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

Dom inion Energy Nuclear Co nnecti cut, Inc.

5000 Dom inion Boul evard, Glen All en, VA 2306 0 Dom inion Energy.com November 20, 2023 ATTN: Document Control Desk Serial No.: 23-290 U.S. Nuclear Regulatory Commission NRA/SS: RO Washington, DC 20555-0001 Docket No.: 50-423 License No.: NPF-49 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

MILLSTONE POWER STATION UNIT 3 REACTOR VESSEL INTERNALS INSPECTIONS AGING MANAGEMENT PROGRAM SUBMITTAL RELATED TO LICENSE RENEWAL COMMITMENT 13 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

Serial No.: 23-290 Docket No.: 50-423 Page 2 of 2 Commitments made by this letter: None

Enclosure:

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

Serial No.: 23-290 Docket No.: 50-423 LICENSE RENEWAL Enclosure, Page 1 of 32 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.

Serial No.: 23-290 Docket No.: 50-423 LICENSE RENEWAL Enclosure, Page 2 of 32 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.

Serial No.: 23-290 Docket No.: 50-423 LICENSE RENEWAL Enclosure, Page 3 of 32 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 Except for the first fuel cycle, MPS3 has Assumption with high-leakage always operated with low leakage fuel satisfied core loading patterns loading patterns. Historical fuel loading followed by patterns have been validated. Relevant implementation of a results include:

low-leakage fuel MPS3 active fuel to upper core plate management strategy distance > 12.2 inches for the remaining 30 MPS3 average core power density <

years of operation, as 124 W/cm3 well as the average MPS3 heat generation figure of core power levels and merit, F < 68 W/cm3 proximity of active fuel to the upper core MPS3 continues to validate these support plate satisfies assumptions for each fuel cycle.

limits described in MRP-227 Rev. 1-A Appendix B Base load operation MPS3 has always operated as a base load Assumption for majority of plant unit and has no plans to alter this practice. satisfied operation

Serial No.: 23-290 Docket No.: 50-423 LICENSE RENEWAL Enclosure, Page 4 of 32 AGING MANAGEMENT PROGRAM Table 2.1-1 MPS3 Reconciliation with MRP-227 Rev. 1-A Applicability Assumptions Assumption MPS3 Specifics Reconciliation No design changes MPS3 is a Westinghouse design. Assumption beyond those Significant design changes to MPS3 satisfied identified in general include:

industry guidance or Control rod guide tube (CRGT) split pin recommended by the replacement with Type 316 pins in 2008 original vendors. 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.

The components and Generically validated as part of MRP-191 Assumption material class of each Rev. 2. satisfied functional component The component materials for MPS3 were are as listed in the validated by Westinghouse latest revision of MRP-191.

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 LICENSE RENEWAL Enclosure, Page 5 of 32 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 LICENSE RENEWAL Enclosure, Page 6 of 32 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 LICENSE RENEWAL Enclosure, Page 7 of 32 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.

Serial No.: 23-290 Docket No.: 50-423 LICENSE RENEWAL Enclosure, Page 8 of 32 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 LICENSE RENEWAL Enclosure, Page 9 of 32 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 LICENSE RENEWAL Enclosure, Page 10 of 32 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.

Serial No.: 23-290 Docket No.: 50-423 LICENSE RENEWAL Enclosure, Page 11 of 32 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.

Serial No.: 23-290 Docket No.: 50-423 LICENSE RENEWAL Enclosure, Page 12 of 32 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 LICENSE RENEWAL Enclosure, Page 13 of 32 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 LICENSE RENEWAL Enclosure, Page 14 of 32 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 LICENSE RENEWAL Enclosure, Page 15 of 32 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 LICENSE RENEWAL Enclosure, Page 16 of 32 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 Aging Management Program Attachment 1, Table 1 Docket No.: 50-423 Enclosure, Page 17 of 32 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.

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Serial No.: 23-290 Aging Management Program Attachment 1, Table 1 Docket No.: 50-423 Enclosure, Page 18 of 32 Table 43 Westinghouse Plants Primary Components Expansion Link Examination Primary Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Examination coverage per the All plants requirements of WCAP17451P, W1.Control Rod Per the requirements of WCAP Loss of Material Revision 1 (Notes 5, 12) See Figure Guide Tube Assembly (See WEC None 17451P, including subsequent (Wear) 411 Guide plates (cards) NSAL171) examinations (Notes 5, 12).

Enhanced visual (EVT1) examination 100% of outer (accessible) CRGT to determine the presence of cracklike lower flange weld surfaces and W2.1.Remaining CRGT surface flaws in flange welds no later 0.25inch of the adjacent base W2.Control Rod assembly lower flange than 2 refueling outages from the Cracking (SCC, Fatigue) metal on the individual periphery Guide Tube Assembly All plants welds beginning of the license renewal Aging Management (IE and TE) CRGT assemblies.

Lower flange welds period and subsequent examination (Note 2)

W2.2.BMI column bodies on a tenyear interval. See Figure 412.

100% of the accessible weld W3.3.lower flange weld length of both surfaces (ID Enhanced visual (EVT1), eddy current (LFW), W3.2.Upper axial surface and OD surface) of the W3.Core Barrel (ET), or volumetric (UT) examination, welds (UAW), and UFW and 3/4 of adjacent base Assembly All plants Cracking (SCC) no later than 2 refueling outages from W3.4.Lower support the beginning of the license renewal metal shall be examined.

Upper flange Weld (UFW) forging or casting period and subsequent examination (Note 11) on a tenyear interval. See Figure 413.

W3a.Core Barrel All plants Cracking (SCC) W3.2. Upper core barrel Enhanced visual (EVT1), 100% of the accessible weld Assembly upper axial welds (UAW) volumetric (UT), or surface (ET) length of both surfaces (ID Upper girth weld (UGW) W3.3. Lower flange weld examination, no later than 2 surface and OD surface) of the (Promoted back to Primary) refueling outages from the UGW and 3/4 of adjacent base (LFW), W3.4. Lower beginning of the license renewal metal shall be examined. If UT is support forging or casting period and subsequent performed, it need only be examination on a tenyear completed from one surface, interval.

either ID or OD. (Note 11).

See Figure 413.

W4.1.Upper core plate, 100% of the accessible weld W4.4.Lower support Enhanced visual (EVT1) examination, length of the OD surface of the column bodies (cast, non W4.Core Barrel volumetric (UT), or surface (ET) no LGW and 3/4 of adjacent base Cracking (SCC, IASCC), Aging cast),

Assembly All plants later than 2 refueling outages from metal shall be examined. (ID Management (IE) W4.2.Middle axial welds surface is inaccessible for Lower girth weld (LGW) the beginning of the license renewal (MAW), visual/ET based surface exams period and subsequent examination W4.3.Lower axial welds on a tenyear interval. due to baffle former assembly).

(LAW) UT is performed from OD surface.

(Note 6)

See Figure 413.

2

Serial No.: 23-290 Aging Management Program Attachment 1, Table 1 Docket No.: 50-423 Enclosure, Page 19 of 32 Table 43 (continued)

Westinghouse Plants Primary Components Expansion Link Examination Primary Item Applicability Effect (Mechanism) Examination Coverage (Note 1) Method/Frequency (Note 1)

Cracking (IASCC, Fatigue) that Bolts and locking devices on high results in Visual (VT3) examination, with fluence seams. 100% of W5.Baffle-Former All plants with Lost or broken locking devices components accessible from core baseline examination between 20 Assembly baffleedge None side.

Failed or missing bolts and 40 EFPY and subsequent Baffleedge bolts bolts examinations on a tenyear interval.

NA for MPS 3 Protrusion of bolt heads Aging See Figure 414.

Management (IE and ISR) (Note 4)

Baseline volumetric (UT) examination W6.Baffle-Former interval is dependent on the plant 100% of accessible All plants Cracking (IASCC, Fatigue) W6.2.Lower support Assembly design (Note 8). Subsequent bolts. (Note 3)

(See WEC Aging Management (IE and ISR) column bolts, Baffleformer bolts examination is dependent on the plant NSAL161) (Note 4) W6.1.Barrelformer bolts (Note 7) design and the results of the baseline See Figure 415.

inspection (Note 9).

Distortion (Void Swelling), or W7.Baffle-Former Core side surface:

Cracking (IASCC) that results in Assembly Visual (VT3) examination to check for High fluence baffle joints Abnormal interaction with fuel Assembly (Includes: Baffle evidence of distortion, with baseline assemblies Top and bottom edge of plates, baffle edge bolts, All plants None examination between 20 and 40 corner bolts, and indirect EFPY and subsequent examinations on baffle plates Gaps between plates effects of void swelling in a tenyear interval.

Bolts and locking devices former plates) Vertical displacement of baffle plates See Figure 416.

Broken or damaged edge bolts Measurements should be taken at Direct measurement of spring height several points around the within three cycles of the beginning of W8.Alignment and All plants with circumference of the spring, with a (before or after) the license renewal Interfacing 304 stainless Distortion (Loss of Load due to Stress statistically adequate number of None period. If the first set of measurements Components steel hold Relaxation) measurements at each point to is not sufficient to assess remaining Internals hold down spring down springs minimize uncertainty.

life, additional spring height NA for MPS 3 measurements will be required.

See Figure 417.

Cracking (Fatigue) or Loss of 100% of accessible surfaces of All plants with Visual (VT3) no later than 2 refueling W9.Thermal Shield Material (Wear) that results in 100% of thermal shield flexures.

thermal shields outages from the beginning of the Assembly thermal shield flexures excessive None (Note 10)

(See WEC license renewal period. Subsequent Thermal shield flexures wear, fracture, or complete TB195) examinations on a tenyear interval.

NA for MPS 3 separation See Figure 418.

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Serial No.: 23-290 Aging Management Program Attachment 1, Table 1 Docket No.: 50-423 Enclosure, Page 20 of 32 Table 43 (continued)

Westinghouse Plants Primary Components Notes to Table 43:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 53.

2. A minimum of 75% of the total identified sample population must be examined.

3. A minimum of 75% of the total bolt population (examined + unexamined), including coverage consistent with the Expansion criteria in Table 53, must be examined for inspection credit.

4. Void swelling effects on this component are managed through management of void swelling on the entire baffleformer assembly.

5. In WCAP17451P the baseline examination schedule has been adjusted for various CRGT designs, the extent of individual CRGT examination modified, and flexible subsequent examination regimens correlating to initial baseline sample size, accuracy of wear estimation, and examination results. Initial inspection prior to the license renewal period may be required. Use WCAP17451P [37], Revision 1, including the modified requirements due to the interim guidance provided in EPRI letter MRP 2018007 dated 3/7/2018 [47] and PWROG letter OG1846 dated 2/20/2018 [38].

6. Examination coverage requires a minimum of 50% of the length of the OD of the weld being examined.

7. Baffleformer bolt inspection includes inspection of the corner plate bolts when applicable.

8. In accordance with MRP 2017009 [39] and MRP 2017010 [41], Tier 1 plants are to perform the baseline UT examination by 20 EFPY or during the next refueling outage after March 1, 2016. Per MRP 2017009 [39], Tier 2 plants are to perform the baseline UT examination at no later than 30 EFPY (initial Tier 2 plant baseline UT exams performed prior to 1/1/2018 are acceptable). All other remaining plants are to perform the baseline UT examination at no later than 35 EFPY.

9. Reexamination periods shall be determined by plantspecific evaluation per the MRP227 Needed Requirement 7.5 as documented and dispositioned in the owners plant corrective action program. If atypical or aggressive baffleformer bolt degradation as defined in MRP 2017009 [39] (i.e., 3% of baffleformer bolts with UT or visual indications or clustering*

for downflow plants and 5% of baffleformer bolts with UT or visual indications or clustering* for upflow plants) is observed, the interim guidance (MRP 2016021 [40] and MRP 2017009 [39]) provides limitations to the permitted reinspection interval (not to exceed 6 years maximum) unless further evaluation is performed to justify a longer interval (See Applicant/Licensee Action Item 1 in the NRC SE for evaluation submittal requirements [35]). If evaluation justifies a longer reinspection interval, it is not permitted to exceed 10 years.

• Clustering is defined per NSAL161 Rev. 1 [42] as three or more adjacent defective baffleformer bolts or more than 40% defective baffleformer bolts on the same baffle plate.

Untestable bolts should be reviewed on a plantspecific basis consistent with WCAP17096NPA for determination if these should be considered when evaluating clustering.

10. See Westinghouse Technical Bulletin TB195 dated 10/9/2019 and MRP 2019017 dated 5/31/2019 for additional details on inspection recommendations.

11. Examination coverage requires a minimum of 75% of the weld length of both the ID and the OD of the weld being examined.

12. (MPS 3 specific) Baseline examinations of the CRGT guide cards were completed in 2019 and were acceptable for greater than 10 EFPY. Subsequent examinations and evaluations will be governed by the requirements of WCAP17451 Rev. 2, consistent with WCAP17096 Rev. 3.

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Serial No.: 23-290 Aging Management Program Attachment 1, Table 2 Docket No.: 50-423 Enclosure, Page 21 of 32 Table 46 Westinghouse Plants Expansion Components Primary Examination Expansion Item Applicability Effect (Mechanism) Link Method/Frequency (Note 1) Examination Coverage (Note 1)

Control Rod Guide All plants Cracking (SCC, W2.CRGT Lower Enhanced visual (EVT1) examination to A minimum of 75% of the CRGT Tube Assembly Fatigue) Flange Welds determine the presence of cracklike assembly lower flange weld surfaces W2.1.Remaining CRGT Aging Management (IE surface flaws in flange welds. and 0.25inch of the adjacent base lower flange welds and TE) Subsequent examination on a tenyear metal for the flange welds not interval. inspected under the primary link.

Bottom Mounted All plants Cracking (Fatigue) W2.CRGT Lower Visual (VT3) examination. Reinspection 100% of BMI column bodies for which Instrumentation including the detection Flange Welds every 10 years following initial difficulty is detected during flux thimble System of completely fractured inspection. insertion/withdrawal.

W2.2.Bottommounted column bodies instrumentation (BMI) Aging Management (IE) See Figure 424.

column bodies Core Barrel Assembly All plants Cracking (SCC) W3.Upper Core Enhanced surface visual (EVT1), eddy 100% of the accessible weld length of W3.2.Upper Axial Weld Barrel Flange current (ET), or volumetric (UT) both surfaces (ID surface and OD (UAW) Weld (UFW) examination. Reinspection every 10 surface) of the UAW and 3/4 of W3a. Upper girth years following initial inspection. adjacent base metal shall be weld (UGW) examined. If UT is performed, it need only be completed from one surface, either ID or OD.(Note 2).

See Figure 413.

100% of the accessible weld length of the OD surface of the LFW and 3/4 of Core Barrel Assembly W3.Upper Core Enhanced visual (EVT1) examination. adjacent base metal shall be W3.3.Lower Flange Weld All plants Cracking (SCC) Barrel Flange Reinspection every 10 years following examined (LFW) Weld (UFW) initial inspection. (Note 5).

See Figure 413.

Lower Internals Minimum of 25% of bottom (noncore Cracking (SCC) Visual (VT3) examination.

Assembly W3.Upper Core Barrel side) surface (Note 3).

All plants Aging Management (TE Reinspection every 10 years W3.4.Lower support Flange Weld (UFW) in Casting) Ref.[34] following initial inspection.

forging or castings See Figure 420.

Minimum of 25% of core side surfaces Upper Internals Cracking (Fatigue), Visual (VT3) examination.

W4.Lower Girth (Note 3).

Assembly All plants Wear, Aging Reinspection every 10 years Weld (LGW)

W4.1.Upper core plate Management (IE) following initial inspection.

See Figure 419.

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Serial No.: 23-290 Aging Management Program Attachment 1, Table 2 Docket No.: 50-423 Enclosure, Page 22 of 32 Table 46 (continued)

Westinghouse Plants Expansion Components Primary Examination Expansion Item Applicability Effect (Mechanism) Link Method/Frequency (Note 1) Examination Coverage (Note 1) 100% of the accessible weld length Core Barrel Assembly Enhanced visual (EVT1) of the OD of the MAW and LAW and W4.2.Middle Axial Welds Cracking (SCC, IASCC) W4.Lower Girth examination. 3/4 of adjacent base metal shall be All plants (MAW) and W4.3.Lower Aging Management (IE) Weld (LGW) Reinspection every 10 years examined (Notes 5 and 6).

Axial Welds (LAW) following initial inspection See Figure 413.

25% of the total number of column assemblies (both visible and nonvisible from above the lower core plate) using Lower Support a VT3 examination from above the Assembly Visual (VT3) examination.

Cracking (IASCC) W4.Lower Girth lower core plate. The inspection W4.4.Lower support All plants Reinspection every 10 years Aging Management (IE) Weld (LGW) coverage must be evenly distributed column bodies (both cast following initial inspection. across the population of column and noncast) assemblies (Notes 3 and 4).

See Figure 423.

100% of accessible barrelformer bolts (Minimum of 75% of the total Cracking (IASCC, W6.Baffleformer Volumetric (UT) examination. population). Accessibility may be Core Barrel Assembly Fatigue)

All plants bolts (also refer to Reinspection every 10 years limited by presence of thermal shield W6.1.Barrelformer bolts Aging Management (IE, or neutron pads.

MRP 2018002) following initial inspection.

Void Swelling and ISR)

See Figure 421.

100% of accessible LSC bolts Lower Support Cracking (IASCC, (Minimum of 75% of the total Volumetric (UT) examination.

Assembly Fatigue) population) or as supported by plant All plants W6.Baffleformer bolts Reinspection every 10 years W6.2.Lower support Aging Management (IE following initial inspection. specific justification.

column bolts and ISR)

See Figure 422 6

Serial No.: 23-290 Aging Management Program Attachment 1, Table 2 Docket No.: 50-423 Enclosure, Page 23 of 32 Table 46 (continued)

Westinghouse Plants Expansion Components Notes to Table 46:

1. Examination acceptance criteria and expansion criteria for the Westinghouse components are in Table 53.

2. Examination coverage requires a minimum of 75% of the weld length of both the ID and the OD of the weld being examined.

3. The stated minimum coverage requirement is the minimum if no significant indications are found. However, the Examination Acceptance criteria in Section 5 require that additional coverage must be achieved in the same outage if significant flaws are found. This contingency should be considered for inspection planning purposes.

4. Justification that adequate distribution of the inspection coverage has been achieved can be based on geometric or layout arguments. Possible examples include, but are not limited to, inspection of all column assemblies in one quadrant of the lower core plate (based on the azimuthal symmetry of the plate) or inspecting every fourth column across the entire plate.

5. A minimum coverage of 75% of the weld length on the surface being examined shall be achieved; however, for welds with limited access (Note 6), a minimum examination coverage of 50% of the weld length on the surface being examined shall be achieved.

6. Accessibility to the MAW and LAW may be limited by the thermal shield or neutron panels - no disassembly to achieve higher weld length coverage is required.

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Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 24 of 32 Aging Management Program Attachment 1, Table 3 Table 49 Westinghouse Plants Existing Programs Components Effect Examination Item Applicability (Mechanism) Reference Examination Method Coverage W10.Core Barrel Assembly All plants Loss of material ASME Code Visual (VT3) exam to All accessible surfaces at Core barrel flange (Wear) Section XI determine general condition specified frequency.

for excessive wear.

W11.Upper Internals All plants Cracking (SCC, ASME Code Visual (VT3) examination. All accessible surfaces at Assembly Fatigue) Section XI specified frequency.

Upper support ring or skirt W12a.Lower Internals All plants Cracking (IASCC, ASME Code Visual (VT3) exam of the All accessible surfaces at Assembly Fatigue) Section XI as lower core plates to detect specified frequency.

Lower core plate Aging Management supplemented by evidence of distortion and/or XL lower core plate (Note 1) (IE) TB164 loss of bolt integrity.

W12b.Lower Internals All plants Loss of material ASME Code Visual (VT3) examination. All accessible surfaces at Assembly (Wear) Section XI as specified frequency.

Lower core plate supplemented by XL lower core plate (Note 1) TB164 W13.Bottom Mounted All plants Loss of material IEB 8809 Surface (ET) examination. Eddy current surface Instrumentation System (Wear) examination as defined Flux thimble tubes in plant response to IEB 8809.

W14.Alignment and All plants Loss of material ASME Code Visual (VT3) examination. All accessible surfaces at Interfacing Components (TB145) (wear) Section XI as specified frequency.

Clevis bearing Stellite wear surface supplemented by Cracking (SCC) Interim Guidance Volumetric (UT) (Note 3) Note 3 (Note 2, 3)

Clevis insert bolts (Note 2, 3)

W15.Alignment and All plants Loss of material ASME Code Visual (VT3) examination. All accessible surfaces at Interfacing Components (Wear) Section XI as specified frequency.

Upper core plate alignment pins supplemented by TB164 8

Serial No.: 23-290 Docket No.: 50-423 Enclosure, Page 25 of 32 Aging Management Program Attachment 1, Table 3 Table 49 (continued)

Westinghouse Plants Existing Programs Components Notes to Table 49:

1. XL = Extra Long referring to Westinghouse plants with 14foot cores.

2. The clevis inserts are attached to integrally welded reactor vessel lugs and the inserts are bolted to the lugs. The ASME Code examination of accessible surfaces is considered to include all details of the clevis configuration, including the bolting and locking devices. The bolting is fabricated from nickelbased materials and is susceptible to stress corrosion cracking (SCC). Although failure of the bolting does not itself cause loss of support function, asset impairment or issues with core barrel removal are a subsequent possibility. Westinghouse technical bulletin TB 145 dated 8/25/2014 provides additional information regarding possible visual indications that clevis bolting failure may have occurred. This information should be reviewed to ensure a heightened awareness of the examiners is applied to this Code inspection.

3. The clevis insert bolting shall be examined, evaluated and dispositioned per NEI 0308 Needed requirements as specified in PWROG Letter OG 21160 dated 9/1/2021.

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Serial No.: 23-290 Aging Management Program Attachment 1, Table 4 Docket No.: 50-423 Enclosure, Page 26 of 32 Table 53 Westinghouse Plants Examination Acceptance and Expansion Criteria Expansion Item Examination Acceptance Primary Item Applicability Expansion Link(s) Expansion Criteria Examination Acceptance Criteria (Note 1)

Criteria W1.Control Rod All plants Per the requirements of WCAP None N/A Per WCAP17451P [37].

Guide Tube Assembly 17451P Guide plates (cards) The specific relevant condition is wear that could lead to loss of control rod alignment and impede control assembly insertion.

W2.Control Rod All plants Enhanced visual (EVT1) W2.1.Remaining Confirmation of surfacebreaking indications in two For BMI column bodies, the Guide Tube Assembly examination. accessible CRGT lower or more CRGT lower flange welds shall require specific relevant condition for the Lower flange welds flange welds visual (EVT1) examination of the remaining VT3 examination is completely The specific relevant condition is a accessible CRGT lower flange welds and visual (VT fractured column bodies.

detectable cracklike surface W2.2.Bottommounted 3) examination of BMI column bodies by the indication. instrumentation (BMI) completion of the next refueling outage.

column bodies W3.Core Barrel All plants Periodic enhanced visual (EVT1), a. The confirmed detection and sizing of a surface The specific relevant condition for Assembly volumetric (UT), or eddy current W3.3.Lower flange breaking indication with a length greater than the expansion core barrel welds Upper flange weld (UFW) (ET) examination. weld (LFW) (Note 2) two inches in the UFW shall require that the (LFW, UAW) and lower support inspection be expanded to include the LFW by forging or casting examinations is The specific relevant condition is a the completion of the next refueling outage. a detectable cracklike surface detectable cracklike surface W3.2.Upper axial indication.

welds (UAW) b. The confirmed detection and sizing of a surface indication.

breaking indication with a length greater than two inches in the LFW shall require that the inspection be expanded to include the UAW by the completion of the next refueling outage.

c. The confirmed detection of a surfacebreaking W3.4. Lower support indication with a length greater than two inches forging/casting in the LFW shall require the inspection of the lower support forging or casting (25% of the non core side surface) within the next three refueling outages. If an indication is found in this inspection of the lower support forging or casting, the examination coverage shall be expanded to 100% of the accessible surface of the noncore side surface of the lower support forging or casting during the same refueling outage.

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Serial No.: 23-290 Aging Management Program Attachment 1, Table 4 Docket No.: 50-423 Enclosure, Page 27 of 32 Table 53 (continued)

Westinghouse Plants Examination Acceptance and Expansion Criteria Expansion Item Examination Acceptance Primary Item Applicability Expansion Link(s) Expansion Criteria Examination Acceptance Criteria (Note 1)

Criteria W3a.Core Barrel All plants Enhanced visual (EVT1), W3.3. Lower flange a. The confirmed detection and sizing of a surface The specific relevant condition for Assembly volumetric (UT), or eddy weld (LFW) (Note 2) breaking indication with a length greater than the expansion core barrel welds Upper girth weld current (ET) examination two inches in the UGW shall require that the (LFW, UAW) and lower support (UGW) inspection be expanded to include the LFW by forging or casting examinations is The specific relevant condition W3.2. Upper axial the completion of the next refueling outage. a detectable cracklike surface is a detectable cracklike Welds (UAW) indication.

surface indication. b. The confirmed detection and sizing of a surface W3.4. Lower support breaking indication with a length greater than forging/casting two inches in the UGW shall require that the inspection be expanded to include the UAW by the completion of the next refueling outage.

c. The confirmed detection of a surface breaking indication with a length greater than two inches in the LFW shall require the inspection of the lower support forging or casting (25% of the noncore side surface) within the next three refueling outages. If an indication is found in this inspection of the lower support forging or casting, the examination coverage shall be expanded to 100% of the accessible surface of the noncore side surface of the lower support forging or casting during the same refueling outage.

W4.Core Barrel All plants Periodic enhanced visual (EVT1) W4.1.Upper core plate a. The confirmed detection and sizing of a surface a. The specific relevant Assembly examination. breaking indication with a length greater than conditions for the inspection Lower girth weld (LGW) W4.4.Lower support two inches in the LGW shall require inspection of the upper core plate are column bodies (cast of the upper core plate (25% of the coreside broken or missing parts of the The specific relevant condition is a and noncast) surface) within the next three refueling outages. plate.

detectable cracklike surface indication. If an indication is found in this inspection of the W4.2.Middle axial upper core plate, the examination coverage b. The specific relevant welds (MAW) shall be expanded to 100% of the accessible conditions for the inspection surface of the coreside surface of the upper of the lower support column W4.3.Lower axial core plate during the same refueling outage. bodies (cast and noncast) are welds (LAW) fractured, misaligned, or

b. The confirmed detection and sizing of a missing columns.

surfacebreaking indication with a length greater than two inches in the LGW shall c. The specific relevant require inspection of the lower support column condition for the expansion bodies (cast and noncast) within the next three MAW and LAW inspections is 11

Serial No.: 23-290 Aging Management Program Attachment 1, Table 4 Docket No.: 50-423 Enclosure, Page 28 of 32 Table 53 (continued)

Westinghouse Plants Examination Acceptance and Expansion Criteria Expansion Item Examination Acceptance Primary Item Applicability Expansion Link(s) Expansion Criteria Examination Acceptance Criteria (Note 1)

Criteria refueling outages. a detectable cracklike surface indication.

The confirmed detection of fractured, misaligned, or missing lower support columns shall require examination of 100% of the accessible uninspected lower support column assemblies using a VT3 examination from above the lower core plate (minimum of 75% of the total population of lower support column assemblies) during the same outage.

c. The confirmed detection and sizing of a surface breaking indication with a length greater than two inches in the LGW shall require that the inspections be expanded to include the lower core barrel cylinder axial welds by the completion of the next refueling outage.

W5.Baffle-Former All plants with Visual (VT3) examination. None N/A N/A Assembly baffleedge bolts Baffleedge bolts The specific relevant conditions are missing or broken locking devices, NA for MPS 3 cracked/failed or missing bolts, and protrusion of bolt heads.

W6.Baffle-Former All plants Volumetric (UT) examination. W6.2.Lower support a. Confirmation that more than 5% of the baffle The examination acceptance Assembly column bolts former bolts actually examined on the four criteria for the UT of the lower Baffleformer bolts The examination acceptance baffle plates at the largest distance from the support column bolts and the W6.1.Barrelformer core (presumed to be the lowest dose locations) barrelformer bolts shall be criteria for the UT of the baffle bolts contain unacceptable indications shall require established as part of the former bolts shall be established as part of the examination technical inspection of the lower support column bolts examination technical justification. within the next three fuel cycles. justification.

b. Confirmation that more than 5% of the lower support column bolts actually examined contain unacceptable indications shall require inspection of the barrelformer bolts within three refueling cycles.

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Serial No.: 23-290 Aging Management Program Attachment 1, Table 4 Docket No.: 50-423 Enclosure, Page 29 of 32 Table 53 (continued)

Westinghouse Plants Examination Acceptance and Expansion Criteria Expansion Item Examination Acceptance Primary Item Applicability Expansion Link(s) Expansion Criteria Examination Acceptance Criteria (Note 1)

Criteria W7.Baffle-Former All plants Visual (VT3) examination. None N/A N/A Assembly The specific relevant conditions are (Includes: Baffle plates, evidence of abnormal interaction baffle edge bolts, with fuel assemblies, gaps along corner bolts, and high fluence baffle plate joints, indirect effects of void vertical displacement of baffle swelling in former plates near high fluence joints, or plates) more than 2 broken or damaged edge bolt locking systems along high fluence baffle plate joints.

W8.Alignment and All plants with Direct physical measurement of None N/A N/A Interfacing 304 stainless spring height.

Components steel hold down Internals hold down springs The examination acceptance spring criterion for this measurement is NA for MPS 3 that the remaining compressible height of the spring shall provide holddown forces within the plant specific design tolerance.

W9.Thermal Shield All plants with Visual (VT3) examination. None N/A N/A Assembly thermal shields The specific relevant conditions for Thermal shield thermal shield flexures are flexures excessive wear, fracture, or NA for MPS 3 complete separation.

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Serial No.: 23-290 Aging Management Program Attachment 1, Table 4 Docket No.: 50-423 Enclosure, Page 30 of 32 Table 53 (continued)

Westinghouse Plants Examination Acceptance and Expansion Criteria Notes to Table 53:

1. The examination acceptance criterion for visual examination is the absence of the specified relevant condition(s).

2. The lower core barrel flange weld may alternatively be designated as the core barreltosupport plate weld in some Westinghouse plant designs.

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Serial No.: 23-290 Docket No.: 50-423 AGING MANAGEMENT PROGRAM Enclosure, Page 31 of 32 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 AGING MANAGEMENT PROGRAM Enclosure, Page 32 of 32 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.