ML20225A152
| ML20225A152 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/11/2020 |
| From: | Dennis Galvin NRC/NRR/DORL/LPL4 |
| To: | Hicks J Luminant Generation Co |
| References | |
| L-2020-LLR-0095, L-2020-LLR-0096 | |
| Download: ML20225A152 (8) | |
Text
From: Galvin, Dennis Sent: Tuesday, August 11, 2020 3:38 PM To: Jack Hicks (Jack.Hicks@luminant.com)
Cc: Dixon-Herrity, Jennifer; Barnette, James; Buford, Angela
Subject:
Verbal Authorization of Comanche Peak Unit 1 Relief Requests SNB-1 and V-3 (L-2020-LLR-0095 and L-2020-LLR-0096)
Attachments: L-2020-LLR-0095 Comanche Peak Unit 1 COVID Snubber Testing Verbal Authorization 2020-08-11.pdf; L-2020-LLR-0096 Comanche Peak Unit 1 COVID IST Interval Extension Verbal Authorization 2020-08-11.pdf
- Jack, Please find the attached written documentation of the verbal authorizations for Comanche Peak Unit 1 Relief Request Relief Requests SNB-1 and V-3. Relief Request SNB-1 is addressed in the first attachment. Relief Request V-3 is addressed in the second attachment.
If you have any questions, please contact me at (301) 415-6256 or Dennis.Galvin@nrc.gov.
Respectfully, Dennis Galvin Project Manager U.S Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operating Reactor Licensing Licensing Project Branch 4 301-415-6256 Docket No. 50-445
Hearing Identifier: NRR_DRMA Email Number: 714 Mail Envelope Properties (SA9PR09MB55197A00EA97EFE4C4E1C727FB450)
Subject:
Verbal Authorization of Comanche Peak Unit 1 Relief Requests SNB-1 and V-3 (L-2020-LLR-0095 and L-2020-LLR-0096)
Sent Date: 8/11/2020 3:38:17 PM Received Date: 8/11/2020 3:38:00 PM From: Galvin, Dennis Created By: Dennis.Galvin@nrc.gov Recipients:
"Dixon-Herrity, Jennifer" <Jennifer.Dixon-Herrity@nrc.gov>
Tracking Status: None "Barnette, James" <James.Barnette@luminant.com>
Tracking Status: None "Buford, Angela" <Angela.Buford@nrc.gov>
Tracking Status: None "Jack Hicks (Jack.Hicks@luminant.com)" <Jack.Hicks@luminant.com>
Tracking Status: None Post Office: SA9PR09MB5519.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 638 8/11/2020 3:38:00 PM L-2020-LLR-0095 Comanche Peak Unit 1 COVID Snubber Testing Verbal Authorization 2020-08-11.pdf 158046 L-2020-LLR-0096 Comanche Peak Unit 1 COVID IST Interval Extension Verbal Authorization 2020-08-11.pdf 155748 Options Priority: Normal Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
VERBAL AUTHORIZATION BY THE NRC OFFICE OF NUCLEAR REACTOR REGULATION FOR 10 CFR 55.55a REQUEST SNB-1, SNUBBER TESTING COMANCHE PEAK NUCLEAR POWER PLANT, UNIT 1 VISTRA OPERATIONS COMPANY LLC DOCKET NO. 50-445 August 11, 2020 Technical Evaluation read by Angela Buford, Branch Chief, Mechanical Engineering and Inservice Testing Branch, Division of Engineering and External Hazards, NRC Office of Nuclear Reactor Regulation By electronic submittal dated July 14, 2020 (Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML20196L873), as supplemented by letter dated August 5, 2020 (ADAMS Accession No. ML20218A853), Vistra Operations Company LLC (the licensee) proposed an alternative to specific inservice testing (IST) program requirements for snubbers (also referred to as dynamic restraints) in the 2004 Edition through 2006 Addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), for Comanche Peak Nuclear Power Plant (Comanche Peak),
Unit 1, pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 55a (10 CFR 50.55a).
In particular, the licensee submitted 10 CFR 50.55a Request SNB-1, on July 14, 2020, which included a request for authorization by the U.S. Nuclear Regulatory Commission (NRC) for a one-time alternative associated with performing snubber functional testing activities from the upcoming refueling outage 1RF21 in the fall 2020 to refueling outage 2RF22 planned for the spring of 2022 due to Coronavirus Disease (referred to as COVID-19) issues for specific snubbers listed in the alternative request.
In its submittal dated July 14, 2020, the licensee provided justification that compliance with the provisions in the ASME OM Code, Subsection ISTD, Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants, paragraphs ISTD-5200, Inservice Operational Readiness Testing, ISTD-5240, Test Frequency, and ISTD-5260, Testing Sample Plans, as incorporated by reference in 10 CFR 50.55a, to conduct snubber functional testing during the upcoming refueling outage would result in a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2).
The licensee considered that the performance of testing of the specific snubbers identified in Request SNB-1, at this time, would represent a hardship due to the occupational health and safety concerns associated with pandemic-related issues pertaining to the COVID-19 outbreak.
For example, the licensee indicated that IST activities, including those related to snubbers, involve close contact with personnel working in tight spaces and thereby limit social distancing capabilities. The licensee also stated that the close contact required to perform the IST activities could be detrimental to the occupational health and safety of the workforce and result in the potential to spread the virus.
In its request, the licensee stated that the 37 Testing Sample Plan, as described in paragraph ISTD-5260 of the ASME OM Code, is used for the balance of the mechanical snubbers specified in Request SNB-1. To support its request, the licensee stated that each snubber listed in its submittal has a predicted service life (2030) determined for refueling outage 1RF20 that is significantly beyond the spring 2022 refueling outage. The licensee stated that this alternative request does not impact the implementation of its service life monitoring (SLM) program required by the ASME OM Code, Subsection ISTD, paragraph ISTD-6000, Service Life Monitoring.
In its request, the licensee reported that in the last 10 years, 241 snubbers have been tested with only snubber CC-1-007-034-A63K failing its test during refueling outage 1RF18 in the fall of 2016. The licensee evaluated this failure and took corrective action as described in Condition Report CR-2016-004366 per the Comanche Peak corrective action program. The licensee reported that its evaluation of this failure concluded that the piping system to which the snubber is attached remained within its design parameters and would have fulfilled its safety functions.
The licensee replaced snubber CC-1-007-034-A63K as part of its corrective action. Due to this snubber failure, the licensee randomly selected 19 more snubbers from the Pacific Scientific (PSA) mechanical snubber population at Comanche Peak, Unit 1, and tested for operational readiness as required by the ASME OM Code, paragraphs ISTD-5412, Additional Sample Size, and ISTD-5420, The 37 Testing Sample Plan Additional Testing. The licensee reported that all of these 19 snubbers passed their tests with no additional failures.
In its letter dated August 5, 2020, the licensee stated that each snubber in the scope of this request will remain within the predicted service life interval, in accordance with ASME OM Code, Subsection ISTD, paragraph ISTD-6100, Predicted Service Life, through refueling outage 1R22 as predicted for refueling outage 1RF20 at Comanche Peak, Unit 1. Further, the licensee stated that all of the 1443 snubbers at Comanche Peak, Unit 1 in the scope of the ASME OM Code, which includes the 37 snubbers listed in Table 1 of the request, were visually inspected during refueling outages 1RF19 and 1RF20, with exception of snubbers CS-1-258-700-A53K and CS-1-259-700-A53K, which were not inspected due to their location in a Locked High Radiation Area. These two snubbers are currently scheduled to be inspected in September of 2020. Out of 1441 snubbers visually inspected, only snubber CC-1-043-013-A43K had unsatisfactory results. The licensee prepared Condition Report CR-2019-003979 and determined that the noted condition had no impact on the operating system. Based on the entire snubber population and the current level of acceptable performance, the licensee considered that there is reasonable assurance that each snubber will continue to be operationally ready to perform its safety functions during the interval of this proposed alternative.
In its request, the licensee stated that the Comanche Peak, Unit 1 operational readiness test history during the past 10 years shows that the snubber population is well maintained within the examination, testing, and SLM program. The licensee reported that there are no planned changes to the snubber environments or operating conditions that would affect the snubbers in a different manner than represented during the past surveillance testing. The licensee indicated that no deficiencies, adverse trends, or maintenance work orders have been identified that would impact or degrade any snubbers performance capability.
Based on the information described above for Comanche Peak, Unit 1, the NRC staff finds that (1) snubber population testing during past ten years indicates the acceptable historical performance of the snubbers within the scope of Request SNB-1; (2) ongoing inservice testing activities have not identified snubber performance concerns, with specific snubber test failures
resolved by the corrective action program; (3) the licensee is implementing its SLM program for all snubbers every refueling outage, and service life maintenance activities are not modified by this request; and (4) a hardship exists for the functional testing of these snubbers at this time that would be contrary to the health and safety of plant personnel.
Therefore, the NRC finds that the licensees proposed alternative for a one-time Snubber Program interval extension for the testing of 37 specific snubbers at Comanche Peak, Unit 1 listed in Request SNB-1, from the upcoming refueling outage 1RF21 in the fall of 2020 to the refueling outage 1RF22 in the spring of 2022 will provide reasonable assurance that the snubbers will be operationally ready to perform their safety functions, in accordance with 10 CFR 50.55a(z)(2). All other ASME OM Code requirements as incorporated by reference in 10 CFR 50.55a for which relief or an alternative was not specifically requested and approved as part of Request SNB-1 remain applicable. If the licensee identifies a performance issue with any of these snubbers, the licensee will be expected to take action to implement the requirements of the applicable ASME OM Code. This authorization will remain in effect until restart from refueling outage 1RF22 for Comanche Peak, Unit 1 in the spring of 2022. The licensees testing plans for these snubbers may be adjusted as appropriate by any subsequent NRC-authorized alternative requests.
Authorization read by Jennifer Dixon-Herrity, Chief of the Plant Licensing Branch IV, Office of Nuclear Reactor Regulation As Chief of the Plant Licensing Branch IV, Office of Nuclear Reactor Regulation, I agree with the conclusions of the Mechanical Engineering and Inservice Testing Branch.
The NRC staff concludes that the proposed alternative for Comanche Peak, Unit 1 will provide reasonable assurance of adequate safety until the next scheduled refueling outage in the spring of 2022 when the testing and examinations of the specific snubbers identified in Request SNB-1 will be performed.
The NRC staff finds that complying with the requirements of the ASME OM Code, as required by 10 CFR 50.55a, 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 all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, effective August 11, 2020, the NRC authorizes the use of the proposed alternative at Comanche Peak, Unit 1 until completion of the next scheduled refueling outage, scheduled for the spring of 2022. All other requirements in ASME OM Code for which relief or an alternative was not specifically requested and approved as part of this request remain applicable.
This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding the proposed alternatives while subsequently preparing the written safety evaluation.
VERBAL AUTHORIZATION BY THE NRC OFFICE OF NUCLEAR REACTOR REGULATION FOR 10 CFR 50.55a REQUEST V-3, REVISION 0, INSERVICE TESTING INTERVAL EXTENSION COMANCHE PEAK NUCLEAR POWER PLANT, UNIT 1 VISTRA OPERATIONS COMPANY LLC DOCKET NO. 50-445 AUGUST 11, 2020 Technical Evaluation read by Angela Buford, Branch Chief, Mechanical Engineering and Inservice Testing Branch, Division of Engineering and External Hazards, NRC Office of Nuclear Reactor Regulation By electronic submittal on July 14, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20196L876), Vistra Operations Company LLC (the licensee) proposed an alternative to specific inservice testing (IST) requirements in the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code), 2004 Edition through 2006 Addenda, for Comanche Peak Nuclear Power Plant (Comanche Peak), Unit 1, pursuant to Title 10 of the Code of Federal Regulations, Part 50, Section 55a (10 CFR 50.55a). In particular, the licensee submitted 10 CFR 50.55a Request V-3, Revision 0, on July 14, 2020, which included a request for authorization by the U.S. Nuclear Regulatory Commission (NRC) for a one-time alternative associated with performing IST Program valve examination and testing activities from the upcoming refueling outage 1RF21 to begin on October 18, 2020, to the next refueling outage 1RF22 scheduled to occur in the spring of 2022 for 4 specific check valves and 9 specific relief valves at Comanche Peak, Unit 1, listed in the request.
In its request on July 14, 2020, the licensee provided justification that compliance with the provisions in ASME OM Code, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, paragraph ISTC-5221, Valve Obturator Movement, subparagraph (c), as incorporated by reference in 10 CFR 50.55a, to perform a sample disassembly examination program for check valves that are impractical to test per subparagraphs (a) or (b) every 2 years; and the provisions in ASME OM Code, Mandatory Appendix I, Inservice Testing of Pressure Relief Devices in Light-Water Reactor Nuclear Power Plants, paragraph I-1320, Test Frequencies, Class 1 Pressure Relief Valves, subparagraph (a), 5-Year Test Interval, and paragraph I-1350, Test Frequency, Classes 2 and 3 Pressure Relief Valves, subparagraph (a), 10-year Test Interval, as incorporated by reference in 10 CFR 50.55a, to conduct testing of relief valves, during the upcoming refueling outage would result in a hardship without a compensating increase in the level of quality and safety in accordance with 10 CFR 50.55a(z)(2). The licensee considered that the performance of disassembly examination for the 4 specific check valves and performance of testing of the 9 specific relief valves identified in its request at this time would represent a hardship due to the occupational health and safety concerns associated with pandemic-related issues pertaining to the Coronavirus Disease 2019 (referred to as COVID-19) outbreak. For example, the licensee indicated that IST activities involve close contact with personnel working in tight spaces and 1
thereby limit social distancing capabilities. The licensee also stated that the close contact required to perform the IST activities could be detrimental to the occupational health and safety of the workforce and result in the potential to spread the virus.
In its request, the licensee reported that its review of the maintenance and test history for the 4 specific check valves at Comanche Peak, Unit 1, showed that the valves had no deficiencies, adverse trends, or open maintenance work orders identified that would impact or degrade each valves performance capability and exclude it from this interval extension request. The licensee reported that the most recent two disassembly examinations for each of the specific check valves at Comanche Peak, Unit 1, verified that the valve internals were structurally sound, and the visual inspection of the valve body, disc, and seat were completed with satisfactory results.
The licensee proposed that these results support an extension of the disassembly examination test interval for these check valves.
The licensee reported that the 7 Main Steam Safety Valves specified in its request are part of a 20-member valve group that is tested in accordance with ASME OM Code, Mandatory Appendix I, subparagraph I-1320(a), which requires a 5-year test interval with a 20 percent sample of the valves in the group to be tested every 24 months. The licensee indicated that over the last two 5-year testing cycles for the 20-member valve group, no additional valves were required to be tested by the acceptance criteria in subparagraph I-1320(c), Requirements for Testing Additional Valves, in Mandatory Appendix I. Of the valves in this 20-member valve group that needed adjustment based on the as-found values, the licensee reported that extrapolation of the setpoint drift would not have reached the as-left acceptance criteria for any of these valves. For these valves listed in the request, the licensee had not identified any deficiencies, adverse trends, or maintenance work orders that would impact or degrade the valves performance capability.
The licensee indicated that two of the relief valves listed in its request are tested in accordance with ASME OM Code, Mandatory Appendix I, paragraph I-1350(a), which requires a 10-year test interval with 20 percent sample of the valves in the group to be tested every 48 months.
With respect to relief valve 1DO-0111, the licensee reported that this valve has performed consistently during the last two successive tests without the need for adjustment with a set pressure drift of only 1 psi over a 9-year interval. With respect to relief valve 1-8855A, the licensee reported that this relief valve has performed consistently over the previous test interval with virtually no set point drift. For these valves listed in its request, the licensee had not identified any deficiencies, adverse trends, or maintenance work orders that would impact or degrade the valves performance capability.
Based on the information described above for the 4 specific check valves and 9 specific relief valves at Comanche Peak, Unit 1, identified in the licensees request, the NRC staff finds that (1) previous testing of these check and relief valves indicate their acceptable historical performance; (2) no current concerns with the performance of these check and relief valves have been identified; (3) periodic maintenance activities are not modified by this request; and (4) a hardship exists for the performance of team-oriented testing of these check and relief valves at this time that would be contrary to the health and safety of plant personnel.
Therefore, the NRC finds that the licensees proposed alternative in its request on July 14, 2020, for a one-time extension of the IST examination and testing interval for the 4 specific check valves and 9 specific relief valves at Comanche Peak, Unit 1, identified in its request, in accordance with 10 CFR 50.55a(z)(2), will provide reasonable assurance that the check and relief valves will be operationally ready to perform their safety functions until the next refueling 2
outage in the spring of 2022. All other ASME OM Code requirements as incorporated by reference in 10 CFR 50.55a for which relief or an alternative was not specifically requested and approved as part of this request remain applicable. If the licensee identifies a performance issue with any of these check or relief valves, the licensee will be expected to take action to implement the requirements of its Technical Specifications. This authorization will remain in effect until restart from the next refueling outage for Comanche Peak, Unit 1, in the spring of 2022. The licensees examination and testing plans for these check and relief valves may be adjusted as appropriate by any subsequent NRC-authorized alternative requests.
Authorization read by Jennifer Dixon-Herrity, Chief of the Plant Licensing Branch IV, Office of Nuclear Reactor Regulation As Chief of the Plant Licensing Branch IV, Office of Nuclear Reactor Regulation, I agree with the conclusions of the Mechanical Engineering and Inservice Testing Branch.
The NRC staff concludes that the proposed alternative for Comanche Peak, Unit 1 will provide reasonable assurance of adequate safety until the next scheduled refueling outage in the spring of 2022 when disassembly examination of 4 specific check valves and the testing of 9 specific relief valves will be performed.
The NRC staff finds that complying with the requirements of the ASME OM Code, as required by 10 CFR 50.55a, 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 all of the regulatory requirements set forth in 10 CFR 50.55a(z)(2).
Therefore, effective August 11, 2020, the NRC authorizes the use of the proposed alternative at Comanche Peak, Unit 1 until completion of the next scheduled refueling outage, scheduled for the spring of 2022. All other requirements in ASME OM Code for which relief or an alternative was not specifically requested and approved as part of this request remain applicable.
This verbal authorization does not preclude the NRC staff from asking additional clarification questions regarding the proposed alternative while subsequently preparing the written safety evaluation.
3