ML23123A088
| ML23123A088 | |
| Person / Time | |
|---|---|
| Issue date: | 07/24/2023 |
| From: | Luke Haeg Licensing Processes Branch |
| To: | |
| References | |
| EPID L-2019-TOP-0049 | |
| Download: ML23123A088 (1) | |
Text
Table of Contents
1.0 INTRODUCTION
2.0 REGULATORY EVALUATION
.............................................................................................. 2.1 NRC Regulations and Licensing Requirements .................................................................. 2.2 NRC Guidance and Industry Guidelines .............................................................................
3.0 TECHNICAL EVALUATION
................................................................................................... 3.1 Overview of Topical Report WCAP-17096, Revision 3 ....................................................... 3.2 Applicability of Topical Report WCAP-17096, Revision 3, to the Current Licensing Basis . 3.2.1 Licensing Considerations - Comanche Peak, Diablo Canyon, and Watts Bar ................... 3.2.2 Holders of First Renewed Operating Licenses.................................................................... 3.2.3 Holders of Subsequent Renewed Operating Licenses ....................................................... 3.3 Bases for Closing Out Previously Issued Conditions from NRCs Safety Evaluation for Topical Report WCAP-17096-A, Revision 2 ....................................................................... 3.3.1 Basis for Closing Out Condition 1 for Group 1 RVI Components - Babcock and Wilcox Designs ............................................................................................................................... 3.3.2 Basis for Closing Out Condition 1 for Group 2 RVI Components - Babcock and Wilcox Designs ............................................................................................................................. 3.3.3 Basis for Closing Out Condition 1 for Group 3 RVI Components - Babcock and Wilcox Designs ............................................................................................................................. 3.3.4 Basis for Closing Out Condition 2 for High Fluence Components - Babcock and Wilcox, Combustion Engineering, and Westinghouse Designs ..................................................... 3.3.6 Basis for Closing Out Condition 3 for Group 2 RVI Components - Combustion Engineering and Westinghouse Designs .............................................................................................. 3.3.8 Basis for Closing Out Condition 3 for Group 4 RVI Components - Combustion Engineering Designs ............................................................................................................................. 3.3.9 Basis for Closing Out Condition 3 for Group 5 RVI Components - Westinghouse Designs - 19 3.3.10 Basis for Closing Out Condition 3 for Group 6 RVI Components - Combustion Engineering and Westinghouse Designs .............................................................................................. 3.3.11 Additional Clarification on Supplemental Analysis for Component ID Item Assessments in Topical Report WCAP-17096, Revision 3 ......................................................................... 3.4 Clarification of Items Addressed in Topical Report WCAP-17096, Revision 3 ................. 3.4.1 Statements in MRP-227, Revision 2 Outside the Scope of Topical Report WCAP-17096, Revision 3 ......................................................................................................................... 3.4.2 Time Dependency Considerations for Topical Report WCAP-17096, Revision 3 ............. 3.4.3 ASME Code Criteria Applicability to Topical Report WCAP-17096, Revision 3 ................ 3.4.4 Reinspection Frequencies for Baffle-Former Bolt Inspections .......................................... 3.4.5 Analyses for Ultrasonic Test Results of Bolted Assemblies .............................................. 3.4.6 Resolution of Operating Experience for Westinghouse-Design Baffle-Former Bolts, Thermal Shield Flexures, and Use of Predictive Bolting Pattern Models or Methods .....................
3.4.7 Acceptance Criteria Methodology and Data Requirements Used for the Evaluation of Component Void Swelling or Distortion ............................................................................ 3.4.8 Combustion Engineering Design Components Assessed by Fatigue Evaluations ........... 3.4.9 Resolution for Technical Justification for Removal in Topical Report WCAP-17906, Revision 3 ......................................................................................................................... 3.4.10 Fracture Toughness Considerations ................................................................................. 3.4.11 Babcock and Wilcox-Design Components Currently Lacking a Formal Inspection Standard or Aging Effect Analysis Methodology .................................................................................. 3.4.12 Westinghouse Design Control Rod Guide Tube Support Plates ....................................... 3.4.13 Relevant Operating Experience and Industry Interim Guidance Methodology Not Addressed in Topical Report WCAP-17096, Revision 3 ..................................................................... 4.0 REFERENCING TOPICAL REPORT WCAP-17096, REVISION 3 IN THE CURRENT LICENSING BASIS ........................................................................................... 5.0 LIMITATIONS AND CONDITIONS ...................................................................................... 6.0
SUMMARY
AND CONCLUSIONS ......................................................................................
7.0 REFERENCES
FINAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR THE PRESSURIZED WATER REACTOR OWNERS GROUP TOPICAL REPORT WCAP-17096-NP, REVISION 3, REACTOR INTERNALS ACCEPTANCE CRITERIA METHODOLOGY AND DATA REQUIREMENTS EPID L-2019-TOP-0049
1.0 INTRODUCTION
By letter dated July 31, 2019 (Reference 1), as supplemented by letters dated April 8, 2021 (Reference 2), and January 14, 2022 (Reference 3), the Pressurized Water Reactor Owners Group (PWROG) submitted Topical Report (TR) WCAP-17096-NP, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements (Reference 4), to the U.S. Nuclear Regulatory Commission (NRC) for review and approval.
The purpose of the subject TR is to provide methodologies for demonstrating PWR reactor vessel internals (RVI) integrity throughout the life of the plant, which includes the extended period authorized by license renewal in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants. The updated acceptance criteria methodology and data requirements for PWR RVI components in the subject TR are based on the PWROGs effort to update the PWR RVI guidelines in WCAP-17096-NP-A, Revision 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements, based on changes made to the augmented inspection and evaluation (I&E) criteria for PWR RVI components in the Electric Power Research Institute (EPRI) Non-Proprietary Report No. 3002017168, Materials Reliability Program [MRP] 227: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, Revision 1 - NRC-approved or (-A) (Reference 5).
In general, the NRC staffs review of the subject TR evaluated how it reinforces NRC regulations and guidelines as noted within this safety evaluation (SE). Table 1, Historical Timeline of NRC-Approved Guidance for Review of RVI Components, below, provides a historical account of NRC-approved documents, approval dates, and corresponding agency records available in the NRC Agencywide Documents Access and Management System (ADAMS). The NRC staff referred to these documents to verify updates to aging management criteria and to confirm the resolution of requests for additional information (RAIs) in previously processed TRs that were reviewed by the NRC.
Table 1 - Historical Timeline of NRC-Approved Guidance for Review of RVI Components Document Title Approval Date Agency Record Reference No.
NUREG-1801, Revision 2 (GALL Report) December 2010 ML103490041 Reference 6 SE (Revision 0) for TR MRP-227, Revision 0 June 22, 2011 ML111600498 Reference 7 SE (Revision 1) for TR MRP-227, Revision 1 December 16, 2011 ML11308A770 Reference 8 TR MRP-227-A, Revision 0 December 2011 ML12017A194 Reference 9 NRCs Endorsement Letter MRP-227-A, February 3, 2012 ML120270374 Reference 10 Revision 0 TR MRP-227-A, Interim Guidance February 18, 2014 ML14274A372 Reference 11 SE for TR WCAP-17096-NP, Revision 2 May 3, 2016 ML16061A243 Reference 12 TR WCAP-17096-NP-A, Revision 2 August 31, 2016 ML16279A320 Reference 13 NRC Approval Letter WCAP-17096-NP-A, January 3, 2017 ML16271A001 Reference 14 Revision 2 NUREG-2191, Vol. 1 (GALL-SLR Report February 7, 2017 ML16274A389 Reference 15 (Vol. 1)
NUREG-2191, Vol. 2 (GALL-SLR Report February 7, 2017 ML16274A399 Reference 16 (Vol. 2)
SE for TR MRP-227, Revision 1 April 25, 2019 ML19081A001 Reference 17 TR MRP-227, Revision 1-A June 22, 2020 ML20175A112 Reference 18 NRCs SLR-Interim Staff Guidance SLR-ISG- January 8, 2021 ML20217L203 Reference 19 2021-01-PWRVI Federal Register Notice (FRN) for SLR- January 13, 2021 ML20217L212 Reference 20 Interim Staff Guidance SLR-ISG-2021 PWRVI This SE provides a summary of NRC staffs review and evaluation of the methodology and acceptance criteria for the WCAP-17096-NP, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements. The NRC staff performed the review based on a comparison of WCAP-17096-NP, Revision 3, to previously approved guidance listed in Table 1, above.
The scope of the NRC staffs review in this SE included an assessment of a given RVI component provided in Appendices A, C, or E of the subject TR where the component identification (ID) item assessment was the subject of an RAI or was identified as needing further staff clarifications. The NRC staffs evaluation does not analyze any new component ID item assessments for specified RVI components in Appendix A, C, or E of the subject TR (where the given component has been defined as a new Primary or Expansion category component as described in MRP-227, Revision 1-A, and where the PWROG has developed a new component ID item assessment for the equivalent component provided in Appendix A, C, or E of the subject TR).
For the review of the subject TR, the NRC staff prepared the SE input based on the following assessment factors:
The PWROGs bases for closure of NRCs conditions that were previously issued in NRCs SE for WCAP-17096-NP, Revision 2, dated May 3, 2016 (Reference 12), and the PWROGs proposed resolution as defined in Section 2.2 of the subject TR, The PWROGs basis in the subject TR for addressing Applicant/Licensee Action Item-01 (A/LAI-01) that was previously issued in NRCs SE dated April 25, 2019 (Reference 17) for MRP-227, Revision 1-A,
Changes that needed to be incorporated into the subject TR as a result of changes made to MRP-227, Revision 1-A, including supplemental EPRI MRP reports, PWR methodologies, or guidelines that apply to the RVI components, and changes that were subsequently issued to the PWROGs issuance of WCAP-17096-NP, Revision 2-A, Items discussed in the subject TR that were not changed from those in WCAP-17096-NP, Revision 2, for which the NRC staff determined that further clarifications were needed related to the updated guidelines in MRP-227, Revision 1-A, Items discussed in the subject TR where the NRC staff determined it needed RAIs, including, but not limited to, RAIs on the disposition of recent PWR RVI operating experience (OE) and RAIs on component methodologies/data requirements that were referenced in analysis involving probabilistic criteria, fracture toughness, or crack growth criteria not yet approved by NRC, and Methodologies presented in the subject TR that continue to be referenced to as implied but yet-to-be developed, industry analysis methodologies, or inspection standards.
Additionally, for the factors presented in this section, the NRC staff made a determination to issue a limitation or condition based on the methodology and criteria within (or outside) the scope of the subject TR such as:
NRC staffs determination that the PWROGs methodology in the subject TR places the responsibility for addressing a component acceptance criterion/data analysis criterion to a licensee that may be implementing the report for its licensing or design basis, NRC staffs determination that the PWROG places a given assessment criterion or PWROG-identified additional actions for a given component in the subject TR (including those for managing specific aging effects or mechanisms) outside the scope of the subject TR, NRC staffs determination that the PWROG-specified acceptance criteria methodology or data requirement basis for a given component ID item assessment in Appendix A, C, or E of the subject TR was based on the PWROGs reference to an implied but yet-to-be developed industry analysis methodology or inspection standard, NRC staffs determination that the PWROGs component methodology or data requirement may be relying on industry methodologies not approved by NRC, including those utilizing probabilistic or risk-informed analysis methods, and
NRCs staffs determination that a limitation or condition may be necessary for the subject TR based on the consequences of a component or generic OE not accounted for or bounded by the component ID item assessment bases used to develop MRP-227, Revision 1-A.
Finally, for the review of the subject TR, the NRC staff did not focus its review on unchanged items or administrative changes (edits) made in the subject TR that the NRC staff considered to be solely administrative in nature.
2.0 REGULATORY EVALUATION
2.1 NRC Regulations and Licensing Requirements Title 10 CFR, Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, sets forth the requirements for renewal of commercial nuclear power plant operating licenses. A license renewal application (LRA) or a subsequent license renewal application (SLRA) submitted pursuant to 10 CFR Part 54 must contain the information specified in 10 CFR 54.19, Contents of applicationgeneral information.
The regulation in 10 CFR 54.21(3) requires that, for each structure and component determined to be subject to an aging management review (AMR), the license renewal applicant demonstrates that the effects of aging will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis (CLB1) for the relevant period of extended operation.
In the subject TR, the regulatory basis for the PWROGs acceptance criteria methodology and data requirements are established from 10 CFR Part 54 and the terms and conditions in renewed licenses or subsequent renewed licenses for the initial or subsequent periods of extended operation for PWR-designed power plants. Specifically, the first2 renewed or subsequent renewed operating licenses for PWR-designed plants include specific terms and conditions that require the NRC licensees of nuclear power plant facilities to implement their aging management programs (AMPs) and time-limited aging analyses (TLAAs) for systems, structures, and components (SSCs) that were within the scope of license renewal and subject to either an AMR or a TLAA in the NRCs review of the LRA or SLRA for the facility. The first renewed or subsequent renewed operating licenses for a given PWR-design light water reactor facility may also include terms and conditions that permit the NRC licensee to make changes to their AMPs (including the PWRVI AMP or equivalent AMP for the RVI components) based on implementation of the 10 CFR 50.59 process.
The regulation of 10 CFR 54.35 establishes (in part) the Commissions requirement that the holders of a renewed or subsequent renewed operating license will continue to comply with the requirements of 10 CFR Part 54 during the licensed term of a renewed or subsequent renewed operating license.
1 As defined in 10 CFR 54.3, current licensing basis (CLB) is the set of NRC requirements applicable to a specific plant and a licensee's written commitments for ensuring compliance with and operation within applicable NRC requirements and the plant-specific design basis (including all modifications and additions to such commitments over the life of the license) that are docketed and in effect.
2 For the purposes of this SE, the initial renewed operating license is known as the first renewed license issued under 10 CFR Part 54.
2.2 NRC Guidance and Industry Guidelines For PWR RVI components that are required to be within the scope of license renewal in accordance with 10 CFR 54.4(a) and subject to an AMR in accordance with 10 CFR 54.21(a)(1), NRC staff developed guidance for PWR RVI AMPs in the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report - Final Report (NUREG-2191, Volumes 1 and 2)
(References 15 and 16), AMP XI.M16A, PWR Vessel Internals, in accordance with NRCs Interim Staff Guidance (ISG), Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors (SLR-ISG-2021-01-PWRVI or SLR-ISG) (References 19 and 20). To assist NRC-licensed PWR facilities in addressing the program element criteria established in GALL-SLR Report AMP XI.M16A, the EPRI MRP developed a generic, sampling based I&E methodology for managing the effects of aging in PWR RVI components, with the current NRC-approved version of the guidance established in MRP-227, Revision 1-A.
The PWROGs evaluation methodology, acceptance criteria, and data analysis criteria in the subject TR are linked to the component I&E criteria established in MRP-227, Revision 1-A. In the SLR-ISG, the NRC updated Section AMP XI.M16A to allow the AMP (through the subsequent period of extended operation) to be based on the EPRIs version of MRP-227, Revision 1-A, as subject to an 80-year gap analysis to account for potential adjustments that might need to be made to I&E criteria in MRP-227, Revision 1-A during a subsequent renewed licensing period. The SLR-ISG allows relevant supplemental methodologies, guidelines, or reports to be included within the scope of a licensees PWR vessel internals AMP, particularly if the reports, guidelines, or methodologies were determined to be applicable to the CLB for the facility. This includes an NRC licensees potential use of the subject TR as a supplemental methodology, even though it is not specifically cited or referenced as a supplemental methodology in the version of GALL-SLR AMP XI.M16A in NUREG-2191, Volume 2, or in the updated version of this AMP in the SLR-ISG.
The industry guidelines referenced in this SE, Nuclear Energy Institute (NEI) NEI 03-08 (Reference 22), which include the recommended classification of each material issue as Mandatory, Needed, or Good Practice, is dependent on the relative level of importance of the issue and associated recommended action. The NRC staff notes that the mention of these types of criteria in the SE are not being established, referenced, or cited as NRC requirements, but rather are established by the NEI for NRC licensee-implemented programs (including PWR VIPs or the equivalent named in MRP-227 program) in alignment with the NEI materials management industry initiatives of NEI 03-08.
3.0 TECHNICAL EVALUATION
This technical evaluation section documents the NRC staff's evaluation of the subject TR against the relevant criteria identified in Section 2.0.
3.1 Overview of Topical Report WCAP-17096, Revision 3 The main objective of the subject TR is to identify consistent, industry-wide analytical methodologies and data requirements for developing: (1) acceptance criteria for the Primary and Expansion components identified in the MRP-227, Revision 1-A, and (2) evaluation procedures for facilities to assess potential safety and functional impacts degradation in components with observed relevant conditions.
3.2 Applicability of Topical Report WCAP-17096, Revision 3, to the Current Licensing Basis The Limitations and Conditions defined in Table A-1 of Attachment A to this SE provide details on the use and generic applicability of the subject TR. The subject TR is dependent on the following factors: (1) the type of plant operating license (i.e., original, first renewed, or subsequent renewed license) held by the NRC-licensed facility which owns the specified PWR unit, (2) the type of AMP (i.e., plant-specific AMP versus generic MRP-227-based AMP) being applied to the PWR RVI components, (3) the type of generic EPRI MRP-based program being applied as the basis for the AMP in the CLB, and (4) the version of MRP-227 being applied to the program.
Limitations and Conditions are additional restrictions imposed by the NRC staff to further frame the scope of applicability of a TR and identify any additional plant-specific actions that will be needed to support the NRC staffs review of a request to implement the subject TR. Limitations and Conditions identify the items that will need to be addressed in subsequent individual plant applications that will reference the subject TR. The NRC staff has enclosed a summary table in Attachment A to this SE named, Table A-1, Summary of Limitations and Conditions for Topical Report WCAP-17096, Revision 3. The table provides a brief synopsis and numerical order of limitation and condition items (e.g., formatted as L-01, C-01, etc.) for the subject TR.
As stated in Section 1.0 of this SE, the PWROGs updated acceptance criteria methodology and data requirements for PWR RVI components in the subject TR presents an update to the previous guidelines in WCAP-17096-NP, Revision 2. The updates are based on changes made to augmented I&E criteria for PWR RVI components in MRP-227, Revision 1-A from the prior version of the EPRI MRP methodology in MRP-227-A. Therefore, the criteria in the subject TR applies to NRC licensee RVI AMPs and similarly applies to the EPRI MRP I&E methodology in MRP-227, Revision 1-A.
Also, the NRC staff noted that, in Section 1 of the subject TR, the PWROG included a statement3 to indicate that the methodology and component criteria in the subject TR may be used in conjunction with the prior EPRI MRP I&E methodology in MRP-227-A.
The NRC staff determined that this could potentially create an inconsistency with the EPRI MRP timing criterion for converting an NRC licensees PWR RVI AMP over to the methodology in MRP-227, Revision 1-A. Specifically, NRC staff noted that, in Section 7 of MRP-227, Revision 1-A, the EPRI MRP establishes an NEI 03-08 Needed criterion that recommends a programmatic AMP conversion to MRP-227, Revision 1-A, by the appropriate time identified in the MRP. Therefore, for NRC licensees that implement EPRI MRP-227 site-specific AMPs as part of their NEI 03-08 initiatives, the licensees AMP would follow the EPRI MRPs implementation criteria in Section 7 of MRP-227 Revision 1-A that called for conversions of program over to the version of MRP-227 in Revision 1-A of the report (by January 1, 2021). The NRC staff addressed this apparent inconsistency in RAI-01 in the NRCs RAI transmittal, dated September 21, 2020 (Reference 23).
In the PWROGs response to RAI-01 (References 24 and 25), it stated that the subject TR is intended to be a companion document to MRP-227 and the subject TR is in alignment with the current version of NEI 03-08 (References 21 and 22). Consequently, the PWROG amended Reference 36 cited in the subject TR to include the updated version of MRP-227 Revision 1-A. Given that the subject TR was updated to account for changes made in MRP-227, Revision 1-A, the NRC staff finds that the PWROGs explanation and its revision to Reference 36 in the subject TR to be acceptable. The NRC 3Note that the methodologies defined herein can also be used in conjunction with Revision 0 of MRP-227-A [1] after accounting for differences in component application, component nomenclature, and component numbering.
licensees following these MRP-227, Revision 1-A guidelines would convert their MRP-227-based RVI management program over to version MRP-227, Revision 1-A, by the appropriate time identified in the guidance. Therefore, RAI-01 is resolved.
The NRC staff also verified that in Section 1.1 of the subject TR, the PWROG added new information to provide the specific basis for implementing the methodology in the subject TR in alignment with: (1) the three implementation categories known as Mandatory, Needed, or Good Practice defined in NEI 03-08, (2) the implementation protocols established in Section 7 of MRP-227, Revision 1-A, and (3) Needed criterion for applicable components as specified in Section 7.0 of MRP-227, Revision 1-A. Specifically, the PWROG states that the methodology in the subject TR does not include any specific NEI 03-08 Mandatory, Needed, or Good Practice criteria, other than those needed to clarify that the methodologies defined in the subject TR which can be used to meet the NEI 03-08 Needed criterion specified in Section 7.5 of MRP-227, Revision 1-A.
In general, the NRC staff concluded that the PWROGs rationale in this regard is acceptable because Section 7 of MRP-227, Revision 1-A permits an NRC licensee to apply the methodology in the subject TR as one of the options that the licensee may have for evaluating the RVI component inspection results. Specifically, Section 7.5 of MRP-227, Revision 1-A establishes the following NEI 03-08 Needed criterion for these types of AMPs:
Needed: Examination results that do not meet the examination acceptance criteria defined in Section 5 of these guidelines shall be recorded and entered in the owners plant corrective action program and dispositioned. Engineering evaluations used to disposition an examination result that does not meet the examination acceptance criteria in Section 5, shall be conducted in accordance with NRC approved evaluation methods (i.e., ASME Code Section XI, PWR Owners Group topical report 17096-NP-A or equivalent method)."
The NRCs updated version of GALL-SLR AMP XI.M16A in the SLR-ISG also accounts for this possibility. Therefore, the NRC staff does not specifically reference the subject TR methodology or the previous methodology in WCAP-17096-NP, Revision 2, in any of the program elements of GALL-SLR AMP XI.M16A (including the updated version of the AMP in the SLR-ISG). As such, the evaluation methodology, acceptance criteria, and data analysis requirements in the subject TR (or in WCAP-17096-NP, Revision 2) only provide one way of demonstrating conformance with the monitoring and trending element criteria in the GALL/GALL-SLR AMP XI.M16A or with the NEI Needed criterion referenced in Section 7.5 of the MRP-227, Revision 1-A.
3.2.1 Licensing Considerations - Comanche Peak, Diablo Canyon, and Watts Bar L-01(a) (WCAP-17096-NP, Revision 3):
For PWR plants with an original operating license in a pre-submittal phase for first-time license renewal or currently undergoing NRC review of a first-time license renewal application, NRC licensees may use the subject TR if they are implementing AMPs in accordance with MRP-227, Revision 1-A.
During the time of NRCs review of the subject TR, all NRC licensees with PWR units had received either a first renewed or a subsequent renewed operating license for their PWR units, except for Comanche Peak Steam Electric Station, Units 1 and 2 (Comanche Peak), Diablo Canyon Power Plant, Units 1 and 2 (Diablo Canyon), and Watts Bar Nuclear Plant, Units 1 and 2 (Watts Bar).
Therefore, because the Comanche Peak, Diablo Canyon, and Watts Bar original operating licenses did not include aging management terms and conditions discussed in Section 2.1 of this SE, the
limitation above has been identified and needs to be addressed in subsequent individual plant licensing applications when referencing the subject TR.
The NRC licensees for Comanche Peak, Diablo Canyon, and Watts Bar may opt to implement (on their own initiative) an RVI AMP for their units based on the EPRI I&E criteria in MRP-227, Revision 1-A. The methodology in the subject TR may be applied as part of the AMP regulatory basis for these units, subject to the limitations and conditions of this SE as defined in Table A-1 of Attachment A to this SE which applies to the Westinghouse (W)-designed PWRs.
3.2.2 Holders of First Renewed Operating Licenses L-01(b) (WCAP-17096-NP, Revision 3):
For PWR plants with first renewed operating licenses, NRC licensees may use the subject TR if they are implementing AMPs in accordance with MRP-227, Revision 1-A.
NRC licensees holding their first renewed operating license for their PWR units may apply the acceptance criteria methodology and data requirements as defined in the subject TR (subject to the limitations and conditions when referencing the subject TR as documented in this SE) where the NRCs safety evaluation report (SER) for the first license renewal references an NRC-approved RVI AMP that is either: (1) based on MRP-227, Revision 1-A, or (2) based on MRP-227-A that has been converted over to MRP-227, Revision 1-A, after the date of issuance of the NRC SER for the license renewal application. This may be used to satisfy the NEI 03-08 Needed criterion as referenced in Section 7.5 of the MRP-227, Revision 1-A. Therefore, the limitation above has been identified and needs to be addressed in subsequent individual plant licensing applications when referencing the subject TR.
3.2.3 Holders of Subsequent Renewed Operating Licenses L-01(c) (WCAP-17096-NP, Revision 3):
For PWR plants with subsequent-renewed operating licenses, NRC licensees may use the subject TR only for those RVI components whose I&E criteria for the components are still consistent with MRP-227, Revision 1-A, and have not been modified by the results of an 80-year RVI gap analysis or a version of MRP-227 covering an 80-year aging assessment period.
For NRC licensees holding subsequent renewed operating licenses of their PWR units, where the NRC SER for the SLR application references an NRC-approved RVI AMP in accordance with MRP-227, Revision 1-A (as supplemented by the 80-year gap analysis as defined in SLR-ISG), or alternatively, an EPRI MRP-based AMP that is based on an NRC-approved 80-year version of the MRP-227 report (e.g. MRP-227, Revision 2-A of the report), NRC licensees may apply the PWROG acceptance criteria methodology and data requirements as defined in the subject TR (subject to the limitations and conditions when referencing the subject TR as documented in this SE) which may be used to satisfy the NEI-03-08 Needed criterion referenced in Section 7.5 of the MRP-227, Revision 1-A, but only in a limited case. Specifically, the PWROG's acceptance criteria and data requirements for the applicable component ID item assessment in the subject TR may only be applied if the I&E criteria for the component type in MRP-227, Revision 1-A remain applicable to the component when considering 80-year aging considerations in a gap analysis or as defined in
MRP-227, Revision 2. For RVI components whose I&E criteria would change under 80-year aging considerations from those for the same components in MRP-227, Revision 1-A, the applicable ID item assessments for the components in the subject TR would not apply to the 80-year aging considerations for the components. Therefore, the limitation identified in this section needs to be addressed in subsequent individual plant licensing applications when referencing the subject TR.
3.3 Bases for Closing Out Previously Issued Conditions from NRCs Safety Evaluation for Topical Report WCAP-17096-A, Revision 2 In the NRCs SE dated May 3, 2016, for WCAP-17096-NP, Revision 2, several limitations and conditions were issued based on the PWROGs previous acceptance criteria and data analysis bases for component RVI locations, as defined in Appendix A, C, or E. In Section 2.2 of the subject TR, the PWROG amended the WCAP-17096 methodology for the purpose of resolving those NRC limitations and conditions identified in NRCs SE for WCAP-17096-NP, Revision 2.
The following subsections provide a discussion of the NRC staffs review of the PWROGs bases for resolving and closing out these conditions as detailed in the subject TR. As discussed below, the NRC staff considers all conditions from NRCs SE, dated May 3, 2016, to have been met (relative to the issuance of the conditions in the May 3, 2016 SE for WCAP-17096-NP, Revision 2) because: (1) the PWROGs basis for closing the stated condition in the subject TR, including supplemental information from the PWROGs RAI responses, was sufficient for closing the condition opened in the May 3, 2016 SE, or (2) the PWROGs basis for closing the stated condition in the subject TR, including supplemental information from the PWROGs RAI responses, was not sufficient to close out the condition identified in NRCs SE for WCAP-17096-NP, Revision 2, and the NRC staff proceeded to open a new condition or limitation in this SE for resolution of the previous condition.
3.3.1 Basis for Closing Out Condition 1 for Group 1 RVI Components - Babcock and Wilcox Designs The components that apply to Condition 1 (C-01) (WCAP-17096-NP, Revision 2) for Group 1 RVI components are identified below:
core barrel cylinder and welds core former plates internal and external baffle-to-baffle (BB) bolts core barrel-to-former (CF) bolts locking devices for CF bolts and external BB bolts core baffle plates The PWROG proposed to resolve C-01 (WCAP-17096-NP, Revision 2) for Group 1 RVI components by inclusion of the following additional action to the applicable section of Appendix A of the subject TR:
"The licensee shall submit the plant-specific analysis for this component item within one year after the inspection of the linked Primary item when the results trigger the expansion criteria in MRP-227."
During its review, the NRC staff discovered two apparent inconsistencies with the PWROG proposed resolution of C-01 (WCAP-17096-NP, Revision 2) for the referenced Group 1 RVI components.
Specifically, the NRC staff observed that the PWROG only modified Appendix A to include the
additional actions for Group 1 RVI components related to EPRI MRP Expansion category components. However, the NRC staff determined that the C-01 (WCAP-17096-NP, Revision 2) for Group 1 RVI components also applies to the Primary core baffle plates, and the PWROG did not include an analogous type of additional actions in Appendix A to the subject TR, Section A.1.9 for the core baffle plates. Furthermore, NRC staff also verified that the EPRI MRPs basis for dispositioning flaws in core baffle plates, or in any of the Group 1 RVI components Expansion category components also permitted disposition for Appendix A (specifically Sections A.2.5 to A.2.9) to be based on component repair or replacement activities as an alternative to a plant-specific analysis of the impacted component. The NRC staff addressed these apparent gaps in information through the issuance of RAI-02 and RAI-03 (see Reference 23).
In its response to RAI-02 (Reference 26), the PWROG amended the additional actions for the referenced Group 1 RVI components Expansion components in Section 2.2 to the subject TR and in Sections A.2.5 to A.2.9 of Appendix A to the subject TR to include component replacement or justified alternative disposition process options in the statement. In its response to RAI-03 (see Reference 26),
the PWROG amended the subject TR Section 2.2 and the data analysis basis in the subject TR, Appendix A, Section A.1.9 for the Primary baffle plates to include the following additional action for the baffle plates:
the licensee is required to submit the LEFM [linear elastic fracture mechanics]
analysis for the baffle plates under normal/upset condition loads, considering IASCC, within one year of the inspection of the Primary component item for NRC to determine whether review and approval are needed if the inspection results indicate aging not meeting the acceptance criteria in Table 5-1 of MRP-227.
The NRC staff finds that the PWROGs incorporation of these additional actions in the subject TR is sufficient to close C-01 (WCAP-17096-NP, Revision 2) for Group 1 RVI components because the timing for submitting the plant-specific or LEFM analysis is appropriately incorporated into the subject TR and these additional actions now form the basis for submitting the plant-specific analysis to the NRC. Section 3.3.11 of this SE provides additional clarification on the use of these additional actions.
Based on these considerations, C-01 (WCAP-17096-NP, Revision 2) for Group 1 RVI components has been met and is closed.
3.3.2 Basis for Closing Out Condition 1 for Group 2 RVI Components - Babcock and Wilcox Designs The components that apply to C-01 (WCAP-17096-NP, Revision 2) for Group 2 RVI components are identified below:
vent valve assembly, top and bottom retaining rings flow distributor (FD) bolts upper thermal shield (UTS) bolts, and their associated locking devices lower thermal shield (LTS) bolts, and their associated locking devices surveillance specimen holder tube (SSHT) bolts The PWROG proposed to resolve C-01 (WCAP-17096-NP, Revision 2) for Group 2 RVI components by inclusion of the following additional actions to the applicable section in Appendix A to the subject TR:
(1) the vent valve top and bottom retaining rings, FD bolts, and bolt locking devices:
The licensee shall submit the plant-specific analysis for this component within one year after the inspection if the results do not meet the acceptance criteria in MRP-227, or, (2) the UTS bolts and bolt locking devices, LTS bolts and bolt locking devices, and the SSHT bolts and bolt locking devices:
The licensee shall submit the plant-specific analysis for this component within one year after the inspection of the linked Primary item when the results trigger the expansion criteria in MRP-227.
The NRC staff finds that the PWROGs incorporation of these additional actions in the subject TR is sufficient to close C-01 (WCAP-17096-NP, Revision 2) for Group 2 RVI components because the timing for submitting the plant-specific analysis is appropriately incorporated into the subject TR and the additional actions now form the basis for submitting the plant-specific analysis to the NRC.
Section 3.3.11 of this SE provides additional clarification on the use of these additional actions. Based on these considerations, C-01 (WCAP-17096-NP, Revision 2) for Group 2 RVI components has been met and is closed.
3.3.3 Basis for Closing Out Condition 1 for Group 3 RVI Components - Babcock and Wilcox Designs The components that apply to C-01 (WCAP-17096-NP, Revision 2) for Group 3 RVI components are identified below:
control rod guide tube (CRGT) assembly - CRGT spacer castings, and incore monitoring instrument (IMI) guide tube spiders and associated spider-to-lower grid rib section welds The PWROG proposed to resolve C-01 (WCAP-17096-NP, Revision 2) for Group 3 RVI components by the inclusion of the following additional action to the applicable sections in the subject TR, Appendix A:
The licensee shall submit the plant-specific analysis for this component within one year after the inspection if the inspection results do not meet the acceptance criteria in MRP-227.
The NRC staff finds that the PWROGs incorporation of the additional action to the subject TR is sufficient to closeout C-01 (WCAP-17096-NP, Revision 2) on the Group 3 RVI components because the basis for submitting the referenced analysis is appropriately incorporated in the subject TR and now forms the basis for submitting the plant-specific analysis to the NRC. Section 3.3.11 of this SE provides additional clarification on the use of this additional action. Based on these considerations, C-01 (WCAP-17096-NP, Revision 2) for Group 3 RVI components has been met and is closed.
3.3.4 Basis for Closing Out Condition 2 for High Fluence Components - Babcock and Wilcox, Combustion Engineering, and Westinghouse Designs As issued in NRCs SE, dated May 3, 2016, for WCAP-17096-NP, Revision 2, Condition 2 (C-02)
(WCAP-17096-NP, Revision 2) pertains to high fluence RVI components that are projected to have neutron fluence exposures in excess of 3x1021 n/cm2 (E > 1.0 MeV) at the end of the licensed service life of the plant or service period of interest. The condition addresses the crack growth rate (CGR) models for the high fluence stainless steel (SS) RVI components in the B&W, Combustion Engineering (CE) and W-design PWRs.
Section 2.2 of the subject TR includes the following additional action in the flaw tolerance section of the component ID item assessment for each of the specified C-02 (WCAP-17096-NP, Revision 2) RVI components that are included in either Appendix A, C, or E of the subject TR:
One appropriate model is available in EPRI Report Models of Irradiation-Assisted Stress Corrosion Cracking of Austenitic SS in Light Water Reactor Environments, Volume 2: Disposition Curves Application [46], which provides the technical basis for ASME Code Case N-889 [52].
The NRC staff determined that additional clarification was needed to verify the basis for using the EPRI report on stress corrosion cracking published in 2014 (cited as Reference 46 in the subject TR) for component CGRs. Specifically, the PWROG indicated that the CGR models in the referenced EPRI report (References 28 and 29) were also used as the CGR criteria basis for the industrys development of ASME Code Case N-889. At the time of NRCs review of the subject TR, staff verified that the cited EPRI report had not yet been submitted for NRC review and that NRCs review of the ASME Code Case N-889 was still pending review under the current 10 CFR 50.55a rulemaking update. Therefore, NRC staff concluded that the EPRI report is not an NRC-approved methodology for establishing component CGRs. The NRC staff addressed the use of unapproved guidance with the issuance of RAI-04 (Reference 23).
By letter dated January 14, 2022 (Reference 24), the PWROG responded to RAI-04 and provided additional information on the technical justification for the CGR model calibration bases of SS weld components.4 The NRC staff evaluated the PWROGs response to RAI-04 as part of the review of the proposed CGRs for SS base metal and weld components that are subject to the neutron fluence ranges shown in Table 2 below.
Table 2 - Neutron Fluence Values Fluence (f) Range Dose n/cm2 (E>1.0 MeV) (dpa) f 5x1020 < 0.75 5x1020 < f 1.4x1022 < 0.75 and 20 f > 1.4x10 22 > 20 The NRC staff verified that the PWROGs response to RAI-04 has placed a selective bounding limit on the CGR basis for the SS component for some of the PWROG-conditional statements regarding selection of the appropriate CGR. In addition, NRC staff reviewed the applicability of the referenced 2014 EPRI report based on a separate NRC review of ASME Code Case N-889 in NRCs Regulatory 4 The NRC staff notes that the PWROG is applying the same bounding CGR bases to crack growth evaluations for stainless steel base metal components.
Guide (RG) 1.147 (Reference 29). In alignment with previously established NRC positions in the SE for WCAP-17096-NP, Revision 2 and given that the NRC has endorsed ASME Code Case N-889 in RG 1.147, Revision 20, as subject to staff conditions on use of the code case, the NRC staff determined that the PWROG-established conditional criteria should be reflected in this Final SE with a condition for the selected CGRs. This condition is intended to ensure that licensees relying on the subject TR would be applying CGR bases of the referenced 2014 EPRI report in a manner that is consistent with the staffs conditions placed on use of ASME Code Case N-889, as established in Table 2 of RG 1.147, Revision 20. Therefore, NRC staff has defined a new Condition 1 (C-01)
(WCAP-17096-NP, Revision 3) in Attachment A of this SE.
C-01 (WCAP-17096-NP, Revision 3):
For stainless steel metal and weld components in the specified neutron fluence exposure ranges as shown in Table 2 of Section 3.3.4 of this SE, the PWROG has indicated that the CGRs are based on the CGR criteria fluence ranges specified in ASME Code Case N-889, as established by the EPRI models in the EPRI Report and cited as Reference 46 in the subject TR. The NRC has endorsed Code Case N-889 in Regulatory Guide 1.147, Revision 20, with conditions for the specified neutron fluence exposure ranges. Therefore, if the CGR models in Reference 46 of the subject TR are followed by a licensee using this TR, then the CGR models are subject to the NRC conditions on Code Case N-889, as defined in Table 2 of Regulatory Guide 1.147, Revision 20.
As a result of the PWROGs response to RAI-04, and its inclusion of the additional action in the subject TR, the NRC staff finds that the item addressed in RAI-04 is resolved and C-02 (WCAP-17096-NP, Revision 2) for High Fluence components for B&W, CE, and W is closed, as subject to the new C-01 (WCAP-17096-NP, Revision 3) for the applicable CGR basis.
3.3.5 Basis for Closing Out Condition 3 for Group 1 RVI Components - Combustion Engineering and Westinghouse Designs The components that apply to Condition 3 (C-03) (WCAP-17096-NP, Revision 2) for Group 1 RVI components are identified below:
CE-design welded core shroud assemblies CE-design bolted core shroud assemblies W-design baffle edge bolts W-design baffle-former assemblies The PWROG proposed to resolve C-03 (WCAP-17096-NP, Revision 2) for Group 1 RVI components by inclusion of the following additional action to the applicable sections in the subject TR, Appendices C and E:
The licensee shall submit the plant-specific or generic FMEA [failure modes and effects analysis] and confirmation of applicability of generic FMEA analysis within one year after any inspection that detects relevant conditions as defined in Table 5-2 (or Table 5-3) of MRP-227.
The NRC staff finds that the PWROGs incorporation of the additional action in the subject TR, Appendices C and E, are sufficient to close C-03 (WCAP-17096-NP, Revision 2) for Group 1 RVI components because the basis for submitting the referenced analysis is appropriately incorporated in
the subject TR and now forms the basis for submitting the plant-specific analysis to the NRC.
Section 3.3.11 of this SE provides additional clarification on the use of this additional action. Based on these considerations, C-03 (WCAP-17096-NP, Revision 2) for Group 1 RVI components has been met and is closed.
3.3.6 Basis for Closing Out Condition 3 for Group 2 RVI Components - Combustion Engineering and Westinghouse Designs The components that apply to C-03 (WCAP-17096-NP, Revision 2) for Group 2 RVI components are identified below:
CE-design core support column bolts in CE designs with bolted core shroud (CS) assemblies5 CE-design core support columns W-design lower support column bolts W-design cast and non-cast lower support column bodies The PWROG proposed to resolve C-03 (WCAP-17096-NP, Revision 2) for Group 2 components by inclusion of the following additional actions in the applicable sections of the subject TR for the components identified below:
(1) for the bolted CE-design core support column bolts and the W-design lower support column bolts:
Note that the licensee shall submit the plant-specific analysis for the acceptable distribution of intact components within one year after any inspection that triggers the expansion criteria as defined in Table 5-2 (or Table 5-3) of MRP-227.
(2) for the CE-design core support columns and the W-design cast or non-cast lower support column bodies:
Note that the licensee shall submit the plant-specific analysis for the acceptable distribution of intact components within one year after any inspection that detects the relevant conditions as defined in Table 5-2 (or Table 5-3) of MRP-227.
(3) for the cast lower support column bodies:
"If plant-specific functional analyses have been performed to respond to applicant/licensee Action Item 7 from MRP-227-A prior to the initial inspection of the components, licensees may compare the previously performed functionality analyses and need not resubmit plant-specific analyses unless the inspection findings are not bounded by the original analysis."
The NRC staff finds that the PWROGs incorporation of these additional actions in the subject TR is sufficient to closeout C-03 (WCAP-17096-NP, Revision 2) for the Group 2 RVI components because 5
The CE-ID assessment for core support column bolts in CE-ID Item 1.1 of Appendix C of the Subject TR was deleted in the PWROGs response to RAI No. 7-(a). Thus, the staff acknowledges that C-03 (Revision 2 of TR) for Group 2 components no longer applies to Group 2 component types (including core support column bolts) in the two U.S. CE-design PWRs with bolted core shroud assembly designs. These PWRs (i.e., the PWR units at Palisades and Ft. Calhoun Stations) have entered permanently defueled, non-power operations per their amended DPR-20 and DPR-40 operating license and technical specification requirements.
the basis for submitting the plant-specific analysis is appropriately incorporated in the subject TR and the additional actions now form the basis for submitting the plant-specific analysis to the NRC.
Section 3.3.11 of this SE provides additional clarification on the use of these additional actions. Based on these considerations, C-03 (WCAP-17096-NP, Revision 2) for Group 2 RVI components has been met and is closed.
3.3.7 Basis for Closing Out Condition 3 for Group 3 RVI Components Items A and B - Combustion Engineering and Westinghouse Designs The components that apply to C-03 (WCAP-17096-NP, Revision 2) for Group 3 RVI components are identified below:
CE-design core support barrel assembly (CSB) upper flange weld CE-design CSB middle girth weld (MGW)
CE-design CSB flexure weld CE-design CSB lower girth weld (LGW)
CE-design CSB upper girth weld (UGW)
CE-design CSB upper axial welds (UAWs)
CE-design lower core support beams CE-design CSB middle axial welds (MAWs)
CE-design CSB lower axial welds (LAWs)
W-design core barrel assembly (CB) upper flange weld (UFW)
W-design CB LGW W-design CB UGW W-design CB UAWs W-design CB lower flange weld (LFW)
W-design CB MAWs W-design CB LAWs In Section 4.0 of the NRCs SE, dated May 3, 2016, for WCAP-17096-NP, Revision 2, the NRC identified several conditions to be implemented when referencing Revision 2 of the TR. In the subject TR, the PWROG provided resolutions for how each condition is addressed. With respect to C-03 (WCAP-17096-NP, Revision 2) for Group 3 RVI components, Item a, the NRC staff established its position for this case (in Section 3.2.2.2.2 of NRCs 2016 SE) where a follow-up inspection for depth sizing of the detected crack in the referenced weld type is only necessary when the visual inspection of the specified weld reveals the presence of a surface-breaking flaw or crack. The NRC staff also established its position that the follow-up inspection of the weld should be performed at the next refueling outage (RFO) after the discovery of the crack.
To address the technical justification for a component re-inspection interval associated with C-03 (WCAP-17096-NP, Revision 2), Group 3 RVI components, Item a, the PWROG incorporated the following additional action to the applicable sections of the subject TR, Appendices C and E:
In order to apply the TR methodology for cracks detected via one-sided visual examinations, if supplementary examinations to determine crack depth were not performed, a follow up examination must be performed no later than the next RFO to confirm the assumed crack growth rates are conservative, unless a technical justification for a longer verification interval is submitted.
The NRC staff finds that with the exception of the issue regarding neutron fluence values discussed below, the PWROGs incorporation of this additional action in the subject TR is sufficient to close C-03 (WCAP-17096-NP, Revision 2) for Group 3 RVI components, Item a, because: (1) the need for performing a follow-up inspection of the applicable components is consistent with the basis of C-03 (WCAP-17096-NP, Revision 2) and (2) the additional action is appropriately incorporated into the subject TR which now forms the basis for submitting the plant-specific analysis to the NRC.
Section 3.3.11 of this SE provides additional clarification on the use of this additional action.
The NRC staff confirmed that the PWROGs reference in the subject TR to the 2014 EPRI report on stress corrosion cracking (see Reference 27) provides the basis for closing out C-03 (WCAP-17096-NP, Revision 2), Group 3 RVI components, Item a, by establishing the CGRs of the referenced SS base metal or weld components. The NRC staff also verified that the methodology in the referenced EPRI report has not been approved by the NRC staff. However, the NRC staff was able to review the applicability of the referenced 2014 EPRI report based on a separate NRC review conducted on code case N-889 in NRCs Regulatory Guide 1.147 (Reference 29). The NRC approved the code case with conditions as previously discussed and evaluated in Section 3.3.4 of this SE.
Based on these considerations, the NRC staff evaluated the PWROGs response to RAI-04 as part of the review of the proposed CGRs for SS base metal and weld components that are subject to the neutron fluence ranges as described in Table 2 in Section 3.3.4 of this SE.
The NRC staff finds that the PWROGs inclusion of the additional actions in the subject TR Appendices C and E, and NRC staffs issuance of the new C-01 (WCAP-17096-NP, Revision 3)6 on CGRs for PWR RVI components with neutron fluence exposures and ranges listed in this section are sufficient to close out C-03 (WCAP-17096-NP, Revision 2), Group 3 RVI components, Item a because the basis for submitting the referenced analysis is appropriately incorporated in the subject TR. Based on these criteria and assessments including the issuance of the new C-01 (WCAP-17096-NP, Revision 3), C-03 (WCAP-17096-NP, Revision 2), Group 3 RVI components, Item a, has been met and is closed.
With respect to C-03 (WCAP-17096-NP, Revision 2), Group 3 RVI components, Item b, the NRC staff established its position from the previous approach applied in WCAP-17096-NP, Revision 2, for the referenced Group 3 Primary components requiring performance of a plant-specific analysis for the components, with the basis for initiating the analysis being dependent on the EPRI MRP category for the component and whether or not the applicable Section 5 of Table 5-2 in MRP-227, Revision 1-A established any defined expansion criteria for the components. The PWROG addressed C-03 (WCAP-17096-NP, Revision 2), Group 3 RVI components, Item b, with the inclusion of the following additional actions for the applicable sections of the subject TR:
(1) CE-design Primary category CSB UFWs and MGWs and the W-design Primary category CB UFWs and LGWs The licensee shall submit the plant-specific crack growth and fracture mechanics analysis justifying the detected flaws, including a justification of the allowable operating period before re-inspection, within one year after any inspection that triggers the expansion criteria as defined in Table 5-2 (or Table 5-3) of MRP-227.
(2) CE-design Primary category CSB flexure welds and all the referenced Group 3 CE-and W-designs Expansion category components in CE-ID Items 5.1, 5.2, 5.3, 5.4, 6.1, and 6 Condition C-01 is addressed in Section 3.3.4 in this SE.
6.2 and W-ID Items 3.1, 3.2, 3.3, 4.2, and 4.3, where the components do not have applicable Expansion acceptance criteria in Table 5-2 or 5-3 of MRP-227, Revision 1-A:
The licensee shall submit the plant-specific crack growth and fracture mechanics analysis justifying the detected flaws, including a justification of the allowable operating period before re-inspection, within one year after any inspection that detects relevant conditions as defined in Table 5-2 (or Table 5-3) of MRP-227.
The NRC staff finds that the PWROGs incorporation of the additional actions in the subject TR is sufficient to closeout C-03 (WCAP-17096-NP, Revision 2) for the Group 3 RVI components, Item b, because the timing for submitting the plant-specific analysis is appropriately incorporated into the subject TR and the additional actions now form the basis for submitting the plant-specific analysis to the NRC. Section 3.3.11 of this SE provides additional clarification on the use of this additional action.
Based on these considerations, C-03 (WCAP-17096-NP, Revision 2) for Group 3 RVI components, Item b, has been met and is closed.
3.3.8 Basis for Closing Out Condition 3 for Group 4 RVI Components - Combustion Engineering Designs The components that apply to C-03 (WCAP-17096-NP, Revision 2) for Group 4 RVI components are identified below:
CE-design core shroud plate-to-former plate welds CE-design core shroud plates CE-design core shroud remaining axial welds (CE-design units with welded core shrouds assembled in two vertical sections)
CE-design core shroud remaining axial welds (CE-design units with welded core shrouds assembled with full height shroud plates)
CE-design core shroud rib and rings In Section 4.0 of NRCs SE for WCAP-17096-NP, Revision 2, NRC identified several conditions to be implemented when referencing Revision 2. In the subject TR, the PWROG provided resolution for how each condition is addressed. The PWROG resolved C-03 (WCAP-17096-NP, Revision 2) for Group 4 with the addition of the following additional actions to the applicable sections in the subject TR, Appendix C:
CE-design Primary category core shroud plate-to-former plate welds and core shroud plates:
The licensee shall submit the plant-specific fracture mechanics analysis justifying the detected flaws within one year after any inspection that triggers the expansion criteria as defined in Table 5-2 of MRP-227.
CE-design Expansion category core shroud remaining axial welds in CE-ID Items 2.1 and 3.1 that do not have applicable Expansion acceptance criteria in Table 5-2 of MRP-227, Revision 1-A:
The licensee shall submit the plant-specific fracture mechanics analysis justifying the detected flaws within one year after any inspection that detects relevant conditions as defined in Table 5-2 of MRP-227.
With the exception of one item discussed below that needs further clarification regarding the PWROGs basis for the core shroud ribs and rings in CE-ID Item 3.2, the NRC staff finds that the PWROGs incorporation of these additional actions in the subject TR, is sufficient to closeout C-03 (WCAP-17096-NP, Revision 2) for the Group 4 RVI components because the basis for submitting the plant-specific analysis is appropriately incorporated in the subject TR and the additional actions now form the basis for submitting the plant-specific analysis to the NRC. Section 3.3.11 of this SE provides additional clarification on the use of these additional actions.
The NRC staff observed that the PWROG did not apply the second of these additional actions to the subject TR to the assessment for the core shroud ribs and rings in CE-ID Item 3.2 (applicable to CE-design PWRs with welded core shrouds using two vertical sections) based on the additions shown below:
The CE component, "Core Support Barrel Assembly (Welded) - Ribs and Rings," has been identified as an inaccessible component for MRP-227, Rev. 1 as discussed in Part C of RAI 12 [41]. This change to MRP-227, Rev. 1 was accepted by the NRC in the draft SE for MRP-227, Rev. 1 [57]. Since a fracture mechanics analysis is no longer applicable to this component, the associated submittal requirement as contained in WCAP-17096-NP-A, Rev. 2 was not included in the methodology defined herein.
The NRC staff also observed that, while the EPRI MRP clearly identified that the CSB ribs and rings are inaccessible (as stated in CE-ID Item 3.2 of MRP-227, Revision 1-A, Table 4-5), the EPRI MRP also identified that the NRC licensee needs to justify further service of the ribs and rings by evaluation or by replacement if expansion to the rib and ring components is triggered by the results of Primary inspections being applied to the core shroud plates in the units. The NRC staff addressed this gap in the information with the issuance of RAI-05 (Reference 23).
In the PWROGs response to RAI-05 (Reference 25), revisions were made to Section 2.2 to the subject TR to indicate that a functionality analysis, if triggered, is applicable to the core shroud ribs and rings in CE-design plants that have welded core shrouds using full height shroud plates. The PWROG also amended the Additional Action(s) in the applicable sections in Appendix C to the subject TR, CE-ID Item 3.2 to include the following additional action for the data analysis criteria that apply to the core shroud ribs and rings:
The licensee shall submit the plant-specific functionality analysis within one year after any inspection that detects relevant conditions as defined in Table 5-2 of MRP-227.
The NRC staff finds that the PWROGs incorporation of these additional actions in the subject TR, is sufficient to closeout C-03 (WCAP-17096-NP, Revision 2) for the Group 4 RVI components because the basis for submitting the plant-specific analysis is appropriately incorporated in the subject TR and the additional actions now form the basis for submitting the plant-specific analysis to the NRC.
Section 3.3.11 of this SE provides additional clarification on the use of these additional actions. Based on all the considerations discussed in this section, C-03 (WCAP-17096-NP, Revision 2) for Group 4 RVI components has been met and is closed.
3.3.9 Basis for Closing Out Condition 3 for Group 5 RVI Components - Westinghouse Designs The components that apply to C-03 (WCAP-17096-NP, Revision 2) for Group 5 RVI components are identified below:
Control rod guide tube lower flange welds Thermal shield flexures only for W-designed PWRs that include internal thermal shield assemblies CRGT LFWs in the remaining (i.e., accessible non-peripheral) CRGT assemblies In Section 4.0 of NRCs SE for Revision 2 of the TR, NRC identified several conditions to be implemented when referencing Revision 2. In the subject TR, the PWROG provided resolutions for how each condition is addressed. The PWROG resolved C-03 for Group 5 by inclusion of the following additional actions to the applicable sections of the subject TR for the components identified below:
W-design peripheral CRGT LFWs and the Expansion category non-peripheral CRGT LFWs:
Plant-specific analysis of acceptable intact weld patterns for this component must be submitted to the NRC to determine if review and approval is needed, within one year after any inspection that triggers the expansion criteria in Table 5-3 of MRP-227, and or W-design thermal shield flexures:
Plant-specific analysis of acceptable dynamic response of the remaining flexures for this component must be submitted to the NRC to determine if review and approval is needed, within one year after any inspection that detects relevant conditions as defined in the "Examination Acceptance Criteria" column in Table 5-3 of MRP-227.
The NRC staff finds that the PWROGs incorporation of these additional actions in the subject TR, is sufficient to closeout C-03 (WCAP-17096-NP, Revision 2) for the Group 5 RVI components because the basis for submitting the plant-specific analysis is appropriately incorporated in the subject TR and the additional actions now form the basis for submitting the plant-specific analysis to the NRC.
Section 3.3.11 of this SE provides additional clarification on the use of these additional actions. Based on these considerations, C-03 (WCAP-17096-NP, Revision 2) for Group 5 RVI components has been met and is closed.
3.3.10 Basis for Closing Out Condition 3 for Group 6 RVI Components - Combustion Engineering and Westinghouse Designs The components that apply to C-03 (WCAP-17096-NP, Revision 2) for Group 6 RVI components are identified below:
CE-design core support plates in the lower support structure CE-design fuel alignment plates in the upper internals assembly W-design lower support column forging or casting in the lower internals assembly W-design upper core plate in the upper internals assembly In Section 4.0 of NRCs SE for WCAP-17096-NP, Revision 2, the NRC staff identified several conditions to be implemented when referencing Revision 2. In the subject TR, the PWROG provided resolutions for how each condition is addressed. The PWROG addressed C-03 for Group 6 by inclusion of the following additional action to the applicable sections of the subject TR for the components referenced CE-ID and W-ID items:
The licensee shall submit the plant-specific analysis for this component within one year after any inspection that detects relevant conditions as defined in Table 5-2 (or Table 5-3) of MRP-227.
The NRC staff finds that the PWROGs incorporation of this additional action in the subject TR, is sufficient to closeout C-03 (WCAP-17096-NP, Revision 2) for the Group 6 RVI components because the basis for submitting the plant-specific analysis is appropriately incorporated in the subject TR and the additional action now forms the basis for submitting the plant-specific analysis to the NRC.
Section 3.3.11 of this SE provides additional clarification on the use of these additional actions. Based on these considerations, C-03 (WCAP-17096-NP, Revision 2) for Group 6 RVI components has been met and is closed.
3.3.11 Additional Clarification on Supplemental Analysis for Component ID Item Assessments in Topical Report WCAP-17096, Revision 3 Some of the PWROG component ID item assessments in the subject TR Appendices A, C, and E included additional actions for applicable components that may instruct NRC licensees to submit a supplemental analysis (when triggered) of the components that are within the scope of a given component ID item assessment for potential NRC review. In many cases, the additional actions for applicable components in the subject TR may specify performance of a specific type of analysis for the specified components. Accordingly, the need for performing these supplemental activities fall within the scope of the NRC licensees own corrective action program, and a licensees decision to determine the proper corrective action resolution that would be used under its 10 CFR Part 50, Appendix B program or NEI 03-08 industry initiative process could result in augmented inspections that are performed in accordance with the methodology in MRP-227, Revision 1-A.
The NRC staff acknowledges that the specific type of analysis that may be referenced in an additional action for an applicable component in the subject TR constitutes only one way of dispositioning levels of age-related degradation that exceed the levels for initiating corrective actions per the criteria that are defined for the applicable component type in MRP-227, Revision 1-A. However, consistent with the implementation criteria in Chapter 7 of MRP-227, Revision 1-A, the licensee has the option of performing and using any type of engineering analysis that differs from the analysis specified in the
subject TR. The licensee may also justify further service of the components with proposed component repair or replacement activities and repair/replacement schedules that would ensure the intended functions of the components.
3.4 Clarification of Items Addressed in Topical Report WCAP-17096, Revision 3 3.4.1 Statements in MRP-227, Revision 2 Outside the Scope of Topical Report WCAP-17096, Revision 3 In Section 1 of the subject TR, the PWROG included the following statement:
the MRP-227 inspection and evaluation guidelines are in development for SLR (80-year operating license), and can be expected to continue to evolve based upon lessons learned from operating experience. Therefore, each utility must remain cognizant of changes made The NRC staff notes that it has not yet approved the MRP-227 report for an 80-year operating license (e.g., MRP-227, Revision 2, or equivalent report submitted by the EPRI MRP for NRC approval).7 Since the NRC has not yet approved MRP-227, Revision 2, for implementation by NRC licensees, the NRC staff finds these statements to be outside the scope of this SE.
3.4.2 Time Dependency Considerations for Topical Report WCAP-17096, Revision 3 In Section 1 of the subject TR, the PWROG clarified that: (1) the RVI AMP applies to both initial and subsequent license renewal (LR/SLR) periods, and (2) the analytical methods defined in the subject TR may be independent of time or consider time dependency when appropriate. The NRC staff determined that some additional explanations are needed for these statements in the subject TR.
Specifically, the methodology in the subject TR accounts for the possibility that an NRC licensee could be performing an RVI component evaluation that is based on a time-dependent parameter defined by the current operating period (Refer to Criterion 3 in 10 CFR 54.3(a) [§54.3(a)]), and that the analysis may conform to all six of the criteria for defining a TLAA in §54.3(a). For those analyses that are determined to conform to the six criteria for TLAAs in §54.3(a), it is the NRC licensees responsibility to ensure that: (1) these analyses are identified and evaluated as TLAAs in the LRA/SLRA in accordance with 10 CFR 54.21(c)(1) [§54.21(c)(1)], and (2) that the TLAAs are appropriately dispositioned in the LRA/SLRA in accordance with one or more of the acceptance criteria for TLAAs, as specified in §54.21(c)(1)(i), (ii) or (iii). For any RVI specific analysis that is determined to conform to the definition of a TLAA, the LRA/SLRA would also be required to include a Final Safety Analysis Report (FSAR)/Updated Final Safety Analysis Report (UFSAR) supplemental summary description for the TLAA in the LRA/SLRA, as required by 10 CFR 54.21(d) [§54.21(d)]. The NRC staff determined that no further review is necessary for the objectives of this SE since the need for identifying and evaluating TLAAs in LRAs/SLRAs is driven and set by the TLAA criteria specified in §54.3a, and the requirements for TLAAs in §54.21(c)(1) and §54.21(d).
7 The staff notes that EPRI MRP submitted an eighty-year version of MRP-227 (as MRP-227, Revision 2) for staff approval on May 9, 2022 (ML22129A140 for the EPRI MRP transmittal letter; ML22129A141 for the MRP-227, Revision 2 report) and the staff accepted MRP-227, Revision 2 for a formal review in a staff-issued email to EPRI MRP dated June 17, 2022 (ML22145A401). The staffs review of MRP-227, Revision 2 is ongoing and the staffs decision on MRP-227, Revision 2 is pending.
3.4.3 ASME Code Criteria Applicability to Topical Report WCAP-17096, Revision 3 In Section 2 of the subject TR, the PWROG provided the following statements clarifying that specific acceptance standards in ASME Boiler and Pressure Vessel Code,Section XI Article IWB-3500 may not apply to the evaluation of PWR RVI component inspection results:
While these acceptance standards are appropriate for the ASME Code Section XI inspections, including those that are specifically highlighted in MRP-227 as existing reactor internals inspection requirements, they are not fully applicable to the MRP-227 inspection recommendations. In particular, the linear flaw standards of IWB-3510, as intended to guard against propagation of cracks through the reactor pressure vessel, are not meaningful when applied to the removable internals components that do not provide a pressure boundary.
The NRC staff determined that the PWROGs general statements in this regard would not be valid for NRC licensees implementing MRP-based PWR RVI AMPs that: (1) rely on MRP-227, Revision 1-A, and (2) include Existing Program criteria for specified ASME Code Class 1 core support structure components or reactor internal interior attachments. Specifically, such criteria would apply to W- and CE-design RVI components that: (1) are defined as Existing Program components in Table 4-8 or 4-9 of MRP-227, Revision 1-A, and (2) are referenced in ASME Code Class inspection criteria by the line item for the specific component type in either Table 4-8 or 4-9 of MRP-227, Revision 1-A.8 Although the NRC staff agrees with the PWROG conclusion that PWR RVI components are not defined by ASME International as ASME Code Class 1 reactor coolant pressure boundary components, they may be defined in the CLB as ASME Code Class 1 interior attachments or core support structure components under ASME Section XI, Table IWB-2500-1, Examination Category B-N-2 or B-N-3 criteria.
In addition, the EPRI MRP makes the following statements in the executive summary of MRP-227, Revision 1-A regarding 10 CFR 50.55a and ASME Code compliance:
These guidelines are not intended to reduce, alter, or otherwise affect current American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI [2] or plant-specific licensing inservice inspection requirements. Where ASME Code Section XI examinations are credited for aging management as Existing Programs, these guidelines provide the specificity considered necessary to ensure that the examinations meet the intent for which they are credited.
Therefore, for PWR RVI components that are defined in the CLB as ASME Code Class 1 Examination Category B-N-2 interior attachments or B-N-3 core support structure components (regardless of the MRP-227, Revision 1-A category classification), the acceptance criteria for the components are established according to one of the following In-Service Inspection (ISI) program bases:
The applicable acceptance criteria for the component in ASME Section XI, Article IWB--3000 8 These criteria do not apply to B&W designed PWRs as the EPRI MRP does not include any RVI components in the Existing Program category of MRP-227, Revision 1-A for B&W designed plants.
Through use of an NRC-approved ASME Section XI Code Case (Reference 30) endorsed in the most recent staff-issued version of RG 1.147 (Reference 29), as referenced in 10 CFR 50.55a(b) and subject to any limitations or conditions placed on use of the Code Case in the RG Through implementation of alternate, NRC-approved ISI acceptance criteria and data requirements that are authorized by NRC through implementation of a 10 CFR 50.55a(z) relief request review and approval process The need to follow the above referenced ASME Section XI or 10 CFR 50.55a process criteria for RVI components that are included as ASME Code Class 1 or Class A components in the plant design is independent of the EPRI MRP category assigned to the component in MRP-227, Revision 1-A or in supplemental EPRI methodologies (e.g., EPRI MRP 2018-022).
3.4.4 Reinspection Frequencies for Baffle-Former Bolt Inspections In Section 2.3 of the subject TR, the PWROG referenced the EPRI MRPs interim inspection criteria for W-design baffle-former bolt (BFB) inspections, as established in MRP-2016-021 (Reference 32) or PWROG-17071-NP, Revision 0 (Reference 34). The PWROG also referenced EPRIs MRP responses to the NRCs RAIs on the Interim Guidance Reports, dated July 13, 2017 (Reference 46),
and NRCs approval basis for accepting NEI 03-08 versions (Reference 36).
In Section 2.3.2 of the topical report PWROG-17071-NP, Revision 0, the PWROG included the following Needed criterion, which was created based on a Westinghouse PWR licensees proposed reinspection interval bases for its BFBs:
Any plant-specific evaluation used to extend the re-inspection interval beyond those defined in MRP-2017-009 is to be submitted to the NRC for information at least one year prior to the end of the current applicable interval for BFB subsequent examination.
The NRC staff previously accounted for the plant-specific evaluation and reinspection need bases of W-design BFBs through the staffs issuance of A/LAI-01 in the NRCs SE dated April 25, 2019, for MRP-227, Revision 1-A. To address the A/LAI-01 in Revision 3 to the TR, the PWROG updated Appendix E to the subject TR, W-ID Item 6 to include the following additional actions for BFBs in the Additional Action(s) section of the component ID item:
If the table in MRP 2017-009 indicates that the subsequent inspection interval is not to exceed 6 years (e.g., downflow plants with 3 percent BFBs with indications or clustering, or upflow plants with 5 percent of BFBs with indications or clustering), the plant-specific evaluation to determine a subsequent inspection interval shall be submitted to the NRC for information within one year following the outage in which the degradation was found. Any evaluation to lengthen the determined inspection interval or to exceed the maximum inspection interval recommended in MRP 2017-009 shall be submitted to the NRC for information at least one year prior to the end of the current applicable interval for BFB subsequent examination.
The NRC staff finds that the supplemental BFB analysis and reinspection criteria is acceptable because they are consistent with the NRCs recommendations for W-design BFB reinspection intervals, as discussed in the NRCs April 25, 2019, SE for MRP-227, Revision 1-A (Reference 17).
For LRA/SLRA of W-design PWRs that include the subject TR within the scope of its LRA/SLRA PWR Vessel Internals AMPs, licensees may address the A/LAI-01-based reinspection intervals for the BFBs in either the monitoring and trending or operating experience program element assessments of their AMPs by referencing the PWROGs statements in the Additional Action(s) section of Appendix E of the subject TR, W-ID Item 6. The timeframe for determining whether a supplemental analysis for BFB is necessary is appropriately defined in the additional action for the components in the subject TR. Section 3.3.11 of this SE provides additional clarification regarding the use of corrective actions for a given component and have been incorporated in the subject TR. Based on these considerations, the NRC staff finds the methodology guidance in Section 2.3 of the subject TR to be acceptable.
3.4.5 Analyses for Ultrasonic Test Results of Bolted Assemblies In Section 3.3 of the subject TR, the PWROG discussed its bases for evaluating the results of ultrasonic test (UT) inspections performed on bolted RVI assemblies, including, but not limited to the evaluation of BFBs used in the fabrication of the W-design baffle-former assemblies. Section 3.3 of the subject TR also includes discussions of acceptable bolting pattern analysis (ABPA) methodologies and use of acceptable bolting patterns (ABPs) that may be used to justify the failure of a prescribed number of bolts without causing structural integrity issues with the RVI assembly containing the bolts or to justify reinspection intervals of the bolts contained in these assemblies. Specifically, the PWROG added the following additional action to the applicable sections of the subject TR to include ABPAs for bolted assembly evaluations:
The use of acceptable bolting patterns as acceptance criteria that allow individual bolt failures is established in the industry. The PWROG (from prior Westinghouse Owners Group [WOG] efforts) has developed acceptable bolting patterns for baffle-to-former bolting for Westinghouse reactor internals designs. The plant-specific applicability of these generic acceptable patterns must be confirmed by the user since certain inputs may have changed over the two decades since these patterns were originally developed. The PWROG has supported development of similar strategies for core barrel bolt inspections in Babcock & Wilcox (B&W) plants. The methodology for performing an acceptable bolting pattern or similar strategy is beyond the scope of the current task.
The NRC staff transmitted RAI-07, Parts (a), (b), and (c) to the PWROG on September 21, 2020 (Reference 23), to address three items which were related to the PWROGs basis for using the ABPA in the evaluation of degraded bolts in RVI bolted connections. In the PWROGs response to RAI-07, Parts (a), (b), and (c) dated January 14, 2022 (Reference 43), PWROGs response to Part (a) discusses the ABPA from the proprietary version of WCAP15029P-A (Reference 37) for CE-design core shroud bolts and barrel-to-shroud bolts; the only two CE-designed power plants that link the inspection and ABPA evaluation criteria are the nuclear reactors at the Fort Calhoun and Palisades nuclear power stations. The PWROG indicated that the licensees for the PWRs at Fort Calhoun Station and Palisades Station have permanently ceased operations of the units. The current version of Operating License No. DPR-20 for Palisades Station indicates that the site is now operating in accordance with permanently defueled plant conditions and technical specification requirements.
Similarly, the current version of Operating License No. DPR-40 for Fort Calhoun Station indicates that the site is now operating in accordance with permanently defueled plant conditions and technical specification requirements.
The PWROG addressed this RAI by amending Appendix C to the subject TR by deleting the component acceptance criteria methodology and data requirements assessments for core shroud bolts and barrel-to-shroud bolts in Appendix C to the subject TR, CE- Item ID 1 and 1.2, along with the corresponding criteria for the linked Expansion category core support column bolts in Appendix C to the subject TR, CE-ID Item 1.1 and associated flow charts for these components in the corresponding component ID item in Appendix D to the subject TR. Given that the NRC staff confirmed that the referenced ABPA acceptance criteria methodology and data requirements are only for the specified bolting types in non-operating CE-design reactors (i.e., Fort Calhoun and Palisades), the NRC staff finds these proposed changes in the subject TR for CE-ID item to be acceptable, as the MRP-227-based AMPs would no longer be implemented at these sites because they have both permanently ceased power operations. Therefore, RAI-07, Part (a) is resolved.
Additionally, in the PWROGs January 14, 2022, response to RAI-07, Part (b), the PWROG confirmed that the methods in WCAP-18034-P are used to supplement acceptance criteria methodology and data requirements in WCAP-15029-P-A, Revision 1, for application to the W-ID item assessments for the BFB in Appendix E to the subject TR, W-ID Item 6 and for the barrel-to-former bolts in Appendix E to the subject TR, W-ID Item 6.1. The NRC staff confirmed that the PWROG amended Appendix E to the subject TR, W-ID Items 6 and 6.1 to include these clarifications. Based on the PWROGs clarifications and the amendment of Appendix E to the subject TR, W-ID Items 6 and 6.1, the NRC staff finds that the subject TR appropriately addresses which NRC-approved ABPA methodologies that will be applied to the assessments of W -design BFBs and core barrel-to-former bolts. Therefore, RAI-07, Part (b) is resolved.
Moreover, in the PWROGs January 14, 2022, response to RAI-07, Part (c), the PWROG restated its position that there are no NEI 03-08 Mandatory, Needed or Good Practice criteria in the subject TR, but indicated that the subject TR can be used to meet the NEI 03-08 Needed criterion as specified in Section 7.5 of MRP-227, Revision 1-A. Therefore, the NRC staff finds this aspect of the response to RAI-07, Part (c) acceptable because it is consistent with the implementation bases in Section 7 of the MRP-227, Revision 1-A.
The PWROG also revised its response to RAI-07, Part (c) in the PWROGs letter of January 14, 2022 (Reference 24 for the transmittal letter; Reference 25 for the revised RAI response), that superseded the PWROGs prior response to RAI-07, Part (c) previously submitted in the PWROGs letter dated March 30, 2021 (see Reference 43). Specifically, the PWROG indicated that the discussion in Section 3.3 of the subject TR regarding the confirmation of the plant-specific applicability of a generic ABPA is a process step that is discussed within the scope of the W-ID Item 6 methodology in the subject TR, rather than Section 3.3 of the subject TR. Therefore, the PWROG revised RAI-07, Part (c) in its response transmitted on February 8, 2022 (see Reference 43) as discussed below:
Replacement of the Approach section statement of W-ID Item 6: generic work completed in previous PWROG program with the statement generic ABPAs that were previously performed should be confirmed to be applicable on a plant-specific basis Deletion of the following statement in Section 3.3 of the subject TR: The plant-specific applicability of the generic acceptable patterns must be confirmed by the user, since certain inputs may have been changed over two decades since these patterns were originally developed."
The NRC staff did not find the PWROGs changes to the RAI responses to be acceptable (i.e., the complete deletion of the statement in Section 3.3 to the subject TR and limited scope on use of
APBAs limited only to the bolting assessments for W design BFB in the subject TR, Appendix E, W-ID Item 6) because, even though changes were made to the subject TR in the PWROGs January 13, 2022, RAI response, the PWROG was still using NRC-approved APBAs, or have potential use of APBAs, in other ID items assessments in the subject TR Appendices A or E (including, but not limited to Appendices A or E):
WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6.1 for W-design Expansion category barrel-to-former bolts (which, like W-ID Item 6 for W-design BFB, references use of the NRC-approved ABPA methodology in Westinghouse Class 2 Proprietary Report No. WCAP-15029-P-A, Revision 1 (Reference 38)).
Multiple B&W-ID Item assessments in WCAP-17096-NP, Revision 3, Appendix A for bolted B&W-design Primary or Expansion category bolting types that referenced potential use of ABPAs for ensuring the integrity of the bolted assemblies (where the PWROG had indicated that ABPA methodology could be developed for the applicable bolting type).
With respect to this observation, the NRC staff has previously approved the use of ABPAs only for W-design baffle-to-former bolts and barrel-to-former bolts (WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6 and 6.1 components) based on NRC staffs past approval of WCAP-15029-P-A, Revision 1, in the NRCs SE (Reference 33). Additionally, in accordance with the basis in the NRCs SE WCAP-15029-P-A, Revision 1, the scope of the NRC approval was limited to only the deterministic ABPA methodology basis which would be used to generate an ABP or set of ABPs for the bolts.
NRCs WCAP-15029-P-A, Revision 1, SE did not cover any methodology for the probabilistic Monte Carlo methods that might be used to generate a set of unit-specific bolting patterns (BPs) and that would then be compared against the NRC-approved ABP or set of ABPs for the W-design baffle-to-former bolts and barrel-to-former bolts. The NRC staff has not approved any ABPA and Monte Carlo methodologies for generating ABPs and unit-specific BPs for other types of W-bolted components or for CE-design or B&W-designed bolted component types. Therefore, with the inclusion of the new Condition 2 (new C-02) (WCAP-17096-NP, Revision 3) (discussed below in Section 3.4.6), if BPs in these other bolted component types are evaluated using unapproved ABPA, these evaluations should be made available to the NRC as discussed in Section 3.4.6 of this SE, which includes the NRC staffs evaluation of the PWROGs response to RAI-13, Part (b) (Reference 26) and the linked resolution of RAI-07, Part (c).
3.4.6 Resolution of Operating Experience for Westinghouse-Design Baffle-Former Bolts, Thermal Shield Flexures, and Use of Predictive Bolting Pattern Models or Methods In Section 2.3 of the subject TR, the PWROG discussed its basis for dispositioning relevant OE with cracking in W-design BFB. The NRC staff observed that due to the timing of the subject TR publication which took place in July of 2019, the PWROG did not address/discuss the BFB degradation that occurred at the Salem, Unit 1 facility during the reinspection for the spring 2019 outage, following the BFB replacement in the unit during the spring of 2016. Indications in the thermal shield flexures were also found during the spring of 2019 reinspection at Salem, Unit 1. WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 9, provides evaluation methods for detected flaws in thermal shield flexures. The spring of 2019 reinspection at Salem, Unit 1 resulted in more than the expected number of degraded BFBs predicted by the methodology in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6, which describes the basic framework for a probabilistic predictive model as a basis for evaluating BFB reinspection intervals. W-ID Item 6 states that the probabilistic model must
predict both the proportion and spatial distribution (i.e., patterns) of failed bolts at certain time intervals. Acceptability of at least 95 percent of predictive model outputs relative to the acceptable bolting pattern is recommended as a basis for setting the reinspection interval. ABPA for as-found inspection results is based on the NRC-approved methodology in WCAP-15029-P-A, Revision 1.
On September 18, 2019, during an NRC closed meeting, Westinghouse (in partnership with the PWROG), gave a presentation (Reference 37) to NRC staff on Westinghouse Baffle-Former Bolt Predictive Methodology. The presentation provided an overview of the proprietary methodology for probabilistic predictive evaluations of BFB degradation as stated in the basic framework of WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6. The presentation included OE on the BFB degradation from the spring of 2019 outage at Salem, Unit 1. On September 21, 2020, the NRC staff issued RAI-13, Parts (a) to (g) (Reference 23) to address the unexpected number of degraded BFBs from the spring 2019 outage at Salem, Unit 1 (including information related to the thermal shield flexure indications) and clarifications provided in the September 2019 Westinghouse presentation on the probabilistic predictive methodology for evaluating BFB degradation.
The NRC staff reviewed the PWROGs September 2019 Westinghouse presentation on BFB degradation OE that provided information on how the predictive, probabilistic evaluation outputs compare to BFB reinspection data. However, the NRC staff determined that it would need further verification that the predictive, probabilistic method for the BFB was suitable for setting reinspection intervals of the bolts based on generating predictive outputs starting from an initial degraded pattern.
In the PWROGs RAI-13 response (see Reference 39), the PWROG responded to RAI-13, Parts (a) -
(g). In the PWROGs response to RAI-13 Part (a), the PWROG explained that the validity of the Westinghouse predictive, probabilistic model is demonstrated by considering the results from previous applications, which were used to predict both initial BFB UT results and reinspection BFB UT results, as shown in the September 2019 Westinghouse presentation. The NRC staff noted from its review of the response to RAI-13 Part (a) that the reinspection BFB UT results were within the bounds of the predictive model, for all except for one plant. Based on considerations and the NRC staffs review of the response to RAI-13, Part (a), the NRC staff finds that the predictive bolting pattern model is an acceptable method for establishing the reinspection interval for a specified BFB because the NRC staff has confirmed that the ABPA model is capable of generating a conservative, predictive set of future bolting pattern outputs starting from an initial degraded bolting pattern (as established from the last inspections performed on the assembly), and therefore, the issue in RAI-13, Part (a) is resolved.
With respect to the issue identified in RAI-13, Part (b), the NRC staff observed that if other vendors used probabilistic methods for bolted assemblies that are different from those described in the September 2019 Westinghouse presentation, these new probabilistic methods would need to be submitted for NRC review and approval. In the PWROGs response to RAI-13, Part (b), the PWROG explained that WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6, specifies that each application of the predictive model methodology in W-ID Item 6 will be submitted to the NRC for information if the percentiles defined in MRP 2017-009 are exceeded, and that based on BWRVIP 2019-016 (Reference 47), the licensee submitting the request has the responsibility for making any proprietary details of the predictive methodology available to the NRC. Also, the PWROG stated that Framatome performs predictive BFB evaluations for W-design plants and that the methodology in Section A.1.8 of the subject TR is applied. The NRC staff finds that for W-design plants, the specification in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6 to submit the plant-specific application of the predictive model methodology defined in W-ID Item 6 is an acceptable way for making the details of the predictive methodology available to the NRC. In accordance with WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6, submittal to the NRC is only for information and is only necessary if: (1)
MRP 2017-009 indicates that the subsequent inspection interval is not to exceed 6 years, or (2) there is an evaluation to lengthen the determined inspection interval or to exceed the maximum inspection
interval recommended in MRP 2017-009. The staff concludes that for W-design plants, the specification in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6, to submit the plant-specific application of the predictive model methodology defined in W-ID Item 6 is an acceptable way for making the details of the predictive methodology available to the NRC.
Furthermore, the NRC staff observed that the PWROGs referenced use of Framatome ABPA and predictive methodologies for B&W-design bolted assembly types have yet to be developed by Framatome at the time of the development of this SE report. The NRC staff would expect that the plant-specific bolting evaluation methods9 for the bolted assembly type to be placed into the NRC licensees corrective actions process consistent with either the requirements in 10 CFR Part 50, Appendix B, Criterion XVI or else the NEI 03-08 Needed corrective action specified in Section 7.5 of the MRP-227, Revision 1-A. The staff also expects these evaluation methods to be fully defined, discussed, and justified in an applicable site-specific document or record. This is part of a new Condition (C-02) (WCAP-17096-NP, Revision 3) as defined and discussed in Table A-1 of Attachment A of this SE, which relates to NRC staffs evaluation items discussed in RAI-07 Part (c) (as carried over in Section 3.3.5 of this SE) and RAI-13 Part (b) as discussed and evaluated above. Based on the addition of C-02 (WCAP-17096-NP, Revision 3), the items discussed in RAI-07, Part (c) and RAI-13, Part (b) are resolved.
C-02 (WCAP-17096-NP, Revision 3):
For bolted assembly analysis methods that use predictive, probabilistic ABPA models, the models and methods should appropriately account for limits on the number of allowable bolts with presumed failed conditions, geometric bolt failure considerations (i.e., bolt clustering considerations), and bolting analysis reliability considerations (i.e., addressing 95% reliability confidence limits.) These models and methods should be appropriately documented and justified in a site-specific or owner-defined record.
With respect to the items discussed in RAI-13 Part (c), the NRC staff observed that potential errors in the UT data of the BFBs, in the inputs into the predictive methodology, or in the methodology itself could have caused the unexpected number of degraded BFBs discovered during the spring of 2019 outage at Salem, Unit 1. In the PWROGs response to RAI-13, Part (c), it discussed the root cause evaluation of the unexpected number of degraded BFBs from the spring of 2019 outage at Salem, Unit 1. Specifically, in its root cause evaluation, the PWROG concluded that the detected degradation in the thermal shield flexures resulted in increased applied stresses to the BFBs as a result of additional vibrational loads disseminating from the thermal shield, which lead to the unexpected number of degraded BFBs at Salem, Unit 1. Based on the conclusion of the root cause evaluation, the PWROG stated that Section 3 of the subject TR will be revised to clarify that the process of evaluating degradation (in general) in the subject TR assumes that each component degradation is independent of degradation of other components. The NRC staff reviewed the excerpts of the root cause evaluation provided in the supplement, and determined that the unexpected number of degraded BFBs in the spring 2019 outage at Salem, Unit 1 was not due to errors in the BFB UT data, the inputs into the predictive methodology, or in the methodology itself because the root cause of the degradation found in the BFB and the number of BFB found with relevant crack-like indications was 9 This is in reference to both: (1) a deterministic ABPA methodology for generating an approved bolting pattern [ABP] or set of ABPs used for structural integrity acceptance criteria objectives of the bolted assembly, and (2) either a deterministic or probabilistic bolting pattern methodology for generating a set of unit specific bolting patterns (BPs) for the assembly that would then be compared to the ABP or set of ABPs.
vibrations coming from the degraded thermal shield flexures. Therefore, the item discussed in RAI-13 Part (c) is resolved.
With respect to the item discussed in RAI-13 Part (d), the NRC staff observed that the 95 percent reliability criterion, i.e., a 95 percent probability of retaining an acceptable bolting pattern at reinspection compared against plant specific ABPAs, means that there is a 5 percent probability that the predictive model would result in an unacceptable bolting pattern at reinspection. In the PWROGs response to RAI-13 Part (d), it states that the 95 percent reliability criterion in terms of an acceptable bolting pattern adds a layer of conservatism to a scenario that is already judged to have low risk significance with respect to core damage frequency. The PWROG also stated that passive failures, such as BFB failures, are not typically modeled in plant-specific probabilistic risk assessments (PRAs) since failure probabilities associated with passive failures are several orders of magnitude lower than other active failures. The PWROG discussed the conservatisms in both the Westinghouse predictive model, and ABPA methodology (in WCAP-15029-P-A, Revision 1) that contribute to the low-risk significance of BFB degradation. The NRC staff especially observed the conservatism in what constitutes as core damage for a given BFB-related event and verified that this core damage criterion is more conservative than that used in PRAs as defined in the PRA Standard. The NRC staff confirmed that this referenced PRA Standard is the same standard for PRA evaluations specified in RG 1.174, Revision 3, for evaluating risk due to changes to the licensing bases. The NRC staff finds that the 95 percent reliability criterion is acceptable when the Westinghouse predictive model and ABPA methodologies are used for evaluating BFB degradation because of the conservatisms in both methodologies. However, the NRC staff determined that if the Framatome predictive model and ABPA methodologies are used for evaluating BFB degradation, the conservatisms in the methodologies would need to be confirmed for the acceptability of the 95 percent reliability criterion. This is addressed as part of the new C-02 (WCAP-17096-NP, Revision 3), as defined and discussed in Table A-1 of Attachment A in this SE. Based on the addition of new C-02 (WCAP-17096-NP, Revision 3), the item in RAI-13 Part (d), is resolved.
The methodology in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6, refers to consideration of a portion of untestable bolts as functional when evaluating a bolting pattern in general or evaluating a bolting pattern with clustering. With respect to the item discussed in RAI-13 Part (e), the NRC staff observed that the subject TR does not include any specific guidance on how a plant should evaluate untestable bolts to determine the extent to which they may be credited as functional structural members. In the PWROGs response to RAI-13 Part (e), it stated that MRP-227, Revision 1-A allows a certain number of untestable bolts to be excluded when performing the MRP-227 inspections of BFBs, and that in general, when evaluating reduced bolting patterns based on inspection results, the most conservative approach is to consider all untestable bolts as failed. The PWROG also stated that a sampling approach was determined appropriate by the NRC, referring to the NRCs SE for WCAP-17096-NP, Revision 2 (Reference 13). The PWROG stated that the sampling approach is also intended to be used when untestable bolts are relevant to the determination of clustering in the reduced bolting pattern, and that Note 1 of Step 2 of WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6, will be revised to reflect this intent. The NRC staff reviewed the response and the NRCs SE of WCAP-17096-NP, Revision 2, regarding treatment of untestable bolts. The NRC staff finds that the methodology summarized by the PWROG for evaluating untestable bolts to determine the extent to which they may be credited as functional structural members acceptable because (1) treating all untestable bolts as failed (meaning they are not credited as a load bearing member) is conservative; and (2) adding the statistically determined number of failed bolts in the untested (i.e., uninspected) population to the number of failed bolts in the tested (i.e., inspected) population provides a conservative means of estimating the total number of failed bolts, as determined in the NRC SE that is included in Revision 2 of the TR. Additionally, consistent with the new C-02 (WCAP-17096-NP, Revision 3), use of predictive bolting pattern methodologies for PWR bolted assembly objectives
should be appropriately defined, discussed, and justified in a site-specific document or record. Based on these considerations and the addition of the new C-02 (WCAP-17096-NP, Revision 3), the items discussed in RAI-13 Part (e), is resolved.
With respect to the item discussed in RAI-13, Part (f), the NRC staff observed that the language on reinspection intervals in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6, is inconsistent with that in MRP-227, Revision 1-A and MRP 2017-009. In its response to RAI-13, Part (f), the PWROG acknowledged the existence of the perceived inconsistency and stated that step (c) of WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 6, will be revised to include a statement that a submittal of a deviation to MRP-227 guidelines would be needed for reinspection intervals that are beyond 10-years.
The NRC staff reviewed the response and finds that the additional statement on submittal of a deviation to MRP-227 guidelines to be acceptable because the PWROGs basis is consistent with EPRI MRP programmatic deviation identification and resolution criteria for these types of programs, as defined in Section 7 of MRP-227, Revision 1-A. The item discussed in RAI-13 Part (f), is resolved.
With respect to the item discussed in RAI-13, Part (g), the NRC staff observed that the guidance in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 9, for thermal shield flexures does not include the inspection guidance from Technical Bulletin (TB)-19-5 that Westinghouse issued after the determination that degraded thermal shield flexures were the root cause of the unexpected number of degraded BFBs at Salem, Unit 1 from the spring of 2019 outage (refer to the earlier discussion of the response to RAI-13 Part (c) in this SE). In the PWROGs response to RAI-13, Part (g), the PWROG explained that Revision 0 of TB-19-5 is referenced in MRP-227, Revision 1-A and that Revision 1 of TB-19-5 will be referenced in MRP-227, Revision 2 instead of in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 9, because the MRP-227 document in general (not the acceptance criteria methodology in the WCAP-17096 document in general) is associated with the inspection of thermal shield flexures and support block cap screws. The PWROG also stated that the thermal shield flexure methodology in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 9, postulates the existence of failed flexures and the likelihood that one failed flexure would result in a cascading failure of additional flexures. The NRC staff reviewed the response and WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 9, and determined that the basis for not including the TB-19-5 inspection guidance in WCAP-17096-NP, Revision 3, Appendix E, W-ID Item 9, is acceptable because the NRC staff has confirmed that Item W9 in Table 4-3 of MRP-227, Revision 1-A, already references TB-19-5 as an inspection standard for W-design thermal shield flexures. The item analyzed in RAI-13 Part (g) is resolved.
Based on the discussion above, the NRC staff finds that the PWROG has adequately dispositioned relevant OE with respect to cracking in W-design BFBs and thermal shield flexures.
3.4.7 Acceptance Criteria Methodology and Data Requirements Used for the Evaluation of Component Void Swelling or Distortion The NRC staff verified that Section 3.5 of the subject TR included the PWROGs bases for using visual inspection techniques as a method for detecting changes in dimension that may occur in PWR RVI components as a result of void swelling or distortion. Section 3.5 also included the bases for evaluating such effects, including the PWROGs discussions with respect to the detection and evaluation of swelling or distortion that may occur in the gap areas of CE-designed core shrouds that are fabricated in two vertical sections, as stated in WCAP-17096-NP, Revision 3, Appendix C, CE-ID Item 4a. The PWROG stated that the evaluation of the gap areas of the shrouds will require additional sensitivity studies to relate the swelling level to the predicted distortion and gap opening in the structure, and that the sensitivity studies may be necessary to evaluate the current condition,
predict future effects, and determine aging management strategies, which are beyond the scope of the TR.
In addition, the NRC staff observed that the PWROG did not address any acceptance criteria or data analysis criteria for managing component changes in dimension due to void swelling or distortion in the subject TR. The NRC staff also observed that, as of the date of this SE, there has been little industry-reported OE with reported indications of changes in dimension, swelling, or distortion in W-design, CE-design, or B&W-design RVI components that the EPRI MRP has identified as being potentially susceptible to swelling or distortion effects. The NRC staff addressed this informational gap with the issuance of September 21, 2020, RAI-11 (Reference 23), in which the staff asked the PWROG to clarify whether the intent of Section 3.5 of the subject TR is for swelling-related distortion assessments to be the responsibility of licensees with the implementation of component analysis and acceptance criteria for these detected aging effects.
In the PWROGs response to RAI-11 (Reference 25), revisions were made to Section 3.5 of the subject TR to clarify that a plant-specific plan should be developed for evaluating mitigating void swelling. Based on the PWROGs response to RAI-11 and its amendment to Section 3.5 of the subject TR, the NRC staff finds that the scope of the subject TR is limited in that it does not include any acceptance and data analysis criteria for evaluating component distortion, or void swelling, or for managing changes in dimension due to distortion, or void swelling in the PWR Primary or Expansion category components that are within the scope of the subject TR. Because the PWROG is placing the responsibility for assessing potential RVI component void swelling or distortion impacts on the license holder of the PWR unit, the use of the subject TR for void swelling or distortion evaluation objectives is subject to L-02 (WCAP-17096-NP, Revision 3), as established and defined in Table A-1 of Attachment A in this SE, List of Limitations and Conditions for WCAP-17096, Revision 3. Based on the addition of L-02 (WCAP-17096-NP, Revision 3), the item discussed in RAI-11 is resolved.
L-02 (WCAP-17096-NP, Revision 3):
The scope of WCAP-17096-NP, Revision 3, is limited in that it does not include any acceptance and data analysis criteria for evaluating component distortion or void swelling, or for managing changes in dimension due to distortion or void swelling in the PWR Primary or Expansion category components that are within the scope of WCAP-17096-NP, Revision 3. If distortion or changes in dimension due to void swelling is detected in an RVI component, a licensee using this TR will need to address distortion or changes in dimension due to void swelling on a component-specific basis per the acceptance criteria and corrective actions program element bases of its PWR Vessel Internals Programs and its 10 CFR Part 50, Appendix B program.
3.4.8 Combustion Engineering Design Components Assessed by Fatigue Evaluations In Section 3.6 of the subject TR, the PWROG discussed how aging management of three CE-design Primary category components may be achieved through use of component fatigue or cyclical loading analyses. The NRC staff determined that, even though the changes to Section 3.6 of the subject TR represent Administrative Edits of the section, the process described in the subject TR warrants further evaluation and discussions with respect to a licensees basis (for CE-designed plants only) for managing these CE-design components during an initial period of extended operation (i.e., licensed 40 to 60 years) or a subsequent period of extended operation (i.e., licensed 60 to 80 years).
The three CE-design Primary category components that are subject to the EPRI MRPs fatigue screening criteria in MRP-227, Revision 1-A, and to the PWROGs corresponding statements in Section 3.6 of the subject TR and the CE-ID Items for the components in the subject TR Appendix C are:
CSB flexure weld located in the CSB assembly of the plant design, core support plate in the lower support structure of the plant design, and fuel alignment plate in the upper internals assembly of the plant design The NRC staff observed that the PWROGs basis for evaluating these components in Section 3.6 of the subject TR (and in the referenced CE-ID item assessments in the subject TR Appendix C) are consistent with those established in the MRP-227, Revision 1-A. The EPRI MRP criteria establish the need for performing enhanced visual (EVT-1) examinations of the referenced CE-design Primary category component only if a fatigue analysis of the specified component in the CLB cannot demonstrate adequate screening against fatigue impacts. However, the concept behind the EPRI MRPs criteria assumes that the CLB for referenced CE-design Primary category components will include at least one form of fatigue or cyclical loading analysis for the fatigue screening objective and that the analysis will be sufficient to demonstrate that: (1) initiation of a fatigue-induced macro crack or fatigue growth coalescence of two or more micro-cracks into a detectable macrocrack is unlikely to occur at the facility for the licensing period of interest, (2) the amount of analyzed and projected growth of a detected or postulated macro-crack in the component will remain within an specified allowable limit on stable flaw size through the end of the licensing period of interest, or (3) the fracture toughness values for the component will not be less than a specified lower-bound limit on acceptable fracture toughness values for the component.10 However, the NRC staff also observed that there were specific regulatory and technical nuances on these fatigue-related Primary bases that require further explanations in this SE.
Specifically, the NRC staff observed that the EPRI MRPs I&E bases defined in MRP-227, Revision 1-A, Table 4-2 (and the corresponding PWROGs discussions in the subject TR in Section 3.6 and in Appendix C, CE-ID Items 7, 8, and 9), did not define the specific types of fatigue or cyclical loading analyses that could be used for performing the referenced fatigue screening objective for the specified component; however, the NRC staff acknowledges that analysis may be a cycle dependent cumulative factor analysis, fatigue waiver or exemption analysis, crack growth or flaw growth analysis, or fracture mechanics analysis. Due to the cyclical nature of these analyses, the NRC staff acknowledges that these types of analyses are often identified by LR/SLR applicants as being TLAAs for their LRAs or SLRAs, which is consistent with the statement in footnote 2 of the subject TR.
The NRC staff also observed that EPRI MRPs criteria for these components are less clearly defined if the CLB for the specified component does not include at least one type of fatigue or cyclical loading analysis that could be credited for the referenced fatigue screening objective. Furthermore, to bypass the Primary EVT-1 examinations of CE-design CSB flexure welds, screening of the flexure welds for stress corrosion cracking (SCC) would also need to be performed, in addition to screening of fatigue, since the EPRI MRP identifies that CSB flexure welds are susceptible to both fatigue and SCC in Item C7 of MRP-227, Revision 1-A, Table 4-2. The NRC staff observed that the PWROG did not 10 For example, a component cumulative usage factor (CUF) analysis or ASME Section III fatigue waiver or exemption analysis in the CLB could be used for satisfying the first numbered clause (1) in this sentence. Similarly, a component, cycle-dependent fatigue flaw growth analysis could be used to satisfy the second numbered clause (2) in this sentence, and a component, cycle-dependent linear elastic or elastic plastic fracture mechanics analysis could be used to satisfy the third numbered clause (3) in this sentence.
address the susceptibility of fatigue and SCC on welds in Section 3.6 of the subject TR (even though the PWROG did account for it in WCAP-17096-NP, Revision 3, Appendix C, CE-ID Item 7, for the CSB flexure welds). The NRC staff addressed these informational gaps with the issuance of September 21, 2020, RAI-08, Parts (a), (b), and (c) (Reference 23). In the PWROGs final responses to RAI-08, Parts (a), (b), and (c) transmitted on January 14, 2022 (see Reference 25), revisions were made to the subject TR in Section 3.6 and in Appendix C, CE-ID Items 7, 8, and 9, to include fatigue screening evaluation criteria for CE-design core support barrel (CSB) flexure welds, core support plates, and fuel alignment plates and SCC screening evaluation criteria for CE-design CSB flexure welds.
Based on the changes to the responses to RAI-08 Parts (a), (b), and (c) (Reference 25), from the previous versions, the NRC staff observed that the PWROG is now placing fatigue screening needs for the referenced CSB flexure welds, core support plates, and fuel alignment plates, and the SCC screening needs for CSB flexure welds outside the scope of the subject TR instead of amending the report to include the appropriate acceptance criteria methodology and data requirements. Specifically, for the Item 7, 8, and 9, assessments evaluated in the PWROGs response to RAI-08 (Reference 25),
the PWROG removed the component-specific screening analysis assessment criteria for evaluating fatigue and SCC in CE-design core barrel flexure welds and fatigue in CE-design core support plates and upper fuel alignment plates from the scope of the referenced ID item assessments in Appendix C of the subject TR.
In order to address how to evaluate component screening assessment items that are outside of the scope of the subject TRs methodology, the NRC staff has identified a limitation on the RAI-08 fatigue and SCC screening topics and the lack of defined fatigue screening bases in the subject TR when referencing CSB flexure welds, core support plate and fuel alignment plates components (as assessed in CE-ID Items 7, 8, and 9) or defining SCC screening bases for the CSB flexures welds (CE-ID Item 7) in the subject TR. Therefore, these screening topics are the subject of L-03 (WCAP-17096-NP, Revision 3) and the clarifications made in the limitation, as established are defined in Table A-1 of Attachment A of this SE. Based on the addition of this limitation, the items discussed in RAI-08, Parts (a), (b), and (c), are resolved.
L-03 (WCAP-17096-NP, Revision 3):
The scope of the WCAP-17096-NP, Revision 3, is limited in that it no longer includes or defines the acceptance criteria or data analysis criteria for performing the fatigue and SCC screening analyses of CE-design core barrel flexure welds or fatigue screening analyses of CE-design core support plates and upper fuel alignment plates. Consistent with the guidelines in MRP-227, Revision 1-A, a licensee using the subject TR should address the acceptance criteria and data analysis criteria for performing the component-specific screening analyses of CE-design core barrel flexure welds, core support plates, and upper fuel alignment plates on a unit-specific or site-specific case-by-case basis.
3.4.9 Resolution for Technical Justification for Removal in Topical Report WCAP-17906, Revision 3 In Appendix A, of the subject TR, the PWROG identified that a technical justification for removal may be necessary for specific types of B&W designed components. The PWROGs statements apply to the component types in the following B&W-ID Item assessments in Appendix A:
Locking Devices for Baffle to Former Bolts and Internal Baffle to Baffle Bolts
Alloy X-750 Dowel-to-Guide Blocks Welds Upper Grid Fuel Assembly Dowel-to-Support Pad Welds Lower Grid Fuel Support Pads Items Alloy X-750 Dowel-to-Support Pad Welds In RAI-09, (Reference 23), the NRC staff informed the PWROG that it was not evident whether the referenced statement was addressing a physical removal of the referenced component through licensee performance of a justifiable design modification or removal of the component from either the Primary or Expansion category inspection category in MRP-227, Revision 1-A and recategorization as No Addition Measures components for the program. In the PWROGs response to RAI-09 (Reference 26), the PWROG clarified that the referenced statement is made in reference to a physical removal of the component in reference. Given that physical removal of the referenced components from the plant design would require the licensee to implement the design change through a 10 CFR 50.59 review, the NRC staff finds the explanation acceptable and RAI-09 is resolved.
3.4.10 Fracture Toughness Considerations In Appendices C and E of the subject TR, the PWROG referenced the following four neutron fluence-dependent, lower bound, linear elastic fracture mechanics (LB-LEFM) acceptance criterion values for specified SS base metal or weld components:
150 ksi-inch for components with project neutron fluence exposures 3x1020 n/cm2 (E > 1.0 MeV; 0.5 dpa) at the end of the service period of interest11 112 ksi-inch for components with project neutron fluence exposures > 3x1020 n/cm2 and 3x1021 n/cm2 (E > 1.0 MeV; > 0.5 dpa and 5 dpa) (E > 1.0 MeV; 0.5 dpa) at the end of the service period of interest 50 ksi-inch for components with project neutron fluence exposures > 3.0x1021 n/cm2 and 1.0x1022 n/cm2 (E > 1.0 MeV; > 5 dpa and 15 dpa) at the end of the service period of interest 34.6 ksi-inch for components with project neutron fluence exposures > 1.0x1022 n/cm2 (E > 1.0 MeV; > 15 dpa) at the end of the service period of interest For the referenced 150 ksi-inch, 112 ksi-inch, and 50 ksi-inch LB-LEFM values, the NRC staff observed that the applicant was indirectly referencing the EPRI BWRVIP-100, Revision 1-A (Reference 41)12 as the source document basis for the referenced LB-LEFM values. However, the NRC staff also observed that the PWROG did not provide the source document for the referenced LB-LEFM value of 34.6 ksi-inch because BWRVIP-100, Revision 1-A does not cover that lower bound LEFM fracture toughness value. In RAI-10 (Reference 23), the NRC staff asked for further justification for use of the 34.6 ksi-inch lower bound LEFM fracture toughness value for SS base metal or weld components. In the PWROGs response to RAI-10 (Reference 25), it cited EPRI MRP Report MRP-11 For example, the service period of interest could be at the end of the first renewed period of extended operation if made in reference to a first license renewal extended period of operation, or the end of the augmented inservice inspection interval of interest if being used for justification of a component reinspection interval that differs from the 10-year reinspection interval bases in MRP-227, Revision 1-A.
12 Updated bases for this EPRI report were approved in EPRI Proprietary Report No. 3002008388 (Reference 37).
21113 as the basis for the 34.6 ksi-inch lower bound LEFM fracture toughness value. In addition, the PWROG amended the source reference for the 34.6 ksi-inch line item entry in the fracture toughness table for CE-ID Item 2, Core Shroud Assembly (Welded), Core Shroud Plate-Former Plate Weld as MRP-211 and indicated that there are plans to amend the corresponding 34.6 ksi-inch based line item entries of the fracture toughness tables in the corresponding ID item assessments to include the MRP-211 in the following:
CE Core Shroud Assembly (Welded) - Core Shroud Plate-to-Former Plate Weld CE Core Shroud Assembly (Welded) - Remaining Axial Welds (CE-design units with welded core shrouds assembled in two vertical sections)
CE Core Shroud Assembly (Welded) - Shroud Plates CE Core Shroud Assembly (Welded), Remaining Axial Welds (CE-design units with welded core shrouds assembled with full height shroud plates)
CE CSB Assembly, UGW CE CSB Assembly, UAW CE CSB Assembly, MGW CE CSB Assembly, MAW CE CSB Assembly, LAW CE CSB Assembly, UFW CE CSB Assembly, LGW CE CSB Assembly, CSB Flexure Weld CE Lower Support Structure, Deep Beams (CE-design units with welded core shrouds assembled with full height shroud plates)
CE Lower Support Structure, Lower Core Support Beams (All CE-design units except those with welded core shrouds assembled with full height shroud plates)
W Core Barrel Assembly, UFW W Core Barrel Assembly, UGW W Core Barrel Assembly, UAW W Core Barrel Assembly, LGW W Core Barrel Assembly, MAW W Core Barrel Assembly, LAW W Core Barrel Assembly, LFW Since the PWROG has yet to adjust the lower bound fracture toughness entries in the ID items in the subject TR, the staff has addressed the gap as part of Condition C-03. From a technical perspective, the NRC staff observed that the 34.6 ksi-inch line item entries in the fracture toughness tables of the applicable ID item assessments continue to identify that the 34.6 ksi-inch based value becomes applicable at a projected neutron fluence range of > 1x1022 n/cm2 (E > 1.0 MeV), whereas as the MRP-211 document actually cites that the 34.6 ksi-inch lower bound LEFM value is applicable at a neutron fluence range somewhat lower than that (i.e., MRP-211 reports it as being applicable at projected neutron fluence exposures in excess of 6.7x1021 n/cm2 (E > 1.0 MeV) or projected doses in excess of 10 dpa). The NRC staff also observed an administrative error in the revised response to RAI-10 for the associated amendments of the fracture toughness tables in the February 8, 2022, RAI response, where the PWROG cited the applicable MRP report in the revised/updated footnote 2 of the tables as being MRP-221, and not as MRP-211 as indicated in the response to RAI-10.
13 For the objectives of the staffs SE, the staff is referencing Revision 1 of the MRP-211 Report (Reference 35).
Additionally, the NRC staff observed that, in the fracture toughness tables of the applicable ID items of
, Appendices C and E of the subject TR, the PWROG uses EPRI proprietary version BWRVIP-100, Revision 1-A as the source record for the cited lower bound LEFM value of 150 ksi-inch, which is being applied for SS components with projected neutron fluence exposures < 3x1020 n/cm2 (E > 1.0 MeV). Similarly, in the fracture toughness tables of the applicable ID items in Appendices C and E of the subject TR, the PWROG uses BWRVIP-100, Revision 1-A as the source record for the lower bound LEFM facture toughness values of 112 ksi-inch (applicable to SS components with projected component fluence exposures > 3x1020 n/cm2 and 3x1021 n/cm2 [E > 1.0 MeV]) and 50 ksi-inch (applicable to SS components with projected component fluence exposures > 3x1021 n/cm2 and 1x1022 n/cm2 [E > 1.0 MeV]). The NRC staff observed that in EPRI BWRVIP Letter No. 2021-030 (Reference 42), the EPRI BWRVIP reported new fracture toughness data may create some potential issues with the lower bound LEFM fracture toughness values reported for SS components with projected neutron fluence exposures in excess of 5x1020 n/cm2 (E > 1.0 MeV).
Based on the discussion in this section, NRC staff concluded that there may be some uncertainty with the accuracy of the fluence ranges reported for lower bound LEFM fracture toughness values reported in the fracture toughness tables provided in Appendices C and E of the subject TR. Therefore, the licensee should take actions to confirm that all reported lower bound LEFM fracture toughness values and the fluence ranges reported for the lower bound LEFM fracture toughness values in the subject TR are valid and accurate or else amend them accordingly. The selection of the applicable lower bound fracture toughness value(s) should include an appropriate justification and identified source record and should be reconciled with any updated industry reports or records used to correlate lower bound fracture toughness value(s) and the most recent neutron fluence range(s) reported for the specified fracture toughness value(s).
The NRC staff has identified this item as Condition C-03 (WCAP-17096-NP, Revision 3), as reflected in Table A-1 of Attachment A.
C-03 (WCAP-17096-NP, Revision 3):
Until such time that the EPRI or PWROG can resolve the EPRI BWRVIP Letter No. 2021-020 Part 21 issue on neutron fluence ranges associated with specified lower bound fracture toughness values reported in an updated version of MRP-210 or MRP-211 (or alternate industry record), licensees will need to justify any lower bound fracture toughness values used as inputs in component-specific flaw tolerance analyses. The selection of the applicable lower bound fracture toughness value(s) should be reported and justified in a site-specific record and reconciled with any updated industry reports or records used to correlate lower bound fracture toughness value(s) and the most recent neutron fluence range(s) reported for the specified fracture toughness value(s).
3.4.11 Babcock and Wilcox-Design Components Currently Lacking a Formal Inspection Standard or Aging Effect Analysis Methodology For B&W-design RVI components corresponding to the applicable ID item in Appendix A of the subject TR, the NRC staff observed that the ID item assessments indicated that the components will either receive Primary VT-3 visual inspections or potential Expansion based VT-3 visual inspections if the expanded inspections of the Expansion category component type is found to be applicable from results of visual inspections performed in the linked Primary component. The NRC staff observed that the PWROG included the following statement from Appendix A Section 1.2
(Control Rod Guide Tube Spacer Castings) in the subject TR for a number of B&W component ID item assessments:
The NDE inspection standard could be developed generically.
The NRC staff observed that for components in the applicable sections of the Appendix A of the subject TR, that included statements of this nature, the report did not firmly establish what the acceptance criteria methodology and data requirements would be for evaluating the results of VT-3 visual inspections applied to the components (or that might be applied to components that are designated as Expansion category components for the program). Therefore, the NRC staff addressed this information gap in RAI-12 (Reference 23).
In the PWROGs response to RAI-12 (Reference 26), it stated that the referenced B&W-ID Items include general provisions for identification of: (1) what visual examination wear indications are considered rejectable and what would require additional examination and evaluation, (2) any additional examination results that are anticipated and the general acceptance criteria for the additional items expected to be in the VT-3 visual examination field of vision. However, the PWROG admitted that these statements did not fully define the acceptance and data analysis criteria for the components in these B&W-ID item assessments. Therefore, the PWROG proposed to replace the statements with the following additional action for B&W-ID Items:
The VT-3 exam acceptance criteria are no relevant conditions. Relevant conditions for the B&W Items of interest are defined in Section 5 of MRP-227-1A. These relevant conditions constitute discontinuities or imperfections (e.g., loose or missing parts, distortion, corrosion, wear, etc.), which are readily detectable with the VT-3 technique.
Note that the VT-3 exam method meets the requirements of ASME B&PV Code Section XI, as specified and conditioned by 10 CFR 50.55a, and the MRP-228 Inspection Standard.
For B&W components in those referenced B&W-ID item assessments calling for VT-3 visual inspections (e.g., the CRGT spacer castings in B&W-ID Item A.1.2, etc.), the PWROGs revised data analysis and acceptance criteria basis in the above reference statement creates a limitation on the scope of the report because it defines relevant VT-3 visual detected conditions as being those that are defined in MRP-227, Revision 1-A and not on those that would be included and defined in the applicable B&W-ID items referenced above. Furthermore, Section 5 of the MRP-227, Revision 1-A only gives examples relevant VT-3 visual detected conditions for aging mechanism parameters inspected by VT-3 visual methods and in many cases leaves the full criteria for defining relevant VT-3 visual detected conditions up to the licensees implementing their PWR Vessel Internals Programs (or equivalent named program for the internals in the CLB). Therefore, the PWROG is holding the licensee responsible for adding the inspected component in the NRC licensees corrective action program based on: (1) a licensees finding of a relevant condition, (2) how the licensee defines relevant conditions for the applicable component to match up to the general relevant condition bases for the component in Table 5-1 of the MRP-227, Revision 1-A (e.g. as generally defined Line Item C2 for the CRGT spacer castings in Table 5-1 of the MRP-227, Revision 1-A), and (3) how many relevant conditions would need to be detected in order to expand the VT-3 visual inspection basis to any Expansion category components that may be linked to the referenced Primary component.
Given that the PWROG is no longer in the position of defining the relevant VT-3 conditions or defining the acceptance criteria methodology and data requirements for relevant VT-3 conditions that may potentially be detected in the referenced B&W-design components of the applicable B&W-ID items, the subject TR is limited in that it places the responsibility on the licensee holding the first renewed or
subsequent renewed operating licensee of the B&W designed PWR to define the relevant VT-3 conditions and the acceptance criteria for relevant conditions of the referenced B&W components. In accordance with its operating license, the licensee should define these aspects of the programs as part of its PWR vessel internals AMP. This is identified as L-04, which is explained in Table A-1 of Attachment A of this SE. Based on the addition of this Limitation, RAI-12 is resolved.
L-04 (WCAP-17096-NP, Revision 3):
The scope of the WCAP-17096-NP, Revision 3, is limited in that it does not define any acceptance criteria or data analysis criteria for evaluating relevant conditions in B&W-design RVI components that are inspected using VT-3 visual inspection methods. Consistent with the guidelines in MRP-227, Revision 1-A, the acceptance criteria and data analysis criteria for evaluating relevant conditions in B&W-design Primary or Expansion category components that are subject to VT-3 inspections are to be addressed on a component-specific case-by-case basis by a licensee using WCAP-17096-NP, Revision 3.
3.4.12 Westinghouse Design Control Rod Guide Tube Support Plates The PWROG established its updated set of acceptance criteria methodology and data requirements for W-design CRGT guide plates (guide cards) in WCAP-17096-NP, Revision 3 Appendix E, W-ID Item 1. The NRC staff found that the acceptance criteria methodology and data requirements for CRGT guide cards in W-ID Item 1 to be acceptable for implementation because they were consistent with those established for the CRGT guide cards in Revision 2 of the TR, with one exception that needed further explanations from the PWROG. Specifically, the NRC staff observed that, in W-ID Item 1 assessment criteria bases, the PWROG included the two following data analysis criteria for the CRGT guide card evaluations:
Guide card innermost hole ligament wear depth (or remaining ligament thickness) or slot opening width if ligament thickness is greater than 100% worn away Continuous guidance member wear if ligament at first guide card above the continuous is projected to wear-through before the time of the next inspection The NRC staff observed that the data analysis criteria did not include the following additional (third) data analysis criterion for W-design CRGT guide cards that was included in the corresponding Data Requirements subsection of the applicable ID item:
Guide tube operational effective full power years (EFPYs) at the time of the inspection or measurement.
The NRC staff addressed this information gap with the issuance of RAI-14 (see Reference 23) which requested that the PWROG reconcile the differences in the data analysis criteria given on the Data Requirements section of W-ID Item 1 in Appendix E of the subject TR, from the corresponding data analysis criteria for the CRGT guide card components in Appendix E of WCAP-17096-NP, Revision 2.
In the PWROGs response to RAI-14 (see Reference 25), it stated that the inclusion of W-ID Item 1 data analysis criterion on innermost hole ligament wear was inadvertently duplicated in the Data Requirements section of the ID item, and that the Data Requirements section should have included the bulleted item on guide tube operational effect full power years. The PWROG amended W-ID
Item 1 in the subject TR to include the applicable statement from WCAP-17096-NP, Revision 3, cited above. Based on this amendment of the subject TR, the NRC staff finds the acceptance criteria methodology and data requirements in W-ID Item 1 of WCAP-17096-NP, Revision 3, Appendix E, to be acceptable and RAI-14 is resolved.
3.4.13 Relevant Operating Experience and Industry Interim Guidance Methodology Not Addressed in Topical Report WCAP-17096, Revision 3 The NRC staff observed that based on EPRI Letter No. MRP 2019-023 dated September 3, 2019 (Reference 40), EPRI established Interim Guidance for the inspections of middle axial welds (MAWs) and lower axial welds (LAWs) located in the core barrels of W-design PWRs or core support barrels of CE-design PWRs. The MRP 2019-023 letter included and docketed EPRIs formal interim guidelines for inspecting the MAWs and LAWs in EPRI Letter No. MRP 2019-009 (Reference 40). Attachment 1 to MRP 2019-009 provided EPRIs rationale for creating Needed interim inspection guidelines for these MAWs and LAWs.
The NRC staff observed that the interim guidelines in EPRI Letter 2019-009 did not formally include or establish any applicable acceptance criteria for the one-time VT-3 visual examinations that would be applied to the MAWs and LAWs using EPRIs interim guidance criteria for the components. The NRC staff also observed that the information in WCAP-17096-NP, Revision 3, Section 2.3, and in Appendix C, CE-ID Items 6.1 and 6.2, and WCAP-17096-NP, Revision 3, Appendix E, W-ID Items 4.2 and 4.3, do not reference the existence of the EPRI MRP 2019-009 and MRP 2019-023 letters or provide any acceptance criteria for the one-time VT-3 visual examination that is recommended for the MAWs and LAWs in MRP 2019-009. The NRC staff addressed these informational gaps with the issuance of RAI-06, Parts (a) and (b) (Reference 23).
In the PWROGs complete response to RAI-06, Parts (a) and (b) (see Reference 25), it clarified that the interim guidelines in MRP-2019-009 that require a one-time VT-3 of the core support barrel/core barrel MAWs and LAWs are only cited as NEI-03-08 Good Practice protocols and are not issued as NEI 03-08-established Needed or Mandatory criterion under the umbrella of the Needed criterion of the I&E tables in Sections 4 and 5 of the MRP-227, Revision 1-A. Based on these clarifications, the NRC staff finds that the PWROG has provided a sufficient basis regarding why the subject TR does not include any acceptance criteria methodology and data requirements in WCAP-17096-NP, Revision 3, Appendix C, CE-ID Items 6.1 and 6.2, or in Appendix E, W-ID Items 4.2 and 4.3, for the one-time VT-3 examinations that may be applied as Good Practices for core support barrel/core barrel MAWs and LAWs. The items discussed in RAI-06, Parts (a) and (b) are resolved.
4.0 REFERENCING TOPICAL REPORT WCAP-17096, REVISION 3 IN THE CURRENT LICENSING BASIS WCAP-17096-NP, Revision 3, and its relationship to the CLB is discussed in Section 3.2 of this SE. In addition, NRC staff has identified Limitation L-01 (WCAP-17096-NP, Revision 3) as discussed in this SE and presented in detail in Attachment A. NRC licensees applying the subject TR to their licensing basis or design basis should follow Limitation L-01 (WCAP-17096-NP, Revision 3) in alignment with their specific type of operating license that is currently in effect for the CLB. Section 3.2.1 of this SE provides information for holders of original operating licenses. Similarly, Section 3.2.2 of this SE gives details for holders of first renewed operating licenses, and Section 3.2.3 of this SE explains the process for holders of subsequent renewed operating licenses.
5.0 LIMITATIONS AND CONDITIONS As a result of its review of the subject TR and supplemental information, the NRC staff has identified Limitations and Conditions in Section 3.0 of this SE which are associated with the implementation of this TR. Evaluations for certain components will require plant specific analyses or generic analyses that will need to be submitted to the NRC staff in the event that the inspection results do not meet the acceptance criteria or trigger the expansion criteria methodology in MRP-227. As stated in Section 3.0, limitations and conditions identify any additional plant-specific action items that will be needed to support the NRCs staffs review of a request to implement the TR. Limitations and conditions identify the items that will need to be addressed in subsequent individual plant applications which plan to reference the subject TR. Based on its review of the subject TR and supplemental information, the NRC staff identified four Limitations and three Conditions in Section 3.0 of this SE regarding the implementation of the TR.
A complete detailed overview of all limitations and conditions are available in Attachment A to this SE, Table A-1, Summary of Limitations and Conditions, for Topical Report WCAP-17096, Revision 3.
6.0
SUMMARY
AND CONCLUSIONS The NRC staff has reviewed WCAP-17096-NP, Revision 3, and concludes that the report, as modified by the limitations and conditions in Table A-1 of Attachment A, describes acceptable methods for assessing aging degradation that may be found in RVI components. For NRC licensees conducting inspections in accordance with MRP-227, Revision 1-A, and using the appropriate evaluation procedures, the NRC staff concludes that WCAP-17096-NP, Revision 3, is acceptable for referencing in plant-specific RVI AMPs, as subject to the limitations and conditions in this SE.
7.0 REFERENCES
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- 3. Powell, Michael, PWROG letter to U.S. Nuclear Regulatory Commission, "Transmittal of the Response to Request for Additional Information RAIs Associated with WCAP-17096, Revision 3, "Reactor Internals Acceptance Criteria Methodology and Data Requirements, (PA-MSC-1567)," January 14, 2022 (ADAMS Package Accession No. ML22187A149).
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July 2019 (ADAMS Package Accession No. ML19218A179).
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- 8. U. S. Nuclear Regulatory Commission (NRC) letter to 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 (PWR) Internals Inspection and Evaluation Guidelines, December 16, 2011 (ADAMS Accession No. ML11308A770).
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- 11. Electric Power Research Institute (EPRI) to Materials Reliability Program, "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A) EPRI, Palo Alto, CA 2011; 1022863 Transmittal of Interim Guidance," February 18, 2014 (ADAMS Accession No. ML14274A372).
- 12. U. S. Nuclear Regulatory Commission (NRC) letter to EPRI, "Final Safety Evaluation of WCAP-17096-NP, Revision 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements," May 3, 2016 (ADAMS Accession No. ML16061A243).
- 13. Westinghouse Electric Company, LLC, "WCAP-17096-NP-A, Revision 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements," August 2016 (ADAMS Accession No. ML16279A320).
- 14. U. S. Nuclear Regulatory Commission (NRC) letter to Electric Power Research Institute (EPRI), "U. S. Nuclear Regulatory Commission Approval Letter for the Electric Power Research Institute Topical Report for WCAP-17096-NP-A, Revision 2, Reactor Internals Acceptance Criteria Methodology and Data Requirements," January 3, 2017 (ADAMS Accession No. ML16271A001).
- 15. U. S. Nuclear Regulatory Commission (NRC), "NUREG- 2191, Vol 1, Generic Aging Lessons Learned for Subsequent License Renewal (GAL-SLR Report) Final Report," February 7, 2017 (ADAMS Accession No. ML16274A389).
- 16. U. S. Nuclear Regulatory Commission (NRC), "NUREG- 2191, Vol 2, Generic Aging Lessons Learned for Subsequent License Renewal (GAL-SLR Report) Final Report," February 7, 2017 (ADAMS Accession No. ML16274A399).
- 17. U. S. Nuclear Regulatory Commission (NRC) letter to EPRI, "Final Safety Evaluation for Electric Power Research Institute Topical Report MRP-227, Revision 1, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guideline," April 25, 2019 (ADAMS Accession No. ML19081A001).
- 18. Electric Power Research Institute (EPRI), "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A) Technical Report," June 2020 (ADAMS Accession No. ML20175A112).
- 19. U. S. Nuclear Regulatory Commission (NRC), "SLR-ISG-2021-01-PWRVI - Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors
- Interim Staff Guidance," January 2021 (ADAMS Accession No. ML20217L203).
- 20. U. S. Nuclear Regulatory Commission (NRC), "Federal Register Notice (FRN) - Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized-Water Reactors," January 13, 2021 (ADAMS Accession No. ML20217L212).
- 21. Nuclear Energy Institute, "NEI 03-08, Rev. 3 - Guideline for the Management of Materials Issues," February 2017 (ADAMS Accession No. ML19079A256).
- 22. Nuclear Energy Institute, "NEI 03-08, Rev. 4 - Guidelines for the Management of Materials Issues," October 2020 (ADAMS Accession No. ML20315A536).
- 23. Fields, Leslie, U.S. Nuclear Regulatory Commission, Email to Chad M. Holderbaum, PWR Owners Group, "PWROG-17096 Request for Additional Information," September 21, 2020 (ADAMS Accession No. ML20268B123).
- 24. Pressurized Water Reactor Owners Group (PWROG) letter to U. S. Nuclear Regulatory Commission, "Transmittal of the Response to Request for Additional Information. RAIs Associated with WCAP-17096, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements," January 14, 2022 (ADAMS Accession No. ML22039A143).
- 25. Westinghouse Electric Company, LLC, Attachment 1 to January 14, 2022, letter to PWR Owners Group, "Responses to NRC Request for Additional Information (RAIs) on WCAP-17096-NP, Revision 3," RAIs 01, 04, 05, 06, 07, 08, 10,11 and 14, January 13, 2022 (ADAMS Accession No. ML22039A144).
- 26. Framatome, "Responses to NRC WCAP-17096, Revision 3 (RAIs) Technical Report," RAIs 02, 03, 09, 12 and 13(b), January 13, 2022 (ADAMS Accession No. ML22039A147).
- 27. Electric Power Research Institute (EPRI), "Models of Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steel in Light-Water Reactor Environments," EPRI, Palo Alto, CA, 2014.
- 28. American Society of Mechanical Engineers, "ASME Code Case N-889 - Reference Stress Corrosion Crack Growth Rate Curves for Irradiated Austenitic Stainless Steel in Light-Water Reactor Environments."
- 29. U. S. Nuclear Regulatory Commission (NRC), "Regulatory Guide (RG) 1.147, Revision 20 -
Inservice Inspection Code Case Acceptability ASME Section XI, Div. 1," December 17, 2021 (ADAMS Accession No. ML21181A222).
- 30. American Society of Mechanical Engineers,Section XI, Division 1, "ASME Boiler and Pressure Vessel Code, Rules for Inservice Inspection of Nuclear Power Plant Components," New York, NY.
- 31. Electric Power Research Institute (EPRI) Letter No. MRP 2016-021, "Transmittal of NEI 03-08 "Needed" Interim Guidance Regarding Baffle Former Bolt Inspections for Tier 1 Plants as Defined in Westinghouse NSAL 16-01," July 27, 2016 (ADAMS Accession No. ML16210A006).
- 32. Electrical Power Research Institute (EPRI) Letter No. MRP 2017-009, "Transmittal of NEI 03-08 "Needed" Interim Guidance Regarding Baffle Former Bolt Inspections for Tier 1 Plants as Defined in Westinghouse NSAL 16-01, Rev. 1," March 15, 2017 (ADAMS Accession No. ML17087A106).
- 33. PWR Owners Group Letter OG-18-226 Transmittal of PWROG-18034-P and PWROG-18034-NP Revision 0, Updates to the Methodology in WCAP-15029-P-A, Revision 1, 'Westinghouse Methodology for Evaluating the Acceptability of Baffle-Former-Barrel Bolting Distributions Under Faulted Load Conditions,' PA-MSC-1519," October 31, 2018 (ADAMS Package Accession No. ML18306A583). REFERENCE NOTE: Errata to the WCAP-15029-P-A, Revision 1 and WCAP-15030-NP-A, Revision 0 are given in ADAMS Package Accession No. ML18306A789.
- 34. PWR Owners Group letter to U.S. Nuclear Regulatory Commission, "Submittal of PWROG-17071-NP, Revision 0, WCAP-17096-NP-A Interim Guidance to the NRC for Information Only (PA-MSC-1567)," July 12, 2018 (ADAMS Accession Nos. ML18204A165 and ML18204A166).
- 35. Electric Power Research Institute (EPRI) Report No. 3002010270, "Materials Reliability Program: PWR Internals Age-Related Material Properties, Degradation Mechanisms, Models and Basis Data State of Knowledge (MRP-211, Revision 1)," October 2017 (ADAMS Package Accession Nos. ML17361A168/ML17361A191 - Proprietary/Non-Public); and ML17361A189, - Non-Proprietary/Public).
- 36. NRC Agency Record providing Office of Nuclear Reactor Regulation Staff Assessment of Electric Power Research Institute NEI 03-08, Revision 2, NEEDED Interim Guidance Regarding Baffle-Former Bolt Inspections in Westinghouse Design Pressurized Water Reactors, November 20, 2017 (ADAMS Accession No. ML17310A861).
- 37. Westinghouse Electric Company Proprietary Class 2, "Westinghouse Baffle-Former Bolt Predictive Methodology," September 18, 2019 (ADAMS Accession Nos. ML19242B369 and ML19274F482 - Proprietary/Non-Public; ML19274F472 -Non-Proprietary/Public).
- 38. U. S. Nuclear Regulatory Commission, letter to Westinghouse Owners Group Steering Committee, "Safety Evaluation of Topical Report WCAP-15029, Westinghouse Methodology
for Evaluating the Acceptability of Baffle-Former-Barrel /bolting Distributions Under Faulted Load Conditions," November 10, 1998 (ADAMS Accession No. ML20195C636).
- 39. Westinghouse Electric Company, LLC, Attachment 1 to January 14, 2022, letter to PWR Owners Group, "PWR Owners Group Responses to NRC Request for Additional Information (RAIs) on WCAP-17096-NP., Rev. 3," RAI 13, January 13, 2022 (ADAMS Accession No. ML22039A146).
- 40. Electric Power Research Institute (EPRI), "Transmittal of MRP-227-A Related Interim Inspection Guidance Regarding PWR Core Barrel," MRP-2019-023 (BWRVIP 2019-009),
September 3, 2019 (ADAMS Accession No. ML19249B102).
- 41. Electric Power Research Institute (EPRI), "2017 Technical Report - BWRVIP-100, Revision 1-A: BWR Vessel and Internals Project, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds," February 2017 (ADAMS Accession No. ML17076A232 - Proprietary/Non-Public).
- 42. Electric Power Research Institute (EPRI), "BWRVIP Letter No. 2021-030 - Potential Non-Conservatism in EPRI Report, BWRVIP-100, Revision 1-A, 30020008388 and Impacted BWRVIP Reports," March 22, 2021 (ADAMS Accession No. ML21084A164).
- 43. Electric Power Research Institute (EPRI), "Transmittal of MRP-227-A-Related Interim Inspection Guidance Regarding PWR Core Barrel," September 3, 2019 (ADAMS Accession No. ML19249B102).
- 44. Pressurized Water Reactor Owners Group (PWROG), "Transmittal of the Response to Request for Additional Information (RAIs) Associated with WCAP-17096, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements (PA-MSC-1567)," January 14, 2022 (ADAMS Accession No. ML22018A068).
- 45. Entergy Nuclear Operations, Inc., Palisades Nuclear Plant, "Unplanned Shutdown Due to Elevated Primary Coolant System Leak Rate and Permanent Plant Shutdown," May 23, 2022 (ADAMS Accession No. ML22143A937).
- 46. Electric Power Research Institute (EPRI) letter to U. S. Nuclear Regulatory Commission (NRC), "Responses to the Questions from the U. S. Nuclear Regulatory Commission Staff on the Baffle-Former Bolt "Needed" Guidance Transmitted in Letter MRP 2017-009," July 13, 2017 (ADAMS Accession No. ML17261B149).
- 47. Electric Power Research Institute (EPRI) Letter No. BWRVIP 2019-016, White Paper on Suggested Content for PFM Submittals to the NRC, February 27, 2019 (ADAMS Accession No. ML19241A545).
Attachments: A. Summary of Limitations and Conditions B. Comment Resolution Table Principal Contributors: James Medoff David Dijamco Chris Sydnor Leslie Fields Luke Haeg Siva Lingam Date: July 24, 2023
Overview of PWROG RAI Response Transmittals Document Title / Description Document Date Agency Record Reference No.
PWR Owners Group Transmittal of the January 14, 2022 ML22018A068 Reference 44 Response to Request for Additional Information, RAIs Associated with WCAP-17096, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements (PA-MSC-1567)
PWR Owners Group Transmittal Letter January 14, 2022 ML22039A143 Reference 24 in Response of Request for Additional Information (RAIs) on WCAP-17096-NP, Revision 3 PWR Owners Group - Response to January 13, 2022 ML22039A144 Reference 25 NRC Requests for Additional Information (RAIs) on WCAP-17096-NP, Revision 3: (RAI-01; RAI-04; RAI-05; RAI-06; RAI-07 RAI-08; RAI-10; RAI-11 and RAI-14)
Westinghouse Response to RAI-13 on January 13, 2022 ML22039A146 Reference 39 WCAP-17096-NP, Revision 3 Framatome Responses to NRC January 13, 2022 ML22039A147 Reference 26 WCAP-17096, Revision 3 (RAIs)
Technical Report (RAI-02; RAI-03; RAI-09; RAI-12; and RAI-13 (b)
- Please Note: On February 8, 2022, the PWROG resubmitted the January 13, 2022, RAI responses to NRC and the transmittal was added to the ADAMS docket. The February 8, 2022, transmittal superseded the January 14, 2022, transmittal, however the RAI document date of January 13, 2022, remains unchanged.
Safety Evaluation, Attachment A Table A-1, Summary of Limitations and Conditions Attachment A
NRC SAFETY EVALUATION FOR TOPICAL REPORT WCAP-17096-NP, REVISION 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements TABLE A-1:
SUMMARY
OF LIMITATIONS AND CONDITIONS Rev. 3 Topical Report Numerical Identifier for Limitations and Conditions Limitations and Conditions (TR) & Safety Eval. (SE)
Sections Number Topic / RAI Technical Basis TR SE Limitation Applicability of The generic applicability of Limitation 1 (L-01)(a): Revision 3 3.2, L-01(a)(b)(c) Revision 3 of Rev. 3 of the TR to the Chapter 1 3.2.1, the TR to the licensing basis for a given SE Section 3.2.1: Licensing Considerations- Comanche Peak, Diablo Canyon, and Watts Bar 3.2.2, current PWR unit is dependent on For PWR plants with an original operating license in a pre-submittal phase for first-time license renewal or currently &
licensing basis the following factors in the undergoing an NRC review of a first-time license renewal application, NRC licensees may use the subject TR if they are 3.2.3 CLB CLB for the unit: implementing AMPs in accordance with MRP-227, Revision 1-A.
(1) the type of plant operating L-01(b):
license (i.e., original, first renewed, or subsequent SE Section 3.2.2: Holders of First Renewed Operating Licenses renewed license) held by the For PWR plants with first renewed operating licenses, NRC licensees may use the subject TR if they are implementing NRC-licensed facility which AMPs in accordance with MRP-227, Revision 1 -A.
owns the specified PWR unit, L-01(c):
(2) the type of aging management program (i.e., SE Section 3.2.3: Holders of Subsequent Renewed Operating Licenses plant-specific AMP versus For PWR plants with subsequent-renewed operating licenses, NRC licensees may use the subject TR only for those RVI generic MRP-227-based components whose I&E criteria for the components are still consistent with MRP-227, Revision 1-A, and have not been AMP) being applied to the modified by the results of an 80-year RVI gap analysis or a version of MRP-227 covering an 80-year aging assessment RVI components in the unit, period.
(3) the type of generic EPRI MRP-based program being applied as the basis for the AMP in the CLB, and (4) the version of MRP-227 being applied to the program.
A-2
NRC SAFETY EVALUATION FOR TOPICAL REPORT WCAP-17096-NP, REVISION 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements TABLE A-1:
SUMMARY
OF LIMITATIONS AND CONDITIONS Rev. 3 Topical Report Numerical Identifier for Limitations and Conditions Limitations and Conditions (TR) & Safety Eval. (SE)
Sections Number Topic / RAI Technical Basis TR SE Limitation In Revision 3 of In the PWROGs response to Limitation 2 (L-02): Revision 3 3.4.7 L-02 the TR, the RAI-11 (Reference 6 of this Section 3.5 PWROGs Final SE), which places the The scope of WCAP-17096-NP, Revision 3, is limited in that it does not include any acceptance and data analysis criteria basis for responsibility for evaluating for evaluating component distortion or void swelling, or for managing changes in dimension due to distortion or void swelling evaluating VT-3 component void swelling or in the PWR Primary or Expansion category components that are within the scope of WCAP-17096-NP, Revision 3. If visual distortion on the NRC distortion or changes in dimension due to void swelling is detected in an RVI component, a licensee using this TR will need inspection licensee implementing its to address distortion or changes in dimension due to void swelling on a component-specific basis per the acceptance results that PWR Vessel Internals criteria and corrective actions program element bases of its PWR Vessel Internals Programs and its 10 CFR Part 50, provide Program (or an equivalently Appendix B program.
evidence of named RVI AMP for the void swelling or CLB).
distortion in components.
/
RAI-11
A-3
NRC SAFETY EVALUATION FOR TOPICAL REPORT WCAP-17096-NP, REVISION 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements TABLE A-1:
SUMMARY
OF LIMITATIONS AND CONDITIONS Rev. 3 Topical Report Numerical Identifier for Limitations and Conditions Limitations and Conditions (TR) & Safety Eval. (SE)
Sections Number Topic / RAI Technical Basis TR SE Limitation Fatigue In the PWROGs response to Limitation 3 (L-03): Rev.3 Section 3.4.8 L-03 screening RAI-08 (Reference 24, this 3.6 requests for SE), NRC staff observed that The scope of the WCAP-17096-NP, Revision 3, is limited in that it no longer includes or defines the acceptance criteria or CE-design core the PWROGs fatigue data analysis criteria for performing the fatigue and SCC screening analyses of CE-design core barrel flexure welds or Rev. 3 support plates screening requests for CSB fatigue screening analyses of CE-design core support plates and upper fuel alignment plates. Consistent with the Appendix C, in lower flexure welds is outside the guidelines in MRP-227, Revision 1-A, a licensee using the subject TR should address the acceptance criteria and data CE-ID Item 7, support scope of the WCAP-17096- analysis criteria for performing the component-specific screening analyses of CE-design core barrel flexure welds, core structures and NP, Revision 3. Therefore, support plates, and upper fuel alignment plates on a unit-specific or site-specific case-by-case basis. Rev. 3 fueling NRC staff has identified a Appendix C, alignment limitation on the RAI-08 CE-ID Item 8 plates in upper fatigue and SCC screening internals topics when licensees Rev. 3 assembly, and reference CSB flexure welds, Appendix C, fatigue and core support plate and fuel CE-ID Item 9 stress alignment plates components corrosion in WCAP-17096-NP, cracking (SCC) Revision 3 (as assessed in screening CE-ID Items 7, 8, and 9) or requests for defining SCC screening CE-design core bases for the CSB flexures support barrel welds (CE-ID Item 7).
(CSB) flexure welds.
/
RAI-08, Parts (a), (b), and (c)
A-4
NRC SAFETY EVALUATION FOR TOPICAL REPORT WCAP-17096-NP, REVISION 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements TABLE A-1:
SUMMARY
OF LIMITATIONS AND CONDITIONS Rev. 3 Topical Report Numerical Identifier for Limitations and Conditions Limitations and Conditions (TR) & Safety Eval. (SE)
Sections Number Topic / RAI Technical Basis TR SE Limitation In Revision 3 of In the PWROGs response to Limitation 4 (L-04): Rev. 3 of the 3.4.11 L-04 the TR, the RAI-12, the referenced B&W- TR Appendix A, PWROGs ID items include general The scope of the WCAP-17096-NP, Revision 3, is limited in that it does not define any acceptance criteria or data analysis bases for provisions for identification of: criteria for evaluating relevant conditions in B&W-design RVI components that are inspected using VT-3 visual inspection B&W ID-Items defining (1) what visual examination methods. Consistent with the guidelines in MRP-227, Revision 1-A, the acceptance criteria and data analysis criteria for A.1.1, A1.2, relevant wear indications are evaluating relevant conditions in B&W-design Primary or Expansion category components that are subject to VT-3 A.1.3, A.1.6, conditions for considered rejectable and inspections are to be addressed on a component-specific case-by-case basis by a licensee using WCAP-17096-NP, A.1.7, A.1.9, B&W-design would require additional Revision 3. A.1.10, A.1.11, Primary examination and evaluation, A.1.12, A.1.14, category and and (2) any additional A.2.1, A.2.3, Expansion examination results that are A.2.4, A.2.10, category RVI anticipated and the general A.2.12 components acceptance criteria for these that are within additional items are expected the scope of to be in the VT-3 visual EPRI MRP VT- examination field of vision.
3 visual However, the PWROG has examinations confirmed that these methods. statements do not fully define
/ the acceptance and data RAI-12 analysis criteria for the components in these B&W-ID item assessments.
A-5
NRC SAFETY EVALUATION FOR TOPICAL REPORT WCAP-17096-NP, REVISION 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements TABLE A-1:
SUMMARY
OF LIMITATIONS AND CONDITIONS Topical Report (TR) &
Numerical Identifier for Limitations and Conditions Limitations and Conditions Safety Eval. (SE)
Sections Number Topic / RAI Technical Basis TR SE Condition Selection and The NRC staffs condition Condition 1 (C-01): Rev. 3 of the 3.3.4 and C-01 justification of reflects the PWROGs TR Section 3.3.7 CGRs for conditional statements in its For stainless steel metal and weld components in the specified neutron fluence exposure ranges as shown in Table 2 in 2.2 stainless steel response to RAI-04 for the Section 3.3.4 of this SE, the PWROG has indicated that the CGRs are based on the CGR criteria fluence ranges specified in metal and subject TR regarding the ASME Code Case N-889, as established by the EPRI models in the EPRI Report cited as Reference 46 in WCAP-17096-NP, Appendix A, weld use of bounding CGRs for Revision 3. The NRC has endorsed Code Case N-889 in Regulatory Guide 1.147, Revision 20, with conditions for the specified B&W-ID components. stainless steel components neutron fluence exposure ranges. Therefore, if the CGR models in Reference 46 of WCAP-17096-NP, Revision 3, are followed Items A.1.9 and welds (Reference 3). by a licensee using WCAP-17096-NP, Revision 3, then the CGR models are subject to the NRC conditions on Code Case N-889, and A.2.11 NRC staffs as defined in Table 2 of Regulatory Guide 1.147, Revision 20.
C-01 RAI-04 applied to 16 high Appendix C, corresponds fluence components. CE-ID Items with the However, in the response 2, 2.1, 3, PWROGs to RAI-04, the PWROG 3.1, 5, 5.1, conditional established bounding 5.2, 5.3 5.4, statements in CGRs for stainless steel 6, 6.1, 6.2, Revision 3 of welds with neutron fluence 7, and 11 the TR exposures shown in (regarding the Table 2 of this Final SE. Appendix E, use of Therefore, Condition 1 is W-ID Items bounding applicable to all 23 3, 3.1, 3.2, CGRs), with components (i.e., 2 B&W, 3.3, 4, 4.2 additional 14 CE, and 7 and 4.3 considerations Westinghouse) that use the when the flaw tolerance evaluation licensee method.
decides to use an alternate The CGRs were based on CGR basis the CGR models in the from that EPRI report cited as defined by the Reference 46 in the subject PWROG. TR, which were also used to develop bounding CGRs in ASME Code Case N-889.
A-6
NRC SAFETY EVALUATION FOR TOPICAL REPORT WCAP-17096-NP, REVISION 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements TABLE A-1:
SUMMARY
OF LIMITATIONS AND CONDITIONS Topical Report (TR) &
Numerical Identifier for Limitations and Conditions Limitations and Conditions Safety Eval. (SE)
Sections Number Topic / RAI Technical Basis TR SE Condition Use of In the NRC staffs Condition 2 (C-02): WCAP- 3.4.5 and C-02 predictive resolution of RAI-13, Parts 17096-NP, 3.4.6 bolting pattern (a), and (b), in this SE, the For bolted assembly analysis methods that use predictive, probabilistic ABPA models, the models and methods should Revision 3, model or staff established its SE appropriately account for limits on the number of allowable bolts with presumed failed conditions, geometric bolt failure Section 2.3 methodology evaluation position that considerations (i.e., bolt clustering considerations), and bolting analysis reliability considerations (i.e., addressing 95% reliability bases for predictive bolting pattern confidence limits). These models and methods should be appropriately documented and justified in a site-specific or owner- WCAP-bolts in bolted model or methods may be defined record. 17096-NP, assemblies. used to establish the Revision 3,
/ reinspection interval for a Section 3.3 RAI-07, specified bolted assembly Part (c), and type or even as a basis for WCAP-RAI-13, Parts justifying further service of 17096-NP, (a), and (b) a bolted assembly when Revision 3, evidence of degradation in Appendices a number of bolts in the A, C, & E:
assembly and the Applicable methodology can support a B&W-ID, finding that there would be CE-ID and a sufficient number of bolts, W-ID items as based on the ability of for bolted the model to generate a assemblies predictive set of future in the bolting pattern outputs subject TR starting from an initial appendix degraded bolting pattern that rely or (as established from the may rely on last inspections performed acceptable on the assembly). bolting pattern bases for structural integrity of the assembly.
A-7
NRC SAFETY EVALUATION FOR TOPICAL REPORT WCAP-17096-NP, REVISION 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements TABLE A-1:
SUMMARY
OF LIMITATIONS AND CONDITIONS Topical Report (TR) &
Numerical Identifier for Limitations and Conditions Limitations and Conditions Safety Eval. (SE)
Sections Number Topic / RAI Technical Basis TR SE Condition Accuracy and As has been discussed in Condition 3 (C-03): WCAP- 3.4.10 C-03 validation of Section 3.4.10 of this SE, 17096-NP, lower bound some neutron fluence Until such time that the EPRI or PWROG can resolve the EPRI BWRVIP Letter No. 2021-020 Part 21 issue on neutron fluence Revision 3, LEFM fracture ranges reported for ranges associated with specified lower bound fracture toughness values reported in an updated version of MRP-210 or MRP-211 Appendix C, values specified lower bound (or alternate industry record), licensees will need to justify any lower bound fracture toughness values used as inputs in CE-ID Items reported in fracture toughness values component-specific flaw tolerance analyses. The selection of the applicable lower bound fracture toughness value(s) should be 2, 2.1, 3, Revision 3 of cited in Rev. 3 of the TR reported and justified in a site-specific record and reconciled with any updated industry reports or records used to correlate lower 3.1, 5, 5.1, TR and the may no longer be valid per bound fracture toughness value(s) and the most recent neutron fluence range(s) reported for the specified fracture toughness 5.2, 5.3, 5.4, validity and EPRIs Part 21 notification value(s). 6, 6.1, 6.2, accuracy of on the lower bound fracture 7, & 11, neutron toughness values, as ranges reported in EPRI BWRVIP and reported for Letter No.
the specified 2021-030 (Reference 42). WCAP-fracture Based on the information in 17096-NP, toughness EPRI BWRVIP Letter No. Revision 3, values. 2021-020, there may be Appendix E,
/ some uncertainty with the W-ID Items RAI-10 accuracy of the fluence 3, 3.1, 3.2, ranges reported for 3.3, 4, 4.2, &
specified lower bound 4.3 fracture toughness values that were reported in the fracture toughness tables of Appendices C and E in Revision 3 of the TR.
A-8
Safety Evaluation, Attachment B Resolution of Comments for the U.S. Nuclear Regulatory Commissions Draft Safety Evaluation of Pressurized Water Reactors Owners Group Topical Report No. WCAP-17096, Revision 3 On September 30, 2022, the staff submitted a copy of the draft safety evaluation (DSE) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22181B009) for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) No. WCAP-17096, Revision 3 , for commentary by members of the PWROG (refer to the staffs transmittal email at ADAMS Accession No. ML22266A280).
The staff received a total of 73 comments regarding the DSE for the subject TR, including: (1) a total of four comments from Framatome as the PWROGs participating Nuclear Steam Supply System (NSSS) vendor for Babcock and Wilcox (B&W)-designed pressurized water reactors (PWRs), (2) a total of 68 comments from Westinghouse Electric Company as the PWROGs participating NSSS vendor for Combustion Engineering (CE)-designed and Westinghouse-designed PWRs, and (3) an additional email comment from the PWROG (Mr. J. Andrachek) on the appropriateness of citing the > 6.7x1021 n/cm2 (E> 1.0 MeV) fluence exposure value and the > than 10 dpa projected dose exposure value as proprietary contents in the DSE for the subject TR. The comments received from Westinghouse Electric Company and Framatome Corporation are in ADAMS package Accession No. ML23009A588.
The comment received via email from the PWROG is located at ADAMS Accession No. ML23173A090 (Non-Public).
The comments and the staffs basis for addressing and resolving comments received on the DSE for the subject TR are provided in the table that follows.
Note: The first four comments were from Framatome. The remaining comments were from Westinghouse except for the final comment via email from the PWROG.
Attachment B
Resolution of Comments Received on the DSE for WCAP-17096, Revision 3 Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1)
Comments Received from Framatome Corporation 1 38 - 40 33 on Accuracy and Basis Comment: The staff did not find Framatome Page 38 Comment 1 to be valid or the
- 1 on The primary purpose of the TR is to provide recommended action to be acceptable Page 40 evaluation methodologies to address relevant for implementation. Specifically, the conditions. However, per Section 5.2 of the TR, staffs placement of Limitation L-04 in for the B&W units, the TR also indicates actions the DSE for WCAP-17096, Revision 3 that might be used to define nondestructive applies to those B&W-design Primary examination (NDE) acceptance standards and and Expansion category components actions that could support analytical that are inspected by VT-3 visual evaluations. inspection methods for conditions induced by aging effects in the The NRC states the TR does not define components. The VT-3 visual methods acceptance criteria or data analysis criteria for look for gross conditions occurring in evaluating relevant conditions for the RV the components (e.g., gross indications internals components inspected using VT-3 of wear, loose parts, missing or severed visual inspection methods. It is unclear why the parts, or distortion).
NRC is stating this. Each component item has a stated analytical methodology to be used for the In its response to RAI 12 (ADAMS acceptance criteria. For example, the Accession No. ML22018A071),
methodology for evaluation for the CRGT Framatome only made the following spacer castings (A.1.2 in the TR) is provided on general statements regarding the pages A-5 and A-6 (i.e., drop time evaluation or acceptance standards for VT-3 reactivity analysis). detected relevant conditions and analytical efforts used to disposition the Directly following this is the statement, The relevant conditions:
general methodology to be used for acceptance criteria for these component items will be The general methodology to be used for development of an NDE inspection standard VT-3 (a general condition monitoring visual that contains examples of acceptable and examination) acceptance criteria for these component items will be development of an unacceptable visual indications and mockups for NDE inspection standard that contains the VT-3 inspection of fractured spacers or explanations of acceptable and missing screws. This is unrelated to the unacceptable visual indications . . .
evaluation methodology. This statement is Analytical efforts may be performed on a addressing NDE acceptance standards (per unit-specific basis due to the different Section 5.2 of the TR). interference. The NE inspection standard could be developed generically.
B Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 1 Suggested Revision: Thus, even after the response to RAI 12 (Continued) was received by the staff, the Section 3.4.11 of the SE should be revised to component-specific ID item state that the Staff accepts the response to RAI assessments for the B&W-design
- 12. Consequently, Limitation L-04 should be components subject to VT-3 in removed. Appendix A of the subject TR did not provide sufficient technical details on Staff Observation on Comment 1: defining specific acceptance criteria or data analysis criteria for analyses that The potential relevancy of this comment should might be performed on B&W-design really be applied to all of Final SE components with non-conforming Section 3.4.11. relevant conditions. That is the reason why the response to RAI 12 was not fully accepted by the staff and why Limitation L-04 was created and issued in the draft SE for commentary.
As stated in Limitation L-04, the licensee should ensure that the non-conforming condition detected by the VT-3 visual examination of the component is appropriately entered, dispositioned, and justified in the licensees corrective actions program.
2 29 47 - 50 Clarification Comment: The staff agrees that the Framatome Comment 2 recommendation is valid for The DSE states: The NRC staff finds that for implementation. Specifically, the W-design plants, the specification in Revision 3 comment applies to W-ID Item 6 in of the TR Appendix E, W-ID Item 6 to submit the Appendix E of the subject TR for plant-specific application of the predictive model Westinghouse-design BF bolts and bolt methodology defined in W-ID Item 6 is an locking devices. The updated W-ID Item acceptable way for making the details of the 6 for the Westinghouse-design BF bolts predictive methodology available to the NRC. and bolt locking devices includes a new analysis submittal statement that was Per TR Appendix E, W-ID Item 6, submittal to added to W-ID Item 6 in resolution of the NRC is for information and is only required if Condition Item 3 for Group 3
- 1) MRP 2017-009 indicates that the subsequent components (which included the W-ID inspection interval is not to exceed 6 years Item 6 Westinghouse-design BF bolts, B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 2 and/or 2) there is an evaluation to lengthen the as referenced in Section 2.2 of the (Continued) determined inspection interval or to exceed the subject TR).
maximum inspection interval recommended in MRP 2017-009. As based on Section 2.2 of the subject TR, W-ID Item 6 in Appendix E of the Suggested Revision: subject TR now calls any analysis used to extend the inspection interval of the The following sentence should be added to the Westinghouse-design BF bolts to be paragraph at the bottom of Page 27 of the Final submitted to the staff. Therefore, the DSE. staff finds the comment to be accurate and the staff will implement and will add Per TR Appendix E, W-ID Item 6, submittal to the referenced sentence identified in the NRC is for information and is only required if Framatome Comment 2 to
- 1) MRP 2017-009 indicates that the subsequent Section 3.4.6 of this Final SE for the inspection interval is not to exceed 6 years subject TR with minor editorial changes.
and/or 2) there is an evaluation to lengthen the determined inspection interval or to exceed the maximum inspection interval recommended in MRP 2017-009.
3 30 8 - 12 Editorial Comment: The staff acknowledges the comment.
Upon further review, the staffs This is the sentence that is the subject of this perspective is that the referenced comment: This would apply even if the bolting sentence still provides an important analysis methodology was approved by NRC clarification, which is that the applicable (e.g., the PWROGs referencing of WCAP- analysis should be placed into the 15029-P-A, Revision 1 as an ABPA licensees corrective action program methodology in Revision 3 of the TR Appendix regardless of whether the analysis E, W-ID Items 6 and 6.1 for W-design BFBs and methodology was staff approved. Thus, barrel-to-former bolts) or if Revision 3 of the TR the staff left the referenced sentence in precluded the plant-specific bolted assembly place in Section 3.4.6 of this Final SE and bolting analysis from being performed using for the subject TR.
an unapproved methodology.
Previous sentences on Page 28 state the NRC staff would expect the bolting evaluation methods to be placed into the NRC licensees corrective action program and to be fully defined, discussed, and justified in a site-B Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 3 specific document/record. Therefore, the (Continued) italicized sentence is not needed.
Suggested Revision:
Remove the italicized sentence.
4 A-7 and C-02 Basis Comment: The stated Condition C-02 applies to A-8 the licensee-specific analytical All aspects of this condition are already present corrective action bases for in the TR (A.1.8 and W-ID: 6). It is unclear why dispositioning flaw indications in B&W this condition is necessary. and Westinghouse and CE design bolting other than baffle-former (BF)
Suggested Revision: bolts or core shroud bolting.
Remove Condition C-02. The objective of Condition C-02 is to ensure that the analysis or alternate corrective action basis (e.g.,
component-specific repair or replacement) used to disposition a non-conforming condition in the bolting type or bolted assembly will be entered into, dispositioned, and justified in the licensees correction actions program.
Condition C-02 remains as a Condition listed in Table A-1 of the Final SE for the subject TR.
Comments Received from Westinghouse Electric Company 1 9 37 Accuracy Comment (including suggested revision): The staff agrees that the comment is valid. The staff will make the change The word methods is missing from the quoted (Final SE Section 3.2) to include the statement (i.e., should be NRC approved word methods.
evaluation methods).
2 14 29 - 30 Clarification Comment (including suggested revision): The staff agrees that the comment is valid. The staff will incorporate the The quoted statement beginning on Line 34 is change recommended in Comment 2 from TR R3 (not revised by RAI-04). Therefore, B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 2 the statement, In PWROGs response to RAI- (located within Section 3.3.4 of this (Continued) 04 (See Reference 25), it stated that Section 2.2 Final SE).
of Revision 3 of the TR was revised to include the following additional action should be changed to Section 2.2 of Revision of the TR includes the following additional action 3 15 5-7 Editorial Comment (including suggested revision): The staff agrees that the comment is valid. The staff adjusted the part of the Delete extraneous figure title, Figure 2 - sentence in this Final SE in accordance Neutron Fluence Values. with the editorial comment and changed the term Figure to Table.
4 15 13 Editorial Comment (including suggested revision): The staff agrees that the comment is valid. The staff capitalized the first Capitalize Code Case. letters of the words Code Case in this Final SE.
5 15 Footnote Editorial Comment (including suggested revision): The staff agrees that the comment is 3 valid. The staff adjusted Footnote 3 Change cracking growth evaluations to accordingly in this Final SE.
crack growth evaluation.
6 16 26 - 27 Clarification Comment (including suggested revision): The staff agrees that the comment is valid considering the change in the Add footnote to CE-design core support column licensing status of the operating bolts in CE design with bolted core shroud (CS) licenses for Ft. Calhoun and Palisades assemblies to clarify that this component has PWR units to permanently defueled been removed from TR R3 per the response to technical specification requirements.
RAI 7(a). The modified version of the requested footnote on the component-specific bullet for the CE-design core support column bolts (DSE Page 16, Lines 26 -
- 27) was added as new Footnote 4 to this Final SE.
7 19 33 - 34 Clarification Comment (including suggested revision): The staff understands the rationale for the comment; however, for SE bullets Add (CE-ID 2.1). involving specified Primary or Expansion components, the assigned B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 7 NRC project manager made an editorial (Continued) decision for the SE format not to include the applicable component ID item number for the specified component type in the applicable DSE bullet. So instead of the adding (CE-ID Item 2.1) to the applicable bullet for the core shroud remaining axial welds, the staff added the words (CE-design units with welded core shrouds assembled in two vertical sections) to the component-specific bullet item within Section 3.3.8 of this Final SE. This is the welded type of CE-design core shroud that is associated with CE-ID Item 2.1.
8 19 43 - 44 Clarification Comment (including suggested revision): The staff understands the rationale for the comment; however, for bullets Add (CE-ID 3.1). involving specified Primary or Expansion components, the assigned NRC project manager made an editorial decision for the SE format not to include the applicable component ID item number for the specified component type in the applicable bullet. Instead of adding CE-ID Item 3.1 to the applicable bullet for the core shroud remaining axial welds, the staff added the words (CE-design units with welded core shrouds assembled with full height shroud plates) to the component-specific bullet within Section 3.3.8 of the Final SE. This is the welded type of CE-design core shroud that is associated with CE-ID Item 3.1.
9 23 14 Accuracy Comment (including suggested revision): The staff agrees that the comment is valid since the indented sentence involves quoted material from the B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1)
Spell out OE as, operating experience as done applicable referenced document. The in the quoted sentence. staff adjusted the indented, italicized sentence located within Section 3.4.1 of this Final SE.
10 24 40 Clarification Comment (including suggested revision): The staff agrees that the comment is valid. The staff corrected the Correct Article WB-3000 to Article typographical error within Section 3.4.3 IWB-3000. of this Final SE.
11 25 8 Editorial Comment (including suggested revision): The staff agrees that the comment is valid. EPRI does not include Correct document number of MRP-2018-022 hyphenation between MRP and the to MRP 2018-022. listed year in these types of EPRI letter references. Thus, it should be in reference to MRP 2018-022 for accuracy. The staff made the editorial change to within this Final SE.
12 25 14 Clarification Comment (including suggested revision): The staff partially agrees that the comment is valid. Westinghouse is Reference 33 in the DSE is an erratum to correct that Reference 33 in DSE correct a typographical error in WCAP-15029-P- Section 7.0 should actually be A/WCAP-15030-NP-A. Based on the context, referencing the submittal of the WCAP-Reference 33 should be updated to the 15029-P-A/WCAP-15030-NP-A reports; submittal of WCAP-15029-P-A/WCAP-15030- but the staffs perspective is that NP-A to the Staff. Reference 33 should be citing the latest staff-approved version of the Staff Comment and Observation: Proprietary report (WCAP-15029-P-A, Revision 1), which was submitted in The comment associated with Comment 12 is ADAMS Accession Package actually in reference to the manner the No. ML18306A583 and transmitted in Reference 33 record (as referenced on DSE PWROG Letter OG-18-226, dated Page 25, Line 14) is worded in DSE Section 7.0 October 18, 2018.
- see related Westinghouse Comment 42. Per Comment 42, the PWROG indicated that The staff amended Final SE Section Reference 33 should be citing: Westinghouse 7.0, Reference 33 accordingly per Reports, WCAP-15029-P-A, Rev. 0 and WCAP- Comment 12 and the corresponding 15030-NP-A, Rev. 0, Westinghouse comment in Westinghouse Comment B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 12 Methodology for Evaluating the Acceptability of 42. A footnote is included in this Final (Continued) Baffle-Former-Barrel Bolting Distributions under SE for Reference 33 regarding the Faulted Load Conditions, December 1998 / errata for the report.
April 1999 (ADAMS Accession Pkg.
No. ML20206G308).
13 25 18 Clarification Comment (including suggested revision): The staff agrees that the comment is valid. Westinghouse is correct that the Reference 38 is the SE for WCAP-15029 which, statement on Page 25, Line 18, of the based upon the context of the sentence, is not DSE should be referencing a new the correct reference. This sentence should be Section 7.0 reference for the staffs SE updated to refer to a new reference (that is not on the Interim BFB Guidance dated currently included in Section 7.0 of the DSE): November 20, 2017 (ADAMS NRC Letter, Office of Nuclear Reactor Accession No. ML17310A861) and not Regulation Staff Assessment of Electric Power Reference 38. However, the staffs Research Institute NEI 03-08, Revision 2, acceptance of Comment 44 also had NEEDED Interim Guidance Regarding Baffle- some bearing on how the staff Former Bolt Inspections in Westinghouse addressed and incorporated the Final Design Pressurized Water Reactors SE changes recommended in both November 20, 2017 (ADAMS Accession Westinghouse Comments 13 and 44.
No. ML17310A861).
Specifically, Comment 44 requests Staff Comment and Observation: deletion of Section 7.0 Reference 36 due to its redundancy with Section 7.0 Westinghouse Electric Companys comment Reference 23, and Comment 13 associated with Comment 18 is in regard to the recommends creation of a new citation of Reference 38 on DSE Page 25, Line reference for the staffs Agency Record 18 (as linked to the staffs SE for the interim used for acceptance of the industrys baffle-to-former bolts (BFBs) when compared to Interim BFB Guidance. Thus, the staff the referencing of Reference 38 in other addressed both comment portions of DSE Section 3.4.4 which cites the recommendations by amending the staff SE for WCAP-15029-P-A (Revision 0). Final SE reference of Reference 38 to Reference 36, and by amending Reference 36 in this Final SE Section 7.0 to be the new Final SE Section 7.0 reference for the staffs Agency Record assessment on the Interim BFB Guidance. Thus, the Final SE Section 7.0, Reference 36 was not B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 13 deleted, but was amended as follows to (Continued) achieve the same objective of the combined recommended actions in Comments 13 and 44:
NRC Agency Record providing Office of Nuclear Reactor Regulation Staff Assessment of Electric Power Research Institute NEI 03-08, Revision 2, NEEDED Interim Guidance Regarding Baffle-Former Bolt Inspections in Westinghouse Design Pressurized Water Reactors, November 20, 2017 (ADAMS Accession No. ML17310A861).
14 26 24 Accuracy Comment (including suggested revision): The staff agrees that the comment is valid. The staff added the missing The acronym is missing from the quoted acronym to Section 3.4.5 of this Final statement. Westinghouse Owners Group SE.
should be Westinghouse Owners Group
[WOG].
15 26 33 - 34 Clarification Comment (including suggested revision): The staff agrees that Comment 13 is valid but observes that the validity and The PWROG is not aware of a separate the needed resolution of Comment 44 transmittal for RAI-07. The RAIs on TR R3 were requested activity is also relevant to the transmitted in Reference 23 dated needed resolution basis for resolving September 21, 2022. Therefore, the date should Comment 15.
be changed from September 24, 2020 to September 21, 2020, and the citation to The staff amended the date reference (Reference 36) should be changed to in this Final SE to cite it as (Reference 23). See related Comments 21 and September 21, 2020. The hyperlinks
- 44. for the reference section were removed.
16 26 37 - 38 Clarification Comment (including suggested revision): The staff agrees that the comment is valid. The hyperlinks to the reference Reference 37 is the Westinghouse predictive section were removed.
model methodology presentation dated B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 16 September 18, 2019. It seems based on the (Continued) context that this should be a reference to WCAP-15029-P-A. Therefore, (Reference 37) should be (Reference 33) per the previous update to Reference 33 provided in Comments 12 and 42.
17 27 12 - 13 Clarification Comment (including suggested revision): The staff agrees that the comment is valid but proposes an alternate solution This statement appears to be referring to to the comment. Westinghouse is submittal of WCAP-17096-NP, Revision 3 to the referring to the clause in Lines 12 and staff rather than issuance of WCAP-15029-P-A, 13 of the sentence currently on DSE Revision 1. Therefore, prior to the PWROGs Page 27, lines 10 - 15, which states:
issuance of the NRC-approved ABPA although WCAP-18034-P was issued methodology in WCAP-15029-P-A, Revision 1, prior to the PWROGs issuance of the should be updated to, prior to the PWROGs approved ABPA methodology in submittal of WCAP-17096-NP, Revision 3 to the WCAP-15029-P-A, Revision 1. . . Upon Staff for approval in July 2019 further review, the staff finds that the clause starting with the word although is not necessary for the context of the referenced sentence; thus, the clause starting with the word although was deleted from the referenced sentence.
18 27 46 Accuracy Comment (including suggested revision): The staff agrees that the comment is valid and made the editorial change A correction is needed to the quoted statement. identified in the comment.
The wording, may have changes over two decades should be, may have been changed over the two decades 19 27 50 Clarification Comment (including suggested revision): The staff agrees that the comment is valid and incorporated the editorial Part C of RAI-07 deleted a sentence in change identified in the comment into Section 3.3 of the TR R3, not the entire section. the Final SE.
Therefore, the statement, the complete deletion of Section 3.3 to Revision 3 of the TR should be updated to, the deletion of B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 19 the statement in section 3.3 of Revision 3 of the (Continued) TR 20 28 2 Clarification Comment (including suggested revision): The staff agrees that the comment is valid with respect to a need of The updated RAI response was dated referencing a proper date for the January 14, 2022. Therefore, February 8, PWROG record containing the 2022 should be changed to January 14, response to RAI-07. The RAI response 2022 is contained in ADAMS Accession No. ML22039A144, but the date of the record is January 13, 2022, and not January 14, 2022 (which instead applies to the transmittal letter). So, the staff amended the reference date of the RAI-07 response to January 13, 2022.
21 29 11 - 12 Clarification Comment (including suggested revision): The staff agrees that the comment is valid. For the RAI-13 record referenced All RAIs on TR R3 were issued via on DSE Page 29, Line 11, the staff Reference 23, dated September 21, 2020. amended the cited date from Therefore, September 24, 2020 should be September 24, 2020 to updated to September 21, 2022, and September 21, 2020.
Reference 36 should be updated to Reference 23. See related comments 15 and For the reference item associated with
- 44. the RAI-13 record referenced on DSE Page 29, Line 12, the staff amended the reference cited to Reference 23.
22 29 41 - 42 Clarification Comment (including suggested revision): The staff agrees that the comment is valid. Upon further review, the staff Reference 40 is MRP 2019-009 (CB weld confirmed and amended this Final SE interim guidance). BWRVIP 2019-016 was cited with a new reference within this Final in TR R3 as an example of what the NRC has SE Section 7.0 that corresponds to the previously accepted for submittal of probabilistic BWRVIP 2019-016 record and that the fracture mechanics. Recommend adding hyperlinked reference item cited in this BWRVIP 2019-016 as an additional reference: Final SE cites and hyperlinks to the EPRI Letter, BWRVIP 2019-016, White Paper Final SE Section 7.0 reference item for on Suggested Content for PFM Submittals to BWRVIP 2019-016. Specifically:
B Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 22 the NRC, February 27, 2019. See related 47. EPRI Letter No. BWRVIP 2019-(Continued) Comment 46. 016, White Paper on Suggested Content for PFM Submittals to the Staff comment and observation: NRC, February 27, 2019 (ADAMS Accession No. ML19241A545).
The staff perceives the comment to be an accuracy based comment with what DSE Refer to the staffs resolution basis for Section 7.0, Reference 40 should be citing in the related comments in joint the hyperlinked Reference 40 record or Comments 45 and 46.
alternatively whether Line 42 on DSE Page 29 should be referencing a new hyperlinked reference item in SE Section 7.0 that corresponds to the BWRVIP 2019-016 record (and whether the Reference item listed on DSE Page 29, Line 42 should be cited and hyperlinked to the new reference item that will be added to Final SE Section 7.0 for the BWRVIP 2019-016 record).
23 32 28 Editorial Comment (including suggested revision): The staff agrees that the comment is valid and made the associated editorial Correct document number TB 19-5 to change to this Final SE to correct the TB-19-5. typographical error in the TB-19-5 reference.
24 35 9 Editorial Comment (including suggested revision): The staff agrees that the comment is valid and made the associated editorial Correct CE ID to CE-ID. change to this Final SE; the CE-ID reference mentioned in Comment 24 is currently located to Page 35, Line 17 of the Final SE.
25 37 2 and 4 Clarification Comment (including suggested revision): The staff agrees that the comment is valid but proposes an alternative Please add the CE-ID number (CE-ID 2.1), to solution for resolution of the comment this component to distinguish between based on the NRC PMs formatting CE-ID 3.1 on Line 4. decision not to include specified ID item references for component-specific bullet items. For the referenced Core Shroud B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 25 Assembly (Welded) - Remaining Axial (Continued) Welds bullet referenced in Comment 25, the staff added the words (CE-design units with welded core shrouds assembled in two vertical sections),
which correspond to the type of welded CE-design unit core shroud associated with the CE-ID Item 2.1 assessment in Appendix C of the subject TR.
Similarly, for the corresponding Core Shroud Assembly (Welded) -
Remaining Axial Welds bullet, the staff deleted the words -ID 3.1 from the bullet item and add the words (CE-design units with welded core shrouds assembled with full height shroud plates), which corresponds to the type of welded CE-design unit core shroud assessed in the CE-ID Item 3.1 assessment of Appendix C of the subject TR.
26 37 5-9 Clarification Comment (including suggested revision): The staff agrees that Comment 26 for the specified CE-design core barrel Please add CE before these component weld components is valid. The staff names to distinguish these from Westinghouse added CE at the beginning of the components. specified CE-design component-specific bullet items. Additionally, the Staff Comment and Observation: The specified Lower Support Structure, Deep comment applies to the five (5) bullets for Beams referenced in DSE Page 37, specified CE-design core support barrel (CSB) Line 10 of the DSE are only applicable weld types (i.e., bullets for CE CSB Assembly to CE units with welded core shrouds UGWs, UAWs, MGWs, MAWs, and LAWs). assembled with full height shroud plates. The staff added to this Final SE a parenthetical clarification to the CE component-specific bullet for the deep beams.
B Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 27 37 11 - 15 Clarification Comment (including suggested revision): The staff agrees that Comment 27 for the specified Westinghouse-design core Please add W before these component names barrel weld components is valid. The to distinguish from CE components. staff added W at the beginning of the specified Westinghouse-design Staff comment and observation: component-specific bullet items in this Final SE.
The specified comment applies to the five (5) bullets for specified Westinghouse-design core barrel weld types (i.e., bullets for Westinghouse Core Barrel Assembly UGWs, UAWs, LGWs, MAWs, and LAWs).
28 37 1 - 15 Clarification Comment (including suggested revision): Staff agrees comment is valid. The staff added following component-specific A total of 21 CE and Westinghouse component bullet items to this Final SE:
methodologies use a flaw tolerance evaluation with the fracture toughness limits that are CE CSB potentially impacted by the Open Item; however, Assembly, UFW only 15 components are listed. The following six CE CSB additional components should be included: Assembly, LGW CE Lower Support CE CSB Assembly, UFW Structure, Lower CE CSB Assembly, LGW Core Support CE Lower Support Beams Structure, Lower Core CE CSB Support Beams Assembly, CSB CE CSB Assembly, CSB Flexure Weld Flexure Weld W Core Barrel W Core Barrel Assembly, Assembly, UFW UFW W Core Barrel W Core Barrel Assembly, Assembly, LFW LFW The new component-specific bullet for See related Comment 53. CE lower support structure, lower core support beams includes the following parenthetical clause: (All CE-design units except those with welded core B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 28 shrouds assembled with full height (Continued) shroud plates).
29 37 22 - 23 Clarification Comment (including suggested revision): Staff agrees with the comment. See the staffs resolution of the similar PWROG These values are from the Executive Summary email comment that is covered by a line of MRP-211, which was made public when this item at the end of this table. The was submitted to the Staff. Therefore, these specified 6.7x1021 n/cm2 (E> 1.0 values do not need to be redacted, and the MeV) fluence exposure and 10 dpa Safety Evaluation can be considered non- projected dose exposure values in proprietary. The transmittal letter for MRP-211 question are now included in this Final is: Electric Power Research Institute (EPRI), SE. The values were downgraded and MRP 2017-036, Transmittal of Revision 1 to listed as non-proprietary contents for EPRI Technical Reports MRP-175 and MRP- this Final SE on the subject TR. In 211, December 18, 2017 (ML17361A187). addition, the Proprietary headers and footers were removed from this Final Staff comment and observation: SE per the staffs acceptance of this comment and the analogous email A similar comment was provided in a PWROG comments on proprietary contents in email to the staff. The staff has included that the DSE.
email-based comment at the end of this table.
30 39 34 Editorial Comment (including suggested revision): The staff agrees that the comment is valid and fixed the typographical error Correct W--D to W-ID. of the acronym in this Final SE. The acronym now reads as W-ID.
31 43 Ref. 2 Clarification Comment (including suggested revision): The staff recognizes that the comment is accurate, but no changes of the Accession number should be ML21099A172. referenced ADAMS Package number needed to be made in response to Comment 31 for Reference 2.
Specifically, the staff entered the same Package records into two different packages in ADAMS: (1) one by the Package in ADAMS Accession No. ML21099A172, as indicated in Comment 31, and (2) again in the Package in ADAMS Accession B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 31 No. ML22187A141, as currently (Continued) indicated in Reference 2 of Section 7.0 of this Final SE.
Thus, the citing of Package Item ML22187A141 in Reference 2 is accurate, and the Package has the same records as the package in ADAMS Package No. ML21099A172.
32 43 Ref. 3 Editorial Comment (including suggested revision): The staff recognizes that the comment is accurate and valid in that the cited The PA number (PA-MSC-1567) is missing title for Reference 3 omitted the (PA-from the document title. The title should be, MSC-1567) term in the record title. The Transmittal of the Response to Request for staff added the (PA-MSC-1567) term Additional Information RAIs Associated with into the title of Reference 3 to make it WCAP-17096, Revision 3, Reactor Internals accurate with the title of the actual Acceptance Criteria Methodology and Data record in ADAMS Package Requirements, (PA-MSC-1567). No. ML22187A149.
33 43 Ref. 3 Clarification Comment (including suggested revision): The staff recognizes that the comment is accurate, but no changes of the Accession number should be ML22039A142. referenced ADAMS Package number needed to be made in response to Comment 33 and Reference 3.
Specifically, the staff entered the same Package records into two different packages in ADAMS: (1) one by the Package in ADAMS Accession No. ML22039A142, as indicated in Comment 33, and (2) again in the Package in ADAMS Accession No. ML22187A149, as currently indicated in Reference 3 of Section 7.0 of this Final SE.
Thus, the citing of Package Item ML22187A149 is accurate, and the Package has the same records as the B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 33 package in ADAMS Package Item (Continued) No. ML22039A142.
34 43 Ref. 4 Clarification Comment (including suggested revision): The staff agrees the comment is accurate and valid. The staff made the The document revision number is missing from stated editorial changes to Reference 4; the citation and the title is incorrect. Revise however, the term Non-Proprietary citation as follows: Class 3 did not need to be removed from the Reference 4 item.
Westinghouse Electric Company Topical Report WCAP-17096-NP, Revision 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements, July 2019 (Pkg ML22187A179).
35 43 Ref. 13 Clarification Comment (including suggested revision): The staff agrees the comment is accurate and valid and that ADAMS Accession number should be ML16279A320. No. ML16279A320 is the proper reference for WCAP-17096-NP-A, Rev.
2 (Reference 13).
The staff confirmed that ADAMS No. ML16274A399 is the ADAMS reference for the NRCs NUREG-2191, Volume 2, Generic Aging Lessons Learned for Subsequent License Renewal GALL-SLR) Report; thus, the citing of ADAMS No. ML16274A399 in DSE Reference 13 is incorrect and the staff corrected the citation of ML16279A320 as the applicable ADAMS number for the Agency record cited for Reference 13 within Section 7.0 of this Final SE.
36 44 Ref. 24 Clarification Comment (including suggested revision): The staff agrees the comment is accurate and valid. The staff corrected Document date should be January 14, 2022. the referenced date for Reference 24 within Section 7.0 of this Final SE.
B Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 37 44 Ref. 25 Clarification Comment (including suggested revision): The staff agrees the comment is accurate and valid. The staff corrected Document date should be January 13, 2022. the referenced date for Reference 25 within Section 7.0 of this Final SE.
38 44 Ref. 27 Editorial Comment (including suggested revision): The staff agrees the comment is accurate and valid. The staff corrected Misspelling of word in title. Should be the misspelling of the word Austenitic Austenitic Stainless Steel in the title of the Agency record that is cited for Reference 27 within Section 7.0 of this Final SE.
39 45 Ref. 30 Editorial Comment (including suggested revision): The staff agrees the comment is accurate and valid. The staff corrected Location of publisher should be, New York, the location of the cited publisher of the NY. Agency Record cited for Reference 30 within Section 7.0 of this Final SE.
40 45 Ref. 31 Clarification Comment (including suggested revision): The staff agrees the comment is accurate and valid. The staff inserted Letter number is missing from the citation. the words Letter No. MRP 2016-021 Please add MRP 2016-021. into the Agency Record cited for Reference 31 within Section 7.0 of this Final SE.
41 45 Ref. 32 Clarification Comment (including suggested revision): The staff agrees the comment is accurate and valid. The staff inserted The letter number is missing from the citation. the words Letter No. MRP 2017-009 Please add MRP 2017-009. into the Agency Record cited for Reference 32 within Section 7.0 of this Final SE.
42 45 Ref. 33 Clarification Comment (including suggested revision): The staff partially agrees that the comment is valid. Westinghouse is Reference 33 should be the original NRC- correct that Reference 33 in DSE approved Westinghouse ABPA methodology. Section 7.0 should have referenced the This reference should be updated to: submittal of the WCAP-15029-P-A/
Westinghouse Reports, WCAP-15029-P-A, Rev. WCAP-15030-NP-A reports, but it 0 and WCAP-15030-NP-A, Rev. 0, should be to the latest approved version B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 42 Westinghouse Methodology for Evaluating the of the Proprietary report (WCAP-15029-(Continued) Acceptability of Baffle-Former-Barrel Bolting P-A, Rev. 1), which was transmitted in Distributions under Faulted Load Conditions, PWROG Letter OG-18-226, dated December 1998 / April 1999 (Pkg. October 8, 2018, and included ADAMS No. ML20206G308). Package No. ML18306A583.
It is recommended that WCAP-15029-P-A, The staff amended Reference 33 within Rev. 0 include a footnote to explain that WCAP- Section 7.0 of this Final SE per the 15029-P-A, Rev. 1 is equivalent to Rev. 0 with corresponding comments in the Safety Evaluation attached. Westinghouse Comments 12 and 42, but as modified by the staffs Staff Comment and Observation: observations for Reference 33 regarding the ADAMS Package number Westinghouse Electric Companys comment for the latest staff-approved version of associated with Comment 42 is actually in WCAP-15029-P-A (currently reference to the manner the Reference 33 Revision 1). A footnote was included in record is worded in DSE Section 7.0 (on SE the amended version of Reference 33 page 43) - see Westinghouse related Comment regarding the errata for the report.
- 12. Per Comment 42, the PWROG indicates that Reference 33 should be citing: Based on acceptance of the Westinghouse Reports, WCAP-15029-P-A, Rev. Comments 12 and 42, the staff 0 and WCAP-15030-NP-A, Rev. 0, confirmed the proper reference for Westinghouse Methodology for Evaluating the Reference 33 is given in PWROG Letter Acceptability of Baffle-Former-Barrel Bolting No. OG-18-226, October 31, 2018 Distributions under Faulted Load Conditions, (ADAMS Package No. ML18306A583),
December 1998 / April 1999 (Pkg. which has the latest staff-approved No. ML20206G308). However, the updated versions of the WCAP-15029-P-A and staff-approved version of the proprietary report WCAP-15030-NP-A reports (i.e.,
(WCAP-15029-P-A, Rev. 1) was submitted in Revision 1 and Revision 0, PWROG Letter No. OG-18-226, Oct. 31, 2018. respectively).
Thus, the staff is of the opinion Reference 33 (now on Final SE Page 45) should be citing the updated staff-approved version in WCAP-15029-P-A, Rev. 1, as included in OG-18-226.
43 45 Ref. 35 Editorial/Clarification Comment (including suggested revision): The staff agrees the comment is valid, and that a number of corrections to The accession number of the non-proprietary Reference 35 were necessary.
version needs to be corrected and should be Reference 35 should be the citation for B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 43 indicated as non-proprietary. Also, the MRP-211, Revision 1 report, as (Continued) ML15223A269 cannot be located in ADAMS. cited and hyperlinked to Reference 35 The corrected accession number info is: on Page 36 of the DSE. But Page 34 (ML080110645 - Non-Proprietary/Public) improperly stated that Reference 35 is for the BWRVIP-100-A report. For the Final SE to be correct, Section 7.0 was corrected to give the Reference 35 citation as the cited record for the EPRI MRP-211 report (EPRI Report 3002010270).
The version of BWRVIP-100 cited in the DSE should be BWRVIP-100, Revision 1-A, which is Reference 41 within Section 7.0 of this Final SE.
44 45 Ref. 36 Clarification Comment (including suggested revision): See the staffs resolution of the suggested action in Westinghouse All of the RAIs for TR R3 were provided in Comment 13, which includes the staffs Reference 23. This ML number cannot be found basis for amending Reference 36 within in ADAMS. This reference should be deleted. Section 7.0 of this Final SE per the joint See related Comments 15 and 21. comment amendment actions recommended in Westinghouse Comments 13 and 44.
45 and 46 46 Ref. 40 Editorial Comments (including suggested revision): The staff agrees the comment is valid and Reference 46 within Section 7.0 of Correct the document number from MRP- this Final SE was revised. However, 2019-023 to MRP 2019-023. BWRVIP 2019- BWRVIP 2019-016 is correct when 016 is not associated with MRP 2019-023. MRP taken in context with Comment 22 and 2019-023 transmits letter MRP 2019-009. the staffs basis for resolving Comment Therefore, (BWRVIP 2019-016) should be 22. The staff made the associated changed to (MRP 2019-009). editorial change to Reference 47 within Section 7.0 of this Final SE.
47 46 Ref. 42 Editorial Comment (including suggested revision): The staff agrees the comments are valid. The staff made the editorial Correct the document title: change BWR-VIP- change to Reference 42 within 100 should be changed to BWRVIP-100. Section 7.0 of this Final SE.
B Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 48 46 Ref. 43 Clarification Comment (including suggested revision): The staff agrees the comment is valid.
The staff deleted Reference 43. This The document is a repeat of Reference 40 and resulted in a shift of the Reference can be deleted. numbers up one for references coming after Reference 43 (e.g., Reference 44 became the new Reference 43, and Reference 45 became the new Reference 44, etc.) The referencing of those references in the body of the Final SE was adjusted accordingly.
49 43 Ref. 44 Editorial Comment (including suggested revision): The staff agrees the comment is valid.
The staff made the editorial change to Correction to title, CWAP-17096 should be Reference 44 of this Final SE which WCAP-17096. became the new Reference 43 after the Comment 48 change was implemented.
50 A-3 L-03 Clarification Comment (including suggested revision): The staff agrees the comment is valid and deleted the extra reference of the Rev. 3 Appendix C, CE-ID Item 8 is listed Ref. 3 Appendix C, CE-ID Item 8 in twice. One instance should be deleted to clarify the last column entry of this Final SE the intent. Table A-1, Limitation L-03 line item.
51 A-5 C-01 Clarification Comment (including suggested revision): The staff agrees the comment is valid.
The following statement in the The following statement is included in the Technical Basis column entry in technical basis, In the response to RAI-04, the Table 1, Condition C-01 line item of the PWROG established bounding CGRs for DSE:
stainless steel welds with neutron fluence exposures shown in Figure 2 of this SE. In the response to RAI-04, the PWROG established bounding CGRs RAI-04 was only related to high fluence for stainless steel welds with neutron components; therefore,16 high fluence fluence exposures in the fluence ranges components were identified in the response to shown in Figure 2 of this SE.
RAI-04. However, since ASME Code Case N-889 applies to all fluence ranges in Figure 2, a Now reads as the following in the Final total of 23 components are affected (2 B&W, 14 SE:
CE, and 7 Westinghouse).
B Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 51 It is proposed that this statement is updated as RAI-04 applied to 16 high fluence (Continued) followed to clarify: components. However, in the response to RAI-04, the PWROG established RAI-04 applied to 16 high fluence components. bounding CGRs for stainless steel However, in the response to RAI-04, the welds with neutron fluence exposures PWROG established bounding CGRs for shown in Table 2 of this SE. Therefore, stainless steel welds with neutron fluence Condition 1 is applicable to all 23 exposures shown in Figure 2 of this SE. components (i.e., 2 B&W, 14 CE, and Therefore, Condition 1 is applicable to all 23 7 Westinghouse) that use the flaw components (i.e., 2 B&W, 14 CE, and tolerance evaluation method.
7 Westinghouse) that use the flaw tolerance evaluation method. See Comment 52.
52 A-5 C-01 Clarification Comment (including suggested revision): The staff agrees the comment is valid.
The staff made the applicable editorial There are two typographical errors in the ID changes to the last column entry for the numbers of the 23 components. The B&W lower Condition C-01 line item in Table A-1 of grid rib section is A.2.11 rather than A.2.3. this Final SE. Refer to the staffs Also, the CE Lower Support Structure is CE-ID resolution of Comment 51 as well.
11 rather than CE-ID 1.
53 A-7 OI-01 Clarification Comment (including suggested revision): The staff agrees that the intent of the comment is valid. However, subsequent The fracture toughness Open Item 1 applies to to receipt of Comment 53, the staff 21 components, which should be identified in determined that it could not issue the Table OI-01. Therefore, CE-ID 5 (CSB final SE with an open item (OI-01).
Assembly - Upper Flange Weld) should be Accordingly, the staff converted OI-01 added. See related comment 28. into a new Condition (new Condition C-03). Thus, the staff points out that the ID item referenced in Comment 53 as previously applying to Open Item OI-01 (deleted in this Final SE) now applies to new Condition C-03 in this Final SE.
54 4 Figure 1 Editorial Comment (including suggested revision): Hyperlinks to the reference section were removed.
Hyperlinks do not work. Correct hyperlinks as necessary.
55 4 Ref. 19 Clarification Comment (including suggested revision): The staff agrees the comment is valid.
in The staff made the applicable editorial B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 55 Figure 1 A correction is needed to the SLR-Interim Staff changes the line items for Interim Staff (Continued) Guidance document title. NRCs SLR-Interim Guidance No. SLR-ISG-2021 Staff Guiance-PWRVI-2021 should be PWRVI and the Federal Register Notice NRCs SLR-Interim Staff Guidance-2021 for the SLR-ISG in Table 1 of this Final PWRVI. SE for the subject TR.
56 5 38 Editorial Comment (including suggested revision): The staff agrees the comment is valid and deleted the extraneous an from Delete extraneous an from sentence to clarify: the referenced bullet item.
NRC staffs determination that the PWROG-specified an acceptance criteria methodology or data requirement basis for a given component ID item assessment in Appendix A, C, or E of Revision 3 of the TR was should be NRC staffs determination that the PWROG-specified acceptance criteria methodology or data requirement basis for a given component ID items assessment in Appendix A, C, or E of Revision 3 of the TR was 57 10 1-6 Clarification Comment (including suggested revision): The staff acknowledges the comment and agrees the comment appears to This paragraph discusses use of the WCAP- make a valid point. Consistent with the 17096-NP acceptance criteria methodology and comment, the staff deleted the data requirements to meet the acceptance paragraph in this Final SE.
criteria for the component item in Section 5 of MRP-227, Rev. 1-A. This seems to be referring to the Examination Acceptance Criteria column in Table 5-1, 5-2, and 5-3. These Examination Acceptance Criteria are the specific relevant (or reportable) conditions used by an inspector when performing the MRP-227 inspections. The MRP-227, Section 5 Examination Acceptance criteria are not related to the WCAP-17096 methodologies, which are intended to create criteria for addressing the relevant conditions that are identified in inspections. Therefore, this paragraph discusses cases that are not relevant B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 57 and may add confusion to the Final SE, and it is (Continued) requested that this paragraph be deleted.
58 11 15 - 19 Clarification Comment (including suggested revision): The staff did not find the Westinghouse Comment 58 to be valid for A change in the inspection categorization would implementation because the point of the not impact the methodologies in WCAP-17096- limitation in Limitation L-01 applies to NP, Rev. 3; however, a change in the type of potential use of applicable component-inspection could possibly impact the method. specific ID item assessments for B&W, Correct inspection categorizations to type of CE, and Westinghouse design Primary inspection. and Expansion components in Appendix A, C, or E of WCAP-17096, Revision 3 when compared to an updated list of B&W, CE, and Westinghouse design Primary and Expansion category components in an 80-year version of the MRP-227 report (e.g., a staff-approved version of MRP-227, Revision 2, which was submitted for staff review in early May of 2022).
Specifically, the list of Primary and Expansion category component ID item assessments in Appendices A, C, and E of the subject TR (WCAP-17096, Rev. 3) correspond to the set of Primary and Expansion category components as approved in the MRP-227, Revision 1-A; however, the Primary and Expansion category components addressed by the ID item assessments in Appendices A, C, and E of the subject TR do not perfectly correspond to the set of Primary and Expansion category components for B&W, CE, and Westinghouse components in the MRP-227, Revision 2 report, where there were some changes to the defined set of Primary and Expansion B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 58 category components in MRP-227, (Continued) Revision 2 from the prior set of Primary and Expansion category components in MRP-227, Revision 1-A. Thus, the staffs statement on Page 11, Lines 15
- 19 of the DSE (as referenced in Westinghouse Comment 58) is appropriately being made in reference to changes in inspection category and not to changes in inspection method.
The sentence was left as written in this Final SE.
59 16 28 Clarification Comment (including suggested revision): The staff agrees the comment is valid and made the editorial change of the CE-designed core support should be CE- referenced bullet item to list the designed core support columns applicable CE-design RVI component as CE-design cores support columns.
60 17 37 Editorial Comment (including suggested revision): The staff agrees the comment is valid and made the editorial change of the W-designed CB middle axial welds should be referenced SE bullet in this Final SE.
W-designed CB MAWs since MAWs was already defined above.
61 17 47 Clarification Comment (including suggested revision): The staff agrees the comment is valid.
The staff made an editorial change Correct one year to next RFO. within this Final SE that replaces the words no later than one year with the words at the next refueling outage (RFO) . . .
62 19 31 Editorial Comment (including suggested revision): The staff agrees the comment is valid and has indented the bullet for the CE-This is the first item being listed and should be design core shroud plate-to-former indented and bulletized like the other four items. plate welds accordingly within this Final SE.
63 30 Footnote Editorial Comment (including suggested revision): The staff agrees the comment is valid 7 and applies throughout the DSE in a manner that is more generic than B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 63 APB should be corrected to ABP in the implied in Westinghouse Comment 63.
(Continued) sentence for the assembly that would then The staff searched the DSE for the term be compared to the APB or set of ABPs, should APB and globally changed that to be for the assembly would be compared to ABP within this Final SE. The staff the ABP or set of ABPs. also searched the DSE for the term APBA and changed that globally to Staff Comment and Observation: ABPA within this Final SE.
The staff observes that the type of typographical error referenced in Westinghouse Comment 63 applies more generically than just the noted typographical error of the acronym APB listed in the referenced SE footnote. It applies to some existing typographical errors of the acronyms for approved bolting patterns (incorrectly listed as APB) or approved bolting pattern analyses (incorrectly listed as APBA) throughout the DSE for commentary, in which the acronyms need to be corrected to ABP or ABPA.
64 33 1-2 Clarification Comment (including suggested revision): The staff acknowledges the comment may be valid. Upon further review, the The operating experience being referred to in staff has determined that the referenced and with the exception of distortion recently parenthetical clause (and with the reported in one of the clevis insert assemblies in exception of distortion in one of the a U.S. W-design PWR was not considered clevis insert assemblies in a U.S.
distortion of the component. The clevis insert design PWR) is not necessary for the at that plant was displaced from its installed referenced sentence on Page 33, Lines location. Discussion of that operating 1 - 2 of the DSE. The staff deleted the experience in the context of void swelling is reference parenthetical clause (and misleading both due to the lack of distortion in with the exception of distortion in one of the experience and due to the fact that the the clevis insert assemblies in a U.S.
clevis insert assembly is located far from the design PWR) from the scope of the core and receives very low radiation doses sentence within this Final SE.
orders of magnitude below the amount needed to cause void swelling distortion. It is recommended that (and with the exception of distortion recently reported in one of the clevis B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 64 insert assemblies in a U.S. W-design PWR) be (Continued) deleted.
65, 66, 38 and P38, Clarification Comment (including suggested revision) applies The staff did not find these comments 67, and 39 Lines 41 to Comments 65, 66, 67, and 68): to be acceptable for implementation in 68 and 46, Section 3.4.11 of the Final SE.
Correct VT-3 to EVT-1, ET, or UT. Specifically, the use of EVT-1, ET, or P39, UT as specified inspection techniques Lines 2 Staff Comment and Observation: for replacing visual inspection methods and 8 in MRP-227 space typically applies to Staff perceives the Westinghouse- the EPRI MRPs expansion of the types recommended SE adjustment action in of inspection methods that may be used Westinghouse Comments 65 - 68 to be for management of cracking in RVI applicable to the staffs referencing of VT-3 structural seam weld types (e.g., those inspection methods throughout DSE Section in the welds of welded core barrel or 3.4.11 and not just to the portions of DSE core support barrel assembly Section 3.4.11 that are located on Pages 38 and structures), as redefined and given in 39 of the Draft SE for Commentary. Thus, the EPRI 80-Year (SLR-based) reports staff will respond to these comments based on (e.g., MRP-227, Revision 2; MRP-231, the referencing of VT-3 inspection methods Revision 4; MRP-232, Revision 2; or throughout Final SE Section 3.4.11. MRP 2018-022). But, the ID item assessments in Appendices A, C and E of the subject TR correspond to those Primary and Expansion category components in MRP-227, Revision 1-A, which is a 60-Year (initial LR-based) report.
In contrast, the staff observes that the referencing of VT-3 visual methods in Section 3.4.11 applies to use of VT-3 visual inspection methods for aging effects in specific Babcock and Wilcox (B&W)-design Primary or Expansion category components other than cracking, as linked to the VT-3 monitoring basis in Section 3.2 of the subject TR and a number of specified component-specific B&W ID item B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1) 65, 66, assessments in Appendix A of the 67, and subject TR. For B&W-design Primary or 68 Expansion category components (Continued) associated with these B&W ID item assessments, the VT-3 methods are appropriate to monitor for conditions that may be indicative of gross conditions occurring in the specified B&W-design components (e.g., evidence of gross indications of wear, loose or missing parts, severed parts, etc.). Thus, the staff does not find recommended action for generically switching VT-3 over to EVT-1, ET, or UT to be appropriate for the referencing those B&W-design ID item components in Appendix A of the subject TR that are the subject of the staffs evaluation in Section 3.4.11 and the VT-3 methodology described in Section 3.2 of the subject TR.
PWROG 35 27 and Clarification Summary of Comment (including suggested The staff agrees that the comment is email 28 revision): valid. The staff confirmed that the Comment PWROG eMail comment 6.7x1021 n/cm2 (E> 1.0 MeV) fluence (from Mr. J. Andrachek) Proprietary values listed on DSE Page 35, Lines exposure and 10 dpa projected dose on inclusion of proprietary 27 and 28 can be downgraded to non- exposure values linked to the cited lower bound fracture proprietary contents. The > 6.7x1021 n/cm2 34.6 ksi-inch lower bound fracture toughness values in the (E> 1.0 MeV) fluence exposure and > than toughness value in the DSE were DSE. An analogous 10 dpa projected dose exposure linked to the included in the Executive Summary of comment is given in cited 34.6 ksi-inch lower bound fracture the EPRI MRP-211, Revision 1 report Westinghouse Comment toughness value in the DSE were included in and that the Executive Summary in the
- 29. the Executive Summary of the Proprietary report is given as publicly available MRP-211 Report (ADAMS ML17361A191); the information. The staff amended this Executive Summary is included in MRP-211, Final SE to downgrade the cited Rev. 1 as publicly available information. 6.7x1021 n/cm2 (E> 1.0 MeV) fluence exposure and 10 dpa projected dose The PWROG recommends downgrading the exposure values as non-proprietary referenced 6.7x1021 n/cm2 (E> 1.0 MeV) values. In addition, the proprietary B
Text Location in Comment Type DSE (Clarification, Editorial, NRC Staff Basis for Responding to Comment Basis, Accuracy, Comment, Including Applicable PWROG Page(s) Line(s) and Resolving Specified Comment No. Proprietary) Suggested Resolution or Revision (See (See Received Note 1) Note 1)
PWROG fluence exposure and 10 dpa projected dose headers and footers contained in the email exposure values to non-proprietary content draft SE for commentary were Comment status in the Final SE, which can result in removed from the scope of this Final (Continued) issuance of the Final SE as a non-proprietary SE in accordance with this comment.
NRC Agency Record in ADAMS.
Consistent with the PWROGs perception, this results in the issuance of this Final SE for WCAP-17096, Rev.
3 as a non-proprietary agency record in ADAMS.
Note: 1. The DSE pages and lines identified by Framatome Corporation or Westinghouse Electric Company in their comments did not account for the fact that the first two pages of MS Word file containing the Draft SE Issued for Commentary were taken up by the Table of Contents. The staff inserted the correct DSE (ML22181B009) page(s) and line number(s) within the Table of Attachment B of this Final SE.
B