ML23032A332
| ML23032A332 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/24/2024 |
| From: | Dawnmathews Kalathiveettil NRC/NRR/DORL/LPL2-1 |
| To: | Gayheart C Southern Nuclear Operating Co |
| Shared Package | |
| ML24114A070 | List: |
| References | |
| EPID L-2022-LLA-0120 | |
| Download: ML23032A332 (1) | |
Text
April 24, 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. 322 AND 267, REGARDING LICENSE AMENDMENT REQUEST TO RELAX REQUIRED NUMBER OF FULLY TENSIONED REACTOR PRESSURE VESSEL HEAD CLOSURE BOLTS IN TECHNICAL SPECIFICATIONS TABLE 1.1-1, MODES (EPID L-2022-LLA-0120)
Dear Jamie Coleman:
The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 322 to Renewed Facility Operating License No. DPR-57 and Amendment No. 267 to Renewed Facility Operating License No. NPF-5 for the Edwin I. Hatch Nuclear Plant (Hatch), Unit Nos. 1 and 2, respectively. The amendments consist of changes to the Hatch Technical Specifications (TSs) in response to your application dated August 19, 2022, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22231B055), as supplemented by letters dated January 20 and October 20, 2023, and January 30, 2024 (ML23020A902, ML23293A235, and ML24030A895, respectively).
The proposed license amendment request would revise TSs for Hatch, Units 1 and 2, to relax the required number of fully tensioned reactor pressure vessel head closure bolts in TS Table 1.1-1, MODES.
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. 322 to DPR-57
- 2. Amendment No. 267 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. 322 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 August 19, 2022, as supplemented by letters dated January 20, and October 20, 2023, and January 30, 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. 322, 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 within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jamie Pelton, Deputy Director 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: April 24, 2024 JAMIE PELTON Digitally signed by JAMIE PELTON Date: 2024.04.24 09:57:41 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 322 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 Page 11 Page 11 TSs TSs 1.1-7 1.1-7 Renewed License No. DPR-57 Amendment No. 322 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. 322, 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.
Renewed License No. DPR-57 Amendment No. 322
- c.
The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, from the date of the most recent successful pressure measurement test.
(11) 10 CFR 50.69 Risk-Informed Categorization Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. 305, dated June 26, 2020.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).
Prior to implementation of the Renewed License Amendment No. 305, dated June 26, 2020, Southern Nuclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as-built, as-operated, and as-maintained plant and shall ensure the risk acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met (12) Reactor Vessel Head Closure Bolts Hatch Nuclear Plant Unit 1 is approved to operate in Modes 1 - 4 with at least 51 reactor vessel head closure bolts fully tensioned. In addition, a reactor vessel head closure bolt cannot be considered fully tensioned unless all applicable ASME Section XI acceptance criteria are met (irrespective of any existing NRC approved alternative to, or relief from, the acceptance criteria). Upon implementation of Amendment No. 322, Southern Nuclear Operating Company shall update the Reactor Vessel Reassembly procedure to include this requirement.
D.
Southern Nuclear shall not market or broker power or energy from Edwin I. Hatch Nuclear Plant, Unit 1.
- 3.
This renewed license is effective as of the date of issuance and shall expire at midnight, August 6, 2034.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION Samuel J. Collins, Director Office of Nuclear Reactor Regulation Attachments:
Appendix A - Technical Specifications Appendix B - Environmental Protection Plan Date of Issuance: January 15, 2002
Definitions 1.1 HATCH UNIT 1 1.1-7 Amendment No. 322 Table 1.1-1 (page 1 of 1)
MODES MODE TITLE REACTOR MODE SWITCH POSITION AVERAGE REACTOR COOLANT TEMPERATURE
(°F) 1 Power Operation Run NA 2
Startup Refuel(a) or Startup/Hot Standby NA 3
Hot Shutdown(a)
Shutdown
> 212 4
Cold Shutdown(a)
Shutdown 212 5
Refueling(b)
Shutdown or Refuel NA (a)
At least 51 reactor vessel head closure bolts fully tensioned, subject to the limitations of license condition 2.C.(12).
(b)
Fewer than 51 reactor vessel head closure bolts fully tensioned.
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. 267 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 August 19, 2022, as supplemented by letters dated January 20, and October 20, 2023, and January 30, 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. 267 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 within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Jamie Pelton, Deputy Director 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: April 24, 2024 JAMIE PELTON Digitally signed by JAMIE PELTON Date: 2024.04.24 09:57:09 -04'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 267 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 Page 11 Page 11 TSs TSs 1.1-8 1.1-8
Renewed License No. NPF-5 Amendment No. 267 (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. 267 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.
Renewed License No. NPF-5 Amendment No. 267 (h)
TSTF-448 Control Room Habitability Upon implementation of the Amendments adopting TSTF-448, Revision 3, the determination of control room envelope (CRE) unfiltered air inleakage as required by SR 3.7.4.4, in accordance with TS 5.5.14.c.(i), the assessment of CRE habitability as required by Specification 5.5.14.c.(ii), and the measurement of CRE pressure as required by Specification 5.5.14.d, shall be considered met. following implementation:
i)
The first performance of SR 3.7.4.4, in accordance with Specification 5.5.14.c.(i), shall be within the next 18 months.
ii)
The first performance of the periodic assessment of CRE habitability, Specification 5.5.14.c.(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, of the next successful tracer gas test.
iii)
The first performance of the periodic measurement of CRE pressure, Specification 5.5.14.d, shall be within 24 months, plus the 6 months allowed by SR 3.0.2, from the date of the most recent successful pressure measurement test.
(i) 10 CFR 50.69 Risk-Informed Categorization Southern Nuclear Operating Company is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) specified in the Renewed License Amendment No. 250 dated June 26, 2020.
Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).
Prior to implementation of the Renewed License Amendment No. 250 dated June 26, 2020, Southern Nuclear Operating Company shall update the Probabilistic Risk Assessment (PRA) models to reflect the as-built, as-operated, and as-maintained plant and shall ensure the risk acceptance guidelines found in Regulatory Guide (RG) 1.174, Revision 3 are met.
(j)
Reactor Vessel Head Closure Bolts Hatch Nuclear Plant Unit 2 is approved to operate in Modes 1 - 4 with at least 54 reactor vessel head closure bolts fully tensioned. In addition, a reactor vessel head closure bolt cannot be considered fully tensioned unless all applicable ASME Section XI acceptance criteria are met (irrespective of any existing NRC approved alternative to, or relief from, the acceptance criteria). Any bolt that is less than fully tensioned shall have at least nine adjacent bolts on either side that are fully tensioned. Upon implementation of Amendment No. 267, Southern Nuclear Operating Company shall update the Reactor Vessel Reassembly procedure to include this requirement.
Definitions 1.1 HATCH UNIT 2 1.1-8 Amendment No. 267 Table 1.1-1 (page 1 of 1)
MODES MODE TITLE REACTOR MODE SWITCH POSITION AVERAGE REACTOR COOLANT TEMPERATURE
(°F) 1 Power Operation Run NA 2
Startup Refuel(a) or Startup/Hot Standby NA 3
Hot Shutdown(a)
Shutdown
> 212 4
Cold Shutdown(a)
Shutdown 212 5
Refueling(b)
Shutdown or Refuel NA (a) At least 54 reactor vessel head closure bolts fully tensioned, subject to the limitations of license condition 2.C.(3)(j).
(b) Fewer than 54 reactor vessel head closure bolts fully tensioned.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 322 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-57 AND AMENDMENT NO. 267 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 August 19, 2022, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22231B055), as supplemented by letters dated January 20, and October 20, 2023, and January 30, 2024 (ML23020A902, ML23293A235, and ML24030A895, respectively), 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). Specifically, the licensee is proposing to relax the required number of fully tensioned reactor pressure vessel (RPV) head closure bolts in Technical Specification (TS) Table 1.1-1, MODES of each of the TSs for the Edwin I. Hatch Nuclear Plant (Hatch) Units 1 and 2. Hereafter, any reference to studs is synonymous with RPV head closure bolts.
The NRC staff participated in a regulatory audit from November 30, 2022, through January 31, 2023, to support its review of the LAR. An audit plan was issued on November 17, 2022 (ML22320A072), to ascertain the information needed to support its review of the application and to develop requests for additional information (RAIs), as needed. On February 21, 2023, the staff issued an audit summary to support the review of the LAR (ML23034A246).
As part of its review, the NRC staff issued RAIs by email correspondence dated December 21, 2022 (ML22355A208) and September 20, 2023 (ML23250A047). SNC provided responses to the RAIs by letters dated January 20 and October 20, 2023. The supplemental letter, dated January 30, 2024, provided additional information that clarified the application, expanded the overall scope of the application as originally noticed, and changed the staffs original proposed no significant hazards consideration determination as published in the Federal Register on November 1, 2022 (87 FR 65834). A subsequent notice of the staffs further no significant hazards consideration determination was published in the Federal Register on February 20, 2024 (89 FR 12876).
2.0 REGULATORY EVALUATION
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 Title 10 of the Code of Federal Regulations (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 of Hatch, Unit 1, and Hatch, Unit 2, were issued on September 30, 1969 (Unit 1), and on December 27, 1972 (Unit 2), respectively. Consequently, Hatch, Unit 2, is licensed in conformance with 10 CFR Part 50, Appendix A, General Design Criteria. Hatch, Unit 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 1, Final Safety Analysis Report (FSAR), Appendix F, Conformance To Atomic Energy Commission Criteria, (ML20063G153) describes the relevant licensing bases for Hatch, Unit 1.
The operating license for Hatch, Unit 1, was issued in 1974, and the operating license for Hatch, Unit 2, was issued in 1978.
Section 3.0, Design of Structures, Components, Equipment, and Systems, of the NRC safety evaluation report (ML20063G115) related to the operation of Hatch, Unit 1, describes the NRC staffs evaluation of the facilitys conformance with the GDC for the original facility operating license. Hatch, Unit 1, was designed and constructed in accordance with the GDC issued for comment in July of 1967. The NRC safety evaluation report for the Unit 1 operating license concluded that there was reasonable assurance that the plant met the intent of the GDC published in the FR in 1971.
Section 3.0, Design of Structures, Components, Equipment, and Systems, (ML20063G191) of the NRC safety evaluation report (NUREG-0411) related to the operation of Hatch, Unit 2 (ML19309G772), describes the NRC staffs evaluation of the facilitys conformance with the GDC for the original facility operating license. Hatch, Unit 2, was designed and constructed in accordance with the amended GDC dated July 7, 1971, and the NRC safety evaluation report for the Unit 2 operating license concluded that the plant design conformed to the amended criteria.
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.
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 § 50.34.
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. The licensees proposed changes pertain to the definitions of plant operating modes shown in TS Table 1.1-1, MODES. Each LCO is required to be met in the applicable operating modes specified in the TS.
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.
10 CFR 50.55a, Codes and standards, specifies standards approved for incorporation by reference. By letters dated August 10, 2017 (ML17205A345), and February 6, 2019 (ML19035A550), the NRC authorized relief from inspection requirements of the American Society of Mechanical Engineers (ASME) Code,Section XI, for Hatch Unit 2 RPV head closure bolt #33. This amendment does not address relief from future inspection of installed RPV head closure bolts that would be required under 10 CFR 50.55a and the ASME Code.
Section 10 CFR Part 50, Appendix A, General Design Criteria [GDC] for Nuclear Power Plants, applies and includes the following GDCs applicable to the licensees LAR:
GDC 1, (Criterion 1), Quality standards and records, requires 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 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
3.1 Summary of the Licensees Proposed Changes Hatch TS 1.0, Use and Application, provides Section 1.1, Definitions, and Table 1.1-1, MODES. Hatch TS Section 1.1 defines a MODE as:
A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
Current TS Table 1.1-1 for Units 1 and 2 states, The licensee proposes the following revisions to the current TSs:
Revised footnotes (a) and (b) to Unit 1 TS Table 1.1-1 would state:
(a)
At least 51 reactor vessel head closure bolts fully tensioned, subject to the limitations of license condition 2.C.(12).
(b)
Fewer than 51 reactor vessel head closure bolts fully tensioned.
Revised footnotes (a) and (b) to Unit 2 TS Table 1.1-1 would state:
(a)
At least 54 reactor vessel head closure bolts fully tensioned, subject to the limitations of license condition 2.C.(3)(j).
(b)
Fewer than 54 reactor vessel head closure bolts fully tensioned.
Proposed Unit 1 license condition, 2.C.(12), Reactor Vessel Head Closure Bolts,' would state:
Hatch Nuclear Plant Unit 1 is approved to operate in Modes 1 - 4 with at least 51 reactor vessel head closure bolts fully tensioned. In addition, a reactor vessel head closure bolt cannot be considered fully tensioned unless all applicable ASME Section XI acceptance criteria are met (irrespective of any existing NRC approved alternative to, or relief from, the acceptance criteria). Upon implementation of Amendment No. 322, Southern Nuclear Operating Company shall update the Reactor Vessel Reassembly procedure to include this requirement.
Proposed Unit 2 license condition, 2.C.(3)(j), Reactor Vessel Head Closure Bolts,' would state:
Hatch Nuclear Plant Unit 2 is approved to operate in Modes 1 - 4 with at least 54 reactor vessel head closure bolts fully tensioned. In addition, a reactor vessel head closure bolt cannot be considered fully tensioned unless all applicable ASME Section XI acceptance criteria are met (irrespective of any existing NRC approved alternative to, or relief from, the acceptance criteria). Any bolt that is less than fully tensioned shall have at least nine adjacent bolts on either side that are fully tensioned. Upon implementation of Amendment No. 267, Southern Nuclear Operating Company shall update the Reactor Vessel Reassembly procedure to include this requirement.
The licensee is proposing to relax the required number of fully tensioned RPV head closure bolts in TS Table 1.1-1, MODES, of each Hatch unit by revising footnotes (a) and (b) in the table, as described in Attachments 1 and 2 of its letter dated January 30, 2024. In that letter, the licensee stated, TS Table 1.1-1 Footnotes (a) and (b) specify the reactor vessel head closure bolt requirements. For the purpose of this RAI [request for additional information] response, reactor vessel head closure bolts as specified in the TSs, are equivalent to reactor pressure vessel head closure studs or studs. The licensees submittal dated August 19, 2022, included calculations performed by its contractor, Dominion Engineering, Inc. (DEI), that determined the impact of out-of-service studs on RPV closure stresses, stud stresses, closure flange separation, and fatigue usage. These calculations are documented in C-037-2201-00-01, Revision 0, Hatch Unit 1 Operation with One Stud Out of Service Evaluation, (Enclosure 2) and C-037-2201-00-02, Revision 0, Hatch Unit 2 Operation with Two Studs Out of Service Evaluation (Enclosure 3).
In Section 2.3 of Enclosure 1 of its letter dated August 19, 2022, the licensee discusses the inspection results on Unit 2 closure bolt #33, its efforts to remove the repair/replace the subject bolt, and industry operating experience. The licensee states, in part, that, As such, SNC is proposing changes to TS Table 1.1-1 to address the increased possibility that a reactor pressure vessel head closure stud may not be able to be fully tensioned and avert the possible need for an exigent or emergency license amendment during the Spring 2023 Refueling Outage. During that refueling outage, stud #33 will be detensioned for reactor pressure vessel head removal and retensioned for operation, thereby increasing the possibility of its failure.
The licensee requested NRC approval of these amendments prior to the Unit 2 spring 2023 Refueling Outage (2R27). In its response to RAI-02 by letter dated January 20, 2023 (ML23020A902), the licensee stated, in part, that, SNC cannot fully characterize the indication and is unable to predict the operational impact of stud #33 (e.g., its continued ability to be fully tensioned).
Furthermore, this is a condition for which there is not any directly applicable operating experience. Given these uncertainties, SNC is requesting approval of the LAR as a contingency in the event that the stud may not be able to be fully tensioned during the refueling outage and future attempts to remove and replace the stud may result in not being able to fully tension the stud. SNC fully intends to employ sound engineering principles to maintain all RPV studs in service. While there is no apparent degradation of any other RPV studs on HNP Unit 1 or Unit 2, circumstances may arise that result in the need to safely operate HNP Units 1 and 2 with a head closure stud(s) not fully tensioned. SNC considered it prudent and reasonable to request the change as a contingency given the engineering calculation included with the LAR supports plant operation with one Unit 1 stud and two Unit 2 studs less than fully tensioned.
SNC is unable to fully characterize the indication in stud #33 to predict the behavior of the indication. Therefore, there are no calculations or inspection results that indicate a potential for the indication to propagate to a point that the tensioning is not achievable. SNC is requesting the LAR as a contingency in the event the stud #33 were to not achieve sufficient tensioning.
The increased potential for a less than fully tensioned stud during an outage is due to an indication detected in RPV head closure bolt #33 of Hatch Unit 2 during the spring 2017 Refueling Outage (2R24) during a volumetric examination required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. A subsequent surface examination is also required, per ASME Code,Section XI, if an indication is detected during the volumetric examination. The licensee stated that due to the location of the indication at or slightly below the reactor flange, removal of RPV head closure bolt #33 was necessary to perform the surface examination. Hence, the licensee submitted and received approval for two relief requests from the required surface examination. The first relief request was for the duration of 2R24 until the beginning of the spring 2019 Refueling Outage (2R25);
and the second was for the duration of the fifth inservice inspection (ISI) interval scheduled to end on December 31, 2025. The NRC staff authorized these relief requests, as documented in NRC safety evaluations dated August 10, 2017, and February 6, 2019 (ML17205A345 and ML19035A550, respectively).
The licensees attempts to remove RPV head closure bolt #33 of Hatch Unit 2 during refueling outages 2R24 and 2R25 were unsuccessful, and, therefore, the closure bolt is expected to remain in place and fully tensioned for the duration of the fifth ISI interval ending on December 31, 2025. The licensee stated that the relief request does not provide relief from TS Table 1.1-1, MODES, which requires all reactor vessel head closure bolts to be fully tensioned, and that if the indication on closure bolt #33 of Hatch Unit 2 becomes worse, or an indication is found on another closure bolt, there is a possibility that a closure bolt(s) might be incapable of full tension. In response to RAI-04 of its supplement dated January 20, 2023 (ML23020A902), the licensee stated, in part, that, There are no repair plans for stud #33 for the Spring 2023 refueling outage for HNP [Hatch Nuclear Plant] Unit 2. A different non-destructive method was being considered since other methods of removal have been unsuccessful. The method of consideration has not been used in the nuclear industry, so additional vetting is required to ensure no adverse collateral impacts to other areas such as the vessel flange. There was not sufficient time to complete these vetting activities prior to the upcoming Unit 2 refueling outage.
In its letter dated August 19, 2022, the licensee also stated, in part, that, Prudent operating and engineering principles are applied to maintain all head closure studs in service. Circumstances may arise that result in the need to safely operate HNP Units 1 and 2 with a head closure stud(s) not fully tensioned.
If the reactor pressure vessel head is to be installed with an untensioned stud(s),
the tensioning pattern will be reviewed to ensure that the sealing o-rings are fully compressed before the stud(s) not being tensioned is encountered in the pattern.
The ASME Code examination requirements would also be reviewed to verify the requirements are being met or if relief is necessary.
3.2 NRC Staffs Evaluation By e-mail dated December 21, 2022 (ML22355A208), the NRC staff transmitted RAIs to the licensee concerning matters discussed during the regulatory audit. The RAIs requested the following information:
RAI-01: (a) a summary of all attempts to inspect and/or remove stud #33 since the 2017 authorization, including techniques applied for removal of each attempt, results achieved, and acceptance criteria for terminating removal efforts, and (b) post-tensioning values for the as-left condition of stud #33 for each outage; RAI-02: (a) the technical basis supporting the assertion that stud #33 or any other stud cannot be fully tensioned, and (b) any estimated calculations or inspection results that would indicate that the known flaw on stud #33 would propagate to a point such that sufficient tensioning is not achievable; RAI-03: The safety basis to have additional studs less than fully tensioned in Unit 1 and 2 when there is no apparent degradation or inspection results supporting such action; RAI-04 (a) confirm outage-specific repair plans for stud #33 that were expected to be completed during the spring 2023 refueling outage for Hatch, Unit 2, and (b) justification for SNCs request for NRC approval and LAR issuance by February 25, 2023, prior to completion of the Spring 2023 outage; RAI-05: (a) available industry operating experience for stuck RPV studs at other boiling water reactor (BWRs) and methods used to repair/replace such studs, and (b) confirm that this is the first request for a BWR to change its design and licensing basis to allow operation with one or more reactor vessel studs less than fully tensioned; RAI-06: clarify that the stud conditions that were analyzed in Table 3 of DEI Calculation C-037-2201-00-02 for Hatch Unit 2 were for two un-tensioned studs separated by nine tensioned studs for Cases B1 and B2 and two studs (separated by nine tensioned studs) that fail in service for Cases C1 and C2; RAI-07: (a) clarify whether CC [ASME Code Case] N-864 is used now or will be used in the future in the ISI program for the Hatch RPV threads-in-flange, and (b) confirm that the impact of the proposed change during preload conditions is bounded by the maximum applied stress intensity factor for the Preload case in Table 6-1 of the EPRI
[Electric Power Research Institute] report, if CC N-864 is used now or will be used in the future in the ISI program; During the review of the LAR, the NRC staff evaluated the calculations in DEI Calculations C-037-2201-00-01, Revision 0, and C-037-2201-00-02, Revision 0, as described below, and assessed the impact of out-of-service studs on remaining stud stresses, RPV closure stresses, RPV closure flange separation, and fatigue usage. The staff also evaluated the impact of the proposed changes on the RPV threads-in-flange.
As part of its evaluation, the NRC staff replicated independently the finite element models described in the DEI calculations to perform independent sensitivity analysis. Specifically, the staff changed element types used for the modeled RPV head closure studs and compared stress results in studs to assess the sufficiency of DEIs finite element models.
In its letter dated October 20, 2023, the licensee stated, in part that, The use of the term out of service was chosen in part for consistency with the evaluation performed by Dominion Engineering, Inc. (DEI) which provided the technical justification for the change. Also, the terms operable and operability are not defined in the regulations related to commercial nuclear power.
For a common reference, the NRC staffs discussion in Sections 3.2.1 and 3.2.2 refers to the DEI analytical methods and calculations in the same context of out-of-service RPV head closure bolts. A discussion of the TS and operability is provided in Section 3.2.3 of this SE.
3.2.1 Evaluation of Structural Integrity The licensee performed a structural integrity evaluation of the Hatch Unit 2 RPV closure flange in 2017 as a technical basis for the two NRC-authorized relief requests, noted in Section 3.1 of this SE, associated with the indication detected in stud #33. That evaluation was documented in DEI Calculation C-3944-00-01, Revision 0, Hatch Unit 2 Operation with One Stud Out of Service Evaluation, and included as Enclosure 2 to the SNC submittals dated February 17, 2017 (ML17048A090) and November 5, 2018 (ML18309A272). The 2017 structural integrity evaluation analyzed the impact of one un-tensioned stud and one stud that fails during service on the design stresses and fatigue usage of the Hatch Unit 2 RPV closure studs and closure flange. The NRC staff approved relief requests based on this evaluation, as documented in the safety evaluations dated August 10, 2017 (ML17205A345), and February 6, 2019 (ML19035A550), respectively.
The structural integrity evaluations of the RPV flange in DEI Calculation C-037-2201-00-01, Revision 0, for Hatch Unit 1 and C-037-2201-00-02, Revision 0, for Hatch Unit 2, follow the same structural integrity evaluation methodology as that performed in the 2017 structural integrity evaluation. The NRC staff compared the methodologies and assumptions in the former and latter evaluations and confirmed consistency between the evaluations. A summary of what the staff compared for each aspect of the structural integrity evaluation is summarized in the following sections.
The structural integrity evaluation of DEI Calculation C-037-2201-00-01, Revision 0, analyzed the impact of one untensioned stud and one stud that fails during service on the design stresses and fatigue usage of the Hatch Unit 1 RPV closure studs and closure flange. The structural integrity evaluation of DEI Calculation C-037-2201-00-02, Revision 0, analyzed the impact of two untensioned studs (separated by at least nine studs) and two studs (separated by at least nine studs) that fail during service on the design stresses and fatigue usage of the Hatch Unit 2 RPV closure studs and closure flange. The NRC staff noted that the impact of two untensioned studs in the current evaluation for Hatch Unit 2 addresses the possibility of finding an indication on another stud, and therefore, the evaluation addresses the potential of up to two studs being incapable of full tension. The staff evaluated the impact of out-of-service studs on the stresses and fatigue usage in the studs and RPV closure flange, as discussed in the following sections.
3.2.1.1 Primary Load Stresses in Remaining Studs As noted above, the NRC staff confirmed independently that the methodologies in DEI Calculations C-037-2201-00-01, Revision 0, and C-037-2201-00-02, Revision 0, were consistent with those in the 2017 structural integrity evaluation. This comparison included confirming consistency of the methodology in the determination of the impact of out-of-service studs on the primary load stresses in the remaining studs. Specifically, the staff confirmed that the governing equations used in the calculation of primary load stresses were consistent with those in the 2017 structural integrity evaluation. Section 3 of Enclosure 1 to its letter dated August 19, 2022 (based on Sections 5.1.3 of DEI Calculations C-037-2201-00-01 and C-037-2201-00-02), states that the maximum primary stress in the remaining studs is less than the allowable stress of 36.3 ksi for both Hatch units.
The NRC staff determined that the impact of one out-of-service stud for Hatch Unit 1 and two out-of-service studs for Hatch Unit 2 is acceptable in terms of primary load stress in the remaining studs because the maximum primary stress values are less than the corresponding allowable stress values.
3.2.1.2 ASME Code Stresses in RPV Closure Flange and Remaining Studs The NRC staffs comparison of methodologies, discussed in Section 3.2.1 of this SE, included confirming consistency of the methodology in the determination of the impact of out-of-service studs on the ASME Code stresses in the remaining RPV closure studs and RPV closure flange.
Specifically, the staff confirmed that the modeling method, boundary conditions, and analyses cases were consistent with those in the 2017 structural integrity evaluation. The licensee provided the results of the ASME Code stress evaluation in Table 3 of DEI Calculations C-037-2201-00-01 and C-037-2201-00-02. These results are for normal and upset conditions. In the supplement (response to request for additional information (RAI)-06), the licensee clarified that Table 3 of DEI Calculation C-037-2201-00-02, Revision 0 summarizes the analysis results of having two studs untensioned, separated by nine tensioned studs. The Table 3 results show that the increase in stress due to out-of-service studs are all below the ASME BPV Code stress limits for normal and upset conditions for both Hatch units.
The DEI Calculations C-037-2201-00-01 and C-037-2201-00-02 also considered emergency and faulted conditions. For Hatch Unit 1, DEI stated in the former calculation that the normal and upset conditions included transients associated with emergency and faulted conditions. For Hatch Unit 2, DEI stated in the latter calculation that the maximum pressure from the emergency and faulted condition is 1.08 times the normal condition pressure and the corresponding allowable stress is 1.2 times the normal condition allowable stress. The NRC staff reviewed the discussion of the emergency and faulted conditions in both DEI calculations and determined that these conditions were duly considered and acceptable because (1) for Hatch Unit 1, the normal and upset conditions included transients associated with emergency and faulted conditions; and (2) for Hatch Unit 2, the higher pressure due to an emergency and faulted condition is countered by a larger increase in corresponding allowable stress value.
On December 1, 2022, the NRC staff conducted an audit of the DEI structural integrity evaluations to gain a better understanding of the stresses resulting from the cases analyzed.
The audit plan and audit summary report are ML22320A072 and ML23034A246, respectively.
Based on its review and the audit it conducted, the NRC staff has determined that the impact of one out-of-service closure bolts for Hatch Unit 1 and two out-of-service closure bolts for Hatch Unit 2 is acceptable in terms of ASME Code stresses in the remaining RPV closure studs and RPV closure flange because the stress values are less than the corresponding ASME Code Section III allowable stress values for all analyzed design-basis conditions.
3.2.1.3 Fatigue Usage The NRC staffs comparison of methodologies, discussed in Section 3.2.1 of this SE, included confirming consistency of the methodology in the determination of the impact of out-of-service studs on fatigue usage in the RPV closure flange and the remaining RPV closure studs.
Specifically, the staff confirmed that the referenced materials and fatigue curves used in the fatigue usage evaluation in the current evaluation were consistent with those in the 2017 structural integrity evaluation. Section 3 of Enclosure 1 to the application states that the fatigue usage values were calculated for the full service life of the subject components and that the increased fatigue usage values for each Hatch unit due to out-of-service studs is below the ASME Code Section III allowable of 1.0.
Based on the above, the NRC staff has determined that the impact of one out-of-service closure bolt for Hatch Unit 1 and two out-of-service closure bolts for Hatch Unit 2 is acceptable in terms of fatigue usage because the increased fatigue usage values is below the ASME Code Section III allowable value of 1.0.
3.2.1.4 Evaluation of Impact on RPV Threads-in-Flange The NRC staff noted that the application did not discuss the impact of the proposed change to the Hatch TS Table 1.1-1 to have less than a full set of tensioned studs on the RPV threads-in-flange. Because the RPV threads-in-flange are the threads into which the studs are installed in the RPV flange, there could be a potential impact of the licensees proposed change on the RPV threads-in-flange. The staff noted that ASME Section XI Code Case (CC) N-864 may be used by licensees to forgo the ISI examination of the RPV threads-in-flange for a certain number of ISI intervals without the need for an alternative request, because the code case is conditionally approved in Regulatory Guide (RG) 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 20 (ML21181A222), which is incorporated by reference in 10 CFR 50.55a. The staff also noted that the proposed change to the Hatch TS Table 1.1-1 to have less than a full set of tensioned studs is inconsistent with the technical basis for CC N-864 since the technical basis assumes a full set of tensioned studs.
In its response to RAI-07 to the letter dated January 20, 2023 (ML23020A902), the licensee stated that, Plant Hatch does not currently use CC N-864 but does not wish to preclude usage of CC N864 with one stud out of service (Unit 1) or two studs out of service (Unit 2).
The licensee also showed that the impact of the proposed change on the RPV threads-in-flange is equal to or bounded by the technical basis of CC N-864. The NRC staff reviewed the licensees response and determined that the impact of the proposed change would have no impact on the Hatch RPV threads-in-flange because the licensee adequately showed that the loads representing the cases, in the proposed change, were equal to or bounded by those in the technical basis of CC N-864.
3.2.2 Evaluation of Leakage Integrity In Section 3 of Enclosure 1 of its application dated August 19, 2022 (ML22231B055), the licensee provided the maximum increased separation values of the RPV flanges due to the out-of-service studs at both Hatch units, as determined in DEI Calculations C-037-2201-00-01 and C-037-2201-00-02. The NRC staff confirmed that the methodology for determining the separation values of the RPV flanges in these two DEI calculations is consistent with that used in the 2017 evaluation. The evaluation in DEI Calculation C-037-2201-00-01 analyzed the impact of one untensioned stud and one stud that fails during service on flange separation of the Hatch Unit 1 RPV closure flange. The evaluation in DEI Calculation C-037-2201-00-02 analyzed the impact of two untensioned studs (separated by at least nine studs) and two studs (separated by at least nine studs) that fail during service on flange separation of the Hatch Unit 2 RPV closure flange. In its letter dated August 19, 2022, the licensee stated, in part, that, The condition with two closure studs out of service has a modest impact on the stresses associated with the stud Maximum Membrane and Maximum Membrane plus Bending stress. Each of these stress values remain below the appropriate ASME Code allowable stresses.
The condition with two closure studs out of service causes an additional flange separation at the inner o-ring of 0.0066 inch. The total flange separation is less than the o-ring minimum springback of 0.010 inch. Due to rotation of the flanges under stud loading, the flange separation at the outer ring is less than the inner o-ring. Therefore, compression on the gasket is maintained, and the flange separation will not result in additional risk of leakage.
The NRC staffs evaluation determined that the increased separation values of the RPV flanges at both Hatch units is acceptable because they are below the O-ring minimum springback value of 0.010 inch is sufficient to demonstrate limited risk of leakage at RPV flange. Accordingly, the staff finds that the licensee has adequately addressed the impact of out-of-service closure bolts on the leak integrity of the RPV flange of both Hatch units.
In Section 2.1 of Enclosure 1 its letter dated August 19, 2022, the licensee stated, in part, that, Reactor vessel top head flange leakage detection is provided for both HNP Units 1 and 2. A connection is provided on the reactor vessel flange into the annulus between the two metallic seal rings used to seal the reactor vessel and top head flanges. This connection permits detection of leakage from the inside of the reactor vessel past the inner seal ring. The connection is piped to a pressure switch having an associated alarm in the main control room. The reactor vessel top head flange leakage detection is described in the HNP Unit 1 FSAR Section 7.8.5.5 (Reference 4) and in the HNP Unit 2 FSAR Sections 5.6.2.5 and 5.2.7.2.2.1 (Reference 5).
A Class 1 system leakage test in accordance with ASME Code IWB-5220 is conducted prior to each plant startup following a reactor refueling outage at a test pressure of 1045 psig.
Plant procedures provide the methodology for tensioning the reactor pressure vessel closure studs, and the studs at HNP Units 1 and 2 are typically tensioned four studs at a time in a four-fold symmetric pattern. If the reactor pressure vessel head is to be installed with an untensioned stud(s), the tensioning pattern will be reviewed to ensure that the sealing o-rings are fully compressed before the stud(s) not being tensioned is encountered in the pattern. The final tensioned condition of the studs is verified by an elongation measurement when all required studs have been tensioned. Elongations must be within acceptance criteria. The acceptance criteria for the final stud elongations are established to ensure that ASME Code stress limits are met for specified service loads for the worst-case flange and stud bending that result from various tensioning patterns, including an untensioned stud(s). The ASME Code examination requirements would also be reviewed to verify the requirements are being met or if relief is necessary.
The NRC staff finds that the leakage detection system of each Hatch unit provides adequate monitoring of leakage at the RPV flange because if there was leakage past the inner O-ring, the pressure switch connection between the two sealing O-rings would detect the leakage before there is leakage past the outer O-ring.
3.2.3 Evaluation of proposed Technical Specifications changes Hatch Units 1 and 2 TSs define the terms operable and operability as:
A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
In its letter dated January 30, 2024, the licensee states, in part, that, The terms operable and operability are not defined in the regulations related to commercial nuclear power. The terms are defined in the TS for each plant and, therefore, only have meaning as used relative to that document. The scope of structures, systems, and components (SSCs) considered within the operability determination process is limited to those SSCs that are required to be operable by a TS limiting condition for operation (LCO). An SSC that is not directly addressed by a TS LCO may perform a necessary support function for an SSC that is directly required by a TS LCO. The SSC providing the support function is itself not subject to operability determination process, but its capabilities are a consideration in determining operability of the supported SSC that is directly required by a TS LCO (Reference 3).
The NRC staff notes that Reference 3 is the Nuclear Energy Institute (NEI) 18-03, Operability Determination, October 2019 (ML19284C872). The licensees statement above is based on Section 2.1 of NEI 18-03 which states, in part, that, Operational impacts of deficient conditions of other SSC not subject to TS LCOs should be considered and evaluated under existing processes (i.e. CAP), but are not subject to the operability process.
The NRC does not endorse NEI 18-03 and so the above language in Section 2.1 of NEI 18-03 was not adopted in the revision of NRC Inspection Manual Chapter (IMC) 0326, Operability Determinations, dated October 1, 2019 (ML19273A898). For these amendments and the associated TS Table 1.1-1 and proposed license conditions for Hatch Units 1 and 2, the NRC relied on the Hatch TS definitions of operable and operability and does not approve a plant-specific precedent use of NEI 18-03.
The NRC staffs review determined that the RPV closure bolts provide a necessary support function to support TS operability and that the licensees proposed license condition ensures the basis for determining if the RPV closure bolts can perform their support function(s). Based on the staffs evaluation in Sections 3.2.1 and 3.2.2 above, the staff finds that the licensees analysis demonstrated reasonable assurance that operating the plant with a relaxed number of required fully tensioned closure bolts (i.e., 51 fully tensioned bolts for Hatch Unit 1 and 54 fully tensioned bolts for Hatch Unit 2) will not adversely affect the operation of Hatch Units 1 and 2.
The proposed license condition will provide additional clarity on operation in Modes 1-4 and applicability for RPV head closure bolt to be considered fully tensioned in accordance with ASME Section XI acceptance criteria. Therefore, the staff finds reasonable assurance that the minimum number of fully tensioned bolts specified for Hatch Units 1 and 2 will continue to ensure the operability of the reactor pressure vessel (RPV), and that the structural integrity of the Hatch RPV closure bolts and RPV closure flange, and the leak tightness of the reactor coolant system pressure boundary.
The NRC staffs evaluation of the licensees analysis as described in Sections 3.2.1 and 3.2.2 concluded that the impact of one not fully tensioned stud for Hatch Unit 1 and two not fully tensioned studs for Hatch Unit 2 (with 9 tensioned studs between them) is acceptable because the structural integrity and leakage integrity of the Hatch RPV closure studs and RPV closure flange would continue to be maintained. Leak tightness of the RPV flange is part of the overall reactor coolant pressure boundary (RCPB), which is necessary for specific TS LCOs to be met.
TS LCO 3.4.1, Recirculation Loops Operating and TS LCO 3.4.4, RCS Operational Leakage are examples of LCOs which rely upon an intact RCPB. The Modes of Applicability Requirements for these LCOs remain unaffected by the modified TS changes. The staff concludes that the proposed TS changes would ensure that the RPV head closure bolts would continue to provide their necessary support function(s) and the affected LCOs would continue to require the lowest functional capability or performance levels of equipment required for safe operation of the facility.
Based on the above, the NRC staff finds the proposed TS changes to Table 1.1-1 and license conditions to Hatch Units 1 and 2 acceptable because the modified TSs would continue to specify the lowest functional capability or performance levels of equipment required for safe operation of the facility in accordance with 10 CFR 50.36. This provides reasonable assurance of adequate protection of public health and safety and promotes the common defense and security and protects the environment.
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 March 21, 2024. The State official confirmed the State of Georgia had no comments on April 8, 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. 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 had previously issued a proposed finding that the amendments involve no significant hazards consideration on November 1, 2022 (87 FR 65834). Due to changes to the scope of the initial submittal, the Commission issued a further proposed finding that the amendments involve no significant hazards consideration on February 20, 2024 (89 FR12876) 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.
Based on the information submitted, the NRC staff finds that the proposed changes to footnotes (a) and (b) of TS Table 1.1-1, MODES, satisfy 10 CFR 50.36 because the TSs would continue to specify the lowest functional capability or performance levels of equipment required for safe operation of the facility. The proposed Unit 1 license condition, 2.C.(12) and Unit 2 license condition, 2.C.(3)(j) would provide additional clarity on operating Mode applicability and RPV head closure bolt tensioning in accordance with ASME Section XI and 10 CFR 50.55a. The NRC staff also finds that the proposed changes to TS Table 1.1-1, MODES, satisfy the requirements of GDCs 1, 14, 30 in Appendix A to 10 CFR 50.
Based on the above, the NRC has reasonable assurance of structural integrity and leakage integrity of the Hatch RPV head closure studs and flange and that leakage will be monitored and maintained. As such, the NRC staff finds that the proposed changes to the TS as described in the application as supplemented, are acceptable.
Principal Contributors: David Dijamco, Ravi Grover, Kaihwa Hsu Date: April 24, 2024
Package, ML24114A070 Amendment, ML23032A332 NCP-2023-004, ML23248A490 OFFICE NRR/DORL/LPL2-1/PM*
NRR/DORL/LPL2-1/LA NRR/DNRL/NVIB/BC NAME DKalathiveettil KGoldstein ABuford DATE 01/31/2023 01/29/2024 02/22/2024 OFFICE NRR/DSS/STSB/BC NRR/DEX/EMIB/BC OGC - NLO NAME SMehta SBailey IMurphy DATE 02/22/2024 02/22/2024 03/28/2024 OFFICE NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/DD NRR/DORL/LPL2-1/PM NAME MMarkley (JPelton For)
JPelton DKalathiveettil DATE 04/24/2024 04/24/2024 04/24/2024