ML24117A248
| ML24117A248 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 05/07/2024 |
| From: | Jennivine Rankin Plant Licensing Branch IV |
| To: | Dent J Nebraska Public Power District (NPPD) |
| Wengert T, NRR/DORL/LPLIV, 415-4037 | |
| References | |
| EPID L-2023-LLR-0035 | |
| Download: ML24117A248 (1) | |
Text
May 7, 2024 COOPER NUCLEAR STATION - AUTHORIZATION AND SAFETY EVALUATION FOR RELIEF REQUEST NO. RC3-02 (EPID L-2023-LLR-0035)
LICENSEE INFORMATION Recipients Name and Address:
John Dent, Jr.
Executive Vice President and Chief Nuclear Officer Nebraska Public Power District Cooper Nuclear Station 72676 648A Avenue P.O. Box 98 Brownville, NE 68321 Licensee:
Nebraska Public Power District Plant Name and Unit:
Cooper Nuclear Station (Cooper)
Docket No.:
50-298 APPLICATION INFORMATION Submittal Date: June 27, 2023 Submittal Agencywide Documents Access and Management System (ADAMS) Accession No.: ML23178A273.
Supplement Date: January 25, 2024 Supplement ADAMS Accession No. ML24025A054 Applicable Containment Inservice Inspection (ISI) and Interval Start/End Dates: The Cooper third 10-year containment ISI interval, current interval began on April 1, 2016, and is scheduled to end on February 28, 2026.
Relief Request Provision: The licensee requested relief under Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(5)(iii), ISI program update: notification of impractical ISI code requirement.
Containment ISI Requirement: American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, subsection IWE-2313, VT-1 Visual Examinations, subparagraph (b), and table IWE-2500-1 Examination Category E-G, Pressure Retaining Bolting.
Applicable Code Edition and Addenda: ASME Code Section XI, 2007 Edition through 2008 Addenda.
Brief Description of the Relief Request: The licensees Relief Request RC3-02, corresponding to the above code requirements, is requested for the containment drywell head bolting VT-1 examinations at Cooper for the third 10-year containment ISI interval. The containment drywell head bolting was examined using a combination of both the ASME Code,Section XI direct and remote VT-1 examinations. There are 76 primary containment drywell head flange bolt assemblies that connect the drywell head flange to the containment drywell shell flange. All 76 bolts, 76 upper spherical washers, 33 nuts, and 6 lower spherical washers, which were disassembled and accessible, were examined by direct VT-1 examinations in accordance with the ASME Code. The 43 nuts and the 70 lower spherical washers, which were partially disassembled and partially accessible, were examined by remote VT-1 examinations using the ASME Code,Section XI, Code qualified Level II examiners and procedure with assistance from the in vessel visual inspection (IVVI) team and cameras, due to the physical limitations experienced.
For additional details on the licensees request, please refer to the documents located at the ADAMS Accession number identified above.
STAFF EVALUATION The licensee requested a relief related to examinations of the containment drywell head bolting at Cooper. Specifically, the request concerns the following examination requirements in table IWE-2500-1 of ASME Code,Section XI, subsection IWE.
Structural Integrity and Leak Tightness of the Containment Drywell Head The U.S. Nuclear Regulatory Commission (NRC) staff evaluated the structural integrity and leak tightness of the drywell head because the licensee is not able to visually examine certain washers and nuts of the head bolting. The NRC staff requested the licensee to discuss the design analysis or qualitative factors of the drywell head bolted connection demonstrating that structural integrity and leak tightness of the drywell is maintained, assuming the bolt connections that could not be examined are degraded. In the January 25, 2024, supplement, the licensee responded to the NRC staffs Request for Additional Information (RAI)-1 that during the reassembly at the end of each refueling outage, the drywell head flange and accessible bolts/nuts/washers are cleaned, visually examined, and threads lubricated prior to assembly as a routine maintenance practice.
The licensee stated in its supplement dated January 25, 2024, that, Once the [drywell] head is in place, the bolting is torqued in three passes to achieve the final 880 ft-lbs [foot pounds] of torque. Degraded threads in the nuts would most likely prevent the bolting assembly from achieving the required torque. The combination of torquing the head bolts to required specifications and conducting the As-Left Local Leak Rate Test (LLRT) of the flange joint each refueling outage, provides reasonable assurance joint integrity is established for the subsequent 24 months of operation.
According to the licensee, the results of the ASME Code,Section XI direct VT-1 inspections of the disassembled bolting were satisfactory.
The nuts/washers that are tack welded to the bottom side of the drywell flange were examined remotely using the General Electric Hitachi VT-1 procedure. The remote visual examination was performed when the refueling cavity was temporarily flooded for refueling operations. The licensee lowers high-definition underwater cameras calibrated to VT-1 requirements from the refueling platform and positioned to examine the accessible portions of each nut/washer including portions of the nut threads that could be observed. In addition, the licensee electronically recorded the examinations allowing for improved ability to compare examination results with future results for improved monitoring of potential degradation changes. The licensee explained that As stated in the Refueling Outage 32 (RE32) remote VT-1 data sheet VT-IWE32-22-033 comments, the VT-1 examination was considered [its] best effort because of the limited access as well as the observed surface condition related to rust, flaked, blistered, and peeled paint.
According to the licensee, The tack welded nuts/washers were not cleaned due to dose considerations and risk of introducing debris into the reactor coolant if not completely removed from the refueling cavity. The licensee estimates a radiation dose of 3.5 rem (roentgen equivalent man) would be accumulated to additionally clean the nuts/washers for little additional gain of examination coverage.
The NRC staff determined that the licensee has disassembled the drywell head, performed the examinations, reassembled the head, and performed the leakage test of the drywell without showing leakage. The licensee has demonstrated that the drywell head bolting has provided leak tightness of the drywell. Therefore, the NRC staff finds that the structural integrity and leak tightness of the drywell head are acceptable.
Leak Rate Test for the Drywell Head to Drywell Shell Flange Joint As part of evaluating the leak tightness of the drywell head, the NRC staff evaluated the licensees leak rate test. The NRC staff notes that the leak rate test is a method to demonstrate the structural integrity and leak tightness of the drywell head, which includes bolting. In the January 25, 2024, supplement, the licensee responded to the NRC staffs question RAI-2 that it conducts a LLRT in accordance with 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, Type B test, which is a pneumatic leakage test of the drywell head to drywell shell flange joint and is conducted after each refueling outage. The licensee stated that per the plant test procedure, the test volume between the drywell head flange O-rings (approximately 0.475 cubic feet (ft3)) is pressurized between 58.0 to 63.8 pounds per square inch gauge (psig) and held until the leakage rate is stabilized. The licensee reported that the leak test performed during the refueling outage, RE32, was completed satisfactory with an actual test pressure of 60.19 psig.
The licensee stated that per ASME [Code],Section XI, IWE-5223.4(b) a Type A, B, or C test in accordance with 10 CFR [Part] 50 Appendix J is considered acceptable pneumatic leakage tests following repair/replacement activities for Class MC [metal containment] components thus providing a high degree of confidence of mechanical joint integrity. The licensee clarified that the disassembly of the drywell bolting and reassembly is a maintenance activity and not a repair/replacement activity.
The licensee stated that the hold time for the leakage test is approximately 10-15 minutes until the pressure stabilizes. The licensee further stated that there is no stipulated wait time as the hold time ultimately is based on how long the test volume takes to stabilize. The leakage rate is recorded and compared to the Administrative Limit and Operability Limit. The licensee completed its leak test for refueling outage RE32 with an as-left leakage rate of 0.042 standard cubic feet per hour (scfh), well below the Administrative Limit of 0.5 scfh.
The licensee explained that if the leakage rate is greater than the Administrative Limit of less than or equal to () 0.5 scfh or the Operability Limit of 1.0 scfh, it will initiate a condition report per the plant procedure and will assess the condition in accordance with the corrective action program. The licensee will enter the corrective action process to determine the cause and take appropriate actions at that time, which could include replacing bolting, as needed. The licensee stated that before plant startup is authorized, a satisfactory test that meets the acceptance criteria of the plant procedure on the LLRT would need to be performed.
The NRC staff determined that the licensee has a plant procedure that provides the necessary provisions for the LLRT. The procedure includes the specific acceptance criteria for the leakage.
If the leakage exceeds the criteria, the event is recorded in the correction action program. The licensee will take corrective action to address the excessive leakage. The NRC staff determined that prior to the plant restart, the licensee needs to perform a satisfactory leak test. Therefore, the NRC staff finds that the licensee performs the leak test in accordance with 10 CFR Part 50, Appendix J, Type B test and therefore the leak test is acceptable.
Examination Results of the Containment Drywell Head Bolting Using Direct and Remote VT-1 Examinations During Third 10-Year Containment ISI interval The licensee utilized the ASME Code Section XI, table IWE-2500-1, direct VT-1 examinations on the drywell head bolting that was disassembled and accessible. The licensee examined all 76 bolts, 33 nuts, and 6 washers by direct VT-1 examinations and concluded that its examination results were acceptable. During a regulatory audit (ML24033A309), the NRC staff reviewed the RE32 direct VT-1 data sheet VT-IWE32-22-032 and the licensees responses to RAI-3, RAI-4, and RAI-5 of the supplement dated January 25, 2024, and confirmed that the examination results of these 76 bolts, 76 upper spherical washers, 33 nuts, and 6 lower spherical washers met the ASME Code IWE requirements. Therefore, the NRC staff considers these 76 bolts, 76 upper spherical washers, 33 nuts, and 6 lower spherical washers to be acceptable.
In addition, the licensee utilized the ASME Code Section XI, table IWE-2500-1, remote VT-1 examinations on the remaining 43 nuts and the 70 lower spherical washers that are tack welded in place to the lower side of the drywell shell flange and are partially inaccessible limiting the ability of the VT-1 qualified examinations from achieving a full code qualified examination. The licensee examined the remaining 43 nuts and the 70 lower spherical washers and concluded that its examination results were acceptable. The NRC staff reviewed the relief request and the licensees responses to RAI-3, RAI-4, and RAI-5 of the supplement dated January 25, 2024, and notes that the licensee examined these remaining nuts and lower spherical washers remotely using the ASME Code,Section XI, Code qualified Level II examiners and VT-1 procedure with assistance from the IVVI team, and a high-definition underwater camera to gain as much coverage as possible. During a regulatory audit, the NRC staff reviewed the RE32 remote VT-1 data sheet VT-IWE32-22-033 and confirms that the examination results of these remaining 43 nuts and the 70 lower spherical washers met the ASME Code IWE requirements except that exterior surfaces of tack welded nuts exhibited signs of light to heavy rust and the
paint coating to be flaking, blistering, and peeling with some signs of under-deposit corrosion.
The licensee evaluated these conditions and concluded in its response to RAI-5 that none of these conditions would reduce the cross-section thickness by more than 5 percent, which met the ASME Code IWE requirements. Therefore, the NRC staff considers the remaining 43 nuts and the 70 lower spherical washers, which are tack welded in place to the lower side of the drywell shell flange, to be acceptable.
Examination Results of the Containment Drywell Head Bolting Using Direct and Remote VT-3 Examinations During Second 10-Year Containment ISI Interval During the previous second 10-year containment ISI interval, the ASME Code,Section XI, table IWE-2500-1, VT-3 examinations of the drywell head bolting were conducted as required per the 2001 Edition/2003 Addenda to determine the general mechanical and structural condition of the bolting. The licensee examined the drywell head bolting and observed light rust and no apparent loss of material thickness. The license concluded in its responses to RAI-6 in the supplement dated January 25, 2024, that the VT-3 examination was satisfactory with no non-conformances identified. The NRC staff reviewed the relief request and the licensees responses to RAI-6, and notes that the licensee examined the disassembled bolting using the ASME Code Section XI, table IWE-2500-1, direct VT-3 examinations, and the nuts and washers that could not be disassembled using the ASME Code Section XI, table IWE-2500-1, remote VT-3 examinations.
During a regulatory audit, the NRC staff reviewed the Refueling Outage, RE28 remote VT-3 data sheet VT-F14-025 and confirms that the VT-3 examination results of the drywell head bolting met the ASME Code IWE requirements. Therefore, the NRC staff considers the drywell head bolting to be acceptable.
Summary Based on the above evaluation, the NRC staff determines that the licensee has demonstrated physical limitations to meet the 100 percent code examination requirement and adequate performance of the drywell flange pressure retaining bolting by presenting plant-specific drywell bolting examination results and operating experience to provide reasonable assurance of structural integrity and leakage integrity of the drywell flange pressure retaining bolting assembly.
The NRC staff grants the relief request for the third 10-year containment ISI interval for the containment drywell head bolting at Cooper, based on the following:
(1) Completion of qualified direct VT-1 examinations for all 76 bolts, 76 upper spherical washers, 33 nuts, and 6 lower spherical washers that were disassembled and accessible.
(2) Completion of best effort remote VT-1 examinations on the remaining 43 nuts and the 70 lower spherical washers that are tack welded in place to the lower side of the drywell shell flange and are partially inaccessible.
The NRC staff finds it reasonable to grant Relief Request No. RC3-02 using a qualified direct VT-1 examination along with a best effort remote VT-1 examination, based on its review of the plant-specific testing and examination results and operating experience provided in the request.
CONCLUSION The NRC staff has determined that the licensees relief request referenced above would provide sufficient technical basis and limitations experienced to support the determination that conformance with a Code requirement is impractical.
The NRC staff concludes that the licensee has adequately addressed the regulatory requirements set forth in 10 CFR 50.55a(g)(5)(iii) and that granting relief pursuant to 10 CFR 50.55a(g)(6)(i), Impractical ISI requirements: Granting of relief, is authorized by law and will not endanger life or property or the common defense and security, and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all regulatory requirements set forth in 10 CFR 50.55a(g)(6)(i).
Therefore the NRC staff grants Relief Request RC3-02 at Cooper, for the containment drywell head bolting for the third 10-year containment ISI interval, which began on April 1, 2016, and is scheduled to end on February 28, 2026, as stipulated in the safety evaluation summary section above.
All other ASME Code,Section XI, requirements for which a relief was not specifically requested and authorized remain applicable, including third -party review by the Authorized Nuclear Inservice Inspector.
Principal Contributors: G. Wang, NRR J. Tsao, NRR D. Dijamco, NRR Date: May 7, 2024 Jennivine K. Rankin, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation cc: Listserv Jennivine K.
Rankin Digitally signed by Jennivine K. Rankin Date: 2024.05.07 13:24:29 -04'00'
ML24117A248 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DEX/ESEB/BC NAME TByrd PBlechman ITseng DATE 04/26/24 4/29/2024 04/18/24 OFFICE NRR/DNRL/NVIB/BC NRR/DORL/LPL4/BC NAME ABuford JRankin DATE 05/02/24 05/07/24