ML24107B120
| ML24107B120 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 04/18/2024 |
| From: | Jeffrey Whited NRC/NRR/DORL/LPL3 |
| To: | Lies Q Indiana Michigan Power Co, Nuclear Generation Group |
| Wall S, NRR/DORL/LPL3 | |
| References | |
| EPID L-2024-LLR-0026 | |
| Download: ML24107B120 (1) | |
Text
April 18, 2024
DONALD C COOK NUCLEAR PLANT, UNIT NO. 2 - AUTHORIZATION AND SAFETY EVALUATION FOR ALTERNATIVE REQUEST NO. ISIR-5-07 (EPID: L-2024-LLR-0026)
LICENSEE INFORMATION
Recipients Name and Address : Q. Shane Lies Senior Vice President and Chief Nuclear Officer Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, MI 49106
Licensee: Indiana Michigan Power Company
Plant Name(s) and Unit(s): Donald C. Cook Nuclear Plant, Unit No. 2
Docket No.: 50-316
APPLICATION INFORMATION
Submittal Date: April 3, 2024
Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML24094A283
Applicable Inservice Inspection (ISI) Program Interval and Interval Start/End Dates:
The Donald C. Cook Nuclear Plant, Unit 2s (CNP-2) fifth 10-year ISI interval began on March 1, 2020, and is scheduled to end on February 28, 2030.
Alternative Provision: The licensee requested an alternative for CNP-2 under Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2) for the duration of one cycle of operation, until the end of the next scheduled refueling outage, U2C29.
ISI Requirement: The regulations in 10 CFR 50.55a(g)(6)(ii)(D)(1), Implementation, requires the licensee augment their ISI programs in accordance with American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Case N-729-6, subject to the conditions specified in paragraphs (2) through (8) of 10 CFR 50.55a(g)(6)(ii)(D). ASME Code Case N-729-6, paragraph -3142.2, requires that nozzles with relevant conditions indicative of possible nozzle leakage undergo repair/replacement or supplemental examinations, in accordance with paragraph -3200(b).
Paragraph -3200(b), Supplemental Examinati ons, of N-729-6 requires supplemental examinations consisting of a volumetric examination of the nozzle tube and surface examination of the partial-penetration weld or surface examination of the nozzle tube inside surface, the partial-penetration weld, and nozzle tube outside surface below the weld.
Applicable Code Edition and Addenda: CNP-2s ASME code of record for the fifth 10-year ISI interval is the 2013 Edition of ASME Code,Section XI, with no addenda.
Brief Description of the Proposed Alternative: As an alternative to performing supplemental examinations required by paragraph -3142.2, the licensee proposes performing the ASME Code Case-required bare metal visual (BMV) examination of the CNP-2 reactor vessel closure head (RVCH) during the next scheduled refueling outage, U2C29, in accordance with the latest revision of ASME Code Case N-729 endorsed in 10 CFR 50.55a. The examination will be conducted in accordance with paragraph -3140 and the results will be evaluated in accordance with paragraph -3142 to ensure that no leakage is occurring from the RVCH nozzle.
During the current outage, the licensee steam cleaned the affected areas of the RVCH surface and performed a post-cleaning BMV examination to verify the structural integrity of the RVCH.
In addition, during the upcoming cycle of operation, the licensee stated that it would monitor for leakage in a manner which will continue to ensure the structural integrity of the RVCH.
The licensee indicated that to perform the supplemental volumetric and surface examinations required by ASME Code Case N-729-6, paragraph -3200(b) would result in approximately 754 mRem (millirem) of additional dose to personnel performing these examinations. The mobilization of personnel and equipment and the completion of the supplemental examinations would take approximately 10 days.
For these reasons, pursuant to 10 CFR 50.55a(z)(2), the supplemental examinations represent a hardship or unusual difficulty without a compensating increase in the level of quality and safety. For additional details on the licensees request, please refer to the April 3, 2024, application.
STAFF EVALUATION
The licensee identified one RVCH penetration, nozzle 76, that had relevant conditions of possible nozzle leakage during the performance of the RVCH BMV examination performed during the current refueling outage, U2C28, at CNP-2. In accordance with 10 CFR 50.55a(g)(6)(ii)(D) which mandates ASME Code Case N-729-6, with conditions, the licensee would be required to perform volumetric and surface examinations of the affected penetration nozzle 76 and its associated weld. As a proposed alternative, under 10 CFR 50.55a(z)(2), the licensee requests to monitor for reactor coolant pressure boundary leakage during the upcoming cycle of operation to ensure structural integrity of the RVCH and perform a BMV examination of the RVCH during the next scheduled refueling outage, U2C29.
The licensee supported the proposed alternative by describing the most likely source of leakage, the location of the deposits, the final bare metal head surface condition at the penetration after cleaning, the crack resistance of the nozzle and weld materials of the RVCH at CNP-2, and the radiological dose hardship of performing the additional required supplemental examinations at CNP-2 during the current refueling outage.
The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the licensees identified hardship and found the licensees estimation of radiological dose that would be imposed on personnel necessary to perform the supplemental volumetric examinations during this outage was consistent with estimates at other facilities. As such, the NRC staff finds that the radiological dose is a hardship on the licensee consistent with 10 CFR 50.55a(z)(2).
The NRC staff reviewed the licensees proposed alternative by evaluating whether the actions identified by the licensee provide reasonable assurance of structural integrity of the RVCH for the next operating cycle without requiring the licensee to perform supplemental volumetric examinations during the current refueling out age. The NRC reviewed photographs of the initial visual examination and the as left condition of the bare metal head surface. The staff notes that the most probable source of the leakage is from the in-core instrumentation assembly seals and not related to an active leakage from the RVCH penetration 76. The NRC staff reviewed operating experience of previous leakage that may have masked penetrations in previous examinations. While there were indications of similar sources of leakage from above for other penetrations, no leakage was previously identified at penetration 76. The licensee noted that due to the repeat nature of this leakage, they have taken corrective actions to prevent reoccurrence of the leakage contacting the RVCH in the future by updating the procedures for reassembly of the seals in question. The NRC staff notes that if masking deposits at the same penetration 76 are observed during the subsequent BMV inspection, further scrutiny will be required of potential sources of leakage. Therefore, the NRC staff finds the licensees corrective actions are necessary to address this concer n for the subsequent BMV examinations. The NRC staff notes that after cleaning there was no indication of significant degradation of the RVCH surface during the current outage. Additionally, the head surface meets the requirements of ASME Code N-729-6 to allow for an effective BMV examination during the next refueling outage.
The NRC staff recognizes that the nozzle and weld material for the RVCH at CNP-2 are fabricated with alloy 690/52/152 materials which have a demonstrated resistance to the cracking mechanism of concern, primary water stress corrosion cracking (PWSCC), in pressurized-water reactor coolant system environments such as those at CNP-2. Extensive crack initiation and crack growth rate testing by NRC contractors and the industry has been performed on these materials nationally and internationally to verify their reduced susceptibility to PWSCC, with no indication of cracking in similar operating environments as those at CNP-2. Confirmatory crack initiation testing has been performed by the NRC s Office of Nuclear Regulatory Research simulating over 20 years of operation in similar environments at CNP-2 with no indications of cracking. This testing supports the conclusion that it is unlikely that the source of leakage was from cracking in these materials in the CNP-2 RVCH penetration 76.
Further, if a crack had initiated and grown to a size to allow minor leakage of a RVCH penetration, the known resistance of alloy 690/52/152 to crack growth provides additional assurance that any cracking currently present would be unlikely to increase to the point of challenging the structural integrity of the RVCH over one additional operating cycle. The licensees identified leakage monitoring actions would also enable detection of the onset or increase in leakage through RVCH penetration 76 prior to it presenting a significant challenge to structural integrity of the RVCH.
Hence, the NRC finds that the licensees propo sed alternative provides reasonable assurance of the structural integrity of the RVCH for the next operating cycle at CNP-2 without requiring the licensee to perform supplemental volumetric and surface examinations during the current refueling outage. Further, there would be limited gain in quality and safety by performing the
required supplemental examinations to verify that no indications of cracking are present for the RVCH penetration 76 during the current refueling outage. Given the hardship, the NRC staff finds that (1) there is reasonable assurance that the licensees proposed alternative has a minimal impact on quality and safety; and (2) the licensees hardship justification is acceptable.
CONCLUSION
As set forth above, the NRC staff determines that the licensee has demonstrated that the proposed alternative in relief request (RR) ISIR-5-07 provides reasonable assurance of structural integrity of the subject components and that complying with the specified ASME Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(z)(2). Therefore, the NRC staff authorizes the use of RR ISIR-5-07 at CNP-2 for one cycle of operation, not to exceed the end of the next refueling outage, U2C29.
All other ASME BPV Code,Section XI, requirements for which an alternative was not specifically requested and authorized remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: Varoujan Kalikian, NRR Jay Collins, NRR
Date: April 18, 2024
Jeffrey A. Whited, Chief Plant Licensing Branch 3 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
cc: Listserv
ML24107B120 NRR-028 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DNRL/NPHP/BC NAME SWall SRohrer MMitchell DATE 04/15/2024 04/17/2024 04/12/2024 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME JWhited SWall DATE 04/18/2024 04/18/2024