid degradation 19 of pump impeller blades when cavitation occurs, (5) wear or galling of sliding '
20 contacts, (6) crevice corrosion,. and (7) fatigue cracking in regions experienc-21 ing tensile stresses. An understanding of age-related degradation requires a l 22 knowledge of which sites degrade by what mechanisms and at what rates. This 23 information is fundamental to deciding where, how, and with what frequency moni- !
24 toring should be implemented to reliably trend and mitigate degradation.
25 A.2.2 Managing Aging Degradation :
26 When the interactive effects of materials, designs, and stressors from 27 environmental and service conditions are understood, the root causes of age- i 28 related degradation can be identified and programs to ensure that SSCs will
]
29 adequately perform their intended functions can be implemented. Inspections
]
30 and surveillance to monitor degradation in SCCS important to license renewal 31 should be regularly performed. Selectively applied condition monitoring and 32 trending can also be useful in this respect. Effective management of aging A-14
+
I
l I
1 will permit timely repair, replacement, or servicing through preventive or 2 corrective maintenance.
3 Effective maintenance programs require understanding of what to maintain, 4 when to maintain, and how to maintain plant SSCs. Depending upon their intended 5
function, these programs take various forms (e.g. , inspections, surveillance, 6
tests, condition monitoring, trending, recordkeeping, predictive maintenance, 7 preventive maintenance, corrective maintenance, and reliability-centered main- ,
8 tenance). The mix of elements in an overall maintenance program should reflect 9 both the technical nature and the potential consequences of the age-related 10 degradation processes that the program is intended to mitigate.
11 From an aging management perspective, the key steps in determining when to :
12 maintain and how to maintain specific SSCs are: I 13 1. Identify monitorable indicators that account for both normal and off-normal 14 conditions that can be trended to show aging effects on the performance or 15 reliability of SSCs important to license renewal.
16 2. Develop and implement methods for monitoring the indicators identified in
(
- 17. number 1 above that will provide the total life history for the specific 18 f
SSCs.
I i
19 3. Retain information acquired by monitoring programs in auditable retrievable l
l 20 form.
l l 21 4. Trend performance measures and functional indicators for each SSC under l 22 observation and analyze the impact of rate of change; retain information 23 in auditable retrievable records.
l 24 5. Determine minimum acceptable functional capability at the end of service i 25 life for normal operation and for accident mitigation. i i
i 26 6. Develop criteria for effective surveillance, maintenance, refurbishment, 27 and replacement programs.
. 28 7. Interpret, analyze, and make decisions for maintenance or replacement.
A-15 I
i l
L 1 Both predictive and preventive maintenance programs are needed to manage 2 aging. The aging management program will provide useful input for making deci-3 sions for the full spectrum of maintenance-related activities, including quality l 4 assurance and quality control, engineering support, and plant modifications. !
5 A.2.2.1 Root-Cause Determination l
6 In order to avoid recurrences of excessive degradation, it is necessary to 7 understand the basic underlying.causes of observed deterioration, i.e., root 8 causes.- Root cause is defined as the most basic reason or collection of reasons 9 for the degradation that, if corrected, will prevent future similar deteriora-10 tion. Root causes may be associated with intrinsic SSC characteristics, such as 11 composition, metallurgical structure, or design features; or they may reflect 12 situational factors, such as departures from design envelopes, extremes in envi-13 ronmental factors or stressors, operational variables, or combinations of these 14 and other factors. An analytical program should exist to evaluate instances of ,
15 unexpected or excessive degradation in terms of their root causes. Root-cause l 16 analysis relies on the availability of accurate, sufficiently detailed, retriev- '!
17 able records to provide the facts needed to evaluate the potential engineering, 18 procedural, operational, and environmental contributors to the observed degrada-19 tion. Given this information, knowledgeable staff can generally track causes 20 and effects to successively more basic levels until the root causes are revealed.
21 When the root causes are understood, methods for preventing recurrence of 22 similar degradation will generally become evident.
23 A.2.2.2 Monitoring Aging Degradation 24 Monitoring and trending of age-related degradation are the bases for 25 predictive maintenance. The overall goal of the predictive maintenance program 26 is to provide information concerning degradation rates and residual lifetimes ,
27 that can'be used to predict and prevent failures. Tools used in doing this 28 include nondestructive examination (NDE), condition monitoring, residual life 29 ~ assessment, and information analysis and trending. Trends and defined action 30 levels provide guidance needed by the preventive maintenance program to schedule 31 services with a frequency that will avoid failure of SSCs important to license 32 renewal. Monitoring and trending of the effects of age-related degradation 33 provide opportunities for identifying and eliminating sources of unnecessary A-16
_m . _ - - -__ ._ _ _ . _ _ . _ _ . _ _ . . _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ . . _ _ _
l' degradation through root-cause analysis and corrective action. Monitoring i
-- 2 programs should provide total life histories for SSCs of concern. Approaches i t
3 to monitoring degradation include the following. .
i 4 A.2.2.2.1 Nondestructive Examination. Various nondestructive techniques k 5 are employed as part of inservice inspection and testing programs to detect and !
6 characterize flaws or other evidence of degradation that may be failure pre-7 cursors. Commonly used methods include visual inspection, dye penetrant and
- 8 magnetic particle treatments, radiography, eddy-current testing, ultrasonic l
9 testing, electrical signature analysis, and acoustic emission monitoring. Each i 10 of these methods has its advantages and limitations. The limitations derive 11 mainly from the fact that NDE techniques were developed primarily as quality i
- 12 control tools for detecting manufacturing flaws. New or improved NDE methods 13 are continuously being developed. Techniques that will provide the quantitative 14 characterizations of flaws required for fracture mechanics analysis and that 15 will allow on-line monitoring of deterioration in mechanical properties during l 16 long-term inservice exposure are expected to be available in the future. ;
I l l
17 A.2.2.2.2 Condition Monitoring. For some SSCs that are important to ,
18 license renewal, integrated monitoring programs that might involve a combination 19 of sensors and evaluation methods to ensure reliability may be in order. Condi-20 tion monitoring should be employed when justified in terms of the consequences i
21 of potential failures.
22 A.2.2.2.3 Surveillance, Testing, and Inspection Programs. Detailed and 23 comprehensive requirements for monitoring degradation in SSCs are conveyed by 24 various regulatory instruments including the Inservice Inspection (ISI) require-25 ments in Section XI of the ASME Boiler and Pressure Vessel Code and the surveil-26 lance testing requirements stated in the plant technical specifications. These
'27 oversight programs can provide useful indications of age-related degradation.
28 These programs are supplemented by nonmandatory surveillance, inspections, and ,
29 tests that reflect good engineering practices.
30 A.2.2.2.4 Residual Life Assessment. For monitored trends in age-related i
31 degradation to have meaning in terms of frequency of service, replacement, or 32 refurbishment, it is necessary to correlate the level of monitored parameters A-17
! i
1 with expected SSC residual lifetimes. These correlations are difficult to 2 establish at best; as a consequence, the technology for assessing residual life 3 is not well developed. Methods employed include surveillance specimen testing, 4 monitoring of operational parameters, evaluation of SSCs that have been in ser-5 vice, and mechanistic and empirical modeling to provide bases for predictions.
6 Improvements in the technology, accruing from more sophisticated and reliable 7 models, better archiving, development of miniature specimen testing and recon-8 stituted specimen testing techniques, as well as in situ monitoring of the 9 effects of age-related degradation, are expected to greatly increase the scope 10 and confidence of future residual lifetime assessments.
11 In summary, degradation monitoring methods, e.g., inspection, surveillance, 12 testing, and condition monitoring, should reflect mechanistic and empirical 13 assessments performed by qualified staff in their efforts to understand and 14 mitigate age-related degradation. These methods should employ state-of-the-art 15 NDE, e.g., ultrasonic testing, signature analysis, vibration analysis, dielec-16 tric performance measurements, and other measuring techniques performed by 17 qualified staff. Measurement results should be documented, trended, and anal-18 yzed with respect to implications for residual SSC lifetime and for frequency 19 and nature of preventive and corrective maintenance.
20 A.2.2.3 Mitigating Aging 21 Timely mitigation of age-related degradation through regular service, 22 repair, refurbishment, or replacement of SSCs is the prime function of the main-23 tenance program. Some or all of the monitoring activities discussed in the 24 preceding sections are generally included under the auspices of the maintenance ,
25 department. For present purposes, mitigation of aging is construed as the col-26 lection of activities that relate directly to physical maintenance of SSCs that 27 are important to license renewal. Adjustments in operating environments and- ,
28 service conditions can also serve to mitigate aging. -
29 Maintenance activities range from simple straightforward tasks to complex 30 activities that require extensive coordination, training, and technical exper-31 tise. The level of oversight and resources devoted to these activities should 32 reflect their complexity and importance to plant safety and reliability. A 33 maintenance program has many important elements. Those considered here as being l 34 particularly relevant to age-related degradation include preventive maintenance, A-18 .
=- . - . _ _ _. _ .- . - . . _ -
1 I
t l l
1 corrective maintenance, reliability-centered maintenance, and recordkeeping and 2 trending. Most of these elements have clear interfaces and interdependencies 3 with the monitoring activities discussed in the preceding sections. In addi-4 tion, the scope and nature of the various maintenance elements should reflect 5 the as-built plant specifications; manufacturer's recommendations; operating 6 experience, both internal and external; relevant recommendations and information 7 from the NRC, the nuclear power industry, and its vendors; and general good 8 engineering practices.
l 9 A.2.2.3.1 Preventive Maintenance. Preventive maintenance includes the j 10 planned and scheduled actions performed to prevent equipment failure. Preven-11 tive maintenance relies heavily on information generated by monitoring programs 12 to define necessary activities and to determine the frequency at which they I l
13 should be performed. In addition to input from monitoring programs, preventive l 14 maintenance action should be based on equipment histories, other plant perfor-15 mance experience, vendor recommendati.ns (to support life extension programs l
l 16 as well as the current licensing basis), and good engineering practice. Pre-17 ventive maintenance conducted to support license renewal should be so identified l 18 and should be comprehensive in nature. Planned actions and schedules should be i 19 documented, and departures from these plans should be justified on technical 20 grounds and subject to management review and approval. Clear, comprehensive l 21 procedures are vital for preventive maintenance and other oversight and main-22 tenance activities.
23 A.2.2.3.2 Corrective Maintenance. Corrective maintenance is performed to 24 restore failed or malfunctioning equipment to service. For some types of equip-25 ment (e.g. , items lacking severe failure consequences), a corrective rather than l 26 preventive approach is preferred. Malfunctions that represent significant ;
27 challenges to plant safety or reliability should be prevented. A major respon- !
28 sibility of the maintenance organization is to be cognizant of the significance 29 of potential malfunctions and to ensure that severe consequence events are l
30 averted by adequate preventive maintenance. As with other maintenance activi-31 ties, corrective maintenance priorities should be based on the relative impor-32 tance of the equipment and on plant safety and reliability objectives. Added ;
33 functions of corrective maintenance are to determine root causes of malfunctions 34 and carry out appropriate corrective action to prevent recurrences.
A-19 i ,
j <
I l
! i l 1 A.2.2.3.3 Reliability-Centered Maintenance. The traditional approach to l
! 2 defining maintenance program objectives and priorities is based on engineering )
3 judgment supported by vendor and industry data, maintenance and operating his- i 4 tories, and regulatory requirements and guidance. These will continue to be 1
5 essential considerations in structuring a maintenance program. They are likely l 6 to be supplemented by new approaches that quantitatively correlate priority with 7 safety significance and reliability as key factors in prioritizing maintenance 8 activities. Reliability-centered maintenance uses formalized decision logic to 9 set preventive maintenance priorities and to limit maintenance and oversight to l 10 those SSCs having low safety or economic consequences of failure. The general 11 product of applying reliability-centered priorities to preventive maintenance 12 will be:
13
- A list of SSCs whose failure or loss of function could have significant 14 safety consequences. These SSCs require scheduled preventive maintenance 15 that may have further priorities based upon risk, operating experience, and 16 expert opinion. ;
1 17
- A list of SSCs whose failure or loss of function would not be self-evident. I l 18 These SSCs should also be subject to scheduled oversight and maintenance.
19
- All other SSCs. Failure or loss of function for these will have economic i 20 consequences only.
21 Results of risk-based analyses can be used to set priorities for l 22 reliability-centered maintenance activities. These methods employ quantitative )
23 failure mode and effect analyses, e.g. , PRA, to quantitatively identify SSCs, I 24 in the context of their service and systems environments, whose malfunction I 25 could jeopardize plant. safety. In this way, SSCs can be ranked in terms of l 26 safety significance, and oversight and maintenance efforts can be commensurately 27 focused upon the most risk-significant equipment.
28 A.2.2.4 Recordkeeping and Trending 29 Recordkeeping and trending are essential elements of both monitoring and 30 maintenance programs. The sole product of monitoring programs is information.
A-20 l
l 1 In order to be useful, this information must be translated into effective main-2 tenance practices. Therefore, (1) the information obtained by monitoring 3 activities must be recorded in adequate, unambiguous detail in a form that 4 allows ready retrievability, and (2) the information must be reliably relatable 5 to specific maintenance practices that effectively address the age-related 6 degradation that is actually occurring. These records are needed to set priori-7 ties for maintenance resources and to correlate actual operating environments i
t 8 and stressors with design assumptions and computed lifetimes so that SSC life-l 9 times and maintenance intervals can be realistically anticipated.
l 10 Maintenance records serve to establish performance histories for the SSCs l 11 in the plant. This information and its continuous feedback are useful in l 12 y ecifying what, how, and when equipment should be maintained; what information 13 should be collected; and how it should be recorded. Maintenance histories and j 14 equipment performance trends should be documented and kept current. Require-15 ments for records retention and retrieval should be established to meet the ;
16 needs of other elements of programs to understand and manage age-related degra-17 dation. These requirements should be consistent with quality assurance program .,
18 requirements related to records.
19 Recordkeeping can be supplemented or requirements offset by conservative l 20 maintenance practices based on equipment history or conservative condition i 21 assessments for selected SSCs; however, detailed, usable, and retrievable 22 records of such practices and condition assessments and supplementary raw data 23 should be maintained. This is a task that, in principle, can be simplified l
24 greatly by modern computer technology, which has enhanced the technical and 25 economic feasibility of maintaining high quality records. Trending of informa-26 tion obtained by monitoring activities may be a straightforward process that 27 leads directly to maintenance recommendations. More often, however, trending 28 intended to lead to improved oversight and control requires considerable initial
. 29 development of the basic trending program and qualification of the measures to 30 be trended if results are to be meaningful. It is particularly important that 31 trending methods take into account off-normal as well as normal operating condi-l 32 tions. Records of component failure data can be trended and monitored to l 33 assess maintenance program effectiveness. Process indicators, such as post-34 maintenance test results, surveillance test results, ratio of preventive to 35 corrective maintenance, maintenance backlog, and rework frequency, should also 36 be trended to provide indications of overall maintenance effectiveness and areas 37 requiring improvement.
A-21 l
APPENDIX B REPRESENTATIVE SYSTEMS, STRUCTURES, AND COMPONENTS l POTENTIALLY IMPORTANT TO LICENSE RENEWAL *
\
I. PRESSURIZED WATER REACTORS Standard Generic Review Plan, Functional Standard Technical NUREG-0800 Tabla III Specifications A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTOOWN AND ACCIDENT MITIGATION l
l
- 1. Reactor Coolant Pressure Boundary l Reactor Vessel 5.3.3 A 3/4.2.1, 4.9.1, 4.10,
. 3/4.9.10 l l Steam Generator 5.4.1, 5.4.2.2 A 3/4.4.5, 3/4.4.6, l l 3/4.4.10 l
) Reactor Coolant Pump 5.4.1 A 3/4.4.1, 3/4.4.10 Piping 5.4.3 A 3/4.4.1.1, 3/4.4.10, Pressurizer 5.4.10 A 3/4.4.3, 3/4.4.10, 3/4.4.10 l Instrumentation 7.1 A 3/4.3 '
l Valves 5.4.12 A 3/4.4.2, 3/4.4.4 j 2. Power Operated Relief Valves, Block Valves, and Interconnected Piping Pressurizer PORV 5.4.13 A 3/4.4.4, 3/4.4.9.3, -
3/4.4.10 Pressurizer Block Valves 5.4.13 A 3/4.4.4, 3/4.4.10 l
Pressurizer Piping 5.4.3 A 3/4.4.10 l
Safety Valves 3/4.4.2, 3/4.4.10 i
- 3. Reactor Protection System J l
Detector 7.2 J 3/4.3.1, 3/4.2, 2.2.1 Signal Comparator 7.2 J 3/4.3.1, 3/4.2, 2.2.1 Logic Circuit 7.2 J 3/4.3.1, 3/4.2, 2.2.1 l
Master Relay 7.2 J 3/4.3.1, 3/4.2, 2.2.1 Slave Relay 7.2 J 3/4.3.1, 3/4.2, 2.2.1 Connecting Wire / Cable 7.2 J 3/4.3.1, 3/4.2, 2.2.1
- 4. Engineered Safety Features Actuation System K l Detector 7.3 K 3/4.3.2, 2.2.1 Signal Comparator 7.3 K 3/4.3.2, 2.2.1 Logic Circuit 7.3 K 3/4.3.2, 2.2.1 Master Relay 7.3 K 3/4.3.2, 2.2.1 Slave Relay 7.3 K 3/4.3.2, 2.2.1 Connecting Wire / Cable 7. 3 K 3/4.3.2, 2.2.1 CReferences to NUREG-0800 and the Standard Technical Specifications are directly relevant only to NPPs that were reviewed against NUREG-0800. For older NPPs, these references should be viewed as illustrative only; the licensee t,hould consult the plant-specific CLB, which includes the current FSAR, for comparable sources of information.
B-1 i
i
Stend:rd Gensric R; view Plan, Functionai Standard Technical NUREG-0800 Table III Spxcifications A. RELIED UPON FOR PRESSURE B0UNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION (contd)
- 5. Control Room and Auxiliary Shutdown 7.4 N Cable 7.4 N 3/4.3.3.5, 2.2.1 Instrumentation 7.4 N 3/4.3.3.5, 2.2.1 '
- 6. Nuclear Instrumentation J,K ,
I Source Range Detectors 7.2 J,K 3/4.3.1 Intermediate Range Detectors 7. 2 J,K 3/4.3.1 i Power Range Detectors 7.2 J,K 3/4.3.1 Connecting Cable J,K 3/4.3.1
- 7. Non-nuclear Instrumentation J,K Temperature, RCS 7.2 J,K 3/4.3.1 l Pressure, RCS 7.2, 7.3 J,K 3/4.3.2,3/4.3.1 '
-Pressurizer Level 7.2 J,K 3/4.3.1 Flow, RCS 7.2 J,K 3/4.3.1 Reactor Vessel Level 7.5 --
3/4.3.3.6 Instrumentation Sub-cooling 7.5 J,K 3/4.3.3.6
- Pressurizer Pressure 7.2, 7.3, 7.4, J,K 3/4.3.1,3/4.3.3, 3/4.3.2,3/4.3.3 Steam Generator Level 7.2, 7.4, 7.5 J,K 3/4.3.1,3/4.3.3.6 Impulse Pressure 7.7 J,K Steam Flow 7.7, 7.2 --
3/4.2 l Feedwater Flow 7.7, 7.2 --
3/4.3.3.5 (AF)
Steam Pressure 7.7, 7.2 J,K 3/4.3.3.5 Feedwater Pressure --
- 8. In-Core Instrumentation J,K Flux Detector 7. 2 --
3/4.3.3.2 Thermocouple 7.5 --
3/4.3.3.6 ;
Drive Assembly 7.2 --
3/4.3.3.2 Transfer Device 7.2 --
3/4.3.3.2 Connecting Tubing 7.2 --
Drive Cable --
3/4.3.3.2 Readout / Control Equipment --
3/4.3.3.2 -
Gas Purge System --
Leak Detection System --
- 9. Seismic Category I Piping, Raceways, Cables, Hangers, l Structures A,C l
Piping 5.2.4, 5.2.3 3/4.4.10,3/4.1.2, A,C i 3/4.4.10 Raceways 5.2.4 C Cables C Hangers 5.2.4 C 3/4.7.9 Structures C B-2
Standard G::n*"ic
' ~
R: view Plan, Fun 1a1 Standard Technical NUREG-0800 Tab. .II Specificatiens A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION i
(contd)
- 10. Auxiliary Feedwater System 7.4/10.49 G l Pumps G 3/4.4.10,3/4.7.1.2 Motor G 3/4.7.1.2 I Turbine G 3/4.7.1.2 Valves G 3/4.4.10,3/4.7.1.2-Piping G 3/4.4.10, 3/4.7.1.2 Pipe Supports G 3/4.7.9 Pipe Restraints G 3/4.7.9 Condensate Storage Tank G (not 3/4.7.1.3, 3/4.4.10 Automatic Steam Generator always) !
Overfill Protection M 3/4.3.1 Control Air '
- 11. Emergency Diesel Generators 8.3.1 Q ;
Diesel Engine Q 3/4.8.1 Alternator Q 3/4.8.1 Starting Air Compressor 9.5.6 Q 3/4.8.1 Aftercooler Q Air Dryer Q Air Receiver Q Filters Q i Valves Q l Piping Q i l
Pipe Supports Q Pipe Restraints Q 3/4.7.9 l Intake Air Filter 9.5.8 Q 3/4.8.1 ,
Silencers Q
~
Intercoolers Q Ducting Q l Turbocharger Q l
- Exhaust Air Silencer 9.5.8 Q 3/4.8.1 !
Fuel Oil Storage Tank 9.5.4 Q 3/4.8.1 Day Tank Q 3/4.8.1 Transfer Pumps 9.5.4 Q 3/4.8.1 Filters Q Strainers Q Piping Q Valves Q InjectorPumps Q 3/4.8.1 Drain Tank Q 3/4.8.1 Drain Tank Pump '
Q 3/4.8.1
( Intercooler Heat Exchanger Q 3/4.8.1 l Jacket Water Heat Exchangers Q 3/4.8.1 Jacket Water Pumps Q 3/4.8.1 Jacket Water Auxiliary Pump Q 3/4.8.1 Lube Oil Cooler Q 3/4.8.1 Valves Q 3/4.8.1 Jacket Water Heaters Q 3/4.8.1 Expansion Tank Q 3/4.8.1 4 Piping Q 3/4.8.1 Instrumentation Q 3/4.8.1 l
l B-3 l l
I
. - , _ _ _. - ~. ...
Standard Generic
, R view Plan, Functiona. Standard Technical ;
NUREG-0800 Table III Spicifications A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATIO_N :
(contd) t
(contd) :
l l Lube Oil Pumps 9.5.7 Q 3/4.8.1 Auxiliary Lube Oil Pump Q 3/4.8.1 I Motor Q 3/4.8.1 i Electric Heater Q 3/4.8.1 Filter Q 3/4.8.1 ,
Strainers Q 3/4.8.1 ;
Valves Q~ 3/4.8.1 ;
Heat Exchangers Q 3/4.8.1 i Auxiliary Tank Q 3/4.8.1 i
- Rocker Lube Oil Pump Q 3/4.8.1 Pre-Lube Pump Q 3/4.8.1 i Motor Q 3/4.8.1 Reservoir Q 3/4.8.1 Gas Ejector Q 3/4.8.1 Separator Q 3/4.8.1
, Sump Q 3/4.8.1 l Tubing Q 3/4.8.1 Instrunntation Q 3/4.8.1
- 12. Station Ratteries and Vital Power -
l Supplies) 8.3.2 Q
! Battery Q 3/4.8.2 l'
Battery Charger Q 3/4.8.2 Cable Q 3/4.8.2 Breakers Q
- 13. Electrical Distribution, Safety Related 8.3.1 Q 3/4.8.3 i All Components with Safety Q Function l 14. Containment Building 0 Containment Lines 3.8.1-3 0 3/4.6.1, 3/4.6.1.7 Shield Building 0 3/4.6.1 Primary Shield Wall 0 i Missile Shield 0 Refueling Cavity 0 Recirculation Sump 0 l Base Mat 0 3/4.6.1 !
Relief Valves 3/4.6.7 .
Tendons 0 3/4.6.1.7 Isolation Valve 0 3/4.6.4,3/4.6.1.2, 3/4.9.9 Air Locks 0 3/4.6.1.3,3/4.9.4 j 15. Containr:nt isolation System 6.2.4 0 3/4.6.1.4 Cable 0 Instrumentation 0 B-4
l Standard .aneric R: view Plan, Functicnal Standard Technical NUREG-0800 Table III_ Specifications i
A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTD0'.fd AND ACCIDENT MITIGATION Icontd) ;
- 16. Containment Spray System 6.5.2 C Containment Spray Pumps C 3/4.6.2.1,3/4.4.10 Spray Additive Tank C 3/4.6.2.2,3/4.4.10 Piping C 3/4.4.10 Nozzles C 3/4.6.2.1 Instrumentation C '
Valves C 3/4.4.10
- 17. Containment Air Cooling System 6.2.2 C Fans C 3/4.6.2.3 !
Motors C Coolers C 3/4.6.2.3 Roughing Filters C 3/4.6.1.9 HEPA Filters C 3/4.6.1.9 Dampers C Ductwork C Instrumentation C Moisture Separator C
Relief Devices C -
Charcoal Filters C 3/4.6.4, 3/4/6.1.9
- 18. Component Coolin g ater System 9.2.1 G Pumps G 3/4.7.3, 3/4.4.10 Heat Exchangers G 3/4/7.3, 3/4.4.10
- Surge Tanks G 3/4.4.10 Valves G 3/4.4.10,.3/4.7.3 Piping G 3/4.4.10, 3/4.7.3 Instrumentation G
- 19. Service Water System, Safety- 9.2.1 G Related Pumps G 3/4.7.4, 3/4.4.10 Strainers G Piping G 3/4.4.10, 3/4.7.4 Valves G 3/4.4.10, 3/4.7.4 Instrumentation -G Cooling Towers G 3/4.7.5
- 20. Emergency Core Cooling System 6. 3 C Accumulators C 3/4.5.1,3/4.4.10 Boron Injection Tank C,M 3/4.5.4.1, 3/4.4.10 Refueling Water Storage Tank C 3/4.5.5, 3/4.4.10 Intermediate Head Injection C 3/4.5.2, 3/4.4.10
, System Low Head Injection System C 3/4.5.2, 3/4.4.10 High Head Injection System C 3/4.1.2.2, 3/4.1.2.4, 3/4.5.2, 3/4.1.2.1, 3/4.4.10 B-5 t'
Sttndard Generic Review Plan, Functional Standard Technical ;
NUREG-0800 Table III Specifications i A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION (contd)
- 20. Emergency Core Cooling System (contd)
Containment Recirculation Sump C 3/4.5.2 Valves C 3/4.5.2,3/4.4.10 i Piping C 3/4.5.2,3/4.4.10 ;
- 21. Residual Heat Removal System 5.4.7 D Pumps D 3/4.5.2,3/4.4.10, ;
Heat Exchangers D 'N:$:$,3/4.4.10, :
3/4.9.8 !
Valves D 3/4.5.2,3/4.4.10, 3/4.9.8 Piping D 3/4.5.2, 3/4.4.10, 3/4.9.8 Instrumentation D 3/4.3.3.5, 3/4.9.8,
- 22. Chemical and Volume Control 9.3.4 D,M System Regenerative Heat Exchanger --
3/4.4.10 Letdown Heat Exchanger --
3/4.4.10 Ion Exchangers --
3/4.4.10 Volume Control Tank --
3/4.1.2, 3/4.4.10 Primary Water Storage Tank --
3/4.1.2,3/4.4.10 Boric Acid Tanks D,M 3/4.1.2,3/4.4.10 -
Boric Acid Batch Tank 0,M 3/4.1.2,3/4.4.10
- Boric Acid Transfer Pumps D,M 3/4.1.2,3/4.4.10 Filter D,M Blender D,M ,
Excess Letdown Heat Exchanger --
3/4.4.10 l Valves D,M 3/4.1.2,3/4.4.10 Piping D,M 3/4.1.2,3/4.4.10 Instrumentation --
Positive Displacement Pump --
3/4.4.10 :
- 23. Combustible Gas Control 6.2.5 C Post-Accident Hydrogen Venting C 3/4.6.5.3 System j Post-Accident Hydrogen Sampling C 3/4.6.5.1 i System Post-Accident Hydrogen Mixing C 3/4.6.5.4 System l Internal Hydrogen Recombiners C 3/4.6.5.2 l External Hydrogen Recombiners C 3/4.6.5.2 l
- 24. HVAC, Control Room and ESF C Purge and Exhaust System 6.4 C 3/4.7.7 Reactor Containment Fan Cooler C 3/4.6.2.3 System B-6 l
I
{ St;ndard s ric l R2 view Pitn, Functional Standtrd Technical i NUREG-0800 Table III Specifications l
l A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION I l (contd) l 24. HVAC, Control Room and ESF (contd) l Containment Activated Charcoal C Filter Units System 3/4.6.1.9, 3/4.6.4 0 '
Reactor Cavity and Excore '
Instrumentation Ventilation --
System Control Rod Drive Mechanism Ventilation System --
Manipulator Crane Ventilation --
3/4.9.12 ;
System i Pressure Vacuum and Relief System --
3/4.6.7 ,
Control Room Ventilation System N 3/4.7.7 l
- 25. Instrument Air System 9.3.1 R Compressors R l
After Cooler R Receiver R !
Dryer / Filter Train R ,
Accumulators R I Instrumentation R l
- 26. Fuel Pool Structure and 9.1.3 D l Cooling System l l
Pumps D 3/4.9.12,3/4.4.10 Heat Exchanger D 3/4.9.12, 3/4.4.10 Purification Pumps --
3/4.9.12, 3/4.4.10 Demineralizer --
3/4.4.10 Piping D 3/4.9.12,3/4.4.10 Strainers, Filters --
3/4.9.12 Valves D 3/4.9.12, 3/4.4.10 l
- 27. Fire Protection 9.5.1 R 1
Pumps --
3/4.7.11.1,3/4.7.11.2 Valves --
3/4.7.11.1,3/4.7.11.2 Piping --
3/4.7.11.1, 3/4.7.11.2 Tanks --
3/4.7.11.1,3/4.7.11.2 Instrumentation --
3/4.3.3.8 '
Halon --
3/4.7.11.4 CO 2
3/4.7.11.3.3
- 28. Ultimate Heat Sink 9.2.5 --
3/4.7.5 NA l B. FAILURE CAN AFFECT FUNCTIONING OF CATEGORY A SSC
- 1. Condensate /Feedwater System, 10.4.7, F,R Including Reheat 10.3.6 l
l Main Condenser 10.4.1 --
3/4.4.10 Condensate Pumps --
3/4.4.10 B-7
Sttndard G:neric t Review Plan, Functional Standard Technical NUREG-0800 Table III Specifications B. FAILURE CAN AFFECT FUNCTIONING OF CATEGORY A SSC (contd)
- 1. Condensate /Feedwater System, 10.4.7 Including Reheat 10.3.6 (contd)
Demineralizers --
3/4.4.10 i LP Feedwater Heaters --
3/4.4.10 Piping --
3/4.4.10 Valves --
3/4.4.10, 3/4.7.1 Main Feed Pumps --
3/4.4.10 l HP Feedwater Heaters --
3/4.4.10 Startup Feedwater System --
3/4.4.10 Heater Drain System --
3/4.4.10 Condensate Storage and Transfer --
3/4.4.10, 3/4.7.1.3 System
- 2. Turbine, Main Generator, 10.2 F,R,5 Controls 1
HP Turbine --
LP Turbines --
Valves -- i 1
Piping --
Gland Steam Condenser --
Condenser Exhausters -- '
Regulators --
l Piping --
Valves --
l Oil Pumps --
Oil Reservoir --
Oil Coolers --
Turning Gear --
Ejector --
Moisture Separator Reheater --
Main Generator --
Excitation System --
Instrumentation --
3/4.3.4
- 3. Main Steam System 10.3 A,F,R,S Steam Generator 3/4.7.2, 3/4.4.5, 3/4.4.10 Piping 3/4.4.10 Valves 3/4.4.10,3/4.7.1
- 4. Reactor Control System Control Rod Drive Mechanism 3.9.4 D,M 3/4.1.3, 3/4.3.3 Logic Cabinet Power Cabinet
, Instrumentation I
l l
B-8 l
! l t
1 l
Stand:rd ( ic R: view Plan, Fui...f onal Sttndard Technical l NUREG-0800 Table III Specifications j
B. FAILURE CAN AFFECT FUNCTIONING OF CATEGORY A SSC (contd) l
- 5. Condenser Cooling System. 10.4.5 F,R
[ l Circulating Water Pumps --
i Valves --
l Piping --
J Condenser --
1 l- 6. Instrument / Service Air .9.3.1 R l Compressors --
! After Coolers --
1 Air Receivers --
Dryer / Filter Train --
Instrumentation --
l C. OTHER SSCs IMPORTANT TO LICENSE RENEWAL r -
l 1. Reactor Post-Accident 7.5 J,K 3/4.3.3.6 l Monitoring System 1
Instrumentation l 1
l 2. Safety Parameter Display System K l
) Computer 7 --
L Instrumentation --
- 3. Waste Systems: Liquid, Gas, 11.4, 11.2 P Solid 11.3 l
- Liquid Subsystems l Solid Subsystems l Gaseous Subsystems j- 4. Fuel Handling Systems L,P 3/4.9 l New Fuel Storage Area Spent Fuel Storage Pool 3/4.4.10,3/4.9.11 Fuel Storage Building Crane Spent Fuel Bridge Crane New Fuel Elevator l- New Fuel Handling Tool -- 1 Spent Fuel Handling Tool --
Refueling Cavity
. Transfer Canal 3/4.4.10 ,
Polar Crane 3/4.9.7 Manipulator Crane 3/4.9.6 4 Red Cluster Assembly Change --
Fixture Reactor Vessel Head Lifting --
Device
. Reactor Internals Lifting Device --
Stud Tensioner --
I Refueling Tools --
l Conveyer Car Assembly i B-9 l 1
Standard Gener. ,
Review Plan, Functional Standard Technical 5 NUREG-0800 Table III Specifications !
C. OTHER SSCs IMPORTANT TO LICENSE RENEWAL (contd) !
Drive Frame Assembly i Lifting Mechanism --
Valve 3/4.3.3.1, 3/4.9.2 !
Instrumentation l Controls ;
i
- 5. Radiation and Environmental Monitoring K l
Containment Air :
Particulate Detector 11.3 --
3/4.3.3.1 !
Containment Noble Gas Monitor 11.3 --
3/4.3.3.1 Containment Purge Exhaust 11.3 --
3/4.3.3.1 l Monitor i Auxiliary Building Ventilation 11.3 --
3/4.3.3.1 '
System Monitor :
Plant Vent Stack Monitor 11.3 --
.3/4.3.3.1 Control Room Air Intake 11.3, 9.4.1 --
3/4.3.3.1 ,
Monitor Condenser Air Ejection Gas 11.3 --
3/4.3.3.1 ;
Monitor ;
Steam Generator Blowdown 11.2 --
Liquid Monitor ~
Component Cooling Water System --
Monitor Service Water Effluent Discharge --
Monitor -
Waste Disposal System Liquid --
Effluent Monitor i Gas Decay Tank Effluent Gas --
Monitor
- 6. Communications Equipment 9.5.2 K 3/4.9.5 i Telephone System --
Radio System --
Page System --
J
- 7. Intrusion Detection -- I Motion Detection System --
Sound Monitoring System --
Television System --
RF Field System --
E-Field System --
- 8. Access Control K Door Control System --
Badging /ID System --
B-10 t i j
f
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,l
'Stkndard : Gt - c- !
Review Plan, Funto.cnal Standard Technical ;
NUREG-0800 Table III Specifications i i
\
C. i
. OTHER SSCs IMPORTANT TO LICENSE RENEWAL (contd) :
i
.9. Guard Response Support !'
- i. Weapons Systems. --
j
~ Communications Systems- '
- 10. Alarm Station Operation J Instrumentation '
-- -i I !
- 11. Area Radiation Monitors J.K '3/4.3.3.1-l Area Radiation Monitoring System --
i
- 12. Radiation Survey Instruments l- Radiation Monitoring Systems --
- 13. Personnel Monitoring Devices l
Radiation Detectors 12.'/4 3 --
3/4.6
- 14. Personnel Protection Barriers
L Machinery- --
I l Structural --
l i
l l
l i
i'
)
k i I
I B-11 '
)-
L
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_. . - . -- o
- _- - -- = _.
II. BOILING WATER REACTC i Standard G:n:ric !
Review Plan, Functional Standard Technical NUREG-0800 Table III Specifications A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION ,
i
- 1. Reactor Coolant Pressure Boundary A Reactor Vessel 5.3.1 A 2.0, 3/4, 4.3 MSIVs 5.'3.3, 5.2.3 A 3/4, 4.7 i Core Spray Isolation Valves 5.3.3, 5.2.3 A '3/4, 3.2, 4.3 l Core Injector Isolation Valves 5.3.3, 5.2.3 A 3/4, 3.2, 4.3 ;
Recirculation Loops 5.3.3, 5.2.3 A 3/4, 3.2, 4.3 CRDM(s) 4.5.1 A 3/4, 3.2, 4.3 Feedwater Isolation Valves 5.3.3, 5.2.3 A 3/4, 3.2, 4.3 Head Spray Isolation Valves 5.3.3, 5.2.3 A 3/4, 3.2, 4.3
- 2. Reactor Protection System J,D l MG Sets 7.2 J,D 3/4, 3.1 4 Detectors (LPRM, APRM, etc.)~ 7.2 J,D 3/4, 3.1 l Divisions Channels 7.2 J,D 3/4, 3.1 Analog Comparator Units (ACU) 7. 2 J,D 3/4, 3.1 l, A.D. Converters 7.2 J,D 3/4, 3.1 l Optical Isolators 7.2 J,D 3/4, 3.1 ;
Logic Circuits 7.2 . J,0 3/4, 3.1 !
Solenoid Control Logic 3.9.4, 7.2 J,D 3/4, 3.1 Scram Air Operated Pilot Valves 3.9.4, 7.2 J,D 3/4, 3.1
- Back Up Solenoid Scram Values 3.9.4, 7.2 J,D 3/4, 3.1 Scram Discharge Volume Pilot ,
Valves 3.9.4, 7.2 J,D 3/4, 3.1 ;
- 3. Control Rod Drive System --
Suction Filters 4.5 H 3/4, 1.3, 1.4 Pumps 4.5 H 3/4, 1.3, 1.4 Isolation Valves 4.5.1 A 3/4, 1.3, 1.4 HCUs 3.9.4 0 3/4, 1.3, 1.4 Accumulators 3.9.4 D 3/4, 1.3, 1.4 Scram Discharge Volume 3.9.4 A,0 3/4, 1.3, 1.4 Control Rod 4.6 D 3/4, 1.3, 1.4
- 4. Standby Liquid Control System M l Storage Tank 9.3.5 M 3/4, 1.5 Pumps 9.3.5 M 3/4, 1.5 :
Squib Valves 9.3.5 M 3/4, 1.5 Neutron Absorption System 9.3.5 M 3/4, 1.5 l
- 5. Control Room and Auxiliary Shutdown D Remote S/D Panel 7.4 D 3/4, 3.7 l
- 6. Neutron Monitoring System J,K Source Range Monitor 7.1 J,K 3/4, 3.7 l Intermediate Range Monitor 7.1 J,K 3/4, 3.7 ,
LPRM/APRM 7.1 J,K 3/4, 3.7 l B-12
Sttndard Ger -
Review Plcn, Funt ,a1 Standard Technical NUREG-0800 Trble III Sp:cifications A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION (contd)
- 7. Seismic Category 1 Piping, Raceways, Cables, Hangers, and Structures to Support Dynamic Loads A Reactor Vessel, System 3.0, 3.10 A,B 3/4, 4.6 Recirculation System 3.0, 3.9, 6, A 3/4, 4.6
' 3.10 Main Steam System 3.0, 3.9 A,C 3/4,4.7 Condensate and Feedwater System 3.0, 3.7 A,G 3/4,4.4 Automatic Reactor Vessel Overfill Protection 3.0, 3.7 M 3/4.3.1 Reactor Core Isolation Cooling System 5.46 A,G 3/4,7.3 Reactor Water Cleanup System 5.4.8 A,P 3/4,4.4 (Category 1 piping and valves) :
- 8. Primary Containment --
Reactor Building Foundation 3.2.1 0,5 3/4, 6.1, 6.2, 6.3, 1 6.4, 6.5 )
!- Drywell 3.2.1 0 3/4, 6.1, 6.2, 6.3, 1 6.4, 6.5 !
l Drywell Access Penetrations 3.2.1 0,5 3/4, 6.1, 6.2, 6.3, '
6.4, 6.5 Drywell Electrical 3. 0 0,5 3/4, 6.1, 6.2, 6.3, l Penetrations 6.4, 6.5 Drywell Pipe Penetrations 3.0 0,5 3/4, 6.1, 6.2, 6.3, )
6.4, 6.5 Horizontal Vents and Weir Wall 3.0 0 3/4, 6.1, 6.2, 6.3, 6.4, 6.5 Containment 3.0 0 3/4, 6.1, 6.2, 6.3, 6.4, 6.5 Fuel Transfer Tube 9.1 0 3/4, 6.1, 6.2, 6.3, 6.4, 6.5 Suppression Pool 9.0, 3.0 0,G 3/4, 6.1, 6.2, 6.3, 6.4, 6.5 ,
l Containment Upper Pool 9.1 0 3/4, 6.1, 0.2, 6.3, l 6.4, 6.5 Primary Containment HVAC 9.4 0 3/4, 6.1, 6.2, 6.3, System 6.4, 6.5 l Primary Containment Auxiliary 9.4 0 3/4, 6.1, 6.2, 6.3, l System 6.4, 6.5 l Containment Spray
~
9.4, 6, 5.2 0,G 3/4, 6.1, 6.2, 6.3, 6.4, 6.5 l
- 9. Containment Air Cooling C '
Drywell Recirculation System 6.2.2 C 3/4,6.7 l Drywell Purge Ventilation 6.2.2 C 3/4, 6.7 System B-13 .
1
1
, Standard Ge ne'.
Review Plan, Functic.. I Standard Technical NUREG-0800 Table III Specifications l l
A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION (contd) ,
Containment Normal Ventilation t System 6.2.2 C 3/4, 6.7 j Containment High Flow Purge I System 6.2.2 C 3/4, 6.7 Containment Recirculation 6.2.2 C 3/4, 6.7 System l'
- 10. Hydrogen Control System C Containment Combustible Gas Control System 6.2.5 C 3/4,6.7 Distributed Igniter System 6.2.5 C 3/4, 6.7 Containment Atmospheric Monitoring System 6.2.5 C 3/4, 6.7
- 11. Station Batteries and Vital Power Supplies Q ,
4.16 KV Switchgear 8.3.1 Q 3/4, 8.1, 8.2, '
8.3, 8.4 l Division 1&2 Diesel Generators 9.5 Q 3/4, 8.1, 8.2, .3 8.3, 8.4 Division 3 Diesel Generators 9.5 Q 3/4, 8.1, 8.2, 8.3, 8.4 ;
480 V Switchgear 8.3.1 Q 3/4, 8.1, 8.2, '
8.3, 8.4 :
Essential AC Power Supplies 8.3.1 Q 3/4, 8.1, 8.2, '
8.3, 8.4 '
Batteries 125, 250, VDC 8.3.2 Q 3/4, 8.1, 8.2, 8.3, 8.4 j Battery Chargers 8.3.2 Q 3/4, 8.1, 8.2, )
8.3, 8.4 j Nuclear System Protection ,
Separate Divisional Power l Supplies 8.1 K 3/4, 8.1, 8.2, 8.3, 8.4 l 12. EDG (Including Air Storage, Fuel Storage and Transmission and Cooling) 'I Cooling Water System 9.5.5 C 3/4, 5.1 -
Lube Oil System 9.5.7 C 3/4,5.1
-Air Compressors 9.5.6 C,I 3/4, 5.1 Air Storage Tanks 9.5.6 C,I 3/4, 5.1 Diesel Engine 9.5.8 3/4, 8.1-8.4 Generator 9.5.8.1 3/4, 8.1-8.4 Intake & Exhaust 9.5.8 3/4, 8.1-8.4
^
Fuel Oil System 9.5.4 3/4, 8.1-8.4 Instrumentation & Control 9.5 3/4, 8.1-8.4 I
i i
B-14
{
Standard r -ic Review Plan, F. sonal Standard Technical NUREG-0800 Table III Sptcifications i
l A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION' (contd) ,
- 13. Electrical Distribution - I Safety Related Q ,
Divisional Power 1,2,3,4,5 8.1 Q 3/4, 8.1-8.4 Static Bypass Switch 8.1 Q 3/4, 8.1-8.4 Inverter 8.1 Q 3/4, 8.1-8.4 ,
- 14. Reactor Core Isolation Cooling System (Including Isolation Condenser) A,C,E,G '
Steam Isolation Valves 5.4.6 A,C,E 3/4,7.3 :
Steam Flow Elements 5.4.6 A,C,E 3/4,7.3 !
Turbine Trip Throttle Valve 5.4.6 A,C,E. 3/4,7.3 Turbine Governor Valve 5.4.6 A,C 3/4, 7.3 !
Turbine 5.4.6 A,C 3/4, 7.3 Turbine Oil System 5.4.6 C 3/4, 7.3 t Gland Seal Elements 5.4.6 G,C 3/4, 7.3 Exhaust Piping 5.4.6 G,C 3/4, 7.3
- Suction Strainer 5.4.6 G,C 3/4, 7.3
- Suction Valves 5.4.6 G,C 3/4, 7.3 ,
! Water Log Pump 5.4.6 G,C 3/4, 7.3 '
RCIC Pump 5.4.6 G,C 3/4,7.3 Auxiliary Equipment Cooling 5.4.6 G,C 3/4, 7.3 Minimum Flow Bypass Line 5.4.6 G,C 3/4, 7.3 Test Recirculator Line 5.4.6 G,C 3/4, 7.3 Testable Check Valve 5.4.6 G,C 3/4, 7.3 Flow Controller 5.4.6 G,C 3/4, 7.3
- 15. High Pressure Cooling Injection System G,C Suction Path 6.3 G,C 3/4, 5.1
- HPCS(I) Pump 6.3 G,C 3/4, 5.1
! Discharge Path 6.3 G,C 3/4,5.1 l HPCS(I) Water Leg Pump 6.3 G,C 3/4, 5.1 l Leak Detection System 6.3 G,C 3/4, 5.1 Valve Interlock 6.3 G,C 3/4,5.1
- 16. Automatic Depressu'rization System A Safety / Relief Valves 6.0 A 3/4, 5.1 Air Supply 6.0 --
3/4, 5.1 i Vacuum Breakers 6.0 --
3/4, 5.1 H
B-15
Sttndard G:nE R view Plan, Functw.41 Standard Technical l NUREG-0800 Table III Sp!cifications ,
I A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION (contd)
- 17. Core Spray System or Low Pressure Injection System --
Suction Path 5.4.7 C 3/4, 5.1 LPCS Pump 5.4.7 C 3/4, 5.1 Discharge Path 5.4.7 C 3/4, 5.1 LPCS Water Leg Pump 5.4.7 C 3/4, 5.1 ,
LPCI System 5.4.7 C 3/4, 5.1 '
- 18. Reactor Circulation System A,H Recirculation Loop Suction 5.2.3 A.H 3/4, 4.1 i Piping I Suction Isolation Valve 5.2.3 A,H 3/4,4.1 Recirculation Pumps 5.4 A,H 3/4, 4.1 Recirculation Pump Shaft Seals 5.4 A,H 3/4,4.1 Recirculation Pump Discharge Piping 5.4 A,H 3/4, 4.1 l Flow Control Valve 5.4 A,H 3/4, 4.1 l Discharge Isolation Valve 5.4 AH 3/4, 4.1 Reactor Water Sample 5.4 A.H 3/4, 4.1 Connnection Discharge Manifold and Risers 4.5.2 A.H 3/4,4.1 '
l Jet Pumps 4.5.2 A,H 3/4,4.1 Recirculation Pump Motors 5.4 A,H 3/4, 4.1 Low Frequency Motor Generator 5.4 H 3/4, 4.1 Sets
- 19. Residual Heat Removal System (Including Drywell Spray) C Suction Strainers 5.4.7 C,D,G 3/4, 4.9, 4.1 RHR Water Leg Pump 5.4.7 C,0,G 3/4, 4.9, 4.1 RHR Pumps 5.4.7 C,D,G 3/4,4.9,4.1 RHR Heat Exchangers 5.4.7 C,D,G 3/4, 4.9, 4.1 i
Motor Operated Valves 5.4.7 C,D,G 3/4, 4.9, 4.1 Testable Check Valves 5.4.7 C,D,G 3/4,4.9,4.1 Containment Spray Spargers 5.4.7 C,D,G 3/4, 4.9, 4.1 Air Operated Control Valves 5.4.7 D 3/4,4.9,4.1 Electro Pneumatic Controllers 5.4.7 D 3/4,4.9,4.1
- 20. RHR/ Shutdown Service Water System --
Ultimate Heat Sink Basin and Towers 9.2.1, 9.2.5 G 3/4,5.1,7.1 Standby Service Water Pumps 9.2.1 G 3/4, 5.1, 7.1 Heat Exchangers 9.2.5 G 3/4, 5.1, 7.1 B-16
- - - - . ._~ - -
Standard G r ~ic Review Plan, Fu- nal Standard Technical NUREG-0800 TrL ill Specifications A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION *
(contd)
- 21. Emergency Equipment Cooling C Heat Exchangers 9.2.1 C 3/4, 7.1 Closed Cooling Water System 9.2.1 C 3/4, 7.1
- 22. HVAC-Control Room and ESF N,R Supply Air Handling Units 6.4 N,R 3/4, 7.2 Recirculation Fans 6.4 N,R 3/4, 7.2 Makeup Air Cleaning Units 6.4 N,R 3/4, 7.2
- 23. Instrument Air--Important to Safety N,R Service and Instrument Air Compressors 9.3.1 N,R N/A Air Receiver 9.3.1 N,R N/A Refrigeration Air Dryers and After Filters 9.3.1 N,R -N/A Dessicant Air Dryers 9.3.1 N,R N/A Booster Instrument Air Compressors 9.3.1 N,R N/A *
- 24. Fuel Pool Structure and Cooling System D,L Skimmers Weirs and Scuppers 9.1 D,L 3/4, 9.1-9.12 FPCC Drain Tank 9.1 D,L 3/4, 9.1-9.12 FPCC Pumps 9.1. 2 D,L 3/4, 9.1-9.12 FPCC Heat Exchangers 9.1.2 D,L 3/4, 9.1-9.12 Filter /Demineralizers 9.1. 2 D,L 3/4, 9.1-9.12 Diffusers 9.1. 2 D,L 3/4, 9.1-9.12 Spent Fuel Pool 9.1. 3 D,L 3/4,9.1-9.12 Transfer Pool 9.1.3 D,L 3/4, 9.1-9.12
- 25. Fire Protection (Including Suppression) R Fresh Water Supplies 9.5.1 R 3/4, APPENDIX "R" Fire Water Supplies 9.5.1 R 3/4, APPENDIX "R" Fire Jockey Pumps 9.5.1 R 3/4, APPENDIX "R" Fire Main 9.5.1 R 3/4, APPENDIX "R" Manual Hose Station 9.5.1 R 3/4, APPENDIX "R" Preaction Type Sprinkler 9.5.1 R 3/4, APPENDIX "R" System i Deluge Type Sprinkler System 9.5.1 R 3/4, APPENDIX "R" Wet Pipe Type System 9.5.1 R 3/4, APPENDIX "R" High Pressure CO 9.5.1 R 3/4, APPENDIX "R" 2
Low Pressure CO 2
9.5.1 R 3/4, APPENDIX "R" Halon System 9.5.1 R 3/4, APPENDIX "R" Heat Detection Systems 9.5.1 R 3/4, APPENDIX "R" Smoke Detectors 9.5.1 R 3/4, APPENDIX "R" i B-17
Standard G:nst R: view Pltn, Functi, Standard Technical NUREG-0800 T:ble III' Speci fications -
i A. RELIED UPON FOR PRESSURE B0UNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION (contd)
- 25. Fire Protection (Including Suppression) (contd)
Flame Detectors 9.5.1 R 3/4, APPENDIX "R"
- 26. Ultimate Heat Sink R Circulating Water Pumps 9.2.5 R 3/4, 7.1 Cooling Towers 9.2.5 R 3/4,7.1 Main Condenser 9.2.5 R 3/4,7.1 MajorValves 9.2.5 R 3/4, 7.1 Major Piping 9.2.5 R 3/4, 7.1 B. FAILURE CAN AFFECT FUNCTIONING OF CATEGORY A SSC
- 1. Condensate /Feedwater System Including Reheat 10.4.7 R 4.4
- 2. Turbine-Generator and Controls 10.2 R,5 3/4, 3.8
- 3. Main Steam System 10.3 R,S 3/4, 4.7 l 4. Reactor Control System 7.1 0,M 3/4,4.1
- 5. Condenser Cooling System l (circulation Water System) 10.4.5 R N/A l 6. Instrument Air / Service Air, Not 5.R. 9.3.1 R N/A )
l 7. Switchyard 8. 2 R 3/4, 8.1-8.4 C. OTHER SSCs IMPORTANT TO LICENSE RENEWAL
! 1. Reactor Post-Accident Monitoring System 9.3.2 J,K 3/4,3.1-3.9 Instrumentation 9.3.2 K 3/4, 3.1-3.9
- 2. Safety Parameter Display System K 3/4,3.1-3.9 Computer 7.1 --
3/4,3.1-3.9 Instrumentation 7.1 --
3/4,3.1-3.9 i 3. Waste Systems: Liquid, Gas, 11.0 P 3/4, 11.1-11.4 Solid i Liquid Subsystems 11.0 --
3/4, 11.1-11.4 Solid Subsystems 11.0 --
3/4, 11.1-11.4 Gaseous Subsystems 11.0 --
3/4, 11.1-11.4 l
l i
B-18 ;
l I
st.xndard G2nr '- '
Rsview Plcn, Func s1 Standard Technical p NUREG-0800 Tabli. .4I Specifications A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION L (contd)
- 4. Fuel Handling Systems 9. 0 L,P 3/4, 9.1-9.12 i
New Fuel Storage Area 9.1.1, 9.1.2, --
3/4, 9.1-9.12 i 9.1.3 --
'3/4, 9.1-9.12 l Spent Fuel Storage Pool 9.1.1, 9.1.2, --
3/4, 9.1-9.12 )
l 9.1.3 --
3/4, 9.1-9.12
- Fuel Storage Bcilding Crane- 9.1.1, 9.1.2, --
3/4,- 9.1-9.12 9.1.3 3/4, 9.1-9.12 1
Spent Fuel Bridge Crane 9.1.1, 9.1.2, --
3/4, 9.1-9.12 l 9.1.3 --
3/4, 9.1-9.12
- New Fuel Elevator 9.1.1, 9.1.2, --
3/4, 9.1-9.12 9.1.3 --
3/4, 9.1-9.12 New Fuel Handling Tool 9.1.1, 9.1.2, --
3/4, 9.1-9.12 ,
9.1.3 --
3/4, 9.1-9.12 1 Spent Fuel Handling Tool 9.1.1, 9.1.2, --
3/4, 9.1-9.12 9.1. 3 --
3/4, 9.1-9.12 Refueling Cavity 9.1.1, 9.1.2, --
3/4, 9.1-9.12 9.1.3 3/4, 9.1-9.12 l
i Transfer Canal 9.1.1, 9.1.2, --
3/4, 9.1-9.12 l
! 9.1.3 --
3/4, 9.1-9.12 -
- l. Polar Crane 9.1.1, 9.1.2, --
3/4, 9.1-9.12
!' 9.1.3 --
3/4, 9.1-9.12 ^~ l l Manipulator Crane 9.1.1, 9.1.2, --
3/4, 9.1-9.12 l 9.1.3 --
3/4, 9.1-9.12
! Red Cluster Assembly Change 9.1.1, 9.1.2, --
3/4, 9.1-9.12 ,
l Fixture . 9.1.3 --
3/4, 9.1-9.12 L Reactor Vessel Head Lifting 9.1.1, 9.1.2, --
3/4, 9.1-9.12 !
! Device 9.1.3 --
3/4,9.1-9.12 l l Reactor Internals Lifting 9.1.1, 9.1.2, --
3/4, 9.1-9.12 .l Device 9.1.3 --
3/4, 9.1-9.12 Stud Tensioner 9.1.1, 9.1.2, --
3/4, 9.1-9.12 9.1.3 --
3/4,9.1-9.12 Refueling Tools 9.1.1, 9.1.2, --
3/4, 9.1-9.12 9.1.3 ' --
3/4,9.1-9.12 i Conveyer Car Assembly 9.1.1, 9.1.2, --
3/4, 9.1-9.12 l 9.1.3 --
3/4, 9.1-9.12 Drive Frame Assembly 9.1.1, 9.1.2, --
3/4, 9.1-9.12 l 9.1. 3 --
3/4, 9.1-9.12 1 l Lifting Mechanism 9.1.1, 9.1.2, --
3/4, 9.1-9.12 '
l 9.1.3 --
3/4, 9.1-9.12 )
Valve 9.1.1, 9.1.2, --
3/4, 9.1-9.12 '
9.1.3 --
3/4, 9.1-9.12
. Instrumentation 9.1.1, 9.1.2, --
3/4, 9.1-9.12 ;*
9.1. 3 --
3/4, 9.1-9.12 '- '
Controls 9.1.1, 9.1.2, --
3/4, 9.1-9.12 9.1.3 --
3/4, 9.1-9.12
! I i
B-19 ;
l . -_ - --
l
- - . - . . - .. - .. - - . - . - . - . - - . ~ .. - .~. .... - - - - . - - - . . - - -
Standard- Gsner- I i Review Plan, Functio. Standard Technical !
! NUREG-0800- Ttble III Specifications t ,
l .
I A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTDOWN AND ACCIDENT MITIGATION (contd) i
- 5. Radiation and Environmental Monitoring K l Containment Air 12,1-12.5 3/4,3.7 j j Particulate Detector 12.1-12.5 --
3/4,3.7 ;
l Containment Noble Gas Monitor 12.1-12.5 --~ 3/4, 3.7 ;
Containment Purge Exhaust 3/4, 3.7 12.1-12.5 --
Monitor Auxiliary Building Ventilation l l System Monitor 12.1-12.5 --
3/4,3.7 l Plant Vent Stack Monitor 12.1-12.5 --
3/4,3.7 ;
Control Room Air Intake 12.1-12.5 --
3/4, 3.7 i l
Monitor :
l Condenser Air Ejection Gas 12.1-12.5 --
3/4, 3.7 l l Monitor ,
Steam Generator Blowdown l Liquid Monitor. 12.1-12.5 --
3/4, 3.7 Component Cooling Water System Monitor 12.1-12.5 --
3/4,3.7 Service Water Effluent ;
Discharge Monitor 12.1-12.5 --
3/4, 3.7 l l Waste Disposal System Liquid ,
i Effluent Monitor 12.1-12.5 --
3/4,3.7 Gas Decay Tank Effluent Gas l l !
l Monitor 12.1-12.5 --
3/4,3.7 l l l
- 6. Communications Equipment 9.5.2 K N/A I l Telephone System 9.5.2 --
N/A I l Radio System 9.5.2 --
N/A 1 9.5.2 Page System --
N/A l
- 7. Intrusion Detection 13.6 --
N/A I Motion Detection System 13.6 --
N/A Sound Monitoring System 13.6 --
N/A Television System 13.6 --
N/A RF Field System 13.6 --
N/A E-Field System 13.6 --
N/A
- 8. Access Control 13.6' K N/A i
l Door Control System 13.6 --
N/A i Badging /ID System 13.6 --
N/A l
- 9. Guard Response Support l
4 Weapons Systems 13.6 --
N/A I
! Communications Systems 13.6 --
N/A i
l 10. Alare Station Operation J Instrumentation 13.6 --
N/A i
B-20 I l
Standard Gr 'c -
t R: view Plan, Fu. enal St:ndard Technical !
t NUREG-0800 Tabie III Specifications ;
A. RELIED UPON FOR PRESSURE BOUNDARY INTEGRITY, SHUTOOWN AND ACCIDENT MITIGATION '
l (contd) 11.' Area Radiation Monitors i J,K Area Radiation Monitoring 12.0 --
3/4, 3.7 I System ,
l 12. Radiation Survey Instruments '
i Radiation Monitoring Systems 12.0 --
3/4, 3.7 e
- 13. Personnel Monitorina Devices >
f Radiation Detectors 12.0 --
3/4, 3.7 ,
- 14. Personnel Protection Barriers t
Machinery 13.6 --
N/A Structural 13.6 --
N/A i
rs l
l l
3 I '~
t I
B-21 l l l
l
\
l 4
1 REGULATORY ANALYSIS l
2 A separate regulatory analysis was not prepared for this draft regulatory 3 guide. The regulatory analysis prepared for the proposed 10 CFR Part 54, ;
4
" Requirements for Renewal of Operating Licenses for Nuclear Power Plants," l 5 NUREG-1362, " Regulatory Analysis for Proposed Rule on Nuclear Power Plant 6
License Renewal," July 1990, provides the regulatory basis for this guide and 7 examines the costs and benefits of the rule as implemented by the guide. A 8 copy of the regulatory analysis is available for inspection and copying for a 9 fee at the NRC Public Document Room, 2120 L Street NW., Washington, DC. !
10 BACKFIT ANALYSIS l
11 This draft regulatory guide presents for the first time NRC staff guidance 12 on complying with a proposed new rule, 10 CFR Part 54, " Requirements for Renewal 13
{
of Operating Licenses for Nuclear Power Plants." Accordingly, publication of 14 this regulatory guide is not a backfit under 10 CFR 50.109, and no backfit 15 analysis is necessary or has been prepared for this regulatory guide. ,
l 4
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1 I UNITED STATES ,c,,,,,,,,,
l NUCLEAR REGULATORY COMMISSION PostAos e rees me
"*""C i WASHINGTON, D.C. 20555 ,
i PenuT me. oc OFFICIAL BUSINESS PENALTY FOR PRfVATE USE,6300 L
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I l I THIS DOCUMENT WAS PRINTED USING RECYCLED PAPER.
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