ML24348A228
| ML24348A228 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/25/2025 |
| From: | Dawnmathews Kalathiveettil Plant Licensing Branch II |
| To: | Coleman J Southern Nuclear Operating Co |
| Kalathiveettil, D | |
| References | |
| EPID L-2024-LLA-0017 | |
| Download: ML24348A228 (21) | |
Text
February 25, 2025 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 AMENDMENT NOS. 327 AND 272, REGARDING LICENSE AMENDMENT REQUEST TO ADOPT AN ALTERNATIVE SEISMIC METHOD FOR CATEGORIZATION OF STRUCTURES, SYSTEMS, AND COMPONENTS (EPID L-2024-LLA-0017)
Dear Jamie Coleman:
The U.S. Nuclear Regulatory Commission (NRC, or the Commission) has issued the enclosed Amendment No. 327 to the Renewed Facility Operating License (RFOL) No. DPR-57, and Amendment No. 272 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 February 20, 2024, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24051A239), as supplemented by letter dated July 26, 2024 (ML24208A169).
The amendments are related to Hatchs adoption of Title 10 of the Code of Federal Regulations (10 CFR) 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. Specifically, the amendments modify Hatch, Unit 1, RFOL license condition 2.C.(11) and Hatch, Unit 2, RFOL license condition 2.C.(3)(i) to allow the use of an alternative approach for evaluating seismic risk for categorization of structures, systems, and components under Hatchs approved 10 CFR 50.69 program, and removes certain pre-program implementation items that have been completed.
J. Coleman 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 Zach Turner for/
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. 327 to DPR-57
- 2. Amendment No. 272 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. 327 Renewed License No. DPR-57 1.
The Nuclear Regulatory Commission (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 February 20, 2024, as supplemented by letter dated July 26, 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. 327, 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.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. 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: February 25, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.02.25 15:31:29 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 327 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of Renewed Facility Operating License No. DPR-57 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating License Remove Insert Renewed License No. DPR-57 Amendment No. 327 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. 327, 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. 327
- 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.
In addition, SNC is approved to implement 10 CFR 50.69 using the alternative seismic approach described in SNC's letter dated February 20, 2024, for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs, as specified in Renewed License Amendment No. 327 dated February 25, 2025.
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).
(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 Attachments:
Samuel J. Collins, Director Office of Nuclear Reactor Regulation Appendix A - Technical Specifications Appendix B - Environmental Protection Plan Date of Issuance: January 15, 2002 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. 272 Renewed License No. NPF-5 1.
The Nuclear Regulatory Commission (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 February 20, 2024, as supplemented by letter dated July 26, 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. 272, 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 90 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. 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: February 25, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.02.25 15:32:40 -05'00'
ATTACHMENT TO LICENSE AMENDMENT NO. 272 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of Renewed Facility Operating License No. NPF-5 with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Renewed Facility Operating License Remove Insert Renewed License No. NPF-5 Amendment No. 272 (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. 272 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. 272 (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.
In addition, SNC is approved to implement 10 CFR 50.69 using the alternative seismic approach described in SNC's letter dated February 20, 2024, for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs, as specified in Renewed License Amendment No. 272 dated February 25, 2025.
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).
(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.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 327 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-57, AND AMENDMENT NO. 272 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 February 20, 2024, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24051A239), as supplemented by letter dated July 26, 2024 (ML24208A169), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC) for the Edwin I. Hatch Nuclear Plant (Hatch), Unit Nos. 1 and 2, 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. The LAR is related to Hatchs adoption of 10 CFR 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors. Specifically, the amendments modify Hatch, Unit 1, Renewed Facility Operating License (RFOL) license condition 2.C.(11) and Hatch, Unit 2, RFOL license condition 2.C.(3)(i) to allow the use of an alternative approach for evaluating seismic risk for categorization of structures, systems and components (SSCs) under Hatchs approved 10 CFR 50.69 program and removes certain pre-program implementation items that have been completed.
The NRC staff participated in a regulatory audit from July 1 through September 30, 2024, to ascertain the information needed to support its review and determine request for additional information, as needed. During the audit, the NRC staff identified information that needed to be submitted on the docket to reach its safety conclusion. The licensee provided this information in its supplement letter dated July 26, 2024. This information clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 16, 2024 (89 FR 26947). On October 31, 2024, the staff issued an audit summary to support the review of the LAR (ML24299A222).
1.1 Background
In the safety evaluation (SE) for Hatch, Unit Nos. 1 and 2, License Amendment Nos. 305 and 250 to RFOL Nos. DPR-57 and NPF-5, respectively, dated June 26, 2020 (ML20077J704), the NRC staff concluded that the licensees 10 CFR 50.69 program was consistent with the NRC-endorsed guidance in the Nuclear Energy Institute (NEI) 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline (ML052910035), and thus satisfied the requirements of 10 CFR 50.69(c). A license condition incorporated into the license as part of the NRC staffs decision to approve the licensees original 10 CFR 50.69 LAR stated that, 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).
In its submittal dated February 20, 2024, the licensee proposed to modify the license conditions to the Hatch RFOLs to allow the use of an alternative seismic approach for evaluating seismic risk in addition to the use of a pre-reviewed, plant-specific seismic probabilistic risk assessment (SPRA) described in the licensees previously approved 10 CFR 50.69 program.
1.2 Description of Proposed Changes A license condition was added to the Hatch RFOLs when the NRC approved the licensees use of its 10 CFR 50.69 program on June 26, 2020. As discussed in 10 CFR 50.69(b), a licensee may voluntarily comply with 10 CFR 50.69 as an alternative to compliance with the following requirements for Risk Informed Safety Class (RISC)-3 and RISC-4 SSCs after it submits, and the NRC approves, an application for a license amendment:
(i) 10 CFR Part 21, Reporting of Defects and Noncompliance; (ii)
A portion of 10 CFR 50.46a, Acceptance criteria for reactor coolant system venting systems, paragraph (b);
(iii) 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants; (iv) 10 CFR 50.55, Conditions of construction permits, early site permits, combined licenses and manufacturing licenses, paragraph (e);
(v)
Certain requirements of 10 CFR 50.55a, Codes and standards; (vi) 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, except for paragraph (a)(4);
(vii) 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors; (viii) 10 CFR 50.73, License event report system; (ix) 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants; (x)
Certain containment leakage testing requirements in 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors; and (xi)
Certain requirements of 10 CFR Part 100, Appendix A, Seismic and Geologic Siting Criteria for Nuclear Power Plants.
The current RFOL license conditions for Hatch, Units 1 and 2 (added by License Amendment Nos. 305 and 250, respectively), states the following [bracketed text denotes unit-specific license amendment numbers]:
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. [305/250], 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. [305/250],
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.
The licensee proposed in the subject LAR to amend the above license condition for Unit Nos. 1 and 2 to read as follows:
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. [305/250], dated June 26, 2020.
In addition, SNC is approved to implement 10 CFR 50.69 using the alternative seismic approach described in SNCs letter dated February 20, 2024, for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs, as specified in Renewed License Amendment No. [XXX/YYY] dated [DATE].
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).
2.0 REGULATORY EVALUATION
2.1 Applicable Regulations The regulations in 10 CFR 50.90 state that whenever a holder of an operating license desires to amend the license, including technical specifications (TSs) in the license, an application for amendment must be filed with the Commission fully describing the changes desired. The regulations at 10 CFR 50.92(a) state that determinations on whether to grant an applied for license amendment are guided by the considerations that govern the issuance of initial licenses to the extent applicable and appropriate. Both the common standards for licenses in 10 CFR 50.40(a) (regarding, among other things, consideration of the operating procedures, the facility and equipment, the use of the facility, and other TSs, or the proposals) and those specifically for issuance of operating licenses in 10 CFR 50.57(a)(3), provide that there must be reasonable assurance that the activities at issue will not endanger the health and safety of the public, and that the applicant will comply with the Commission's regulations.
The regulations in 10 CFR 50.69 allow a licensee voluntarily to comply with the requirements in 10 CFR 50.69 as an alternative to compliance with specific requirements for RISC-3 SSCs (i.e.,
safety-related SSCs that perform low safety significant (LSS) functions) and RISC-4 SSCs (i.e.,
nonsafety-related SSCs that perform LSS functions). The approval standards are set forth in 10 CFR 50.69(b)(3) and 50.69(b)(4) and state:
(3) The Commission will approve a licensee's implementation of this section if it determines that the process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs satisfies the requirements of § 50.69(c) by issuing a license amendment approving the licensee's use of this section.
(4) An applicant choosing to implement this section shall include the information in § 50.69(b)(2) as part of application. The Commission will approve an applicant's implementation of this section if it determines that the process for categorization of RISC-1, RISC-2, RISC-3, and RISC-4 SSCs satisfies the requirements of § 50.69(c).
In addition, the regulations in 10 CFR 50.69 contain requirements regarding how a licensee categorizes SSCs using a risk-informed process, adjusts treatment requirements consistent with the relative significance of the SSC, and manages the process over the lifetime of the plant. A risk-informed categorization process is employed to determine the safety significance of SSCs and place the SSCs into one of four RISC categories.
Categorization of SSCs does not allow for the elimination of SSC functional requirements, nor does it allow for equipment that is required by the deterministic design basis to be removed from the facility. Instead, 10 CFR 50.69 enables licensees to focus their resources on SSCs that make a significant contribution to plant safety. For SSCs that are categorized as high safety significant (HSS), existing treatment requirements are maintained or potentially enhanced.
Conversely, for SSCs categorized as LSS that do not significantly contribute to plant safety on an individual basis, the regulation allows for an alternative risk-informed approach to treatment that provides a reasonable level of confidence that these SSCs will satisfy functional requirements. Implementation of 10 CFR 50.69 allows licensees to improve focus on equipment that is HSS.
2.2 Applicable Regulatory Guidance The NRC staff considered the following regulatory guidance during its review of the proposed changes:
Regulatory Guide (RG) 1.201, Revision 1, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (ML061090627);
RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities (ML090410014);
RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (ML17317A256); and NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk Informed Decisionmaking (ML17062A466).
3.0 TECHNICAL EVALUATION
The NRC staff reviewed the proposed change with respect to the previously approved Hatch 10 CFR 50.69 program. In its submittal, the licensee stated, in part, that the proposed amendment to Hatch RFOLs would allow the use of an alternative approach (in addition to the current use of an SPRA for evaluating seismic risk) and that previously approved 10 CFR 50.69 categorization methods would not be impacted. The NRC staff confirmed that the LAR did not impact or change any other aspect of the licensees categorization process except for the addition of the alternative seismic approach as a categorization method to consider the seismic risk. Therefore, the NRC staffs determinations in the letter dated June 26, 2020, concerning the licensees categorization process, other than the addition of the alternative seismic approach related to the consideration of the seismic risk, remain unchanged and valid. Further, in Section 3 of the Enclosure to the LAR, the licensee stated that it would use a single approach (i.e., either SPRA or the proposed alternative seismic approach) for the categorization of an entire system. Consequently, the NRC staff did not review the licensees categorization process other than the change requested in the subject LAR.
As stated in RG 1.201, if a licensee wishes to change its categorization approach, the NRC staffs review of the resulting LAR will focus on the acceptability of the methodology and analyses relied upon in the application. Section 3.1 below summarizes the NRC staffs review of the acceptability of the proposed alternative seismic approach as described in the licensees submittal, as supplemented.
In Section 2.3 of the LAR Enclosure, Description of the Proposed Change, the licensee stated that the implementation items in the previously approved amendment were completed.
Section 3.2 below summarizes the NRC staffs review of the PRA implementation items in the previously approved amendment that would be removed from the license condition as described in the LAR.
3.1 Alternative Seismic Approach As part of its proposed process to categorize SSCs according to safety significance, the licensee proposed to use a non-PRA method to consider seismic hazards. The licensee provided a description of its proposed alternative seismic approach for considering seismic risk in the categorization process and described how the proposed alternative seismic approach would be used in the categorization process in Section 3.1 of the Enclosure to the LAR. In Section 3.1 of the Enclosure to the LAR, Categorization Process Description (10 CFR 50.69(b)(2)(i)), the licensee stated that the proposed approach for Tier 2 sites is documented in Electric Power Research Institute (EPRI) Report 3002017583, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization (ML21082A170), with the EPRI markups provided in Attachment 2 of ADAMS Accession Nos. ML20290A791 and ML21022A130. The NRC staff notes that use of the alternative seismic approach is a method not endorsed by the NRC in NEI 00-04. Therefore, a detailed NRC staff review of the licensees proposed plant-specific alternative seismic approach is provided below.
The licensee based its plant-specific evaluation, in part, on the test cases described in EPRI Report 3002017583 to determine the extent and type of unique HSS SSCs from SPRAs. The licensee stated that the test cases are applicable to Hatch and demonstrated that there are very few, if any, SSCs that would be designated HSS for seismic unique purposes. Furthermore, the test cases identified that the unique seismic insights were typically associated with seismically correlated failures and led to unique HSS SSCs. The licensee further stated that the categorization team will evaluate correlated seismic failures and seismic interactions between SSCs as described in EPRI Report 3002017583 and the referenced markups. The licensee stated that the determination of seismic insights will make use of the full power internal events PRA model supplemented by focused seismic walkdowns, and the licensee described the steps of the process used to determine the importance of SSCs for mitigating seismic events on a system basis. Based on the above, the NRC staffs independent review finds that the requirements in 10 CFR 50.69(b)(2)(ii) for the proposed plant-specific alternative seismic approach are met because the licensee sufficiently described the measures taken to assure the quality and level of detail of their proposed alternative seismic approach to evaluate the plant for the categorization of SSCs.
In Section 3.1 of the Enclosure to the LAR, the licensee cited a precedent for its proposed alternative seismic approach. The precedent, the LaSalle alternative seismic approach, as described in Exelons letter dated January 31, 2020, and all its subsequent associated supplements, as specified in License Amendment No. 249 dated May 27, 2021 (ML21082A422), includes a combination of qualitative and quantitative considerations of the mitigation capabilities as well as seismic failure modes of SSCs in the categorization process.
These considerations are based on plant-specific walkdowns for the SSCs undergoing categorization, quantification of the impact of seismic failure of SSCs subject to correlated or interaction failures, and insights obtained from prior seismic evaluations performed at LaSalle.
The licensee identified differences between Hatch and the NRC staffs approval of the precedent as documented in the LaSalle SE. The licensee described that, aside from two exceptions (i.e., the use of EPRI Report 3002017583 rather than EPRI Report 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization (not publicly available), and the LaSalle request for additional information responses that are outside the scope of the LAR), the licensee would follow the same alternative seismic approach in its proposed 10 CFR 50.69 categorization process as the approved plant-specific LaSalle 10 CFR 50.69 license amendment.
In the licensees July 26, 2024, supplement to the LAR, the licensee described its site-specific PRA configuration maintenance process and clarified that the proposed 10 CFR 50.69 categorization process follows the same approach as the previously approved LaSalle approach, with the exception of site-specific LaSalle information. Based on information provided by the licensee in the LAR Enclosure, the NRC staff understands that the technical criteria in EPRI Report 3002017583 are unchanged from its predecessor EPRI Report 3002012988, and that the test cases are applicable to Hatch and are, therefore, justified for use in the licensees proposed plant-specific alternative seismic approach. The NRC staff confirmed that the test cases in EPRI Report 3002017583 used by the licensee to support its proposed alternative seismic approach provided a sufficient plant-specific evaluation of the applicability and differences for Hatch as compared to the plant-specific approach approved by the NRC for LaSalle. The NRC staffs independent review finds that the information presented in the licensees LAR and licensees analysis for items referenced in EPRI Report 3002017583 provides a sufficient description of, and basis for, acceptability of the evaluations to be conducted to satisfy 10 CFR 50.69(c)(1)(iv) for the Hatch plant-specific alternative seismic approach. The NRC staff determined that there is reasonable confidence that evaluated LSS safety-related SSCs would have sufficient safety margins maintained, and that any potential increases in core damage frequency and large early release frequency resulting from the changes in SSC treatment are small. Based on the above, the NRC staffs independent review finds that the requirements in 10 CFR 50.69(b)(2)(iv) are met for the proposed plant-specific alternative seismic approach.
Based on the information provided in the licensees submittal, as supplemented, the NRC staffs independent review finds that the licensee provided acceptable bridging analysis for the Hatch plant-specific alternative seismic approach, in part, based on analysis supporting approval of LaSalles plant-specific alternative seismic approach because: (1) the differences between the licensees proposed alternative seismic approach and the alternative seismic approach previously approved for LaSalle are addressed; (2) there are no differences in the technical criteria used in EPRI Report 3002017583 and its predecessor, EPRI Report 3002012988, for use in this application; and (3) all references needed to support the NRC staffs finding on the proposed alternate seismic approach have been cited by the LAR.
3.1.1 Evaluation of Technical Acceptability of the PRAs Used for Test Cases Supporting the Proposed Alternative Seismic Approach In Section 3.1 of the Enclosure to the LAR, the licensee discussed the precedent, including the test cases (also referred to as case studies in Reference 1), mapping approach, and conclusions on the determination of unique HSS SSCs from the test cases, which were used by the licensee to support its proposed plant-specific alternative seismic approach. The licensee stated that Hatch is using test case information from EPRI Report 3002017583 which demonstrated that seismic risk is adequately addressed for Tier 2 sites. The licensee stated that Tier 2 seismic demand sites have a lower likelihood of seismically induced failures and fewer challenges to plant systems than test case plants. Section 3.1.2 below provides the NRC staffs independent evaluation of the Tier 2 criteria for Hatch. The NRC staff reviewed and evaluated the technical acceptability of the PRAs used in the test cases for Plants A, C, and D in EPRI Report 3002017583, as well as the applicability of these test cases to Hatch. The NRC staff also evaluated the peer review process, resolution of peer review findings, and key assumptions and sources of uncertainty for Plants A, C, and D.
Based on the above, the NRC staffs independent review finds the technical acceptability of PRAs used for the Plant A, C, and D test cases in EPRI Report 3002017583, the mapping approach used in those test cases, and the conclusions on the determination of unique HSS SSCs from the test cases in the precedent are applicable to Hatchs proposed plant-specific alternative seismic approach. Therefore, the NRC staff concludes that the Plant A, C, and D PRAs were technically acceptable and applicable for use in support of the licensees proposed alternative seismic approach, the mapping of SSCs between the SPRA, the full power internal events PRA, and, as applicable, the fire PRA (FPRA) for the Plant A, C, and D test cases. The NRC staffs independent review further finds that the licensees plant-specific evaluation is technically justified to support conclusions on the determination of unique HSS SSCs from SPRAs in the Plant A, C, and D test cases in EPRI Report 3002017583, and that the licensees proposed plant-specific alternative seismic approach is applicable to Hatch.
3.1.2 Evaluation of the Criteria for the Proposed Alternative Seismic Approach In Section 3.1 of the Enclosure to the LAR, the licensee provided the basis for Hatch being a Tier 2 plant. The licensee referred to the following criteria provided in EPRI Report 3002017583:
Tier 1: Plants where the ground motion response spectrum (GMRS) peak acceleration is at or below approximately 0.2g or where the GMRS is below or approximately equal to the safe shutdown earthquake (SSE) between 1.0 Hz and 10 Hz.
Tier 2: Plants where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is greater than in Tier 1 but not high enough to be treated as Tier 3. At these sites, the unique seismic categorization insights are expected to be limited.
Tier 3: Plants where the GMRS to SSE comparison between 1.0 Hz and 10 Hz is high enough that the NRC required the plant to perform an SPRA to respond to the 10 CFR 50.54(f) letter regarding the Fukushima Dai-Ichi Accident (ML12053A340).
The licensee explained that the Tier 2 criterion is based on a comparison of the GMRS to SSE; however, the design basis earthquake (DBE) is the licensing basis earthquake for Hatch which is used for this comparison. The licensee stated that this substitution was acknowledged in Section 3.1, Plant Seismic Design Basis, of the NRC staffs assessment of the information provided pursuant to the 10 CFR 50.54(f) request relating to Near-Term Task Force (NTTF)
Recommendation 2.1 (ML15097A424). The regulations in 10 CFR 100 recognize that SSE defines the earthquake commonly referred to as the DBE; therefore, the licensees comparison of GMRS to DBE is acceptable.
The licensee compared the GMRS to the DBE and explained that the GMRS to DBE comparison is above the Tier 1 threshold under criteria in the EPRI report but is not high enough for the NRC to require the plant to perform an SPRA to respond to NTTF Recommendation 2.1.
This determination on the licensees plant-specific evaluation is supported by its response to NTTF Recommendation 2.1 (ML14092A017). The NRC staff reviewed the plant-specific evaluation and concludes that the proposed Tier 2 criteria to determine the applicability and use of the proposed alternative seismic approach is acceptable. Based on its review of the GMRS to DBE comparison, the NRC staff finds that the licensees seismic hazard meets the criteria for its proposed alternative seismic approach.
3.1.3 Evaluation of the Implementation of Conclusions The categorization conclusions supporting the request for Hatch indicated that seismic specific failure modes resulted in HSS categorization uniquely from SPRAs. Seismic specific failure modes, such as correlated failures, relay chatter, and passive component structural failure modes, can influence the categorization process. The NRC staff reviewed the proposed alternative seismic approach for Hatch to evaluate whether the categorization-related conclusions were assessed appropriately and provisions for implementation were sufficient.
In Section 3.1 of the Enclosure to the LAR, the licensee described the implementation of the proposed alternative seismic approach. For HSS SSCs uniquely identified by the PRA models but having design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events, the licensee stated that it will address these SSCs using non-PRA based qualitative assessments in conjunction with any seismic insights provided by the PRA. For components that are HSS due to the FPRA, but not HSS due to the full power internal events PRA, the licensee stated that the categorization team will review design-basis functions during seismic events or functions credited for mitigation and prevention of severe accidents caused by seismic events and characterize these as additional qualitative inputs to the Integrated Decision-making Panel (IDP).
The licensee described that the categorization team will evaluate plant-specific seismic reviews.
The licensee stated that the objective of the seismic reviews is to identify plant-specific seismic insights that might include potentially important impacts such as:
Impact of relay chatter
Implications related to potential seismic interactions such as with block walls
Seismic failures of passive SSCs such as tanks and heat exchangers
Any known structural or anchorage issues with a particular SSC
Components implicitly part of PRA-modeled functions (including relays)
Hatch stated that it will follow the same categorization process that the NRC approved as precedent for LaSalle and incorporates the LaSalle RAI responses. Hatch described its exceptions to the precedent, including LaSalle site-specific information, as evaluated above.
The NRC staff reviewed the licensees proposed alternative seismic approach and determined that the NRC-approved precedent is applicable to the proposed alternative seismic approach and the plant-specific evaluation on the implementation of the alternative seismic approach is acceptable. Based on its independent review of the licensees proposed alternative seismic approach, in conjunction with the requirements in 10 CFR 50.69 and the corresponding statement of consideration (ML15097A424), the NRC staff finds that the proposed alternative seismic approach provides reasonable confidence in the evaluations required by 10 CFR 50.69(c)(1)(ii) as well as 10 CFR 50.69(c)(1)(iv) because:
The proposed alternative seismic approach includes qualitative consideration of seismic events at several steps of the categorization process, including documentation of the information for presentation to the IDP as part of the integrated, systematic process for categorization.
The proposed alternative seismic approach presents system specific seismic insights to the IDP for consideration as part of the IDP review process as each system is categorized, thereby providing the IDP a means to consider potential impacts of seismic events in the categorization process.
The insights presented to the IDP include potentially important seismically induced failure modes, as well as mitigation capabilities of SSCs during seismically induced design basis and severe accident events consistent with the conclusions on the determination of unique HSS SSCs from SPRAs in EPRI Report 3002017583. The insights will use prior plant-specific seismic evaluations, and therefore, in conjunction with performance monitoring for the proposed alternative seismic approach, reasonably reflect the current plant configuration. Further, the recommendation for categorizing civil structures in the alternative seismic approach provides appropriate consideration of such failures from a seismic event.
The proposed alternative seismic approach includes qualitative considerations and insights related to the impact of a seismic event on SSCs for each SSC that is categorized and does not limit the scope to SSCs from the test cases supporting this application.
3.1.4 Consideration of Changes to Seismic Hazard An important input to the NRC staffs evaluation of the proposed alternative seismic approach is the current knowledge of the seismic hazard at Hatch. The possibility exists for the seismic hazard at the site to increase or decrease such that the criteria for use of the proposed alternative seismic approach are challenged. In such a situation, the categorization process may be impacted from a seismic risk perspective either solely due to the seismic risk or by the integrated importance measure determination.
In Section 3.1 of the Enclosure to the LAR, the licensee stated, in part, that:
If the [Hatch] seismic hazard changes from medium risk (i.e., Tier 2) at some future time, prior NRC approval, under 10 CFR 50.90, will be requested even if the [Hatch] feedback process determines that a process different from the proposed alternative seismic approach is warranted for seismic risk consideration in categorization under 10 CFR 50.69.
The licensee also stated that it will follow its categorization review and adjustment process to review the changes to the plant and update, as appropriate, the SSC categorization in accordance with 10 CFR 50.69(e) and the EPRI Report 3002017583 SSC categorization criteria for the updated tier, and that this includes use of the licensees corrective action process. The licensee described the processes it will follow if the seismic hazard is reduced such that it meets the criteria for Tier 1 in EPRI Report 3002017583 or is increased to a degree that an SPRA becomes necessary to demonstrate adequate seismic safety.
The NRC staffs independent review finds that the consideration of changes to the seismic hazard in the licensees proposed alternative seismic approach is consistent with the applicable regulations and the precedent approved by NRC staff for LaSalle. Therefore, the NRC staffs evaluation of the proposed changes to Hatchs seismic hazard against the requirements in 10 CFR 50.69(e)(1), 10 CFR 50.69(e)(3), and 10 CFR 50.69(d)(2)(ii), as well as the resulting conclusion on consideration of changes to the seismic hazard for the NRC approved precedent is applicable to this licensees proposed alternative seismic approach. Consequently, the NRC staff finds that the consideration of changes to the seismic hazard at Hatch is acceptable because: (1) the criteria for use of the proposed alternative seismic approach are clear and traceable, (2) the proposed alternative seismic approach includes periodic reconsideration of the seismic hazard as new information becomes available, (3) the proposed alternative seismic approach satisfies the requirements in 10 CFR 50.69 as discussed above, and (4) the licensee has included a proposed license condition in the LAR to require NRC approval for a change to the specified seismic categorization approach.
3.1.5 Monitoring of Inputs to and Outcome of Proposed Alternative Seismic Approach In Section 3.5 of the Enclosure to the LAR, Feedback and Adjustment Process, the licensee described its feedback and adjustment (i.e., performance monitoring) process. The licensee described its performance monitoring process, configuration control process, and problem identification and resolution process. In the LAR, the licensee stated, in part, that:
To more specifically address the feedback and adjustment (i.e., performance monitoring) process as it pertains to the proposed alternative seismic method for Tier 2 sites [], implementation of the SNC design control and corrective action programs provides assurance that the inputs for the qualitative determinations for seismic continue to remain valid to maintain compliance with the requirements of 10 CFR 50.69(e).
Furthermore, the licensee then goes on to state, in part, that:
Scheduled periodic reviews no more frequent than once every two refueling outages will evaluate new insights resulting from available risk information (i.e.,
PRA model or other analysis used in the categorization) changes, design changes, operational changes, and SSC performance. If it is determined that these changes have affected the risk information or other elements of the categorization process such that the categorization results are more than minimally affected, then the risk information and the categorization process will be updated.
The NRC staffs independent review finds that consideration of the feedback and adjustment process in the licensees proposed alternative seismic approach is acceptable because: (1) the licensees programs provide reasonable assurance that the existing seismic capacity of LSS components would not be significantly impacted, and (2) the monitoring and configuration control program ensures that potential degradation of seismic capacity would be detected and addressed before significantly impacting the plants risk profile. Therefore, the NRC staffs independent review finds that the potential impact of the seismic hazard on the categorization of RISC-3 SSCs is maintained acceptably low and the requirements in 10 CFR 50.69(c)(1)(iv) are met for the proposed alternative seismic approach.
3.2 PRA Implementation Items As described in Section 2.2 of this SE above, the current license condition for Hatch, Units 1 and 2, (added by License Amendment Nos. 305 and 250) states the following [bracketed text denotes unit-specific license amendment numbers]:
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. [305/250], 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. [305/250], 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.
In Section 2.3 of the Enclosure to the LAR, the licensee stated that the implementation items identified in the last paragraph of the license condition were completed as required prior to the implementation of the 10 CFR 50.69 categorization process, which began in June 2020. The licensee proposed deleting the paragraph specific to the implementation items in the revised license condition because the paragraph is no longer applicable.
3.3 Technical Conclusion Based on the NRC staffs independent evaluation in the above sections, the staff finds the licensees non-PRA methods for assessing risk for seismic hazards, a deviation from NEI 00-04, acceptable and determines that the licensees proposed 10 CFR 50.69 program, with the proposed license condition, would continue to meet the requirements in 10 CFR 50.69.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the State of Georgia official was notified of the proposed issuance of the amendments on December 20, 2024. The State official confirmed the State of Georgia had no comments on December 20, 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 in the Federal Register on April 16, 2024 (89 FR 26947), 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 to 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 Contributors: D. Silverstein, NRR S. Alferink, NRR R. Wang, NRR Date: February 25, 2025
ML24348A228 OFFICE NRR/DORL/LPL2-1/PM*
NRR/DORL/LPL2-1/LA NRR/DRA/APLC/BC NAME DKalathiveettil KZeleznock ANeuhausen (A)
DATE 12/12/2024 12/20/2024 9/03/2024 OFFICE OGC - NLO NRR/DEX/EXHB NRR/DORL/LPL2-1/BC NAME ALeatherman BHayes MMarkley DATE 1/16/2025 2/14/2025 2/25/2025 OFFICE NRR/DORL/LPL2-1/PM NAME DKalathiveettil (ZTurner for)
DATE 2/25/2025