ML24281A097
| ML24281A097 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/09/2024 |
| From: | Dawnmathews Kalathiveettil Plant Licensing Branch II |
| To: | Coleman J Southern Nuclear Operating Co |
| Kalathiveettil, D | |
| References | |
| EPID L-2024-LLA-0054 | |
| Download: ML24281A097 (1) | |
Text
December 9, 2024 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.
3535 Colonnade Parkway Birmingham, AL 35243
SUBJECT:
EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENTS NOS. 323 AND 268, REGARDING LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS TO INCREASE SAFETY/RELIEF VALVE SETPOINTS (EPID L-2024-LLA-0054)
Dear Jamie Coleman:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 323 to Renewed Facility Operating License (RFOL) No. DPR-57 and Amendment No. 268 to RFOL No. NPF-5 for the Edwin I. Hatch Nuclear Plant (Hatch), Unit Nos. 1 and 2, respectively, in response to your application dated April 19, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24110A098 [public]
and ML24110A097 [non-public]), as supplemented by letter dated August 23, 2024 (ML24236A746).
The amendments revise Hatch, Units 1 and 2, Technical Specifications (TS) surveillance requirement (SR) 3.4.3.1 to increase the reactor coolant system nuclear pressure relief system safety/relief valve (S/RV) nominal mechanical relief setpoints. The proposed changes are intended to reduce the potential for S/RV pilot leakage. As a result of the increased S/RV setpoints, the amendments also change SR 3.1.7.7 to increase the minimum standby liquid control (SLC) pump discharge pressure.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commissions monthly Federal Register notice. If you have any questions, please contact me at dawnmathews.kalathiveettil@nrc.gov or 301-415-5905.
Sincerely,
/RA/
Dawnmathews T. Kalathiveettil, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-321 and 50-366
Enclosures:
- 1. Amendment No. 323 to DPR-57
- 2. Amendment No. 268 to NPF-5
- 3. Safety Evaluation cc: Listserv
SOUTHERN NUCLEAR OPERATING COMPANY, INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 323 Renewed License No. DPR-57
- 1.
The Nuclear Regulatory Commission (NRC, the Commission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 1 (the facility) Renewed Facility Operating License No. DPR-57 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated April 19, 2024, as supplemented by letter dated August 23, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I;
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is hereby amended by the new license condition and page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-57 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 323, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented prior to startup from the Unit 1 refueling outage in 2026.
FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. DPR-57 and Technical Specifications Date of Issuance: December 9, 2024 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2024.12.09 07:18:11 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 323 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Pages Insert Pages License License Page 4 Page 4 TSs TSs 3.1-20 3.1-20 3.4-6 3.4-6 Renewed License No. DPR-57 Amendment No. 323 for sample analysis or instrument calibration, or associated with radioactive apparatus or components (6)
Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed license shall be deemed to contain, and is subject to, the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below:
(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady-state reactor core power levels not in excess of 2,804 megawatts thermal.
(2)
Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 323, are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated before the time and condition specified:
SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled performance.
(3)
Fire Protection Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, August 9, October 7, and December 13, 2019, and February 5, and March 13, 2020, and as approved in the NRC safety evaluation (SE) dated June 11, 2020. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
SLC System 3.1.7 HATCH UNIT 1 3.1-20 Amendment No. 323 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate 41.2 gpm at a discharge pressure 1251 psig.
In accordance with the INSERVICE TESTING PROGRAM SR 3.1.7.8 Verify flow through one SLC subsystem from pump into reactor pressure vessel.
In accordance with the Surveillance Frequency Control Program SR 3.1.7.9 Verify all heat traced piping between storage tank and pump suction is unblocked.
In accordance with the Surveillance Frequency Control Program AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after pump suction piping temperature is restored within the Region A limits of Figure 3.1.7-2 SR 3.1.7.10 Verify sodium pentaborate enrichment is 60.0 atom percent B-10.
Prior to addition to SLC tank
S/RVs 3.4.3 HATCH UNIT 1 3.4-6 Amendment No. 323 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the S/RVs are as follows:
Number of Setpoint S/RVs (psig) 11 1160 34.8 Following testing, lift settings shall be within 1%.
In accordance with the INSERVICE TESTING PROGRAM
SOUTHERN NUCLEAR OPERATING COMPANY, INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 268 Renewed License No. NPF-5
- 1.
The Nuclear Regulatory Commission (NRC, the Commission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit No. 2 (the facility) Renewed Facility Operating License No. NPF-5 filed by Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated April 19, 2024, as supplemented by letter dated August 23, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations set forth in 10 CFR Chapter I;
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is hereby amended by the new license condition and page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-5 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No. 268 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented prior to startup from the Unit 2 refueling outage 2025.
FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to Renewed Facility Operating License No. NPF-5 and Technical Specifications Date of Issuance: December 9, 2024 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2024.12.09 07:18:59 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 268 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the License and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Pages Insert Pages License License Page 4 Page 4 TSs TSs 3.1-19 3.1-19 3.4-6 3.4-6
Renewed License No. NPF-5 Amendment No. 268 (6)
Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This renewed license shall be deemed to contain, and is subject to, the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions2 specified or incorporated below:
(1)
Maximum Power Level Southern Nuclear is authorized to operate the facility at steady sate reactor core power levels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.
(2)
Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B); as revised through Amendment No. 268 are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission.
(a)
Fire Protection Southern Nuclear Operating Company shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c),
as specified in the licensee amendment request dated April 4, 2018, supplemented by letters dated May 28, August 9, October 7, and December 13, 2019, and February 5, and March 13, 2020, and as approved in the NRC safety evaluation (SE) dated June 11, 2020.
Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would 2
The original licensee authorized to possess, use, and operate the facility with Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.
SLC System 3.1.7 HATCH UNIT 2 3.1-19 Amendment No. 268 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each SLC subsystem manual and power operated valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.
In accordance with the Surveillance Frequency Control Program SR 3.1.7.7 Verify each pump develops a flow rate 41.2 gpm at a discharge pressure 1251 psig.
In accordance with the INSERVICE TESTING PROGRAM SR 3.1.7.8 Verify flow through one SLC subsystem from pump into reactor pressure vessel.
In accordance with the Surveillance Frequency Control Program SR 3.1.7.9 Verify all heat traced piping between storage tank and pump suction is unblocked.
In accordance with the Surveillance Frequency Control Program AND Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after pump suction piping temperature is restored within the Region A limits of Figure 3.1.7-2 SR 3.1.7.10 Verify sodium pentaborate enrichment is 60.0 atom percent B-10.
Prior to addition to SLC tank
S/RVs 3.4.3 HATCH UNIT 2 3.4-6 Amendment No. 268 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1 Verify the safety function lift setpoints of the S/RVs are as follows:
Number of Setpoint S/RVs (psig) 11 1160 34.8 Following testing, lift settings shall be within +/- 1%.
In accordance with the INSERVICE TESTING PROGRAM
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 323 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-57 AND AMENDMENT NO. 268 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-5 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-321 AND 50-366
1.0 INTRODUCTION
By application dated April 19, 2024, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24110A098 [public] and ML24110A097 [non-public]), as supplemented by letter dated August 23, 2024 (ML24236A746), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.90, Application for amendment of license, construction permit, or early site permit, to the U. S. Nuclear Regulatory Commission (NRC). The amendments revise Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2, Technical Specification (TS) surveillance requirement (SR) 3.4.3.1 to increase the reactor coolant system nuclear pressure relief system safety/relief valve (S/RV) nominal mechanical relief setpoints.
The proposed changes are intended to reduce the potential for S/RV pilot leakage. As a result of the increased S/RV setpoints, the amendments also change SR 3.1.7.7 to increase the minimum standby liquid control (SLC) pump discharge pressure.
The licensees August 23, 2024, response (ML24211A016) to the NRC staffs request for additional information (RAI), provided additional information that clarified the application, but did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on July 2, 2024 (89 FR 54864).
2.0 REGULATORY EVALUATION
2.1
System Description
There are 11 S/RVs at each Hatch unit, located on the main steam lines within the drywell between the reactor vessel and the first isolation valve. Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. The S/RVs have two modes of operation: the safety mode and the relief mode. In the safety mode (or spring mode of operation), the spring-loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve.
In the relief mode of operation, pneumatic pressure is used to open the valve, initiated by switches located in the control room or by pressure-sensing instrumentation. S/RVs that are used in the relief mode are used as part of the low-low set (LLS) and Automatic Depressurization System (ADS) functions. The proposed setpoint change is for the safety mode of operation and does not change the opening pressure for either the LLS or ADS functions.
The SLC system (SLCS) is a backup method of manually shutting down the reactor to a cold subcritical state and is independent of the control rod system. It consists of a storage tank and two parallel trains of pumps, valves and associated piping used to deliver borated water from the storage tank to the reactor vessel. The SLC system is manually initiated from the control room as required by the emergency operating procedures. Since Hatchs construction permits were issued prior to 1984, the requirement in 10 CFR 50.62(c)(4) stating that SLCS initiation must be automatic, is not required for the Hatch units.
2.2 Licensees Proposed Changes As stated in Enclosure 1 of the LAR, the licensee has experienced S/RV pilot leakage during normal plant operation. To decrease the amount of leakage, the licensee is proposing to increase the nominal mechanical relief setpoint. This will increase the difference between the vessel steam dome pressure and the S/RV opening pressure, or what is commonly referred to as the simmer margin.
The proposed TS changes are applicable to both Hatch units. The licensee proposed to revise S/RV setpoint in SR 3.4.3.1 from 1,150 +/- 34.5 pounds per square inch gage (psig) to 1,160 +/-
34.8 psig (note that the setpoint tolerance remains at +/- 3%). In addition, the licensee also proposed to revise the SLC pump discharge pressure in SR 3.1.7.7 from 1,232 psig to 1,251 psig.
2.3 Regulatory Requirements On July 11, 1967, the Atomic Energy Commission (AEC) published for public comment in the Federal Register (32 FR 10213), a revised and expanded set of 70 draft General Design Criteria (GDC). On February 20, 1971, the AEC published in the Federal Register (36 FR 3255) a final rule that added Appendix A (final GDC) to 10 CFR Part 50, which was amended on July 7, 1971 (36 FR 12733). Differences between the 1967 draft GDC and the final GDC included a consolidation from 70 to 64 criteria.
The construction permits for Hatch, Unit No. 1, and Hatch, Unit No. 2, were issued on September 30, 1969, and on December 27, 1972, respectively. Consequently, Hatch, Unit No.
2, is licensed in conformance with 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants. Hatch, Unit No. 1, is licensed in conformance with the 1967 version of 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plant Construction Permits (ML043310029). Hatch, Unit No. 1, Final Safety Analysis Report (FSAR), Appendix F, Conformance to Atomic Energy Commission (AEC) Criteria, describes the relevant licensing bases for Hatch, Unit No. 1. The facility operating license for Hatch, Unit No. 1, was issued in 1974, and the facility operating license for Hatch, Unit No. 2, was issued in 1978.
Section 3.0, Reactor, of the Hatch, Unit No. 1, FSAR describes the facilitys conformance with the GDC. Hatch, Unit No. 1, was designed and constructed in accordance with the GDC issued for comment in July 1967. The NRC staff concluded that there was reasonable assurance that the facility met the intent of these GDC.
Section 3.0, Design of Structures, Components, Equipment, and Systems, of the Hatch, Unit No. 2, FSAR describes the facilitys conformance with the GDC. Hatch, Unit No. 2, was designed and constructed in accordance with the amended GDC dated July 7, 1971. The NRC staff concluded that the facility design conformed to these GDC.
Paragraph 10 CFR 50.92(a) states, in part, that in determining whether an amendment to a license will be issued, the NRC is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The common standards for licenses in 10 CFR 50.40(a), and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that the NRC must find reasonable assurance that the activities authorized can be conducted without endangering the health and safety of the public, and that such activities will be conducted in compliance with the Commissions regulations.
Paragraph 10 CFR 50.36(b) states, in part, that, the technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to Section 50.34. The TSs are included as Appendix A to the Hatch renewed facility operating licenses. Changes to the TSs are made under the provisions of Section 50.90, Application for amendment of license, construction permit, or early site permit.
Paragraph 10 CFR 50.36(c)(2) requires that TSs include limiting conditions for operation (LCOs). Per 10 CFR 50.36(c)(2)(i), [LCOs] are the lowest functional capability or performance levels of equipment required for safe operation of the facility. The regulation also requires that, when an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.
Paragraph 10 CFR 50.36(c)(3) states that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. The S/RVs are part of the primary success path and are credited in the Final Safety Analysis Report (FSAR) accident and safety analyses to mitigate the effects of the design basis transients and accidents. Since the setpoints in the SRs for the S/RVs are proposed to be increased, the licensee is required to provide sufficient analyses to ensure the proposed TS changes will continue to meet 10 CFR 50.36(c)(3).
Paragraph 10 CFR 50.55a, Codes and standards, specifies standards approved for incorporation by reference. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code requires that each vessel designed to meet Section III be protected from overpressure. The ASME BPV Code provides for a peak allowable pressure of 110 percent of vessel design pressure. As shown in Table 5.2-1 of the Hatch FSAR, the reactor pressure vessel design pressure is 1,250 psig. This results in a peak allowable pressure of 1,375 psig. In addition, Hatch TS 2.1.2, Reactor Coolant System (RCS) Pressure SL [Safety Limit], states that reactor steam dome pressure shall be 1,325 psig.
10 CFR 50.62 Requirements for reduction of risk from anticipated transients without scram (ATWS) events for light-water-cooled nuclear power plants, paragraphs (c)(4) and (c)(6),
provide, in part, that, each boiling water reactor must have a standby liquid control system (SLCS) and provide information sufficient to demonstrate the adequacy of items in such systems to the Commission. Any changes to the SLC system SRs require the licensee to provide acceptable analyses to support the adequacy of proposed TS changes.
10 CFR Part 50, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, includes the following GDCs applicable to the licensees LAR:
GDC 1, (Criterion 1), Quality standards and records, requires, in part, that structures, systems and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.
GDC 14 (Criterion 14), Reactor coolant pressure boundary, requires that the reactor coolant pressure boundary be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, or rapidly propagating failure, and of gross rupture.
GDC 15 (Criterion 15), Reactor coolant system design, requires that the reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
GDC 30 (Criterion 30), Quality of reactor coolant pressure boundary, requires, in part, that components which are part of the reactor coolant pressure boundary be designed, fabricated, erected, and tested to the highest quality standards practical.
3.0 TECHNICAL EVALUATION
In determining whether an amendment to a license will be issued, the Commission is guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. The staff evaluated the licensees application to determine if the proposed changes are consistent with the regulations, guidance, and licensing and design basis information discussed in Section 2.0 of this safety evaluation. The NRC staff reviewed the licensees statements in the LAR, the relevant sections of the licensees FSAR, and TS Bases to determine if the proposed changes are acceptable.
The main objective of the nuclear pressure relief system is to prevent overpressurization of the nuclear system. To meet this objective, it must be demonstrated that the peak pressure remains below acceptable limits. The proposed change to increase the S/RV setpoint from 1,150 psig to 1,160 psig will increase the peak pressure during overpressure events. In addition to vessel overpressure, there are several other items that are potentially affected by this change, including: anticipated operational occurrences (AOOs), design basis accidents, high pressure makeup systems, containment response and dynamic loading, and ATWS response.
Overpressure events are discussed below in Section 3.1, high pressure system performance is discussed in Section 3.2, containment evaluation in Section 3.3, ATWS in Section 3.4, standby liquid control system in Section 3.5 and emergency core cooling system (ECCS)/loss of coolant accident (LOCA) in Section 3.6.
3.1 Overpressure Considerations With the proposed increased setpoint from 1,150 psig to 1,160 psig, and the existing allowable
+/-3% tolerance, the S/RVs could open at a pressure as high as 1,194.8 psig. The existing Hatch cycle-specific analyses to confirm vessel overpressure margin is maintained is performed assuming an S/RV opening of 1,195 psig. Since the value of 1,195 psig is larger than the revised nominal setpoint plus maximum allowable drift tolerance of 1,194.8 psig (1160 psig + 34.8 psig), the NRC staff finds that the existing analysis is bounding and the proposed increase in setpoint does not result in changes to the overpressure analysis.
In Enclosure 1 to the LAR, the licensee makes cites a previous analysis and TS change where the S/RV setpoints were increased to their current value of 1,150 psig, and states that the evaluations documented in technical report NEDC-32041P, Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2, Updated Safety/Relief Valve Performance Requirements, Revision 2, provided justification for the S/RVs to open as high as 1,195 psig with one S/RV inoperable and at least 50 psi margin to the ASME BPV Code upset limit of 1,375 psig. As part of this review, the NRC staff examined NEDC-32041P, and notes that Section 2.2 states that the margin allows for variations in the peak vessel pressure which were calculated for past cycles and may be predicted for future fuel cycles. This statement infers that some of the margin can be used, and peak pressure could potentially be larger than 1,325 psig. However, the Hatch TS Safety Limit in TS 2.1.2 states that reactor steam dome pressure shall be 1,325 psig. In its August 23, 2024, supplement, the licensee clarified that, as part of its reload licensing analysis process, the overpressurization analysis is performed using the approved methodology with an S/RV opening setpoint of 1,195 psig and one S/RV out-of-service to confirm that both the peak pressure in the steam dome will be 1,325 psig (TS SL) and the overall peak pressure (i.e., at the bottom of the vessel) will be 1,375 psig (ASME limit). Therefore, the NRC staff finds that the existing overpressure analysis remains applicable and bounding given the proposed increase in S/RV setpoint and demonstrates the overpressure criteria will not be exceeded and is, therefore, acceptable.
3.2 High pressure system performance If the S/RVs actuate at a higher opening pressure, as proposed in the LAR, the vessel may experience a higher pressure during transients, AOOs or accidents. Consequently, the pressures in the piping, pumps, and turbines which supply the high-pressure makeup systems may also increase. For Hatch, the high-pressure systems include high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC). These systems are utilized for core cooling during postulated design basis transients and accidents as well as inventory control during events in which the normal feedwater is unavailable.
to the LAR states that the high-pressure system performance is discussed in Section 4.0 of the GE Hitachi Nuclear Energy (GEH) report NEDC-34126P, Edwin I. Hatch Nuclear Power Plant Units 1 and 2, Safety/Relief Valve Setpoint Increase, Revision 0, March 2024. This report, included as Attachment 4 (non-public) and Attachment 5 to the LAR, evaluated the HPCI and RCIC performance with a conservative S/RV opening pressure of 1,195 psig. This analysis is conservative as the opening pressure assumed is larger than the proposed allowable opening pressure in TS SR 3.4.3.1 of 1,194.8 psig, which includes the +3%
tolerance allowed for setpoint drift. The GEH analysis examined the pump performance curves and determined that the HPCI and RCIC systems can provide the rated flow as required given the increase in system pressure. The analysis also considered the fact that the pumps would operate at a slightly higher speed and found that the increased speed is within the turbine speed limits of each system and maintains margin to the mechanical overspeed trip.
In addition, as part of their design process, the licensee evaluated the impact on MOVs due to the potential for increased reactor vessel and system pressure as a result of the increase in the S/RV nominal opening setpoint in accordance with NRC Generic Letter 89-10 (ML031150300).
In its August 23, 2024, supplement, the licensee clarified that it follows the industry standard design process which requires a Design Attribute Review (DAR) be completed for all modifications, and that the design process, along with site specific procedures, have requirements that ensure all necessary changes to the affected systems and components are implemented prior to or concurrent with the overall change being made, which in this case, is the S/RV setpoint increase. The licensee stated that, The on-going design change will identify the necessary calculations, documentation, and/or valve setup changes necessary to implement the new S/RV setpoint while also ensuring the impacted MOVs will maintain sufficient margin to meet Generic Letter 96-05 [ML031110010] as well as conforming to testing requirements as specified in the generic letter. In addition, the licensee stated that, Although formal evaluations have not yet been performed, the design change process will ensure the continued operability of all affected systems and components, including impacted MOVs, at the increased S/RV setpoints. Based upon its independent review, the NRC staff finds that the licensee has a process to evaluate any impact of the new S/RV setpoint allowed in this LAR on the performance of MOVs at Hatch, Units 1 and 2.
The NRC staff reviewed the analysis provided in Section 3.0 of Enclosure 1 to the LAR and Section 4.0 of GEH NEDC-34126P for high pressure system performance. The licensee used a conservative opening pressure for the S/RVs that was larger than the allowable opening pressure in TS SR 3.4.3.1 and determined that operation of HPCI and RCIC is within the design limits for the system piping, pumps, and turbines. The licensees analysis considered both startup and steady state operation of the HPCI and RCIC systems for both loss of feedwater events as well as LOCAs. Based on the above, the NRC staff finds that there is reasonable assurance that the high-pressure systems will continue to perform their safety functions with the proposed increase to the S/RV setpoint and is, therefore, acceptable.
3.3 Containment Evaluation to the LAR states that the containment evaluation is discussed in Section 5.0 of GEH report NEDC-34126P. This report considered five containment related evaluations including: ATWS, design basis accident (DBA) LOCA, small steam line break (SSLB) for equipment qualification (EQ), Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, to 10 CFR Part 50, and station blackout (SBO). The licensee performed evaluations with the same methodologies as the current licensing bases for the individual events. The NRC staff discusses each specific event in the following for its effects on the containment response. In addition, the licensee evaluated the S/RV discharge line loads which is summarized below.
Anticipated Transients Without Scram Consistent with previous licensing basis analysis, the ODYN, STEMP and SHEX methods are used for ATWS analysis. The evaluation performed demonstrated that the peak wetwell pressure and temperature with the proposed S/RV setpoint change were equal to or bounded by the current analysis of record. Based on its independent review, the NRC staff finds that the existing ATWS containment analysis remains applicable given the proposed increase in S/RV setpoint.
Design Basis Accident (DBA) Loss-of-Coolant Accident The licensees analysis considered both short-and long-term LOCA loads and found that DBA LOCA analyses are unaffected by the proposed S/RV setpoint increase from 1,150 to 1,160 psig. For larger break sizes, the S/RVs would not open as the system pressure rapidly decreases upon the break. For the smaller break sizes, the S/RVs may open after the main steam isolation valves (MSIVs) close; however, the existing analysis assumed an S/RV opening pressure of 1,195 psig, which is larger, and thus more conservative, than the proposed allowable value of 1,194.8 psig (includes +3% allowable setpoint drift). Based on its independent review, the NRC staff finds that the DBA LOCA containment loading is not affected by the proposed increase in S/RV setpoint and is, therefore, acceptable.
Small Steam Line Break (SSLB) for Equipment Qualification The SSLB containment analyses were based on using limiting values from both Hatch, Units 1 and 2, so the results are applicable to both units. The licensees analysis resulted in negligible changes (+/-1°F) in the drywell temperature response for the various break sizes where the results remain below the EQ acceptance criteria. Based on its independent review, the NRC staff finds that the EQ for small steam line breaks is acceptable given the proposed increase in S/RV setpoint and is, therefore, acceptable.
Appendix R to 10 CFR Part 50 - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979 The licensee states that Hatch, Units 1 and 2, current licensing basis is a risk-informed, performance-based program based on National Fire Protection Association (NFPA) standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, for fire protection. However, the deterministic Appendix R containment response evaluation was assessed by the licensee for impact. The licensees analysis for Appendix R is a non-break event where the reactor pressure vessel is isolated by closure of the MSIVs. In this case, once the first S/RV would open at the proposed 1,160 psig setpoint, the remaining S/RV actuations will open based on the LLS logic which opens the S/RVs at pressures below the proposed S/RV setpoint. The increased S/RV setpoint results in only a minor delay in the first S/RV opening. The licensees analysis found that the effect on the suppression pool temperature response due to S/RV setpoint increase was negligible and, therefore, the effect on containment temperature and pressure is negligible. Based on its independent review, the NRC staff finds that the existing containment loading analysis for an Appendix R event remains applicable given the increase in S/RV setpoint and is, therefore, acceptable.
Station Blackout As with the Appendix R event described above, the station blackout is a non-break event where the reactor pressure vessel is isolated by closure of the MSIVs. In this case, once the first S/RV opens at the proposed 1,160 psig setpoint, the remaining S/RV actuations will open based on the LLS logic which then maintains vessel pressure for the duration of the SBO event. The increased S/RV setpoint results in only a minor delay in the first S/RV opening.Based on its independent review, the NRC staff finds that containment loading will not be challenged for an SBO event given the increase in S/RV setpoint and is, therefore, acceptable.
S/RV discharge line loads Section 3 of Enclosure 1 to the LAR states that an assessment of the impact of increasing the S/RV nominal setpoint to 1,160 psig on the S/RV discharge line loads for Hatch, Units 1 and 2, was performed. The licensee stated that the updated analyses for both units demonstrated that the current configuration of all 11 S/RV discharge line piping portions located within the vent pipes and torus meet ASME BPV Code requirements for all loading combinations. Based on its independent review, the NRC staff finds that the S/RV discharge line loads will not be challenged given the increase in S/RV setpoint and is, therefore, acceptable.
3.4 ATWS to the LAR states that the ATWS evaluations are discussed in Section 6.0 of GEH report NEDC-34126 P. This report examined the two limiting licensing basis ATWS events including MSIV closure and a pressure regulator failure open. The analysis was performed to demonstrate compliance with the following:
ASME BPV Code service level C pressure limit (1500 psig)
Containment pressure design limit (plant specific limit)
Suppression pool temperature design limit (plant specific limit) 10 CFR 50.46 peak cladding temperature (PCT) limit ( 2,200°F) 10 CFR 50.46 maximum cladding oxidation thickness nowhere exceed 0.17 times the total cladding thickness before oxidation The ATWS analysis performed used a combination of operating conditions from Hatch, Units 1 and 2, that are bounding for this specific ATWS analysis, so the results are applicable to both units. The analysis also considered mixed cores of GNF2 and GNF3, as well as full cores of GNF3. The results of the GEH analysis show that all the acceptance criteria are met. Given that the analyses were performed with the same methodologies as the current bases for these events and used the same assumptions, the NRC staff finds that there is reasonable assurance that the ATWS acceptance criteria would not be violated by the proposed increase to the S/RV setpoint.
3.5 Standby Liquid Control System TS SR 3.1.7.7 specifies the flow rate and discharge pressure necessary for each SLC pump to meet the requirements of 10 CFR 50.62(c)(4). To meet these requirements with the proposed increase in S/RV setpoint to 1,160 psig, the licensee is proposing to increase the required pump discharge pressure from 1,232 psig to 1,251 psig. As stated in Section 3 of Enclosure 1 to the LAR, the required SLC pump discharge pressure is based on the limiting peak pressure at the injection location (lower plenum injection) during an MSIV closure event at the beginning of cycle. The licensees analysis with the increased S/RV setpoint determined the resulting pressure at the SLC injection location is 1,203.3 psig at the SLC system injection time. When the SLC system losses of 47 psi are included, the required SLC pump discharge pressure is the proposed SR 3.1.7.7 value of 1,251 psig. As stated in Section 3 of Enclosure 1 of the LAR, the SLC pumps are positive displacement pumps which deliver a constant flow rate regardless of the discharge pressure and are adequate for the increase in operating pressure. In addition, as stated in Section 4.2.3.4.3 of the Hatch Unit 2 FSAR, the SLC system and pumps have sufficient pressure margin, up to the system relief valve setting of ~1,400 psig, to ensure solution injection into the reactor above the normal pressure in the bottom of the reactor. Based on its independent review, the NRC staff finds that there is reasonable assurance that the SLC system would continue to perform its safety functions with the proposed increase to the S/RV setpoint.
3.6 ECCS/LOCA to the LAR states that the ECCS/LOCA evaluation is discussed in Section 3.0 of GEH NEDC-34126P. This evaluation discusses the effect of the S/RV setpoint change on the peak cladding temperatures for LOCAs. Both Hatch, Unit 1 and Unit 2, are licensed to the TRACG-LOCA best estimate plus uncertainty ECCS/LOCA evaluation methodology. GEH performed an analysis using the licensed methodology by selecting representative limiting break locations to determine the effect of increasing the S/RV opening setpoint by running the break spectra for those break locations. GEH found that the licensing basis ECCS/LOCA results are not affected by increasing the S/RV opening setpoint nominal value from 1,150 psig to 1,160 psig. This is expected as for most breaks, the system pressure falls such that the S/RVs do not open. For small breaks, the system pressure may increase to the S/RV setpoint as the MSIVs close; however, the first S/RV opening would be slightly delayed with the proposed S/RV setpoint and would not be expected to change the overall system response. In addition, the ADS will provide system depressurization regardless of the S/RV opening setpoint. Based on its independent review, the NRC staff finds that there is reasonable assurance that the existing ECCS/LOCA analysis would not be affected by the proposed increase to the S/RV setpoint and is, therefore, acceptable.
3.7 Technical Conclusion The licensee proposed changes to TS SR 3.4.3.1 to increase the S/RV setpoint from 1,150 psig to 1,160 psig and to SR 3.1.7.7 to increase the minimum SLC pump discharge pressure from 1,232 psig to 1,251 psig. The NRC staff has reviewed the licensees submittal, as supplemented, and analysis provided in Enclosure 1 to the LAR as well as GEH report NEDC-34126P (included as an attachment to the LAR) and finds the proposed increase to the S/RV setpoint acceptable, and notes that:
The reactor coolant system pressure safety limit in TS 2.1.2 will be protected, The ASME BPV Code requirement that peak allowable pressure of 110 percent vessel design pressure would not be exceeded, The ECCS/LOCA analysis is not affected, The high-pressure systems (HPCI and RCIC) will continue to perform their safety functions, Containment loadings would remain within the design basis limits, The ATWS acceptance criteria would not be violated; and The SLC system would continue to perform its safety function.
Based on its independent evaluation, as described above in section 3.0 of this safety evaluation, the NRC staff concludes that the requirements of 10 CFR 50.36(c)(3) will continue to be met because the SRs will continue to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation would be met. Based on the above, the staff concludes that there is reasonable assurance that the applicable plant design criteria and the ASME BPV Code, as incorporated by reference in 10 CFR 50.55a, related to vessel overpressure would continue to be met.
Therefore, the NRC staff finds the proposed changes to be acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the State of Georgia official was notified of the proposed issuance of the amendment on October 21, 2024. The State official confirmed the State of Georgia had no comments on October 25, 2024.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration on July 2, 2024 (89 FR 54864), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Robert Beaton, NRR Thomas Scarbrough, NRR Date: December 9, 2024
ML24281A097 OFFICE NRR/DORL/LPL2-1/PM*
NRR/DORL/LPL2-1/LA NRR/DSS/SNSB/BC NAME DKalathiveettil KGoldstein DMurdock DATE 10/04/2024 10/08/2024 12/03/2024 OFFICE NRR/DSS/STSB/BC NRR/DEX/EMIB/BC NRR/DEX/EICB/BC NAME SMehta SBailey FSacko DATE 12/06/2024 12/04/2024 12/05/2024 OFFICE OGC - NLO NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/PM NAME MChwedczuk MMarkley DKalathiveettil DATE 12/03/2024 12/09/2024 12/09/2024