ML15222A852

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ENT000641 - Final License Renewal Interim Staff Guidance LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors (May 28, 2013)
ML15222A852
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 08/10/2015
From:
Entergy Nuclear Operations
To:
Atomic Safety and Licensing Board Panel
SECY RAS
References
RAS 28134, ASLBP 07-858-03-LR-BD01, 50-247-LR, 50-286-LR
Download: ML15222A852 (159)
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ENT000641 Submitted: August 10, 2015 FINAL LICENSE RENEWAL INTERIM STAFF GUIDANCE LR-ISG-2011-04 UPDATED AGING MANAGEMENT CRITERIA FOR REACTOR VESSEL INTERNAL COMPONENTS FOR PRESSURIZED WATER REACTORS INTRODUCTION This license renewal interim staff guidance (LR-ISG) updates the U.S. Nuclear Regulatory Commission (NRCs) guidance in NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, (SRP-LR) and NUREG-1801, Revision 2, Generic Aging Lessons Learned Report (GALL Report). This LR-ISG is primarily based on the issuance of Revision 1 to the Final Safety Evaluation (SE) of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines, by letter dated December 16, 2011 (SE, Revision 1, on MRP-227) (ADAMS Accession No. ML11308A770). After the issuance of the staffs SE, Revision 1, on MRP-227, EPRI Technical Report No. 1022863 (MRP-227-A) was published in January 2012. MRP-227-A is the NRC-endorsed version of MRP-227, which incorporates the NRC staffs SE, Revision 1, on MRP-227. Specifically, this LR-ISG revises the recommendations in the GALL Report and the NRC staffs acceptance criteria and review procedures in the SRP-LR to ensure consistency with MRP-227-A. This LR-ISG also provides a framework to ensure that PWR license renewal applicants will adequately address age-related degradation and aging management of reactor vessel internal (RVI) components during the term of the renewed license.

DISCUSSION Current Regulatory Framework Pursuant to Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, Section 21(a)(3), of Title 10 of the Code of Federal Regulations (10 CFR 54.21(a)(3)), a license renewal applicant is required to perform an integrated plant assessment (IPA) that demonstrates that the effects of aging on structures and components subject to an aging management review (AMR) are adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB) for the period of extended operation.

The NRCs guidance in SRP-LR Section 3.0.1 defines the AMR as the identification of the structure and component materials, environments, aging effects, and aging management programs (AMPs) credited for managing the aging effects. In turn, SRP-LR Section A.1.2.3 defines an acceptable AMP as consisting of 10 elements. In addition, 10 CFR 54.21(d) requires the license renewal application (LRA) to contain a final safety analysis report (FSAR) supplement that includes a summary description of the programs and activities for managing the effects of aging and the evaluation of time-limited aging analyses (TLAAs) for the period of extended operation.

GALL Report AMP XI.M16A, PWR Vessel Internals, provides recommendations for an AMP to manage the effects of aging for PWR RVI components. In addition, the GALL Report provides component-specific AMR items for PWR RVI components in the following tables:

  • Table IV.B2 for Westinghouse-designed RVI components
  • Table IV.B3 for CE-designed RVI components
  • Table IV.B4 for B&W-designed RVI components SRP-LR Table 3.1-1 provides the specific commodity group-based AMR items for PWR RVI components. SRP-LR Sections 3.1.2.2.1, 3.1.2.2.3, 3.1.2.2.9, 3.1.2.2.10, 3.1.2.2.12, 3.1.2.2.13, and 3.1.2.2.14 provide the aging management review results for which further evaluation is recommended by the GALL Report for PWR RVI components. Finally, SRP-LR Table 3.0-1 provides an example of the type of information to be included in the FSAR Supplement for an AMP for PWR RVI components.

Basis for Issuing Interim Guidance On January 12, 2009, EPRI submitted Technical Report No. 1016596, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 0), for NRC staff review and approval. On June 22, 2011, the NRC staff issued its SE on MRP-227, Revision 0, which contained specific topical report condition items (TRCIs) on the use of MRP-227 and Applicant/Licensee Action Items (A/LAIs) that must be addressed by those applicants or licensees utilizing this topical report. The staff issued a revision of its SE on the report methodology (i.e., SE, Revision 1, on MRP-227) by letter dated December 16, 2011.

MRP-227-A, the NRC-endorsed version of MRP-227, was later published in January 2012 and provides guidance for a PWR licensee or license renewal applicant to use in the development and implementation of an AMP for RVI components. MRP-227-A also incorporates A/LAIs that are to be addressed if this report is referenced to satisfy the requirements of 10 CFR 54.21(a)(3) for demonstrating that the effects of aging on the RVI components, within the scope of MRP-227, will be adequately managed. The staff recommends that a PWR license renewal applicant provide its responses to these A/LAIs in Appendix C of the LRA. The use of MRP-227-A by a PWR license renewal applicant is not a substitute for performing a plant-specific IPA to identify those structures and components subject to an aging management review, in accordance with 10 CFR 54.21(a)(1).

Regulatory Issue Summary (RIS) 2011-07, License Renewal Submittal Information For Pressurized Water Reactor Internals Aging Management, dated July 21, 2011, was issued, in part, to facilitate a predictable and consistent method for reviewing the aging management of RVI components for commercial PWR LRAs. An inspection plan is one aspect of the A/LAIs of the staffs SE, Revision 1, for MRP-227. This inspection plan provides information about the RVI components to be inspected and a description of how they will be managed for age-related degradation. Details of an inspection plan for those PWR plant licensees that have not submitted but plan to submit an LRA in the future will be incorporated into the LRA as part of the 10-element aging management program and aging management review line items. Thus, consistent with RIS 2011-07, these future license renewal applicants need not submit a separate document that contains an inspection plan in response to the A/LAIs of the staffs SE, Revision 1, for MRP-227.

Prior to the completion of its review and issuance of the SE on MRP-227, the staff issued SRP-LR, Revision 2, and GALL Report, Revision 2, in December 2010. Since SRP-LR, Revision 2, and GALL Report, Revision 2, were based on MRP-227, Revision 0, the relevant

portions of the SRP-LR, Revision 2, and GALL Report, Revision 2, are now being updated with this LR-ISG to reconcile any differences with MRP-227-A.

ACTION This LR-ISG updates the GALL Report, Revision 2, and SRP-LR, Revision 2, to ensure consistency with MRP-227-A for the aging management of age-related degradation for PWR RVI components during the term of a renewed operating license. Appendix A, Revisions to the GALL Report and SRP-LR, to this LR-ISG shows these changes. The majority of these changes result in the incorporation of MRP-227-A within the SRP-LR, Revision 2, and the GALL Report, Revision 2. To better show these changes, a mark-up is shown in Appendix B, Mark-Up of Changes to the GALL Report and SRP-LR, to this LR-ISG.

On March 20, 2012, at Volume 77, page 16270, of the Federal Register (77 FR 16270), the NRC requested public comments on draft LR-ISG-2011-04. Subsequently, as noticed on April 19, 2012, at Volume 77, page 23513, of the Federal Register (77 FR 23513), the NRC issued an editorial correction to the original notice to specifically identify the ADAMS Accession Nos. for additional documents associated with draft LR-ISG-2011-04.

The staff received comments on draft LR-ISG-2011-04 by letters from EPRI and the Pressurized Water Reactor Owners Group Materials Subcommittee (ADAMS Accession No. ML12146A267) and from the Nuclear Energy Institute (ADAMS Accession No. ML12144A147). The staff considered all comments, and its evaluation of these comments is contained in Appendix C, Staff Response to Public Comments on Draft License Renewal Interim Staff Guidance 2011-04, of this LR-ISG. The guidance described in this final LR-ISG supersedes the affected sections of the SRP-LR, Revision 2, and the GALL Report, Revision 2, and is approved for use by the NRC staff and stakeholders.

NEWLY IDENTIFIED SYSTEMS, STRUCTURES, AND COMPONENTS UNDER 10 CFR 54.37(b)

Any structures and components identified in this LR-ISG as requiring aging management, which were not previously identified in earlier versions of the SRP-LR, Revision 2, or GALL Report, Revision 2, are considered by the staff to be newly-identified structures and components under 10 CFR 54.37(b).

BACKFITTING AND ISSUE FINALITY This LR-ISG contains guidance on one acceptable approach for managing the associated aging effects during the PEO for components within the scope of license renewal. The staff's discussion on compliance with the requirements of the Backfit Rule, 10 CFR 50.109 is presented below.

Compliance with the Backfit Rule and Issue Finality Issuance of this LR-ISG does not constitute backfitting as defined in 10 CFR 50.109(a)(1), and the NRC staff did not prepare a backfit analysis for issuing this LR-ISG. There are several rationales for this conclusion, depending on the status of the nuclear power plant licensee.

Licensees currently in the license renewal process - The backfitting provisions in 10 CFR 50.109 do not protect an applicant, as backfitting policy considerations are not applicable to an

applicant. Therefore, issuance of this LR-ISG does not constitute backfitting as defined in 10 CFR 50.109(a)(1). There currently are no combined licenses (i.e., 10 CFR Part 52) license renewal applicants; therefore, the changes and new positions presented in the LR-ISG may be made without consideration of the issue finality provisions in 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants.

Licensees who already hold a renewed license - This guidance is nonbinding and the LR-ISG would not require current holders of renewed licenses to take any action (i.e., programmatic or plant hardware changes for managing the associated aging effects for components within the scope of this LR-ISG). Current holders of renewed licenses should treat this guidance as operating experience and take actions as appropriate to ensure that applicable aging management programs are, and will remain, effective. If, in the future, the NRC decides to take additional action and impose requirements for managing the associated aging effects for components within the scope of this LR-ISG, then the NRC would follow the requirements of the Backfit Rule.

Current operating license or combined license holders who have not yet applied for renewed licenses - The backfitting provisions in 10 CFR 50.109 do not protect any future applicant, as backfitting policy considerations are not applicable to a future applicant. Therefore, issuance of this LR-ISG does not constitute backfitting as defined in 10 CFR 50.109(a)(1). The issue finality provisions of 10 CFR Part 52 do not extend to the aging management matters covered by 10 CFR Part 54, as evidenced by the requirement in 10 CFR 52.107, Application for Renewal, stating that applications for renewal of a combined license must be in accordance with 10 CFR Part 54.

APPENDICES Appendix A provides the staffs revisions to the SRP-LR, Revision 2, and the GALL Report, Revision 2, for managing aging in PWR RVI components and includes the following sections:

  • Section 1 - Revised version of the GALL Report
  • Section 2 - Revised version of the SRP-LR Appendix B provides a mark-up of the SRP-LR, Revision 2, and GALL Report, Revision 2, to better show the changes made as a result of LR-ISG-2011-04 and includes the following sections:
  • Section 1 - Mark-up of changes to the GALL Report
  • Section 2 - Mark-up of changes to the SRP-LR Appendix C provides the staffs bases for resolving comments that were received on the draft LR-ISG-2011-04.

REFERENCES

1. U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, Title 10, Energy.
2. U.S. Code of Federal Regulations, Licenses, Certifications, and Approvals for Nuclear Power Plants, Part 52, Chapter I, Title 10, Energy.
3. U.S. Code of Federal Regulations, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, Part 54, Chapter I, Title 10, Energy.
4. U.S. Nuclear Regulatory Commission, Generic Aging Lessons Learned (GALL) Report, NUREG-1801, Revision 2, December 2010, ADAMS Accession No. ML103490041.
5. U.S. Nuclear Regulatory Commission, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, NUREG-1800, Revision 2, December 2010, ADAMS Accession No. ML103490036.
6. U.S. Nuclear Regulatory Commission, Final Safety Evaluation of EPRI Report, Materials Reliability Program Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, June 22, 2011, ADAMS Accession No.

ML111600498.

7. U.S. Nuclear Regulatory Commission, Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP)

Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, December 16, 2011, ADAMS Accession No. ML11308A770.

8. Electric Power Research Institute, EPRI Technical Report No. 1016596, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227 Revision 0), December 2008, ADAMS Accession No.

ML090160204 (Cover letter from EPRI MRP) and ADAMS Accession No. ML090160206 (Final Report).

9. Electric Power Research Institute, EPRI Technical Report No. 1022863, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), December 2011, ADAMS Accession No. ML12017A193 (Transmittal letter from the EPRI-MRP) and ADAMS Accession Nos. ML12017A194, ML12017A196, ML12017A197, ML12017A191, ML12017A192, ML12017A195 and ML12017A199 (Final Report).
10. U.S. Nuclear Regulatory Commission, Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors, Federal Register, Vol. 77, No. 54, March 20, 2012, pp. 16270-16271.
11. U.S. Nuclear Regulatory Commission, Updated Aging Management Criteria for Reactor Vessel Internal Components of Pressurized Water Reactors, Federal Register, Vol. 77, No. 76, April 19, 2012, pp. 23513.
12. T. Wells and E. Fernandez, Electric Power Research Institute Materials Reliability Program and the Pressurized Water Reactor Owners Group Materials Subcommittee, letter to Document Control Desk, U.S. Nuclear Regulatory Commission, May 21, 2012, ADAMS Accession No. ML12146A267.
13. M. Richter, Nuclear Energy Institute, letter to Cindy K. Bladey, U.S. Nuclear Regulatory Commission, May 21, 2012, ADAMS Accession No. ML12144A147.
14. U.S. Nuclear Regulatory Commission, Nuclear Regulatory Commission Regulatory Issue Summary 2011-07, License Renewal Submittal Information For Pressurized Water

Reactor Internals Aging Management, July 21, 2011, ADAMS Accession No. ML111990086.

15. U.S. Nuclear Regulatory Commission. 2008. Memorandum from Dale E. Klein, Chairman, to Hubert T. Bell, Office of the Inspector General, Response to Recommendation 8 of 9/6/07 Audit Report on NRCs License Renewal Program.

(April 1, 2008). ADAMS Accession No. ML080870286.

Appendix A REVISIONS TO THE GALL REPORT AND SRP-LR A-1

Appendix A, Section 1 - Revised version of the GALL Report (1) Revised version of GALL Report AMP XI.M16A XI.M16A PWR VESSEL INTERNALS Program Description This program relies on implementation of the Electric Power Research Institute (EPRI) Technical Report No. 1022863, Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines, (MRP-227-A) and EPRI Technical Report No. 1016609, Materials Reliability Program: Inspection Standard for PWR Internals, (MRP-228) to manage the aging effects on the pressurized water reactor (PWR) reactor vessel internal (RVI) components. The recommended activities in MRP-227-A and additional plant-specific activities not defined in MRP-227-A are implemented in accordance with Nuclear Energy Institute (NEI) 03-08, Guideline for the Management of Materials Issues. The staff approved the augmented inspection and evaluation (I&E) criteria for PWR RVI components in NRC Safety Evaluation (SE), Revision 1, on MRP-227 by letter dated December 16, 2011.

This program is used to manage the effects of age-related degradation mechanisms that are applicable in general to the PWR RVI components at the facility. These aging effects include:

(a) cracking, including stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), and cracking due to fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging or neutron irradiation embrittlement; (d) changes in dimensions due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

The program applies the guidance in MRP-227-A for inspecting, evaluating, and, if applicable, dispositioning non-conforming RVI components at the facility. These examinations provide reasonable assurance that the effects of age-related degradation mechanisms will be managed during the period of extended operation. The program includes expanding periodic examinations and other inspections, if the extent of the degradation identified exceeds the expected levels.

MRP-227-A guidance for selecting RVI components for inclusion in the inspection sample is based on a four-step ranking process. Through this process, the RVIs for all three PWR designs were assigned to one of the following four groups: Primary, Expansion, Existing Programs, and No Additional Measures. Definitions of each group are provided in Generic Aging Lessons Learned Report (GALL Report), Revision 2, Chapter IX.B.

The result of this four-step sample selection process is a set of Primary internals component locations for each of the three plant designs that are inspected because they are expected to show the leading indications of the degradation effects, with another set of Expansion internals component locations that are specified to expand the sample should the indications be more severe than anticipated.

The degradation effects in a third set of internals locations are deemed to be adequately managed by Existing Programs, such as American Society of Mechanical Engineers A-2

(ASME) Code,Section XI,11 Examination Category B-N-3, examinations of core support structures. A fourth set of internals locations are deemed to require No Additional Measures.

Evaluation and Technical Basis

1. Scope of Program: The scope of the program includes all RVI components based on the plants applicable nuclear steam supply system design. The scope of the program applies the methodology and guidance in MRP-227-A, which provides an augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants designed by Babcock & Wilcox (B&W),

Combustion Engineering (CE), and Westinghouse. The scope of components considered for inspection in MRP-227-A includes core support structures, those RVI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii). In addition, ASME Code,Section XI includes inspection requirements for PWR removable core support structures in Table IWB-2500-1, Examination Category B-N-3, which are in addition to any inspections that are implemented in accordance with MRP-227-A.

The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation. The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are managed in accordance with an applicants AMP that corresponds to GALL AMP XI.M1, ASME Code,Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.

2. Preventive Actions: MRP-227-A relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program, as described in GALL AMP XI.M2, Water Chemistry.
3. Parameters Monitored/Inspected: The program manages the following age-related degradation effects and mechanisms that are applicable in general to RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclic loading; (b) loss of material induced by wear; (c) loss of fracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimensions due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

For the management of cracking, the program monitors for evidence of surface-breaking linear discontinuities if a visual inspection technique is used as the non-destructive examination (NDE) method, or for relevant flaw presentation signals if a volumetric ultrasonic testing (UT) method is used as the NDE method. For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components. For the management of loss of preload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement. Instead, the impact of loss of 11 Refer to the GALL Report, Chapter I, for applicability of various editions of the ASME Code,Section XI.

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fracture toughness on component integrity is indirectly managed by: (1) using visual or volumetric examination techniques to monitor for cracking in the components, and (2) applying applicable reduced fracture toughness properties in the flaw evaluations, in cases where cracking is detected in the components and is extensive enough to necessitate a supplemental flaw growth or flaw tolerance evaluation. The program uses physical measurements to monitor for any dimensional changes due to void swelling or distortion.

Specifically, the program implements the parameters monitored/inspected criteria consistent with the applicable tables in Section 4, Aging Management Requirements, in MRP-227-A.

4. Detection of Aging Effects: The inspection methods are defined and established in Section 4 of MRP-227-A. Standards for implementing the inspection methods are defined and established in MRP-228. In all cases, well-established inspection methods are selected. These methods include volumetric UT examination methods for detecting flaws in bolting and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). VT-3 visual methods may be applied for the detection of cracking in non-redundant RVI components only when the flaw tolerance of the component, as evaluated for reduced fracture toughness properties, is known and the component has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. VT-3 visual methods are acceptable for the detection of cracking in redundant RVI components (e.g., redundant bolts or pins used to secure a fastened RVI assembly).

In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss of preload caused by thermal and irradiation-enhanced stress relaxation and creep.

The program adopts the guidance in MRP-227-A for defining the Expansion Criteria that need to be applied to the inspection findings of Primary components and for expanding the examinations to include additional Expansion components. RVI component inspections are performed consistent with the inspection frequency and sampling bases for Primary components, Existing Programs components, and Expansion components in MRP-227-A.

In some cases (as defined in MRP-227-A), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for changes in dimensions due to void swelling or distortion.

Inspection coverages for Primary and Expansion RVI components are implemented consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227.

5. Monitoring and Trending: The methods for monitoring, recording, evaluating, and trending the data that result from the programs inspections are given in Section 6 of MRP-227-A and its subsections. Flaw evaluation methods, including recommendations for flaw depth sizing and for crack growth determinations as well as for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications, are defined in MRP-227-A. The A-4

examination and re-examinations that are implemented in accordance with MRP-227-A, together with the criteria specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel, provide for timely detection, reporting, and implementation of corrective actions for the aging effects and mechanisms managed by the program.

The program applies applicable fracture toughness properties, including reductions for thermal aging or neutron embrittlement, in the flaw evaluations of the components in cases where cracking is detected in a RVI component and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program includes criteria to evaluate the aging effects in the inaccessible portions of the components and the resulting impact on the intended function(s) of the components. For redundant components (such as redundant bolts, screws, pins, keys, or fasteners, some of which are accessible to inspection and some of which are not accessible to inspection), the program includes criteria to evaluate the aging effects in the population of components that are inaccessible to the applicable inspection technique and the resulting impact on the intended function(s) of the assembly containing the components.

6. Acceptance Criteria: Section 5 of MRP-227-A, which includes Table 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-designed RVIs, provides the specific examination and flaw evaluation acceptance criteria for the Primary and Expansion RVI component examination methods. For RVI components addressed by examinations performed in accordance with the ASME Code,Section XI, the acceptance criteria in IWB-3500 are applicable. For RVI components covered by other Existing Programs, the acceptance criteria are described within the applicable reference document. As applicable, the program establishes acceptance criteria for any physical measurement monitoring methods that are credited for aging management of particular RVI components.
7. Corrective Actions: Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. The implementation of the guidance in MRP-227-A, plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Other alternative corrective actions bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Alternative corrective actions not approved or endorsed by the NRC will be submitted for NRC approval prior to their implementation.

8. Confirmation Process: Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the recommendations of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable. The implementation of the guidance in MRP-227-A, in conjunction with NEI 03-08 and other guidance documents, reports, or methodologies referenced in this AMP, provides an acceptable level of quality and an acceptable basis for confirming the quality of inspections, flaw evaluations, and corrective actions.

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9. Administrative Controls: The administrative controls for these types of programs, including their implementing procedures and review and approval processes, are implemented in accordance with the recommended industry guidelines and criteria in NEI 03-08, and are under existing site 10 CFR 50 Appendix B, Quality Assurance Programs, or their equivalent, as applicable. The evaluation in Section 3.5 of the NRCs SE, Revision 1, on MRP-227 provides the basis for endorsing NEI 03-08. This includes endorsement of the criteria in NEI 03-08 for notifying the NRC of any deviation from the I&E methodology in MRP-227-A and justifying the deviation no later than 45 days after its approval by a licensee executive.
10. Operating Experience: The review and assessment of relevant operating experience for its impacts on the program, including implementing procedures, are governed by NEI 03-08 and Appendix A of MRP-227-A. Consistent with MRP-227-A, the reporting of inspection results and operating experience is treated as a Needed category item under the implementation of NEI 03-08.

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry operating experience, as discussed in Appendix B of the GALL Report, which is documented in LR-ISG-2011-05.

References 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, Office of the Federal Register, National Archives and Records Administration, 2011.

10 CFR Part 50.55a, Codes and Standards, Office of the Federal Register, National Archives and Records Administration, 2011.

ASME Boiler & Pressure Vessel Code,Section V, Nondestructive Examination, 2004 Edition, American Society of Mechanical Engineers, New York, NY.

ASME Boiler & Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, The ASME Boiler and Pressure Vessel Code, 2004 edition as approved in 10 CFR 50.55a, The American Society of Mechanical Engineers, New York, NY.

EPRI Technical Report No. 1016596, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 0), Electric Power Research Institute, Palo Alto, CA: 2008.

EPRI Technical Report No.1022863, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), December 2011, ADAMS Accession No. ML12017A193 (Transmittal letter from the EPRI-MRP) and ADAMS Accession Nos. ML12017A194, ML12017A196, ML12017A197, ML12017A191, ML12017A192, ML12017A195 and ML12017A199, (Final Report).

EPRI 1016609, Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228), Electric Power Research Institute, Palo Alto, CA, July 2009 (Non-publicly available ADAMS Accession No. ML092120574). The non-proprietary version of the report may be accessed by members of the public at ADAMS Accession No. ML092750569.

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NRC Interim Staff Guidance LR-ISG-2011-05, Ongoing Review Of Operating Experience, March 16, 2012, (ADAMS Accession No. ML12044A215).

Nuclear Energy Institute (NEI) Report No. 03-08, Revision 2, Guideline for the Management of Materials Issues, ADAMS Accession No. ML101050334).

NRC Safety Evaluation from Robert A. Nelson (NRC) to Neil Wilmshurst (EPRI), Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, December 16, 2011, ADAMS Accession No. ML11308A770.

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(2) Revised version of GALL Report Chapter IV.B2 B2. REACTOR VESSEL INTERNALS (PWR) - WESTINGHOUSE Systems, Structures, and Components This section addresses the Westinghouse pressurized-water reactor (PWR) vessel internals, which consist of components in the upper internals assembly, the control rod guide tube assembly, the core barrel assembly, the baffle/former assembly, the lower internals assembly, lower support assembly, thermal shield assembly, bottom mounted instrumentation system, and alignment and interfacing components.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (IV.A2).

A-8

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-300 IV.B2-33 Alignment and Stainless steel Reactor coolant and Loss of preload Chapter XI.M16A, PWR Vessel No (R-108) interfacing neutron flux due to thermal and Internals components: internals irradiation enhanced hold down spring stress relaxation; changes in dimensions due to void swelling or distortion; loss of material due to wear IV.B2.RP-301 IV.B2-40 Alignment and Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-112) interfacing neutron flux due to stress corrosion Internals and Chapter XI.M2, components: upper cracking Water Chemistry core plate alignment pins IV.B2.RP-299 IV.B2-34 Alignment and Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No (R-115) interfacing neutron flux due to wear Internals components: upper core plate alignment pins IV.B2.RP-271 IV.B2-10 Baffle-to-former Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-125) assembly: baffle-to- neutron flux due to irradiation- Internals and Chapter XI.M2, former bolts assisted stress Water Chemistry (for SCC corrosion cracking or mechanisms only) fatigue A-9

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-272 IV.B2-6 Baffle-to-former Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-128) assembly: baffle-to- neutron flux toughness Internals former bolts due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep IV.B2.RP-270 IV.B2-1 Baffle-to-former Stainless steel Reactor coolant and Changes in Chapter XI.M16A, PWR Vessel No (R-124) assembly: baffle and neutron flux dimensions Internals former plates due to void swelling or distortion IV.B2.RP-270a IV.B2-1 Baffle-to-former Stainless steel Reactor coolant and Cracking due to Chapter XI.M16A, PWR Vessel No (R-124) assembly: baffle and neutron flux irradiation-assisted Internals and Chapter XI.M2, former plates stress corrosion Water Chemistry cracking IV.B2.RP-275 IV.B2-6 Baffle-to-former Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-128) assembly: baffle-edge neutron flux due to irradiation- Internals and Chapter XI.M2, bolts (all plants with assisted stress Water Chemistry (for SCC baffle-edge bolts) corrosion cracking or mechanisms only) fatigue A-10

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-354 Baffle-to-former Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No assembly: baffle-edge neutron flux toughness Internals bolts (all plants with due to neutron baffle-edge bolts) irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep IV.B2.RP-273 IV.B2-10 Baffle-to-former Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-125) assembly: barrel-to- neutron flux due to irradiation- Internals and Chapter XI.M2, former bolts assisted stress Water Chemistry (for SCC corrosion cracking or mechanisms only) fatigue IV.B2.RP-274 IV.B2-6 Baffle-to-former Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-128) assembly: barrel-to- neutron flux toughness Internals former bolts due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep A-11

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-284 IV.B2-12 Bottom mounted Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No (R-143) instrument system: (with or without neutron flux due to wear Internals or Chapter flux thimble tubes chrome plating) XI.M37,Flux Thimble Tube Inspection IV.B2.RP-293 IV.B2-24 Bottom-mounted Stainless steel Reactor coolant and Cracking due to fatigue Chapter XI.M16A, PWR Vessel No (R-138) instrumentation neutron flux Internals system: bottom-mounted instrumentation (BMI) column bodies IV.B2.RP-292 Bottom-mounted Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No instrumentation neutron flux toughness Internals system: bottom- due to neutron mounted instrument irradiation (BMI) column bodies embrittlement IV.B2.RP-296 Control rod guide Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No tube (CRGT) neutron flux due to wear Internals assemblies: CRGT guide plates (cards)

IV.B2.RP-298 IV.B2-28 Control rod guide Stainless steel Reactor coolant and Cracking due to stress Chapter XI.M16A, PWR Vessel No (R-118) tube (CRGT) neutron flux corrosion cracking or Internals and Chapter XI.M2, assemblies: CRGT fatigue Water Chemistry (for SCC lower flange welds mechanisms only)

IV.B2.RP-297 Control rod guide Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No tube (CRGT) (including CASS) neutron flux toughness Internals assemblies: CRGT due to thermal aging lower flange welds and neutron irradiation embrittlement and for CASS, due to thermal aging embrittlement A-12

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-355 Control rod guide Stainless steel; Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No tube (CRGT) nickel alloy neutron flux due to stress corrosion Internals and Chapter XI.M2, assemblies: guide cracking or fatigue Water Chemistry (for SCC tube support pins mechanisms only)

(split pins)

IV.B2.RP-356 Control rod guide Stainless steel; Reactor coolant and Loss of material due to Chapter XI.M16A, PWR Vessel No tube (CRGT) nickel alloy neutron flux wear Internals assemblies: guide tube support pins (split pins)

IV.B2.RP-387 Core barrel assembly: Stainless steel Reactor coolant and Cracking due to stress Chapter XI.M16A, PWR Vessel No upper core barrel and neutron flux corrosion cracking or Internals and Chapter XI.M2, lower core barrel irradiation-assisted Water Chemistry (for SCC circumferential (girth) stress corrosion mechanisms only) welds cracking or fatigue IV.B2.RP-387a Core barrel assembly: Stainless steel Reactor coolant and Cracking due to stress Chapter XI.M16A, PWR Vessel No upper core barrel and neutron flux corrosion cracking or Internals and Chapter XI.M2, lower core barrel irradiation-assisted Water Chemistry (for SCC vertical (axial) welds stress corrosion mechanisms only) cracking or fatigue IV.B2.RP-388 Core barrel assembly: Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No upper core barrel and neutron flux toughness Internals lower core barrel due to neutron circumferential (girth) irradiation welds embrittlement IV.B2.RP-388a Core barrel assembly: Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No upper core barrel and neutron flux toughness Internals lower core barrel due to neutron vertical (axial) welds irradiation embrittlement A-13

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-345 Core barrel assembly: Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No core barrel flange neutron flux due to wear Internals IV.B2.RP-278 IV.B2-8 Core barrel assembly: Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-120) core barrel outlet neutron flux due to stress Internals and Chapter XI.M2, nozzle welds corrosion Water Chemistry (for SCC cracking mechanisms only) or fatigue IV.B2.RP-278a Core barrel assembly: Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No core barrel outlet neutron flux toughness Internals nozzle welds due to neutron irradiation embrittlement IV.B2.RP-280 IV.B2-8 Core barrel assembly: Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-120) lower core barrel neutron flux due to stress corrosion Internals and Chapter XI.M2, flange weld cracking or fatigue Water Chemistry (for SCC mechanisms only)

IV.B2.RP-276 IV.B2-8 Core barrel assembly: Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-120) upper core barrel neutron flux due to stress corrosion Internals and Chapter XI.M2, flange weld cracking and Water Chemistry irradiation-assisted stress corrosion cracking IV.B2.RP-285 IV.B2-14 Lower internals Nickel alloy Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No (R-137) assembly: clevis neutron flux due to wear; loss of Internals insert bolts or screws preload due to thermal and irradiation enhanced stress relaxation or creep A-14

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-399 Lower internals Stainless steel; Reactor coolant and Cracking due to Chapter XI.M16A, PWR Vessel No assembly: clevis nickel alloy neutron flux primary water stress Internals and Chapter XI.M2, insert bolts or screws corrosion cracking, Water Chemistry (for SCC irradiation-assisted mechanisms only) stress corrosion cracking or fatigue IV.B2.RP-289 IV.B2-20 Lower internals Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-130) assembly: lower core neutron flux due to irradiation- Internals and Chapter XI.M2, plate and extra-long assisted stress Water Chemistry (for SCC (XL) lower core plate corrosion cracking or mechanisms only) fatigue IV.B2.RP-288 IV.B2-18 Lower internals Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-132) assembly: lower core neutron flux toughness Internals plate and extra-long due to neutron (XL) lower core plate irradiation embrittlement; loss of material due to wear IV.B2.RP-291 IV.B2-24 Lower support Cast austenitic Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-138) assembly: lower stainless steel neutron flux due to irradiation- Internals and Chapter XI.M2, support column assisted stress Water Chemistry bodies (cast) corrosion cracking IV.B2.RP-290 IV.B2-21 Lower support Cast austenitic Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-140) assembly: lower stainless steel neutron flux toughness Internals support column due to thermal aging bodies (cast) and neutron irradiation embrittlement A-15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-291a Lower support Stainless steel Reactor coolant and Cracking due to stress Chapter XI.M16A, PWR Vessel No assembly: lower neutron flux corrosion cracking or Internals and Chapter XI.M2, support forging or fatigue Water Chemistry (for SCC casting mechanisms only)

IV.B2.RP-290a Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No assembly: lower neutron flux toughness due to Internals support forging or neutron irradiation casting embrittlement (and thermal aging embrittlement for CASS, PH SS, and martensitic SS)

IV.B2.RP-294 IV.B2-24 Lower support Stainless steel Reactor coolant and Cracking due to Chapter XI.M16A, PWR Vessel No (R-138) assembly: lower neutron flux irradiation-assisted Internals and Chapter XI.M2, support column stress corrosion Water Chemistry bodies (non-cast) cracking IV.B2.RP-295 Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No assembly: lower neutron flux toughness due to Internals support column neutron irradiation bodies (non-cast) embrittlement IV.B2.RP-286 IV.B2-16 Lower support Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No (R-133) assembly: lower neutron flux due to irradiation- Internals and Chapter XI.M2, support column bolts assisted stress Water Chemistry (for SCC corrosion cracking or mechanisms only) fatigue A-16

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-287 IV.B2-17 Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-135) assembly: lower neutron flux toughness Internals support column bolts due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep IV.B2.RP-303 IV.B2-31 Reactor vessel Stainless steel; Reactor coolant and Cumulative fatigue Fatigue is a time-limited aging Yes - TLAA (R-53) internal components nickel alloy neutron flux damage due to fatigue analysis (TLAA) to be evaluated for the period of extended operation. See the SRP, Section 4.3 Metal Fatigue, for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

IV.B2.RP-24 IV.B2-32 Reactor vessel Stainless steel; Reactor coolant and Loss of material Chapter XI.M2, Water No (RP-24) internal components nickel alloy neutron flux due to pitting and Chemistry crevice corrosion A-17

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-382 IV.B2-26 Reactor vessel Stainless steel; Reactor coolant and Cracking Chapter XI.M1, ASME Section No (R-142) internals: ASME nickel alloy neutron flux due to fatigue, stress XI Inservice Inspection,Section XI, corrosion cracking, or Subsections IWB, IWC, and Examination Category irradiation-assisted IWD or Chapter XI.M16A, B-N-3 core support stress corrosion PWR Vessel Internals, by structure components cracking; invoking applicable (not already identified loss of material 10 CFR 50.55a and ASME as Existing due to wear Section XI inservice inspection Programs requirements components in MRP-227-A)

IV.B2.RP-302 Thermal shield Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No assembly: thermal neutron flux due to fatigue Internals shield flexures IV.B2.RP-302a Thermal shield Stainless steel Reactor coolant and Loss of material due to Chapter XI.M16A, PWR Vessel No assembly: thermal neutron flux wear Internals shield flexures IV.B2.RP-265 Reactor internal No Stainless steel; Reactor coolant and No additional aging Chapter XI.M16A, PWR Vessel No Additional Measures nickel alloy neutron flux management for Internals components reactor internal No Additional Measures components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists IV.B2.RP-291b Upper Internals Stainless steel Reactor coolant and Cracking due to fatigue Chapter XI.M16A, PWR Vessel No Assembly; upper core neutron flux Internals plate A-18

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Further Item Link Material Environment Component Mechanism Program (AMP) Evaluation IV.B2.RP-290b Upper Internals Stainless steel Reactor coolant and Loss of material due to Chapter XI.M16A, PWR Vessel No Assembly; upper core neutron flux wear Internals plate IV.B2.RP-346 Upper Internals Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No Assembly: upper neutron flux due to stress corrosion Internals and Chapter XI.M2, support ring or skirt cracking or fatigue Water Chemistry (for SCC mechanisms only)

A-19

(3) Revised version of GALL Report Chapter IV.B3 B3. REACTOR VESSEL INTERNALS (PWR) - COMBUSTION ENGINEERING Systems, Structures, and Components This section addresses the Combustion Engineering (CE) pressurized-water reactor (PWR) vessel internals, which consist of components in the upper internals assembly, the control element assembly (CEA), the core support barrel assembly, the core shroud assembly, and the lower support structure assembly, and encore instrumentation (ICI) components.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (IV.A2).

A-20

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-312 IV.B3-2 Control Element Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-149) Assembly (CEA): steel and neutron flux due to stress Internals and Chapter XI.M2, Water instrument guide corrosion cracking Chemistry (for SCC mechanisms only) tubes in peripheral or fatigue CEA assemblies IV.B3.RP-313 Control Element Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No Assembly (CEA): steel and neutron flux due to stress Internals and Chapter XI.M2, Water remaining corrosion cracking Chemistry (for SCC mechanisms only) instrument guide or fatigue tubes in CEA assemblies IV.B3.RP-320 IV.B3-9 Core shroud Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-162) assemblies (all steel and neutron flux due to fatigue Internals plants): guide lugs; guide lug inserts and bolts IV.B3.RP-319 IV.B3-9 Core shroud Stainless Reactor coolant Loss of material Chapter XI.M16A, "PWR Vessel No (R-162) assemblies (all steel and neutron flux due to wear; loss of Internals plants): guide lugs; preload due to guide lug inserts thermal and and bolts irradiation enhanced stress relaxation or creep A-21

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-358 Core shroud Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No assemblies (for steel and neutron flux due to irradiation- Internals and Chapter XI.M2, Water bolted core shroud assisted stress Chemistry assemblies): corrosion cracking assembly components, including shroud plates and former plates IV.B3.RP-318 IV.B3-8 Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No (R-163) assemblies (for steel and neutron flux toughness Internals" bolted core shroud due to neutron assemblies): irradiation assembly embrittlement; components, changes in including shroud dimensions plates and former due to void swelling plates or distortion IV.B3.RP-316 IV.B3-9 Core shroud Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-162) assemblies (for steel and neutron flux due to irradiation- Internals and Chapter XI.M2, Water bolted core shroud assisted stress Chemistry (for SCC mechanisms only) assemblies): corrosion cracking barrel-shroud bolts or fatigue A-22

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-317 IV.B3-7 Core shroud Stainless Reactor coolant Loss of preload Chapter XI.M16A, "PWR Vessel No (R-165) assemblies (for steel and neutron flux due to thermal and Internals" bolted core shroud irradiation enhanced assemblies): stress relaxation or barrel-shroud bolts creep; loss of fracture toughness due to neutron irradiation embrittlement IV.B3.RP-314 IV.B3-9 Core shroud Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-162) assemblies (for steel and neutron flux due to irradiation- Internals and Chapter XI.M2, Water bolted core shroud assisted stress Chemistry (for SCC mechanisms only) assemblies): core corrosion cracking shroud bolts or fatigue IV.B3.RP-315 IV.B3-7 Core shroud Stainless Reactor coolant Loss of preload Chapter XI.M16A, PWR Vessel No (R-165) assemblies (for steel and neutron flux due to thermal and Internals bolted core shroud irradiation enhanced assemblies): core stress relaxation or shroud bolts creep; loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion A-23

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-359 Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel No assembly (designs steel and neutron flux toughness Internals assembled in due to neutron two vertical irradiation sections): core embrittlement; shroud plate-to- changes in former plate welds dimensions due to void swelling or distortion IV.B3.RP-322 Core shroud Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No assembly (designs steel and neutron flux due to irradiation- Internals and Chapter XI.M2, Water assembled in two assisted stress Chemistry vertical sections): corrosion cracking core shroud plate-to-former plate welds IV.B3.RP-326 Core shroud Stainless Reactor coolant Changes in Chapter XI.M16A, "PWR Vessel No assembly (designs steel and neutron flux dimensions Internals" assembled in two due to void swelling vertical or distortion; loss of sections): fracture toughness assembly due to neutron components, irradiation including embrittlement monitoring of the gap opening at the core shroud re-entrant corners A-24

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-326a Core shroud Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel No assembly (designs steel and neutron flux stress corrosion Internals and Chapter XI.M2, Water assembled in cracking or fatigue Chemistry (for SCC mechanisms only) two vertical sections):

assembly components, including monitoring of the gap opening at the core shroud re-entrant corners IV.B3.RP-323 Core shroud Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No assembly (designs steel and neutron flux due to irradiation- Internals and Chapter XI.M2, Water assembled in assisted stress Chemistry two vertical corrosion cracking sections):

remaining axial welds IV.B3.RP-359a Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel No assembly (designs steel and neutron flux toughness Internals assembled in due to neutron two vertical irradiation sections): embrittlement; remaining axial changes in welds dimensions due to void swelling or distortion A-25

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-324 Core shroud Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No assembly (designs steel and neutron flux due to irradiation- Internals and Chapter XI.M2, Water assembled assisted stress Chemistry with full-height corrosion cracking shroud plates):

shroud plate axial weld seams at the core shroud re-entrant corners IV.B3.RP-360 Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No assembly (designs steel and neutron flux toughness Internals" assembled due to neutron with full-height irradiation shroud plates): embrittlement shroud plate axial weld seams at the core shroud re-entrant corners IV.B3.RP-325 Core shroud Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No assembly (designs steel and neutron flux due to irradiation- Internals and Chapter XI.M2, Water assembled assisted stress Chemistry with full-height corrosion cracking shroud plates):

remaining axial welds, ribs, and rings A-26

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-361 Core shroud Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No assembly (designs steel and neutron flux toughness Internals" assembled due to neutron with full-height irradiation shroud plates): embrittlement remaining axial welds, ribs, and rings IV.B3.RP-362 Core support barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No assembly: lower steel and neutron flux toughness Internals" cylinder due to neutron circumferential irradiation (girth) welds embrittlement IV.B3.RP-362a Core support barrel Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel No assembly: lower steel and neutron flux stress corrosion Internals and Chapter XI.M2, Water cylinder cracking or Chemistry circumferential irradiation-assisted (girth) welds stress corrosion cracking IV.B3.RP-362b Core support barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No assembly: lower steel and neutron flux toughness Internals" cylinder vertical due to neutron (axial) welds irradiation embrittlement A-27

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-362c Core support barrel Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel No assembly: lower steel and neutron flux stress corrosion Internals and Chapter XI.M2, Water cylinder vertical cracking or Chemistry (axial) welds irradiation-assisted stress corrosion cracking IV.B3.RP-329 IV.B3-15 Core support barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-155) assembly: upper steel and neutron flux due to stress Internals and Chapter XI.M2, Water cylinder (base corrosion cracking Chemistry metal and welds) and upper core barrel flange (flange base metal)

IV.B3.RP-333 Core support barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No assembly: lower steel and neutron flux due to stress Internals and Chapter XI.M2, Water flange corrosion cracking Chemistry (for SCC mechanisms only) or fatigue IV.B3.RP-328 IV.B3-15 Core support barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-155) assembly: lower steel and neutron flux due to stress Internals and Chapter XI.M2, Water core barrel flange corrosion cracking Chemistry (for SCC mechanisms only) weld or fatigue IV.B3.RP-332 IV.B3-17 Core support barrel Stainless Reactor coolant Loss of material Chapter XI.M16A, PWR Vessel No (R-156) assembly: upper steel and neutron flux due to wear Internals core barrel flange IV.B3.RP-327 IV.B3-15 Core support barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-155) assembly: upper steel and neutron flux due to stress Internals and Chapter XI.M2, Water core barrel flange corrosion cracking Chemistry weld A-28

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-357 Incore instruments Zircaloy-4 Reactor coolant Loss of material Chapter XI.M16A, PWR Vessel No (ICI): ICI thimble and neutron flux due to wear Internals tubes - lower IV.B3.RP-336 IV.B3-22 Lower support Stainless Reactor coolant Loss of material Chapter XI.M16A, PWR Vessel No (R-170) structure (designs steel and neutron flux due to wear; Internals assembled in two loss of fracture vertical sections): toughness fuel alignment pins due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep IV.B3.RP-334 IV.B3-23 Lower support Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-167) structure (designs steel and neutron flux due to stress Internals and Chapter XI.M2, Water assembled in two corrosion cracking, Chemistry (for SCC mechanisms only) vertical sections or irradiation-assisted with full-height stress corrosion shroud plates): cracking, or fatigue fuel alignment pins A-29

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-334a IV.B3-22 Lower support Stainless Reactor coolant Loss of material Chapter XI.M16A, PWR Vessel No (R-170) structure (designs steel and neutron flux due to wear; Internals assembled in two loss of fracture vertical sections or toughness with full-height due to neutron shroud plates): irradiation fuel alignment pins embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep IV.B3.RP-364 Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No structure (all steel and neutron flux toughness Internals" plants): core (including due to neutron support column CASS) irradiation welds embrittlement and for column welds made from CASS, thermal embrittlement IV.B3.RP-363 Lower support Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel No structure (all steel and neutron flux stress corrosion Internals and Chapter XI.M2, Water plants): core cracking, irradiation- Chemistry (for SCC mechanisms only) support column assisted stress welds corrosion cracking, or fatigue IV.B3.RP-330 IV.B3-23 Lower support Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-167) structure: core steel and neutron flux due to irradiation- Internals and Chapter XI.M2, Water support column assisted stress Chemistry (for SCC mechanisms only) bolts corrosion cracking or fatigue A-30

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-331 Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel No structure: core steel and neutron flux toughness Internals support column due to neutron bolts irradiation embrittlement IV.B3.RP-335 IV.B3-23 Lower support Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No (R-167) structure (designs steel and neutron flux due to stress Internals and Chapter XI.M2, Water except those corrosion cracking Chemistry (for SCC mechanisms only) assembled with or fatigue full-height shroud plates):

lower core support beams IV.B3.RP-365 Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel No structure (designs steel and neutron flux toughness Internals" with a core support due to neutron plate): core support irradiation plate embrittlement IV.B3.RP-343 Lower support Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No structure (designs steel and neutron flux due to fatigue Internals with a core support plate): core support plate A-31

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-342 Lower support Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No structure (designs steel and neutron flux due to stress Internals and Chapter XI.M2, Water with core shrouds corrosion cracking, Chemistry (for SCC mechanisms only) assembled with full irradiation-assisted height shroud stress corrosion plates): deep cracking, or fatigue beams IV.B3.RP-366 Lower support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel No structure (designs steel and neutron flux toughness Internals with core shrouds due to neutron assembled with full irradiation height shroud embrittlement plates): deep beams IV.B3.RP-339 IV.B3-24 Reactor vessel Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, TLAA (R-53) internal steel; nickel and neutron flux damage (TLAA) to be evaluated for the period of components alloy due to fatigue extended operation. See the SRP, Section 4.3 Metal Fatigue, for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

IV.B3.RP-306 Reactor internal Stainless Reactor coolant No additional aging Chapter XI.M16A, PWR Vessel No No Additional steel; nickel and neutron flux management for Internals Measures alloy reactor internal No components Additional Measures components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists A-32

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure Aging Effect/ Aging Management Program Further Item Link and/or Material Environment Mechanism (AMP) Evaluation Component IV.B3.RP-24 IV.B3-25 Reactor vessel Stainless Reactor coolant Loss of material Chapter XI.M2, Water Chemistry No (RP-24) internal steel; nickel and neutron flux due to pitting and components alloy crevice corrosion IV.B3.RP-382 IV.B3-22 Reactor vessel Stainless Reactor coolant Cracking Chapter XI.M1, ASME Section XI No (R-170) internals: ASME steel; nickel and neutron flux due to fatigue, Inservice Inspection, Subsections IWB,Section XI, alloy stress corrosion IWC, and Examination cracking, or IWD or Chapter XI.M16A, PWR Vessel Category irradiation-assisted Internals, by invoking applicable B-N-3 core support stress corrosion 10 CFR 50.55a and ASME Section XI structure cracking; inservice inspection requirements components (not loss of material already identified due to wear as Existing Programs components in MRP-227-A)

IV.B3.RP-338 Upper internals Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel No assembly (designs steel and neutron flux due to fatigue Internals with core shrouds assembled with full height shroud plates): fuel alignment plate IV.B3.RP-400 Core Support Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel No Barrel Assembly: steel and neutron flux stress corrosion Internals and Chapter XI.M2, Water thermal shield cracking, irradiation- Chemistry (for SCC mechanisms only) positioning pins assisted stress corrosion cracking or fatigue; loss of material due to wear A-33

(4) Revised version of GALL Report Chapter IV.B4 B4. REACTOR VESSEL INTERNALS (PWR) - BABCOCK AND WILCOX Systems, Structures, and Components This section addresses the Babcock and Wilcox (B&W) pressurized-water reactor (PWR) vessel internals, which consist of components in the plenum cover assembly, the upper grid assembly, the control rod guide tube (CRGT) assembly, the core support shield assembly, the core barrel assembly, the lower grid assembly, incore monitoring instrumentation (IMI) guide tube assembly, and the flow distributor assembly.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (IV.A2).

A-34

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-242 IV.B4-4 Control rod guide Cast Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-183) tube (CRGT) austenitic and neutron flux toughness assembly: CRGT stainless due to thermal aging spacer castings steel embrittlement IV.B4.RP-242a Control rod guide Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No tube (CRGT) steel and neutron flux stress corrosion and Chapter XI.M2, Water Chemistry (for assembly: CRGT (including cracking or fatigue SCC mechanisms only) spacer castings CASS)

IV.B4.RP-245 IV.B4-13 Core barrel Nickel alloy Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-194) assembly and neutron flux due to stress and Chapter XI.M2, Water Chemistry (applicable to corrosion cracking Crystal River Unit 3 or Davis Besse only):

surveillance specimen holder tube (SSHT) studs/nuts or bolts IV.B4.RP-245a Core barrel Nickel alloy Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly and neutron flux fatigue (applicable to Crystal River Unit 3 or Davis Besse only):

surveillance specimen holder tube (SSHT) stud or bolt locking devices A-35

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-245b Core barrel Nickel alloy Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No assembly and neutron flux to wear; changes in (applicable to dimensions due to CR-3 or DB void swelling or only): distortion surveillance specimen holder tube (SSHT) stud or bolt locking devices IV.B4.RP-247 IV.B4-13 Core barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-194) assembly: lower steel; nickel and neutron flux due to stress and Chapter XI.M2, Water Chemistry core barrel (LCB) alloy corrosion cracking bolts IV.B4.RP-247a Core barrel Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly: lower steel; nickel and neutron flux fatigue core barrel (LCB) alloy bolt locking devices IV.B4.RP-247b Core barrel Stainless Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No assembly: lower steel; nickel and neutron flux to wear; changes in core barrel (LCB) alloy dimensions due to bolt locking void swelling or devices distortion IV.B4.RP-249 IV.B4-12 Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-196) assembly: baffle steel and neutron flux toughness plates due to neutron irradiation embrittlement IV.B4.RP-249a Core barrel Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly: baffle steel and neutron flux irradiation-assisted and Chapter XI.M2, Water Chemistry plates stress corrosion cracking A-36

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-241 IV.B4-7 Core barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-125) assembly: baffle- steel and neutron flux due to stress and Chapter XI.M2, Water Chemistry (for to-former bolts corrosion cracking, SCC mechanisms only) and screws irradiation-assisted stress corrosion cracking, fatigue, and overload IV.B4.RP-241a Core barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No assembly: steel and neutron flux due to stress and Chapter XI.M2, Water Chemistry (for locking devices corrosion cracking, SCC mechanisms only)

(including locking irradiation-assisted welds) of baffle- stress corrosion to-former bolts cracking, fatigue, and internal and overload baffle-to-baffle bolts IV.B4.RP-240 IV.B4-1 Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-128) assembly: baffle- steel and neutron flux toughness to-former bolts due to neutron and screws irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear A-37

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-240a Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No assembly: steel and neutron flux toughness locking devices due to neutron (including locking irradiation welds) of baffle- embrittlement; to-former bolts loss of material and internal due to wear baffle-to-baffle bolts IV.B4.RP-250 IV.B4-12 Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-196) assembly: core steel and neutron flux toughness barrel cylinder due to neutron (including vertical irradiation and embrittlement circumferential seam welds);

former plates IV.B4.RP-250a Core barrel Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly: core steel and neutron flux irradiation-assisted and Chapter XI.M2, Water Chemistry (for barrel cylinder stress corrosion SCC mechanisms only)

(including vertical cracking or fatigue and circumferential seam welds);

former plates IV.B4.RP-375 Core barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No assembly: steel and neutron flux due to irradiation- and Chapter XI.M2, Water Chemistry (for internal baffle-to- assisted stress SCC mechanisms only) baffle bolts corrosion cracking, fatigue, or overload A-38

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-375a Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No assembly: steel and neutron flux toughness internal baffle-to- due to neutron baffle bolts irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear IV.B4.RP-244 IV.B4-7 Core barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-125) assembly; steel and neutron flux due to irradiation- and Chapter XI.M2, Water Chemistry (for external baffle- assisted stress SCC mechanisms only) to-baffle bolts corrosion cracking, and core barrel- fatigue, and to-former bolts; overload IV.B4.RP-244a Core barrel Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No assembly: steel and neutron flux due to irradiation- and Chapter XI.M2, Water Chemistry (for locking devices assisted stress SCC mechanisms only)

(including welds) corrosion cracking, of external baffle- or fatigue to-baffle bolts and core barrel-to-former bolts A-39

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-243 IV.B4-1 Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No (R-128) assembly: steel and neutron flux toughness external baffle- due to neutron to-baffle bolts irradiation and core barrel- embrittlement; to-former bolts loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear IV.B4.RP-243a Core barrel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No assembly: steel and neutron flux toughness locking devices due to neutron (including welds) irradiation of external baffle- embrittlement; loss to-baffle bolts of material and core barrel- due to wear to-former bolts IV.B4.RP-248 IV.B4-12 Core support Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-196) shield (CSS) steel; nickel and neutron flux due to stress and Chapter XI.M2, Water Chemistry assembly: upper alloy corrosion cracking core barrel (UCB) bolts IV.B4.RP-248a Core support Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No shield (CSS) steel; nickel and neutron flux fatigue assembly: upper alloy core barrel (UCB) bolt locking devices A-40

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-248b Core support Stainless Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No shield (CSS) steel; nickel and neutron flux to wear; changes in assembly: upper alloy dimensions due to core barrel void swelling or (UCB) bolt distortion locking devices IV.B4.RP-252 IV.B4-16 Core support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-188) shield (CSS) steel, and neutron flux toughness assembly: CSS including due to thermal aging vent valve top CASS and embrittlement and bottom PH steels retaining rings (valve body components)

IV.B4.RP-252a IV.B4-16 Core support Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No (R-188) shield (CSS) steel and neutron flux stress corrosion and Chapter XI.M2, Water Chemistry (for assembly: cracking or fatigue SCC mechanisms only)

CSS vent valve top and bottom retaining rings; vent valve locking devices (valve body components)

IV.B4.RP-251 IV.B4-15 Core support Stainless Reactor coolant Loss of material Chapter XI.M16A, PWR Vessel Internals No (R-190) shield (CSS) steel and neutron flux due to wear; assembly: loss of preload CSS top flange (wear)

IV.B4.RP-251a IV.B4-15 Plenum cover Stainless Reactor coolant Loss of material Chapter XI.M16A, PWR Vessel Internals No (R-190) assembly: steel and neutron flux due to wear; plenum cover loss of preload weldment rib (wear) pads and plenum cover support flange A-41

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-256 IV.B4-25 Flow distributor Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-210) assembly: flow steel; nickel and neutron flux due to stress and Chapter XI.M2, Water Chemistry distributor bolts alloy corrosion cracking IV.B4.RP-256a Flow distributor Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly: flow steel; nickel and neutron flux fatigue distributor bolt alloy locking devices IV.B4.RP-256b Flow distributor Stainless Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No assembly: flow steel; nickel and neutron flux to wear; changes in distributor bolt alloy dimensions due to locking devices distortion or void swelling or distortion IV.B4.RP-259 IV.B4-31 Incore Monitoring Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-205) Instrument (IMI) steel and neutron flux toughness guide tube due to neutron assembly: IMI irradiation guide tube embrittlement spider-to-lower grid rib section welds IV.B4.RP-259a Incore Monitoring Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No.

Instrument (IMI) steel and neutron flux stress corrosion and Chapter XI.M2, Water Chemistry guide tube cracking, irradiation-assembly: assisted stress IMI guide tube corrosion cracking spider-to-lower or fatigue grid rib sections welds A-42

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-258 IV.B4-4 Incore Monitoring Cast Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-183) Instrument (IMI) austenitic and neutron flux toughness guide tube stainless due to thermal aging assembly: IMI steel and neutron guide tube irradiation spiders embrittlement (castings)

IV.B4.RP-258a Incore Monitoring Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No Instrumentation steel and neutron flux stress corrosion and Chapter XI.M2, Water Chemistry (IMI) guide tube cracking, irradiation-assembly: assisted stress IMI guide tube corrosion cracking spiders or fatigue IV.B4.RP-254 IV.B4-25 Lower grid Nickel alloy Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-210) assembly: and neutron flux due to stress and Chapter XI.M2, Water Chemistry alloy X-750 lower corrosion cracking grid shock pad bolts (Three Mile Island Unit 1, only)

IV.B4.RP-254a Lower grid Nickel alloy Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly: and neutron flux fatigue alloy X-750 lower grid shock pad bolt locking devices (Three Mile Island Unit 1, only)

IV.B4.RP-254b Lower grid Nickel Alloy Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No assembly: and neutron flux to wear; changes in alloy X-750 lower dimensions due to grid shock pad void swelling or bolt locking distortion devices (Three Mile Island Unit 1, only)

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-246 IV.B4-12 Lower grid Stainless Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-196) assembly: steel; nickel and neutron flux due to stress and Chapter XI.M2, Water Chemistry upper thermal alloy corrosion cracking shield (UTS) bolts and lower thermal shield (LTS) bolts IV.B4.RP-246a Lower grid Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly: steel; nickel and neutron flux fatigue upper thermal alloy shield (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices IV.B4.RP-246b Lower grid Stainless Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No assembly: steel; nickel and neutron flux to wear; changes in upper thermal alloy dimensions due to shield (UTS) bolt void swelling or locking devices distortion and lower thermal shield (LTS) bolt locking devices IV.B4.RP-260 IV.B4-31 Lower grid fuel Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals" No (R-205) assembly: (a) steel; nickel and neutron flux toughness pads; (b) pad-to- alloy due to neutron rib section welds; irradiation (c) alloy X-750 embrittlement dowels, cap screws and locking devices A-44

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-260a Lower grid fuel Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly: (a) steel; nickel and neutron flux stress corrosion and Chapter XI.M2, Water Chemistry (for pads; (b) pad-to- alloy cracking or fatigue SCC mechanisms only) rib section welds; (c) alloy X-750 dowels, cap screws and locking devices IV.B4.RP-262 IV.B4-32 Lower grid Nickel alloy Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-203) assembly: alloy and neutron flux due to stress and Chapter XI.M2, Water Chemistry X-750 dowel-to- corrosion cracking lower fuel assembly support pad locking welds IV.B4.RP-261 IV.B4-32 Lower grid Nickel alloy Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-203) assembly: alloy and neutron flux due to stress and Chapter XI.M2, Water Chemistry X-750 dowel-to- corrosion cracking guide block welds IV.B4.R-53 IV.B4-37 Reactor vessel Stainless Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, TLAA (R-53) internal steel; nickel and neutron flux damage (TLAA) to be evaluated for the period of components alloy due to fatigue extended operation. See the SRP, Section 4.3 Metal Fatigue, for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

IV.B4.RP-24 IV.B4-38 Reactor vessel Stainless Reactor coolant Loss of material Chapter XI.M2, Water Chemistry No (RP-24) internal steel; nickel and neutron flux due to pitting and components alloy crevice corrosion A-45

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-376 Reactor vessel Stainless Reactor coolant Reduction in ductility Ductility - Reduction in Fracture Yes, TLAA internal steel; nickel and neutron flux and fracture Toughness is a TLAA (BAW-2248A) to be components alloy toughness evaluated for the period of extended due to neutron operation. See the SRP, Section 4.7, irradiation "Other Plant-Specific TLAAs," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

IV.B4.RP-382 IV.B4-42 Reactor vessel Stainless Reactor coolant Cracking Chapter XI.M1, ASME Section XI No (R-179) internals: ASME steel; nickel and neutron flux due to fatigue, Inservice Inspection, Subsections IWB,Section XI, alloy stress corrosion IWC, and IWD or Chapter XI.M16A, Examination cracking, or PWR Vessel Internals, by invoking Category irradiation-assisted applicable 10 CFR 50.55a and ASME B-N-3 core stress corrosion Section XI inservice inspection support structure cracking; requirements components loss of material due to wear IV.B4.RP-352 Upper grid Nickel alloy Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No assembly: alloy and neutron flux due to stress and Chapter XI.M2, Water Chemistry X-750 dowel-to- corrosion cracking upper fuel assembly support pad welds (all plants except Davis-Besse)

A-46

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure Aging Effect/ Further Item Link and/or Material Environment Aging Management Program (AMP)

Mechanism Evaluation Component IV.B4.RP-236 Reactor internal Stainless Reactor coolant No additional aging Chapter XI.M16A, PWR Vessel Internals No No Additional steel; nickel and neutron flux management for Measures alloy reactor internal No components Additional Measures components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists IV.B4.RP-400 Core support Stainless Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No shield assembly: steel and neutron flux stress corrosion and Chapter XI.M2, Water Chemistry upper (top) cracking flange weld IV.B4.RP-401 Core support Stainless Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No shield assembly: steel and neutron flux toughness due to upper (top) neutron irradiation flange weld embrittlement A-47

(5) Revised version of GALL Report Chapter IX.C and IX.G IX.C Selected Definitions & Use of Terms for Describing and Standardizing MATERIALS Stainless steel Products grouped under the term stainless steel include austenitic, ferritic, martensitic, precipitation-hardened (PH),

or duplex stainless steel (Cr content >11%). These stainless steels may be fabricated using a wrought or cast process. These materials are susceptible to a variety of aging effects and mechanisms, including loss of material due to pitting and crevice corrosion, and cracking due to stress corrosion cracking. In some cases, when an aging effect is applicable to all of the various stainless steel categories, it can be assumed that the term stainless steel in the Material column of an AMR line-item in the GALL Report encompasses all stainless steel types. Cast austenitic stainless steel (CASS) is quite susceptible to loss of fracture toughness due to thermal and neutron irradiation embrittlement. In addition, MRP-227-A indicates that PH stainless steels or martensitic stainless steels may be susceptible to loss of fracture toughness by a thermal aging mechanism. Therefore, when loss of fracture toughness due to thermal and neutron irradiation embrittlement is an applicable aging effect and mechanism for a component in the GALL Report, the CASS, PH stainless steel, or martensitic stainless steel designation is specifically identified in an AMR line-item.

Steel with stainless steel cladding also may be considered stainless steel when the aging effect is associated with the stainless steel surface of the material, rather than the composite volume of the material.

Examples of stainless steel designations that comprise this category include A-286, SA193-Gr. B8, SA193-Gr. B8M, Gr.

660 (A-286), SA193-6, SA193-Gr. B8 or B-8M, SA453, Type 416, Type 403, 410, 420, and 431 martensitic stainless steels, Type 15-5, 17-4, and 13-8-Mo PH stainless steels, and SA-193, Grade B8 and B8M bolting materials.

Examples of wrought austenitic stainless materials that comprise this category include Type 304, 304NG, 304L, 308, 308L, 309, 309L, 316 and 347. Examples of CASS that comprise this category include CF3, CF3M, CF8 and CF8M. [Ref. 6, 7, 30]

A-48

IX.G References

30. Welding Handbook, Seventh Edition, Volume 4, Metals and Their Weldability, American Welding Society, 1984, p.76-145.

A-49

Appendix A, Section 2 - Revised version of the SRP-LR (1) Revised version of SRP-LR Table 3.0-1 Table 3.0-1 FSAR Supplement for Aging Management of Applicable Systems GALL GALL Description of Program Implementation Applicable GALL Chapter Program Schedule Report and SRP-LR Chapter References XI.M16A PWR Vessel The program relies on implementation Program GALL IV / SRP 3.1 Internals of the inspection and evaluation should be guidelines in EPRI Technical Report implemented No. 1022863 (MRP-227-A) and EPRI prior to period Technical Report No. 1016609 of extended (MRP-228) to manage the aging operation effects on the reactor vessel internal components. This program is used to manage (a) cracking, including stress corrosion cracking, primary water stress corrosion cracking, irradiation-assisted stress corrosion cracking, and cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging, neutron irradiation embrittlement, or void swelling; (d) dimensional changes due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

(2) Revised version of SRP-LR Section 3.1.2, Acceptance Criteria 3.1.2.2.9 Removed as a result of LR-ISG-2011-04 3.1.2.2.10 Removed as a result of LR-ISG-2011-04 3.1.2.2.12 Removed as a result of LR-ISG-2011-04 3.1.2.2.13 Removed as a result of LR-ISG-2011-04 3.1.2.2.14 Removed as a result of LR-ISG-2011-04 (3) Revised version of SRP-LR Section 3.1.3, Review Procedures 3.1.3.2.9 Removed as a result of LR-ISG-2011-04 3.1.3.2.10 Removed as a result of LR-ISG-2011-04 3.1.3.2.12 Removed as a result of LR-ISG-2011-04 3.1.3.2.13 Removed as a result of LR-ISG-2011-04 3.1.3.2.14 Removed as a result of LR-ISG-2011-04 A-50

(5) Revised version of SRP-LR Table 3.1-1 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 3 BWR/ Stainless steel or nickel Cumulative fatigue damage Fatigue is a TLAA Yes, TLAA (See IV.B1.R-53 IV.B1-14 (R-53)

PWR alloy reactor vessel internal due to fatigue evaluated for the period of subsection 3.1.2.2.1) IV.B2.RP-303 IV.B2-31 (R-53) components exposed to extended operation (See IV.B3.RP-339 IV.B3-24 (R-53)

IV.B4.R-53 IV.B4-37 (R-53) reactor coolant and neutron SRP, Section 4.3 Metal flux Fatigue, for acceptable methods to comply with 10 CFR 54.21(c)(1) 15 PWR Stainless steel Babcock & Reduction of ductility and Ductility - Reduction in Yes, TLAA (See IV.B4.RP-376 N/A Wilcox (including CASS, fracture toughness due to fracture toughness is a subsection 3.1.2.2.3.3) martensitic SS, and PH SS) neutron irradiation TLAA to be evaluated for and nickel alloy reactor embrittlement, and for the period of extended vessel internal components CASS, martensitic SS, and operation, See the SRP, exposed to reactor coolant PH SS due to thermal aging Section 4.7, Other Plant-and neutron flux embrittlement Specific TLAAs, for acceptable methods of meeting the requirements of 10 CFR 54.21(c).

28 PWR Stainless steel Combustion Loss of material due to Chapter XI.M16A, PWR No IV.B3.RP-400 N/A Engineering Existing wear; cracking due to stress Vessel Internals, and Programs components corrosion cracking, Chapter XI.M2, Water exposed to reactor coolant irradiation-assisted stress Chemistry (for SCC and neutron flux corrosion cracking, or mechanisms only) fatigue 32 PWR Stainless steel, nickel alloy, Cracking, or loss of material Chapter XI.M1, ASME No IV.B2.RP-382 IV.B2-26 (R-142) or CASS reactor vessel due to wear Section XI Inservice IV.B3.RP-382 IV.B3-22 (R-170) internals, core support Inspection, Subsections IV.B4.RP-382 IV.B4-42 (R-179) structure (not already IWB, IWC, and IWD or referenced as ASME Chapter XI.M16A, PWR Section XI Examination Vessel Internals, invoking Category B-N-3 core applicable 10 CFR 50.55a support structure and ASME Section XI components in MRP-227- inservice inspection A), exposed to reactor requirements for these coolant and neutron flux components A-51

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 51a PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B4.RP-241 IV.B4-7 (R-125) alloy Babcock & Wilcox corrosion cracking, Vessel Internals, and IV.B4.RP-241a N/A reactor internal Primary irradiation-assisted stress Chapter XI.M2, Water IV.B4.RP-242a N/A IV.B4.RP-247 IV.B4-13 (R-194) components exposed to corrosion cracking, or Chemistry (for SCC IV.B4.RP-247a N/A reactor coolant and neutron fatigue mechanisms only) IV.B4.RP-248 IV.B4-25 (R-210) flux IV.B4.RP-248a N/A IV.B4.RP-249a N/A IV.B4.RP-252a N/A IV.B4.RP-256 IV.B4-25 (R-210)

IV.B4.RP-256a N/A IV.B4.RP-258a N/A IV.B4.RP-259a N/A IV.B4.RP-261 IV.B4-32 (R-203)

IV.B4.RP-400 N/A 51b PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B4.RP-244 IV.B4-7 (R-125) alloy Babcock & Wilcox corrosion cracking, Vessel Internals, and IV.B4.RP-244a N/A reactor internal Expansion irradiation-assisted stress Chapter XI.M2, Water IV.B4.RP-245 IV.B4-13 (R-194)

IV.B4.RP-245a N/A components exposed to corrosion cracking, fatigue, Chemistry (for SCC IV.B4.RP-246 IV.B4-12 (R-196) reactor coolant and neutron or overload mechanisms only) IV.B4.RP-246a N/A flux IV.B4.RP-254 IV.B4-25 (R-210)

IV.B4.RP-254a N/A IV.B4.RP-260a N/A IV.B4.RP-262 IV.B4-32 (R-203)

IV.B4.RP-352 N/A IV.B4.RP-250a N/A IV.B4.RP-375 N/A 52a PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B3.RP-312 IV.B3-2 (R-149) alloy Combustion corrosion cracking, Vessel Internals, and IV.B3.RP-314 IV.B3-9 (R-162)

Engineering reactor irradiation-assisted stress Chapter XI.M2, Water IV.B3.RP-322 N/A IV.B3.RP-324 N/A internal Primary corrosion cracking, or Chemistry (for SCC IV.B3.RP-326a N/A components exposed to fatigue mechanisms only) IV.B3.RP-327 IV.B3-15 (R-155) reactor coolant and neutron IV.B3.RP-328 IV.B3-15 (R-155) flux IV.B3.RP-342 N/A IV.B3.RP-358 N/A IV.B3.RP-362a N/A IV.B3.RP-363 N/A IV.B3.RP-338 N/A IV.B3.RP-343 N/A A-52

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 52b PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B3.RP-313 NA alloy Combustion corrosion cracking, Vessel Internals, and IV.B3.RP-316 IV.B3-9 (R-162)

Engineering reactor internal irradiation-assisted stress Chapter XI.M2, Water IV.B3.RP-323 N/A IV.B3.RP-325 N/A Expansion components corrosion cracking, or Chemistry (for SCC IV.B3.RP-329 IV.B3-12 (R-155) exposed to reactor coolant fatigue mechanisms only) IV.B3.RP-330 IV.B3-23 (R-167) and neutron flux IV.B3.RP-333 N/A IV.B3.RP-335 IV.B3-23 (R-167)

IV.B3.RP-362c N/A 52c PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B3.RP-320 IV.B3-9 (R-162) alloy Combustion corrosion cracking, Vessel Internals, and IV.B3.RP-334 IV.B3-23 (R-167)

Engineering reactor internal irradiation-assisted stress Chapter XI.M2, Water Existing Programs corrosion cracking, or Chemistry (for SCC components exposed to fatigue mechanisms only) reactor coolant and neutron flux 53a PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B2.RP-270a N/A alloy Westinghouse corrosion cracking, Vessel Internals, and IV.B2.RP-271 IV.B2-10 (R-125) reactor internal Primary irradiation-assisted stress Chapter XI.M2, Water IV.B2.RP-275 IV.B2-6 (R-128)

IV.B2.RP-276 IV.B2-8 (R-120) components exposed to corrosion cracking, or Chemistry (for SCC IV.B2.RP-280 IV.B2-8 (R-120) reactor coolant and neutron fatigue mechanisms only) IV.B2.RP-298 IV.B2-28 (R-118) flux IV.B2.RP-302 N/A IV.B2.RP-387 N/A 53b PWR Stainless steel Cracking due to stress Chapter XI.M16A, PWR No IV.B2.RP-273 IV.B2-10 (R-125)

Westinghouse reactor corrosion cracking, Vessel Internals, and IV.B2.RP-278 IV.B2-8 (R-120) internal Expansion irradiation-assisted stress Chapter XI.M2, Water IV.B2.RP-286 IV.B2-16 (R-133)

IV.B2.RP-291 IV.B2-24 (R-138) components exposed to corrosion cracking, or Chemistry (for SCC IV.B2.RP-291a N/A reactor coolant and neutron fatigue mechanisms only) IV.B2.RP-291b N/A flux IV.B2.RP-293 IV.B2-24 (R-138)

IV.B2.RP-294 IV.B2-24 (R-138)

IV.B2.RP-387a N/A 53c PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B2.RP-289 IV.B2-20 (R-130) alloy Westinghouse corrosion cracking, Vessel Internals, and IV.B2.RP-301 IV.B2-40 (R-112) reactor internal Existing irradiation-assisted stress Chapter XI.M2, Water IV.B2.RP-345 N/A IV.B2.RP-346 N/A Programs components corrosion cracking, or Chemistry (for SCC IV.B2.RP-399 N/A exposed to reactor coolant fatigue mechanisms only) IV.B2.RP-355 N/A and neutron flux A-53

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 54 PWR Stainless steel bottom Loss of material due to Chapter XI.M16A, PWR No IV.B2.RP-284 IV.B2-13 (R-145) mounted instrument wear Vessel Internals, or system flux thimble Chapter XI.M37, Flux tubes (with or without Thimble Tube Inspection chrome plating) exposed to reactor coolant and neutron flux (Westinghouse Existing Programs components) 55a PWR Stainless steel or nickel No additional aging Chapter XI.M16A, PWR No IV.B4.RP-236 NA alloy Babcock and Wilcox management for reactor Vessel Internals reactor internal No internal No Additional Additional Measures Measures components components exposed to unless required by ASME reactor coolant and neutron Section XI, Examination flux Category B-N-3 or relevant operating experience exists 55b PWR Stainless steel or nickel No additional aging Chapter XI.M16A, PWR No IV.B3.RP-306 NA alloy Combustion management for reactor Vessel Internals Engineering reactor internal internal No Additional No Additional Measures Measures components components exposed to unless required by ASME reactor coolant and neutron Section XI, Examination flux Category B-N-3 or relevant operating experience exists 55c PWR Stainless steel or nickel No additional aging Chapter XI.M16A, PWR No IV.B2.RP-265 NA alloy Westinghouse management for reactor Vessel Internals reactor internal No internal No Additional Additional Measures Measures components components exposed to unless required by ASME reactor coolant and neutron Section XI, Examination flux Category B-N-3 or relevant operating experience exists 56a PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B3.RP-315 IV.B3-7 (R-165) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B3.RP-318 IV.B3-8 (R-163) martensitic SS) or nickel embrittlement and for CASS, IV.B3.RP-359 N/A IV.B3.RP-360 N/A alloy Combustion martensitic SS, and PH SS A-54

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended Engineering reactor internal due to thermal aging IV.B3.RP-362 N/A Primary components embrittlement; or changes in IV.B3.RP-364 N/A exposed to reactor coolant dimensions due to void IV.B3.RP-366 N/A IV.B3.RP-365 N/A and neutron flux swelling or distortion; or loss IV.B3.RP-326 N/A of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 56b PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B3.RP-317 IV.B3-7 (R-165) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B3.RP-331 N/A martensitic SS) embrittlement and for CASS, IV.B3.RP-359a N/A IV.B3.RP-361 N/A Combustion Engineering martensitic SS, and PH SS IV.B3.RP-362b N/A Expansion reactor internal due to thermal aging components exposed to embrittlement; or changes in reactor coolant and neutron dimensions due to void flux swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 56c PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B3.RP-319 IV.B3-9 (R-162) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B3.RP-332 IV.B3-17 (R-156) martensitic SS) or nickel embrittlement and for CASS, IV.B3.RP-334a N/A IV.B3.RP-336 IV.B3-22 (R-170) alloy Combustion martensitic SS, and PH SS IV.B3.RP-357 N/A Engineering reactor internal due to thermal aging Existing Programs embrittlement; or changes in components exposed to dimensions due to void reactor coolant and neutron swelling or distortion; or loss flux of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 58a PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B4.RP-240 IV.B4-1 (R-128) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B4.RP-240a N/A martensitic SS) or nickel embrittlement and for CASS, IV.B4.RP-242 IV.B4-4 (R-183)

IV.B4.RP-247b N/A alloy Babcock & Wilcox martensitic SS, and PH SS IV.B4.RP-248b N/A reactor internal Primary due to thermal aging IV.B4.RP-249 IV.B4-12 (R-196) components exposed to embrittlement; or changes in IV.B4.RP-251 IV.B4-15 (R-190) reactor coolant and neutron dimensions due to void IV.B4.RP-251a N/A flux swelling or distortion; or loss IV.B4.RP-252 IV.B4-16 (R-188)

A-55

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended of preload due to wear; or IV.B4.RP-254b N/A loss of material due to wear IV.B4.RP-256b N/A IV.B4.RP-258 IV.B4-4 (R-183)

IV.B4.RP-259 IV.B4-31 (R-205)

IV.B4.RP-401 N/A 58b PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B4.RP-245b N/A including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B4.RP-246b N/A martensitic SS) or nickel embrittlement and for CASS, IV.B4.RP-254b N/A IV.B4.RP-260 IV.B4-31 (R-205) alloy Babcock & Wilcox martensitic SS, and PH SS IV.B4.RP-243 IV.B4-1 (R-128) reactor internal Expansion due to thermal aging IV.B4.RP-243a N/A components exposed to embrittlement; or changes in IV.B4.RP-250 IV.B4-12 (R-196) reactor coolant and neutron dimensions due to void IV.B4.RP-375a N/A flux swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 59a PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B2.RP-270 IV.B2-1 (R-124) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B2.RP-272 IV.B2-6 (R-128) martensitic SS) or nickel embrittlement and for CASS, IV.B2.RP-296 N/A IV.B2.RP-297 N/A alloy Westinghouse reactor martensitic SS, and PH SS IV.B2.RP-302a N/A internal Primary due to thermal aging IV.B2.RP-354 N/A components exposed to embrittlement; or changes in IV.B2.RP-388 N/A reactor coolant and neutron dimensions due to void IV.B2.RP-300 N/A flux swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 59b PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B2.RP-274 IV.B2-6 (R-128) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B2.RP-278a N/A martensitic SS) embrittlement and for CASS, IV.B2.RP-287 IV.B2-17 (R-135)

IV.B2.RP-290 IV.B2-21 (R-140)

Westinghouse reactor martensitic SS, and PH SS IV.B2.RP-290a N/A internal Expansion due to thermal aging IV.B2.RP-290b N/A components exposed to embrittlement; or changes in IV.B2.RP-292 IV.B2-21 (R-140) reactor coolant and neutron dimensions due to void IV.B2.RP-295 IV.B2-22 (R-141) flux swelling or distortion; or loss IV.B2.RP-388a N/A of preload due to thermal and irradiation enhanced stress relaxation or creep; or A-56

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended loss of material due to wear 59c PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B2.RP-285 IV.B2-14 (R-137) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B2.RP-288 IV.B2-18 (R-132) martensitic SS) or nickel embrittlement and for CASS, IV.B2.RP-299 IV.B2-34 (R-115)

IV.B2.RP-356 N/A alloy Westinghouse reactor martensitic SS, and PH SS internal Existing due to thermal aging Programs components embrittlement; or changes in exposed to reactor coolant dimensions due to void and neutron flux swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear A-57

Appendix B MARK-UP OF CHANGES TO THE GALL REPORT AND SRP-LR B-1

Appendix B, Section 1 - Mark-up of Changes to the GALL Report In the mark-up, strikethrough text indicates a deletion and underline text indicates an insertion.

Double strikethrough text indicates the original location of the moved text and a double underline text indicates the final location of the moved text.

(1) Mark-up of changes to GALL Report AMP XI.M16A XI.M16A PWR VESSEL INTERNALS Program Description This program relies on implementation of the Electric Power Research Institute (EPRI)

Technical Report No. 10165961022863, Materials Reliability Program: Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines, (MRP-227-A) and EPRI Report Technical No. 1016609, Materials Reliability Program: Inspection Standard for PWR Internals, (MRP-228) to manage the aging effects on the pressurized water reactor (PWR) reactor vessel internal (RVI) components. The recommended activities in MRP-227-A and additional plant-specific activities not defined in MRP-227-A are implemented in accordance with Nuclear Energy Institute (NEI) 03-08, Guideline for the Management of Materials Issues. The staff approved the augmented inspection and evaluation (I&E) criteria for PWR RVI components in NRC Safety Evaluation (SE),

Revision 1, on MRP-227 by letter dated December 16, 2011.

This program is used to manage the effects of age-related degradation mechanisms that are applicable in general to the PWR RVI components at the facility. These aging effects include :

(a) various forms of cracking, including stress corrosion cracking (SCC), which also encompasses primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), orand cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging or neutron irradiation embrittlement; (d) changes in dimensions due to void swelling or distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

The program applies the guidance in MRP-227-A for inspecting, evaluating, and, if applicable, dispositioning non-conforming RVI components at the facility. The program conforms to the definition of a sampling-based condition monitoring program, as defined by the Branch Technical Position RSLB-1, with periodic examinations and other inspections of highly-affected internals locations. These examinations provide reasonable assurance that the effects of age-

-related degradation mechanisms will be managed during the period of extended operation.

The program includes expanding periodic examinations and other inspections, if the extent of the degradation effects identified exceeds the expected levels.

The MRP-227-A guidance for selecting RVI components for inclusion in the inspection sample is based on a four-step ranking process. Through this process, the reactor internalsRVIs for all three PWR designs were assigned to one of the following four groups: Primary, , Expansion,

, Existing Programs,, and No Additional Measures components. Definitions of each group are provided in Generic Aging Lessons Learned Report (GALL Report), Revision 2, Chapter IX.B.

The result of this four-step sample selection process is a set of Primary Iinternals Ccomponent locations for each of the three plant designs that are inspected because they are expected to show the leading indications of the degradation effects, with another set of Expansion B-2

Iinternals Ccomponent locations that are specified to expand the sample should the indications be more severe than anticipated.

The degradation effects in a third set of internals locations are deemed to be adequately managed by Existing Programs, such as American Society of Mechanical Engineers (ASME)

Code,Section XI,11 Examination Category B-N-3 examinations of core support structures. A fourth set of internals locations are deemed to require Nno aAdditional mMeasures. As a result, the program typically identifies 5 to 15% of the RVI locations as Primary Component locations for inspections, with another 7 to 10% of the RVI locations to be inspected as Expansion Components, as warranted by the evaluation of the inspection results. Another 5 to 15% of the internals locations are covered by Existing Programs, with the remainder requiring no additional measures. This process thus uses appropriate component functionality criteria, age-related degradation susceptibility criteria, and failure consequence criteria to identify the components that will be inspected under the program in a manner that conforms to the sampling criteria for sampling-based condition monitoring programs in Section A.1.2.3.4 of NRC Branch Position RLSB-1. Consequently, the sample selection process is adequate to assure that the intended function(s) of the PWR reactor internal components are maintained during the period of extended operation.

The programs use of visual examination methods in MRP-227 for detection of relevant conditions (and the absence of relevant conditions as a visual examination acceptance criterion) is consistent with the ASME Code,Section XI rules for visual examination. However, the programs adoption of the MRP-227 guidance for visual examinations goes beyond the ASME Code,Section XI visual examination criteria because additional guidance is incorporated into MRP-227 to clarify how the particular visual examination methods will be used to detect relevant conditions and describes in more detail how the visual techniques relate to the specific RVI components and how to detect their applicable age-related degradation effects.

The technical basis for detecting relevant conditions using volumetric ultrasonic testing (UT) inspection techniques can be found in MRP-228, where the review of existing bolting UT examination technical justifications has demonstrated the indication detection capability of at least two vendors, and where vendor technical justification is a requirement prior to any additional bolting examinations. Specifically, the capability of programs UT volumetric methods to detect loss of integrity of PWR internals bolts, pins, and fasteners, such as baffle-former bolting in B&W and Westinghouse units, has been well demonstrated by operating experience.

In addition, the programs adoption of the MRP-227 guidance and process incorporates the UT criteria in MRP-228, which calls for the technical justifications that are needed for volumetric examination method demonstrations, required by the ASME Code,Section V.

The program also includes future industry operating experience as incorporated in periodic revisions to MRP-227. The program thus provides reasonable assurance for the long-term integrity and safe operation of reactor internals in all commercial operating U.S. PWR nuclear power plants.

Age-related degradation in the reactor internals is managed through an integrated program.

Specific features of the integrated program are listed in the following ten program elements.

Degradation due to changes in material properties (e.g., loss of fracture toughness) was considered in the determination of inspection recommendations and is managed by the requirement to use appropriately degraded properties in the evaluation of identified defects. The integrated program is implemented by the applicant through an inspection plan that is submitted to the NRC for review and approval with the application for license renewal.

Evaluation and Technical Basis 11 Refer to the GALL Report, Chapter I, for applicability of various editions of the ASME Code,Section XI.

B-3

1. Scope of Program: The scope of the program includes all RVI components at the [as an administrative action item for the AMP, the applicant to fill in the name of the applicants nuclear facility, including applicable units], which [is/are] built to a [applicant to fill in Westinghouse, CE, or B&W, as applicable] based on the plants applicable nuclear steam supply system NSSS design. The scope of the program applies the methodology and guidance in MRP-227-Athe most recently NRC-endorsed version of MRP-227, which provides an augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S. PWR nuclear power plants designed by Babcock & Wilcox (B&W, ),

Combustion Engineering (CE,), and Westinghouse. The scope of components considered for inspection underin MRP-227 guidance includes-A include core support structures (typically denoted as Examination Category B-N-3 by the ASME Code,Section XI),, those RVI components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1), and other RVI components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii). In addition, ASME Code,Section XI includes inspection requirements for PWR removable core support structures in Table IWB-2500-1, Examination Category B-N-3, which are in addition to any inspections that are implemented in accordance with MRP-227-A.

The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review (AMR), as defined by the criteria set in 10 CFR 54.21(a)(1).. The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are adequately managed in accordance with an applicants AMP that corresponds to GALL AMP XI.M1, ASME Code,Section XI Inservice Inspection, Subsections IWB, IWC, and IWD.

The scope of the program includes the response bases to applicable license renewal applicant action items (LRAAIs) on the MRP-227 methodology, and any additional programs, actions, or activities that are discussed in these LRAAI responses and credited for aging management of the applicants RVI components. The LRAAIs are identified in the staffs safety evaluation on MRP-227 and include applicable action items on meeting those assumptions that formed the basis of the MRPs augmented inspection and flaw evaluation methodology (as discussed in Section 2.4 of MRP-227), and NSSS vendor-specific or plant-specific LRAAIs as well. The responses to the LRAAIs on MRP-227 are provided in Appendix C of the LRA.

The guidance in MRP-227 specifies applicability limitations to base-loaded plants and the fuel loading management assumptions upon which the functionality analyses were based.

These limitations and assumptions require a determination of applicability by the applicant for each reactor and are covered in Section 2.4 of MRP-227.

2. Preventive Actions: The guidance in MRP-227-A relies on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of material induced by general, pitting corrosion, crevice corrosion, or stress corrosion cracking or any of its forms [SCC, PWSCC, or IASCC]). Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry Program, as described. The program B-4

description, evaluation, and technical basis of water chemistry are presented in GALL AMP XI.M2, Water Chemistry.

3. Parameters Monitored/Inspected: The program manages the following age-related degradation effects and mechanisms that are applicable in general to the RVI components at the facility: (a) cracking induced by SCC, PWSCC, IASCC, or fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness induced by either thermal aging or neutron irradiation embrittlement; (d) changes in dimensions due to void swelling and irradiation growth, or distortion, or deflection; and (e) loss of preload caused due toby thermal and irradiation--enhanced stress relaxation or creep.

For the management of cracking, the program monitors for evidence of surface breaking linear discontinuities if a visual inspection technique is used as the non-destructionve examination (NDE) method, or for relevant flaw presentation signals if a volumetric ultrasonic testing (UT) method is used as the NDE method. For the management of loss of material, the program monitors for gross or abnormal surface conditions that may be indicative of loss of material occurring in the components. For the management of loss of preload, the program monitors for gross surface conditions that may be indicative of loosening in applicable bolted, fastened, keyed, or pinned connections. The program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement, or by void swelling and irradiation growth; instead. Instead, the impact of loss of fracture toughness on component integrity is indirectly managed by: (1) using visual or volumetric examination techniques to monitor for cracking in the components, and by(2) applying applicable reduced fracture toughness properties in the flaw evaluations if, in cases where cracking is detected in the components and is extensive enough to warrant necessitate a supplemental flaw growth or flaw tolerance evaluation under the MRP-227 guidance or ASME Code,Section XI requirements.. The program uses physical measurements to monitor for any dimensional changes due to void swelling or, irradiation growth, distortion, or deflection..

Specifically, the program implements the parameters monitored/inspected criteria for [as an administrative action item for the AMP, applicant is to select one of the following to finish the sentence, as applicable to its NSSS vendor for its internals: for B&W designed Primary Components in Table 4-1 of MRP-227; for CE designed Primary Components in Table 4-2 of MRP-227; and for Westinghouse designed Primary Components in Table 4-3 of MRP-227]. Additionally, the program implements the parameters monitored/inspected criteria for

[as an administrative action item for the AMP, applicant is to select one of the following to finish the sentence, as applicable to its NSSS vendor for its internals: for B&W designed Expansion Components in Table 4-4 of MRP-227; for CE designed Expansion Components in Table 4-5 of MRP-227; and for Westinghouse designed Expansion Components in Table 4-6 of MRP-227]. The parameters monitored/inspected for Existing Program Components follow the bases for referenced Existing Programs, such as the requirements for ASME Code Class RVI components in ASME Code,Section XI, Table IWB-2500-1, Examination Categories B-N-3, as implemented through the applicants ASME Code,Section XI program, or the recommended program for inspecting Westinghouse-designed flux thimble tubes in GALL AMP XI.M37, Flux Thimble Tube Inspection. No inspections, except for those specified in ASME Code,Section XI, are required for components that are identified as requiring No Additional Measures, in accordance with the analyses reported in MRP-227.

Specifically, the program implements the parameters monitored/inspected criteria consistent with the applicable tables in Section 4, Aging Management Requirements, in MRP-227-A.

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4. Detection of Aging Effects: The detection of aging effects is covered in two places: (a) the guidance in Section 4 of MRP-227 provides an introductory discussion and justification of the examination The inspection methods selected for detecting the aging effects of interest; and (b) standards for examinationare defined and established in Section 4 of MRP-227-A. Standards for implementing the inspection methods, procedures, are defined and personnel are providedestablished in a companion document, MRP-228. In all cases, well-established inspection methods are were selected. These methods include volumetric UT examination methods for detecting flaws in bolting, physical measurements for detecting changes in dimension, and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from general conditions to detection and sizing of surface-breaking discontinuities.

Surface examinations may also be used as an alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, and fatigue is monitored/inspected by either VT-1 or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). The VT-

-3 visual methods may be applied for the detection of cracking in non-redundant RVI components only when the flaw tolerance of the component or affected assembly, as evaluated for reduced fracture toughness properties, is known and the component has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. VT-3 visual methods are acceptable for the detection of cracking in redundant RVI components (e.g.,

redundant bolts or pins used to secure a fastened RVI assembly).

In addition, VT-3 examinations are used to monitor/inspect for loss of material induced by wear and for general aging conditions, such as gross distortion caused by void swelling and irradiation growth or by gross effects of loss of preload caused by thermal and irradiation-enhanced stress relaxation and creep.

In addition, theThe program adopts the recommended guidance in MRP-227-A for defining the Expansion criteriaCriteria that need to be applied to the inspection findings of Primary Components and Existing Requirement Components components and for expanding the examinations to include additional Expansion Components. As a result, components. RVI component inspections performed on the RVI components are performed consistent with the inspection frequency and sampling bases for Primary Components, components, Existing Requirement ComponentsPrograms components, and Expansion Components components in MRP-227-A, which have been demonstrated to be in conformance with the inspection criteria, sampling basis criteria, and sample Expansion criteria in Section A.1.2.3.4 of NRC Branch Position RLSB-1.

Specifically, the program implements the parameters monitored/inspected criteria and bases for inspecting the relevant parameter conditions for [as an administrative action item for the AMP, applicant is to select one of the following to finish the sentence, as applicable to its NSSS vendor for its internals: B&W designed Primary Components in Table 4-1 of MRP-227; CE designed Primary Components in Table 4-2 of MRP-227; or Westinghouse designed Primary Components in Table 4-3 of MRP-227] and for [as an administrative action item for the AMP, applicant is to select one of the following to finish the sentence, as applicable to its NSSS vendor for its internals: for B&W designed Expansion Components in Table 4-4 of MRP-227; for CE designed expansion components in Table 4-5 of MRP-227; and for Westinghouse designed Expansion Components in Table 4-6 of MRP-227].

The program is supplemented by the following plant-specific Primary Component and Expansion Component inspections for the program (as applicable): [As a relevant license renewal applicant action item, the applicant is to list (using criteria in MRP-227) each B-6

additional RVI component that needs to be inspected as an additional plant-specific Primary Component for the applicants program and each additional RVI component that needs to be inspected as an additional plant-specific Expansion Component for the applicants program.

For each plant specific component added as an additional primary or Expansion Component, the list should include the applicable aging effects that will be monitored for, the inspection method or methods used for monitoring, and the sample size and frequencies for the examinations].

In addition, in some cases (as defined in MRP-227-A), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for changes in dimensions due to void swelling, deflection or distortion. The physical measurements methods applied in accordance with this program include [Applicant to input physical measure methods identified by the MRP in response to NRC RAI No. 11 in the NRCs Request for Additional Information to Mr. Christen B. Larson, EPRI MRP on Topical Report MRP-227 dated November 12, 2009].

Inspection coverages for Primary and Expansion RVI components are implemented consistent with Sections 3.3.1 and 3.3.2 of the NRC SE, Revision 1, on MRP-227.

5. Monitoring and Trending: The methods for monitoring, recording, evaluating, and trending the data that result from the programs inspections are given in Section 6 of MRP-227-A and its subsections. The Flaw evaluation methods include, including recommendations for flaw depth sizing and for crack growth determinations as well as for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications., are defined in MRP-227-A. The examinationsexamination and re-examinations required by thethat are implemented in accordance with MRP-227 guidance-A, together with the requirementscriteria specified in MRP-228 for inspection methodologies, inspection procedures, and inspection personnel, provide timely detection, reporting, and corrective actions with respect to the effects of the age-related degradation mechanisms within the scope of the program. The extent of the examinations, beginning with the sample of susceptible PWR internals component locations identified as Primary Component locations, with the potential for inclusion of Expansion Component locations if the effects are greater than anticipated, plus the continuation of the Existing Programs activities, such as the ASME Code,Section XI, Examination Category B-N-3 examinations for core support structures, provides a high degree of confidence in the totalfor timely detection, reporting, and implementation of corrective actions for the aging effects and mechanisms managed by the program.

The program applies applicable fracture toughness properties, including reductions for thermal aging or neutron embrittlement, in the flaw evaluations of the components in cases where cracking is detected in a RVI component and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation.

For singly-represented components, the program includes criteria to evaluate the aging effects in the inaccessible portions of the components and the resulting impact on the intended function(s) of the components. For redundant components (such as redundant bolts, screws, pins, keys, or fasteners, some of which are accessible to inspection and some of which are not accessible to inspection), the program includes criteria to evaluate the aging effects in the population of components that are inaccessible to the applicable inspection technique and the resulting impact on the intended function(s) of the assembly containing the components.

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6. Acceptance Criteria: Section 5 of MRP-227-A, which includes Table 5-1 for B&W-designed RVIs, Table 5-2 for CE-designed RVIs, and Table 5-3 for Westinghouse-designed RVIs, provides the specific examination and flaw evaluation acceptance criteria for the Primary and Expansion Component examinations. For RVI component examination methods. For RVI components addressed by examinations referenced to performed in accordance with the ASME Code,Section XI, the IWB-3500 acceptance criteria apply.in IWB-3500 are applicable. For otherRVI components covered by other Existing Programs,, the examination acceptance criteria are described within the Existing Programapplicable reference document.

The guidance in MRP-227 contains three types of examination As applicable, the program establishes acceptance criteria:

  • For visual examination (and surface examination as an alternative to visual examination),

the examination acceptance criterion is the absence of any of the specific, descriptive relevant conditions; in addition, there are requirements to record and disposition surface breaking indications that are detected and sized for length by VT-1/EVT-1 examinations;

  • For volumetric examination, the examination acceptance criterion is the capability for reliable detection of indications in bolting, as demonstrated in the examination Technical Justification; in addition, there are requirements for system-level assessment of bolted or pinned assemblies with unacceptable volumetric (UT) examination indications that exceed specified limits; and For physical measurements, the examination acceptance criterion for the acceptable tolerance in the measured differential height from the top of the plenum rib pads to the vessel seating surface in B&W plants are given in Table 5-1 of MRP-227. The acceptance criterion for physical measurements performed on the height limits of the Westinghouse-designed hold-down springs are [The incorporation of this sentence is a license renewal applicant action item for Westinghouse PWR applicants only - insert the applicable sentence incorporating the specified any physical measurement criteria only if the applicants facility is based on a Westinghouse NSSS design: the Westinghouse applicant is to incorporate the applicable language and then specify the fit up limits on the hold down springs, as established on a plant-specific basis for the design of the hold-down springs at the applicants Westinghouse-designed facility].monitoring methods that are credited for aging management of particular RVI components.
7. Corrective Actions: Corrective actions following the detection of unacceptable conditions are fundamentally provided for in each plants corrective action program. Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection.

The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. Examples of methodologies that can be used to analytically disposition unacceptable conditions are found in the ASME Code,Section XI or in Section 6 of MRP-227. Section 6 of MRP-227 describes the options that are available for disposition of detected conditions that exceed the examination acceptance criteria of Section 5 of the report. These include engineering evaluation methods, as well as supplementary examinations to further characterize the detected condition, or the alternative of component repair and replacement procedures. The latter are subject to the requirements of the ASME Code,Section XI. The implementation of the guidance in MRP-227 The implementation of the guidance in MRP-227-A, plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in B-8

accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

Other alternative corrective actionactions bases may be used to disposition relevant conditions if they have been previously approved or endorsed by the NRC. Examples of previously NRC-endorsed alternative corrective actions bases include those corrective actions bases for Westinghouse-design RVI components that are defined in Tables 4-1, 4-2, 4-3, 4-4, 4-5, 4-6, 4-7 and 4-8 of Westinghouse Report No. WCAP-14577-Rev. 1-A, or for B&W-designed RVI components in B&W Report No. BAW-2248. Westinghouse Report No. WCAP-14577-Rev. 1-A was endorsed for use in an NRC SE to the Westinghouse Owners Group, dated February 10, 2001. B&W Report No. BAW-2248 was endorsed for use in an SE to Framatome Technologies on behalf of the B&W Owners Group, dated December 9, 1999. Alternative corrective action basesactions not approved or endorsed by the NRC will be submitted for NRC approval prior to their implementation.

8. Confirmation Process: Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the recommendations of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable. It is expected that the The implementation of the guidance in MRP-227 will provide-A, in conjunction with NEI 03-08 and other guidance documents, reports, or methodologies referenced in this AMP, provides an acceptable level of quality and an acceptable basis for inspectionconfirming the quality of inspections, flaw evaluation, and other elements of aging management of the PWR internals that are addressed in accordance with the 10 CFR Part 50, Appendix B, or their equivalent (as applicable), confirmation process, and administrative controlsevaluations, and corrective actions.
9. Administrative Controls: The administrative controls for suchthese types of programs, including their implementing procedures and review and approval processes, are implemented in accordance with the recommended industry guidelines and criteria in NEI 03-08, and are under existing site 10 CFR 50 Appendix B, Quality Assurance Programs, or their equivalent, as applicable. Such The evaluation in Section 3.5 of the NRCs SE, Revision1, on MRP-227 provides the basis for endorsing NEI 03-08. This includes endorsement of the criteria in NEI-03-08 for notifying the NRC of any deviation from the I&E methodology in MRP-227-A and justifying the deviation no later than 45 days after its approval by a program is thus expected to be established with a sufficient level of documentation and administrative controls to ensure effective long-term implementationlicensee executive.
10. Operating Experience: Relatively few incidents of PWR internals aging degradation have been reported in operating U.S. commercial PWR plants. A summary of observations to date is provided in Appendix A of MRP-227-A. The applicant is expected to review subsequentand assessment of relevant operating experience for impactits impacts on itsthe program or to participate in industry initiatives that perform this function.

The application of the MRP-227 guidance will establish a considerable amount of , including implementing procedures, are governed by NEI 03-08 and Appendix A of MRP-227-A.

Consistent with MRP-227-A, the reporting of inspection results and operating experience over the next few years. Section 7 of MRP-227 describes the reporting requirements for these applications, and the plan for evaluating the accumulated additional operating experienceis treated as a Needed category item under the implementation of NEI 03-08.

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The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry operating experience, as discussed in Appendix B of the GALL Report, which is documented in LR-ISG-2011-05.

References 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants, Office of the Federal Register, National Archives and Records Administration, 20092011.

10 CFR Part 50.55a, Codes and Standards, Office of the Federal Register, National Archives and Records Administration, 2009.2011.

ASME Boiler & Pressure Vessel Code,Section V, Nondestructive Examination, 2004 Edition, American Society of Mechanical Engineers, New York, NY.

ASME Boiler & Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, The ASME Boiler and Pressure Vessel Code, 2004 edition as approved in 10 CFR 50.55a, The American Society of Mechanical Engineers, New York, NY.

B&W Report No. BAW-2248, Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, Framatome Technologies (now AREVA Technologies), Lynchburg VA, July 1997. (NRC Microfiche Accession Number A0076, Microfiche Pages 001 - 108).

EPRI 1014986, PWR Primary Water Chemistry Guidelines, Volume 1, Revision 6, Electric Power Research Institute, Palo Alto, CA, December 2007. (Non-publicly available ADAMS Accession Number ML081140278). The non-proprietary version of the report may accessed by members of the public at ADAMS Accession Number ML081230449 EPRI 1016596, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, -Revision. 0), Electric Power Research Institute, Palo Alto, CA: 2008.

EPRI Technical Report No. 1022863, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A), December 2011, ADAMS Accession No. ML12017A193 (Transmittal letter from the EPRI-MRP) and ADAMS Accession Nos. ML12017A194, ML12017A196, ML12017A197, ML12017A191, ML12017A192, ML12017A195 and ML12017A199, (Final Report).

EPRI Technical Report No. 1016609, Materials Reliability Program: Inspection Standard for PWR Internals (MRP-228), Electric Power Research Institute, Palo Alto, CA, July 2009.

(Non-publicly available ADAMS Accession No.umber ML092120574). The non-proprietary version of the report may accessed by members of the public at ADAMS Accession No.umber ML092750569.

NRC RAI No. 11 in the NRCs Request for Additional Information to the Mr. Christen B. Larson, EPRI MRP on Topical Report MRP-227 dated November 12, 2009.

NRC Safety Evaluation from C. I. Grimes [NRC] to R. A, Newton [Chairman, Westinghouse Owners Group], Acceptance for Referencing of Generic License Renewal Program Topical Report Entitled License Renewal Evaluation: Aging Management for Reactor Internals, WCAP-14577, Revision 1, February 10, 2001. (ADAMS Accession Number ML010430375).

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NRC Safety Evaluation from C. I. Grimes [NRC] to W. R. Gray [Framatome Technologies],

Acceptance for Referencing of Generic License Renewal Program Topical Report Entitled Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, February 10, 2001. (ADAMS Accession Number ML993490288).

NUREG-1800, Revision 2, Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants, Appendix A.1, Aging Management Review - Generic (Branch Technical Position RLSB-1), U.S. Nuclear Regulatory Commission, Washington, DC, 2010.

Westinghouse Non-Proprietary Class 3 Report No. WCAP-14577-Rev. 1-A, License Renewal Evaluation: Aging Management for Reactor Internals, Westinghouse Electric Company, Pittsburgh, PA [March 2001]. Report was submitted to the NRC Document Control Desk in a letter dated April 9, 2001. (ADAMS Accession Number ML011080790).

NRC Interim Staff Guidance LR-ISG-2011-05, Ongoing Review Of Operating Experience, March 16, 2012, (ADAMS Accession No. ML12044A215).

Nuclear Energy Institute (NEI) Report No. 03-08, Revision 2, Guideline for the Management of Materials Issues, ADAMS Accession No. ML101050334).

NRC Safety Evaluation from Robert A. Nelson (NRC) to Neil Wilmshurst (EPRI), Revision 1 to the Final Safety Evaluation of Electric Power Research Institute (EPRI) Report, Materials Reliability Program (MRP) Report 1016596 (MRP-227), Revision 0, Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, December 16, 2011, ADAMS Accession No. ML11308A770.

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(2) Mark-up of changes to GALL Report Chapter IV.B2 B2. REACTOR VESSEL INTERNALS (PWR) - WESTINGHOUSE Systems, Structures, and Components This section addresses the Westinghouse pressurized -water reactor (PWR) vessel internals and consists of, which consist of components in the upper internals assembly, the control rod guide tube assembliesassembly, the core barrel assembly, the baffle/former assembly, the lower internal assembly, and theinternals assembly, lower support assembly, thermal shield assembly, bottom mounted instrumentation support structures. Based on Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants, all structures and components that comprise the reactor vessel are governed by Group A or B Quality Standards.system, and alignment and interfacing components.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (IV.A2).

Inspection Plan An applicant will submit an inspection plan for reactor internals to the NRC for review and approval with the application for license renewal in accordance with Chapter XI.M16A, PWR Vessel Internals.

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-300 IV.B2-33 Alignment and Stainless steel Reactor coolant and Loss of preload Chapter XI.M16A, PWR Vessel No (R-108) interfacing components: neutron flux due to thermal and Internals internals hold down irradiation enhanced Primary components (identified in spring stress relaxation; changes the "Structure and Components" in dimensions column) due to void swelling or no Expansion components distortion; loss of material due to wear IV.B2.RP-301 IV.B2-40 Alignment and Stainless steel Reactor coolant and Cracking 'Chapter XI.M2, Water Chemistry, No (R-112) interfacing components: neutron flux due to stress -corrosion and upper core plate cracking Chapter XI.M16A, PWR Vessel alignment pins Internals and Chapter XI.M2, Water Chemistry Existing Program components (identified in the "Structure and Components" column) no Expansion components IV.B2.RP-299 IV.B2-34 Alignment and Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No (R-115) interfacing components: neutron flux due to wear Internals upper core plate Existing Program components alignment pins (identified in the "Structure and Components" column) no Expansion components IV.B2.RP-271 IV.B2-10 Baffle-to-former Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-125) assembly: accessible neutron flux due to irradiation-assisted and baffle-to-former bolts stress -corrosion cracking Chapter XI.M16A, PWR Vessel andor fatigue Internals Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, Water Chemistry (for Expansion components see AMR Items IV.B2.RP-273 and IV.B2.RP-286)SCC mechanisms only)

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-272 IV.B2-6 Baffle-to-former Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-128) assembly: accessible neutron flux toughness Internals baffle-to-former bolts due to neutron irradiation Primary components (identified in embrittlement; the "Structure and Components" changechanges in column) dimensions (for Expansion components see due to void swelling or AMR Items IV.B2.RP-274 and distortion; IV.B2.RP-287) loss of preload due to thermal and irradiation enhanced stress relaxation or creep IV.B2.RP-270 IV.B2-1 Baffle-to-former Stainless steel Reactor coolant and Changes in dimensions Chapter XI.M16A, PWR Vessel No (R-124) assembly: baffle and neutron flux due to void swelling or Internals former plates distortion Primary components (identified in the "Structure and Components" column) no Expansion components IV.B2.RP-270a IV.B2-1 Baffle-to-former Stainless steel Reactor coolant and Cracking due to Chapter XI.M16A, PWR Vessel No (R-124) assembly: baffle and neutron flux irradiation-assisted stress Internals and Chapter XI.M2, former plates corrosion cracking Water Chemistry IV.B2.RP-275 IV.B2-6 Baffle-to-former Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-128) assembly: baffle-edge neutron flux due to irradiation-assisted and bolts (all plants with stress -corrosion cracking Chapter XI.M16A, PWR Vessel baffle-edge bolts) andor fatigue Internals and Chapter XI.M2, Water Chemistry (for SCC mechanisms only)

Primary components (identified in the "Structure and Components" column) no Expansion components B-14

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-354 Baffle-to-former Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No assembly: baffle-edge neutron flux toughness Internals bolts (all plants with due to neutron irradiation Primary components (identified in baffle-edge bolts) embrittlement; the "Structure and Components" changes in dimensions column) due to void swelling or no Expansion components distortion; loss of preload due to thermal and irradiation enhanced stress relaxation or creep IV.B2.RP-273 IV.B2-10 Baffle-to-former Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-125) assembly: barrel-to- neutron flux due to irradiation-assisted and former bolts stress -corrosion cracking Chapter XI.M16A, PWR Vessel andor fatigue Internals Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, Water Chemistry (for Primary components see AMR Item IV.B2.RP-271)SCC mechanisms only)

IV.B2.RP-274 IV.B2-6 Baffle-to-former Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-128) assembly: barrel-to- neutron flux toughness Internals former bolts due to neutron irradiation Expansion components (identified embrittlement; in the "Structure and Components" changes in dimensions column) due to void swelling or (for Primary components see AMR distortion; Item IV.B2.RP-272) loss of preload due to thermal and irradiation enhanced stress relaxation or creep B-15

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-284 IV.B2-12 Bottom mounted Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No (R-143) instrument system: flux (with or without neutron flux due to wear Internals thimble tubes chrome plating) Existing Program components (identified in the "Structure and Components" column)

No expansion components; and or Chapter XI.M37, ",Flux Thimble Tube Inspection" IV.B2.RP-293 IV.B2-24 Bottom-mounted Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-138) instrumentation system: neutron flux due to fatigue and bottom-mounted Chapter XI.M16A, PWR Vessel instrumentation (BMI) Internals column bodies Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-298)

IV.B2.RP-292 IV.B2-21(R- Bottom-mounted Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No 140) instrumentation system: neutron flux toughness Internals bottom-mounted due to neutron irradiation Expansion components (identified instrumentationinstrume embrittlement in the "Structure and Components" nt (BMI) column bodies column)

(for Primary components see AMR Item IV.B2.RP-297)

IV.B2.RP-296 Control rod guide tube Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No (CRGT) assemblies: neutron flux due to wear Internals CRGT guide plates Primary Components (identified in (cards) the "Structure and Components" column) (for Expansion components see AMR Line Item IV.B2.RP-386)

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-298 IV.B2-28 Control rod guide tube Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-118) (CRGT) assemblies: neutron flux due to stress -corrosion and CRGT lower flange cracking andor fatigue Chapter XI.M16A, PWR Vessel welds (accessible) Internals Primary components (identified in the "Structure and Components" column) and Chapter XI.M2, Water Chemistry (for Expansion components see AMR Items IV.B2.RP-291 and IV.B2.RP-293)SCC mechanisms only)

IV.B2.RP-297 Control rod guide tube Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (CRGT) assemblies: (including CASS) neutron flux toughness Internals CRGT lower flange due to thermal aging and Primary components (identified in welds (accessible) neutron irradiation the "Structure and Components" embrittlement and for column)

CASS, due to thermal (for Expansion components see aging embrittlement AMR Items IV.B2.RP-290 and IV.B2.RP-292)

IV.B2.RP-386 Control rod guide tube Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No (CRGT) assemblies: C- neutron flux due to wear Internals tubes and sheaths Expansion components (identified in the "Structure and Components" column) are only the components associated with a primary component that exceeded the acceptance limit.

(for Primary components see AMR Item IV.B2.RP-296)

IV.B2.RP-355 Control rod guide tube NickelStainless Reactor coolant and Cracking A plant-specific aging management Yes, plant-IV.B2.RP-355 (CRGT) assemblies: steel; nickel alloy neutron flux due to stress -corrosion program is to be evaluatedChapter specificNo guide tube support pins cracking andor fatigue XI.M16A, PWR Vessel Internals (split pins) and Chapter XI.M2, Water Chemistry (for SCC mechanisms only)

B-17

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-356 Control rod guide tube NickelStainless Reactor coolant and Loss of material A plant-specific aging management Yes, plant-(CRGT) assemblies: steel; nickel alloy neutron flux due to wear program is to be evaluatedChapter specificNo guide tube support pins XI.M16A, PWR Vessel Internals (split pins)

IV.B2.RP-387 Core barrel assembly: Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No upper core barrel neutron flux due to stress -corrosion and axialand lower core cracking, and or Chapter XI.M16A, PWR Vessel barrel circumferential irradiation-assisted stress Internals (girth) welds -corrosion cracking or Expansion components (identified fatigue in the "Structure and Components" column) and Chapter XI.M2, Water Chemistry (for Primary components see AMR Item IV.B2.RP-276SCC mechanisms only)

IV.B2.RP-387a Core barrel assembly: Stainless steel Reactor coolant and Cracking due to stress Chapter XI.M16A, PWR Vessel No upper core barrel and neutron flux corrosion cracking or Internals and Chapter XI.M2, lower core barrel vertical irradiation-assisted stress Water Chemistry (for SCC (axial) welds corrosion cracking or mechanisms only) fatigue IV.B2.RP-388 Core barrel assembly: Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No upper core barrel neutron flux toughness Internals axialand lower core due to neutron irradiation Expansion components (identified barrel circumferential embrittlement in the "Structure and Components" (girth) welds column)

(for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP- IV.B2-8(R- Core barrel assembly: Stainless steel Reactor coolant and CrackingLoss of fracture Chapter XI.M2, Water Chemistry, No 282388a 120) upper core barrel neutron flux toughness and flangeand lower core due to stress corrosion Chapter XI.M16A, PWR Vessel barrel vertical (axial) cracking and Internals welds fatigueneutron irradiation Expansion components (identified embrittlement in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-276)

B-18

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-345 Core barrel assembly: Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No core barrel flange neutron flux due to wear Internals Existing Program components (identified in the "Structure and Components" column) no Expansion components IV.B2.RP-278 IV.B2-8 Core barrel assembly: Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-120) core barrel outlet nozzle neutron flux due to stress corrosion and welds cracking and fatigue Chapter XI.M16A, PWR Vessel Cracking Internals due to stress Expansion component (identified in corrosion the "Structure and Components" cracking column) or fatigue and Chapter XI.M2, Water Chemistry (for Primary components see AMR Item IV.B2.RP-276)SCC mechanisms only)

IV.B2.RP- IV.B2-8(R- Core barrel assembly: Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No 280278a 120) lower core barrel flange neutron flux due to stress corrosion and weldoutlet nozzle welds cracking and irradiation- Chapter XI.M16A, PWR Vessel assisted stress corrosion Internals cracking Expansion component (identified in Loss of fracture the "Structure and Components" toughness column) due to neutron (for Primary components see AMR irradiation Item IV.B2.RP-276) embrittlement IV.B2.RP-281280 IV.B2-98 Core barrel assembly: Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-122120) lower core barrel flange neutron flux toughnessCracking Internals weld due to neutron irradiation Expansion Components (identified embrittlementstress in the "Structure and Components" corrosion cracking or column) fatigue Chapter XI.M2, Water Chemistry (for Primary components see AMR Item IV.B2.RP-276)SCC mechanisms only)

B-19

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-276 IV.B2-8 Core barrel assembly: Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-120) upper core barrel flange neutron flux due to stress -corrosion and weld cracking and irradiation- Chapter XI.M16A, PWR Vessel assisted stress corrosion Internals and Chapter XI.M2, cracking Water Chemistry Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B2.RP-278, IV.B2.RP-280, IV.B2.RP-282, IV.B2.RP-294, IV.B2.RP-295,IV.B2.RP-281, IV.B2.RP-387, and IV.B2.RP-388)

IV.B2.RP-285 IV.B2-14 Lower internals Nickel alloy Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel No (R-137) assembly: clevis insert neutron flux due to weardue to wear; Internals bolts or screws loss of preload due to Existing Program components thermal and irradiation (identified in the "Structure and enhanced stress Components" column) relaxation or creep no Expansion components IV.B2.RP-399 Lower internals Stainless steel; Reactor coolant and Cracking due to primary Chapter XI.M16A, PWR Vessel No assembly: clevis insert nickel alloy neutron flux water stress corrosion Internals and Chapter XI.M2, bolts or screws cracking, irradiation- Water Chemistry (for SCC assisted stress corrosion mechanisms only) cracking or fatigue IV.B2.RP-289 IV.B2-20 Lower internals Stainless steel Reactor coolant and Cracking 'Chapter XI.M2, Water Chemistry, No (R-130) assembly: lower core neutron flux due to irradiation-assisted and plate and extra-long (XL) stress -corrosion cracking, Chapter XI.M16A, PWR Vessel lower core plate and or fatigue Internals and Chapter XI.M2, Water Chemistry (for SCC mechanisms only)

Existing Program components (identified in the "Structure and Components" column) no Expansion components B-20

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-288 IV.B2-18 Lower internals Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-132) assembly: lower core neutron flux toughness Internals plate and extra-long (XL) due to neutron irradiation Existing Program components lower core plate embrittlement; (identified in the "Structure and loss of material Components" column) due to wear no Expansion components IV.B2.RP-291 IV.B2-24 Lower support Cast austenitic Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-138) assembly: lower support stainless steel neutron flux due to irradiation-assisted and column bodies (cast) stress -corrosion cracking Chapter XI.M16A, PWR Vessel Internals and Chapter XI.M2, Water Chemistry Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-298)

IV.B2.RP-290 IV.B2-21 Lower support Cast austenitic Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-140) assembly: lower support stainless steel neutron flux toughness Internals column bodies (cast) due to thermal aging and Expansion components (identified neutron irradiation in the "Structure and Components" embrittlement column)

(for Primary components see AMR Item IV.B2.RP-297)

IV.B2.RP-291a Lower support Stainless steel Reactor coolant and Cracking due to stress Chapter XI.M16A, PWR Vessel No assembly: lower support neutron flux corrosion cracking or Internals and Chapter XI.M2, forging or casting fatigue Water Chemistry (for SCC mechanisms only)

IV.B2.RP-290a Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No assembly: lower support neutron flux toughness due to neutron Internals forging or casting irradiation embrittlement (and thermal aging embrittlement for CASS, PH SS, and martensitic SS)

B-21

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-294 IV.B2-24 Lower support Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-138) assembly: lower support neutron flux due to irradiation-assisted and column bodies (non- stress -corrosion cracking Chapter XI.M16A, PWR Vessel cast) Internals and Chapter XI.M2, Water Chemistry Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP-295 IV.B2-22(R- Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No 141) assembly: lower support neutron flux toughness Internals column bodies (non- due to neutron irradiation Expansion Components (identified cast) embrittlement in the "Structure and Components" column)

(for Primary components see AMR Item IV.B2.RP-276)

IV.B2.RP-286 IV.B2-16 Lower support Stainless steel; Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, No (R-133) assembly: lower support nickel alloy neutron flux due to irradiation-assisted and column bolts stress -corrosion cracking Chapter XI.M16A, PWR Vessel andor fatigue Internals Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, Water Chemistry (for Primary components see AMR Item IV.B2.RP-271)SCC mechanisms only)

IV.B2.RP-287 IV.B2-17 Lower support Stainless steel; Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No (R-135) assembly: lower support nickel alloy neutron flux toughness Internals column bolts due to neutron irradiation Expansion component (identified in embrittlement; the "Structure and Components" loss of preload column) due to thermal and (for Primary components see AMR irradiation enhanced Item IV.B2.RP-272) stress relaxation or creep B-22

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP-303 IV.B2-31 Reactor vessel internal Stainless steel; Reactor coolant and Cumulative fatigue Fatigue is a time-limited aging Yes, - TLAA (R-53) components nickel alloy neutron flux damage analysis (TLAA) to be evaluated for due to fatigue the period of extended operation.

See the SRP, Section 4.3 Metal Fatigue, for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

IV.B2.RP-24 IV.B2-32 Reactor vessel internal Stainless steel; Reactor coolant and Loss of material Chapter XI.M2, Water Chemistry No (RP-24) components nickel alloy neutron flux due to pitting and crevice corrosion IV.B2.RP-268382 IV.B2-26 Reactor vessel Stainless steel; Reactor coolant and Cracking Chapter XI.M2, Water Yes, if (R-142) internalinternals: ASME nickel alloy neutron flux due to fatigue, stress Chemistry,M1, ASME Section XI accessible Section XI, Examination corrosion cracking, andor Inservice Inspection, Subsections Primary, Category B-N-3 core irradiation- assisted stress IWB, IWC, and Expansion or support structure corrosion cracking; IWD or Chapter XI.M16A, PWR Existing components loss of material Vessel Internals , by invoking program (inaccessible due to wear applicable components locations)not already 10 CFR 50.55a and ASME Section indicate aging identified as Existing XI inservice inspection effects that need Programs components requirements managementNo in MRP-227-A)

IV.B2.RP-302 Thermal shield Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No assembly: thermal shield neutron flux due to fatigue Internals flexures B-23

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP- Reactor vessel internal Stainless steel; Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel Yes, if 269302a components nickel alloy neutron flux toughness Internals accessible (inaccessible due to neutron irradiation Primary, locations)Thermal shield embrittlement; Expansion or assembly: thermal shield change in dimension Existing flexures due to void swelling; program loss of preload components due to thermal and indicate aging irradiation enhanced effects that need stress relaxation; managementNo loss of material due to wear IV.B2.RP-265 Reactor internal No Stainless steel; Reactor coolant and No additional aging Chapter XI.M2, Water Chemistry, No Additional Measures nickel alloy neutron flux management for reactor and Chapter XI.M16A, PWR componentsReactor internal No Additional Vessel Internals Note:

vessel internal Measures components Components with no additional components with no unless required by ASME measures are not uniquely additional measures Section XI, Examination identified in GALL tables -

Category B-N-3 or Components with no additional relevant operating measures are defined in Section experience existsCracking 3.3.1 of MRP-227, "Materials due to stress corrosion Reliability Program: Pressurized cracking, and irradiation- Water Reactor Internals Inspection assisted stress corrosion and Evaluation Guidelines" cracking IV.B2.RP- Reactor vessel internal Stainless steel; Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel No 267291b components with no nickel alloy neutron flux toughness Internals additional measures Cracking due to neutron Note: Components with no Upper Internals irradiation embrittlement; additional measures are not Assembly; upper core change in dimension uniquely identified in GALL tables -

plate due to void swelling; Components with no additional loss of preload measures are defined in Section due to thermal and 3.3.1 of MRP-227, "Materials irradiation enhanced Reliability Program: Pressurized stress relaxationfatigue; Water Reactor Internals Inspection loss of material and Evaluation Guidelines" due to wear IV.B2.RP-382 IV.B2-26(R- Reactor vessel internals: Stainless steel; Reactor coolant and Cracking, or Chapter XI.M1, ASME Section XI No 142) core support structure nickel alloy; cast neutron flux Loss of material Inservice Inspection, Subsections austenitic stainless due to wear IWB, IWC, and IWD steel B-24

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 Reactor Vessel Internals (PWR) - Westinghouse Structure and/or Aging Effect/ Aging Management Program Further Item Link Material Environment Component Mechanism (AMP) Evaluation IV.B2.RP- Upper Internals Stainless steel Reactor coolant and Cracking Chapter XI.M16A, PWR Vessel No 302290b Assembly; upper core neutron flux due to fatigue; Internals plateThermal shield lossLoss of material Primary components (identified in assembly: thermal shield due to wear the "Structure and Components" flexures column) no Expansion components IV.B2.RP-346 Upper iInternals Stainless steel Reactor coolant and Cracking 'Chapter XI.M2, Water Chemistry, No aAssembly: upper neutron flux due to stress corrosion and support ring or skirt cracking and or fatigue Chapter XI.M16A, PWR Vessel Internals and Chapter XI.M2, Water Chemistry (for SCC mechanisms only)

Existing Program components (identified in the "Structure and Components" column) no Expansion components B-25

(3) Mark-up of changes to GALL Report Chapter IV.B3 B3. REACTOR VESSEL INTERNALS (PWR) - COMBUSTION ENGINEERING Systems, Structures, and Components This section addresses the Combustion Engineering (CE) pressurized -water reactor (PWR) vessel internals and consists of, which consist of components in the upper internals assembly, the control element assembly (CEA) shrouds,), the core support barrel assembly, the core shroud assembly, and the lower internal assembly. Based on Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants, all structures and components that comprise the reactor vessel are governed by Group A or B Quality Standards support structure assembly, and encore instrumentation (ICI) components.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (IV.A2).

Inspection Plan An applicant will submit an inspection plan for reactor internals to the NRC for review and approval with the application for license renewal in accordance with Chapter XI.M16A, PWR Vessel Internals.

B-26

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-312 IV.B3-2 Control Element Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and No (R-149) Assembly (CEA): neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals shroud assemblies: corrosion cracking Primary components (identified in the instrument guide andor fatigue "Structure and Components" column) tubes in peripheral and Chapter XI.M2, Water Chemistry (for CEA assemblies Expansion components see AMR Item IV.B3.RP-313)SCC mechanisms only)

IV.B3.RP-313 Control Element Stainless steel Reactor coolant and Cracking Chapter XI.M2, "Water Chemistry," and No Assembly (CEA): neutron flux due to stress - Chapter XI.M16A, "PWR Vessel Internals" shroud assemblies: corrosion cracking Expansion components (identified in the remaining instrument andor fatigue "Structure and Components" column) guide tubes in CEA and Chapter XI.M2, Water Chemistry (for assemblies Primary components see AMR Item IV.B3.RP-312)SCC mechanisms only)

IV.B3.RP-320 IV.B3-9 Core shroud Stainless steel Reactor coolant and Cracking 'Chapter XI.M2, Water Chemistry, and No (R-162) assemblies (all neutron flux due to fatigue Chapter XI.M16A, PWR Vessel Internals plants): guide lugs Existing Program components (identified in and; guide lug the "Structure and Components" column) insertinserts and no Expansion components bolts IV.B3.RP-319 IV.B3-9 Core shroud Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, "PWR Vessel Internals" No (R-162) assemblies (all neutron flux due to wear; loss of Existing Program components (identified in plants): guide lugs preload due to thermal the "Structure and Components" column) and; guide lug and irradiation no Expansion components insertinserts and enhanced stress bolts relaxation or creep B-27

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-358 Core shroud Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and No assemblies (for neutron flux due to irradiation- Chapter XI.M16A, PWR Vessel Internals bolted core shroud assisted stress - and Chapter XI.M2, Water Chemistry assemblies): corrosion cracking Expansion components (identified in the (a)assembly "Structure and Components" column) components, (for Primary component see AMR Item including shroud IV.B3.RP-314) plates and (b) former plates IV.B3.RP-318 IV.B3-8 Core shroud Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No (R-163) assemblies (for neutron flux toughness Primary components (identified in the bolted core shroud due to neutron "Structure and Components" column) assemblies): irradiation no Expansion components (a)assembly embrittlement; components, changes in dimensions including shroud due to void swelling or plates and (b) former distortion plates IV.B3.RP-316 IV.B3-9 Core shroud Stainless steel Reactor coolant and Cracking Chapter XI.M2, "Water Chemistry," and No (R-162) assemblies (for neutron flux due to irradiation- Chapter XI.M16A, "PWR Vessel Internals" bolted core shroud assisted stress - Expansion components (identified in the assemblies): barrel- corrosion cracking or "Structure and Components" column) shroud bolts with fatigue and Chapter XI.M2, Water Chemistry (for neutron exposures Primary components see AMR Item greater than 3 dpa IV.B3.RP-314)SCC mechanisms only)

IV.B3.RP-317 IV.B3-7 Core shroud Stainless steel; Reactor coolant and Loss of preload Chapter XI.M16A, "PWR Vessel Internals" No (R-165) assemblies (for nickel alloy neutron flux due to thermal and Expansion components (identified in the bolted core shroud irradiation enhanced "Structure and Components" column) assemblies): barrel- stress relaxation or (for Primary components see AMR Item shroud bolts with creep; IV.B3.RP-315) neutron exposures loss of fracture greater than 3 dpa toughness due to neutron irradiation embrittlement B-28

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-314 IV.B3-9 Core shroud Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and No (R-162) assemblies (for neutron flux due to irradiation- Chapter XI.M16A, PWR Vessel Internals bolted core shroud assisted stress - Primary components (identified in the assemblies): core corrosion cracking "Structure and Components" column) shroud bolts andor fatigue and Chapter XI.M2, Water Chemistry (for (accessible) Expansion components see AMR Items IV.B3.RP-316, IV.B3.RP-330, and IV.B3.RP-358)SCC mechanisms only)

IV.B3.RP-315 IV.B3-7(R- Core shroud Stainless steel Reactor coolant and Loss of preload Chapter XI.M16A, PWR Vessel Internals, No 165) assemblies (for neutron flux due to thermal and Primary components (identified in the bolted core shroud irradiation enhanced "Structure and Components" column) assemblies): core stress relaxation or (for Expansion components see AMR Items shroud bolts creep; IV.B3.RP-317, and IV.B3.RP-331)

(accessible) loss of fracture toughness due to neutron irradiation embrittlement; changes in dimensions due to void swelling or distortion IV.B3.RP-359 Core shroud Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel Internals, No assemblies (welded): neutron flux toughness Primary components (identified in the (assembly (designs due to neutron "Structure and Components" column) assembled in irradiation no Expansion components two vertical embrittlement; sections): core changes in dimensions shroud plates and (b) due to void swelling or plate-to-former distortion platesplate welds B-29

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-322 Core shroud Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and No assembly (for welded neutron flux due to irradiation- Chapter XI.M16A, PWR Vessel Internals core shroudsdesigns assisted stress - Primary components (identified in the assembled in two corrosion cracking "Structure and Components" column) vertical sections): (for Expansion components see AMR Item Core shroud plate- IV.B3.RP-323)and Chapter XI.M2, Water former plate weld (a) Chemistry The axial and horizontal weld seams at the core shroud re-entrant corners as visible from the core side of the shroud, within six inches of the central flange and horizontal stiffeners, and (b) the horizontal stiffeners in core shroud plate-to-former plate weldwelds IV.B3.RP-326 Core shroud Stainless steel Reactor coolant and Changes in Chapter XI.M16A, "PWR Vessel Internals" No assembly (for welded neutron flux dimensions Primary components (identified in the core shroudsdesigns due to void swelling or "Structure and Components" column) assembled in two distortion; loss of no Expansion components vertical fracture toughness due sections): gap to neutron irradiation betweenassembly embrittlement components, including monitoring of the upper and lower platesgap opening at the core shroud re-entrant corners B-30

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP- Core shroud Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and No 323326a assembly (for welded neutron flux due to irradiation- Chapter XI.M16A, PWR Vessel Internals core shroudsdesigns assisted stress - Expansion components (identified in the assembled in corrosion cracking or "Structure and Components" column) two vertical fatigue and Chapter XI.M2, Water Chemistry (for sections): remaining Primary components see AMR Item axial welds in IV.B3.RP-322)SCC mechanisms only) assembly components, including monitoring of the gap opening at the core shroud plate-to-former platere-entrant corners IV.B3.RP-324323 Core shroud Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and No assembly (for welded neutron flux due to irradiation- Chapter XI.M16A, PWR Vessel Internals core shrouds with assisted stress - Primary components (identified in the full-height shroud corrosion cracking "Structure and Components" column) plates): axial weld (for Expansion components see AMR Item seams at the core IV.B3.RP-325)and Chapter XI.M2, Water shroud re-entrant Chemistry corners, at the core mid-plane (+3 feet in height) as visible from the core side of the shroud Core shroud assembly (designs assembled in two vertical sections): remaining axial welds B-31

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP- Core shroud Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No 360359a assembly (for welded neutron flux toughness Primary components (identified in the core shrouds with due to neutron "Structure and Components" column) full-height shroud irradiation (for Expansion components see AMR Item plates): axial weld embrittlement; IV.B3.RP-361) seams at the core changes in dimensions shroud re-entrant due to void swelling or corners, at the core distortion mid-plane (+3 feet in height) as visible from the core side of the shroud Core shroud assembly (designs assembled in two vertical sections): remaining axial welds IV.B3.RP-325324 Core shroud Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and No assembly (for welded neutron flux due to irradiation- Chapter XI.M16A, PWR Vessel Internals core shrouds designs assisted stress - and Chapter XI.M2, Water Chemistry assembled corrosion cracking Expansion components (identified in the with full-height "Structure and Components" column) shroud plates): (for Primary components see AMR Item remainingshroud IV.B3.RP-324) plate axial welds, ribs, and ringsweld seams at the core shroud re-entrant corners B-32

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-361360 Core shroud Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No assembly (for welded neutron flux toughness Expansion components (identified in the core shrouds designs due to neutron "Structure and Components" column) assembled irradiation (for Primary components see AMR Item with full-height embrittlement IV.B3.RP-360) shroud plates):

remainingshroud plate axial welds, ribs, and ringsweld seams at the core shroud re-entrant corners IV.B3.RP-362325 Core support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No barrelshroud neutron flux toughnessCracking Expansion components (identified in the assembly: lower due to neutron "Structure and Components" cylinder (designs irradiation column)Chapter XI.M2, Water Chemistry assembled embrittlement-assisted (for Primary components see AMR Item with full-height stress corrosion IV.B3.RP-327) shroud plates): cracking remaining axial welds, ribs, and rings IV.B3.RP-329361 IV.B3-15(R- Core support Stainless steel Reactor coolant and CrackingLoss of Chapter XI.M2, Water Chemistry, and No 155) barrelshroud neutron flux fracture toughness Chapter XI.M16A, "PWR Vessel Internals" assembly: lower due to stress corrosion Expansion components (identified in the cylinder (designs crackingneutron "Structure and Components" column) assembled irradiation (for Primary components see AMR Item with full-height embrittlement IV.B3.RP-327) shroud plates):

remaining axial welds, ribs, and remaining core barrel assembly weldsrings IV.B3.RP-333362 Core support barrel Stainless steel Reactor coolant and CrackingLoss of Chapter XI.M2, Water Chemistry, and Yes, evaluate to assembly: lower neutron flux fracture toughness Chapter XI.M16A, "PWR Vessel Internals" determine the flange weld, if fatigue due to fatigueneutron Primary components (identified in the potential locations life cannot be irradiation "Structure and Components" column) and extent of demonstrated by embrittlement no Expansion components fatigue TLAAcylinder crackingNo circumferential (girth) welds B-33

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-389 Core support barrel Stainless steel Reactor coolant and Cumulative fatigue Chapter XI.M16A, PWR Vessel Internals Yes, TLAANo 362a assembly: lower neutron flux damage and Chapter XI.M2, Water Chemistry flange weld (if fatigue due to fatigueCracking analysis due to stress corrosion exists)cylinder cracking or irradiation-circumferential (girth) assisted stress welds corrosion cracking IV.B3.RP- IV.B3-15(R- Core support barrel Stainless steel Reactor coolant and CrackingLoss of Chapter XI.M2, Water Chemistry, and No 328362b 155) assembly: surfaces neutron flux fracture toughness Chapter XI.M16A, "PWR Vessel Internals of the lower core due to stress corrosion Primary components (identified in the barrel flange weld cracking and "Structure and Components" column)

(accessible fatigueneutron no Expansion components" surfaces)cylinder irradiation vertical (axial) welds embrittlement IV.B3.RP- IV.B3-17(R- Core support barrel Stainless steel Reactor coolant and Loss of material Chapter XI.M16A, PWR Vessel Internals No 332362c 156) assembly: upper core neutron flux due to wearCracking Existing Program components (identified in barrel flangelower due to stress corrosion the "Structure and Components" column) cylinder vertical cracking or irradiation- no Expansion componentsChapter XI.M2, (axial) welds assisted stress Water Chemistry corrosion cracking IV.B3.RP-327329 IV.B3-15(R- Core support barrel Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and No 155) assembly: upper neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals cylinder (base metal corrosion cracking Primary components (identified in the and welds) and upper "Structure and Components" column) core support barrel (for Expansion components see AMR Items flange weld IV.B3.RP-329, IV.B3.RP-335, IV.B3.RP-362, (accessible IV.B3.RP-363, IV.B3.RP-364)and Chapter surfaces)(flange base XI.M2, Water Chemistry metal)

IV.B3.RP-357333 Incore Zircaloy- Reactor coolant and Loss of A plant-specific aging management program Yes, plant-instrumentation (ICI): 4Stainless steel neutron flux materialCracking is to be evaluatedChapter XI.M16A, PWR specificNo ICI thimble tubes - due to wearstress Vessel Internals and Chapter XI.M2, Water lowerCore support corrosion cracking or Chemistry (for SCC mechanisms only) barrel assembly: fatigue lower flange B-34

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-336328 IV.B3- Lower support Stainless steel Reactor coolant and CrackingLoss of Chapter XI.M16A, PWR Vessel Internals No 2215(R- structure: A286 fuel neutron flux material Existing Program components (identified in 170)155) alignment pins (all due to wear; the "Structure and Components" column) plants with core loss of fracture no Expansion componentsChapter XI.M2, shroud assembled in toughness Water Chemistry (for SCC mechanisms two vertical due to neutron only) sections)Core irradiation support barrel embrittlement; loss of assembly: lower core preload barrel flange weld due to thermal and irradiation enhanced stress relaxationcorrosion cracking or fatigue IV.B3.RP-334332 IV.B3- Lower support Stainless steel Reactor coolant and CrackingLoss of 'Chapter XI.M2, Water Chemistry, and No 2317(R- structure: A286 fuel neutron flux material Chapter XI.M16A, PWR Vessel Internals 167)156) alignment pins (all due to irradiation- Existing Program components (identified in plants with core assisted stress the "Structure and Components" column) shroud assembled corrosion cracking and no Expansion components with full-height fatiguewear shroud plates)Core support barrel assembly: upper core barrel flange IV.B3.RP-364327 IV.B3-15(R- LowerCore support Cast austenitic Reactor coolant and Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No 155) structure:barrel stainlessStainless neutron flux toughnessCracking Expansion components (identified in the assembly: upper core steel due to neutron "Structure and Components" column) support columnbarrel irradiation and thermal (for Primary components see AMR Item flange weld embrittlementstress IV.B3RP-327)Chapter XI.M2, Water corrosion cracking Chemistry IV.B3.RP-363357 Lower support Stainless Reactor coolant and Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No structure: core steelZircaloy-4 neutron flux toughnessmaterial Expansion components (identified in the support columnIncore due to neutron "Structure and Components" column) instruments (ICI): ICI irradiation (for Primary components see AMR Item thimble tubes - lower embrittlementwear IV.B3RP-327)

B-35

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-330336 IV.B3- Lower support Stainless steel Reactor coolant and Loss of Chapter XI.M2, Water Chemistry, and No 2322(R- structure: core neutron flux materialCracking Chapter XI.M16A, PWR Vessel Internals 167)170) support column bolts due to wear; Expansion components (identified in the (designs assembled loss of fracture "Structure and Components" column) in two vertical toughness (for Primary components see AMR Item sections): fuel due to neutron 'IV.B3.RP-314) alignment pins irradiation-assisted embrittlement; loss of preload due to thermal and irradiation enhanced stress corrosion cracking and fatiguerelaxation or creep IV.B3.RP-331334 IV.B3-23(R- Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel Internals No 167) structure: core neutron flux toughnessCracking Expansion components (identified in the support column bolts due to neutronstress "Structure and Components" column)

(designs assembled corrosion cracking, Chapter XI.M2, Water Chemistry (for in two vertical irradiation Primary components see AMR Item sections or with full- embrittlement-assisted 'IV.B3.RP-315)SCC mechanisms only) height shroud plates): stress corrosion fuel alignment pins cracking, or fatigue IV.B3.RP- IV.B3- Lower support Stainless steel Reactor coolant and CrackingLoss of Chapter XI.M2, Water Chemistry, and No 335334a 2322(R- structure: core neutron flux material Chapter XI.M16A, PWR Vessel Internals 167)170) support column due to stress corrosion Expansion components (identified in the welds, applicable to cracking,wear; "Structure and Components" column) all plants except loss of fracture (for Primary components see AMR Item those (designs toughness IV.B3.RP-327) assembled in two due to neutron vertical sections or irradiation-assisted with full-height stress corrosion shroud plates): fuel cracking, alignment pins embrittlement; loss of preload due to thermal and fatigueirradiation enhanced stress relaxation or creep B-36

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-365364 Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No structure: (all plants): (including CASS) neutron flux toughness Primary component (identified in the core support due to neutron "Structure and Components" column) platecolumn welds irradiation no Expansion components embrittlement and for column welds made from CASS, thermal embrittlement IV.B3.RP-343363 Lower support Stainless steel Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and Yes, evaluate to structure: (all plants): neutron flux due to stress corrosion Chapter XI.M16A, PWR Vessel Internals determine the core support plate cracking, irradiation- and Chapter XI.M2, Water Chemistry (for potential locations (applicable to plants assisted stress SCC mechanisms only) and extent of with a core support corrosion cracking, or Primary components (identified in the fatigue plate), if fatigue life fatigue "Structure and Components" column) crackingNo cannot be no Expansion components demonstrated by TLAAcolumn welds IV.B3.RP-390330 IV.B3-23(R- Lower support Stainless steel Reactor coolant and CumulativeCracking Fatigue is a time-limited aging analysis Yes, TLAANo 167) structure: core neutron flux due to irradiation- (TLAA) to be evaluated for the period of support plate assisted stress extended operation. See the SRP, Section (applicable to plants corrosion cracking or 4.3 Metal Fatigue, for acceptable methods with a core support fatigue damage for meeting the requirements of 10 CFR plate), if fatigue due to fatigue 54.21(c)(1).Chapter XI.M16A, PWR Vessel analysis Internals and Chapter XI.M2, Water existscolumn bolts Chemistry (for SCC mechanisms only)

IV.B3.RP-342331 Lower support Stainless steel Reactor coolant and CrackingLoss of Chapter XI.M2, Water Chemistry, and No structure: deep neutron flux fracture toughness Chapter XI.M16A, PWR Vessel Internals beams (applicable due to stress corrosion Primary components (identified in the assemblies with full cracking,neutron "Structure and Components" column) height shroud irradiation-assisted no Expansion components plates)core support stress corrosion column bolts cracking, and fatigue embrittlement B-37

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-366335 IV.B3-23(R- Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel Internals No 167) structure: deep neutron flux toughnessCracking Primary components (identified in the beams (applicable due to neutron "Structure and Components" column) assemblies (designs irradiation no Expansion componentsChapter XI.M2, except those embrittlementstress Water Chemistry (for SCC mechanisms assembled with full - corrosion cracking or only) height shroud fatigue plates)):

lower core support beams IV.B3.RP-365 Lower support Stainless steel Reactor coolant and Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No structure (designs neutron flux toughness Primary component (identified in the with a core support due to neutron "Structure and Components" column) plate): core support irradiation no Expansion components plate embrittlement IV.B3.RP-24343 IV.B3- Reactor vessel Stainless steel; Reactor coolant and Loss of Chapter XI.M2, Water ChemistryM16A, No 25(RP-24) internal nickel alloy neutron flux materialCracking PWR Vessel Internals componentsLower due to pitting and support structure crevice (designs with a core corrosionfatigue support plate): core support plate IV.B3.RP-309342 Reactor vessel Stainless steel; Reactor coolant and Cracking Chapter XI.M2, Water Chemistry, and Yes, if accessible internal components nickel alloy neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals Primary, (inaccessible corrosion cracking, and Chapter XI.M2, Water Chemistry (for Expansion or locations)Lower and irradiation- SCC mechanisms only) Existing program support structure assisted stress - components (designs with core corrosion cracking, or indicate aging shrouds assembled fatigue effects that need with full height managementNo shroud plates): deep beams B-38

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-311366 Reactor vessel Stainless steel; Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel Internals Yes, if accessible internal components nickel alloy neutron flux toughness Primary, (inaccessible due to neutron Expansion or locations)Lower irradiation Existing program support structure embrittlement; components (designs with core change in dimension indicate aging shrouds assembled due to void swelling; effects that need with full height loss of preload managementNo shroud plates): deep due to thermal and beams irradiation enhanced stress relaxation; loss of material due to wear IV.B3.RP-339 IV.B3-24(R- Reactor vessel Stainless steel; Reactor coolant and Cumulative fatigue Fatigue is a time-limited aging analysis Yes, TLAA

53) internal components nickel alloy neutron flux damage (TLAA) to be evaluated for the period of due to fatigue extended operation. See the SRP, Section 4.3 Metal Fatigue, for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

IV.B3.RP-306 Reactor internal No Stainless steel; Reactor coolant and No additional aging Chapter XI.M2, Water Chemistry, and No Additional Measures nickel alloy neutron flux management for Chapter XI.M16A, PWR Vessel Internals componentsReactor reactor internal No Note: Components with no additional vessel internal Additional Measures measures are not uniquely identified in GALL components with no components unless tables - Components with no additional additional measures required by ASME measures are defined in Section 3.3.1 of Section XI, MRP-227, "Materials Reliability Program:

Examination Category Pressurized Water Reactor Internals B-N-3 or relevant Inspection and Evaluation Guidelines" operating experience existsCracking due to stress corrosion cracking, and irradiation-assisted stress corrosion cracking B-39

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-30724 IV.B3- Reactor vessel Stainless steel; Reactor coolant and Loss of fracture Chapter XI.M16A, PWR Vessel Internals No 25(RP-24) internal components nickel alloy neutron flux toughnessmaterial Note: Components with no additional with no additional due to neutron measures are not uniquely identified in GALL measures irradiation tables - Components with no additional embrittlement; measures are defined in Section 3.3.1 of change in dimension MRP-227, "Materials Reliability Program:

due to void swelling; Pressurized M2, Water Reactor Internals loss of preload Inspection and Evaluation due to thermal and Guidelines"Chemistry irradiation enhanced stress relaxation; loss of material due to wearpitting and crevice corrosion IV.B3.RP-382 IV.B3-22(R- Reactor vessel Stainless steel; Reactor coolant and Cracking Chapter XI.M1, ASME Section XI Inservice No 170) internals: ASME nickel alloy; cast neutron flux due to fatigue, stress Inspection, Subsections IWB, IWC, and Section XI, austenitic corrosion cracking, or IWD or Chapter XI.M16A, PWR Vessel Examination stainless steel irradiation-assisted Internals, by invoking applicable Category stress corrosion 10 CFR 50.55a and ASME Section XI B-N-3 core support cracking; inservice inspection requirements structure components Lossloss of material (not already identified due to wear as Existing Programs components in MRP-227-A)

IV.B3.RP-338 Upper internals Stainless steel Reactor coolant and Cracking 'Chapter XI.M2, Water Chemistry, and Yes, evaluate to assembly: fuel neutron flux due to fatigue Chapter XI.M16A, PWR Vessel Internals determine the alignment plate Primary components (identified in the potential locations (applicable to plants "Structure and Components" column) and extent of (designs with core no Expansion components fatigue shrouds assembled crackingNo with full height shroud plates), if fatigue life cannot be demonstrated by TLAA): fuel alignment plate B-40

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B3 Reactor Vessel Internals (PWR) - Combustion Engineering Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B3.RP-391400 Upper internals Stainless steel Reactor coolant and Cumulative fatigue Fatigue is a time-limited aging analysis Yes, TLAANo assembly: fuel neutron flux damage (TLAA) to be evaluated for the period of alignment plate due to fatigueCracking extended operation. See the SRP, Section (applicable to plants due to stress corrosion 4.3 Metal Fatigue, for acceptable methods with core shrouds cracking, irradiation- for meeting the requirements of 10 CFR assembled with full assisted stress 54.21(c)(1).Chapter XI.M16A, PWR Vessel height shroud plates), corrosion cracking or Internals and Chapter XI.M2, Water if fatigue analysis fatigue; loss of Chemistry (for SCC mechanisms only) exists Core Support material due to wear Barrel Assembly:

thermal shield positioning pins B-41

(4) Mark-up of changes to GALL Report Chapter IV.B4 B4. REACTOR VESSEL INTERNALS (PWR) - BABCOCK AND WILCOX Systems, Structures, and Components This section addresses the Babcock and Wilcox (B&W) pressurized -water reactor (PWR) vessel internals and consists, which consist of components in the plenum cover and plenum cylinderassembly, the upper grid assembly, the control rod guide tube (CRGT) assembly, the core support shield assembly, the core barrel assembly, the lower grid assembly, and the flow distributor assembly. Based on Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants, all structures and components that comprise the reactor vessel are governed by Group A or B Quality Standards.incore monitoring instrumentation (IMI) guide tube assembly, and the flow distributor assembly.

Common miscellaneous material/environment combinations where aging effects are not expected to degrade the ability of the structure or component to perform its intended function for the period of extended operation are included in IV.E.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (IV.A2).

Inspection Plan An applicant will submit an inspection plan for reactor internals to the NRC for review and approval with the application for license renewal in accordance with Chapter XI.M16A, PWR Vessel Internals.

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP-242 IV.B4-4 Control rod guide tube Cast austenitic Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-183) (CRGT) assembly: stainless steel and neutron flux toughness Expansion components (identified in the accessible surfaces at due to thermal aging "Structure and Components" column) four screw locations embrittlement (for Primary components see AMR Items (every 90 degrees) for IV.B4.RP-253 and IV.B4.RP-258)

CRGT spacer castings IV.B4.RP-242a Control rod guide tube Stainless steel Reactor coolant Cracking due to stress Chapter XI.M16A, PWR Vessel Internals No (CRGT) assembly: CRGT (including CASS) and neutron flux corrosion cracking or and Chapter XI.M2, Water Chemistry (for spacer castings fatigue SCC mechanisms only)

IV.B4.RP-245 IV.B4-13 Core barrel assembly: (a) Stainless steel; Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No (R-194) upper thermal shield nickelNickel alloy and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals bolts; (b) (applicable to corrosion cracking and Chapter XI.M2, Water Chemistry Crystal River Unit 3 or Expansion components (identified in the Davis Besse only): "Structure and Components" column) surveillance specimen (for Primary components see AMR Items holder tube bolts (Davis- IV.B4.RP-247 and IV.B4.RP-248)

Besse, only); (c) surveillance specimen tube holder(SSHT) studs, and /nuts (Crystal River Unit 3, only)or bolts IV.B4.RP-245a Core barrel assembly Nickel alloy Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No (applicable to Crystal and neutron flux fatigue River Unit 3 or Davis Besse only): surveillance specimen holder tube (SSHT) stud or bolt locking devices B-43

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP-245b Core barrel assembly Nickel alloy Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No (applicable to CR-3 or DB and neutron flux to wear; changes in only): surveillance dimensions due to specimen holder tube void swelling or (SSHT) stud or bolt distortion locking devices IV.B4.RP-247 IV.B4-13 Core barrel assembly: Stainless steel; Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No (R-194) accessible lower core nickel alloy and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals barrel (LCB) bolts and corrosion cracking and Chapter XI.M2, Water Chemistry locking devices Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.RP-245, IV.B4.RP-246, IV.B4.RP-254, and IV.B4.RP-256)

IV.B4.RP-247a Core barrel assembly: Stainless steel; Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No lower core barrel (LCB) nickel alloy and neutron flux fatigue bolt locking devices IV.B4.RP-247b Core barrel assembly: Stainless steel; Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No lower core barrel (LCB) nickel alloy and neutron flux to wear; changes in bolt locking devices dimensions due to void swelling or distortion IV.B4.RP-249 IV.B4-12 Core barrel assembly: Stainless steel Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-196) baffle plate accessible and neutron flux toughness Primary components (identified in the surfaces within one inch due to neutron "Structure and Components" column) around each baffle plate irradiation (for Expansion components see AMR Item flow and bolt holeplates embrittlement IV.B4.RP-250)

IV.B4.RP-249a Core barrel assembly: Stainless steel Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No baffle plates and neutron flux irradiation-assisted and Chapter XI.M2, Water Chemistry stress corrosion cracking B-44

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP-241 IV.B4-7 Core barrel assembly: Stainless steel Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-125) baffle-to-former bolts and and neutron flux due to stress and Chapter XI.M2, Water Chemistry (for screws corrosion cracking, SCC mechanisms only) irradiation-assisted stress corrosion cracking, fatigue, and overload IV.B4.RP- IV.B4-7(R- Core barrel assembly: Stainless steel Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No 241241a 125) baffle/former assembly: and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals (a) accessible baffle-to- corrosion cracking, Primary Components (identified in the former bolts and screws; irradiation-assisted "Structure and Components" column)

(b) accessible locking stress -corrosion and Chapter XI.M2, Water Chemistry (for devices (including locking cracking, fatigue, and Expansion components see AMR Items welds) of baffle-to-former overload IV.B4.RP-244 and IV.B4.RP-375)SCC bolts and internal baffle- mechanisms only) to-baffle bolts IV.B4.RP-240 IV.B4-1 Core barrel assembly: Stainless steel Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel No (R-128) baffle/former assembly: and neutron flux toughness Internals.

(a) accessible baffle-to- due to neutron Primary components (identified in the former bolts and screws; irradiation "Structure and Components" column)

(b) accessible locking embrittlement; (for Expansion components see AMR Item devices (including welds) loss of preload IV.B4.RP-243.)

of baffle-to-former bolts due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear IV.B4.RP-240a Core barrel assembly: Stainless steel Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No locking devices (including and neutron flux toughness locking welds) of baffle-to- due to neutron former bolts and internal irradiation baffle-to-baffle bolts embrittlement; loss of material due to wear B-45

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP-250 IV.B4-12 Core barrel assembly: Stainless steel Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-196) core barrel cylinder and neutron flux toughness Expansion components (identified in the (including vertical and due to neutron "Structure and Components" column) circumferential seam irradiation (for Primary components see AMR Item welds); former plates embrittlement IV.B4.RP-249)

IV.B4.RP-250a Core barrel assembly: Stainless steel Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No core barrel cylinder and neutron flux irradiation-assisted and Chapter XI.M2, Water Chemistry (for (including vertical and stress corrosion SCC mechanisms only) circumferential seam cracking or fatigue welds); former plates IV.B4.RP-375 Core barrel assembly: Stainless steel Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No internal baffle-to-baffle and neutron flux due to irradiation- Chapter XI.M16A, PWR Vessel Internals bolts assisted stress Expansion components (identified in the corrosion cracking, "Structure and Components" column) fatigue, or overload and Chapter XI.M2, Water Chemistry (for Primary components see AMR Item IV.B4.RP-241SCC mechanisms only)

IV.B4.RP-375a Core barrel assembly: Stainless steel Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No internal baffle-to-baffle and neutron flux toughness bolts due to neutron irradiation embrittlement; loss of preload due to thermal and irradiation enhanced stress relaxation or creep; loss of material due to wear IV.B4.RP-244 IV.B4-7 Core barrel assembly; Stainless steel Reactor coolant Cracking Chapter XI.M16A, PWR Vessel Internals No (R-125) external baffle-to-baffle and neutron flux due to irradiation- and Chapter XI.M2, Water Chemistry (for bolts and core barrel-to- assisted stress SCC mechanisms only) former bolts; corrosion cracking, fatigue, and overload B-46

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP- IV.B4-7(R- Core barrel assembly; (a) Stainless steel Reactor coolant Cracking Chapter XI.M2, "Water Chemistry," and No 244244a 125) external baffle-to-baffle and neutron flux due to irradiation- Chapter XI.M16A, "PWR Vessel Internals" bolts; (b) core barrel-to- assisted stress - Expansion components (identified in the former bolts; (c): locking corrosion cracking, or "Structure and Components" column) devices (including welds) fatigue and Chapter XI.M2, Water Chemistry of external baffle-to-baffle (for Primary components see AMR Item bolts and core barrel-to- IV.B4.RP-241)SCC mechanisms only) former bolts IV.B4.RP-243 IV.B4-1 Core barrel assembly; (a) Stainless steel Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No (R-128) external baffle-to-baffle and neutron flux toughness Expansion components (identified in the bolts; (b) core barrel-to- due to neutron "Structure and Components" column) former bolts; (c) locking irradiation (for Primary components see AMR Item devices (including welds) embrittlement; IV.B4.RP-240) of: external baffle-to-baffle loss of preload bolts and core barrel-to- due to thermal and former bolts; (d) internal irradiation enhanced baffle-to-baffle bolts stress relaxation or creep; loss of material due to wear IV.B4.RP-243a Core barrel assembly: Stainless steel Reactor coolant Loss of fracture Chapter XI.M16A, "PWR Vessel Internals" No locking devices (including and neutron flux toughness welds) of external baffle- due to neutron to-baffle bolts and core irradiation barrel-to-former bolts embrittlement; loss of material due to wear IV.B4.RP-248 IV.B4-12 Core support shield Stainless steel; Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No (R-196) (CSS) assembly: nickel alloy and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals accessible upper core corrosion cracking and Chapter XI.M2, Water Chemistry barrel (UCB) bolts and Primary components (identified in the locking devices "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.RP-245, IV.B4.RP-246, IV.B4.RP-254, IV.B4.RP-247, and IV.B4.RP-256)

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IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP-248a Core support shield Stainless steel; Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No (CSS) assembly: upper nickel alloy and neutron flux fatigue core barrel (UCB) bolt locking devices IV.B4.RP-248b Core support shield Stainless steel; Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No (CSS) assembly: upper nickel alloy and neutron flux to wear; changes in core barrel (UCB) bolt dimensions due to locking devices void swelling or distortion IV.B4.RP- IV.B4-2116 Core support shield Cast austenitic Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No 253252 (R-191)188) (CSS) assembly: (a) CSS stainlessStainless and neutron flux toughness Primary components (identified in the cast outlet nozzles steel, including due to thermal aging "Structure and Components" column)

(Oconee Unit 3 and CASS and PH embrittlement (for Expansion components see AMR Item Davis-Besse, only); (b) steels IV.B4.RP-242)

CSS vent valve discstop and bottom retaining rings (valve body components)

IV.B4.RP- IV.B4-16 Core support shield Stainless steel Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No 252252a (R-188) (CSS) assembly: (a) CSS and neutron flux toughness Primary components (identified in the vent valve disc shaft or Cracking due to "Structure and Components" column) hinge pin (b) thermal aging Chapter XI.M2, Water Chemistry (for CSS vent valve top embrittlementstress SCC mechanisms only) retaining ring (c) CSS corrosion cracking or No Expansion components vent valve and bottom fatigue retaining ringrings; vent valve locking devices (valve body components)

IV.B4.RP-251 IV.B4-15 Core support shield Stainless steel Reactor coolant Loss of material Chapter XI.M16A, PWR Vessel Internals No (R-190) (CSS) assembly: and neutron flux due to wear; CSS top flange loss of preload (wear)

B-48

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP- IV.B4-15 Core support shield Stainless steel Reactor coolant Loss of material Chapter XI.M16A, PWR Vessel Internals No 251251a (R-190) (CSS) assembly: CSS top and neutron flux due to wear; Primary component (identified in the flange; plenumPlenum loss of preload (wear) "Structure and Components" column) cover assembly: plenum No Expansion components cover weldment rib pads and plenum cover support flange IV.B4.RP-256 IV.B4-25 Flow distributor assembly: Stainless steel; Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No (R-210) flow distributor bolts and nickel alloy and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals, locking devices corrosion cracking Expansion components (identified in the "Structure and Components" column) and Chapter XI.M2, Water Chemistry (for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP-248)

IV.B4.RP-256a Flow distributor assembly: Stainless steel; Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No flow distributor bolt nickel alloy and neutron flux fatigue locking devices IV.B4.RP-256b Flow distributor assembly: Stainless steel; Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No flow distributor bolt nickel alloy and neutron flux to wear; changes in locking devices dimensions due to distortion or void swelling or distortion IV.B4.RP-259 IV.B4-31 Incore Monitoring Stainless steel; Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-205) InstrumentationInstrument nickel alloy and neutron flux toughness Primary components (identified in the (IMI) guide tube due to thermal aging, "Structure and Components" column) assembly: accessible top neutron irradiation (for Expansion components see Item surfaces of IMI guide tube embrittlement IV.B4.RP-260) spider-to-lower grid rib sectionssection welds IV.B4.RP-259a Incore Monitoring Stainless steel Reactor coolant Cracking due to stress Chapter XI.M16A, PWR Vessel Internals No.

Instrument (IMI) guide and neutron flux corrosion cracking, and Chapter XI.M2, Water Chemistry tube assembly: irradiation-assisted IMI guide tube spider-to- stress corrosion lower grid rib sections cracking or fatigue welds B-49

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP-258 IV.B4-4 Incore Monitoring Cast austenitic Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No (R-183) InstrumentationInstrument stainless steel and neutron flux toughness Primary components (identified in the (IMI) guide tube due to thermal aging, "Structure and Components" column) assembly: accessible top and neutron irradiation (for Expansion components see Item surfaces of IMI Incore embrittlement IV.B4.RP-242) guide tube spider spiders (castings )

IV.B4.RP-258a Incore Monitoring Stainless steel Reactor coolant Cracking due to stress Chapter XI.M16A, PWR Vessel Internals No Instrumentation (IMI) and neutron flux corrosion cracking, and Chapter XI.M2, Water Chemistry guide tube assembly: irradiation-assisted IMI guide tube spiders stress corrosion cracking or fatigue IV.B4.RP-254 IV.B4-25 Lower grid assembly: Nickel alloy Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No (R-210) alloy X-750 lower grid and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals, shock pad bolts and corrosion cracking Expansion components (identified in the locking devices (Three "Structure and Components" column) and Mile Island Unit -1, only) Chapter XI.M2, Water Chemistry (for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP-248)

IV.B4.RP-254a Lower grid assembly: Nickel alloy Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No alloy X-750 lower grid and neutron flux fatigue shock pad bolt locking devices (Three Mile Island Unit 1, only)

IV.B4.RP-254b Lower grid assembly: Nickel Alloy Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No alloy X-750 lower grid and neutron flux to wear; changes in shock pad bolt locking dimensions due to devices (Three Mile void swelling or Island Unit 1, only) distortion IV.B4.RP-246 IV.B4-12 Lower grid assembly: Stainless steel; Reactor coolant Cracking 'Chapter XI.M2, Water Chemistry, and No (R-196) upper thermal shield nickel alloy and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals (UTS) bolts and lower corrosion cracking and Chapter XI.M2, Water Chemistry thermal shield (LTS) bolts Expansion components (identified in the "Structure and Components" column)

(for Primary components see AMR Items IV.B4.RP-247 and IV.B4.RP-248)

B-50

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP-246a Lower grid assembly: Stainless steel; Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No upper thermal shield nickel alloy and neutron flux fatigue (UTS) bolt locking devices and lower thermal shield (LTS) bolt locking devices IV.B4.RP-246b Lower grid assembly: Stainless steel; Reactor coolant Loss of material due Chapter XI.M16A, PWR Vessel Internals No upper thermal shield nickel alloy and neutron flux to wear; changes in (UTS) bolt locking devices dimensions due to and lower thermal shield void swelling or (LTS) bolt locking devices distortion IV.B4.RP-260 IV.B4-31 Lower grid fuel assembly: Stainless steel; Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals" No (R-205) (a) accessible pads; (b) nickel alloy and neutron flux toughness Expansion components (identified in the accessible pad-to-rib due to neutron "Structure and Components" column) section welds; (c) irradiation (for Primary components see AMR Item accessible alloy X-750 embrittlement IV.B4.RP-259) dowels, cap screws and locking devices IV.B4.RP-260a Lower grid fuel assembly: Stainless steel; Reactor coolant Cracking due to stress Chapter XI.M16A, PWR Vessel Internals No (a) pads; (b) pad-to-rib nickel alloy and neutron flux corrosion cracking or and Chapter XI.M2, Water Chemistry (for section welds; (c) alloy X- fatigue SCC mechanisms only) 750 dowels, cap screws and locking devices IV.B4.RP-262 IV.B4-32 Lower grid assembly: Nickel alloy Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No (R-203) accessible alloy X-750 and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals dowel-to-lower fuel corrosion cracking and Chapter XI.M2, Water Chemistry assembly support pad Expansion components (identified in the locking welds "Structure and Components" column)

(for Primary components see AMR Item IV.B4.RP-261)

IV.B4.RP-261 IV.B4-32 Lower grid assembly: Nickel alloy Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No (R-203) alloy X-750 dowel-to- and neutron flux due to stress - Chapter XI.M16A, PWR Vessel Internals guide block welds corrosion cracking Primary components (identified in the "Structure and Components" column)

(for Expansion components see AMR Items IV.B4.RP-262 and IV.B4.RP-352)and Chapter XI.M2, Water Chemistry B-51

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.R-53 IV.B4-37 Reactor vessel internal Stainless steel; Reactor coolant Cumulative fatigue Fatigue is a time-limited aging analysis Yes, TLAA (R-53) components nickel alloy and neutron flux damage (TLAA) to be evaluated for the period of due to fatigue extended operation. See the SRP, Section 4.3 Metal Fatigue, for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

IV.B4.RP-24 IV.B4-38 Reactor vessel internal Stainless steel; Reactor coolant Loss of material Chapter XI.M2, Water Chemistry No (RP-24) components nickel alloy and neutron flux due to pitting and crevice corrosion IV.B4.RP-376 Reactor vessel internal Stainless steel; Reactor coolant Reduction in ductility Ductility - Reduction in Fracture Yes, TLAA components nickel alloy and neutron flux and fracture Toughness is a TLAA (BAW-2248A) to be toughness evaluated for the period of extended due to neutron operation. See the SRP, Section 4.7, irradiation "Other Plant-Specific TLAAs," for acceptable methods for meeting the requirements of 10 CFR 54.21(c)(1).

IV.B4.RP- IV.B4-42 Reactor vessel Stainless steel; Reactor coolant Cracking Chapter XI.M2, Water Chemistry,M1, Yes, if accessible 238382 (R-179) internalinternals: ASME nickel alloy and neutron flux due to fatigue, stress ASME Section XI Inservice Inspection, Primary,Section XI, Examination corrosion cracking, Subsections IWB, IWC, and Expansion or Category andor irradiation- IWD or Chapter XI.M16A, PWR Vessel Existing program B-N-3 core support assisted stress Internals, by invoking applicable 10 CFR components structure components corrosion cracking; 50.55a and ASME Section XI inservice indicate aging (inaccessible locations) loss of material inspection requirements effects that need due to wear managementNo B-52

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP- Upper grid assembly: Stainless steel; Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals Yes, if accessible 239352 alloy X-750 dowel-to- nickelNickel alloy and neutron flux toughnessCracking and Chapter XI.M2, Water Chemistry Primary, upper fuel assembly due to neutron Expansion or support pad welds (all irradiation Existing program plants except Davis- embrittlement; components Besse)Reactor vessel change in dimension indicate aging internal components due to void swelling; effects that need (inaccessible locations) loss of preload managementNo due to thermal and irradiation enhanced stress relaxation; loss of material due to wear-corrosion cracking IV.B4.RP-236 Reactor internal No Stainless steel; Reactor coolant No additional aging Chapter XI.M2, Water Chemistry and No Additional Measures nickel alloy and neutron flux management for Chapter XI.M16A, PWR Vessel Internals componentsReactor reactor internal No Note: Components with no additional vessel internal Additional Measures measures are not uniquely identifies in components with no components unless GALL tables - Components with no additional measures required by ASME additional measures are defined in Section Section XI, 3.3.1 of MRP-227, "Materials Reliability Examination Category Program: Pressurized Water Reactor B-N-3 or relevant Internals Inspection and Evaluation operating experience Guidelines" existsCracking due to stress corrosion cracking, and irradiation-assisted stress corrosion cracking IV.B4.RP-400 Core support shield Stainless steel Reactor coolant Cracking due to Chapter XI.M16A, PWR Vessel Internals No assembly: upper (top) and neutron flux stress-corrosion and Chapter XI.M2, Water Chemistry flange weld cracking B-53

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B4 Reactor Vessel Internals (PWR) - Babcock & Wilcox Structure and/or Aging Effect/ Further Item Link Material Environment Aging Management Program (AMP)

Component Mechanism Evaluation IV.B4.RP- Reactor vessel internal Stainless steel; Reactor coolant Loss of fracture Chapter XI.M16A, PWR Vessel Internals No 237401 components with no nickel alloy and neutron flux toughness Note: Components with no additional additional measuresCore due to neutron measures are not uniquely identified in support shield assembly: irradiation GALL tables - Components with no upper (top) flange weld embrittlement; additional measures are defined in Section change in dimension 3.3.1 of MRP-227, "Materials Reliability due to void swelling; Program: Pressurized Water Reactor loss of preload Internals Inspection and Evaluation due to thermal and Guidelines" irradiation enhanced stress relaxation; loss of material due to wear IV.B4.RP-382 IV.B4-42(R- Reactor vessel internals: Stainless steel; Reactor coolant Cracking, or Chapter XI.M1, ASME Section XI No 179) core support structure nickel alloy; cast and neutron flux Loss of material Inservice Inspection, Subsections IWB, austenitic due to wear IWC, and IWD stainless steel IV.B4.RP-352 Upper grid assembly: Nickel alloy Reactor coolant Cracking Chapter XI.M2, Water Chemistry, and No alloy X-750 dowel-to- and neutron flux due to stress Chapter XI.M16A, PWR Vessel Internals upper fuel assembly corrosion cracking Expansion components (identified in the support pad welds (all "Structure and Components" column) plants except Davis- (for Primary components see AMR Item Besse) IV.B4.RP-261)

B-54

(5) Mark-up of changes to GALL Report Chapter IX.C and IX.G IX.C Selected Definitions & Use of Terms for Describing and Standardizing MATERIALS Stainless steel Products grouped under the term stainless steel include wrought or forged austenitic, ferritic, martensitic, precipitation-hardened (PH), or duplex stainless steel (Cr content >11%). These stainless steels may be fabricated using a wrought or cast process. These materials are susceptible to a variety of aging effects and mechanisms, including loss of material due to pitting and crevice corrosion, and cracking due to stress corrosion cracking.

In some cases, when the recommended AMPan aging effect is applicable to all of the same for PHvarious stainless steel or castcategories, it can be assumed that the term stainless steel in the Material column of an AMR line-item in the GALL Report encompasses all stainless steel types. Cast austenitic stainless steel (CASS) as for stainless steel, PH stainless steel or CASS are included as a part of the stainless steel classification.

However, CASS is quite susceptible to loss of fracture toughness due to thermal and neutron irradiation embrittlement. Therefore, when this aging effect is being considered, CASS In addition, MRP-227-A indicates that PH stainless steels or martensitic stainless steels may be susceptible to loss of fracture toughness by a thermal aging mechanism. Therefore, when loss of fracture toughness due to thermal and neutron irradiation embrittlement is an applicable aging effect and mechanism for a component in the GALL Report, the CASS, PH stainless steel, or martensitic stainless steel designation is specifically identified designated in an AMR line-item.

Steel with stainless steel cladding also may be considered stainless steel when the aging effect is associated with the stainless steel surface of the material, rather than the composite volume of the material.

Examples of stainless steel designations that comprise this category include A-286, SA193-Gr. B8, SA193-Gr. B8M, Gr.

660 (A-286), SA193-6, SA193-Gr. B8 or B-8M, SA453, Type 416, Type 403, 410, 420, and Types431 martensitic stainless steels, Type 15-5, 17-4, and 13-8-Mo PH stainless steels, and SA-193, Grade B8 and B8M bolting materials.

Examples of wrought austenitic stainless materials that comprise this category include Type 304, 304NG, 304L, 308, 308L, 309, 309L, 316, and 347, 403, and 416..

Examples of CASS designationsthat comprise this category B-55

include CF-3, -8, -3M,CF3, CF3M, CF8 and -8M.CF8M.

[Ref. 6, 7], 30]

IX.G References

30. Welding Handbook, Seventh Edition, Volume 4, Metals and Their Weldability, American Welding Society, 1984, p.76-145.

B-56

Appendix B, Section 2 - Mark-up of Changes to the SRP-LR In the mark-up, red or green strikethrough text indicates a deletion and blue underline text indicates an insertion. Green text indicates a move, where a double strikethrough indicates the original location of the text and a double underline indicates the final location of the moved text.

(1) Mark-up of changes to SRP-LR Table 3.0-1 Table 3.0-1 FSAR Supplement for Aging Management of Applicable Systems GALL GALL Description of Program Implementation Applicable GALL Chapter Program Schedule Report and SRP-LR Chapter References XI.M16A PWR Vessel The program relies on implementation Program should GALL IV / SRP 3.1 Internals of the inspection and evaluation be implemented guidelines in EPRI Technical Report prior to period of No. 10165961022863 (MRP-227-A) extended and EPRI Technical Report No. operation 1016609 (MRP-228) to manage the aging effects on the reactor vessel internal components. This program is used to manage (a) various forms of cracking, including stress corrosion crackingSCC, primary water stress corrosion crackingPWSCC, irradiation-assisted stress corrosion cracking (IASCC), orand cracking due to fatigue/cyclical loading; (b) loss of material induced by wear; (c) loss of fracture toughness due to either thermal aging or, neutron irradiation embrittlement, or void swelling; (d) dimensional changes and potential loss of fracture toughness due to void swelling and irradiation growthor distortion; and (e) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.

(2) Mark-up of changes to SRP-LR Section 3.1.2, Acceptance Criteria 3.1.2.2.9 Removed as a result of LR-ISG-2011-04Cracking due to Stress Corrosion Cracking and Irradiation-Assisted Stress Corrosion Cracking Cracking due to SCC and irradiation-assisted stress corrosion cracking (IASCC) could occur in inaccessible locations for stainless steel and nickel-alloy Primary and Expansion PWR reactor vessel internal components. If aging effects are identified in accessible locations, the GALL Report recommends further evaluation of the aging effects in inaccessible locations on a plant-specific basis to ensure that this aging effect is adequately managed. Acceptance criteria are described in Branch Technical Position RLSB-1 (Appendix A.1 of this SRP-LR).

B-57

3.1.2.2.10 Removed as a result of LR-ISG-2011-04Loss of Fracture Toughness due to Neutron Irradiation Embrittlement; Change in Dimension due to Void Swelling; Loss of Preload due to Stress Relaxation; or Loss of Material due to Wear Loss of fracture toughness due to neutron irradiation embrittlement, change in dimension due to void swelling, loss of preload due to stress relaxation, or loss of material due to wear could occur in inaccessible locations for stainless steel and nickel-alloy Primary and Expansion PWR reactor vessel internal components. If aging effects are identified in accessible locations, the GALL Report recommends further evaluation of the aging effects in inaccessible locations on a plant-specific basis to ensure that this aging effect is adequately managed. Acceptance criteria are described in Branch Technical Position RLSB-1 (Appendix A.1 of this SRP-LR).

3.1.2.2.12 Removed as a result of LR-ISG-2011-04Cracking due to Fatigue EPRI 1016596, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0) identifies cracking due to fatigue as an aging effect that can occur for the lower flange weld in the core support barrel assembly, fuel alignment plate in the upper internals assembly, and core support plate lower support structure in PWR internals designed by Combustion Engineering. The GALL Report recommends that inspection for cracking in this component be performed if acceptable fatigue life cannot be demonstrated by TLAA through the period of extended operation as defined in 10 CFR 54.3.

3.1.2.2.13 Removed as a result of LR-ISG-2011-04Cracking due to Stress Corrosion Cracking and Fatigue Cracking due to stress corrosion cracking and fatigue could occur in nickel alloy control rod guide tube assemblies, guide tube support pins exposed to reactor coolant, and neutron flux. The GALL Report, AMR Item IV.B2.RP-355, recommends further evaluation of a plant-specific AMP to ensure this aging effect is adequately managed. Acceptance criteria are described in Branch Technical Position RLSB-1 (Appendix A.1 of this SRP-LR).

3.1.2.2.14 Removed as a result of LR-ISG-2011-04Loss of Material due to Wear Loss of material due to wear could occur in nickel alloy control rod guide tube assemblies, guide tube support pins and in Zircaloy-4 incore instrumentation lower thimble tubes exposed to reactor coolant, and neutron flux. The GALL Report, AMR Items IV.B2.RP-356 and IV.B3.RP-357, recommends further evaluation of a plant-specific AMP to ensure this aging effect is adequately managed. Acceptance criteria are described in Branch Technical Position RLSB-1 (Appendix A.1 of this SRP-LR).

(3) Mark-up of changes to SRP-LR Section 3.1.3, Review Procedures 3.1.3.2.9 Removed as a result of LR-ISG-2011-04Cracking due to Stress Corrosion Cracking and Irradiation-Assisted Stress Corrosion Cracking The GALL Report recommends further evaluation of cracking due to SCC and IASCC for inaccessible locations for Primary and Expansion PWR reactor vessel internal components if aging effects are identified for these components in accessible locations. The reviewer reviews the applicants proposed program on a case-by-case basis to ensure that an adequate program B-58

will be in place for the management of these aging effects consistent with the action item documented in the staffs safety evaluation for MRP-227, Revision 0..

3.1.3.2.10 Removed as a result of LR-ISG-2011-04Loss of Fracture Toughness due to Neutron Irradiation Embrittlement; Change in Dimension due to Void Swelling; Loss of Preload due to Stress Relaxation; or Loss of Material due to Wear The GALL Report recommends further evaluation of loss of fracture toughness due to neutron irradiation embrittlement, change in dimension due to void swelling, loss of preload due to stress relaxation, or loss of material due to wear for inaccessible locations for Primary and Expansion PWR reactor vessel internal components, if aging effects are identified for these components in accessible locations. The reviewer reviews the applicants proposed program on a case-by-case basis to ensure that an adequate program will be in place for the management of these aging effects consistent with the action item documented in the staffs safety evaluation for MRP-227, Revision 0.

3.1.3.2.12 Removed as a result of LR-ISG-2011-04Cracking due to Fatigue The GALL Report recommends further evaluation of cracking due to fatigue in the lower flange weld in the core support barrel assembly, fuel alignment plate in the upper internals assembly, and core support plate in the lower support structure in PWR internals designed by Combustion Engineering. The reviewer determines whether a TLAA has been performed for each component, consistent with the action item documented in the staffs safety evaluation for MRP-227, Revision 0. If a TLAA has not been performed, the reviewer determines whether the applicant has performed an evaluation to identify the potential location and extent of fatigue cracking for each component consistent with the action item documented in the staffs safety evaluation for MRP-227, Revision 0.

3.1.3.2.13 Removed as a result of LR-ISG-2011-04Cracking due to Stress Corrosion Cracking and Fatigue The GALL Report recommends further evaluation of cracking due to stress corrosion cracking and fatigue in the nickel alloy control rod guide tube assemblies, guide tube support pins exposed to reactor coolant, and neutron flux. The reviewer reviews the applicants proposed program on a case-by-case basis to ensure that an adequate program will be in place for the management of these aging effects consistent with the action item documented in the staffs safety evaluation for MRP-227, Revision 0.

3.1.3.2.14 Removed as a result of LR-ISG-2011-04Loss of Material due to Wear The GALL Report recommends further evaluation of loss of material due to wear in nickel alloy control rod guide tube assemblies, guide tube support pins and in Zircaloy-4 incore instrumentation lower thimble tubes exposed to reactor coolant, and neutron flux. The reviewer reviews the applicants proposed program on a case-by-case basis to ensure that an adequate program will be in place for the management of these aging effects consistent with the action item documented in the staffs safety evaluation for MRP-227, Revision 0.

B-59

(4) Mark-up of changes to SRP-LR Table 3.1-1 Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 3 BWR/ Stainless steel or nickel Cumulative fatigue damage Fatigue is a TLAA Yes, TLAA (See IV.B1.R-53 IV.B1-14 (R-53)

PWR alloy reactor vessel internal due to fatigue evaluated for the period of subsection 3.1.2.2.1) IV.B2.RP-303 IV.B2-31 (R-53) components exposed to extended operation (See IV.B3.RP-339 IV.B3-24 (R-53)

IV.B4.R-53 IV.B4-37 (R-53) reactor coolant and neutron SRP, Section 4.3 Metal IV.B3.RP-389 N/A flux Fatigue, for acceptable IV.B3.RP-390 N/A methods to comply with IV.B3.RP-391 N/A 10 CFR 54.21(c)(1) 15 PWR Stainless steel Babcock & Reduction inof ductility and Ductility - Reduction in Yes, TLAA (See IV.B4.RP-376 N/A Wilcox (including CASS, fracture toughness due to fFracture tToughness is a subsection 3.1.2.2.3.3) martensitic SS, and PH SS) neutron irradiation TLAA to be evaluated for and nickel alloy reactor embrittlement, and for the period of extended vessel internal components CASS, martensitic SS, and operation., See the SRP, exposed to reactor coolant PH SS due to thermal aging Section 4.7, Other Plant-and neutron flux embrittlement Specific TLAAs, for acceptable methods forof meeting the requirements of 10 CFR 54.21(c)(1).).

23 PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR Yes, if accessible IV.B2.RP-268 N/A alloy PWR reactor vessel corrosion cracking, and Vessel Internals," and Primary, Expansion or IV.B3.RP-309 N/A internal components irradiation-assisted stress Chapter XI.M2, Water Existing program IV.B4.RP-238 N/A (inaccessible locations) corrosion cracking Chemistry components indicate exposed to reactor coolant aging effects that need and neutron flux management (See subsection 3.1.2.2.9) 24 PWR Stainless steel or nickel Loss of fracture toughness Chapter XI.M16A, PWR Yes, if accessible IV.B2.RP-269 N/A alloy PWR reactor vessel due to neutron irradiation Vessel Internals Primary, Expansion or IV.B3.RP-311 N/A internal components embrittlement; or changes in Existing program IV.B4.RP-239 N/A (inaccessible locations) dimension due to void components indicate exposed to reactor coolant swelling; or loss of preload aging effects that need and neutron flux due to thermal and management (See irradiation enhanced stress subsection 3.1.2.2.10) relaxation; or loss of material due to wear B-60

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 26 PWR Stainless steel Combustion Cracking due to fatigue Chapter XI.M16A, PWR Yes, evaluate to IV.B3.RP-333 N/A Engineering core support Vessel Internals," and determine the potential IV.B3.RP-338 barrel assembly: lower Chapter XI.M2, Water locations and extent of IV.B3.RP-343 flange weld exposed to Chemistry, if fatigue life fatigue cracking (See reactor coolant and neutron cannot be confirmed by subsection 3.1.2.2.12) flux; Upper internals TLAA assembly: fuel alignment plate (applicable to plants with core shrouds assembled with full height shroud plates) exposed to reactor coolant and neutron flux; Lower support structure: core support plate (applicable to plants with a core support plate) exposed to reactor coolant and neutron flux 27 PWR Nickel alloy Westinghouse Cracking due to stress A plant-specific aging Yes, plant-specific (See IV.B2.RP-355 N/A control rod guide tube corrosion cracking and management program is subsection 3.1.2.2.13) assemblies, guide tube fatigue to be evaluated support pins exposed to reactor coolant and neutron flux 28 PWR Nickel alloy Westinghouse Loss of material due to wear A plant-specific aging Yes, plant-specific (See IV.B2.RP-356 N/A control rod guide tube management program is subsection 3.1.2.2.14) IV.B3.RP-357 N/A assemblies, guide tube to be evaluated support pins, and Zircaloy-4 Combustion Engineering incore instrumentation thimble tubes exposed to reactor coolant and neutron flux 28 PWR Stainless steel Combustion Loss of material due to Chapter XI.M16A, PWR No IV.B3.RP-400 N/A Engineering Existing wear; cracking due to stress Vessel Internals, and Programs components corrosion cracking, Chapter XI.M2, Water exposed to reactor coolant irradiation-assisted stress Chemistry (for SCC and neutron flux corrosion cracking, or mechanisms only) fatigue B-61

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 32 PWR Stainless steel, nickel alloy, Cracking, or loss of material Chapter XI.M1, ASME No IV.B2.RP-382 IV.B2-26 (R-142) or CASS reactor vessel due to wear Section XI Inservice IV.B3.RP-382 IV.B3-22 (R-170) internals, core support Inspection, Subsections IV.B4.RP-382 IV.B4-42 (R-179) structure (not already IWB, IWC, and IWD or referenced as ASME Chapter XI.M16A, PWR Section XI Examination Vessel Internals, invoking Category applicable 10 CFR 50.55a B-N-3 core support and ASME Section XI structure components in inservice inspection MRP-227-A), exposed to requirements for these reactor coolant and neutron components flux 51 PWR Stainless steel or nickel- Cracking due to stress Chapter XI.M16A, PWR No IV.B4.RP-236 N/A alloy Babcock & Wilcox corrosion cracking, Vessel Internals," and IV.B4.RP-241 IV.B4-7(R-125) reactor internal irradiation-assisted stress Chapter XI.M2, Water IV.B4.RP-244 IV.B4-7(R-125)

IV.B4.RP-245 IV.B4-13(R-194) components exposed to corrosion cracking, or Chemistry IV.B4.RP-246 IV.B4-12(R-196) reactor coolant and neutron fatigue IV.B4.RP-247 IV.B4-13(R-194) flux IV.B4.RP-248 IV.B4-12(R-196)

IV.B4.RP-254 IV.B4-25(R-210)

IV.B4.RP-256 IV.B4-25(R-210)

IV.B4.RP-261 IV.B4-32(R-203)

IV.B4.RP-262 IV.B4-32(R-203)

IV.B4.RP-352 N/A IV.B4.RP-375 N/A 52 PWR Stainless steel or nickel- Cracking due to stress Chapter XI.M16A, PWR No IV.B3.RP-306 N/A alloy Combustion corrosion cracking, Vessel Internals," and IV.B3.RP-312 IV.B3-2(R-149)

Engineering reactor internal irradiation-assisted stress Chapter XI.M2, Water IV.B3.RP-313 N/A IV.B3.RP-314 IV.B3-9(R-162) components exposed to corrosion cracking, or Chemistry IV.B3.RP-316 IV.B3-9(R-162) reactor coolant and neutron fatigue IV.B3.RP-320 IV.B3-9(R-162) flux IV.B3.RP-322 N/A IV.B3.RP-323 N/A IV.B3.RP-324 N/A IV.B3.RP-325 N/A IV.B3.RP-327 IV.B3-15(R-155)

IV.B3.RP-328 IV.B3-15(R-155)

IV.B3.RP-329 IV.B3-15(R-155)

IV.B3.RP-330 IV.B3-23(R-167)

IV.B3.RP-334 IV.B3-23(R-167)

IV.B3.RP-335 IV.B3-23(R-167)

IV.B3.RP-342 N/A IV.B3.RP-358 N/A B-62

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 53 PWR Stainless steel or nickel- Cracking due to stress Chapter XI.M16A, PWR No IV.B2.RP-265 N/A alloy Westinghouse reactor corrosion cracking, Vessel Internals," and IV.B2.RP-271 IV.B2-10(R-125) internal components irradiation-assisted stress Chapter XI.M2, Water IV.B2.RP-273 IV.B2-10(R-125)

IV.B2.RP-275 IV.B2-6(R-128) exposed to reactor coolant corrosion cracking, or Chemistry IV.B2.RP-276 IV.B2-8(R-120) and neutron flux fatigue IV.B2.RP-278 IV.B2-8(R-120)

IV.B2.RP-280 IV.B2-8(R-120)

IV.B2.RP-282 IV.B2-8(R-120)

IV.B2.RP-286 IV.B2-16(R-133)

IV.B2.RP-289 IV.B2-20(R-130)

IV.B2.RP-291 IV.B2-24(R-138)

IV.B2.RP-293 IV.B2-24(R-138)

IV.B2.RP-294 IV.B2-24(R-138)

IV.B2.RP-298 IV.B2-28(R-118)

IV.B2.RP-301 IV.B2-40(R-112)

IV.B2.RP-346 N/A IV.B2.RP-387 N/A 51a PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B4.RP-241 IV.B4-7 (R-125) alloy Babcock & Wilcox corrosion cracking, Vessel Internals, and IV.B4.RP-241a N/A reactor internal Primary irradiation-assisted stress Chapter XI.M2, Water IV.B4.RP-242a N/A IV.B4.RP-247 IV.B4-13 (R-194) components exposed to corrosion cracking, or Chemistry (for SCC IV.B4.RP-247a N/A reactor coolant and neutron fatigue mechanisms only) IV.B4.RP-248 IV.B4-25 (R-210) flux IV.B4.RP-248a N/A IV.B4.RP-249a N/A IV.B4.RP-252a N/A IV.B4.RP-256 IV.B4-25 (R-210)

IV.B4.RP-256a N/A IV.B4.RP-258a N/A IV.B4.RP-259a N/A IV.B4.RP-261 IV.B4-32 (R-203)

IV.B4.RP-400 N/A 51b PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B4.RP-244 IV.B4-7 (R-125) alloy Babcock & Wilcox corrosion cracking, Vessel Internals, and IV.B4.RP-244a N/A reactor internal Expansion irradiation-assisted stress Chapter XI.M2, Water IV.B4.RP-245 IV.B4-13 (R-194)

IV.B4.RP-245a N/A components exposed to corrosion cracking, fatigue, Chemistry (for SCC IV.B4.RP-246 IV.B4-12 (R-196) reactor coolant and neutron or overload mechanisms only) IV.B4.RP-246a N/A flux IV.B4.RP-254 IV.B4-25 (R-210)

IV.B4.RP-254a N/A B-63

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended IV.B4.RP-260a N/A IV.B4.RP-262 IV.B4-32 (R-203)

IV.B4.RP-352 N/A IV.B4.RP-250a N/A IV.B4.RP-375 N/A 52a PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B3.RP-312 IV.B3-2 (R-149) alloy Combustion corrosion cracking, Vessel Internals, and IV.B3.RP-314 IV.B3-9 (R-162)

Engineering reactor internal irradiation-assisted stress Chapter XI.M2, Water IV.B3.RP-322 N/A IV.B3.RP-324 N/A Primary components corrosion cracking, or Chemistry (for SCC IV.B3.RP-326a N/A exposed to reactor coolant fatigue mechanisms only) IV.B3.RP-327 IV.B3-15 (R-155) and neutron flux IV.B3.RP-328 IV.B3-15 (R-155)

IV.B3.RP-342 N/A IV.B3.RP-358 N/A IV.B3.RP-362a N/A IV.B3.RP-363 N/A IV.B3.RP-338 N/A IV.B3.RP-343 N/A 52b PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B3.RP-313 NA alloy Combustion corrosion cracking, Vessel Internals, and IV.B3.RP-316 IV.B3-9 (R-162)

Engineering reactor internal irradiation-assisted stress Chapter XI.M2, Water IV.B3.RP-323 N/A IV.B3.RP-325 N/A Expansion components corrosion cracking, or Chemistry (for SCC IV.B3.RP-329 IV.B3-12 (R-155) exposed to reactor coolant fatigue mechanisms only) IV.B3.RP-330 IV.B3-23 (R-167) and neutron flux IV.B3.RP-333 N/A IV.B3.RP-335 IV.B3-23 (R-167)

IV.B3.RP-362c N/A 52c PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B3.RP-320 IV.B3-9 (R-162) alloy Combustion corrosion cracking, Vessel Internals, and IV.B3.RP-334 IV.B3-23 (R-167)

Engineering reactor irradiation-assisted stress Chapter XI.M2, Water internal Existing corrosion cracking, or Chemistry (for SCC Programs components fatigue mechanisms only) exposed to reactor coolant and neutron flux B-64

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 53a PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B2.RP-270a N/A alloy Westinghouse corrosion cracking, Vessel Internals, and IV.B2.RP-271 IV.B2-10 (R-125) reactor internal Primary irradiation-assisted stress Chapter XI.M2, Water IV.B2.RP-275 IV.B2-6 (R-128)

IV.B2.RP-276 IV.B2-8 (R-120) components exposed to corrosion cracking, or Chemistry (for SCC IV.B2.RP-280 IV.B2-8 (R-120) reactor coolant and neutron fatigue mechanisms only) IV.B2.RP-298 IV.B2-28 (R-118) flux IV.B2.RP-302 N/A IV.B2.RP-387 N/A 53b PWR Stainless steel Cracking due to stress Chapter XI.M16A, PWR No IV.B2.RP-273 IV.B2-10 (R-125)

Westinghouse reactor corrosion cracking, Vessel Internals, and IV.B2.RP-278 IV.B2-8 (R-120) internal Expansion irradiation-assisted stress Chapter XI.M2, Water IV.B2.RP-286 IV.B2-16 (R-133)

IV.B2.RP-291 IV.B2-24 (R-138) components exposed to corrosion cracking, or Chemistry (for SCC IV.B2.RP-291a N/A reactor coolant and neutron fatigue mechanisms only) IV.B2.RP-291b N/A flux IV.B2.RP-293 IV.B2-24 (R-138)

IV.B2.RP-294 IV.B2-24 (R-138)

IV.B2.RP-387a N/A 53c PWR Stainless steel or nickel Cracking due to stress Chapter XI.M16A, PWR No IV.B2.RP-289 IV.B2-20 (R-130) alloy Westinghouse corrosion cracking, Vessel Internals, and IV.B2.RP-301 IV.B2-40 (R-112) reactor internal Existing irradiation-assisted stress Chapter XI.M2, Water IV.B2.RP-345 N/A IV.B2.RP-346 N/A Programs components corrosion cracking, or Chemistry (for SCC IV.B2.RP-399 N/A exposed to reactor coolant fatigue mechanisms only) IV.B2.RP-355 N/A and neutron flux 54 PWR Stainless steel bottom Loss of material due to Chapter XI.M16A, PWR No IV.B2.RP-284 IV.B2-12(R-143) mounted instrument wear Vessel Internals, and or IV.B2-13 (R-145) system flux thimble Chapter XI.M37, "Flux tubes (with or without Thimble Tube Inspection" chrome plating) exposed to reactor coolant and neutron flux (Westinghouse Existing Programs components) 55 PWR Stainless steel thermal Cracking due to fatigue; Chapter XI.M16A, PWR No IV.B2.RP-302 N/A shield assembly, thermal Loss of material due to wear Vessel Internals shield flexures exposed to reactor coolant and neutron flux 55a PWR Stainless steel or nickel No additional aging Chapter XI.M16A, PWR No IV.B4.RP-236 NA alloy Babcock and Wilcox management for reactor Vessel Internals reactor internal No internal No Additional Additional Measures Measures components B-65

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended components exposed to unless required by ASME reactor coolant and neutron Section XI, Examination flux Category B-N-3 or relevant operating experience invalidates MRP-227-A.

55b PWR Stainless steel or nickel No additional aging Chapter XI.M16A, PWR No IV.B3.RP-306 NA alloy Combustion management for reactor Vessel Internals Engineering reactor internal internal No Additional No Additional Measures Measures components components exposed to unless required by ASME reactor coolant and neutron Section XI, Examination flux Category B-N-3 or relevant operating experience invalidates MRP-227-A.

55c PWR Stainless steel or nickel No additional aging Chapter XI.M16A, PWR No IV.B2.RP-265 NA alloy Westinghouse management for reactor Vessel Internals reactor internal No internal No Additional Additional Measures Measures components components exposed to unless required by ASME reactor coolant and neutron Section XI, Examination flux Category B-N-3 or relevant operating experience invalidates MRP-227-A.

56 PWR Stainless steel or nickel- Loss of fracture toughness Chapter XI.M16A, PWR No IV.B3.RP-307 N/A alloy Combustion due to neutron irradiation Vessel Internals IV.B3.RP-315 IV.B3-7(R-165)

Engineering reactor internal embrittlement; or changes in IV.B3.RP-317 IV.B3-7(R-165)

IV.B3.RP-318 IV.B4-8(R-163) components exposed to dimension due to void IV.B3.RP-319 IV.B3-9(R-162) reactor coolant and neutron swelling; or loss of preload IV.B3.RP-326 N/A flux due to thermal and IV.B3.RP-331 N/A irradiation enhanced stress IV.B3.RP-332 IV.B3-17(R-156) relaxation; or loss of material IV.B3.RP-336 IV.B3-22(R-170) due to wear IV.B3.RP-359 N/A IV.B3.RP-360 N/A IV.B3.RP-361 N/A IV.B3.RP-362 N/A IV.B3.RP-363 N/A IV.B3.RP-364 N/A IV.B3.RP-365 N/A B-66

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended IV.B3.RP-366 N/A 58 PWR Stainless steel or nickel- Loss of fracture toughness Chapter XI.M16A, PWR No IV.B4.RP-237 N/A alloy Babcock & Wilcox due to neutron irradiation Vessel Internals IV.B4.RP-240 IV.B4-1(R-128) reactor internal embrittlement; or changes in IV.B4.RP-242 IV.B4-4(R-183)

IV.B4.RP-243 IV.B4-1(R-128) components exposed to dimension due to void IV.B4.RP-249 IV.B4-12(R-196) reactor coolant and neutron swelling; or loss of preload IV.B4.RP-250 IV.B4-12(R-196) flux due to thermal and IV.B4.RP-251 IV.B4-15(R-190) irradiation enhanced stress IV.B4.RP-252 IV.B4-16(R-188) relaxation; or loss of material IV.B4.RP-253 IV.B4-21(R-191) due to wear IV.B4.RP-258 IV.B4-4(R-183)

IV.B4.RP-259 IV.B4-31(R-205)

IV.B4.RP-260 IV.B4-31(R-205) 59 PWR Stainless steel or nickel- Loss of fracture toughness Chapter XI.M16A, PWR No IV.B2.RP-267 N/A alloy Westinghouse reactor due to neutron irradiation Vessel Internals IV.B2.RP-270 IV.B2-1(R-124) internal components embrittlement; or changes in IV.B2.RP-272 IV.B2-6(R-128)

IV.B2.RP-274 IV.B2-6(R-128) exposed to reactor coolant dimension due to void IV.B2.RP-281 IV.B2-9(R-122) and neutron flux swelling; or loss of preload IV.B2.RP-285 IV.B2-14(R-137) due to thermal and IV.B2.RP-287 IV.B2-17(R-135) irradiation enhanced stress IV.B2.RP-288 IV.B2-18(R-132) relaxation; or loss of material IV.B2.RP-290 IV.B2-21(R-140) due to wear IV.B2.RP-292 IV.B2-21(R-140)

IV.B2.RP-295 IV.B2-22(R-141)

IV.B2.RP-296 N/A IV.B2.RP-297 N/A IV.B2.RP-299 IV.B2-34(R-115)

IV.B2.RP-300 IV.B2-33(R-108)

IV.B2.RP-345 N/A IV.B2.RP-354 N/A IV.B2.RP-386 N/A IV.B2.RP-388 N/A 56a PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B3.RP-315 IV.B3-7 (R-165) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B3.RP-318 IV.B3-8 (R-163) martensitic SS) or nickel embrittlement and for CASS, IV.B3.RP-359 N/A IV.B3.RP-360 N/A alloy Combustion martensitic SS, and PH SS IV.B3.RP-362 N/A Engineering reactor internal due to thermal aging IV.B3.RP-364 N/A Primary components embrittlement; or changes in IV.B3.RP-366 N/A exposed to reactor coolant dimensions due to void IV.B3.RP-365 N/A and neutron flux swelling or distortion; or loss IV.B3.RP-326 N/A of preload due to thermal B-67

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended and irradiation enhanced stress relaxation or creep; or loss of material due to wear 56b PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B3.RP-317 IV.B3-7 (R-165) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B3.RP-331 N/A martensitic SS) embrittlement and for CASS, IV.B3.RP-359a N/A IV.B3.RP-361 N/A Combustion Engineering martensitic SS, and PH SS IV.B3.RP-362b N/A Expansion reactor internal due to thermal aging components exposed to embrittlement; or changes in reactor coolant and neutron dimensions due to void flux swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 56c PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B3.RP-319 IV.B3-9 (R-162) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B3.RP-332 IV.B3-17 (R-156) martensitic SS) or nickel embrittlement and for CASS, IV.B3.RP-334a N/A IV.B3.RP-336 IV.B3-22 (R-170) alloy Combustion martensitic SS, and PH SS IV.B3.RP-357 N/A Engineering reactor internal due to thermal aging Existing Programs embrittlement; or changes in components exposed to dimensions due to void reactor coolant and neutron swelling or distortion; or loss flux of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 58a PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B4.RP-240 IV.B4-1 (R-128) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B4.RP-240a N/A martensitic SS) or nickel embrittlement and for CASS, IV.B4.RP-242 IV.B4-4 (R-183)

IV.B4.RP-247b N/A alloy Babcock & Wilcox martensitic SS, and PH SS IV.B4.RP-248b N/A reactor internal Primary due to thermal aging IV.B4.RP-249 IV.B4-12 (R-196) components exposed to embrittlement; or changes in IV.B4.RP-251 IV.B4-15 (R-190) reactor coolant and neutron dimensions due to void IV.B4.RP-251a N/A flux swelling or distortion; or loss IV.B4.RP-252 IV.B4-16 (R-188) of preload due to wear; or IV.B4.RP-254b N/A loss of material due to wear IV.B4.RP-256b N/A IV.B4.RP-258 IV.B4-4 (R-183)

IV.B4.RP-259 IV.B4-31 (R-205)

IV.B4.RP-401 N/A B-68

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended 58b PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B4.RP-245b N/A including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B4.RP-246b N/A martensitic SS) or nickel embrittlement and for CASS, IV.B4.RP-254b N/A IV.B4.RP-260 IV.B4-31 (R-205) alloy Babcock & Wilcox martensitic SS, and PH SS IV.B4.RP-243 IV.B4-1 (R-128) reactor internal Expansion due to thermal aging IV.B4.RP-243a N/A components exposed to embrittlement; or changes in IV.B4.RP-250 IV.B4-12 (R-196) reactor coolant and neutron dimensions due to void IV.B4.RP-375a N/A flux swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 59a PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B2.RP-270 IV.B2-1 (R-124) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B2.RP-272 IV.B2-6 (R-128) martensitic SS) or nickel embrittlement and for CASS, IV.B2.RP-296 N/A IV.B2.RP-297 N/A alloy Westinghouse reactor martensitic SS, and PH SS IV.B2.RP-302a N/A internal Primary due to thermal aging IV.B2.RP-354 N/A components exposed to embrittlement; or changes in IV.B2.RP-388 N/A reactor coolant and neutron dimensions due to void IV.B2.RP-300 N/A flux swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 59b PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B2.RP-274 IV.B2-6 (R-128) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B2.RP-278a N/A martensitic SS) embrittlement and for CASS, IV.B2.RP-287 IV.B2-17 (R-135)

IV.B2.RP-290 IV.B2-21 (R-140)

Westinghouse reactor martensitic SS, and PH SS IV.B2.RP-290a N/A internal Expansion due to thermal aging IV.B2.RP-290b N/A components exposed to embrittlement; or changes in IV.B2.RP-292 IV.B2-21 (R-140) reactor coolant and neutron dimensions due to void IV.B2.RP-295 IV.B2-22 (R-141) flux swelling or distortion; or loss IV.B2.RP-388a N/A of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear 59c PWR Stainless steel (SS, Loss of fracture toughness Chapter XI.M16A, PWR No IV.B2.RP-285 IV.B2-14 (R-137) including CASS, PH SS or due to neutron irradiation Vessel Internals IV.B2.RP-288 IV.B2-18 (R-132) martensitic SS) or nickel embrittlement and for CASS, IV.B2.RP-299 IV.B2-34 (R-115)

IV.B2.RP-356 N/A alloy Westinghouse reactor martensitic SS, and PH SS B-69

Table 3.1-1 Summary of Aging Management Programs for Reactor Vessel, Internals, and Reactor Coolant System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Aging Management Further Evaluation Rev2 Item Rev1 Item Effect/Mechanism Programs Recommended internal Existing due to thermal aging Programs components embrittlement; or changes in exposed to reactor coolant dimensions due to void and neutron flux swelling or distortion; or loss of preload due to thermal and irradiation enhanced stress relaxation or creep; or loss of material due to wear B-70

Appendix C STAFF RESPONSE TO PUBLIC COMMENTS ON DRAFT LICENSE RENEWAL INTERIM STAFF GUIDANCE 2011-04 C-1

Source of Comments I. Comments from Jean Smith, Electric Power Research Institute Materials Reliability Program (EPRI-MRP) and the Pressurized Water Reactor Owners Group Materials Subcommittee (PWROG-MSC) (ADAMS Accession No. ML12146A267)

II. Comments from Mark Richter, Nuclear Energy Institute (NEI) (ADAMS Accession No. ML12144A147)

Source

  1. Summary of Comment Response ID 1 I-1 The NRC reviewed and approved with limitations MRP-227 Revision The staff agrees with the comment, in part, that it is not necessary to have 0, and subsequently, MRP-227-A was published to incorporate the the level of detail included in LR-ISG-2011-04 issued for public comment SER additions. All needed actions for licensees are contained in regarding PWR reactor vessel internal (RVI) components. However, the MRP-227-A. As a result, it is appropriate for the NRC to review a staff does not agree that the final LR-ISG-2011-04 should only reference licensees PWR reactor internals aging management program MRP-227-A; instead reference to the topical report should be made only against the criteria contained in MRP-227-A. As such, it is not when it is appropriate. Revisions were made to eliminate duplication of necessary to include all the details currently in NUREG-1800 and information for RVIs that is detailed in MRP-227-A. The following is a NUREG-1801 regarding PWR reactor internals, and instead, only a summary of the revisions that have been incorporated into final LR-ISG-reference to MRP-227-A should be made. Outlining the 2011-04 as a result of this comment:

requirements for reactor internals in the Interim Staff Guidance may lead to confusion with respect to the implementation of duplicate Revision to GALL Report Aging Management Program (AMP) XI.M16A requirements, may cause undue NRC staff burden reconciling the In general, GALL Report AMP XI.M16A, PWR Vessel Internals, in final documents each time MRP-227 is revised by the industry, and will LR-ISG-2011-04 references MRP-227-A in the program elements and does likely lead to human errors in document alignment through future not delineate the MRP-227-A inspection and evaluation guidelines for PWR revisions. RVIs. In addition, areas resolved in the staffs safety evaluation (SE),

Revision 1, for MRP-227 and Applicant/Licensee Action Items (A/LAI) are not addressed in GALL Report AMP XI.M16A in final LR-ISG-2011-04.

Revision to SRP-LR Table 3.1-1 Final LR-ISG-2011-04 does not incorporate specific reference to Primary Category, Expansion Category, or Existing Program inspection and evaluation guidelines into the Rev. 2 Item column in the aging management review (AMR) line items for PWR RVIs in SRP-LR Table 3.1-1. In addition, the Component column for PWR RVIs in SRP-LR Table 3.1-1 in final LR-ISG-2011-04 is based on the commodity groups and inspection categories in MRP-227-A.

Revision to GALL Tables IV.B2, IV.B3, and IV.B4 GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 do not reference inspection categories and MRP-227-A inspection and evaluation guidelines.

Revision to SRP-LR Further Evaluation Recommendations for PWR RVIs Areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04 (i.e., these SRP-LR sections were deleted and do appear in final LR-ISG-2011-04). In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the C-2

Source

  1. Summary of Comment Response ID A/LAIs for MRP-227-A in Appendix C of the LRA.

2 I-2 Commenter referenced statement in Section 3.1.2.2.9.A.1 of The staff agrees with the comment that the responses to A/LAIs are to be Appendix A of LR-ISG 2011-04. provided in Appendix C of the LRAs. Thus, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their This statement requires that licensees include responses to responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In applicant action items in both Appendix C of the LRA and in addition, as a result of the staffs resolution of Source ID I-1, areas resolved appropriate further evaluation sections of the LRA. This duplication in the staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in of information provides no significant value to the reviewers. It is the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As recommended that all A/LAI responses be included only in Appendix a result of the staffs resolution of Source ID I-1, final LR-ISG-2011-04 does C, so they are in an easily-referenced location. Any additional not incorporate SRP-LR Section 3.1.2.2.9.A.1.

discussion of the A/LAIs in the further evaluation sections of the SRP should be limited to identifying each of the items requiring responses and any details necessary to ensure responses are adequate. Any other items requiring discussion of the A/LAI responses in further evaluation sections of the LRA should be deleted or reference made to Appendix C of the LRA.

3 I-3 In NUREG-1801 Revision 2 XI.M16A Program Description, last The staff agrees, in part, with the comment in that better guidance paragraph, as well as in ISG-LR-2011-04 Section 3.1.2.2.9.A.2, both regarding the inspection plan is needed to avoid confusion. Regulatory an aging management program and an inspection plan are required Issue Summary (RIS) 2011-07, License Renewal Submittal Information to be submitted as part of an applicants license renewal application. For Pressurized Water Reactor Internals Aging Management, dated July 21, 2011, provides the staffs expectations for Category D plants (PWR However nowhere in these two documents is there any clear plant licensees that had not submitted their LRAs but plan to submit an guidance on the information that should be included in an inspection LRA in the future) to submit, for NRC staff review and approval, an AMP for plan. This ambiguity could lead to applicants submitting information vessel internals that is consistent with MRP-227-A.

that might not meet NRC needs in this area.

An inspection plan is one aspect of satisfying A/LAI No. 8 of the staffs In order to address this situation it is requested that the aging SE, Revision 1, for MRP-227. An inspection plan provides information management program and inspection plan for an applicant be about the RVI components to be inspected and a description of how they clearly defined. It is proposed that the aging management program will be managed for age-related degradation (e.g., examination method, address the 10 program element recommendations for PWR RVI frequency, acceptance criteria, coverage, etc.). The staff expects that the components in GALL AMP XI.M16A, PWR Vessel Internals (AMP details of an inspection plan for Category D plants will be incorporated XI.M16A in NUREG-1801, Revision 2). The inspection plan could into the LRA submittal as part of the 10-element AMP and AMR line items.

be included within a program (i.e. a program/plan) or be a separate Thus, consistent with RIS 2011-07, the staff does not expect Category D document if submitted with a license renewal application. plants to provide a separate document that contains an inspection plan in response to A/LAI No. 8.

The industry believes these elements are satisfied by the applicable line items from Tables 4-1 through 4-9 and Tables 5-1 through 5-3 In order to avoid duplication and confusion, as part of the resolution to of MRP-227-A. The inspection plan submitted as part of a license Source ID I-1, areas resolved in the staffs SE, Revision 1, for MRP-227 renewal application (LRA) should be included in Appendix C of the and A/LAIs are not addressed in the Further Evaluation sections of the LRA along with the responses to the A/LAI items since it is a SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 requirement of A/LAI No. 8. recommends that license renewal applicants for PWRs provide their C-3

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  1. Summary of Comment Response ID responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In doing this, the explicit reference to an inspection plan is avoided in the body of the AMP, and inspection plan is only referenced as part of A/LAI No. 8 in the staffs SE, Revision 1, for MRP-227.

However, the staff does not agree with the Commenters general claim with respect to what satisfies an inspection plan per A/LAI No. 8, as additional guidance is outlined in the SE, Revision 1, for MRP-227, and fulfillment of that action item will depend on each applicants plant-specific review.

4 I-4 The stipulation of appropriate inspection methodologies for these The staff agrees with the comment, in part, that final LR-ISG-2011-04 reactor internals components has already been addressed in the address the acceptability of VT-3 as a management approach for certain review of MRP-227-A. The recommended inspection methods have components. Thus, final LR-ISG-2011-04 does not incorporate SRP-LR already been reviewed and found to be adequate to detect the Sections 3.1.2.2.9.A.7, 3.1.2.2.9.C.1, and 3.1.2.2.9.C.4. However, the relevant conditions. The AMP attribute that is at issue is not staffs position on the use of VT-3 to detect cracking will continue to be detection of aging effects; instead, the issue is the applicants documented in the Detection of Aging Effects program element in GALL corrective action program, and the disposition of relevant conditions Report AMP XI.M16A, which states, in part, the following:

through supplemental examination or engineering evaluation, both of which are outside the scope of the Mandatory or Needed ...VT 3 visual methods may be applied for the detection of cracking in non-requirements of MRP-227-A. Standards for engineering evaluation redundant RVI components only when the flaw tolerance of the component, are addressed in Section 6 of MRP-227-A and in the methodologies as evaluated for reduced fracture toughness properties, is known and the described in WCAP-17096. These recommendations are based on component has been shown to be tolerant of easily detected large flaws, the practice used in Section XI of the ASME code and are consistent even under reduced fracture toughness conditions.

with existing aging management programs. Further justification for the use of the VT-3 examination is not necessary and should not be required by the ISG.

It is recommended that Acceptance Criteria Item 3.1.2.2.9.A.7 (Use of VT-3 Visual Inspection Techniques for Detection of Cracking) be completely eliminated and replaced by a limited requirement to address the acceptability of VT-3 as a management approach for components that 1) were not already considered for aging management in the development of MRP-227-A, 2) are evaluated to require active aging monitoring, and 3) are non-redundant. The Commenter provided justification for its recommendation.

5 I-5 Appendix A - Section 2, Acceptance Criteria Item 3.1.2.2.9.A.9 The staff agrees with the comment, in part, that the evaluation of (Identification of TLAAs for PWR-Design RVI Components) on Page environmental effects for PWR RVI core support structures should not be A-20 and A-21 stipulates that, in order to satisfy the requirements of incorporated in SRP-LR Section 3.1.2.2.9.A.9 in final LR-ISG-2011-04.

the ASME Code,Section III, Subsections NG-2160 and NG-3121, However, the staff does not agree with the commenters statement that the license renewal applicants demonstrating acceptability of RVI evaluation of time-limited aging analyses for the reactor internals should be components with design-basis cumulative usage factor (CUF) addressed in accordance with the existing 10 CFR Part 54 requirements analyses that are TLAAs should include the effects of the reactor without the need to include environmental effects.

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  1. Summary of Comment Response ID coolant system water environment in the fatigue CUF analyses.

The Commenter provided its justification for removal of this last As a result of the staffs resolution of Source ID I-1, areas resolved in the sentence. staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. Final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of the staffs resolution of Source ID I-1, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.9.

To the extent that the commenter does not agree with the need to address evaluation of environmental effects, the staffs SE, Revision 1 for MRP-227-A documents the basis for limitations and conditions being placed on the use of MRP-227 as well as A/LAIs that shall be addressed by applicants/licensees who choose to implement the NRC-approved version of MRP-227. Specifically, the topic of environmentally-assisted fatigue for PWR RVIs is addressed in A/LAI No. 8, Item 5 of MRP-227-A. Thus, the intent of LR-ISG-2011-04 is not to supplement or modify the evaluation in the staffs SE, Revision 1.

6 I-6 The component-specific AMR items described in Appendix-A, The staff does not agree with the comment recommending that NUREG-Sections 4, 5 and 6 are based on migration from NUREG-1801. As 1801 refer existing AMR line items to the applicable MRP-227-A table and a result the listing is more complex than the approved MRP-227-A retain detail only for those items which may be beyond the scope of MRP-tables. For example, there are approximately 25 items in Section 5 227-A.

that classify as Primary component examinations, whereas the equivalent component list in MRP-227-A contains only 13 items. In accordance with 10 CFR 54.21(a)(3) for each structure and component The component content is very similar but the breakdown is identified as part of the integrated plant assessment (IPA), the LRA is to complex. A key advantage of aligning license renewal commitments demonstrate that the effects of aging will be adequately managed so that to the MRP-227-A format is to facilitate important, industry-wide the intended function(s) will be maintained consistent with the current program updates based on Operating Experience through the NEI licensing basis for the period of extended operation. The IPA is 03-08 process. The alignment between MRP-227-A and NUREG- independent of the line items in MRP-227-A and the GALL Report and may 1801 is compromised by embedding item detail in the ISG format. It also result in additional components beyond the generic lists in these is recommended that NUREG-1801 refer existing AMR items to the documents. This requires that the LRA provide a complete listing of AMR applicable MRP-227-A table and retain detail only for those items line items, which may include items consistent with MRP-227-A and the which may be beyond the scope of MRP-227-A. This will GALL Report and may also result in additional components beyond the significantly reduce applicant and NRC staff burden, and improve generic lists in these documents. Thus, the number and content of AMR integration of evolutionary changes through the NEI 03-08 process. line items in the inspection tables of MRP-227-A are not the only basis for determining the AMR line items in the GALL Report. Similarly, the AMR line items in the GALL Report are not the only basis for determining the aging effects requiring management for components or establishing the AMR line items that are included in an LRA.

However, final LR-ISG-2011-04 incorporates revisions to SRP-LR Table 3.1-1 and GALL Tables IV.B2, IV.B3, and IV.B4 as summarized in the C-5

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  1. Summary of Comment Response ID staffs resolution to Source ID I-1.

7 I-7 ISG implementation of Applicant/Licensee Action items from the The staff agrees with the comment that associating A/LAIs with each MRP-227-A SER is by way of notes to AMR items listed in Sections individual AMR line item increases the burden for both the applicant and 4, 5 and 6. This could be addressed by reference to the appropriate NRC staff reviewer.

SER action items. It is recommended that the required evaluations would be documented in a single location specified by the ISG As part of the resolution to Source ID I-1, final LR-ISG-2011-04 rather than associated with individual items. Associating these incorporates revisions to SRP-LR Table 3.1-1 and GALL Tables IV.B2, actions with each individual AMR item increases the burden for both IV.B3, and IV.B4. Specifically, GALL Tables IV.B2, IV.B3, and IV.B4 were the applicant and NRC staff reviewer. revised to be consistent with the format of AMR items in the GALL Report for non-RVI components and the footnotes in the Further Evaluation column of these tables were deleted.

8 I-8 The draft ISG requires Applicants to develop and submit evaluation The staff agrees with the comment that it is not necessary to provide an of inaccessible Reactor Vessel Internal components in accordance evaluation of inaccessible RVI components, with the exception of A/LAI with Note 3 to Sections 4 and 5, and Note 2 to Section 6. With the No. 6 of MRP-227-A. As part of the resolution to Source ID I-7, final LR-exception of A/LAI #6 of the MRP-227-A SER, these evaluations ISG-2011-04 incorporates revisions to delete the further evaluation have been addressed during review and approval of the Industry footnotes from GALL Tables IV.B2, IV.B3, and IV.B4.

program. The requirement to develop, submit and review the inspection basis is unnecessary. It is recommended that this note As a result of staffs resolution to Source ID I-1, areas resolved in the staffs be eliminated. SE, Revision 1, for MRP-227 and A/LAIs are not redundantly addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04.

9 I-9 MRP-227-A provides applicants with an alternative to the defined The staff agrees with the comment that LR-ISG-2011-04 refer to MRP-227-inspection requirements when plant-specific analyses of A and the associated A/LAI discussions for alternatives or deviations to the accumulated fatigue usage are performed. Applicants may choose inspection and evaluation guidelines in MRP-227-A.

to either inspect in accordance with the approved MRP-227-A schedules, or perform analyses. In cases where Applicants perform It is the responsibility of the license renewal applicant to demonstrate in analyses to relax MRP-227-A requirements, those analyses would accordance with 10 CFR 54.21(a)(3) that it can adequately manage aging be submitted for NRC staff approval in accordance with A/LAI 8. of RVIs for the period of extended operation, whether through the use of The ISG is unclear regarding these alternatives. For example item MRP-227-A or alternatives. If a TLAA exists for a RVI, in accordance with IV.B3.RP-343 appears to require physical examinations to support 10 CFR 54.21(c)(1)(iii), an applicant may choose to demonstrate the acceptance of the TLAA. The industry recommends that the ISG effects of aging on the intended function of the component will be refer to MRP-227-A and the associated A/LAI requirement adequately managed for the period of extended operation. It is incumbent discussions. on the license renewal applicant to provide this demonstration of aging management, which can include the use of MRP-227-A or an appropriate alternative.

In order to avoid redundancy, areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed again in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA.

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  1. Summary of Comment Response ID 10 I-10 Item 9.B.1 of the ISG notes that Section 3.2.5.3 of the NRC SE The staff agrees with the comment that there is an inconsistency between (Revision 1) on MRP-227 Revision 0 recommends that the applicant SRP-LR Section 3.1.2.2.9.B.1 in draft LR-ISG-2011-04 and Section 3.2.5.3 consider replacement or inspection activities with regard to the of the staffs SE, Revision 1, for MRP-227. As part of the staffs resolution Control Rod Guide Tube (CRGT) split pins if the pins are currently to Source ID I-1, areas resolved in the staffs SE, Revision 1, for MRP-227 fabricated with Alloy X-750 or Type 316 stainless steel material. A and A/LAIs are not addressed in the Further Evaluation sections of the review of the referenced section of the SE does not reach the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 conclusion that this specificity of action is required; the SE recommends that license renewal applicants for PWRs provide their requirement is to evaluate the adequacy of the plant-specific responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a existing program to ensure that the aging degradation is adequately result of the staffs resolution of Source ID I-1, final LR-ISG-2011-04 does managed during the extended period of operation. The SE direction not incorporate SRP-LR Section 3.1.2.2.9.B.1.

is on evaluation of the performance of the existing program and does not suggest that it should be changed to include inspections. Specific to the Westinghouse CRGT split pins, A/LAI No. 3 recommends an Therefore the industry considers the specificity of direction provided evaluation to consider the need to replace the Alloy X-750 split pins, if in the SE to be sufficient and the ISG should not provide alternate applicable, or an inspection of the replacement type 316 stainless steel direction. split pins to ensure that cracking has been mitigated and that aging degradation is adequately monitored during the extended period of operation. Thus, the intent of LR-ISG-2011-04 is not to supplement or modify the evaluation in the staffs SE, Revision 1, but rather, to recommend that the response to A/LAI No. 3 of MRP-227-A be appropriately documented in Appendix C of the LRA.

11 I-11 Section C.3, page A23 of LR-ISG 2011-04 states that per MRP-227- The staff agrees with the comment that the discussion related to A, ...EVT-1 inspections of certain CE-design components would be CE-designed lower core flange welds, core support plates, and fuel necessary if the design basis fatigue TLAAs for the components alignment plates in SRP-LR Section 3.1.2.2.9.C.3 in draft LR-ISG-2011-04 could not demonstrate that fatigue-induced cracking would be is not clear.

adequately managed... This statement does not accurately represent MRP-227-A Table 4-2, because it assumes that the As a result of the staffs resolution of Source ID I-1, areas resolved in the fatigue evaluations required by the MRP-227-A table item already staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the exist and are part of the current licensing basis, and therefore are Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a formally classifiable as TLAAs. In fact, many, if not all, of the older result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-CE design reactor internals were not qualified to the fatigue rules of LR Section 3.1.2.2.9.C.3.

ASME III, so TLAAs as defined in 10 CFR Part 54 do not exist.

Further, page A24 of the draft ISG states Otherwise, CE-design 10 CFR 54.21(a)(1) requires that license renewal application contain an applicants for renewal are requested to credit the MRPs EVT-1 IPA that must, for those systems, structures, and components within the basis in MRP-227-A as the applicable aging management basis if scope of Part 54, identify and list those structures and components subject either: (1) the CLB does not include applicable CUF or It fatigue to an aging management review (AMR). The components evaluated in analyses for these components; This statement appears to MRP-227-A may not fully encompass the components identified in an IPA, compel the applicant who does not have a current licensing basis as required by 10 CFR 54.21(a)(1), and therefore, should not be TLAA to perform EVT-1 inspections. MRP-227-A clearly does not considered a substitute for performance of an IPA.

require inspections based solely on the lack of a current licensing basis TLAA. In fact, it only requires that a fatigue evaluation be The aging effects requiring management for RVIs are not governed by C-7

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  1. Summary of Comment Response ID performed to determine if a fatigue issue might exist; and if so, MRP-227-A; rather the content of MRP-227-A serves to assist a PWR where would inspection be focused to manage it. The method of the license renewal applicant. In accordance with 10 CFR 54.21(a)(3), the fatigue evaluation was intended to be the usual engineering effects of aging are to be managed for all applicable aging effects for a practice, for example by comparison of the number expected particular component, which may be broader than the aging effects operating transient cycles to those specified by design, or by stress identified in MRP-227-A and the GALL Report for RVIs. It is the analysis if required. responsibility of the license renewal applicant to demonstrate that it can adequately manage aging of RVIs for the period of extended operation, whether through the use of MRP-227-A or alternatives.

Therefore, if the CE-designed lower core flange welds, core support plates, and fuel alignment plates are subject to an AMR and fatigue is an applicable aging effect, regardless if there is a TLAA, the LRA must demonstrate that fatigue will be adequately managed in accordance with 10 CFR 54.21(a)(3).

12 I-12 For A/LAI No. 2, when comparing the licensee renewal AMR from The staff agrees with the comment to include an AMR line item for cracking BAW-2248A to the tables in MRP-189, the locking devices for the of B&W vent valve locking devices made from Alloy 600 materials in GALL vent valve were identified as a possible Primary component. The Table IV.B4 of draft LR-ISG-2011-04. Final LR-ISG-2011-04 incorporates original vent valves located next to outlet nozzles failed due to flow the core support shield vent valve top and bottom retaining rings to be induced vibration, and those valves next to the nozzles were managed for cracking in GALL AMR Item IV.B4.RP-252a.

replaced with locking devices made containing Alloy 600.

It is recommended that Table IV Reactor Vessel, Internals, and Reactor Coolant System, B4 Reactor Vessel Internals (PWR) -

Babcock and Wilcox on page A-124 of LR-ISG 2011-04 be revised to include a line item addressing Alloy 600 replacement vent valve locking devices, which are subject to aging degradation due to primary water stress corrosion cracking (PWSCC).

13 I-13 In Item 8 on page A-11 of the LR-ISG, the second sentence appears The staff agrees with the comment that the sentence in the Confirmation to be incomplete with respect to the statement pertaining to Process program element in GALL Report AMP XI.M16A of draft confirming that the quality of inspections, flaw evaluations, and LR-ISG-2011-04 is incomplete. Final LR-ISG-2011-04 completes this corrective actions performed under this program. It is sentence in the Confirmation Process program element.

recommended that the revised statements be reviewed for completeness.

14 I-14 Item 3 on page A-16 of the LR-ISG should reference NRC SE The staff agrees with the comment that SRP-LR Section 3.1.2.2.9.A.3 in Section 3.2.5.1 and not Section 3.5.1. It is recommended that this draft LR-ISG 2011-04 should reference NRC SE Section 3.2.5.1 and not reference be revised. Section 3.5.1.

As a result of the staffs resolution of Source ID I-1, areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a C-8

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  1. Summary of Comment Response ID result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.3.

15 I-15 Item D.1 on page A-25 discusses evaluation Acceptance Criteria The staff agrees with the comment to use the terminology stress relief recommendations applicable to Babcock and Wilcox reactor process consistently throughout SRP-LR Section 3.1.2.2.9.D.1 of draft internals. In general, A/LAI 4 is not specific relative to the wording LR-ISG-2011-04. Final LR-ISG-2011-04 does not use the term post-weld for the manner in which the items were stress relieved, and it was heat treatment and this term is replaced with the term stress relief stated that a stress relief process was used. In Item D.1, the process. In addition, as a result of the staffs resolution of Source ID I-1, wording used in some cases implies a post-weld heat treatment areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are process. The words stress relief process should be used not addressed in the Further Evaluation sections of the SRP-LR in final LR-consistently without the implication of a heat treatment process only. ISG-2011-04. As a result of these revisions, final LR-ISG-2011-04 does In addition, the requirements in Item D.1 appear to go beyond the not incorporate SRP-LR Section 3.1.2.2.9.D.1.

requirements of the A/LAI as it was written and approved by the MRP-227-A SER.

16 II-1 Page A IV.B2.RP-300 The staff agrees with the comment that the possibility of thermal embrittlement for Type 403 martensitic stainless steel hold down springs is Alignment and interfacing components, such as hold down springs, not addressed. Final LR-ISG-2011-04 does not use the term (Aust. SS are addressed in MRP-227-A. Based on MRP-227-A, the intent of Material) in the Material column in GALL AMR Item IV.B2.RP-300.

the GALL was only to apply to hold down springs made from Type Furthermore, the use of the term Stainless Steel in GALL AMR Item 304 Stainless Steel (SS). The possibility of thermal embrittlement of IV.B2.RP-300 is generic and includes all grades of stainless steel as hold down springs made from Type 403 martensitic SS is not defined in GALL Table IX.C, Selected Definitions & Use of Terms for addressed. The issue is, however, discussed in the proposed SRP Describing and Standardizing - MATERIALS. With these revisions hold section 3.1.2.2.9.A.6 and in applicant action item 7 of the SER down springs made from Type 403 martensitic SS are addressed in GALL (Revision1). AMR Item IV.B2.RP-300.

Proposed Change: Include the words applicable to hold down springs fabricated from Type 304 SS and add a line item to address thermal embrittlement for hold down springs fabricated for Type 403 stainless steel.

17 II-2 Page A Section 3.1.2.2.9.A.3, second paragraph The staff does not agree with the comment, in particular the inference that, unless a utility implemented modifications beyond that recommended by There is little guidance on Applicant Action Item #2 related to the vendor of the RVI, all of the piece parts of the RVI were considered additional RVI piece parts and what was used during the during the development of MRP-189, 191 and 227-A. The methodology development of MRP 191. Utilities are left to draw a conclusion that and results of a topical report, such as MRP-227-A, cannot be assumed to unless the utility implemented a modification beyond the vendor's be generically bounding for every plant.

recommendation, all of the piece parts in the reactor vessel were considered during the development of MRP-189, 191 and 227-A. The IPA described in the response to Source ID I-11 is a plant-specific evaluation performed by a license renewal applicant. Thus, the Proposed Change: Add verbiage to provide additional guidance to components evaluated in MRP-189, 191 and 227-A may not fully allow utilities to make the assumption that unless a utility encompass the components identified in an applicants IPA and therefore, implemented modifications beyond that recommended by the should not be considered a substitute for performance of an IPA. The C-9

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  1. Summary of Comment Response ID vendor of the RVI, then all of the piece parts of the RVI were aging effects requiring management for RVIs may be broader than the considered during the development of MRP-189, 191 and 227-A. aging effects identified in MRP-189, 191 or 227-A. It is the responsibility of the license renewal applicant to demonstrate, in accordance with 10 CFR 54.21(a)(3), that it can adequately manage aging of RVIs for the period of extended operation, whether through the use of MRP-227-A or alternatives. The content in MRP-189, 191 or 227-A only serves to assist a PWR license renewal applicant.

However, as addressed in the staffs resolution to Source ID I-1, in order to avoid redundancy, areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.3.

18 II-3 Page A Section 3.1.2.2.9.A.3 The staff agrees with the comment that portions of SRP Section 3.1.2.2.9.A.3 are confusing.

The words in the third paragraph are confusing and it is not clear what is meant by plant specific AMR line items or why Note E would As a result of the staffs resolution of Source ID I-1, areas resolved in the be appropriate. For those applicants whose plant-specific review staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the results in identification of additional components for inspection or Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a different component inspection categories from those identified in result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-MRP-227-A, the applicant is requested to identify the changes in the LR Section 3.1.2.2.9.A.3.

component inspection categories as either plant-specific AMR line items or NEI Note E consistent with GALL AMR items (whichever is applicable) in their Table 2 AMR line items for their PWR RVI components.

Proposed Change: It is suggested that if only a component line item or two that is not in GALL is being added then an exception can be taken to the program and justification be added that includes inspection specifics such as method and acceptance criteria such that the whole program doesnt have to be evaluated as a plant specific program.

19 II-4 Page A last paragraph The staff agrees that draft LR-ISG-2011-04 does not provide clear direction Page A Section 3.1.2.2.9.1.2 as to what goes into an inspection plan but does not agree with the commenters proposed change. The staff does not agree with the The document does not provide clear direction as to what goes into Commenters general claim with respect to what satisfies an inspection an inspection plan. plan per A/LAI No. 8, as additional guidance is outlined in the SE, Revision 1, for MRP-227, and fulfillment of that action item will depend on each C-10

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  1. Summary of Comment Response ID Proposed Change: Add verbiage to allow utilities to better applicants plant-specific review.

determine what the inspection plan should consist of (e.g., A Westinghouse design plant should provide unit specific information See the staffs resolution to Source ID I-3. RIS 2011-07 provides the staffs in the Inspection Plan consistent with tables 4-3, 4-6 and 4-9 of expectations for Category D plants (PWR plant licensees that have not MRP-227-A and the A/LAIs). submitted their LRAs but plan to submit an LRA in the future) to submit, for NRC staff review and approval, an AMP for vessel internals that is consistent with MRP-227-A. As an inspection plan is one aspect of satisfying A/LAI No. 8 of the staffs SE, Revision 1, for MRP-227. An inspection plan provides information about the RVI components to be inspected and a description of how they will be managed for age-related degradation (e.g., examination method, frequency, acceptance criteria, coverage, etc.). The staff expects that the details of an inspection plan for Category D plants will be incorporated into the LRA submittal as part of the 10-element AMP and AMR line items. Thus, consistent with RIS 2011-07, the staff does not expect Category D plants to provide a separate document that contains an inspection plan in response to A/LAI No. 8.

To avoid confusion, final LR-ISG-2011-04 avoids explicit reference to an inspection plan in the body of the AMP, and inspection plan is only referenced as part of A/LAI No. 8 in the staffs SE, Revision 1, for MRP-227.

20 II-5 Page A Table 3.1-1 Item 27a The staff agrees with the comment that Table 3.1-1, Item 27a, of draft LR-ISG-2011-04 does not clearly address Type 304 stainless steel hold It is not clear that this line item is only applicable to hold down down springs.

springs fabricated from Type 304 SS.

Final LR-ISG-2011-04 does not include this item, but Westinghouse Type Proposed Change: Add Type 304 SS hold down springs. 304 stainless steel hold down springs were incorporated into Table 3.1-1, Item 59a, in final LR-ISG-2011-04, which uses the generic terminology stainless steel.

21 II-6 Page A Table 3.1-1 Item 3 It refers to the parameter being calculated for the cyclical loading analyses. In later editions of the ASME Code Section III, these analyses Under 'Further Evaluation Recommended' column, it is not clear were referred to as cumulative usage factor (CUF) analyses. Thus, the It what It stands for? parameter is analogous to the CUF parameter required for Class 1 components designed to more recent editions of the ASME Code, Section Proposed Change: Provide an explanation. III. The subscripted t was removed in the formatting during the issuance of draft LR-ISG-2011-04 for public comment.

As a result of the staffs resolution of Source ID I-1, SRP-LR Table 3.1-1 Item 3 does not incorporate the reference to the It parameter in final LR-ISG-2011-04.

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  1. Summary of Comment Response ID 22 II-7 Page A IV.B2.RP-280 Page 3-11 of MRP-227-A states that [t]he lower support forging is welded to and supported by the core barrel, which transmits vertical loads to the There is confusion regarding what comprises the lower core barrel vessel through the core barrel flange. In addition, Table 5-3 of MRP-227-flange weld for Westinghouse designed plants. This component is A provides the Examination Acceptance Criteria and Expansion Criteria still listed in MRP-191, and 227-A for Westinghouse designed for the Core Barrel Assembly - Lower core barrel flange weld. Footnote 2 plants. MRP-227-A indicates it may be the weld between the core of Table 5-3 states that [t]he lower core barrel flange weld may barrel and the lower support forging or casting. alternatively be designated as the core barrel-to-support plate weld in some Westinghouse plant designs.

Proposed Change: Provide an explanation regarding what this component is. No revisions were made as a result of this comment.

23 II-8 Page A Section 3.1.2.2.9.A.4 The staff agrees with the comment that the referenced text in SRP-LR Section 3.1.2.2.9.A.4 of draft LR-ISG-2011-04 is not clear. As a result of In the subject paragraph, it appears the NRC wanted an exception the staffs resolution of Source ID I-1, areas resolved in the staffs SE, not an enhancement: For those component inspections that do not Revision 1, for MRP-227 and A/LAIs are not addressed in the Further achieve the inspection coverage criteria stated in the NRC SE Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, (Revision 1) on MRP-227, the applicant is requested to take a final LR-ISG-2011-04 recommends that license renewal applicants for deviation from the MRP-defined inspection criteria and describe the PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C process and type of evaluation that will be implemented to evaluate of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not the impact of the aging effects on the inaccessible regions of the incorporate SRP-LR Section 3.1.2.2.9.A.4.

components. In this case, the applicant is requested to identify this process as an applicable enhancement of the monitoring and trending program element of its RVI Program.

Proposed Change: Clarify what is expected.

24 II-9 Page A IV.B2.RP-280 The staff agrees with the comment that the applicability of Note 3 to GALL AMR Item IV.B2.RP-280 is not clear. As a result of the staffs resolution of It is not clear how Note 3 in the Further Evaluation column is Source ID I-1, Final LR-ISG-2011-04 incorporates revisions to GALL applicable to this GALL Line Item. Tables IV.B2, IV.B3, and IV.B4 as summarized above.

Proposed Change: Clarify the applicability.

25 II-10 Page 3 The staff agrees with the comment that Tables 3-2 and 3-3 should be referenced in the last paragraph of the Discussion section. However, the In the last paragraph of the Discussion section only table 3-1 is Discussion section of final LR-ISG-2011-04 no longer references Table 3-listed for justification of TE for the materials. Tables 3-2 and 3-3 1 in MRP-227-A.

should be mentioned since 3-1 is only for B&W internals.

Proposed Change: Add Tables 3-2 and 3-3.

26 II-11 Page A-7 The staff agrees with the comment to change the terminology to Aging Management Requirement tables in the Parameters Monitored/Inspected C-12

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  1. Summary of Comment Response ID The second paragraph in this Section refers to condition monitoring program element. The Parameters Monitored/Inspected program tables in MRP-227-A. There are no tables with this title in MRP- element in final LR-ISG-2011-04 states the following:

227-A Specifically, the program implements the parameters monitored/inspected Proposed Change: Change to Aging Management Requirement criteria consistent with the applicable tables in Section 4, Aging tables. Management Requirement, in MRP-227-A 27 II-12 Page A-9 The staff agrees with the comment to add references to Table 5-2 and 5-3 of MRP-227-A for the Acceptance Criteria program element. The Only Table 5-1 is listed for acceptance criteria when MRP-227-A Acceptance Criteria program element of GALL Report AMP XI.M16A in contains three tables, 5-1 thru 5-3 final LR-ISG-2011-04 references Table 5-1 through 5-3 of MRP-227-A.

Proposed Change: Change to read Section 5 and Tables 5-1 thru 5-3 of MRP-227 28 II-13 Page A-10 The staff agrees with the comment that Section 3.3.5 of MRP-227-A does not specify acceptance criteria for physical measurements. However, as The first paragraph on the page says The program adopts the a result of the staffs resolution of Source ID I-1, areas resolved in the acceptance criteria for the physical measurement monitoring staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in methods recommended in MRP-227-A, as qualified in Section 3.3.5 GALL Report AMP XI.M16A in final LR-ISG-2011-04.

and A/LAI No. 5 in Revision 1 of the NRC SE on MRP-227. Section 3.3.5 of the MRP does not specify acceptance criteria so there is The Acceptance Criteria program element in final LR-ISG-2011-04 nothing to be adopted. It only requires it be developed as discussed states that, in general, the AMP establishes appropriate acceptance in footnote 3. criteria for any physical measurement monitoring methods that are credited for aging management of RVIs.

Proposed Change: Change sentence to read The program includes acceptance criteria for the physical measurement monitoring methods as recommended in MRP-227-A, Section 3.3.5 and A/LAI No. 5 in Revision 1 of the NRC SE on MRP-227.

29 II-14 Page A-12 The staff agrees with the comment that the sentence associated to the notification criteria already exists in the Administrative Controls program The following sentence relates to notification criteria: The element and does not need to be repeated in the Operating Experience evaluation in Section 3.5 of Revision 1 of the SE on MRP-227 program element of GALL Report AMP XI.M16A. The Operating provides the staffs basis for endorsing the NEI 03-08 Experience program element of GALL Report AMP XI.M16A in final LR-implementation process for these programs. This includes NRCs ISG-2011-04 does not incorporate this sentence associated with the endorsement of the NEI 03-08 criteria for notifying the NRC of any notification criteria.

deviation from the I&E methodology in MRP-227-A and justification of the deviation no later than 45 days after approval by a licensee executive.

Proposed Change: Delete this sentence as it already is discussed in element 9 where it is appropriate.

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  1. Summary of Comment Response ID 30 II-15 Page A-8 The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER The justification required for the use of VT-3 to detect cracking over and MRP are met. As a result of the staffs resolution of Source ID I-4, final that specified in MRP-227A and approved by the staff in the SE that LR-ISG-2011-04 does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, allows its use without the additional limitations and analyses is not 3.1.2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staffs position on the use of needed. VT-3 for the detection of cracking will continue to be documented in the Detection of Aging Effects program element in GALL Report AMP Proposed Change: Eliminate need for additional justification if XI.M16A.

requirements as specified in SER and MRP are met.

31 II-16 Page A Section 3.1.2.2.9.C.1 The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER The justification required for the use of VT-3 to detect cracking over and MRP are met. As a result of the staffs resolutions to Source ID I-4 that specified in MRP-227A and approved by the staff in the SE that and ID II-15, final LR-ISG-2011-04 does not incorporate SRP-LR Sections allows its use without the additional limitations and analyses is not 3.1.2.2.9.A.7, 3.1.2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staffs needed. position on the use of VT-3 for the detection of cracking will continue to be documented in the Detection of Aging Effects program element in GALL Proposed Change: Eliminate need for additional justification if Report AMP XI.M16A.

requirements as specified in SER and MRP are met.

32 II-17 Page A Section 3.1.2.2.9.C.3 The staff agrees with the comment that the discussion related to CE-designed lower core flange welds, core support plates, and fuel alignment The option presented as (3), as an alternative basis for accepting plates in SRP-LR Section 3.1.2.2.9.C.3 of draft LR-ISG-2011-04 is not the design basis fatigue analyses in accordance with the TLAA clear.

acceptance requirement in 10 CFR 54.21(c)(1)(iii) does not make sense when compared to options 1 and 2 As a result of the staffs resolution of Source ID I-1, areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Proposed Change: Add the word the EVT-1 is used at the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a beginning result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.C.3.

Also see the staffs resolution to Source ID I-11, in which the staff clarified that, if the CE-designed lower core flange welds, core support plates, and fuel alignment plates are subject to an AMR and fatigue is an applicable aging effect, regardless if there is a TLAA, then in accordance with 10 CFR 54.21(a)(3), the LRA must demonstrate that fatigue will be adequately managed.

33 II-18 Page A Section 3.1.2.2.9.D.1 The staff agrees with the comment that there is not a need for a plant-specific enhancement of the Preventive Actions program element There is no need for a plant-specific enhancement of the discussed in SRP-LR Section 3.1.2.2.9.D.1 of draft LR-ISG-2011-04, which C-14

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  1. Summary of Comment Response ID preventative actions program element for their RVI Program is associated with A/LAIs No. 4 of MRP-227-A.

enhancement to be identified if an applicant confirms that the welds were appropriately stress-relieved. An enhancement doesn't seem As a result of the staffs resolution of Source ID I-1, areas resolved in the appropriate since the action has already been taken. staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In Proposed Change: Eliminate the need for an enhancement addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.D.1.

34 II-19 Page A Table 3.0-1 The staff agrees with the comment that the further evaluation acceptance criteria do not need to be specified as part of a Safety Analysis Report There is no need for the words or to applicable NRC further description. Final LR-ISG-2011-04 does not incorporate this second evaluation acceptance criteria recommendations in Section 3.1.2.2 paragraph in the Description of Program column for GALL Report AMP of the SRP-LR (i.e., the latest NRC issued version of NUREG- XI.M16A in SRP-LR Table 3.0-1. However, 10 CFR 54.21(d) provides the 1800). Specific acceptance criteria do not need to be part of a SAR requirements for a Final Safety Analysis Report supplement and states, in description. If it is an enhancement it will already be a commitment. part, that it must contain a summary description of the programs and activities for managing the effects of aging. The specificity of such Proposed Change: Delete descriptions will depend on the program proposed by each license renewal applicant.

35 II-20 Page A Table IV.B2 The staff agrees with the comment that there is no need for specifying the Examination Technique in the Aging Management Program column of There is no need for specifying the Examination technique in the GALL Table IV.B2. GALL Tables IV.B2, IV.B3 and IV.B4 in final Program column. LR-ISG-2011-04 do not incorporate a summary of the examination techniques from the Aging Management Program column.

Proposed Change: Delete 36 II-21 A Footnotes The staff agrees with the comment that additional justification for the use of VT-3 to detect cracking is not needed if requirements specified in the SER For note 6, see comments 15 and 16 above on why no justification and MRP are met. As a result of the staffs resolutions of Source ID I-4 for using VT-3 exam is required when it was acceptable in SER for and ID II-15, final LR-ISG-2011-04 does not incorporate SRP-LR Sections 227. This applies to CE and B&W tables that also contain a similar 3.1.2.2.9.A.7, 3.1.2.2.9.C.1, and 3.1.2.2.9.C.4. However, the staffs note. position on the use of VT-3 for the detection of cracking will continue to be documented in the Detection of Aging Effects program element in GALL Proposed Change: Delete the note Report AMP XI.M16A.

In addition, as part of the staffs resolution to Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-RVI components and does not incorporate the footnotes in the Further Evaluation column of these tables.

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  1. Summary of Comment Response ID 37 II-22 Page A-102 - Footnote #1 The staff agrees with the comment that in conjunction with was an editorial error in Note 1. However, as part of the staffs resolution to In conjunction is repeated in the second sentence. Source ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-Proposed Change: Delete second in conjunction RVI components and does not incorporate the footnotes in the Further Evaluation column of these tables. As a result of these revisions, the referenced Note 1 is not incorporated in final LR-ISG-2011-04.

38 II-23 Page A-104 - Footnote #8 The staff agrees with the comment that there is a typographical error in Note 8 of page A-103. However, as part of the staffs resolution to Source 4th line No.2 above, and is so should be and if so. ID I-8, the format of GALL Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR items in the GALL Report for non-Proposed Change: Correct RVI components and does not incorporate the footnotes in the Further Evaluation column of these tables. As a result of these revisions, the referenced Note 8 is not incorporated in final LR-ISG-2011-04.

39 II-24 Page A Table IV.B2 The staff agrees with the comment that GALL Report AMP X.M2, Water Chemistry, is not listed in GALL Table IV.B2. GALL Table IV.B2, IV.B3 Water chemistry is not listed as an AMP, with the aging effect of and IV.B4 of final LR-ISG-2011-04 include GALL Report AMP X.M2, stress corrosion cracking (SCC) and irradiation-assisted stress Water Chemistry, as a recommendation to manage cracking by SCC, corrosion cracking (IASCC) such as in line items IV.B2.RP-270a, PWSCC, or IASCC, or loss of material due to pitting or crevice corrosion of 345, 399, 299a. This mainly occurs in new line items and also RVIs.

shows up in Table IV.B3 and IV.B4 Proposed Change: Add XI.M2 as an AMP 40 II-25 Page A IV.B2.RP-399 The staff agrees with the comment that Table 4-9 of MRP-227-A did not identify cracking as an aging effect requiring management for As indicated in Table 4-9 of MRP-227-A and the associated note 2, Westinghouse-design clevis insert bolts of screws but does not agree with the clevis insert bolts are inspected for wear. To the extent cracking the commenters proposed change.

would be visible in the VT-3 inspection, it would of course be addressed; but, the intent of the inspection is to look for wear. Relevant operating experience associated with aging may exist that has not been accounted for in MRP-227-A. AMR item IV.B2.RP-399 for Proposed Change: Eliminate this line as an existing inspection cracking of Westinghouse-design clevis insert bolts and screws was program element, or change the AMP description to note the included in LR-ISG-2011-04 based on industry operating experience.

inspection is for gross effects of cracking Appendix A of MRP-227-A states, in part, that [f]ailures of Alloy X-750 clevis insert bolts were reported by one Westinghouse-designed plant in 2010 and [a]lthough the failed clevis insert bolts were not removed for metallurgical examination, it can be surmised that the most likely cause of failure was PWSCC. No revisions were made as a result of this comment.

41 II-26 Page A IV.B2.RP-285 The staff agrees with the comment to delete the aging mechanism of loss of fracture toughness from AMR item IV.B2.RP-285. Since the clevis bolts C-16

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  1. Summary of Comment Response ID As described in MRP-191, the clevis bolts and inserts are not in a and inserts are not in a high flux region, GALL AMR Item IV.B2.RP-285 in high flux region and irradiation embrittlement is not a significant final LR-ISG-2011-04 does not incorporate the aging effect of loss of aging mechanism. As indicated in Table 4-9 of MRP-227-A and the fracture toughness due to neutron irradiation embrittlement.

associated note 2, the clevis insert bolts are inspected for wear.

Also, Note 5 is applied to the further evaluation column; however, As a result of the staffs resolution of Source ID I-8, the format of GALL Note 5 refers to reduction of fracture toughness due to thermal Tables IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with embrittlement in stainless steel components, while the material AMR items in the GALL Report for non-RVI components and deleted the listed for this line is nickel alloy. footnotes in the Further Evaluation column of these tables.

Proposed Change: Eliminate the aging mechanism of loss of fracture toughness from this line and remove note 5 from the further evaluation column.

42 II-27 Page A IV.B2.RP-345 The staff agrees with the comment to delete base metal cracking from GALL AMR Item IV.B2.RP-345 of draft LR-ISG-2011-04 since MRP-227-A As indicated in Table 5-1 of MRP-191, cracking of the core barrel identifies that the adjacent base metal is part of the examination coverage flange is a concern for the weld rather than the base metal. Table 4- for the Core Barrel Assembly - Lower core barrel flange weld.

3 specifically identifies the welds as primary components to be inspected for cracking. While inspections of the welds would identify Thus, GALL AMR Item IV.B2.RP-345 in final LR-ISG-2011-04 does not cracking in the adjacent base metal, separately adding cracking as reference cracking of the core barrel flange (base metal). GALL AMR an aging effect to the base metal as an existing component is not IV.B2.RP-345 continues to identify loss of material due to wear for the core consistent with MRP-227-A or existing inspections. barrel flange (base metal).

Proposed Change: Eliminate base metal cracking as an aging effect in this line.

43 II-28 Page A-9 The staff agrees with the comment that the sentence is incomplete. This sentence in the Monitoring and Trending program element of GALL Flaw evaluation methods include recommendations for flaw depth Report AMP XI.M16A in final LR-ISG-2011-04 is complete.

sizing and for crack growth determinations as well for performing applicable load limit. It should read growth determinations as well as for performing.

Proposed Change: Change to include missing as.

44 II-29 Page A-49 The staff acknowledges that the component type neutron pad assembly is not addressed in GALL Table IV.B2 or MRP-227-A. However, the intent of In the first paragraph under Systems, Structures, and Components this LR-ISG is not to supplement such aspects that are not covered in thermal shield assembly should be changed to thermal shield or MRP-227-A. Thus, no revisions were made as a result of this comment.

neutron pad assembly to address the newer Westinghouse plants.

Also, the component type neutron pad is not addressed in Table B2 If a PWR license renewal applicant identifies during the IPA that its plant or MRP-227. design contains a neutron pad assembly (instead of a thermal shield assembly) and is subject to an AMR, the license renewal applicant must C-17

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  1. Summary of Comment Response ID Proposed Change: Address recommended change. identify this assembly in its LRA and propose an adequate means of aging management.

45 II-30 Table IV.B2 The staff does not agree with the comment to note exceptions with regard to use the term neutron flux in GALL AMR items in the GALL Report.

The environment Reactor coolant and neutron flux is used for all line items/components in Table B2, however not all the components The GALL Report generically and conservatively assumes that PWR RVIs listed in Table B2 will experience a neutron fluence exceeding 1017 are exposed to an environment of reactor coolant and neutron flux n/cm2 (E>1MeV) at the end of the period of extended operation. regardless of the fluence level. The staff anticipates that applicants will The environment should be more specific based on the location address their plant-specific data in their IPA and identify appropriate AMR (fluence) of the components. items. No revisions were made as a result of this comment.

Proposed Change: The Table should note exceptions to the neutron fluence level.

46 II-31 Page A Table 3.1-1 Item 27 The component in SRP-LR Table 3.1-1, Item 27, which refers to control rod guide tube (CRGT) split pins (support pins), is applicable to both nickel Component was changed to nickel alloy guide tube support pins, alloy and stainless steel materials. SRP-LR Table 3.1-1 Item 27 in draft however associated Table B2 line items IV.B2.RP-355 and LR-ISG-2011-04 was removed and incorporated into Table 3.1-1 Item 53c IV.B2.RP-356 were changed to include both nickel alloy and in final LR-ISG-2011-04.

stainless steel.

Proposed Change: Clarify 47 II-32 A Last paragraph The staff agrees with the proposed change, however, as a result of the staffs resolution of Source ID I-1 the referenced sentence is not Sentence EPR MRP methodology left some... should be changed. incorporated in final LR-ISG-2011-04.

Proposed Change: Should read EPRI MRP methodology left some...

48 II-33 The following acronyms are used but not included in Appendix B of The staff agrees with the comment; however, draft LR-ISG-2011-04 was this ISG; CUF, NRC, SE, and USAR. revised to remove the full list of acronyms in LR-ISG-2011-04, Appendix B.

Final LR-ISG-2011-04, Appendix B, was revised to document the mark-up Proposed Change: Update Appendix B to include all acronyms. of changes to the GALL Report and SRP-LR. Acronyms in final LR-ISG-2011-04 are defined the first time they are used.

49 II-34 The page numbers for Appendix B are A-165 and A-166, the last The staff agrees with the comment and final LR-ISG-2011-04 includes the page of Appendix A is A-144. correct page numbers.

Proposed Change: Verify correct pagination.

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  1. Summary of Comment Response ID 50 II-35 Page A Section 3.1.2.2.9.A.5 The staff agrees with the comment that if an inspection frequency is not consistent with MRP-227-A, an exception must be identified and justified.

For re-inspection greater than 10 years, further evaluation is redundant and inconsistent with standard GALL AMR and AMP Furthermore, Section 4.0 of the staffs SE, Revision 1, for MRP-227 formatting and presentation. Inspection frequencies would be provides the Conditions And Limitations And Applicant/Licensee Plant-evaluated in AMP element 4 for consistency with MRP-227-A Specific Action Items, which specifically states that the re-examination chapter 4 primary, expansion, and existing components inspection frequency for Primary inspection category components shall be on a tables. If the inspection frequency is identified that is not consistent maximum 10-year interval, unless a plant-specific analysis providing with MRP-227-A Chapter 4 tables, an exception must be identified justification for an extended examination frequency is submitted to and and justified. approved by the NRC.

Proposed Change: Delete further evaluation 3.1.2.2.9.A item 5. As a result of the staffs resolution of Source ID I-1, areas resolved in the Item to be addressed by AMP element 4. staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.5.

51 II-36 Page A Section 3.1.2.2.9.A.7 The staff agrees with the comment that VT-3 inspection requirements should be addressed as part of GALL Report AMP XI.M16A.

For VT-3 Inspection, further evaluation is redundant and inconsistent with standard GALL AMR and AMP formatting and As a result of the staffs resolution of Source ID I-4, final LR-ISG-2011-04 presentation. VT-3 inspection requirements should be addressed as does not incorporate SRP-LR Sections 3.1.2.2.9.A.7, 3.1.2.2.9.C.1, and part of AMP element 3 for consistency with MRP-227-A 3.1.2.2.9.C.4 related to VT-3 inspections. In addition, the staffs position on requirements. Potential enhancements noted by the ISG further the use of VT-3 for the detection of cracking will continue to be evaluation would be addressed by an AMP enhancement. documented in the Detection of Aging Effects program element in GALL Report AMP XI.M16A.

Proposed Change: Delete further evaluation 3.1.2.2.9.A item 7.

Item to be addressed by AMP element 3.

52 II-37 Page A Section 3.1.2.2.9.B.2 The staff agrees with the comment that physical measurement techniques and the inspection acceptance criteria for Westinghouse hold down springs For Westinghouse Hold Down Springs, further evaluation is are to be defined in an AMP.

redundant and inconsistent with standard GALL AMR and AMP formatting and presentation. Definition of physical measurement The staffs SE, Revision 1, for MRP-227 documents the basis for limitations techniques for Westinghouse hold down springs should be and conditions placed on the use of MRP-227 as well as licensee/applicant addressed as part of AMP element 3. Acceptance criteria for the action items that shall be addressed by applicants/licensees who choose to hold down spring inspections would be addressed by AMP element implement the NRC-approved version of MRP-227. Specifically, A/LAI No.

6. 5 of MRP-227-A addresses physical measurements of Westinghouse hold down springs.

Proposed Change: Delete further evaluation 3.1.2.2.9.B item 2.

Item to be addressed by AMP elements 3 and 6. As a result of the staffs resolution of Source ID I-1, areas resolved in the C-19

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  1. Summary of Comment Response ID staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.B.2.

53 II-38 Page A IV.B2.RP-297 The staff agrees with the comment to add cast austenitic stainless steel (CASS) as a material in GALL AMR Item IV.B2.RP-297. In final For CASS CRGT Lower Flanges, the ISG revision to the stainless LR-ISG-2011-04 the Material column of GALL AMR Item IV.B2.RP-297 steel definition in GALL Section IX.C requires that CASS be states stainless steel, including CASS and the Aging Effect/Mechanism specifically designated in an AMR line item when thermal and column states Loss of preload due to neutron irradiation embrittlement, neutron embrittlement susceptibility are identified. MRP-227-A and for CASS due to thermal aging embrittlement.

Table 3-3 identifies the material of construction for CRGT lower flanges as CF-8 and thermal and neutron embrittlement identified as considerations for primary component classification.

Proposed Change: Identify CASS as an additional material in GALL IB.B2.RP-297 54 II-39 Page A IV.B2.RP-268 The staff agrees with the comment to delete IV.B2.RP-268. As a result of the staffs resolution of Source ID I-7 and ID I-8, the format of GALL Tables It appears that the primary purpose for the Inaccessible Locations IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR AMR line item is to provide a further evaluation of inaccessible items in the GALL Report for non-RVI components. In addition, the locations in partially accessible components susceptible to cracking footnotes in the Further Evaluation column of these tables are not due to SCC and IASCC using further evaluation note 3 (SRP-LR incorporated into final LR-ISG-2011-04. GALL AMR Items IV.B2.RP-268, Section 3.1.2.2.9A Part A). This further evaluation is redundant to IV.B3.RP-309 and IV.B4.RP-238 for Westinghouse, Combustion the note 3 further evaluation required by other AMR lines. Engineering and Babcock and Wilcox designed plants, respectively, are not incorporated in final LR-ISG-2011-04.

Proposed Change: Delete IV.B2.RP-268 55 II-40 Page A IV.B2.RP-269 The staff agrees with the comment to delete IV.B2.RP-269. As a result of the staffs resolution of Source ID I-7 and ID I-8, the format of GALL Tables It appears that the primary purpose for the Inaccessible Locations IV.B2, IV.B3, and IV.B4 in final LR-ISG-2011-04 is consistent with AMR AMR line item is to provide a further evaluation of inaccessible items in the GALL Report for non-RVI components. In addition, the locations in partially accessible components susceptible to Loss of footnotes in the Further Evaluation column of these tables are not fracture toughness due to neutron and irradiation embrittlement incorporated into final LR-ISG-2011-04. As a result, GALL AMR Items using further evaluation note 3 (SRP-LR Section 3.1.2.2.9A Part A). IV.B2.RP-269, IV.B3.RP-311 and IV.B4.RP-239 for Westinghouse, This further evaluation is redundant to the note 3 further evaluation Combustion Engineering and Babcock and Wilcox designed plants, required by other AMR lines respectively, are not incorporated into final LR-ISG-2011-04.

Proposed Change: Delete IV.B2.RP-269 C-20

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  1. Summary of Comment Response ID 56 II-41 Page A IV.B2.RP-265 The staff does not agree with the comment to change GALL AMR Item IV.B2.RP-265 to be consistent with other GALL AMR none-none line No additional measures (Cracking due to SCC and IASCC) in items and the statement that there are no aging effects requirement Section 3.3.1 of MRP-227-A defines the no additional measures management.

category as: those PWR internals for which the effects of all eight aging mechanisms are below the screening criteria. Additional The No Additional Measures category of components in MRP-227-A does components were placed in the No Additional Measures group as a not equate to such components not having an aging effect requiring result of FMECA and the functionality assessment. No further management; it only indicates that MRP-227-A does not include guidance action is required by the MRP-227-A for managing the aging of the to manage aging for components categorized as No Additional Measures.

No Additional Measures components. Simply put, there are no Thus, the staff agrees with the commenters following statement that [n]o aging effects requiring aging management. further action is required by MRP-227-A for managing the aging of the No Additional Measures components. The IPA is independent of MRP-227-A Proposed Change: Change the aging effect column and AMP and may identify applicable aging effects to manage, which may be column for IV.B2.RP-265 to be consistent with other GALL AMR broader than the aging effects identified in MRP-227-A for RVIs. Thus, the none-none line items and move the lines to GALL Section IV.E, No Additional Measures category of components in MRP-227-A does not Common Miscellaneous Material Environment Combinations. alleviate the requirements in 10 CFR 54.21(a)(3).

In any event, the staff acknowledges that GALL AMR Items IV.B2.RP-265, IV.B2.RP-267, IV.B3.RP-306, IV.B3.RP-307, IV.B4.RP-236 and IV.B4.RP-237 caused confusion; thus, final LR-ISG-2011-04 does not incorporate GALL AMR Items IV.B3.RP-307, IV.B4.RP-236 and IV.B4.RP-237. In addition, GALL AMR Items IV.B2.RP-265, IV.B2.RP-267 and IV.B3.RP-306 in final LR-ISG-2011-04 clarify that there is no additional aging management for reactor internal No Additional Measures components unless required by ASME Section XI, Examination Category B-N-3 or relevant operating experience exists.

57 II-42 Page A IV.B2.RP-267 The staff does not agree with the comment to change GALL AMR Item IV.B2.RP-267 to be consistent with other GALL AMR none-none line No additional measures (Loss of fracture toughness due to neutron items and that there are no aging effects requirement management.

and irradiation embrittlement) in Section 3.3.1 of MRP-227-A defines the no additional measures category as: those PWR internals for As a result of the staffs resolution for Source ID II-41, final LR-ISG-2011-which the effects of all eight aging mechanisms are below the 04 does not incorporate GALL AMR Items IV.B3.RP-307, IV.B4.RP-236 screening criteria. Additional components were placed in the No and IV.B4.RP-237. In addition, GALL AMR Items IV.B2.RP-265, IV.B2.RP-Additional Measures group as a result of FMECA and the 267 and IV.B3.RP-306 in final LR-ISG-2011-04 clarify that there is no functionality assessment. No further action is required by the MRP- additional aging management for reactor internal No Additional Measures 227-A for managing the aging of the No Additional Measures components unless required by ASME Section XI, Examination Category components. Simply put, there are no aging effects requiring aging B-N-3 or relevant operating experience exists.

management.

Proposed Change: Change the aging effect column and AMP column for IV.B2.RP-267 to be consistent with other GALL AMR none-none line items and move the lines to GALL Section IV.E, C-21

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  1. Summary of Comment Response ID Common Miscellaneous Material Environment Combinations.

58 II-43 Page A-6 The staff noted draft LR-ISG-2011-04 caused confusion between the relationship of the SRP-LR and the GALL Report for PWR RVI Clarification is needed relative to the relationship between the SRP- components. As a result, final LR-ISG-2011-04 does not reference the LR and the GALL documents. SRP-LR in GALL Report AMP XI.M16A in order to be consistent with the format of other AMPs in the GALL Report.

59 II-44 Page A-11 The staff agrees with the comment to delete the word that from the Confirmation Process program element of GALL Report AMP XI.M16A in Wording awkward draft LR-ISG-2011-04. The staff revised this program element to state, in part, for confirming the quality of inspections, flaw evaluations, and Proposed Change: Delete that at the beginning of line 8. corrective actions performed under this program.

60 II-45 Page A-17 The staff agrees with the comment that there is a concern the monitoring and trending and corrective actions program elements are buried in the There is a concern that monitoring and trending program elements Acceptance Criteria section.

and corrective action program elements are buried in the Acceptance Criteria section. As a result of the staffs resolution of Source ID I-1 and II-8, the staff revised LR-ISG-2011-04 so that areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.4.

61 II-46 Page A-20 and A-21 The staff does not agree with the comment to alter the referenced statement, as it comes from the staffs SE, Revision 1, on MRP-227. The The statement To satisfy the requirements of ASME Code Section topic of environmentally-assisted fatigue for PWR RVIs is addressed in III. is confusing if not all plants are committed to Subsection NG. A/LAI No. 8, Item 5 of MRP-227-A. Section 3.0 of the staffs SE, Revision 1, on MRP-227 documents the basis for limitations and conditions being Proposed Change: The statement should be modified to include a placed on the use of MRP-227 as well as licensee/applicant action items qualifying statement like if the plant is committed to Subsection that shall be addressed by applicants/licensees who choose to implement NG. the NRC-approved version of MRP-227. Revisions to the conditions and limitations, applicant/licensee plant-specific action items, and conclusions of the staffs SE, Revision 1, for MRP-227 are not within the scope of LR-ISG-2011-04.

However, as a result of the staffs resolution of Source ID I-1, areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license renewal applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.9.

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  1. Summary of Comment Response ID 62 II-47 Page A Section 3.1.2.2.9.D.1 The staff agrees that the referenced sentence in SRP-LR Section 3.1.2.2.9.D.1 is not clear. However, as a result of the staffs resolution of The intended meaning of the word appropriately in D.1, second Source ID I-1, areas resolved in the staffs SE, Revision 1, for MRP-227 paragraph. Is not clear. and A/LAIs are not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a result of these revisions, final LR-Proposed Change: Clarify meaning ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.D.1.

63 II-48 Page A-142 The staff noted that as a result of the resolution to Source ID I-7, final LR-ISG-2011-04 does not incorporate the referenced variable names and Spell out variable name in 1.0 of Further Evaluation further evaluation footnotes in GALL Tables IV.B2, IV.B3, and IV.B4.

Recommendations 64 II-49 Page A-5 The staff disagrees with the comment because revisions to the conditions and limitations, applicant/licensee plant-specific action items and Each of the following documents provide information for submittal of conclusions of the staffs SE, Revision 1, for MRP-227 are not within the an AMP and inspection plan: scope of LR-ISG-2011-04.

  • GALL Revision 2 (page XI.M16A-2) and LR-ISG-2011-04 (draft, As a result of the staffs resolution of Source ID I-1, areas resolved in the page A-5) staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the
  • Section 3.5.1 of the Safety Evaluation (page 25) applicants for PWRs provide their responses to the A/LAIs for MRP-227-A in Appendix C of the LRA. In addition, see the staffs resolution of Source It is unclear what actually goes into the LRA and the format. The ID I-3, in which the staff discusses the staffs position/guidance regarding above verbiage implies that the AMP and inspection plan are inspection plans that is documented in RIS 2011-07 dated July 21, 2011.

separate documents that are submitted with the application but are The staff expects that the details of an inspection plan for Category D reviewed and approved by the NRC as unique documents. A quick plants (defined in RIS 2011-07) will be incorporated into the LRA submittal search of the GALL indicates that PWR Vessel Internals is the only as part of the 10-element AMP and AMR line items. Thus, consistent with program that requires the AMP and an inspection plan to be RIS 2011-07, the staff does not expect Category D plants to provide a submitted for NRC review and approval. separate document that contains an inspection plan in response to A/LAI No. 8.

Proposed Change: Commenter provided revisions to Section 3.5.1 of Safety Evaluation Revision 1 for MRP-227.

65 II-50 Page A-6 The staff agrees with the comment that it is unclear where the A/LAIs should be addressed. As a result of the staffs resolution of Source ID I-1, GALL Rev 2 (page XI M16A-3) states: The responses to the LR areas resolved in the staffs SE, Revision 1, for MRP-227 and A/LAIs are A/LAIs on MRP-227 are provided in Appendix C of the LRA. not addressed in the Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. In addition, final LR-ISG-2011-04 recommends that license LR-ISG-2011-04 (page A-6) deleted this requirement, however LR- renewal applicants for PWRs provide their responses to the A/LAIs for ISG-2011-04 (page A-14, 15) states to provide responses to the MRP-227-A in Appendix C of the LRA. As a result of these revisions, final A/LAIs in Appendix C of the LRA, and to address SRP-LR further LR-ISG-2011-04 does not incorporate SRP-LR Section 3.1.2.2.9.A.1 and evaluation acceptance criteria that are based on these A/LAIs. It is also does not incorporate a discussion of A/LAIs in GALL AMP XI.M16A.

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  1. Summary of Comment Response ID unclear where the licensee action items should be addressed.

Wording implies that the applicant/licensee action items should be addressed in Appendix C and in the associated further evaluation section.

Proposed Change: The Commenter provided revisions to LR-ISG-2011-04 (page A-6).

66 II-51 XI.M16A, PWR Vessel Internals elements 1. Scope of Program, 5. The staff agrees with the comment to delete any reference to Monitoring and Trending, and 6. Acceptance Criteria refer to the WCAP-17096-NP since the report has been submitted for review but not latest NRC approved version of WCAP-17096-NP and the approved by staff. Final LR-ISG-2011-04 does not reference WCAP-associated applicant/licensee action items. It is our understanding 17096-NP.

that WCAP-17096-NP has been submitted for approval however it has not been approved at this time. A program cannot be developed based on an unapproved document or unknown A/LAIs.

Proposed Change: Remove any reference to WCAP-17096-NP or delay issuance of LR-ISG-2011-04 until WCAP-17096-NP is approved by the NRC.

67 II-52 Many of the A/LAIs specified in the Acceptance Criteria section of The staff disagrees with the comment to revise A/LAIs in LR-ISG-2011-04, LR-ISG-2011-04 request that the applicant make enhancements or as it is a direct reference to the staffs SE, Revision 1, for MRP-227. See augmented enhancements to various program elements as a result the staffs resolution to Source ID II-49, in which the staff explains that of the responses to the further evaluations. It would be simpler if revisions to the conditions and limitations, applicant/licensee plant-specific the NRC specified an acceptable method of addressing an issue in action items and conclusions of the staffs SE, Revision 1, for MRP-227 are the XI.M16A program elements and then if the licensee/applicant not within the scope of LR-ISG-2011-04. No revisions were made as a wanted to do something different take an exception rather than result of this comment.

requiring each licensee/applicant to develop a unique set of enhancements for their program.

Proposed Change: Revise A/LAIs to clearly state that additional justification/information is only required to be included in the further evaluation responses if the applicant/licensee is deviating from the requirements of MRP-227-A.

68 II-53 A simplified method of addressing reactor internals in the GALL The staff agrees with the comment, in part, that GALL AMR Tables IV.B2, tables B.2, B.3, and B.4 would be to have line items based on IV.B3 and IV.B4 can be simplified. However, the staff does not agree with component classification (Primary, Expansion, Existing, and No the commenters proposed change.

Additional Measures) as defined in MRP-227-A rather than individual component types (Alignment and Interfacing components: As explained in the staffs resolution of Source ID I-6, the AMR line items in internals hold down spring, Alignment and interfacing components: the GALL Report and MRP-227-A do not solely serve as the basis for upper core plate alignment pins, etc). Making this change would determining components or aging effects that require management or allow multiple line items to apply to several component types and establish the AMR line items to be included in an LRA. The IPA required C-24

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  1. Summary of Comment Response ID reduce the number of Table 2 line items simplifying this section. by 10 CFR 54.21(a) is independent of the AMR line items provided in MRP-227-A and the GALL Report. It is not necessary for the staff to Proposed Change: Revise NUREG-1801 tables B.2, B.3, and B.4 to correlate the number and contents of AMR items in GALL Tables IV.B2, have line items associated with component classification (Primary, IV.B3, and IV.B4 exactly to the number and contents of inspection items in Expansion, Existing Program, and No Additional Measures) and MRP-227-A.

refer to MRP-227-A for the specific components in the four classification groups. In any event, as a result of the staffs resolution of Source ID I-1, GALL AMR Tables IV.B2, IV.B3 and IV.B4 were revised. See resolution of Source ID I-1 for a summary of the revisions.

69 II-54 Several of the applicant/licensee action items (A/LAI) identified in The staff disagrees with the comment to revise A/LAIs in LR-ISG-2011-04, the Safety Evaluation for MRP-227 (pages 32 - 34) required plant- as it is a direct reference to the staffs SE on MRP-227-A. See the staffs specified evaluations or analyses to be submitted as part of the resolution to Source ID II-49, in which the staff explains that revisions to the application. A/LAI Number 5 requires the applicant/licensee include conditions and limitations, applicant/licensee plant-specific action items and its proposed acceptance criteria and an explanation of how the conclusions of the staffs SE, Revision 1, for MRP-227 are not within the proposed acceptance criteria are consistent with the plants scope of LR-ISG-2011-04. No revisions were made as a result of this licensing basis and the need to maintain the functionality of the comment.

component being inspected under all licensing basis conditions of operation during the period of extended operation as part of their submittal to apply the approved version of MRP-227. A/LAI Number 7 requires a plant-specific analysis to be performed on Westinghouse lower support column bodies made of CASS be included as part of their submittal to apply the approved version of MRP-227.

Proposed Change: Revise A/LAI Numbers 5 and 7 to allow for the applicants/licensees to commit performing an analysis prior to the period of extended operation. This would allow applicants/licensees that are submitting in the near future (2013 timeframe) to perform the analyses on normal schedule rather than an expedited schedule.

70 II-55 Page A-17 The staff agrees with the comment that the referenced statement in SRP-LR Section 3.1.2.2.9.A.4 of draft LR-ISG-2011-04 is more In the last paragraph it states: For those component inspections appropriately addressed in the AMP.

that do not achieve the inspection coverage criteria stated in the NRC SE (Revision 1) on MRP-227, the applicant is requested to As a result of the staffs resolution of Source ID I-1, areas resolved in the identify a deviation from the MRP-defined inspection criteria and staffs SE, Revision 1, for MRP-227 and A/LAIs are not addressed in the describe the process and type of evaluation that will be implemented Further Evaluation sections of the SRP-LR in final LR-ISG-2011-04. As a to evaluate the impact of the aging effects on the integrity of those result of these revisions, final LR-ISG-2011-04 does not incorporate SRP-components in the population that were inaccessible to the LR Section 3.1.2.2.9.A.4.

inspection technique, and to identify this process as an applicable enhancement of the monitoring and trending program element of its RVI Program.

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  1. Summary of Comment Response ID The SRP is specifying actions for applicants to perform as part of an aging management program which is more appropriately addressed within program requirements.

Proposed Change: These actions should be included as program elements, not in the further evaluation sections of the SRP.

71 II-56 Page A-19 and A Section 3.1.2.2.9.A.9 The staff does not agree with the comment, in particular the inference that, the effects of the reactor coolant water environment on metal fatigue are In the third paragraph the further evaluation states: To satisfy the not applicable to RVI components since they do not form a portion of the requirements of the ASME Code,Section III, Subsection NG-2160 reactor coolant pressure boundary.

and NG-3121, the existing fatigue CUF analysis shall include the effects of the reactor coolant water environment. See the staffs resolution to Source ID II-46, in which the staff explains that environmentally-assisted fatigue for PWR RVIs is addressed specifically in The December 26, 1999, Generic Safety Issue (GSI) 190 close-out A/LAI No. 8, Item 5 of the staffs SE, Revision 1, on MRP-227. Final LR-memorandum from Ashok C. Thadani, Director of the Office of ISG-2011-04 recommends that license renewal applicants for PWRs Regulatory Research, to William D. Travers, Executive Director for provide their responses to the A/LAIs for MRP-227-A in Appendix C of the Operations, provides the basis for consideration of the effects of the LRA. As a result of these revisions, final LR-ISG-2011-04 does not reactor coolant water environment. It should be noted that GSI-190 incorporate SRP-LR Section 3.1.2.2.9.A.9 concerns are limited to pipe leakage, which is not applicable to RVI components since they do not form a portion of the reactor coolant pressure boundary and are therefore not subject to leakage.

Proposed Change: Delete the referenced sentence in the third paragraph of Further Evaluation A.9.

72 II The comments submitted by EPRI-MRP, the PWROG-MSC, and The NRC staff acknowledges the complexity of the issues captured in NEI are extensive and involve complex issues. EPRI and PWROF, MRP-227-A. However, LR-ISG-2011-04 is not intended to supplement, along with NEI, respectfully request a follow-up meeting with the modify or further resolve the issues raised in MRP-227-A, but rather to NRC staff to discuss resolution of the comments and, if appropriate, reference MRP-227-A and the associated staffs SE, Revision 1, for MRP-an additional comment period. 227, in the usable format of a generic aging management program, as described in the GALL Report. To the extent that comments provided suggestions to clarify or simplify the format of LR-ISG-2011-04 for ease of use, the staff was able to incorporate those changes into the final document. However, to the extent that comments proposed changes to the actual content of the staffs SE, Revision 1, for MRP-227, the staff did not incorporate those comments, as it is beyond the scope and intent of LR-ISG-2011-04. The staff also does not perceive further benefits from an additional public meeting and comment period to resolve the latter set of comments, as it is beyond the scope of LR-ISG-2011-04.

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