ML24087A011

NRR E-mail Capture - Draft RAIs for Reactor Coolant Pump (RCP) Relief Request (L-2024-LLR-0022)

ML24087A011
Person / Time
Site:	North Anna
Issue date:	03/23/2024
From:	Ed Miller NRC/NRR/DORL/LPL2-1
To:	Sinha S Dominion Generation
References
L-2024-LLR-0022
Download: ML24087A011 (6)

Text

From: Ed Miller Sent: Saturday, March 23, 2024 2:51 PM To: Shayan.Sinha@dominionenergy.com

Subject:

Draft RAIs for RCP Relief Request (L-2024-LLR-0022)

Attachments: DRAI North Anna N1_I5_NDE-007.docx

Mr. Sinha, Attached is the NRC staffs draft RAI for the subject request. The questions are being transmitted to you to determine 1) If the questions clearly convey the NRC information needs, 2)

Whether the regulatory basis for the questions are clear, and 3) If the information has already been provided in existing docketed correspondence. Additionally, review of the draft question will allow you to determine what response time you can support. After youve had a chance to review, please let me know if you would like to have a clarification call or public meeting to discuss. Thank you.

Ed Miller (301) 415-2481 Hearing Identifier: NRR_DRMA Email Number: 2441

Mail Envelope Properties (SA1PR09MB748726183716D198F45E06F1E9302)

Subject:

Draft RAIs for RCP Relief Request (L-2024-LLR-0022)

Sent Date: 3/23/2024 2:50:30 PM Received Date: 3/23/2024 2:50:00 PM From: Ed Miller

Created By: Ed.Miller@nrc.gov

Recipients:

"Shayan.Sinha@dominionenergy.com" <Shayan.Sinha@dominionenergy.com>

Tracking Status: None

Post Office: SA1PR09MB7487.namprd09.prod.outlook.com

Files Size Date & Time MESSAGE 639 3/23/2024 2:50:00 PM DRAI North Anna N1_I5_NDE-007.docx 37298

Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

DRAFT REQUEST FOR ADDITIONAL INFORMATION ALTERNATIVE REQUEST N1-I5-NDE-007 ALTERNATIVE EXAMINATION OF REACTOR COOLANT PUMP VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION, UNIT 1 DOCKET NO. 50-338 L-2024-LLR-0022

1.0 INTRODUCTION

By letter dated March 22, 2024 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML24082A274), Virginia Electric and Power Company (Dominion Energy Virginia, the licensee) requested relief from the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section XI, Subsection IWB-2430, for North Anna Power Station Unit 1. The proposed alternative would move the inspection of a reactor coolant pump from the Spring 2024 refueling outage to the Fall 2025 refueling outage.

Pursuant to Title 10, Code of Federal Regulations, Part 50, 10 CFR 50.55a(z)(2), the licensee submitted for Nuclear Regulatory Commission (NRC) review and approval Alternative Request N1-I5-NDE-007 for North Anna, Unit 1.

The NRC staff requests the following additional information (RAI) to complete its review of the alternative request.

2.0 REGULATORY EVALUATION

The inservice inspection (ISI) of ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code and applicable editions and addenda as required by 10 CFR 50.55a(g), Preservice and inservice inspection requirements.

Pursuant to 10 CFR 50.55a(g)(4), lnservice inspection standards requirement for operating plants, ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year ISI interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a)(1)(ii), 18 months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b)(2).

Pursuant to 10 CFR 50.55a(z), Alternatives to codes and standards requirements, alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

3.0 REQUEST FOR ADDITIONAL INFORMATION

1. Section 5 of Attachment 1 to the March 22, 2024 submittal states that This request proposes an alternative to performing the additional examination of an RCP casing during the current outage in accordance with IWB-2430(a)(1)(a) to performing the additional examination of an RCP casing during the next refueling outage. The next scheduled pump replacement is currently planned for the next refueling outage, N1R31 (Fall 2025) :

(1) Clarify whether the additional examination will be performed on either 1-RC-P-1B RCP or 1-RC-P-1C RCP for the Fall 2025 refueling outage.

(2) Some parts of the submittal mention the pump refurbishment. Clarify whether pump replacement means the refurbishment of hydraulic parts of the RCP such as impeller, turning vane diffuser, diffuser adapter, and casing.

(3) Discuss whether pump casing will be replaced as part of periodic pump refurbishment.

2. Section 6 of Attachment 1 to the March 22, 2024 submittal states that The proposed alternative is requested to complete the additional examination of an RCP casing as required by IWB-2430 during the fall 2025 refueling outage, which is before the end of the current 5th ISI Interval, which ends on April 30, 2029

Clarify the exact duration of the proposed alternative, i.e., whether the proposed alternative is requested for the duration up to the fall 2025 refueling outage or up to the end of the 5th ISI interval.

3. By letter dated August 24, 2020 (ML20246G703), the licensee submitted the subsequent license renewal application (SLRA) for North Anna Units 1 and 2. Section 4.7.6 of the SLRA discusses time limited aging analysis of reactor coolant pump inspection respect of ASME Code Case N-481. As a result of the degradation at 1-RC-P-1A RCP:

(1) discuss how the SLRA will be supplemented to include the subject operating experience.

(2) discuss whether Section 4.7.6 in the North Anna SLRA is still valid or it needs to be revised/updated.

4. Page 4 of Attachment 2, Engineering Evaluation ETE-NA-2024-0033, last sentence of the first paragraph states that It should also be noted that 1-RC-P-1B and 1-RC-P-1C were removed from their casings in 1982 following the identification of the broken diffuser cap screws on 1-RC-P-1A and none of the cap screws on either pump had failed but were replaced as a precaution IAW with Westinghouse Recommendations :

(1) Discuss what year was the cap screws replaced on 1 -RC-P-1B and 1-RC-P-1C pumps (i.e., were the cap screws replaced in recent years or in 1982?).

(2) If the cap screws were replaced in 1982, discuss the likelihood of failure of the cap screws in 1-RC-P-1B or 1-RC-P-1C pumps prior to the 2025 refueling outage.

5. Page 5 of Attachment 2, Engineering Evaluation ETE-NA-2024-0033, states that The interior of the casing, including the areas where all of the indications were found, does not participate in a measurable way in the hydraulic performance of the pump The NRC staff understands that RCPs have vibration limits. Discuss whether excessive vibration at RCP 1-RC-P-1A were detected by the vibration sensor or abnormal occurrence as a result of loose parts during the plant operation.

6. Page 5 of Attachment 2, states that if all of the adapter socket head cap screws had failed, significant operation degradation of the reactor coolant pump would not have resulted A loose adapter would initially drop loop flow about 0.2% which is much less than the existing flow margin of approximately 5% above core thermal design flow. In addition, the automatic low flow reactor trip would prevent operation below core thermal design flow

The NRC staff understands that a loose adapter would reduce the reactor coolant system loop flow. If the loop flow is lower than the core thermal design flow, the reactor would trip which would cause the RCP to trip. As such, there is an automatic low flow reactor trip sensor to protect the reactor operation. Also, there is a RCP vibration limit to trip RCPs. Discuss any other defense-in-depth measures that would trip the RCPs to protect RCPs, the reactor vessel, and reactor coolant system components.

7. Page 8 of Attachment 2, second paragraph discusses a Leak Before Break (LBB) analysis of the pump casing in WCAP-11517. The licensee demonstrated that a postulated through-wall crack in the pump casing remains stable based on a detailed fracture mechanics evaluation and that a gross failure of the casing is not feasible.

With respect to that, discuss the following:

(1) The size of the through wall crack based on applied loading; (2) The critical crack size of the pump casing; (3) The leakage crack size if different from the through wall crack size; (4) The reactor coolant system leakage detection capabilities (e.g., what is the smallest leak rate the sensors can detect); and (5) Whether the fracture mechanics analysis has satisfied (1) the margin of 2 between the crack size and the critical crack size and (2) margin of 10 for the leakage crack size as compared to the leak rate of the leakage detection system as specified in NRCs Standard Review Plan 3.6.3.

8. Page 8 of Attachment 2 discusses WCAP-13045 for the pump casing analysis. The NRC staff noted that WCAP-13045 is valid only for 40-year, not 80-year of plant life. The licensee also referenced WCAP-15555 which applies to the pump casing analysis involving the initial license renewal period up to 60 years of plant life. The licensee has applied to renew its operating licensee up to 80 years for North Anna units. The NRC staff noted that it has approved generic use of topical report PWROG-17033, Revision 1, Update for Subsequent License Renewal: WCAP-13045, Compliance to ASME Code Case N-481 of the Primary Loop Pump Casings of Westinghouse Type Nuclear Steam Supply Systems, for the application of subsequent license renewal up to 80 years (ML19319A188). Based on the licensees conclusion on page 13 of Attachment 2, it appears that 1-RC-P-1A pump casing will not be replaced or repaired for the remaining life of the plant.

(1) Demonstrate by analysis and/or inspection that the degraded casing will maintain its structural integrity to the end of 80-year plant life, i.e., the degraded areas of 1-RC-P-1A pump casing will not grow to challenge the structural integrity of the pump casing.

(2) Discuss whether RCP 1-RC-P-1A is acceptable for use for the 80-year plant life per PWROG-17033, Revision 1.

9. Page 10 of Attachment 2 states that the limiting material fracture toughness, JIC = 750 in-lbf/in2. With respect to that:

(1) Discuss the Japplied value based on the actual crack driving force.

(2) Discuss whether the JIC = 750 in-lbf/in2 is applicable to 80 years.

10. Pages 11 and 12 of Attachment 2 discuss fracture mechanics analysis of the impression/dent on the pump casing wall. The NRC staff noted that the licensees analysis does not address the possibility of growth of the impression/dent. Discuss whether the impression areas would grow in depth and width to a size to challenge the structural integrity of the pump casing at the end of 80 years of operation.

11. Discuss Dominion Energy efforts to address the potential for common -cause failure between pumps in terms of hardware (e.g., procurement material batches, dedication, etc.) and work practices (e.g., procedures, training, common work crews, quality control, etc.). Please include whether there is any additional information or observations of RCPs 1B and 1C that can inform whether a similar condition exists in those pumps.

12. The NRC staff noted what appears to be a minor typographic error on the first line on Pages 2 and 3 of the Attachment 1 which states Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii). Please confirm whether Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) was the intended citation.