RS-24-090, Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds
| ML24257A032 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point, Limerick, LaSalle  |
| Issue date: | 09/12/2024 |
| From: | Para W Constellation Energy Generation |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| RS-24-090, NMP2L2887, EPID - L-2024-LLR-0031 | |
| Download: ML24257A032 (1) | |
Text
200 Energy Way Kennett Square, PA 19348 www.ConstellationEnergy.com 10 CFR 50.55a RS-24-090 NMP2L2887 September 12, 2024 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374 Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353 Nine Mile Point Nuclear Station, Unit 2 Renewed Facility Operating License No. NPF-69 NRC Docket No. 50-410
Subject:
Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds
References:
- 1) Letter from Dave T. Gudger (Constellation Energy Generation, LLC) to U.S.
Nuclear Regulatory Commission, Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds, dated May 2, 2024 (ML24123A119)
- 2) Electronic mail message from Scott P. Wall (U.S. Nuclear Regulatory Commission) to Jesse Brown (Constellation Energy Generation, LLC),
FINAL RAI - Constellation Energy Generation, LLC - Fleet Request -
Proposed Alternative Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds (L-2024-LLR-0031) dated August 21, 2024 (ML24234A277)
By letter dated May 2, 2024 (Reference 1), and in accordance with 10 CFR 50.55a(z)(1),
Constellation Energy Generation, LLC (CEG) requested approval of a proposed alternative for examination of Reactor Pressure Vessel circumferential shell welds for the following stations:
LaSalle County Station, Units 1 and 2 Limerick Generating Station, Units 1 and 2 Nine Mile Point Nuclear Station, Unit 2
Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds September 12, 2024 Page 2 By email dated August 21, 2024 (Reference 2), the U.S Nuclear Regulatory Commission requested additional information necessary to complete the requested review.
The attachment to this letter provides the request for additional information as well as the CEG response.
There are no regulatory commitments contained in this letter.
If you have any questions concerning this letter, please contact Jesse Brown at 267-533-6355.
Respectfully, Wendi Para Sr. Manager - Licensing Constellation Energy Generation, LLC
Attachment:
Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds cc:
Regional Administrator, NRC Region I Regional Administrator, NRC Region III USNRC Senior Resident Inspector, LaSalle County Station USNRC Senior Resident Inspector, Limerick Generating Station USNRC Senior Resident Inspector, Nine Mile Point Nuclear Station NRC Project Manager - LaSalle County Station NRC Project Manager - Limerick Generating Station NRC Project Manager - Nine Mile Point Nuclear Station Illinois Emergency Management Agency - Division of Nuclear Safety A. L. Peterson - NYSERDA
ATTACHMENT Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds LaSalle County Station, Units 1 and 2 NRC Docket Nos. 50-373 and 50-374 Limerick Generating Station, Units 1 and 2 NRC Docket Nos. 50-352 and 50-353 Nine Mile Point Nuclear Station, Unit 2 NRC Docket No. 50-410
Attachment Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds Page 1 of 2 RAI-NVIB
=
Background===
CEG has previously utilized BWRVIP-74-A, BWR Vessel and Internals Project BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal, or BWRVIP-05, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations, to obtain examination relief of reactor pressure vessel (RPV) beltline circumferential welds and to address embrittlement of RPV beltline axial welds during extended plant operation. The NRC safety evaluation (SE) of BWRVIP-74-A issued October 18, 2001 (ML012920549), specified actions to obtain examination relief of RPV beltline circumferential welds and to monitor embrittlement of RPV beltline axial welds, both of which were based on results of probabilistic fracture mechanics (PFM) analyses in the staffs SE and supplemental SE of BWRVIP-05.
Applicable Regulation and Guidance The NRC has established requirements in 10 CFR Part 50 to protect the structural integrity of structures and components in nuclear power plants. Among these requirements are the inservice inspection (ISI) requirements of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a to ensure that adequate structural integrity of the reactor pressure vessels (including their welds) is maintained through the service life of the vessels. Therefore, the regulatory basis for the following request for additional information has to do with demonstrating that the alternative request from the ISI requirements would ensure adequate structural integrity of the licensees subject reactor pressure vessel circumferential welds, and thereby would provide an acceptable level of quality and safety per 10 CFR 50.55a(z)(1).
Request for Additional Information The licensees referenced probabilistic and fracture analyses in BWRVIP-329-A to demonstrate that RPVs in the subject units have margin against failure and provides justification for relief from the examination of RPV circumferential welds. The licensee presented plant-specific information to demonstrate that the referenced analyses in BWRVIP-329-A would bound the subject circumferential welds until the end of the current licenses.
The NRCs SE of BWRVIP-74-A specified actions to obtain examination relief of RPV beltline circumferential welds based on results of PFM analyses in the staffs SE and supplemental SE of BWRVIP-05. Specifically, NRCs SE of BWRVIP-74-A, Section 4.1: Renewal Applicant Action Item (11) states, and (2) that they have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the staffs FSER [Final Safety Evaluation Report]. This is in addition to providing justification that RPV circumferential welds will satisfy the limiting conditional failure frequency in Action Item 11, which the staff noted is the action being addressed by the licensees alternative request through the use of BWRVIP-329-A. The NRC staff also noted in its SE for the Limerick license renewal application (ML14276A156), Section 4.2.6, that the Limerick units will use the same operator procedures and training in the period of extended operation that has been the practice during
Attachment Response to Request for Additional Information - Relief Request Concerning Extension of Permanent Relief from Ultrasonic Examination of Reactor Pressure Vessel Circumferential Shell Welds Page 2 of 2 the original licensing period. However, the staff noted that the licensee did not provide sufficient information that operator training has been implemented for LaSalle and NMP-2.
RAI-NVIB-01 Address licensees plan or confirm licensees activities for implementing operator training and established procedures for limiting the frequency of cold overpressure events for LaSalle for 54 EFPY and NMP-2 for 54 EFPY.
RESPONSE
LaSalle Units 1 and 2 previously applied for relief from the circumferential weld inservice inspection requirements by letter dated June 25, 2003 (ML031880829). The NRC authorized this relief request by Safety Evaluation dated January 28, 2004 (ML033300147).
Nine Mile Point Unit 2 previously applied for relief from the circumferential weld inservice inspection requirements by letter dated November 16, 2006 (ML063320461) and supplemented by letter dated April 20, 2007 (ML071210245). The NRC authorized this relief request by Safety Evaluation dated November 5, 2007 (ML072830047).
These previously authorized relief requests were based upon compliance with the requirements of BWRVIP-05, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05) and Generic Letter 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds. Criterion (2) of Generic Letter 98-05 required licensees to demonstrate that they have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the staffs safety evaluation (ML20236V555). Criterion (2) of Generic Letter 98-05 is addressed by LaSalle Units 1 and 2 and Nine Mile Point Unit 2 in Section 5 of the relief requests dated June 25, 2003, and November 16, 2006, respectively.
This operator training and procedures to limit the frequency of cold overpressure events will continue during the period of extended operation, through 54 effective full power years, for LaSalle Units 1 and 2 and Nine Mile Point Unit 2 as required by the staffs approval of the original relief requests.
Additionally, LaSalle Units 1 and 2 and Nine Mile Point Unit 2 technical specifications include requirements to maintain compliance with the operating limit curves from the pressure temperature limit report (PTLR) at all times. Compliance with the operating limit curves for hydrostatic pressure tests and leak tests, core not critical operation, and core critical operation ensures compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 providing assurance that the station will not operate in a cold overpressure condition.