ML24143A041
| ML24143A041 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/28/2024 |
| From: | Turner Z Plant Licensing Branch II |
| To: | Coleman J Southern Nuclear Operating Co |
| Turner, Zachary | |
| References | |
| EPID L-2024-LLA-0052 | |
| Download: ML24143A041 (8) | |
Text
May 28, 2024
Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Company 3535 Colonnade Parkway Birmingham, AL 35243
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 - EVALUATION OF SUBMITTAL FOR USING THE RISK-INFORMED PROCESS FOR EVALUATION AND SUPPLEMENTAL INFORMATION NEEDED FOR ACCEPTANCE OF LICENSE AMENDMENT REQUEST TO CHANGE TECHNICAL SPECIFICATION 3.6.5, CONTAINMENT AIR TEMPERATURE, ACTIONS (EPID L-2024-LLA-0052)
Dear Jamie Coleman:
By letter dated April 19, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24110A126), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) to change Technical Specification (TS) 3.6.5, Containment Air Temperature, actions for the Joseph M. Farley Nuclear Plant (Farley), Units 1 and 2, using Temporary Staff Guidance TSG-DORL-2021-001, Revision 3, Risk-Informed Process for Evaluation (ML23122A014). SNC proposes to revise TS 3.6.5 Required Action and Completion Time A.1, add Required Actions and Completion Times A.2, A.3, and A.4, as well as remove an expired note in TS 3.6.5 Limiting Conditions for Operation (LCO). Specifically, the change would modify the TS 3.6.5 Actions if containment average air temperature exceeds the LCO of 120 degrees Fahrenheit (°F) to allow continued operation for up to 30 cumulative days per calendar year provided that the containment average air temperature does not exceed 122°F (verified wi thin eight hours of containment average air temperature exceeding 120°F and once per eight hours thereafter) and that refueling water storage tank temperature remains less than or equal to 100°F (verified within eight hours of containment average air temperature exceeding 120°F and once per eight hours thereafter).
Consistent with Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), an application for an amendment to a license (inclu ding the technical specifications) must fully describe the changes requested, and following as far as applicable, the form prescribed for original applications. Section 50.34 of 10 CFR addresses the content of technical information required. This section stipulates that the submittal address the design and operating characteristics, unusual or novel design features, and principal safety considerations.
The U. S. Nuclear Regulatory Commission (NRC) staff has reviewed your application and concluded that supplemental information is needed for the NRC to accept the submittal.
J. Coleman
This information is needed to assess acceptance into the Risk-Informed Process for Evaluation or whether it is sufficiently complete to proceed as a technical review using LIC-206, Revision 1, Integrated Risk-Informed Decision-Making for Licensing Reviews, dated June 26, 2020 (ML19263A645), while leveraging risk insights in the staffs review.
The staff notes that on August 24, 2023, per SNCs LAR, the staff approved a similar temporary change to TS 3.6.5 (ML23235A296). Although the staffs approval for this temporary change was made under an emergent circumstance where the staff concluded that: (1) the licensee used its best efforts to make a timely application; (2) the licensee could not reasonably have avoided the situation; and (3) the licensee had not misused the provisions of 10 CFR 50.91(a)(5). In addition, due to the emergent circumstance, the staff conducted an independent evaluation and concluded that there was reasonable assurance that the proposed changes were bounded by the plant analysis of record. This was acceptable for the limited period of the one-time request where weather patterns were known for the time requested; and no other plant changes were being pursued that would impact the plants safety margins. However, for a permanent change, the licensee must provide an appropriate analysis to demonstrate that adequate safety margins are maintained and identify that appropriate measures are in place to capture any future changes that could impact existing margins in the licensing basis.
In order to make the application complete, the NRC staff requests that SNC supplement the application to address the information requested in the enclosure by June 14, 2024. This will enable the NRC staff to make a determination and begin its review. If the information responsive to the NRC staffs request is not received by the above date, the application will not be accepted for review pursuant to 10 CFR 2.101, Filing of application, and the NRC will cease its activities associated with the application. If the application is subsequently accepted for review, you will be advised of any further information needed to support the NRC staffs detailed technical review by separate correspondence.
The information requested and associated time frame in this letter were discussed with you and Ryan Joyce of your staff on May 28, 2024.
If you have any questions, please contact me at 301-415-2258 or via email at Zachary.Turner@nrc.gov.
Sincerely,
/RA/
Zachary M. Turner, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Docket No. 50-348 and 50-364
Enclosure:
Supplemental Information Needed
cc: Listserv SUPPLEMENTAL INFORMATION NEEDED
LICENSE AMENDMENT REQUEST TO CHANGE TECHNICAL SPECIFICATION 3.6.5,
CONTAINMENT AIR TEMPERATURE, ACTIONS USING THE RISK-INFORMED PROCESS
FOR EVALUATIONS
SOUTHERN NUCLEAR OPERATING COMPANY
The proposed license amendment request (LAR) change would modify the Technical Specification (TS) 3.6.5 Actions if containment average air temperature exceeds the Limiting Conditions for Operation (LCO) of 120 degrees Fahrenheit (°F) to allow continued operation for up to 30 cumulative days per calendar year provided that the containment average air temperature does not exceed 122°F (verified with in eight hours of containment average air temperature exceeding 120°F and once per eight hours thereafter) and that refueling water storage tank (RWST) temperature remains less than or equal to 100°F (verified within eight hours of containment average air temperature exceeding 120°F and once per eight hours thereafter).
The U. S. Nuclear Regulatory Commission (NRC) staff has reviewed your application and has concluded that supplemental information is necessary to enable the staff to make an independent assessment regarding the acceptability of the proposed LAR in terms of using the Risk-Informed Process for Evaluation (RIPE) or whether it is sufficiently complete to proceed separately via the license amendment review process, while leveraging risk insights in the staffs review.
Regulatory Requirements and Guidance
Title 10 of the Code of Federal Regulation (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, includes the following GDCs applicable to the licensees LAR:
Criterion 4 - GDC 4, Environmental and dynamic effects design bases, states, in part, that structures, systems, and components im portant to safety shall be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.
Criterion 16 - GDC 16, Containment design requires, in part, that reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as postulated accident conditions require.
Criterion 38 - GDC 38, Containment heat removal requires, in part, that a system to remove heat from the reactor containment shall be provided. The system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.
Criterion 50 - GDC 50, Containment design basis, requires, in part, that the reactor containment structure, including access openings, penetrations, and the containment heat removal system shall be designed so that the containment structure and its internal
Enclosure
compartment can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from a design basis loss-of-coolant accident (LOCA).
The regulation 10 CFR 50.36, Technical specifications, provides the regulatory requirements related to the contents of technical specifications. The regulations in 10 CFR 50.36(b) require that:
Each license authorizing operation of a production or utilization facility [] will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 [Contents of applications; technical information]. The Commission may include such additional technical specifications as the Commission finds appropriate.
The regulations in 10 CFR 50.36(c)(2) establish the requirements for LCOs.
Criterion 2 from 10 CFR 50.36(c)(2)(ii)(B) requires the establishment of LCOs for any process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
NUREG 1431, Revision 5.0, Standard Technical S pecifications, Westinghouse Plants, Volume 1, Specifications, and Volume 2, Bases (ML21259A155 and ML21259A159, respectively),
contain the improved Standard Technical Specifications (STS) for Westinghouse plants. The improved STS were developed based on the criteria in the Final Policy Statement of Technical Specifications Improvements for Nuclear Power Reactors, dated July 22, 1993 (58 FR 39132),
which was subsequently codified by changes to 10 CFR 50.36 (60 FR 36953).
The NRC staffs guidance for the review of engineered safety features is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:
LWR [light-water reactor] Edition (SRP), Chapter 6, Engineered Safety Features Rev. 2 dated March 2007 (ML063190010).
The NRC staffs guidance for the review of TSs is in NUREG -0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425).
NUMARC 87-00, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors Rev. 1 dated November 1987 (ML12137A732), notes that temperatures up to 120°F would likely not adversely affect operability of mechanical and electrical equipment and instrumentation for a short period of time (i.e., 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />). Based on this, 120°F has generally been considered the threshold for electrical equipment long-term functionality.
The Regulatory Guide (RG) 1.155, Station Blackout (ML003740034), describes a means acceptable to the NRC staff for meeting the requirements of 10 CFR 50.63, Loss of all alternating current power. NUMARC 87-00, Guidelines and Technical Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors, also provides guidance acceptable to the staff for meeting these requirements. NUMARC 87-00, Section 2.7.1, identifies the assumptions associated with loss of ventilation and indicates Equipment
Operability Inside Containment Temperatures re sulting from the loss of ventilation are enveloped by the LOCA and high energy line break environmental profiles.
NRC Temporary Staff Guidance TSG-DORL-2021-01, Revision 4, Risk-Informed Process for Evaluations (ML23354A150), provides the framework for streamlined processing of LARs and exemptions from NRC requirements that are submitted under RIPE.
Request for Supplemental Information
In accordance with TSG-DORL-2021-01, Revision 4, the staff performed a high-level no technical objection review to determine whether the request is appropriate under the RIPE process and provides deterministic insights that could be missed by the risk review. Specifically, the staff reviewed the licensees application against the four proposed criteria considered in developing any technical objections: (1) whether the licensees assumptions in the submitted analysis are reasonable, (2) whether the licensee has used an appropriate methodology, (3) whether the licensee fully considered the technica l aspects of the issue under consideration to support the IDPs determination, and (4) whether the screening questions were answered adequately by the licensees IDP. As a result of this review, the staff developed the following requests for supplemental information in order to further consider the application under the RIPE process.
Request for Supplemental Information 1 - Adherence to 10 CFR 50.36
The information provided should demonstrate that the requested TS changes are derived from the analysis contained in the SAR, as updated, as required by 10 CFR 50.36(b).
The current LCO temperature limit of 120 °F repr esents the lowest functional capability required for safe operation of the system as required by 10 CFR 36(c)(2)(i) under all conditions in which the TS is applicable.
- a. Explain why the proposed Required Actions which provide a higher containment temperature (i.e., 122 °F) and lower RWST temperature are not an alternate set of conditions that represent the lowest functional capability of the system, as required by the regulations, that should be included in the LCO instead of in the Required Actions.
- b. Provide a basis for allowing the containment to be in an elevated temperature (i.e., 122
°F) in conditions that are outside its design basis conditions for 30 days instead of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> as required by the existing Completion Time and the Westinghouse STS.
Request for Supplemental Information 2 - Impacts on Net Positive Suction Head (NPSH)
The current submittal does not address the effect of the increase in containment air temperature on the available NPSH for the pumps that dr aw water from the sump during the LOCA recirculation phase. The NRC staff requests that information be provided detailing the adverse impact of the present request on the LOCA sump temperature response, available NPSH during LOCA recirculation phase, and if any (or additional) containment accident pressure is needed during the transient to maintain positive NPSH margin.
Further, the submittal references Table 6.2-3 of the Final Safety Analysis Report that sets the inside temperature of containment at 127 °F for the containment pressure analysis, which is a
biased input for the peak pressure and temperature response; however, this table also shows the accumulator tank water temperature is 120 °F. To maximize the sump temperature response during LOCA recirculation phase for NPSH analysis, the information requested above should include biased inputs for a conservative analysis.
Request for Supplemental Information 3 - Impacts on Peak Cladding Temperature (PCT)
The NRC staff requests an analysis using an appropriate methodology for addressing the impact on PCT of this proposed change. The PCT impact is qualitatively assessed as a direct change relative to the increase in containment air temperature in the submittal. The NRC staff experience has shown this conclusion is not necessarily true (meaning a 2°F increase in containment air temperature does not necessarily limit PCT increase to only 2°F). A quantitative discussion of PCT margin relative to the 10 CFR 50.46(b)(1) PCT safety limit of 2200°F that compares the effects of the proposed changes to the current analysis of 120°F is requested to determine the magnitude of the PCT change and to adequately assess the effects on dose consequences of the respective accident sequences.
Confirmatory Information 1: Consideration of Non-Environmentally Qualified Equipment
The NRC staffs question spans from the station blackout (SBO) rule which indicates that the agency has accepted the position that most electrical equipment can function indefinitely at temperatures of 120 °F and below.
By letter dated August 22, 2023 (ML23234A151), SNC requested an emergency LAR for Farley, Units 1 and 2. By email dated August 22, 2023 (ML23236A002),
the NRC submitted a request for additional information (RAI). By letter dated August 23, 2023 (ML23235A288), SNC responded to the RAI. Although the NRC staffs RAI explicitly stated non-Environmental Qualification (non-EQ) equipment, the licensees response seem ed to be limited to equipment important to safety within the EQ program (i.e., 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants ) and did not appear to address non-EQ electrical equipment.
As SNCs proposed LAR dated April 19, 2024, is similar to the emergency LAR but requests a permanent versus a one-time temporary change to the average containment air temperature, confirm there is no non-EQ electric equipment (i.e., electric equipment not subject to the requirements in 10 CFR 50.49) within containment that is expected to perform a design function under normal operation (e.g., electrical equipment that is either relied upon by the plant operators to inform operational decisions or prov ides a signal input to other plant systems or processes) whose failure could mislead a plant operator or cause a plant transient.
Confirmatory Information 2: Reactor Coolant Pump (RCP) Seal LOCA Mitigation Strategy
The NRC approved amendments to adopt Initiative 4b (pre-TSTF-505, Revision 2)
(ML19175A243) for Risk-Informed Completion Times and 10 CFR 50.69 (ML21137A247) for risk-informed categorization and treatment of structures, systems, and components based on Regulatory Guide 1.200 Revision 2.
In its submittal dated April 19, 2024, the licensee stated, in part, that, It is noted that a newer version of the peer review process has been developed as NEI 17-07 and has been endorsed for use by RG 1.200. SNC elected to use the process as documented in Appendix X. The
differences do not impact the validity of the review, and that, Updates to FLEX modeling including adoption of PWROG-18042 FLEX equipment data and addition of credit for FLEX pumps to mitigate certain RCP seal leakage sequences if the low leakage RCP seal fails. The updates were performed consistent with the 2022 NRC staff memorandum for crediting FLEX strategies in PRA.
Please confirm, per your submittal, that; (1) Farley has installed Generation III RCP seals and adhering to PWROG-14001-P, Revision 1, "PRA Model for the Generation Ill Westinghouse Shutdown Seal," including exceptions for Limitations and Conditions, as approved in amendments 217 and 214 (ML17261A087), respectively, for the containment leakage rate testing program, and (2) SNC is not proposing to use, in total or in part, newly developed PRA methods that would be associated with NRC approval of amendments to adopt TSTF-591, Revise Risk-Informed Completion Time (RICT) Program," Regulatory Guide 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed 25 Activities, requirements in PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review," and NEI 17-07, Revision 2, "P erformance of PRA Peer Reviews Using the ASME/ANS PRA Standard."
ML24143A041 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-2/LA NRR/DEX/ELTB/BC NAME JLamb KGoldstein JPaige DATE 05/22/2024 05/22/2024 05/22/2024 OFFICE NRR/DSS/STSB/BC NRR/DSS/SNSB/BC NRR/DSS/SCPB/BC NAME SMehta PSahd MValentin DATE 05/22/2024 05/22/2024 05/22/2024 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME MMarkley ZTurner DATE 05/28/2024 05/28/2024