ML23324A422

Reactor Vessel Internals Inspections Aging Management Program Submittal Related to License Renewal Commitment 13

ML23324A422
Person / Time
Site:	Millstone
Issue date:	11/20/2023
From:	James Holloway Dominion Energy Nuclear Connecticut
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
23-290
Download: ML23324A422 (1)

Text

Dominion Energy Nuclear Connecticut, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 Dominion Energy.com November 20, 2023 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 REACTOR VESSEL INTERNALS INSPECTIONS AGING MANAGEMENT PROGRAM SUBMITTAL RELATED TO LICENSE RENEWAL COMMITMENT 13 Serial No.:

NRA/SS:

Docket No.:

23-290 RO 50-423 License No.:

NPF-49 Commitment Item No. 13 of NUREG-1838, the Safety Evaluation Report (SER) for the renewed operating license for Millstone Power Station Unit 3 (MPS3), requires the submission of a revised aging management program (AMP) description for enhanced inspections of the reactor vessel internals to the Nuclear Regulatory Commission (NRC).

In accordance with Commitment Item No. 13, the inspection program for the reactor vessel internals has been updated to include the ten program elements of NUREG-1801 and industry guidance from Electric Power Research Institute (EPRI) guidance document EPRI Report 1022863, "Material Reliability Program: Pressurized Water Reactor Inspection and Evaluation Guidelines (MRP-227, Revision 1-A)." The MPS3 reactor vessel internals inservice inspection AMP description is provided in the enclosure for NRC review and approval.

If you have any questions or require additional information, please contact Shayan Sinha at (804) 273-4687.

Sine;~~~

James E. Holloway Vice President - Nuclear Engineering & Fleet Support

Commitments made by this letter: None

Enclosure:

Serial No.: 23-290 Docket No.: 50-423 Page 2 of 2 Aging Management Program Description, lnservice Inspection: Reactor Vessel Internals cc:

Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road, Suite 102 King of Prussia, Pennsylvania 19406-1415 Richard. V. Guzman NRC Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E3 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector Millstone Power Station

Serial No.: 23-290 Docket No.: 50-423 ENCLOSURE Aging Management Program Description Inservice Inspection: Reactor Vessel Internals DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3

LICENSE RENEWAL AGING MANAGEMENT PROGRAM 1

BACKGROUND The reactor internals are passive structures within the reactor vessel that provide support for the reactor core as well as guiding the cooling water flow and providing guidance for control rod insertion. The materials are primarily stainless steel and the environment is the reactor coolant system primary water plus irradiation. The aging effects of concern are loss of material, cracking, loss of pre-load, change in dimensions due to void swelling, and loss of fracture toughness. The consequences of these aging effects as well as the potential mechanisms that may cause them have been the subject of study of an industry program, the Materials Reliability Program (MRP). This program is an industry-wide initiative conducted by the Electric Power Research Institute (EPRI) with the participation of utilities and vendors. A result of these MRP studies is the current inspection guideline MRP-227 Rev. 1-A (Reference 8), which has been adopted by the nuclear industry as the common program for managing the anticipated aging effects of reactor internals.

Periodic inspections of the internals are performed as part of the existing MPS3 inservice inspection (ISI) program, which meets the requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI (Reference 1). Until the beginning of the period of extended operation (PEO), the ISI program provided the primary inspection requirements to ensure that the integrity is maintained for the components comprising the reactor internals. In many instances the ISI program remains adequate for management of aging effects and is therefore referenced by this Aging Management Program (AMP) document. The additional examinations beyond the requirements of ASME Section XI are governed by this AMP. The technical requirements for such augmented examinations are further specified in EPRI inspection standard, MRP-228 (Reference 9).

In addition to its use as a means of satisfying license renewal commitments, MRP-227 Rev. 1-A has been issued as a guideline with Needed and Mandatory requirements as defined by NEI 03-08. Per the NEI-03-08 guidelines, expected utility actions are classified relative to the level of importance such that: (1) Mandatory is to be implemented at all plants where applicable, and (2) Needed is to be implemented whenever possible but alternative approaches are acceptable. A Dominion Energy fleet procedure (Reference

12) mandates implementation of this protocol.

The NRC reviewed Revision 1 of MRP-227 and agreed that it formed the basis for an acceptable program for aging management of reactor internals, with certain conditions.

In their Safety Evaluation (SE), the NRC identified certain changes to requirements for incorporation in a revised version of the document, and one licensee action item (LAI) related to baffle-former bolting (Reference 11). MRP-227 Rev. 1-A incorporates the changes required by the NRC. The LAI required by the SE is not a direct requirement of MRP-227 Rev. 1-A but must be addressed in the plant-specific inspection plan. For Millstone Power Station Unit 3 (MPS3), this inspection plan must be submitted to the NRC for review and approval.

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LICENSE RENEWAL AGING MANAGEMENT PROGRAM 2

DISCUSSION This section comprises a description of the program. The Reactor Vessel Internals Inspection Program is a plant-specific program that implements aging management activities for the reactor internals as recommended by the relevant industry guideline, MRP-227 Rev. 1-A. The program is coordinated with the ASME XI ISI Program (Reference 2, 3) and supplements its requirements. The current inspection plan under MRP-227 Rev. 1-A for MPS3 is contained in the Attachment 1 tables beginning on page 22 of this attachment. Any changes to the program or its implementation that constitute a deviation to the Mandatory or Needed requirements of MRP-227 Rev. 1-A will be processed in accordance with corporate procedures for implementing NEI 03-08 requirements (References 12-14). Accordingly, any such changes will be reported to the NRC.

The scope of this AMP includes those passive reactor vessel internal subcomponents that support the safety related functions of the reactor and subcomponents whose failure could challenge these safety related functions. This scope is fully consistent with MRP-227 Rev. 1-A. The subject materials include reactor internals composed of austenitic stainless steels, nickel based alloys, and cast austenitic stainless steels (CASS). All materials are subject to the primary water environment. The program manages the aging effects of (1) cracking due to stress corrosion cracking (SCC), primary water stress corrosion cracking (PWSCC), irradiation-assisted stress corrosion cracking (IASCC), and fatigue; (2) loss of material due to wear; (3) loss of fracture toughness due to neutron irradiation embrittlement and thermal aging embrittlement; (4) changes in dimension due to void swelling; and (5) loss of preload due to thermal and irradiation-enhanced stress relaxation or creep. The program uses a leading indicator and sampling approach for inspections. In this approach, certain components which are considered to have a higher susceptibility or greater safety consequence are chosen as Primary inspection items. If the Primary inspections detect significant issues, then other components are subject to Expansion inspection. Some items are considered to be adequately managed through Existing Programs, including ASME Section XI B-N-3 (see below). The remaining reactor internals component items have no specified inspections under this AMP but could become subject to augmented inspection if indicated by operating experience.

ASME Section XI inservice inspections are performed to demonstrate the long-term integrity and continued functionality of the reactor internals. The MPS3 ISI Program is broadly described in an administrative procedure, Dominion Inservice Inspection Program (Reference 2). Specific requirements for inservice inspections are identified in the MPS3 ISI Program Manual1 (Reference 3). For the reactor internals, these include (1) inservice inspections performed in accordance with Examination Categories B-N-3 for all core support structures made accessible by removal of the reactor internals, and (2) augmented examinations not required by ASME Section XI.

The Inspection Plan for the reactor internals components governed by this AMP is summarized in the attached tables. The tables are based on a selection of the table line 1 MPS3 inservice inspections are currently performed to the 2013 Edition of ASME Section XI.

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LICENSE RENEWAL AGING MANAGEMENT PROGRAM items from MRP-227 Rev. 1-A, Sections 4 and 5. The selection includes all line items applicable to MPS3. There have been no modifications of the MRP-227 Rev. 1-A individual line items unless specifically noted.

In addition to the ISI Program, the monitoring and control of reactor coolant water chemistry ensures the long-term integrity and continued functionality of the reactor internals. The monitoring and control of primary chemistry is discussed in AMP MP-LR-4702, Chemistry Control for Primary Systems (Reference 4).

2.1 APPLICABILITY OF MRP-227 REV. 1-A MRP-227 Rev. 1-A, Section 2.4 provides a listing of assumptions used in developing its requirements. A brief reconciliation of MPS3 design, operation, and modifications, with respect to the assumptions is provided in the tables below. (Note that the statement of assumptions is abbreviated in the following table, however the reconciliation has been performed with respect to the full statement in MRP-227 Rev. 1-A Section 2.4.)

Table 2.1-1 MPS3 Reconciliation with MRP-227 Rev. 1-A Applicability Assumptions Assumption MPS3 Specifics Reconciliation 30 years of operation with high-leakage core loading patterns followed by implementation of a low-leakage fuel management strategy for the remaining 30 years of operation, as well as the average core power levels and proximity of active fuel to the upper core support plate satisfies limits described in MRP-227 Rev. 1-A Appendix B Except for the first fuel cycle, MPS3 has always operated with low leakage fuel loading patterns. Historical fuel loading patterns have been validated. Relevant results include:

MPS3 active fuel to upper core plate distance > 12.2 inches MPS3 average core power density <

124 W/cm3 MPS3 heat generation figure of merit, F < 68 W/cm3 MPS3 continues to validate these assumptions for each fuel cycle.

Assumption satisfied Base load operation for majority of plant operation MPS3 has always operated as a base load unit and has no plans to alter this practice.

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LICENSE RENEWAL AGING MANAGEMENT PROGRAM Table 2.1-1 MPS3 Reconciliation with MRP-227 Rev. 1-A Applicability Assumptions Assumption MPS3 Specifics Reconciliation No design changes beyond those identified in general industry guidance or recommended by the original vendors.

MPS3 is a Westinghouse design.

Significant design changes to MPS3 include:

Control rod guide tube (CRGT) split pin replacement with Type 316 pins in 2008 Partial replacement of flux thimble tubes 7% Stretch Power Uprate (implemented at the end of cycle 12 in 2008, no physical modifications to reactor internals)

MUR Power Uprate with no modifications of reactor internals Westinghouse has developed or evaluated these design changes.

Assumption satisfied The components and material class of each functional component are as listed in the latest revision of MRP-191.

Generically validated as part of MRP-191 Rev. 2.

The component materials for MPS3 were validated by Westinghouse Assumption satisfied MRP-227 Rev. 1-A Section 2.4 also requires that its assumptions regarding the effectiveness of existing programs, other than the ASME Section XI inspection program, be verified as acceptable to manage anticipated aging effects. Currently, the only such program for MPS3 is one to manage wear and potential for cracking of bottom mounted instrumentation neutron flux thimbles, established in response to NRC Bulletin 88-09. The current program consists of scheduled inspections, trending, and adjustment or replacement of individual flux thimbles as required. As noted above, this AMP is in compliance with industry guidance.

2.2 ITEMS SELECTED FOR INSPECTION Under ASME XI B-N-3, core support structures are specified for inspection, and unless a more specific delineation of items listed in MRP-227 Rev. 1-A applies, this Code scope governs. As provided by MRP-227 Rev. 1-A, the attached Existing Programs tables provide additional specificity of examination for certain items within the scope of the ASME XI. Under MRP-227 Rev. 1-A, the items selected for inspection are listed in the Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 4 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM attached tables for Primary and Expansion items. All selected table entries and associated notes are taken from the corresponding MRP-227 Rev. 1-A table, except as noted where subsequent Needed guidance per NEI 03-08 has supplemented certain table entries. The sampling and extent of examination for each item is listed in the Examination Coverage column. Inspection of Expansion items is required only when criteria listed in the Examination Acceptance and Expansion Criteria" invoke them.

2.3 INSPECTION SCHEDULE Visual inservice inspections are implemented in accordance with the schedule required for Category B-N-3, Removable Core Support Structures, of ASME Section XI, Subsection IWB. The inspection schedule of MRP-227 Rev. 1-A items is in accordance with MRP-227 Rev. 1-A as shown in the attached tables. Exceptions are noted where subsequent Needed guidance per NEI 03-08 has supplemented certain table entries. The tables refer to the start of the license renewal period, also known as the PEO. Per the MPS3 operating license, the units PEO begins at midnight, November 25, 2025. The second refueling outage after this milestone, by which time MRP-227 Rev. 1-A requires most baseline inspections to be complete, is anticipated to occur at 3R25 in the spring of 2028.

2.4 INSPECTION STANDARDS AND EXAMINATION ACCEPTANCE STANDARDS ASME XI visual Inservice Inspections are implemented by the ISI program in accordance with Category B-N-3, Removable Core Support Structures, of ASME Section XI, Subsection IWB. The examination acceptance standards for the visual examinations (VT-

3) of Category B-N-3 are summarized in IWB-3520.2, Visual Examination, VT-3 (Reference 1).

The visual inspections (VT-3, VT-1 and EVT-1) of items listed in the attached tables are performed in compliance with the industry standards established in MRP-228 (Reference 9). For certain bolting specified by MRP-227 Rev. 1-A, ultrasonic test (UT) examinations are performed. MRP-228 also defines the requirements for a Technical Justification of the UT examination and the definition of a relevant indication.

The examination acceptance standards for items inspected under MRP-227 Rev. 1-A are described in MRP-227 Rev. 1-A Section 5. Standards specific to MPS3 are excerpted in, Table 4, (MRP-227 Rev. 1-A Table 5-3) which is attached to this AMP. As noted, certain entries have been supplemented by subsequent Needed guidance per NEI 03-08.

All relevant conditions and relevant indications identified in both the ASME XI ISI Program and the MRP-227 Rev. 1-A inspections are entered into the corrective action program (Reference 6).

2.5 DISPOSITION OF INSPECTION RESULTS All adverse inspection results are entered into the corrective action program for disposition. Engineering evaluation methodologies used to disposition relevant Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 5 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM conditions identified by MRP-227 Rev. 1-A inspections are in accordance with WCAP-17096-NP-A Rev.3, including conditions imposed by the NRC (Reference 10). Evaluation methodologies for relevant conditions identified by the ASME XI ISI program are in accordance with ASME XI requirements or approved alternatives.

Any repair/replacement activities required as a result of disposition will be in accordance with ASME XI requirements or approved alternatives.

2.6 REPORTING OF INSPECTION RESULTS The results of inspection and dispositions required are reported to the MRP in accordance with MRP-227 Rev. 1-A Section 7.6. The results of the ASME XI inspections are reported to the NRC in accordance with ASME XI requirements.

2.7 DISCUSSION OF GENERIC ISSUES Generic issues that are directly addressed by MRP-227 Rev. 1-A program include void swelling, IASCC, and thermal and irradiation embrittlement of CASS. These effects are adequately managed by the requirements of MRP-227 Rev. 1-A, together with the associated SER Licensee Action Item. Detailed discussion of these issues is therefore removed from this section of the program.

2.8 ONGOING ACTIVITIES Ongoing activities include maintaining awareness of continuing industry activities that could affect the future requirements of this AMP. Activities also include an assessment of related industry experience. The industry efforts on reactor internals aging effects will be followed and the appropriate recommendations from these efforts will be implemented for MPS3.

Industry operating experience has been assessed and incorporated in MRP-227 Rev. 1-A as appropriate. A synopsis of such OE was included as Appendix A of the document.

The EPRI MRP has an ongoing program to gather and assess industry operating experience, including available experience from non-domestic reactors. A broad summary of industry experience is included in Section 3.10 of this AMP. Industry experience through spring 2023 has been considered in development of this AMP.

3 EVALUATION USING NUREG-1801 REV. 2 AND LR-ISG-2011-04, GENERIC AGING LESSONS LEARNED (GALL) REPORT ELEMENTS Note: The existing license renewal basis for MPS3 is GALL Rev. 0 (Reference 7). This version of the GALL was referenced in previous versions of this AMP. At the time of issuance of MRP-227-A, the NRC SE (Reference 11) of MRP-227-A requested that comparisons of the AMP be made to GALL Rev. 2 (Reference 15),Section XI.M16A, PWR Vessel Internals. Subsequently, the recommendations of GALL Rev. 2 were modified by LR-ISG-2011-04, (Reference 16), still with reference to MRP-227-A. For Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 6 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM subsequent license renewal (SLR) applications the NRC issued the GALL-SLR (Reference 17), which also referenced MRP-227-A. Approval of SLR has not been requested for MPS3, therefore, the SLR-related guidance is not applicable.

Additionally, the NRC SE (Reference 11) for MRP-227 Rev. 1-A concludes that the updated MRP guidance document is acceptable for referencing in LR applications but does not require a direct comparison with a GALL-like document. The SE also states in Section 3.6.8 that no licensee action item for submittal of the AMP or the inspection plan is required because each licensee with an approved license renewal has a commitment to submit the AMP and inspection plan two years prior to the PEO. Thus, LR-ISG-2011-04 is the most relevant document for describing the 10 element aging management program for the MPS3 reactor internals. The following subsections are patterned after the XI.M16A PWR VESSEL INTERNALS contained in Appendix A of LR-ISG-2011-04.

A comparison of the GALL Rev. 2, as supplemented by LR-ISG-2011-04, is contained in Section 4 of this AMP.

3.1 SCOPE The scope of the program includes all reactor internals components for MPS3. The scope of the program applies the methodology and guidance in the most recently NRC-endorsed version of MRP-227, MRP-227 Rev. 1-A, which provides augmented inspection and flaw evaluation methodology for assuring the functional integrity of safety-related internals in commercial operating U.S. pressurized water reactor (PWR) nuclear power plants designed by Babcock and Wilcox (B&W), Combustion Engineering (CE), and Westinghouse. The scope of components considered for inspection in MRP-227 Rev. 1-A include core support structures, those reactor internals components that serve an intended license renewal safety function pursuant to criteria in 10 CFR 54.4(a)(1), and other Reactor Vessel Internals (RVI) components whose failure could prevent satisfactory accomplishment of any of the functions identified in 10 CFR 54.4(a)(1)(i), (ii), or (iii). In addition, ASME Code,Section XI includes inspection requirements for PWR removable core support structures in Table IWB-2500-1, Examination Category B-N-3, which are in addition to any inspections that are implemented in accordance with MRP-227 Rev. 1-A.

The scope of the program does not include consumable items, such as fuel assemblies, reactivity control assemblies, and nuclear instrumentation. The scope of the program also does not include welded attachments to the internal surface of the reactor vessel because these components are considered to be ASME Code Class 1 appurtenances to the reactor vessel and are managed in accordance with the ISI Program: Systems, Components, and Supports (Reference 5).

Reactor vessel closure head penetration thermal sleeves are not within the scope of this AMP. However, wear of the sleeves and associated support flanges has been noted in industry operating experience and related guidance, References 23 and 24. MPS3 has performed inspections consistent with this guidance and maintains an on-going effort within the corrective actions program to monitor wear of these components.

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LICENSE RENEWAL AGING MANAGEMENT PROGRAM A review for applicability of MRP-227 Rev. 1-A has been performed in accordance with Section 2.4 of the guideline, and it has been determined that the baseline assumptions of MRP-227 Rev. 1-A are satisfied by the reactor internals inspection program for MPS3.

3.2 PREVENTIVE ACTIONS This AMP and MRP-227 Rev. 1-A rely on PWR water chemistry control to prevent or mitigate aging effects that can be induced by corrosive aging mechanisms (e.g., loss of material induced by general corrosion, pitting corrosion, crevice corrosion, or stress corrosion cracking [including its various forms of SCC, PWSCC, and IASCC]). Reactor coolant water chemistry is monitored and maintained in accordance with the Chemistry Control for Primary Systems AMP (Reference 4).

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

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

Specifically, the program implements the parameters monitored/inspected criteria consistent with the tables in MRP-227 Rev. 1-A for Westinghouse designs.

3.4 DETECTION OF AGING EFFECTS The inspection methods are defined and established in Section 4 of MRP-227 Rev. 1-A.

Standards for implementing the inspection methods are defined and established in MRP-228 Rev. 4. In all cases, well-established inspection methods are selected. These methods include volumetric UT examination methods for detecting flaws in bolting, and various visual (VT-3, VT-1, and EVT-1) examinations for detecting effects ranging from Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 8 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM general conditions to detection and sizing of surface-breaking discontinuities. Surface examinations may also be used as an alternative to visual examinations for detection and sizing of surface-breaking discontinuities.

Cracking caused by SCC, IASCC, and fatigue is monitored/inspected in this AMP by either VT-1 or EVT-1 examination (for internals other than bolting) or by volumetric UT examination (bolting). VT-3 visual methods are utilized for the detection of cracking in non-redundant RVI components only when the flaw tolerance of the component has been shown to be tolerant of easily detected large flaws, even under reduced fracture toughness conditions. VT-3 visual methods are acceptable for the detection of cracking in redundant RVI components (e.g., redundant bolts or pins used to secure a fastened RVI assembly). Eddy current (ET) surface examinations or UT volumetric examinations may also be used in lieu of visual examinations for certain internals components if they have been specified or optionally permitted by MRP-227 Rev. 1-A or related industry guidance.

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

The program adopts the guidance in MRP-227 Rev. 1-A for defining the Expansion Criteria that need to be applied to the inspection findings of Primary components and for expanding the examinations to include additional Expansion components. RVI component inspections are performed consistent with the inspection frequency and sampling bases for Primary components, Existing Programs components, and Expansion components in MRP-227 Rev. 1-A.

In some cases (as defined in MRP-227 Rev. 1-A), physical measurements are used as supplemental techniques to manage for the gross effects of wear, loss of preload due to stress relaxation, or for changes in dimensions due to void swelling or distortion. Qualified photogrammetric methods for physical measurement may also be used, such as measurement of control rod guide tube guide card wear.

Inspection coverages for Primary and Expansion RVI components are implemented consistent with the notes to Tables 4-3 and 4-6 of MRP-227 Rev. 1-A.

3.5 MONITORING AND TRENDING The methods for monitoring, recording, evaluating, and trending the data that result from the programs inspections are given in Section 6 of MRP-227 Rev. 1-A. Flaw evaluation methods including recommendations for flaw depth sizing and for crack growth determinations as well as for performing applicable limit load, linear elastic and elastic-plastic fracture analyses of relevant flaw indications, are defined in MRP-227 Rev. 1-A and WCAP-17096-NP-A Rev. 3. The examination and reexaminations that are implemented in accordance with MRP-227 Rev. 1-A, together with the criteria specified in MRP-228 Rev. 4 for inspection methodologies, inspection procedures, and inspection Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 9 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM personnel, provide timely detection, reporting, and implementation of corrective actions for aging effects and mechanisms managed by the program.

As required by reference to WCAP-17096-NP-A Rev. 3, the program applies applicable fracture toughness properties, including reductions for thermal aging or neutron embrittlement, in the flaw evaluations of the components in cases where cracking is detected in a RVI component and is extensive enough to warrant a supplemental flaw growth or flaw tolerance evaluation For singly-represented components, the program includes criteria to evaluate the aging effects in the inaccessible portions of the components and the resulting impact on the intended function(s) of the components. For redundant components (such as redundant bolts, screws, pins, keys, or fasteners, some of which are accessible to inspection and some of which are not accessible to inspection), the program includes criteria to evaluate the aging effects in the population of components that are inaccessible to the applicable inspection technique, and the resulting impact on the intended function(s) of the assembly containing the components.

3.6 ACCEPTANCE CRITERIA Section 5 of MRP-227 Rev. 1-A, which includes Table 5-3 for Westinghouse-designed RVIs, provides the specific examination and flaw evaluation acceptance criteria for the Primary and Expansion RVI component examination methods. For RVI components addressed by examinations performed in accordance with the ASME Code,Section XI, the acceptance criteria in IWB-3500 are applicable. For RVI components covered by other Existing Programs, the acceptance criteria are described within the reference document.

As applicable, the program establishes acceptance criteria for physical measurement monitoring methods that are credited for aging management of particular RVI components.

3.7 CORRECTIVE ACTIONS Any detected conditions that do not satisfy the examination acceptance criteria are required to be dispositioned through the plant corrective action program, which may require repair, replacement, or analytical evaluation for continued service until the next inspection. The disposition will ensure that design basis functions of the reactor internals components will continue to be fulfilled for all licensing basis loads and events. The implementation of the guidance in MRP-227 Rev. 1-A, plus the implementation of any ASME Code requirements, provides an acceptable level of aging management of safety-related components addressed in accordance with the corrective actions of 10 CFR Part 50, Appendix B or its equivalent, as applicable.

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

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LICENSE RENEWAL AGING MANAGEMENT PROGRAM 3.8 CONFIRMATION PROCESS Site quality assurance procedures, review and approval processes, and administrative controls are implemented in accordance with the recommendations of NEI 03-08 and the requirements of 10 CFR Part 50, Appendix B, or their equivalent, as applicable. The implementation of the guidance in MRP-227 Rev. 1-A, in conjunction with NEI 03-08 and other guidance documents, reports, or methodologies referenced in this AMP provides an acceptable level of quality and an acceptable basis for confirming the quality of inspections, flaw evaluations, and corrective actions.

3.9 ADMINISTRATIVE CONTROLS The administrative controls (References 2, 12-14) for these types of programs, including their implementing procedures, and review and approval processes, are implemented in accordance with the recommended industry guidelines and criteria in NEI 03-08, and are under existing site 10 CFR 50 Appendix B Quality Assurance Programs, or their equivalent, as applicable. The evaluation in Section 3.5 of the NRCs SE on MRP-227 Rev. 1-A provides the basis for endorsing NEI 03-08. This includes endorsement of the criteria in NEI 03-08 for notifying the NRC of any deviation from the inspection and evaluation (I&E) methodology in MRP-227 Rev. 1-A and justifying the deviation no later than 45 days after its approval by a licensee executive.

3.10 OPERATING EXPERIENCE MPS3 is directed by corporate procedures to identify and review relevant operating experience (OE) for reactor internals [14]. The procedures have provisions to modify the program as required to consider future operating experience events. The following summarizes both MPS3-specific and external operating experience.

The review and assessment of relevant OE for its impacts on this AMP, including implementing procedures, are governed by NEI 03-08 protocols. Consistent with MRP-227 Rev. 1-A, the reporting of inspection results and operating experience is treated as a Needed category item under the implementation of NEI 03-08. The compiled results are periodically published in EPRI document MRP-219 (Reference 18). The latest available version of this document, which is Revision 12, has been reviewed for OE applicable to MPS3. The specified examinations in the Attachment 1 tables will be capable of detecting any significant occurrences of these effects at MPS3.

MPS3 is currently in the fourth 10-year interval of its ASME Section XI ISI program.

Examinations of reactor internals were completed as part of the third ten-year interval of the program in 2016, which included augmented examination criteria listed in Reference

25. These examinations did not identify aging-related degradation, detect cases of cracking of the reactor internals components, or find any cases of excessive wear of mating surfaces. In addition, the inspection did not detect any issues with core barrel radial support clevis insert bolting. For the present program, OE regarding clevis insert bolting failures is addressed by the Needed guidance (Reference 26) noted for existing program item W14 in Table 3.

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LICENSE RENEWAL AGING MANAGEMENT PROGRAM As noted in Section 2.1, the original CRGT split pins of a nickel based alloy fabrication were replaced in 2008, with a Type 316 stainless steel pin design that reduces stress concentrations. No failures of the removed original split pins were noted.

The condition of incore instrumentation flux thimbles (Table 3, item W13) have been monitored by ECT inspections at periodic intervals. Ten worn flux thimbles have been replaced with chrome treated thimbles that were expected to be less susceptible to wear.

Subsequent inspections have identified reduced wear of the replacement items, and the wear rate of the non-replaced flux thimble tubes has been moderate. The flux thimbles will continue to be monitored at periodic intervals by this existing program.

Industry operating experience has identified reactor vessel head penetration CRDM thermal sleeves as items subject to wear. In response to Needed industry guidance (Reference 23), the MPS3 thermal sleeves (including their supporting flange) have been inspected for excessive wear, with acceptable results. A reinspection is planned for the last refueling outage prior to entering the PEO. Industry operating experience has also identified that thermal sleeve designs at certain plants may be susceptible to cracking (Reference 24), however MPS3 was not included on that list.

In 2019, MPS3 performed the baseline inspection of the CRGT guide cards in accordance with Attachment 1, Table 1, item W1. The evaluation of inspection results was performed based on Reference 22. The wear was moderate and all CRGT met the acceptance criteria. Reference 21 includes industry CRGT guide card inspection results, including those for MPS3. Future reinspections for MPS3 will be scheduled in accordance with the referenced table item W1.

From 1999 through 2001, inspections of baffle-former bolting in Westinghouse designs were performed at Point Beach Unit 2, Ginna, and Farley Units 1 and 2. Inspections at Point Beach Unit 2 and Ginna indicated the presence of only a small number of non-functional (i.e., cracked) bolts. Ultrasonic inspection at Farley determined that the bolting was not defective, but replacements were installed as a pre-emptive measure. More recent baffle bolting inspections performed per MRP requirements have shown that 4 loop downflow plant designs are particularly susceptible to significant baffle bolt failures, including incidences of clustering of failed bolts. As specified in Table 1 item W6, the baffle-former bolts will be examined and compared with acceptance criteria. Referenced notes 8 and 9 for Table 1, which were taken from MRP-227 Rev. 1-A, establish criteria for accelerated degradation and limits to reinspection intervals. Note 9 requires that justification for alternate reinspection intervals shall be submitted to the NRC. The Needed provision of MRP-227 Rev. 1-A, Section 7.5 also requires use of NRC approved evaluation methods, and MPS3 will comply with these provisions.

At least one of the 4 loop downflow plants with significant baffle former bolting failures has also identified cracking and failure of thermal shield flexures and failure of thermal shield support block bolting. MPS3 is designed with neutron panels in lieu of a thermal shield, so this thermal shield OE is not applicable to MPS3.

Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 12 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM In recent years there has been OE regarding core barrel welds with detected cracking.

This has resulted in Needed guidance per NEI 03-08 (Reference 27), which provided updates to the MRP-227 Rev. 1-A program table items for core barrel welds. These updates have been incorporated in the attached program tables for MPS3.

Finally, License Renewal Interim Staff Guidance LR-ISG-2011-04, Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors (Reference 16), has been reviewed for relevant operating experience. There were no required changes to this AMP as a result of this review.

As required by applicable corporate procedures (References 12-14), industry operating experience will continue to be monitored and this AMP will be updated as needed.

4 AGING MANAGEMENT PROGRAM COMPARISON: MILLSTONE PROGRAM AND NUREG-1801 (GALL REPORT)

The license renewal application and subsequent approval for Millstone Power Station was developed with reference to Chapter XI of Revision 0 of the GALL Report (Reference 7).

Subsequently, Revision 1 of the report was issued in 2005, and Revision 2 of the report (Reference 15) was issued to accommodate the inspection strategies developed in MRP-227-A, which incorporated changes identified in the NRC SE (published within MRP-227-A). The SE required submittal of aging management programs, in accordance with commitments for plants that have received renewed licenses, with reference to the GALL Report, Revision 2. The later revision resolves many of the issues that were outstanding in Revision 0 of the GALL. A subsequent interim staff guidance document, LR-ISG-2011-04, Reference 16, updated the expectations for the content of the aging management program for reactor internals, XI.M16A. Thus, the program description of Section 3 in this attachment was written to be in alignment with Revision 2 of the GALL Report as updated by LR-ISG-2011-04 for ease of review.

The MPS3 ISI Program: Reactor Vessel Internals complies without deviation to MRP-227 Rev. 1-A (Reference 8) and related Needed requirements subsequently issued under NEI 03-08. The program is compatible with the aging management programs described in Chapter XI of GALL Report, Revision 2 as updated by LR-ISG-2011-04. The specific GALL sections and corresponding titles are as follows:

Section XI.M16A, PWR Vessel Internals Exceptions to the GALL (Section XI.M16A)

None.

Enhancements to GALL None Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 13 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM 5

SUMMARY

LR-4711, Inservice Inspection: Reactor Vessel Internals AMP ensures that the effects of aging associated with the in-scope components will be adequately managed. As a result, there is reasonable assurance that aging effects will not prevent in-scope components from performing their intended licensing basis functions during the PEO.

Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 14 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM 6

REFERENCES

1. Rules for Inservice Inspection of Nuclear Power Plant Components,Section XI, American Society of Mechanical Engineers, New York, NY.

2. ER-AA-ISI-100, Dominion Inservice Inspection Program, Rev. 13, Dominion.

3. U3-24-ISI-PRG-Interval 4, Millstone Unit 3 Inservice Inspection Program Manual, Fourth Ten-Year Interval, Revision 2, Program Manual, Millstone Unit 3, Dominion.

4. ETE-MP-2013-1041, Rev. 0, Chemistry Control for Primary Systems, License Renewal Aging Management Program (MP-LR-3702/MP-LR-4702), Dominion Engineering Technical Evaluation.

5. ETE-MP-2013-1040, Rev.1. Inservice Inspection Program: Systems, Components, and Supports; License Renewal Aging Management Program (MP-LR-3701/MP-LR-4701), Dominion Engineering technical Evaluation.

6. PI-AA-200, Rev. 41, Corrective Action, Nuclear Fleet Administrative Procedure, Dominion Nuclear Connecticut.

7. NUREG-1801, Generic Aging Lessons Learned (GALL) Report, US Nuclear Regulatory Commission, July 2001.

8. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A), Electric Power Research Institute, Palo Alto, CA dated December 2019. 3002017168

9. Materials Reliability Program: Inspection Standard for Pressurized Water Reactor Internals - 2020 Update (MRP-228, Rev. 4). EPRI, Palo Alto, CA, dated December 2020. 3002018245

10. Westinghouse WCAP-17096-NP-A Rev. 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements, August 2023, Westinghouse Electric Company LLC, Pittsburgh, PA.

11. NRC SE: Final Safety Evaluation for Electric Power Research Institute Topical Report, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluations Guidelines (CAC NO.MF7223; EPID L-2016-TP-0001) dated 4/25/2019 [included within MRP-227 Rev. 1-A].

12. ER-AA-MAT-10, Reactor Coolant System Materials Degradation Management Program, Rev. 10, Dominion.

13. ER-AA-RII-10, Fleet Reactor Internals Inspection Program Description, Rev. 9, Dominion.

14. ER-AA-RII-101, Fleet Reactor Internals Inspection Program, Rev. 10, Dominion.

15. NUREG-1801 Rev. 2, Generic Aging Lessons Learned (GALL) Report, US Nuclear Regulatory Commission, December 2010 Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 15 of 32

LICENSE RENEWAL AGING MANAGEMENT PROGRAM

16. LR-ISG-2011-04, License Renewal Interim Staff Guidance LR-ISG-2011-04:

Updated Aging Management Criteria for Reactor Vessel Internal Components for Pressurized Water Reactors, May 28, 2013 (Interim Staff Guidance to NUREG-1801 Rev. 2), U.S. Nuclear Regulatory Commission.

17. NUREG-2191, Vol. 1 and 2, Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report, US Nuclear Regulatory Commission, July 2017

18. Materials Reliability Program: Inspection Data Survey Results (MRP-219, Revision12). Electric Power Research Institute, Palo Alto, CA: 2018.

3002007933.

19. MRP 2019-032, PWR Reactor Internals Inspection and Evaluation Guidelines (MRP-227 Revision 1-A), letter dated 12/10/2019

20. MRP 2020-011, Notification of Recent Operating Experience (OE) Related to Displaced PWR Clevis Insert Assembly, EPRI letter dated 4/28/2020

21. MRP 2022-017, 2022 Biennial Report of Recent MRP-227-A Reactor Internals Inspection Results, EPRI letter dated 9/30/2022

22. WCAP-17451-P Rev. 2, Reactor Internals Guide Tube Wear - Westinghouse Domestic Fleet Operational Projections, November 2018, Westinghouse Electric Company LLC

23. MRP 2018-033, Transmittal of NEI-03-08 Needed Interim Guidance for PWR CRDM Thermal Sleeve Wear, EPRI Letter dated 9/5/2018

24. OG-20-113 NEI 03-08 Needed and Good Practice Guidance: Thermal Sleeve Cross-Sectional Failure - Westinghouse Nuclear Safety Advisory Letter NSAL-20-1, 4/13/2020, Westinghouse Electric Company LLC [ADAMS Acc. No.

ML21287A184]_

25. TB-14-5, Reactor Internals Lower Radial Support Clevis Insert Cap Screw Degradation, 8/25/2014, Westinghouse Electric Company LLC

26. OG-21-160, NEI 03-08 Needed Guidance: PWR Lower Radial Support Clevis Insert X-750 Bolt Inspection Requirements, PWROG Owners Group, 9/1/2021

27. MRP 2023-005, MRP-227 NEI 03-08 Needed Interim Guidance for WEC/CE Core Barrel Inspections, EPRI Letter dated 5/19/2023 Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 16 of 32

Aging Management Program, Table 1 1

ATTACHMENT 1 - MPS3 MRP-227 REV. 1-A TABLES MILLSTONE UNIT 3 (MPS3) REACTOR INTERNALS INSPECTION PLAN IN ACCORDANCE WITH MRP-227 REV. 1-A The following tables excerpted from MRP-227 Rev. 1-A (MRP-227) are applicable to Millstone Unit 3 (MPS3). The noted references to figures and notes are to the MRP-227 document. MRP-228 Rev. 4 (MRP-228) is also applicable to the required inspections. Certain table entries, as noted, have been updated to include additional requirements currently in effect due to issuance of Needed guidance per NEI 03-08. These updates affect inspection items W1, updated to note a completed inspection per Reference 22; W3, W3a, W3.2, W4, updated per Reference 27; and W14, updated per Reference 26.

, Table 1 - MRP-227 Table 4-3 Inspection of listed components is required in accordance with MRP-227 and MRP-228.

Note: The baseline inspection of Primary item W.1, Control Rod Guide Tube Assembly, was completed in April 2019 with acceptable results, in accordance with WCAP-17451-P Rev. 2, Ref. 22. Guide card wear management for MPS3 is subject to ongoing review and adjustment within the requirements and limits imposed by approved industry guidance.

, Table 2 - MRP-227 Table 4-6, Westinghouse Plants Expansion Components Inspection of listed components is required in accordance with MRP-227 and MRP-228 when indicated by inspection results of Primary Components

, Table 3 - MRP-227 Table 4-9, Westinghouse Plants Existing Programs Components Inspection of all listed components is required in accordance with the referenced existing program. For references to the ASME Section XI ISI program, that program governs and this AMP specifies augmented requirements or is for reference only.

, Table 4 - MRP-227 Table 5-3, Westinghouse Plants Examination Acceptance and Expansion Criteria Contains the examination expansion criteria for the results of the Primary inspection components, and examination acceptance standards for Primary and Expansion components.

Note: MRP-227 Rev. 1-A inspection plan Table line items that are not applicable to MPS3 are marked NA for MPS3 and are also greyed out.

Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 17 of 32

Aging Management Program, Table 1 Table43 WestinghousePlantsPrimaryComponents 2

Primary Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage

W1.Control Rod Guide Tube Assembly Guideplates(cards)

Allplants (SeeWEC NSAL171)

LossofMaterial (Wear)

None

PertherequirementsofWCAP 17451P, including subsequent examinations(Notes5,12).

Examinationcoverageperthe requirementsofWCAP17451P, Revision1(Notes5,12)SeeFigure 411

W2.Control Rod Guide Tube Assembly Lowerflangewelds

Allplants

Cracking(SCC,Fatigue)

AgingManagement(IEandTE)

W2.1.RemainingCRGT assemblylowerflange welds

W2.2.BMIcolumnbodies Enhancedvisual(EVT1)examination todeterminethepresenceofcracklike surfaceflawsinflangeweldsnolater than2refuelingoutagesfromthe beginningofthelicenserenewal periodandsubsequentexamination onatenyear interval.

100%ofouter(accessible)CRGT lowerflangeweldsurfacesand 0.25inchoftheadjacentbase metalontheindividualperiphery CRGTassemblies.

(Note2)

SeeFigure412.

W3.Core Barrel Assembly UpperflangeWeld(UFW)

Allplants

Cracking(SCC)

W3.3.lowerflangeweld (LFW),W3.2.Upperaxial welds(UAW),and W3.4.Lowersupport forgingorcasting

Enhancedvisual(EVT1),eddycurrent (ET),orvolumetric(UT)examination, nolaterthan2refuelingoutagesfrom thebeginningofthelicenserenewal periodandsubsequentexamination onatenyearinterval.

100%oftheaccessibleweld lengthofbothsurfaces(ID surfaceandODsurface)ofthe UFWand3/4ofadjacentbase metalshallbeexamined.

(Note11)

SeeFigure413.

W3a.Core Barrel Assembly Uppergirthweld(UGW)

(PromotedbacktoPrimary)

Allplants Cracking(SCC)

W3.2.Uppercorebarrel upperaxialwelds(UAW)

W3.3.Lowerflangeweld (LFW),W3.4.Lower supportforgingorcasting Enhancedvisual(EVT1),

volumetric(UT),orsurface(ET) examination,nolaterthan2 refuelingoutagesfromthe beginningofthelicenserenewal periodandsubsequent examinationonatenyear interval.

100%oftheaccessibleweld lengthofbothsurfaces(ID surfaceandODsurface)ofthe UGWand3/4ofadjacentbase metalshallbeexamined.IfUTis performed,itneedonlybe completedfromonesurface, eitherIDorOD.(Note11).

SeeFigure413.

W4.Core Barrel Assembly Lowergirthweld(LGW)

Allplants

Cracking(SCC,IASCC),Aging Management(IE)

W4.1.Uppercoreplate, W4.4.Lowersupport columnbodies(cast,non cast),

W4.2.Middleaxialwelds (MAW),

W4.3.Loweraxialwelds (LAW)

Enhancedvisual(EVT1)examination, volumetric(UT),orsurface(ET)no laterthan2refuelingoutagesfrom thebeginningofthelicenserenewal periodandsubsequentexamination onatenyearinterval.

100%oftheaccessibleweld lengthoftheODsurfaceofthe LGWand3/4ofadjacentbase metalshallbeexamined.(ID surfaceisinaccessiblefor visual/ETbasedsurfaceexams duetobaffleformerassembly).

UTisperformedfromOD surface.

(Note6)

SeeFigure413.

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Aging Management Program, Table 1 Table43(continued)

WestinghousePlantsPrimaryComponents 3

Primary Item Applicability Effect (Mechanism)

Expansion Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage

W5.Baffle-Former Assembly Baffleedgebolts

NAforMPS3

Allplantswith baffleedge bolts Cracking(IASCC,Fatigue)that resultsin Lostorbrokenlocking devices Failedormissing bolts ProtrusionofboltheadsAging Management(IEandISR)(Note 4)

None

Visual(VT3)examination,with baselineexaminationbetween20 and40EFPYandsubsequent examinationsonatenyearinterval.

Boltsandlockingdevicesonhigh fluenceseams.100%of componentsaccessiblefromcore side.

SeeFigure414.

W6.Baffle-Former Assembly Baffleformerbolts (Note7)

Allplants (SeeWEC NSAL161)

Cracking(IASCC,Fatigue)

AgingManagement(IEandISR)

(Note4)

W6.2.Lowersupport columnbolts, W6.1.Barrelformerbolts Baselinevolumetric(UT)examination intervalisdependentontheplant design(Note8).Subsequent examinationisdependentontheplant designandtheresultsofthebaseline inspection(Note9).

100%ofaccessible bolts.(Note3)

SeeFigure415.

W7.Baffle-Former Assembly Assembly(Includes:Baffle plates,baffleedgebolts, cornerbolts,andindirect effectsofvoidswellingin formerplates)

Allplants Distortion(VoidSwelling),or Cracking(IASCC)thatresultsin Abnormalinteractionwithfuel assemblies Gapsbetweenplates Verticaldisplacementofbaffle plates Brokenordamagededgebolts

None

Visual(VT3)examinationtocheckfor evidenceofdistortion,withbaseline examinationbetween20and40 EFPYandsubsequentexaminationson atenyearinterval.

Coresidesurface:

Highfluencebafflejoints Topandbottomedgeof baffleplates Boltsandlockingdevices

SeeFigure416.

W8.Alignment and Interfacing Components Internalsholddownspring NAforMPS3

Allplantswith 304stainless steelhold downsprings

Distortion(LossofLoadduetoStress Relaxation)

None Directmeasurementofspringheight withinthreecyclesofthebeginningof (beforeorafter)thelicenserenewal period.Ifthefirstsetofmeasurements isnotsufficienttoassessremaining life,additionalspringheight measurementswillberequired.

Measurementsshouldbetakenat severalpointsaroundthe circumferenceofthespring,witha statisticallyadequatenumberof measurementsateachpointto minimizeuncertainty.

SeeFigure417.

W9.Thermal Shield Assembly Thermalshieldflexures NAforMPS3 All plants with thermal shields (SeeWEC TB195)

Cracking(Fatigue)orLossof Material(Wear)thatresultsin thermalshieldflexuresexcessive wear,fracture,orcomplete separation

None Visual(VT3)nolaterthan2refueling outagesfromthebeginningofthe licenserenewalperiod.Subsequent examinationsonatenyearinterval.

100%ofaccessiblesurfacesof 100%ofthermalshieldflexures.

(Note10)

SeeFigure418.

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Aging Management Program, Table 1 Table43(continued)

WestinghousePlantsPrimaryComponents 4

NotestoTable43:

1.

ExaminationacceptancecriteriaandexpansioncriteriafortheWestinghousecomponentsareinTable53.

2.

Aminimumof75%ofthetotalidentifiedsamplepopulationmustbeexamined.

3.

Aminimumof75%ofthetotalboltpopulation(examined+unexamined),includingcoverageconsistentwiththeExpansioncriteriainTable53,mustbeexaminedforinspection credit.

4.

Voidswellingeffectsonthiscomponentaremanagedthroughmanagementofvoidswellingontheentirebaffleformerassembly.

5.

InWCAP17451PthebaselineexaminationschedulehasbeenadjustedforvariousCRGTdesigns,theextentofindividualCRGTexaminationmodified,andflexiblesubsequent examinationregimenscorrelatingtoinitialbaselinesamplesize,accuracyofwearestimation,andexaminationresults.Initialinspectionpriortothelicenserenewalperiodmaybe required.UseWCAP17451P[37],Revision1,includingthemodifiedrequirementsduetotheinterimguidanceprovidedinEPRIletterMRP2018007dated3/7/2018[47]and PWROGletterOG1846dated2/20/2018[38].

6.

Examinationcoveragerequiresaminimumof50%ofthelengthoftheODoftheweldbeingexamined.

7.

Baffleformerboltinspectionincludesinspectionofthecornerplateboltswhenapplicable.

8.

InaccordancewithMRP2017009[39]andMRP2017010[41],Tier1plantsaretoperformthebaselineUTexaminationby20EFPYorduringthenextrefuelingoutageafterMarch 1,2016.PerMRP2017009[39],Tier2plantsaretoperformthebaselineUTexaminationatnolaterthan30EFPY(initialTier2plantbaselineUTexamsperformedpriorto1/1/2018 areacceptable).AllotherremainingplantsaretoperformthebaselineUTexaminationatnolaterthan35EFPY.

9.

ReexaminationperiodsshallbedeterminedbyplantspecificevaluationpertheMRP227NeededRequirement7.5asdocumentedanddispositionedintheownersplantcorrective actionprogram.IfatypicaloraggressivebaffleformerboltdegradationasdefinedinMRP2017009[39](i.e.,3%ofbaffleformerboltswithUTorvisualindicationsorclustering*

fordownflowplantsand5%ofbaffleformerboltswithUTorvisualindicationsorclustering*forupflowplants)isobserved,theinterimguidance(MRP2016021[40]andMRP 2017009[39])provideslimitationstothepermittedreinspectioninterval(nottoexceed6yearsmaximum)unlessfurtherevaluationisperformedtojustifyalongerinterval(See Applicant/LicenseeActionItem1intheNRCSEforevaluationsubmittalrequirements[35]).Ifevaluationjustifiesalongerreinspectioninterval,itisnotpermittedtoexceed10years.

• ClusteringisdefinedperNSAL161Rev.1[42]asthreeormoreadjacentdefectivebaffleformerboltsormorethan40%defectivebaffleformerboltsonthesamebaffleplate.

UntestableboltsshouldbereviewedonaplantspecificbasisconsistentwithWCAP17096NPAfordeterminationiftheseshouldbeconsideredwhenevaluatingclustering.

10. SeeWestinghouseTechnicalBulletinTB195dated10/9/2019andMRP2019017dated5/31/2019foradditionaldetailsoninspectionrecommendations.

11. Examinationcoveragerequiresaminimumof75%oftheweldlengthofboththeIDandtheODoftheweldbeingexamined.

12. (MPS3specific)BaselineexaminationsoftheCRGTguidecardswerecompletedin2019andwereacceptableforgreaterthan10EFPY.Subsequentexaminationsandevaluationswill begovernedbytherequirementsofWCAP17451Rev.2,consistentwithWCAP17096Rev.3.

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Aging Management Program, Table 2 Table46 WestinghousePlantsExpansionComponents 5

Expansion Item

Applicability

Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage Control Rod Guide Tube Assembly W2.1.RemainingCRGT lowerflangewelds Allplants Cracking(SCC, Fatigue)

AgingManagement(IE andTE)

W2.CRGTLower FlangeWelds Enhancedvisual(EVT1)examinationto determinethepresenceofcracklike surfaceflawsinflangewelds.

Subsequentexaminationonatenyear interval.

Aminimumof75%oftheCRGT assemblylowerflangeweldsurfaces and0.25inchoftheadjacentbase metalfortheflangeweldsnot inspectedundertheprimarylink.

Bottom Mounted Instrumentation System W2.2.Bottommounted instrumentation(BMI) columnbodies Allplants Cracking(Fatigue) includingthedetection ofcompletelyfractured columnbodies AgingManagement(IE)

W2.CRGTLower FlangeWelds Visual(VT3)examination.Reinspection every10yearsfollowinginitial inspection.

100%ofBMIcolumnbodiesforwhich difficultyisdetectedduringfluxthimble insertion/withdrawal.

SeeFigure424.

Core Barrel Assembly W3.2.UpperAxialWeld (UAW)

Allplants Cracking(SCC)

W3.UpperCore BarrelFlange Weld(UFW)

W3a.Uppergirth weld(UGW)

Enhancedsurfacevisual(EVT1),eddy current(ET),orvolumetric(UT) examination.Reinspectionevery10 yearsfollowinginitialinspection.

100%oftheaccessibleweldlengthof bothsurfaces(IDsurfaceandOD surface)oftheUAWand3/4of adjacentbasemetalshallbe examined.IfUTisperformed,itneed onlybecompletedfromonesurface, eitherIDorOD.(Note2).

SeeFigure413.

Core Barrel Assembly W3.3.LowerFlangeWeld (LFW)

Allplants

Cracking(SCC)

W3.UpperCore BarrelFlange Weld(UFW)

Enhanced visual (EVT1) examination.

Reinspection every 10 years following initialinspection.

100%oftheaccessibleweldlengthof theODsurfaceoftheLFWand3/4of adjacentbasemetalshallbe examined (Note5).

SeeFigure413.

Lower Internals Assembly W3.4.Lowersupport forgingorcastings

Allplants Cracking(SCC)

AgingManagement(TE inCasting)Ref.[34]

W3.UpperCoreBarrel FlangeWeld(UFW)

Visual(VT3)examination.

Reinspectionevery10years followinginitialinspection.

Minimumof25%ofbottom(noncore side)surface(Note3).

SeeFigure420.

Upper Internals Assembly W4.1.Uppercoreplate

Allplants Cracking(Fatigue),

Wear,Aging Management(IE)

W4.LowerGirth Weld(LGW)

Visual(VT3)examination.

Reinspectionevery10years followinginitialinspection.

Minimumof25%ofcoresidesurfaces (Note3).

SeeFigure419.

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Aging Management Program, Table 2 Table46(continued)

WestinghousePlantsExpansionComponents 6

Expansion Item

Applicability

Effect (Mechanism)

Primary Link (Note 1)

Examination Method/Frequency (Note 1)

Examination Coverage

Core Barrel Assembly W4.2.MiddleAxialWelds (MAW)andW4.3.Lower AxialWelds(LAW)

Allplants

Cracking(SCC,IASCC)

AgingManagement(IE)

W4.LowerGirth Weld(LGW)

Enhancedvisual(EVT1) examination.

Reinspectionevery10years followinginitialinspection 100%oftheaccessibleweldlength oftheODoftheMAWandLAWand 3/4 ofadjacentbasemetalshallbe examined(Notes5and6).

SeeFigure413.

Lower Support Assembly W4.4.Lowersupport columnbodies(bothcast andnoncast)

Allplants

Cracking (IASCC)

AgingManagement(IE)

W4.LowerGirth Weld(LGW)

Visual(VT3)examination.

Reinspectionevery10years followinginitialinspection.

25%ofthetotalnumberofcolumn assemblies(bothvisibleandnonvisible fromabovethelowercoreplate)using aVT3examinationfromabovethe lowercoreplate.Theinspection coveragemustbeevenlydistributed acrossthepopulationofcolumn assemblies (Notes3and4).

SeeFigure423.

Core Barrel Assembly W6.1.Barrelformerbolts

Allplants

Cracking(IASCC, Fatigue)

AgingManagement(IE, VoidSwellingandISR)

W6.Baffleformer bolts(alsoreferto MRP2018002)

Volumetric (UT) examination.

Reinspection every 10 years followinginitialinspection.

100%ofaccessiblebarrelformerbolts (Minimumof75%ofthetotal population).Accessibilitymaybe limitedbypresenceofthermalshield orneutronpads.

SeeFigure421.

Lower Support Assembly W6.2.Lowersupport columnbolts

Allplants Cracking (IASCC, Fatigue)

AgingManagement(IE andISR)

W6.Baffleformerbolts

Volumetric (UT) examination.

Reinspection every 10 years followinginitialinspection.

100%ofaccessibleLSCbolts (Minimumof75%ofthetotal population)orassupportedbyplant specificjustification.

SeeFigure422

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Aging Management Program, Table 2 Table46(continued)

WestinghousePlantsExpansionComponents 7

NotestoTable46:

1. ExaminationacceptancecriteriaandexpansioncriteriafortheWestinghousecomponentsareinTable53.

2. Examinationcoveragerequiresaminimumof75%oftheweldlengthofboththeIDandtheODoftheweldbeingexamined.

3. Thestatedminimumcoveragerequirementistheminimumifnosignificantindicationsarefound.However,theExaminationAcceptancecriteriainSection5requirethatadditional coveragemustbeachievedinthesameoutageifsignificantflawsarefound.Thiscontingencyshouldbeconsideredforinspectionplanningpurposes.

4. Justificationthatadequatedistributionoftheinspectioncoveragehasbeenachievedcanbebasedongeometricorlayoutarguments.Possibleexamplesinclude,butarenotlimited to,inspectionofallcolumnassembliesinonequadrantofthelowercoreplate(basedontheazimuthalsymmetryoftheplate)orinspectingeveryfourthcolumnacrosstheentireplate.

5. Aminimumcoverageof75%oftheweldlengthonthesurfacebeingexaminedshallbeachieved;however,forweldswithlimitedaccess(Note6),aminimumexaminationcoverage of50%oftheweldlengthonthesurfacebeingexaminedshallbeachieved.

6. AccessibilitytotheMAWandLAWmaybelimitedbythethermalshieldorneutronpanels-nodisassemblytoachievehigherweldlengthcoverageisrequired.

Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 23 of 32

Aging Management Program, Table 3 Table49 WestinghousePlantsExistingProgramsComponents 8

Item

Applicability Effect (Mechanism)

Reference

Examination Method Examination Coverage W10.Core Barrel Assembly Corebarrelflange Allplants Lossofmaterial (Wear)

ASMECode SectionXI Visual(VT3)examto determinegeneralcondition forexcessivewear.

Allaccessiblesurfacesat specifiedfrequency.

W11.Upper Internals Assembly Uppersupportringorskirt Allplants Cracking(SCC, Fatigue)

ASMECode SectionXI Visual(VT3)examination.

Allaccessiblesurfacesat specifiedfrequency.

W12a.Lower Internals Allplants Cracking(IASCC, ASMECode Visual(VT3)examofthe Allaccessiblesurfacesat Assembly Fatigue)

SectionXIas lowercoreplatestodetect specifiedfrequency.

Lowercoreplate AgingManagement supplementedby evidenceofdistortionand/or XLlowercoreplate(Note1)

(IE)

TB164 lossofboltintegrity.

W12b.Lower Internals Allplants Lossofmaterial ASMECode Visual(VT3)examination.

Allaccessiblesurfacesat Assembly (Wear)

SectionXIas specifiedfrequency.

Lowercoreplate supplementedby XLlowercoreplate(Note1)

TB164 W13.Bottom Mounted Instrumentation System Allplants Lossofmaterial (Wear)

IEB8809 Surface(ET)examination.

Eddycurrentsurface examinationasdefined Fluxthimbletubes inplantresponseto IEB8809.

W14.Alignment and Interfacing Components ClevisbearingStellitewearsurface Allplants (TB145)

Lossofmaterial (wear)

Cracking(SCC)

ASMECode SectionXIas supplementedby InterimGuidance (Note2,3)

Visual(VT3)examination.

Volumetric(UT)(Note3)

Allaccessiblesurfacesat specifiedfrequency.

Note3 Clevisinsertbolts(Note2,3)

W15.Alignment and Allplants Lossofmaterial ASMECode Visual(VT3)examination.

Allaccessiblesurfacesat Interfacing Components (Wear)

SectionXIas specifiedfrequency.

Uppercoreplatealignmentpins supplementedby TB164

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Aging Management Program, Table 3 Table49(continued)

WestinghousePlantsExistingProgramsComponents 9

NotestoTable49:

1. XL=ExtraLongreferringtoWestinghouseplantswith14foot cores.

2. Theclevisinsertsareattachedtointegrallyweldedreactorvessellugsandtheinsertsareboltedtothelugs.TheASMECodeexaminationofaccessiblesurfacesisconsidered toincludealldetailsoftheclevisconfiguration,includingtheboltingandlockingdevices.Theboltingisfabricatedfromnickelbasedmaterialsandissusceptibletostress corrosioncracking(SCC).Althoughfailureoftheboltingdoesnotitselfcauselossofsupportfunction,assetimpairmentorissueswithcorebarrelremovalareasubsequent possibility.WestinghousetechnicalbulletinTB145dated8/25/2014providesadditionalinformationregardingpossiblevisualindicationsthatclevisboltingfailuremayhave occurred.ThisinformationshouldbereviewedtoensureaheightenedawarenessoftheexaminersisappliedtothisCodeinspection.

3. Theclevisinsertboltingshallbeexamined,evaluatedanddispositionedperNEI0308NeededrequirementsasspecifiedinPWROGLetterOG21160dated9/1/2021.

Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 25 of 32

Aging Management Program, Table 4 Table53 WestinghousePlantsExaminationAcceptanceandExpansionCriteria 10

Primary Item

Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Expansion Item Examination Acceptance Criteria W1.Control Rod Guide Tube Assembly Guideplates(cards)

Allplants PertherequirementsofWCAP 17451P Thespecificrelevantconditionis wearthatcouldleadtolossof controlrodalignmentandimpede controlassemblyinsertion.

None N/A PerWCAP17451P[37].

W2.Control Rod Guide Tube Assembly Lowerflangewelds Allplants Enhancedvisual(EVT1) examination.

Thespecificrelevantconditionisa detectablecracklikesurface indication.

W2.1.Remaining accessibleCRGTlower flangewelds

W2.2.Bottommounted instrumentation(BMI) columnbodies Confirmationofsurfacebreakingindicationsintwo ormoreCRGTlowerflangeweldsshallrequire visual(EVT1)examinationoftheremaining accessibleCRGTlowerflangeweldsandvisual(VT 3)examinationofBMIcolumnbodiesbythe completionofthenextrefuelingoutage.

ForBMIcolumnbodies,the specificrelevantconditionforthe VT3examinationiscompletely fracturedcolumnbodies.

W3.Core Barrel Assembly Upperflangeweld(UFW)

Allplants Periodicenhancedvisual(EVT1),

volumetric(UT),oreddycurrent (ET)examination.

Thespecificrelevantconditionisa detectablecracklikesurface indication.

W3.3.Lowerflange weld(LFW)(Note2)

W3.2.Upperaxial welds(UAW)

W3.4.Lowersupport forging/casting

a. Theconfirmeddetectionandsizingofasurface breakingindicationwithalengthgreaterthan twoinchesintheUFWshallrequirethatthe inspectionbeexpandedtoincludetheLFWby thecompletionofthenextrefuelingoutage.

b. Theconfirmeddetectionandsizingofasurface breakingindicationwithalengthgreaterthan twoinchesintheLFWshallrequirethatthe inspectionbeexpandedtoincludetheUAWby thecompletionofthenextrefuelingoutage.

c. Theconfirmeddetectionofasurfacebreaking indicationwithalengthgreaterthantwoinches intheLFWshallrequiretheinspectionofthe lowersupportforgingorcasting(25%ofthenon coresidesurface)withinthenextthreerefueling outages.Ifanindicationisfoundinthis inspectionofthelowersupportforgingor casting,theexaminationcoverageshallbe expandedto100%oftheaccessiblesurfaceof thenoncoresidesurfaceofthelowersupport forgingorcastingduringthesamerefueling outage.

Thespecificrelevantconditionfor theexpansioncorebarrelwelds (LFW,UAW)andlowersupport forgingorcastingexaminationsis adetectablecracklikesurface indication.

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Aging Management Program, Table 4 Table53(continued)

WestinghousePlantsExaminationAcceptanceandExpansionCriteria 11

Primary Item

Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Expansion Item Examination Acceptance Criteria W3a.Core Barrel Assembly Uppergirthweld (UGW)

Allplants Enhancedvisual(EVT1),

volumetric(UT),oreddy current(ET)examination Thespecificrelevantcondition isadetectablecracklike surfaceindication.

W3.3.Lowerflange weld(LFW)(Note2)

W3.2.Upperaxial Welds(UAW)

W3.4.Lowersupport forging/casting

a. Theconfirmeddetectionandsizingofasurface breakingindicationwithalengthgreaterthan twoinchesintheUGWshallrequirethatthe inspectionbeexpandedtoincludetheLFWby thecompletionofthenextrefuelingoutage.

b. Theconfirmeddetectionandsizingofasurface breakingindicationwithalengthgreaterthan twoinchesintheUGWshallrequirethatthe inspectionbeexpandedtoincludetheUAWby thecompletionofthenextrefuelingoutage.

c. Theconfirmeddetectionofasurfacebreaking indicationwithalengthgreaterthantwoinches intheLFWshallrequiretheinspectionofthe lowersupportforgingorcasting(25%ofthe noncoresidesurface)withinthenextthree refuelingoutages.Ifanindicationisfoundinthis inspectionofthelowersupportforgingor casting,theexaminationcoverageshallbe expandedto100%oftheaccessiblesurfaceof thenoncoresidesurfaceofthelowersupport forgingorcastingduringthesamerefueling outage.

Thespecificrelevantconditionfor theexpansioncorebarrelwelds (LFW,UAW)andlowersupport forgingorcastingexaminationsis adetectablecracklikesurface indication.

W4.Core Barrel Assembly Lowergirthweld(LGW)

Allplants Periodicenhancedvisual(EVT1) examination.

Thespecificrelevantconditionisa detectablecracklikesurface indication.

W4.1.Uppercoreplate

W4.4.Lowersupport columnbodies(cast andnoncast)

W4.2.Middleaxial welds(MAW)

W4.3.Loweraxial welds(LAW)

a. Theconfirmeddetectionandsizingofasurface breakingindicationwithalengthgreaterthan twoinchesintheLGWshallrequireinspection oftheuppercoreplate(25%ofthecoreside surface)withinthenextthreerefuelingoutages.

Ifanindicationisfoundinthisinspectionofthe uppercoreplate,theexaminationcoverage shallbeexpandedto100%oftheaccessible surfaceofthecoresidesurfaceoftheupper coreplateduringthesamerefuelingoutage.

b. Theconfirmeddetectionandsizingofa surfacebreakingindicationwithalength greaterthantwoinchesintheLGWshall requireinspectionofthelowersupportcolumn bodies(castandnoncast)withinthenextthree

a. Thespecificrelevant conditionsfortheinspection oftheuppercoreplateare brokenormissingpartsofthe plate.

b. Thespecificrelevant conditionsfortheinspection ofthelowersupportcolumn bodies(castandnoncast)are fractured,misaligned,or missingcolumns.

c. Thespecificrelevant conditionfortheexpansion MAWandLAWinspectionsis Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 27 of 32

Aging Management Program, Table 4 Table53(continued)

WestinghousePlantsExaminationAcceptanceandExpansionCriteria 12

Primary Item

Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Expansion Item Examination Acceptance Criteria refuelingoutages.

Theconfirmeddetectionoffractured, misaligned,ormissinglowersupportcolumns shallrequireexaminationof100%ofthe accessibleuninspectedlowersupportcolumn assembliesusingaVT3examinationfrom abovethelowercoreplate(minimumof75%of thetotalpopulationoflowersupportcolumn assemblies)duringthesameoutage.

c. Theconfirmeddetectionandsizingofasurface breakingindicationwithalengthgreaterthan twoinchesintheLGWshallrequirethatthe inspectionsbeexpandedtoincludethelower corebarrelcylinderaxialweldsbythe completionofthenextrefuelingoutage.

adetectablecracklike surfaceindication.

W5.Baffle-Former Assembly Baffleedgebolts NAforMPS3 Allplantswith baffleedgebolts Visual(VT3)examination.

Thespecificrelevantconditionsare missingorbrokenlockingdevices, cracked/failedormissingbolts,and protrusionofboltheads.

None N/A N/A W6.Baffle-Former Assembly Baffleformerbolts Allplants Volumetric(UT)examination.

Theexaminationacceptance criteriafortheUTofthebaffle formerboltsshallbeestablishedas partoftheexaminationtechnical justification.

W6.2.Lowersupport columnbolts

W6.1.Barrelformer bolts

a. Confirmationthatmorethan5%ofthebaffle formerboltsactuallyexaminedonthefour baffleplatesatthelargestdistancefromthe core(presumedtobethelowestdoselocations) containunacceptableindicationsshallrequire inspectionofthelowersupportcolumnbolts withinthenextthreefuelcycles.

b. Confirmationthatmorethan5%ofthelower supportcolumnboltsactuallyexaminedcontain unacceptableindicationsshallrequire inspectionofthebarrelformerboltswithin threerefuelingcycles.

Theexaminationacceptance criteriafortheUTofthelower supportcolumnboltsandthe barrelformerboltsshallbe establishedaspartofthe examinationtechnical justification.

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Aging Management Program, Table 4 Table53(continued)

WestinghousePlantsExaminationAcceptanceandExpansionCriteria 13

Primary Item

Applicability Examination Acceptance Criteria (Note 1)

Expansion Link(s)

Expansion Criteria Expansion Item Examination Acceptance Criteria W7.Baffle-Former Assembly (Includes:Baffleplates, baffleedgebolts, cornerbolts,and indirecteffectsofvoid swellinginformer plates)

Allplants Visual(VT3)examination.

Thespecificrelevantconditionsare evidenceofabnormalinteraction withfuelassemblies,gapsalong highfluencebaffleplatejoints, verticaldisplacementofbaffle platesnearhighfluencejoints,or morethan2brokenordamaged edgeboltlockingsystemsalong highfluencebaffleplatejoints.

None N/A N/A W8.Alignment and Interfacing Components Internalsholddown spring NAforMPS3 Allplantswith 304stainless steelholddown springs Directphysicalmeasurementof springheight.

Theexaminationacceptance criterionforthismeasurementis thattheremainingcompressible heightofthespringshallprovide holddownforceswithintheplant specificdesigntolerance.

None N/A N/A W9.Thermal Shield Assembly Thermal shield flexures NAforMPS3 Allplantswith thermalshields Visual(VT3)examination.

Thespecificrelevantconditionsfor thermalshieldflexuresare excessivewear,fracture,or completeseparation.

None N/A N/A Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 29 of 32

Aging Management Program, Table 4 Table53(continued)

WestinghousePlantsExaminationAcceptanceandExpansionCriteria 14 NotestoTable53:

1. Theexaminationacceptancecriterionforvisualexaminationistheabsenceofthespecifiedrelevantcondition(s).

2. ThelowercorebarrelflangeweldmayalternativelybedesignatedasthecorebarreltosupportplateweldinsomeWestinghouseplantdesigns.

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AGING MANAGEMENT PROGRAM ATTACHMENT 2 - LICENSEE ACTION ITEMS Millstone Power Station Unit 3 Applicant/Licensee Action Item 1 1.0 Background and Purpose The current Millstone Power Station Unit 3 (MPS3) aging management program (AMP) for the reactor internals is based on the Materials Reliability Program (MRP) specifications contained in MRP-227 Rev. 1-A, Reference 1 below. The applicable NRC Safety Evaluation (SE) (published within Ref. 1) requires compliance with the following A/LAI.

2.0 SE A/LAI 1: (Submittal of evaluations for degraded baffle-former bolting)

The action item text from the SE contained in MRP-227 Rev. 1-A [1] states:

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.

MPS3 Compliance EPRI Letter MRP 2017-009, as applicable to MPS3 (a tier 4 plant), requires A. Baseline volumetric (UT) examination performed no later than 35 EFPY B. If the number of baffle bolts with indications is less than 5% and there is no clustering, the next inspection must be scheduled within 10 years. If there is clustering of indications or the number of bolts with indications is equal or greater than 5%, the next inspection must be schedule within 6 years.

C. Optionally, baffle bolt replacements may be performed to achieve acceptable bolting patterns and a longer reinspection interval, when justified in accordance with established methodologies.

MPS3 compliance with these requirements is specified by the notes 8 and 9 of Table 1 of of this AMP (Table 1 reproduces appliable portions of MRP-227 Rev. 1-A, Table 4-3.) The included Note 8 requires inspection within 35 EFPY and Note 9 requires a bolting reinspection schedule depending on the severity bolting degradation found; the Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 31 of 32

AGING MANAGEMENT PROGRAM note requires justification for longer reinspection intervals and is consistent with and references the MRP 2017-009 letter. MRP-227 Section 7.5, Examination Results Requirement, states that engineering evaluations shall be conducted in accordance with NRC approved evaluation methods. The basis for engineering evaluations of baffle former bolting in the MPS3 program is WCAP-17096-NP-A Rev. 3 (Reference 2),

including the limitations and conditions stated in the NRC safety evaluation of the report.

On page E-70 of Appendix E of the report, it states that any evaluation to lengthen the interval to the subsequent baffle bolt inspection 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. Therefore, by observing the Needed requirements of WCAP-17096-NP-A Rev. 3, the MPS3 program is in compliance with A/LAI 1.

3.0 References

1. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227, Revision 1-A), Electric Power Research Institute, Palo Alto, CA dated December 2019. 3002017168

2. Westinghouse WCAP-17096-NP-A Rev. 3, Reactor Internals Acceptance Criteria Methodology and Data Requirements, August 2023, Westinghouse Electric Company LLC, Pittsburgh, PA.

Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 32 of 32