L-PI-23-023, Baffle Former Bolts Alternate Aging Management Strategy

Baffle Former Bolts Alternate Aging Management Strategy

ML23254A391
Person / Time
Site:	Prairie Island
Issue date:	09/11/2023
From:	Conboy T Northern States Power Company, Minnesota
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-PI-23-023
Download: ML23254A391 (1)

Text

1717 Wakonade Drive Welch, MN 55089 September 11, 2023 L-PI-23-023 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Docket Nos. 50-282 and 50-306 Renewed Facility Operating License DPR-42 and DPR-60 Prairie Island Unit 1 and Unit 2 Baffle Former Bolts Alternate Aging Management Strategy

References:

1) NSPM Letter L-PI-22-037, Updated Approach for Prairie Island Unit 1 and Unit 2 Baffle Former Bolts, dated September 20, 2022.

(NRC ADAMS Accession Number ML22263A469)

2) NSPM Letter L-PI-19-001, Voluntary Submittal of Plant-Specific Evaluation to Extend the Re-Inspection Interval for Baffle Former Bolts, dated May 24, 2019. (NRC ADAMS Accession Number ML19144A244)

3) EPRI MRP-227, Revision 1-A, Materials Reliability Program:

Pressurized Water Reactor Internals Inspection and Evaluation Guidelines. (NRC ADAMS Accession Number ML19339G350)

In Reference 1, Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy (hereafter NSPM), informed the NRC that Prairie Island Unit 1 and Unit 2 would be performing proactive replacement of an engineered pattern of baffle former bolts (BFBs) in 2023 and 2024 in lieu of continued management by inspection for degradation as described in Electric Power Research Institute (EPRI) report MRP-227, Revision 1-A. Because the BFBs are a primary examination item for management of irradiation-assisted stress corrosion cracking (IASCC) and fatigue, it becomes necessary to devise an alternate aging management strategy for these mechanisms, in the unavailability of the normal primary item. NSPM and the original equipment manufacturer (Westinghouse) have used the guidance in EPRI MRP-227 Revision 2, Appendix C (currently under NRC review) to develop such a strategy, and the purpose of this letter is to communicate the results of this effort.

Document Control Desk L-PI-23-023 Page 2 As noted in Reference 3:

If components are repaired, modified or replaced such that the effects of aging are mitigated, then the demonstration of the adequacy of repair, replacement, or modification activities is the responsibility of the owner. In addition, repair, replacement or modification activities may warrant revision to the scope and/or frequency of the generic requirements stated in these guidelines. This includes re-establishing the technical basis for the replaced components (if not fully mitigated) and the technical basis for examination (and reinspection interval) of any linked Expansion components, which was developed on the basis of expert panel solicitation. Individual utilities will be responsible for the technical justification of such activities to demonstrate their acceptability for different requirements than those stated in these guidelines.

The probabilistic model provided with Reference 2 was used to determine whether in the upcoming 2023 and 2024 refueling outages, expansion to the lower support column (LSC) bolts and/or barrel bolts would have likely been triggered, given the known rates and patterns of BFB indications from prior ultrasonic (UT) examination. Westinghouse determined that expansion to barrel bolts would not have been triggered (95%

probability of not triggered). Westinghouse could not rule out expansion to LSC bolts, and LSC bolts were proposed and evaluated for elevation to the new primary component for management of IASCC in the fasteners of the lower internals. Since the LSC bolts have already been evaluated as being the logical successor to BFBs under the waterfall methodology of the following EPRI reports:

MRP-134, Materials Reliability Program: Framework and Strategies for Managing Aging Effects in Reactor Internals, MRP-175, "Materials Reliability Program: PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values, MRP-191, Materials Reliability Program: Screening, Categorization and Ranking of Reactor Internals of Westinghouse and Combustion Engineering PWR Designs, and MRP-232, Materials Reliability Program: Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals, it was shown that they met the criteria in Appendix C of MRP-227, Revision 2, for elevation to primary examination item. The timing for the examination will be adopted from the LSC bolt expansion component entry from the existing MRP-227 table 5-3, as within three 18-month refueling cycles, which for a plant on 24-month cycles equates to within 2 cycles. If no degradation is identified by UT of the LSC bolts during the initial examination, then reevaluation of returning BFB to the Primary examination item will be conducted at that time, rather than continuing with LSC bolt UT exams in the long term.