ML25101A289
| ML25101A289 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 04/23/2025 |
| From: | Tony Nakanishi NRC/NRR/DORL/LPL4 |
| To: | Schuetz R Energy Northwest |
| Chawla M | |
| References | |
| EPID L-2024-LLR-0074 | |
| Download: ML25101A289 (1) | |
Text
April 23, 2025 Mr. Robert Schuetz Chief Executive Officer Energy Northwest Mail Drop 1023 76 North Power Plant Loop P.O. Box 968 Richland, WA 99352
SUBJECT:
COLUMBIA GENERATING STATION - AUTHORIZATION AND SAFETY EVALUATIONS FOR ALTERNATIVE REQUESTS 5ISI-03 AND 5ISI-04 FOR THE FIFTH INSERVICE INSPECTION INTERVAL (EPID L-2024-LLR-0075 AND EPID L-2024-0074)
Dear Mr. Schuetz:
By letter dated November 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24323A191), Energy Northwest (the licensee) submitted two relief requests to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the upcoming fifth inservice inspection (ISI) interval at Columbia Generating Station (Columbia).
Specifically, in Relief Request (RR) Nos. 5ISI-03 and 5ISI-04, the licensee, requests relief pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(z)(2), due to the determination that complying with the requirements would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff has reviewed the subject requests and concludes, as set forth in the enclosed safety evaluations, that the alternatives proposed by the licensee in RR Nos. 5ISI-03 and 5ISI-04, provide reasonable assurance of structural integrity and leak tightness of the standby liquid control system piping and reasonable assurance of structural integrity and leak tightness of the reactor pressure vessel head flange seal leak-off detection line. 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 alternative requests 5ISI-03 and 5ISI-04, at Columbia, for the fifth 12-year ISI interval, which is scheduled to begin on December 13, 2025, and is scheduled to end on December 12, 2037.
All other ASME 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.
R. Schuetz If you have any questions please contact Mahesh Chawla, at (301) 415-8371 or via email at Mahesh.Chawla@nrc.gov.
Sincerely, Tony Nakanishi, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-397
Enclosures:
- 2. Safety Evaluation for RR No. 5ISI-04 cc: Listserv TONY NAKANISHI Digitally signed by TONY NAKANISHI Date: 2025.04.23 07:03:49 -04'00' SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE REQUEST 5ISI-03 FIFTH INSERVICE INPSECTION INTERVAL ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By letter dated November 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24323A191), Energy Northwest (the licensee) submitted two relief requests to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the upcoming fifth inservice inspection (ISI) interval at Columbia Generating Station (Columbia).
Specifically, in Relief Request No. 5ISI-03, the licensee requests relief pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2), Hardship without a compensating increase in quality and safety, due to the determination that complying with the requirements would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
The ASME Code requirements applicable to this request originate in Section XI, IWC-2500 and IWA-5000.
Table IWC 2500 1, Examination Category C-H, Item No. C7.10 requires that Class 2 pressure-retaining components are subjected to a system leakage test in accordance with IWC-5220, a VT-2 visual examination in accordance with IWA-5240, and acceptance standard in accordance with IWC-3516 at each inspection period.
IWA-5213(a)(2) states, in part, that the Class 2 components in standby systems (or portions of standby systems) that are not operated routinely except for testing, a 10-minute holding time is required after attaining test pressure.
The Code of record for the fifth 12-year ISI interval at Columbia is the 2019 Edition of the ASME Code,Section XI
3.0 TECHNICAL EVALUATION
3.1 Licensees Alternative Request 5ISI-03 The standby liquid control (SLC) system piping runs from the SLC storage and test tank to the SLC pump and continues to the reactor vessel. The licensee proposed an alternative only for a section of the SLC system piping. The licensee stated that the subject pipe section is classified as ASME Code Class 2 and composed of two loops. SLC Loop A extends from pump SLC-P-1A to gate valve SLC-V-3A and continues to relief valve SLC-RV-29A. SLC Loop B extends from pump SLC-P-1B to gate valve SLC-V-3B and continues to relief valve SLC-RV-29B. For each loop, the gate valve is used to isolate the section of the system piping upstream of the SLC pump from the section of the system piping downstream of the SLC pump. The design condition for the piping upstream and downstream of the SLC pump is 150 pounds per square inch gauge (psig) and 1400 psig, respectively. Each loop consists of approximately 51/2 feet long SA 312 TP304 stainless steel Schedule 80S pipe with nominal pipe size of 1 inch and 11/2 inches, a 1-inch relief valve, a 11/2-inch check valve, and a 11/2-inch gate valve. The SLC system piping is not insulated.
The proposed alternative is to eliminate the 10-minute hold time requirement of the ASME Code,Section XI, IWA-5213(a)(2) for the section of the discharge piping from pump SLC-P-1A to valve SLC-V-3A in Loop A and from pump SLC-P-1B to valve SLC-V-3B in Loop B that cannot be pressurized using a system hydrostatic test. There are no vents, drain lines, or test connections available in the subject pipe section to connect a hydrostatic test pump. The licensee stated that use of a system hydrostatic test would require disconnecting the pumps and installing blind flanges with fittings, which require major maintenance activities and system modification. The system leakage test of the subject pipe section for each loop is performed using the SLC pump during the pump operability test. Once the SLC pump starts, the system pressure increases rapidly to 1240 psig operating pressure. The pumps will run for 3 to 5 minutes to maintain test pressure of 1240 psig to support the associated VT-2 visual examination. In lieu of the 10-minute holding time prior to the VT-2, the licensee proposed to perform the VT-2 during the 3 to 5 minutes that the SLC pump is operating.
In its evaluation, the NRC staff assessed whether the licensee provided adequate description and technical information to support the basis for a hardship or unusual difficulty if it were required to comply with the ASME Code-required system leakage test. The licensees bases for hardship are as follows:
Use of a hydro pump to pressurize the SLC piping section from SLC pumps SLC-P-1A and SLC-P-1B to gate valves SLC-V-3A and SLC-V-3B is not an option because there are no vent, drain lines, or test connections available to connect a hydro pump. Use of a hydro pump requires disconnecting pumps SLC-P-1A and SLC-P-1B and installing blind flanges with fittings. This would create major maintenance activity due to system modification and could pose safety hazards to the maintenance personnel.
Use of the SLC pump operability test to support leakage testing is limited to running the pump for only 3 to 5 minutes to avoid overheating the circulating fluid, causing erratic pump discharge, chattering relief valves, and damaging the relief valves. Operating experience has shown that chattering relief valves has caused damage to the sealing surfaces in the past.
Based on above, the NRC staff notes that the use of a hydro pump or a SLC pump with a 10-minute holding time to perform the ASME Code required system leakage test could potentially result in unnecessary system modification, damaging SLC pump and relief valves, and unnecessary repair/replacement activities. Therefore, the NRC staff finds that imposing system modification, repair, replacement, and safety hazards to the maintenance personnel constitute a justifiable hardship or unusual difficulty, without a compensating increase in the level of quality and safety.
3.2
NRC Staff Evaluation
The NRC staff evaluated alternative request 5ISI-03 pursuant to 10 CFR 50.55a(z)(2). The NRC staff focused on whether compliance with the specified requirements of 10 CFR 50.55a(g), or portions thereof, would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.
In its evaluation, the NRC staff assessed whether the licensee used the highest achievable test pressure to conduct system leakage testing and the way in which the licensee adequately preformed the testing and VT-2 visual examination of the section of the discharge piping from pump to the gate valve (i.e., 51/2-feet long pipe) for leakage. The NRC staff noted that,
The pressure of 1240 psig created by the SLC pump during the pump operability test will be used to pressurize the subject section of SLC pipe, and this pressure will be maintained for 3 to 5 minutes to complete the pump operability test.
The licensee will perform the VT-2 visual examination during the 3 to 5 minutes of pump operability test to identify any through-wall leak in the system.
The NRC staff determines that the licensees alternative system leakage test is adequate because (a) the licensee will utilize a test pressure as high as 1240 psig, (b) any through-wall flaw, if existed, would reveal itself by a leak under this pressure, and (c) the licensees VT-2 visual examination performed in accordance with IWA-5240 would identify the leak, and the licensee would take appropriate corrective action. Complying with the ASME Code requirement would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.
In addition, the NRC staff considered whether the licensees proposed alternative provided reasonable assurance of structural integrity and leak tightness of the SLC system piping based on the presence or absence of known active degradation mechanisms and the significance of a leak and/or structural failure. The NRC staff notes that the material of construction is stainless steel. As such, fatigue and/or stress corrosion cracking could be potential degradation mechanisms. The plants operating experience has shown that this piping has not experienced degradation and/or through-wall leak. Therefore, it is expected that any significant degradation of the line would be detected by the leakage testing performed.
Based on the above, the NRC staff concludes that the proposed system leakage test is adequate to provide reasonable assurance of structural integrity and leak tightness of the subject pipe segment. Complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff has determined that complying with the specified requirements described in the licensees request would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The proposed alternative provides reasonable assurance of structural integrity and leak tightness of the SLC system piping. 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 alternative request 5ISI-03, at Columbia, for the fifth 12-year ISI interval, which is scheduled to begin on December 13, 2025, and is scheduled to end on December 12, 2037.
All other ASME 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 Contributor: A. Rezai, NRR Date: April 23, 2025 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVE REQUEST 5ISI-04 FOR THE FIFTH INSERVICE INSPECTION INTERVAL ENERGY NORTHWEST COLUMBIA GENERATING STATION DOCKET NO. 50-397
1.0 INTRODUCTION
By letter dated November 18, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24323A191), Energy Northwest (the licensee) submitted two relief requests to the U.S. Nuclear Regulatory Commission (NRC) for the use of alternatives to certain inservice inspection (ISI) requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the upcoming fifth inservice inspection (ISI) interval at Columbia Generating Station (Columbia).
Specifically, in Relief Request No. 5ISI-04, the licensee, requests relief pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.55a(z)(2), Hardship without a compensating increase in quality and safety, due to the determination that complying with the requirements would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
The ASME Code requirements applicable to this alternative request originate in Section XI, IWB-2500 and IWC-2500.
Table IWB-2500-1, Examination Category B-P, Item No. B15.20 requires that the Class 1 pressure-retaining components are subjected to a system leakage test in accordance with IWB-5220, a VT-2 visual examination in accordance with IWA-5240, and acceptance standard in accordance with IWB-3522. Note 3 to Examination Category B-P in Table IWB-2500-1 states that the system leakage test in accordance with IWB-5220 of the boundary specified in IWB-5222(b) shall be performed at or near the end of each inspection interval.
Table IWC 2500 1, Examination Category C-H, Item No. C7.10 requires that the Class 2 pressure-retaining components are subjected to a system leakage test in accordance with IWC-5220, a VT-2 visual examination in accordance with IWA-5240, and acceptance standard in accordance with IWC-3516 at each inspection period.
IWA-5246 states that in lieu of the requirements of IWB-5220, IWC-5220, or IWD-5220, the Class 1, 2, or 3 portion of the reactor vessel head flange seal leak detection system shall be examined using the VT-2 visual examination method. The test shall be conducted at ambient conditions after the refueling cavity has been filled to its normal refueling water level for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
3.0 TECHNICAL EVALUATION
3.1 Licensees Alternative Request 5ISI-04 The licensee stated that the Columbia reactor pressure vessel (RPV) head flange seal leak-off detection line is composed of ASME Code Class 1 and 2 piping and fittings made of carbon steel with nominal pipe size of 1 inch and 3/4 inch, respectively. The Class 1 portion extends from RPV nozzle N17 up to, and including, valves MS-V-14, MS-V-13, and MS-V-753. The Class 2 portion extends from valve MS-V-753 up to, and including, valves MS-V-764 and MS-PS-34.
The Class 2 portion is less than 2 feet long.
The licensee proposed an alternative examination frequency for system leakage testing of ASME Code Class 2 portion (i.e., at or near the end of each inspection interval in lieu of at each inspection period). Specifically, the licensee stated that it will perform the system leakage testing of the Class 2 portion jointly with the Class 1 portion in accordance with IWA-5246 at or near the end of each inspection interval when the RPV head is removed, the refueling cavity is filled to its normal refueling water level, and the minimum static pressure head in the leak-off detection line is 24.5 feet of water for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the VT-2 visual examination.
3.2
NRC Staff Evaluation
The NRC staff evaluated alternative request 5ISI-04 pursuant to 10 CFR 50.55a(z)(2). The NRC staff focused on whether compliance with the specified requirements of 10 CFR 50.55a(g), or portions thereof, would result in hardship or unusual difficulty, without a compensating increase in the level of quality and safety.
In its evaluation, the NRC staff assessed whether the licensee provided adequate description and technical information to support the basis for a hardship or an unusual difficulty if it were required to comply with the ASME Code. The licensees basis for hardship is the exposure of test personnel to additional radiation dose during performance of the leakage test of the Class 2 portion of RPV head flange seal leak-off detection line individually at the required frequency of every inspection period.
In the 2015 refueling outage, the licensee accomplished the Code-required leakage test of the Class 2 portion of the line individually by closing valve MS-V-753 and pressurizing the Class 2 line with an external source at the test connection point. The licensee did not identify any through-wall leakage in the Class 2 portion of the line during the test conducted. The licensee estimated that the radiation dose received by the test personnel during the 2015 leakage testing was 64 millirem (mrem). The licensee estimated that performing the leakage test of the Class 2 portion of the line individually each inspection period to meet the Code-required examination frequency would impose more dose than performing the leakage test of Class 1 and 2 portions of the line jointly in accordance with IWA-5246 once at or near the end of inspection interval.
Based on above, the NRC staff finds that concerns from as low as reasonably achievable (ALARA) radiation exposure for the test personnel constitutes a justifiable hardship or unusual difficulty without a compensating increase in the level of quality and safety.
In its evaluation, the NRC staff assessed whether the licensee used the highest achievable test pressure to conduct the system leakage test and the way in which the licensee adequately preformed the testing and VT-2 visual examination of Class 2 portion of the RPV head flange seal leak-off detection line. The NRC staff found that the licensee will specifically conduct the leakage test of Class 1 and 2 portions of the line jointly at or near the end of the inspection interval as follows:
The fuel cavity is filled to its normal refueling water level, the leak-off detection line are clear of air and filled with water.
The static head pressure developed in the leak-off detection line is at least 24.5 feet of water for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to the VT-2 visual examination.
The VT-2 visual examination is performed in accordance with IWA-5240 after a hold-time of at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to identify any through-wall leak in the line.
The NRC staff determines that the licensees system leakage test is adequate because the Class 1 and 2 portions of the RPV head flange seal leak-off detection line will be pressurized to a test pressure that is as high as reasonably achievable, and a VT-2 visual examination will be performed after a hold-time of at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is achieved.
In addition, the NRC staff considered whether the licensees proposed alternative provided reasonable assurance of structural integrity and leak tightness of the Columbia Class 2 portion of the RPV head flange seal leak-off detection line based on the presence or absence of known active degradation mechanisms and the significance of a leak and/or structural failure. The NRC staff notes that the material of construction of the Class 2 portion is carbon steel, and fatigue and/or stress corrosion cracking could be potential degradation mechanisms. The plants operating experience has shown that this line has not experienced degradation and/or through-wall leak. Therefore, it is expected that any significant degradation of the line would be detected by the leakage testing performed at or near the end of the inspection interval.
Furthermore, the NRC staff notes that Columbia has leakage detection capabilities such as detection of increase in drywell temperature and pressure, detection of an increase in drywell floor drain leakage, containment radiation monitors and monitoring of drywell fission products, which provide warning to the control room operator in the unlikely event of a through-wall leak in the RPV flange seal leak-off detection line concurrent with leak or failure of the RPV head flange inner seal during normal operation.
Based on above, the NRC staff concludes that the system leakage test of the Class 2 portion of the RPV head flange seal leak-off detection line performed concurrently with Class 1 portion in accordance with IWA-5246 once at or near the end of the inspection interval is adequate to provide reasonable assurance of structural integrity and leak tightness of the line. Complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
As set forth above, the NRC staff has determined that complying with the specified requirements described in the licensees request referenced above would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The proposed alternative provides reasonable assurance of structural integrity and leak tightness of the RPV head flange seal leak-off detection line. 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 alternative request 5ISI-04, at Columbia, for the fifth 12-year ISI interval which is scheduled to begin on December 13, 2025, and is scheduled to end on December 12, 2037.
All other ASME 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 Contributor: A. Rezai, NRR Date: April 23, 2025
ML25101A289 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DNRL/NPHP/BC NRR/DORL/LPL4/BC NAME MChawla (WOrders for) PBlechman MMitchell TNakanishi DATE 4/19/2025 4/18/2025 4/22/2025 4/23/2025