Text

Issuance of Amendment Nos. 230 and 212, Regarding Alternative Source Term Using Regulatory Guide 1.183 Revision 1
	ML26062A818
Person / Time
Site:	Vogtle ; (NPF-068, NPF-081)
Issue date:	03/26/2026
From:	John Lamb; Plant Licensing Branch II
To:	Coleman J; Southern Nuclear Operating Co
References
	EPID L-2025-LLA-0080
	Download: ML26062A818 (0)
	v • d • e

March 26, 2026 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway, N 274 EC Birmingham, AL 35243

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 230 AND 212, REGARDING ALTERNATIVE SOURCE TERM USING REGULATORY GUIDE 1.183, REVISION 1 (EPID L-2025-LLA-0080)

Dear Ms. Coleman:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 230 to Renewed Facility Operating License NPF-68 and Amendment No. 212 to Renewed Facility Operating License NPF-81 for the Vogtle Electric Generating Plant (Vogtle),

Units 1 and 2, respectively. The amendments consist of changes in response to your application dated May 12, 2025, as supplemented by letter dated February 16, 2026.

The amendments revise the licensing basis to support a full scope application of an Alternative Source Term methodology consistent with the guidance of Regulatory Guide (RG) 1.183, Revision 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

A copy of the related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's monthly Federal Register notice.

J. Coleman If you have questions, please contact me at 301-415-3100 or via email at John.Lamb@nrc.gov.

Sincerely,

/RA/

John G. Lamb, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425

Enclosures:

1. Amendment No. 230 to NPF-68

2. Amendment No. 212 to NPF-81

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. 50-424 VOGTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 230 Renewed License No. NPF-68 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Renewed Facility Operating License No. NPF-68 filed by the 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 May 12, 2025, as supplemented by letter dated February 16, 2026 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, by Amendment 230, the license is hereby amended to authorize revision to the Updated Final Safety Analysis Report, as set forth in the application dated May 12, 2025, as supplemented by letter dated February 16, 2026.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days. The Updated Final Safety Analysis Reports for Vogtle, Units 1 and 2, are updated by October 31 of every even-numbered year.

FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: March 26, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.03.26 10:23:33 -04'00' SOUTHERN NUCLEAR OPERATING COMPANY, INC.

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-425 VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 212 Renewed License No. NPF-81 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Renewed Facility Operating License No. NPF-81 filed by the 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 May 12, 2025, as supplemented by letter dated February 16, 2026, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, by Amendment 212, the license is hereby amended to authorize revision of the Updated Final Safety Analysis Report, as set forth in the application dated May 12, 2025, as supplemented by letter dated February 16, 2026.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 120 days. The Updated Final Safety Analysis Reports for Vogtle, Units 1 and 2, are updated by October 31 of every even-numbered year.

FOR THE NUCLEAR REGULATORY COMMISSION Michael Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: March 26, 2026 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2026.03.26 10:24:17 -04'00' SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 AMENDMENT NO. 230 TO RENEWED FACILITY OPERATING LICENSE NPF-68 AMENDMENT NO. 212 TO RENEWED FACILITY OPERATING LICENSE NPF-81 SOUTHERN NUCLEAR OPERATING COMPANY, INC.

DOCKET NOS. 50-424 AND 50-425

1.0 INTRODUCTION

By letter dated May 12, 2025 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML25132A313), as supplemented by letter dated February 16, 2026 (ML26047A009), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) to the U.S. Nuclear Regulatory Commission (NRC, the Commission) for changes to the Updated Final Safety Analysis Report (UFSAR) for Vogtle Electric Generating Plant (Vogtle), Units 1 and 2.

The proposed amendments would revise the licensing basis to support full implementation of the alternative source term (AST) radiological analysis methodology following the guidance in NRC Regulatory Guide (RG) 1.183, Revision (Rev.) 1, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Plants, October 2023 (ML23082A305).

The proposed amendments would revise the licensing basis to use an AST in evaluating the offsite and control room (CR) radiological consequences of the Vogtle, Units 1 and 2, design basis accidents (DBAs) as allowed by Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.67, Accident source term, and described in NRC RG 1.183, Rev. 1.

The supplement dated February 16, 2026, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on July 8, 2025 (90 FR 30110).

2.0 REGULATORY EVALUATION

2.1 Proposed Change The proposed amendments would revise the licensing basis to adopt the method and full implementation of AST in evaluating the offsite and CR radiological consequences of the Vogtle, Units 1 and 2, DBAs.

2.2 AST Regulatory Evaluation SNCs request was submitted pursuant to 10 CFR 50.67, Accident source term, which provides a mechanism for licensed power reactors to replace the traditional source term used in the radiological consequence analyses of DBAs discussed in Section 3.0 below, as described in NRC RG 1.183, Rev. 1. Vogtle, Units 1 and 2, current DBA radiological consequence analyses are based on the source term from U.S. Atomic Energy Commission Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, dated March 23, 1962 (ML021720780).

The NRC staff evaluated SNCs analysis of the radiological consequences of the affected DBAs for implementation of the AST methodology, against the radiological dose requirements specified in 10 CFR 50.67(b)(2), and dose limits specified in 10 CFR Part 50, Appendix A, General Design Criteria (GDC) for Nuclear Power Plants, Criterion 19, Control Room.

Section 50.67(b)(2) states:

The NRC may issue the amendment only if the applicants analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv (25 rem)1 total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) TEDE.

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

The NRC staffs AST evaluation is based on the following regulations, RGs, and standards:

The regulations in 10 CFR 50.67, Accident source term, The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, GDC 19, Control room, NRC RG 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982 (Reissued February 1983) (ML003740205),

The regulation 10 CFR 50.49, Environmental qualification of electric equipment important to safety for nuclear power plants, 1 The use of 0.25 SV (25 rem) TEDE is not intended to imply that this value constitutes an acceptable limit for emergency doses to the public under accident conditions. Rather, this 0.25 Sv (25 rem) TEDE value has been stated in this section as a Reference value, which can be used in the evaluation of proposed design basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.

NRC RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 1, October 2023 (ML23082A305)

NRC RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, Revision 0, June 2003 (ML031530505)

NRC RG 1.196, Control Room Habitability at Light-Water Nuclear Power Reactors, Revision 1, January 2007 (ML063560144)

NRC RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants, Revision 1, March 2007 (ML070350028)

NRC Safety Guide 23, Onsite Meteorological Programs, February 17, 1972 (ML020360030)

NUREG-0800, Standard Review Plan (SRP) Section 6.4, Control Room Habitability System, Revision 3, March 2007 (ML070550069)

NUREG-0800, SRP Section 6.5.2, Containment Spray as a Fission Product Cleanup System, Revision 4, March 2007 (ML070190178)

NUREG-0800, SRP Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, Revision 0, July 2000 (ML003734190).

This safety evaluation (SE) addresses the impact of the proposed changes on previously analyzed DBAs radiological consequences and the acceptability of the revised analysis results.

The regulatory requirements from which the NRC staff based its acceptance are the accident radiation dose values in 10 CFR 50.67, and the accident specific guideline values in Section 4.4 of RG 1.183 and Table 1 of SRP Section 15.0.1. The licensee has not proposed any significant deviation or departure from RG 1.183.

2.3 Meteorological Data and Atmospheric Dispersion Factors Regulatory Evaluation The NRC staff's evaluation of the proposed meteorological data and offsite and onsite atmospheric dispersion factors is based upon the following regulations and RGs:

The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 19, Control room NRC NUREG-0800, SRP Section 2.3.4, Short-Term Atmospheric Dispersion Estimates for Accident Releases (ML070730398)

NRC RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants (ML070350028)

NRC RG 1.145, Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants, Revision 1, November 1982 (Reissued February 1983) (ML003740205)

NRC RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 1, October 2023 (ML23082A305)

NRC RG 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants (ML031530505)

3.0 TECHNICAL EVALUATION

3.1 Radiological Consequences of DBAs Consistent with the current licensing basis design-basis accident analyses, the licensee applied a core thermal power level of 3636 megawatts thermal (MWt) in the accident analysis. This value is based on the licensed rated thermal power of 3625.6 MWt with an additional 0.3 percent margin for accident dose calculations. The application of this analytical margin in the accident dose calculations does not impact the licensed rated thermal power of 3625.6 MWt (ML052840233 and ML052840235 for Vogtle, Units 1 and 2, respectively).

SNC has proposed a licensing basis change for its offsite and CR DBA dose consequence analysis for Vogtle, Units 1 and 2. The proposed change will implement the AST methodology provided in RG 1.183, Rev. 1, for determining DBAs offsite and CR doses. For full implementation of the AST DBAs analysis methodology, the dose acceptance criteria specified in 10 CFR 50.67 provides a voluntary alternative to the previous whole body and thyroid dose guidelines stated in 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance, and GDC 19.

As discussed in RG 1.183, Rev. 1, Regulatory Position 1.2.1, Full Implementation, states, in part, that:

Full implementation is a modification of the facility design basis that addresses all characteristics of the AST: specifically, the composition and magnitude of the radioactive material, its chemical and physical form, and the timing of its release.

Full implementation revises the plant licensing basis to specify the AST in place of the previous accident source term and establishes the TEDE as the new acceptance criterion.

This applies not only to the analyses performed in this LAR, but also to all future design basis dose consequence analyses at Vogtle, Units 1 and 2. At a minimum for full implementation of the AST, the DBA maximum hypothetical accident (MHA) loss-of-coolant accident (LOCA), as described in RG 1.183, Rev. 1, must be reanalyzed. Any change from an approved AST methodology to a different AST methodology that is not approved for use at Vogtle, Units 1 and 2, would require a LAR under 10 CFR 50.67.

SNC performed analyses for full implementation of the AST, as provided in RG 1.183, Rev. 1, and in accordance with the guidance in the RG, and Section 15.0.1 of the SRP. Also, SNCs AST analyses were based on the pressurized-water reactor (PWR) DBAs identified in RG 1.183, Rev. 1, that could potentially result in significant CR and offsite doses.

As stated in its letter dated May 12, 2025, the SNC LAR is a full-scope application of the AST methodology for full implementation. SNCs LAR seeks AST implementation for radiological consequences of major DBAs, specifically: MHA LOCA, fuel handling accident (FHA), main steam line break (MSLB), steam generator tube rupture (SGTR), control rod ejection (CRE), and locked rotor analysis (LRA). The licensee is not proposing physical changes to the plant.

RG 1.183, Rev. 1, Regulatory Position 1.3.2, Reanalysis Guidance, states, in part, that:

The NRC staff does not expect a complete recalculation of all facility radiological analyses but does expect licensees to evaluate all impacts of the proposed changes and to update the affected analyses and design bases appropriately.

The NRC considers an analysis to be affected if the proposed modification changes one or more assumptions or inputs used in the analysis, in such a way that the results, or the conclusions drawn from the results, are no longer valid.

The only changes of assumptions or inputs are for the DBAs contained in the LAR, and those analyses results have been submitted. All other existing analyses remain the same and, as mentioned above, there are no physical changes to the plant being proposed.

A full implementation of the AST is proposed for Vogtle, Units 1 and 2. Therefore, to support the licensing, and plant operation changes discussed in the LAR, SNC analyzed the following accidents employing the AST as described in RG 1.183, Rev. 1.

MHA LOCA FHA MSLB SGTR CRE LRA The DBA dose consequence analyses evaluated the integrated TEDE dose at the exclusion area boundary (EAB) for the worst 2-hour period following the onset of the accident. The integrated TEDE doses at the outer boundary of the low population zone (LPZ) during the entire period of the passage of the radioactive cloud resulting from postulated release of fission products, and the integrated dose to a Vogtle, Units 1 and 2, CR operator were evaluated for the duration of the accident. The LOCA, FHA, MSLB, SGTR, control rod ejection accident (CREA), and LRA dose consequence analyses was performed by SNC using the RADTRAD:

Simplified Model for RADionuclide Transport and Removal and Dose Estimation (RADTRAD)

Version 3.10. The development of the RADTRAD radiological consequence computer code was sponsored by the NRC, as described in NUREG/CR-6604 (ML15092A284 and was developed by Sandia National Laboratories for the NRC. The code estimates transport and removal of radionuclides and radiological consequence doses at selected receptors. The NRC staff performed independent confirmatory dose evaluations using the RADTRAD version 5.1.1 computer code.

Each DBA radiological source term used in the AST analyses was developed based on a core power level of 3,636 megawatts thermal (MWt) including measurement uncertainties. The use of 3,636 MWt for the AST DBA radiological source term analyses bounds the current licensed reactor core thermal power level of 3,625.6 MWt and is, therefore, acceptable to the NRC staff for use in implementation of the RG 1.183, Rev. 1, AST at Vogtle, Units 1 and 2.

Guidance in RG 1.183, Rev. 1, Regulatory Position 3.1, Fission Product Inventory, states, in part, that:

The core inventory should be determined using an appropriate computer code for calculating isotope generation and depletion. The core inventory factors (in curies per megawatt thermal) provided in TID-14844 and used in some analysis computer codes were derived for low-burnup, low-enrichment fuel and should not be used with higher burnup or higher enrichment fuels. The code should model the fuel geometries, material composition, and burnup, and the cross section libraries used should be applicable to the projected fuel burnup.

SNC developed the equilibrium core activity inventory using the Oak Ridge National Laboratory ORIGEN-ARP isotope generation and depletion computer code. The determination of core inventory is dependent upon full-power, core average conditions. The nominal inventory was based on an equilibrium cycle modeled with lead rod burnup of 68 gigawatt-days per metric ton of uranium (GWd/MTU). The NRC staff finds this approach acceptable.

SNC used committed effective dose equivalent and effective dose equivalent dose conversion factors (DCFs) from Federal Guidance Reports (FGR) 11 and 12 to determine the TEDE dose in accordance with AST evaluations. The use of ORIGEN-ARP to determine core inventory and utilizing the DCFs from FGR-11 and FGR-12 is an accepted practice in AST analysis and is part of their current licensing basis (CLB) and is, therefore, acceptable.

3.2 Meteorology and Atmospheric Dispersion This portion of the LAR relies on a provision in Section 5.3 of RG 1.183, Rev. 1, that allows the use of previously approved dispersion modeling results. The licensee indicates that both the offsite and onsite atmospheric dispersion factors (/Qs) referenced in the current LAR remain unchanged.

Consequently, the NRC staffs review of those offsite and onsite accident-related atmospheric dispersion modeling results and the related meteorology (Met) data input to those analyses were limited in scope. No confirmatory checks of that dispersion modeling or the corresponding Met data were made. Nevertheless, the NRC staff took into consideration a further statement in Section 5.3 of RG 1.183, Rev. 1, in its review that reads Licensees should ensure that any previously approved values remain accurate and do not include any misapplication of a methodology or calculational errors in the identified values. The sections below further address the scope and results of the staffs assessment.

3.2.1 Meteorological Data Section 3.1.1 of the LAR does not identify the periods of record (PORs) of Met data used to estimate the /Q values input to the offsite and onsite dispersion analyses. Rather, the text states The same meteorological data used to calculate the /Qs applied in the current licensing basis [CLB] radiological consequence analyses were used.

The licensee clarified in response to audit questions from NRC staff that the offsite /Qs at the EAB and LPZ are from the original licensing of Vogtle, Units 1 and 2. The POR of onsite Met data included in those dispersion calculations was a composite of measurements from December 4, 1972 through December 4, 1973; April 4, 1977 through April 4, 1979; and April 1, 1980 through March 31, 1981, as suggested by Section 2.3.3 of the UFSAR and reflected in UFSAR Table 2.3.4-1 (ML24297A648). The licensee also clarified that the wind measurements during that POR were obtained using mechanical wind instruments and represented scalar rather than vector hourly averages. The NRC staff notes that scalar averaging of wind data is appropriate for straight line Gaussian dispersion modeling such as in RG 1.145 as discussed further in Section 3.2.2 below.

The onsite /Qs referenced in the current LAR (ML25132A313) were obtained from an earlier approved LAR (ML22181B066) that followed guidance in Revision 0 of RG 1.183. This earlier LAR likewise does not identify the POR of Met data used in that onsite dispersion modeling analysis. However, the NRC staff found that Section 3.2.1 of the corresponding safety evaluation report (SER) (i.e., ML23158A018) stated that SNC provided the hourly onsite meteorological data from calendar years 1998 through 2000, that was used in the analysis. The licensee clarified that was the case in its audit responses.

For the onsite dispersion modeling analysis, the licensee clarified in responses to audit questions from NRC staff that the wind measurements during January 1, 1998 through December 31, 2000, POR were made using mechanical wind instruments and represented scalar hourly average values. The NRC staff finds use of scalar averages appropriate for straight line Gaussian dispersion modeling such as in RG 1.194 and as discussed further in Section 3.2.3 below.

Pursuant to Rev. 1 of RG 1.183, SNC provided NRC staff with documentation comparing 1998 to 2000 meteorological data with more recent observations made from 2015 to 2020. Based on the analysis provided by the licensee, the NRC staff agrees with SNCs conclusion that the meteorological observations used in the CLB remain representative of current dispersion conditions.

3.2.2 Offsite and Onsite Atmospheric Dispersion Factors The accident-related radiological dose analyses accompanying this LAR require atmospheric dispersion factors (i.e., relative concentrations or /Q values) as direct inputs and, therefore, are relevant to the NRC staffs determination of their acceptability. These /Q values are based on using appropriate regulatory guidance and dispersion models that rely, in part, on the input of representative Met data. The dispersion analyses for this LAR consider both offsite and onsite impacts. Offsite /Q values are estimated at the EAB and the outer boundary of the LPZ to evaluate potential impacts to the public. Onsite /Q values are estimated at the normal and/or emergency air intake locations to evaluate potential impacts to CR habitability.

3.2.3 Offsite /Qs at the EAB and LPZ This section addresses the NRC staffs limited scope evaluation of the atmospheric dispersion factors (or /Qs) at the offsite (i.e., EAB and outer boundary of the LPZ) for Vogtle, Units 1 and 2, as referenced in the current LAR.

The /Q values input to the offsite radiological dose consequence analyses for receptors on the EAB and the LPZ are consistent with the CLB in Revision 25 of the UFSAR for Vogtle, Units 1 and 2 (ML24297A648). The applicable results are listed in UFSAR Table 2.3.4-1 under the heading Based on 3 Years of VEGP Site Data, and correspond to the earliest 3 years of Met data indicated in Section 3.2.1 above. Further, SNC clarified for the NRC staff that the calculation of EAB and LPZ /Q values for the original licensing basis were based on hourly Met data rather than joint frequency distributions typically input to the PAVAN dispersion model, which had not yet been released at that time. Nevertheless, the assessment in the UFSAR was indicated to be consistent with the equations in RG 1.145 which are implemented by the PAVAN code.

The /Q values for the EAB and LPZ are shown in Table 3.1 of the current LAR. The NRC staff review found that most of the accident scenarios evaluated by the licensee using these /Q values are more than an order of magnitude below the applicable accident dose limits of either 25, 6.3, or 2.5 rem (see Section 3.9.4 of this SE for appropriate pairings). Of those few offsite accidents dose analysis types not so markedly below their corresponding limits, only one is within about 30 to 40 percent of its limit, while the others are about one-half and one-fourth of their respective limits (again, see Section 3.9.4 of this SE for appropriate pairings).

Given the above, as well as conservative assumptions inherent in the individual dose analyses themselves, it is reasonable to assume that the atmospheric dispersion conditions that contribute to the highest /Qs passed on to the dose calculations (e.g., windspeeds less than 3 to 4 meters per second and frequencies of stability classes F and G) would not have significantly changed since the POR used in the original licensing basis dispersion modeling.

Therefore, the NRC staff accepts SNCs continued use of the offsite /Q values at the EAB and LPZ in the CLB as referenced in this LAR.

3.2.3 Onsite /Qs at Intakes to the Control Room This section addresses the NRC staffs limited scope evaluation of the atmospheric dispersion factors at the onsite air intakes to the CR areas for Vogtle, Units 1 and 2, as referenced in the current LAR.

For the current LAR, SNC recomputed /Q values for potential accidental releases of radioactive material from Units 1 and 2 that could impact CR habitability. The licensee used the ARCON 2.0 dispersion model, a subsequent update to the ARCON96 code used in the previous LAR to implement Revision 0 of RG 1.183 (ML22181B066). Section 3.1.3 of the current LAR indicates that the updated model incorporates the same algorithms as in ARCON96. The NRC staff agrees with that observation.

The remodeled /Q values are summarized in current LAR Table 3.2 and represent a subset of the modeling results listed in Table 3.3 of the previous LAR. For brevity, Table 3.2 includes only those /Qs associated with a potential release from a given unit (i.e., Unit 1 or 2) that impacts the intake for that same unit because, in this case, the cross-unit /Qs were lower than the /Qs impacting the air intake belonging to the same unit. The NRC staff verified that to be the case by comparison.

Potential accident releases were evaluated from Containment (shown as U1 or U2 Reactor), the containment hatch door, refueling water storage tank (RWST), plant vent, and Fuel Handling Building, as well as accident releases due to an MSLB, SGTR, CRE, and LRA (which could be released from points designated in the tables as U1 North and U1 South, and from U2 North and U2 South) corresponding to secondary side releases from the main steam valve rooms.

The licensee indicates that most of the /Qs are the same as those in the onsite dispersion modeling analysis in the previous LAR. However, the licensee further explains in Section 3.1.3 of the current LAR that:

The /Qs for the [Unit 1 and 2] Plant Vent release point differ from the /Qs in the precedent LAR (ADAMS Accession No. ML22181B066) due to a correction in the upper wind speed measurement height input in the ARCON cases. The non-Safety-Related Plant Vent release is not credited in any DBA; it is provided for information only. No other release points were affected by the upper wind speed measurement change. The hatch door distances and directions were conservatively re-assessed but the hatch doors remain non-limiting release locations that are not used in any analysis.

The NRC staff acknowledges the licensees assessment but notes that the staff made no confirmatory modeling runs because the highest resultant dose among all potential accidents evaluated in the current LAR is only 60 percent of the corresponding 5 rem limit (i.e., for the MHA LOCA). All other accident scenarios are much lower relative to the 5 rem limit.

The NRC staff cross-checked other portions of the LAR and its attachments that reiterated the controlling (highest) onsite accident-related /Qs for consistency as well as those listed in Table 15A-2 of the UFSAR (ML24297A651) labeled as Reactor Building, Fuel Handling Building, and North MSIV [Main Steam Isolation Valve] Room.

Based on the above, as well as conservative assumptions inherent in the individual dose analyses themselves, the NRC staff concludes SNCs continued use of the onsite /Q values as referenced in the current LAR (ML25132A313) is acceptable.

3.2.4 Meteorology and Atmospheric Dispersion Conclusion The NRC staff finds that the accident-related offsite and onsite /Q values used as input to the respective dose calculations in the LAR dated May 12, 2025 (ML25132A313), are acceptable because they are consistent with a provision in Regulatory Position 5.3 of RG 1.183 that allows the use of previously approved dispersion modeling results. The NRC staff notes that the previous acceptability of the onsite /Qs was also due in part to the conservative approach used for their selection. That is, for a given accident scenario, the highest /Q value was chosen from among the Unit 1 or Unit 2 dispersion modeling results.

3.3 CR Habitability for all DBAs Vogtle CR is common to Units 1 and 2. The Control Room Emergency Filtration System (CREFS) is designed to maintain the CR envelope (CRE) at a positive pressure relative to the surrounding area, following postulated accidents. The CREFS is activated on a safety injection (SI) signal and/or high radiation in the normal outside air intakes. CREFS is designed to automatically isolate the CR and start one train of the emergency air filtration system upon a valid signal.

Also, initiation of the CR isolation (CRI) closes the isolation dampers between the normal and emergency systems. Two redundant and physically separated air handling unit trains with a moisture eliminator, an electric preheater, high-efficiency particulate air (HEPA) filters, and charcoal adsorbers are provided for each unit to process intake airflow and recirculated airflow in the combined CR. In emergency operation recirculation mode, the CR maintains a positive pressure of 1/8-inches water gauge (wg) relative to all adjacent areas.

During normal operation, CR unfiltered makeup flow rate is maintained less than or equal to 2575 cubic feet per minute (cfm) with the normal CR HVAC [heating, ventilation, and air conditioning] in service. On detection of a valid CRI signal, the normal outside air intakes for the CR are automatically isolated and the emergency operation/recirculation mode is initiated with 31,000 cfm filtered recirculation flow rate. The CLB for filtered makeup flow rate for all DBAs was 1,800 cfm. This application changes the filtered makeup flow rate to a lower value for 1,467 cfm for the FHA and SGTR, while maintaining the filtered makeup flow rate of 1,800 for all other DBAs. This change to the FHA and STGR accident analysis has a very slight impact on the calculated dose. This change in filtered makeup flow rate causes the calculated dose in the CR to increase slightly and is a more conservative value. This combination of makeup flow rate and recirculation flow help maintain a positive pressure in the CR. Air within the CR is recirculated continuously through the emergency air conditioning units, which contain upstream HEPA filters, charcoal adsorbers, downstream HEPA filters, cooling coil, and fan, to control the room temperature and airborne radioactivity. The outside air required for pressurization is mixed with the return air before it enters the filtration unit. The CR unfiltered in-leakage remains unchanged at 190 cfm, which contains 10 cfm for CR ingress and egress. The NRC staff performed sensitivity studies on the CR dose for the FHA and the SGTR in response to the proposed changes. The change in CR TEDE due to the change in the filtered makeup flow rate for the FHA and SGTR had minimal effect on the calculated dose and is, therefore, acceptable.

The CREFS is designed to maintain the CR envelope at a positive pressure relative to the surrounding area, following a valid CRI signal. The CR pressurization flow is routed through charcoal and HEPA filters. The CR pressurization charcoal filter efficiency for all iodine species is 99 percent and the HEPA filter efficiency for particulates is also 99 percent. The supply and recirculation fans use the same filters with 99 percent efficiency for all iodine species and particulates. The breathing rates and occupancy factors for the CR are in accordance with the values in Section C, Regulatory Position 4.2.6 of RG 1.183, Rev. 1.

The proposed LAR does not alter the CLB of the design or operation of the CR envelope or the CREFS, except for the filtered makeup flow rate for the FHA and SGTR which has been reviewed using confirmatory calculations and sensitivity studies and found to have negligible impacts on the final dose values. CR habitability is common to and applicable to all DBAs.

Inputs, assumptions, and initial conditions that are unique to DBAs analyzed in this LAR are discussed in their respective sections. Section 6.4, Habitability Systems, of the Vogtle, Units 1 and 2, UFSAR describes the CLB CR normal and emergency ventilation systems.

The NRC staffs review determined that SNC used analysis assumptions and inputs supporting CR habitability which are consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions, found acceptable to the NRC staff, are summarized within the LAR as Table 1: Control Room Parameters. The NRC staff finds that the inputs, analyses, and assumptions used in determining the hypothetical maximum exposed individual who is present in the CR are consistent with the NRC-approved methodology and are, therefore, acceptable.

3.4 Maximum Hypothetical Loss-of-Coolant Accident (MHA LOCA)

The MHA LOCA is that accident whose consequences, as measured by the radiation exposure of the surrounding public, would not be exceeded by any other accident whose occurrence during the lifetime of the facility would appear to be credible. The MHA LOCA, as used in this guide, refers to a loss of core cooling resulting in substantial meltdown of the core with subsequent release into containment of appreciable quantities of fission products. These evaluations assume containment integrity with offsite hazards evaluated based on design-basis containment leakage. The MHA LOCA, like all DBAs, is a conservative surrogate accident that is intended to challenge aspects of the facility design. Separate mechanistic analyses are performed using a spectrum of break sizes to evaluate fuel and emergency core cooling system (ECCS) performance for conformance with 10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. Section 15.6.5, Loss of Coolant Accidents, of the Vogtle, Units 1 and 2, UFSAR describes the CLB DBA.

The fission product release is assumed to occur in phases over a 4.5-hour period. When using RG 1.183, Rev. 1, for the evaluation of an MHA LOCA for a PWR, it is assumed that the initial fission product gap release to the containment will last for 30 seconds and will consist of the radioactive materials dissolved or suspended in the reactor coolant system (RCS) liquid. After 30 seconds, fuel damage is assumed to begin and is characterized by clad damage that releases the fission product inventory assumed to reside in the fuel gap. The fuel gap release phase is assumed to continue until 0.23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months after the initial breach of the RCS. As core damage continues, the gap release phase ends and the early in-vessel release phase begins.

The early in-vessel release phase continues until the 4.5-hour mark from the beginning of the event. SNC used the LOCA source term release fractions, timing characteristics, and radionuclide grouping as specified in RG 1.183, Rev. 1, for evaluation of the AST.

In the evaluation of the LOCA design basis radiological analysis, SNC considered dose contributions from the following potential activity release pathways:

Containment leakage directly to the atmosphere, Release from the containment mini-purge, Engineered-Safety-Feature (ESF) systems leakage outside of containment, and RWST (refueling water storage tank) leakage to the atmosphere.

3.4.1 Source Term SNC followed all aspects of the guidance outlined in RG 1.183, Rev. 1, Regulatory Position 3, Accident Source Term, regarding the fission product inventory, release fractions, timing of the release phases, radionuclide composition, and chemical form for the evaluation of the LOCA.

For the DBA LOCA, the licensee uses the core average inventory, as discussed above, and assumes that all the fuel assemblies in the core are affected. The LOCA analysis assumes that iodine will be removed from the containment atmosphere by both containment sprays and natural deposition to the containment walls. As a result of these removal mechanisms, a large fraction of the released activity will be deposited in the containment sump. The sump water will retain soluble gases and soluble fission products such as iodine and cesium, but not noble gases.

The licensee also updated its LOCA source term release characteristics to align with RG 1.183, Rev. 1, values provided in PWR Core Inventory Fraction Released into Containment Atmosphere (Table 2), MHA LOCA Release Phases (Table 5), and Radionuclide Groups (Table 6). The changes described in the amendment request, which align with the updated values in Rev. 1, were verified to be correctly used in the RADTRAD calculations.

The guidance from RG 1.183, Rev. 1, specifies that the iodine deposited in the sump water can be assumed to remain in solution as long as the containment sump pH is maintained at or above 7.0. SNC notes in Table A, Regulatory Section A-1.1, that the containment sump pH has been evaluated for the impact of the alternative source term and confirms that the sump pH remains greater than 7.0. In addition, Vogtle, Units 1 and 2, use trisodium phosphate to create a buffered sump solution that is resistant to change in pH.

The NRC staff reviewed the portion of the amendment dealing with the licensees analysis for maintaining suppression pool pH7 for 30 days following a LOCA. According to RG 1.183, Rev. 1, maintaining pH basic will minimize re-evolution of iodine from the suppression pool water. According to NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants, iodine released from the damaged core to the containment after a LOCA is composed of 95 percent cesium iodide which is a highly ionized salt, soluble in water. Iodine in this form does not present radiological problems since it remains dissolved in the sump water and does not enter the containment atmosphere. However, in the radiation field existing in the containment, some of this iodine could be transformed from the ionic to the elemental form, which is scarcely soluble in water and can be, therefore, released to the containment atmosphere. Conversion of iodine to the elemental form depends on several parameters, of which pH is very important. Maintaining pH basic in the sump water will ensure that this conversion will be minimized. Trisodium phosphate (TSP) is used as the buffering agent at Vogtle Units 1, and 2. The TSP is introduced into the containment sump fluid via stationary baskets following a postulated LOCA. After LOCA, several acids are either generated or are added to the containment. Relative amounts of these acids and that of TSP determine the value of pH reached by the containment sump water. The buffering action provided by the TSP is intended to maintain basic pH in the suppression pool despite the presence of strong acids.

The NRC staff verified that this LAR did not impact CLB containment sump pH analysis and determined that it is applicable to the AST. The existing analysis determines that the sump pH can be maintained 7.0 for 30 days. As such, no re-evolution of iodine from the sump water is assumed in the LOCA analysis.

3.4.2 Assumptions on Transport in the Primary Containment 3.4.2.1 Containment Mixing, Natural Deposition, and Leak Rate In accordance with RG 1.183, Rev. 1, SNC assumed that the activity released from the fuel is mixed instantaneously and homogeneously throughout the free air volume of the containment.

The licensee used the core release fractions and timing as specified in RG 1.183, Rev. 1, with the termination of the release into containment set at the end of the early in-vessel phase.

SNC credited the reduction of airborne radioactivity in the containment by natural deposition.

RG 1.183, Rev. 1, Appendix A, Regulatory Position A-2.2 states, in part, that:

Reduction in airborne radioactivity in the containment due to natural deposition within the containment may be credited. Section 6.5.2, Containment Spray as a Fission Product Cleanup System, of NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP) describes an acceptable model for removal of iodine and aerosols. The NRC staff no longer accepts the prior practice of deterministically assuming that a 50 percent plateout of iodine is released from the fuel, or the aerosol reductions (decontamination) calculated in NUREG/CR-6189, A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments, issued July 1996,

because the deterministic 50 percent plateout value and the aerosol reductions calculated in NUREG/CR-6189 are inconsistent with the characteristics of the revised source terms. However, the methods used in NUREG/CR-6189 may be credited on a case-by-case basis if they are adjusted to incorporate the revised MHA LOCA source term in this revision of Regulatory Guide (RG) 1.183. When these adjusted NUREG/CR-6189 methods are used, the DBA analyses should use the 10th-percentile values unless otherwise justified.

In its LAR dated May 12, 2025, Table B of Attachment 3 states SNCs conformance with RG 1.183, Rev. 1, Appendix A. Table B states that SNCs analysis for RG 1.183, Regulatory Position 2.2 is: Conforms-Aerosol natural deposition is credited based on adjustments to the 10th percentile correlations defined in NUREG/CR-6189 to account for the different source term characteristics defined in RG 1.183 Rev. 1.

SNC revised the natural deposition removal coefficients to align with Regulatory Position A-2.2 in RG 1.183, Rev.1. The licensee properly applied the 10th percentile values over the appropriate time intervals and utilized the most conservative time interval values where overlaps existed between RG 1.183, Rev. 1, and NUREG/CR-6189 phase time intervals. The NRC staff verified the accuracy of the final calculated values used in RADTRAD when applying the appropriate correlation as described in Table 36, Correlation of effective decontamination coefficients for radiological design basis accidents with reactor thermal power, from NUREG/CR-6189. The NRC staff did note a minor typographical error in Table 3.4-Modified NUREG/CR-6189 Correlations for RG 1.183, Rev. 1, of the application in the final row where the numerical value of 0.86 was used in the table. This had no effect on dose results, in that within the RADTRAD Natural Deposition Models, the use of the correct value from Table 36 of NUREG/CR-6189 was verified by staff independent review.

RG 1.183, Rev. 1, Regulatory Position A-2.7 states that the primary containment should be assumed to leak at the peak pressure technical specification (TS) leak rate for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and that for PWRs, the leak rate may be reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50 percent of the TS leak rate. Accordingly, the licensee assumed a containment leak rate of 0.20 percent per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , after which the containment leak rate is reduced to 0.10 percent per day for the duration of the accident. In the docketed calculation, the licensee included an additional 5 percent margin for conservatism (CLB calculations used containment leak rate of 0.21 percent per day for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , after which the containment leak rate is reduced to 0.105 percent per day for the duration of the accident.). In this LAR and associated calculations, the licensee used the TS values with no additional margin. The NRC staff finds the use of the CLB TS values without the additional margin acceptable. The licensee assumes the leakage is from both the sprayed and unsprayed regions of the containment to the environment.

3.4.2.2 Containment Spray Assumptions RG 1.183, Rev. 1, Appendix A, Regulatory Position A-2.3 states, in part, that:

The evaluation of the containment sprays should address areas within the primary containment that are not covered by the spray droplets. The mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building, provided that adequate flow exists between these regions, is assumed to be two turnovers of the unsprayed regions per hour, unless other rates are justified The containment building atmosphere may be considered a single well-mixed volume if the spray covers at least 90 percent of the containment building space and an engineered-safety-feature (ESF) ventilation system is available for adequate mixing of the unsprayed compartments.

For SNC, the volume of the sprayed region is 2.30x106 cubic feet (ft3) and the volume of the unsprayed region is 6.30x105 ft3. A flow rate of 21,000 ft3 per minute is used between the sprayed and unsprayed volume. This correlates to two turnovers of the unsprayed region volume per hour. In accordance with RG 1.183, Rev. 1, Appendix A, Section 2.3, SNC used the mixing rate attributed to natural convection between sprayed and unsprayed regions of the containment building of two turnovers of the unsprayed regions per hour. Since the sprayed region is less than 90 percent of the total containment volume, the licensee used a two-volume model to represent the sprayed and unsprayed regions of the containment. The above values for containment remain unchanged from the CLB. The CLB for containment sprays models the sprays terminating at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after initiation. This was to align with the guidance in RG 1.183, Rev. 0, which models the active release (early in-vessel phase) terminating at 1.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after event initiation. In this application the licensee utilizes containment sprays operating for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months to cover RG 1.183, Rev. 1, early in-vessel release duration terminating at 4.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The licensee states, There are no other design basis plant analyses, processes, or procedures that restrict the use of containment spray beyond two (2) hours or drive the operators to turn off sprays in the timeframe of interest once they have been actuated.

Using the guidance from SRP 6.5.2, Containment Spray as a Fission Product Cleanup System, and NUREG/CR-5966, A Simplified Model of Aerosol Removal by Containment Sprays (ML063480542), SNC determined the aerosol removal rate from the effects of the containment spray system is 5.34 per hour. This removal rate is consistent with their CLB and is acceptable.

Section 2.3 of Appendix A of RG 1.183, Rev.1, states, in part, that:

As provided in the SRP, the maximum DF [decontamination factor] for elemental iodine is based on the maximum iodine activity in the primary containment atmosphere when the sprays actuate, divided by the activity of iodine remaining at some time after decontamination. The SRP also states that the particulate iodine removal rate should be reduced by a factor of 10 when a DF

[decontamination factor] of 50 is reached. The reduction in the removal rate is not required if the removal rate is based on the calculated time-dependent airborne aerosol mass. There is no specified maximum DF for aerosol removal by sprays.

Using the guidance from SRP 6.5.2, SNC determined the elemental iodine removal rate from the effects of the containment spray system when in operation is 13.7 per hour. However, in accordance with the guidance in SRP 6.5.2, the licensee limited the removal rate constant for elemental iodine to zero when the elemental iodine DF reaches a value of 200. The aerosol removal coefficient is reduced by a factor of 10 when the aerosol DF reaches 50. SNC applied the removal rates in the radiological dose analysis from the time of spray actuation until 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after actuation. No credit is taken for organic iodine removal in the containment. These removal rates are consistent with the CLB.

The licensee appropriately models the DF limits and the spray duration in the RADTRAD files.

The staff verified the revised spray duration, and the appropriate DF limits as applied in the calculations. The single change of spray duration to the CLB with respect to containment sprays is acceptable in that there is no limitation that restricts the use of containment sprays beyond the CLB value of two hours.

The NRC staff reviewed SNCs application of credit for iodine removal from the operation of the containment spray system and found that the analysis follows the applicable regulatory guidance and is, therefore, acceptable.

3.4.3 Assumptions on ESF System Leakage Most of the initial conditions and assumptions associated with ESF leakage are consistent with the docketed calculation associated with the licensees adaptation of RG 1.183, Rev. 0, and/or the CLB.

The licensee has not changed the atmospheric dispersion coefficients (/Q) associated with the ESF leakage, but has rearranged the timing of the /Qs to follow the guidance in RG 1.183, Rev. 1, Section 5.3, which states, in part:

To ensure a conservative dose analysis, the period of the most adverse release of radioactive materials to the environment, with respect to doses, should be assumed to occur coincident with the period of most unfavorable atmospheric dispersion.

The NRC staff reviewed the amendment request and confirmed in the RADTRAD calculations that the licensee appropriately applied the previously approved atmospheric dispersion modeling and meteorology assumptions as provided in the Regulatory Position.

The application also modifies the ECCS leakage initiation time and the ESF recirculation start time from 30 minutes in the docketed calculation to 40 minutes to align with their CLB. The NRC staff conducted confirmatory calculations and sensitivity studies and determined this change has negligible effects on dose at the three receptor locations. This change is acceptable.

The changes to the contribution to doses due to ESF leakage in this application are made to either reflect the CLB or to align with the guidance in RG 1.183, Rev. 1, and are, therefore, acceptable.

3.4.5 CR Habitability There are no changes from the CLB to the CR habitability initial conditions, inputs, or assumptions proposed in this application.

3.4.5.1 Direct Shine Dose Evaluations There are no changes from the CLB to the CR direct shine dose evaluation initial conditions, inputs, or assumptions proposed in this application.

3.4.6 Conclusions SNC evaluated the radiological consequences resulting from the postulated LOCA and concluded that the radiological consequences at the EAB, LPZ, and CR are within the radiation dose reference values provided in 10 CFR 50.67 and the accident specific dose criteria specified in SRP Section 15.0.1. The NRC staff finds that SNC used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions associated with the MHA LOCA DBA accident dose consequence analysis are summarized within the LAR as Table 2: MHA LOCA Inputs and Assumptions, and SNCs calculated dose results are provided in Table 3.5: Calculated MHA LOCA Radiological Consequences. The NRC staff performed independent review of all inputs, assumptions and initial conditions in the SNC dose assessment files, and performed independent confirmatory calculations using RADTRAD Version 5.1.1, as necessary, to ensure a thorough understanding of the licensee's methods and to verify that values used in the dose assessment code were in line with values provided in the LAR dated May 12, 2025. The NRC staff concludes that the EAB, LPZ, and CR radiological doses for the MHA LOCA meet the applicable accident dose criteria, and provide reasonable assurance of adequate protection and are, therefore, acceptable 3.5 FHA The FHA involves the drop of a fuel assembly 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months after shutdown, onto another fuel assembly in the fuel building. The 60-hour decay time is less than, and, therefore, results in a higher source term and higher TEDEs at the receptor locations than the 70-hour decay time assumed in the CLB. In its submittal dated May 12, 2025, SNC states that the significant differences in the atmospheric dispersion factors of the fuel building and the containment demonstrate that the accident occurring in the fuel building is the bounding case with respect to radiological consequences in the three receptor areas (CR, EAB, LPZ). The radiological consequences of a FHA in containment are not limiting regardless of the containment configuration compared to a release from the fuel building. The release from the fuel building as the bounding event with respect to an FHA is consistent with the CLB.

The mechanical part of SNCs analysis remains unchanged from the CLB. It assumes that the total number of failed fuel rods is 314 due to the accident. This includes 100 percent of the 264 rods in the dropped fuel assembly and 50 rods in a second assembly which is struck in the drop for a total of 314 damaged fuel rods. A radial peaking factor of 1.7 is applied to the fission product inventory of the damaged fuel rods. The water above damaged fuel is not less than 23 feet and is controlled by TS 3.7.15 and TS 3.9.7. These values remain unchanged from the CLB. Following reactor shutdown, decay of short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. The proposed AST amendment takes credit for the normal decay of irradiated fuel. Section 15.7.4, Fuel Handling Accidents, of the Vogtle, Units 1 and 2, UFSAR describes the CLB DBA.

Appendix B, Assumptions for Evaluating the Radiological Consequences of a Fuel handling Accident, of RG 1.183, Rev.1, models the FHA gap activity release as occurring during two phases.

Phase 1 is the instantaneous release from the rising bubbles (from start of accident to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).

Elemental iodine and organic iodine are conservatively assumed to be in vapor form. Elemental iodine is subsequently decontaminated by passage through the overlying pool of water into the building atmosphere.

Phase 2 is the protracted release due to re-evolution as elemental iodine (starts at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and ends at 30 days). CsI [Cesium Iodide] is conservatively assumed to completely dissociate into the pool water. Because of the low pH of the pool water, CsI (as well as phase 1 absorbed elemental iodine within the pool) slowly re-evolves as elemental iodine into the building atmosphere.

3.5.1 Source Term The fission product inventory that constitutes the source term for this event is the gap activity in the fuel rods assumed to be damaged because of the postulated design basis FHA. Volatile constituents of the core fission product inventory migrate from the fuel pellets to the gap between the pellets and the fuel rod cladding during normal power operations. The fission product inventory in the fuel rod gap of the damaged fuel rods is assumed to be released in two phases over the duration of the accident (30 days).

Fission products released from the damaged fuel are decontaminated by passage through the overlaying water in the reactor cavity or spent fuel pool (SFP) depending on their physical and chemical form. Following the guidance in RG 1.183, Rev. 1, Appendix B, Regulatory Position B-1.3, the licensee assumes: (1) that the chemical form of radioiodine released from the fuel to the SFP consists of 95 percent cesium iodide (CsI), 4.85 percent elemental iodine, and 0.15 percent organic iodide; (2) in phase 1 of the accident progression (from time = 0 to time = 2 hr),

there is an instantaneous release of rising bubbles from the damaged assembly into the SFP in which elemental and organic iodine are assumed to be in vapor form; and (3) in phase 2 of the accident progression (time = 2 hr to time = 30 days), there is a protracted release into the fuel building and subsequently into the environment due to the re-evolution of the iodine species into elemental iodine. Due to the low pH of the pool water, the CsI re-evolves, and releases elemental iodine. This results in a final iodine distribution of 99.85 percent elemental iodine and 0.15 percent organic iodine.

SNCs analysis of the source term for an FHA is consistent with applicable Appendix B of RG 1.183, Rev. 1, which identifies acceptable radiological analysis assumptions for an FHA.

The NRC staff reviewed the licensees source term and identified the corresponding values in licensees calculations and finds SNCs analysis is acceptable.

3.5.1.1 Gap Release Fractions For adaptation of Rev.1 of RG 1.183, the LAR includes updates of the gap release fractions in the FHA analysis. This change from the CLB aligns with Table 4, PWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap, in RG 1.183, Rev. 1.

The NRC finds these assumptions acceptable because they are consistent with the NRC guidance.

Release Phases Phase 1 Release-Initial Gaseous Release and Water Depth The licensee maintains an SFP water level above damaged fuel as 23 feet, consistent with its CLB. In this release phase, the source term released to the fuel building and subsequently the environment is a function of the inventory of radionuclides in the fuel gap, and the elemental iodine DF which is calculated in B-2 of RG 1.183, Rev. 1. If the water depth is between 19 and 23 feet, the DF for elemental iodine can be computed based on a best-estimate rod pin pressure. In this case, the LAR and associated calculations use a pin pressure of 1400 psig.

This value is consistent with the bounding pin pressure used in the technical basis for this portion of the FHA calculation. The resulting DF of 437 for elemental iodine was confirmed in independent calculations using the SFP water level and the assumed pin pressure.

Phase 2 Release - Re-evolution Release Vpool = total pool free volume = 57,000 gal Spool = total pool surface area 200 m2 (bounding)

Qrecirc = volumetric flow of recirculation system (to evaluate effects of filtration) = 0 NI-131gap = fuel pin radioactive iodine in gap (moles) = 4.0E-03 NI-127gap = fuel pin nonradioactive iodine in gap (moles) = 7.3E-02 KL = mass transfer coefficient = 3.66x10-6 m/s pH = bounding design acidity value of the pool = 3.5 The final calculation phase of RG 1.183, Rev. 1, FHA analysis in contained in Step-4 of Appendix B. Utilizing above values, the licensee calculated a final time-dependent concentration of radionuclides released from the pool (Calculated release rate from SPF to Environment) as 1.93 cfm.

The NRC staff performed independent calculations of intermediary inputs to the RADTRAD calculations, such as the DF for elemental iodine and the time-dependent concentration of radionuclides released from the pool, and obtained results similar to the LAR. The NRC staff then used the updated RADTRAD version 5.1.1 which provides updates to the code to incorporate RG 1.183, Rev. 1, changes. The NRC staffs confirmatory calculations yielded results for the three dose receptor locations similar to the values provided in the LAR at the EAB, LPZ, and CR.

3.5.2 Transport 3.5.2.1 FHA in SFP area of Auxiliary Building Releases from an FHA in SFP area are modeled to release to the environment at the fuel building roof at the closet point to the CR intake to maximize the release to the CR. During normal operation, the system is in normal mode with two process radiation detectors monitoring the effluent. On alarm signal, the SFP air supply and exhaust dampers close and the ESF emergency filtration system is placed into service as the system is placed into emergency mode. In emergency mode, the fuel building ventilation automatically reconfigures and exhausts through ESF emergency filtration system charcoal and HEPA filters to remove halogens and particulates prior to discharging to the atmosphere via the plant vent. The licensee assumes a conservative 10-minute delay time to reach the CR emergency filtration system actuation setpoint. Although the ESF emergency filtration system will remove halogens and particulates, no credit is taken for filtration from the Fuel Handling Building (FHB) post-accident exhaust filters. Analysis of the FHA in the fuel building takes no credit for either filtration, or holdup in the fuel building. The FHB post-accident exhaust system is designed to maintain slightly negative pressure within the FHB following an FHA. This model is consistent with the CLB.

3.5.2.2 FHA Atmospheric Dispersion Values (/Q)

The only appreciable change to the /Q values for the FHA are utilization of CLB /Q values that cover the extended period of release in the updated FHA model in RG 1.183, Rev. 1. The release is extended to 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months which require utilization of /Qs not used in previous models to provide input to the calculations for this longer phase 2 release. These /Qs are provided in the CLB UFSAR in Table 15A-2, Limiting Short-Term Atmospheric Dispersion Factors for Accident Analysis for VEGP (s/m3). This change is acceptable in that it conforms with the revised FHA model.

3.5.3 CR Habitability for FHAs All inputs, assumptions, and initial conditions discussed in Section 3.3, CR Habitability for All DBAs, are applicable to the FHA DBA. SNC evaluated CR habitability for the CRE assuming the CREFS automatically transfers to the isolation and pressurization mode of operation upon SI or high radiation in the CR signal.

The initial conditions, inputs, and assumptions for CR habitability for an FHA are unchanged from the CLB with one exception. The licensee performed a sensitivity study on the CR filtered makeup flow rate into the CR during the DBA. The CLB value of 1800 cfm was reduced to 1467 cfm. The proposed value of 1467 cfm does result in slightly higher (more conservative) CR TEDE result. This negligible change in dose was repeated in confirmatory calculations performed by the NRC staff and is, therefore, acceptable.

3.5.4 Conclusion SNC evaluated the radiological consequences resulting from a postulated FHA and concluded that the radiological consequences at the EAB, LPZ, and CR are within the radiological dose guidelines provided in 10 CFR 50.67 and accident specific dose criteria specified in SRP Section 15.0.1. The NRC staff finds that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions associated with the FHA dose consequence analysis are summarized within the LAR as Table 3: FHA Inputs and Assumptions, and SNCs calculated dose results are provided in Table 3.7 - FHA Analysis Results. The NRC staff performed independent review of all inputs, assumptions and initial conditions in the SNC dose assessment files, and performed independent confirmatory calculations using RADTRAD version 5.1.1, as necessary, to ensure a thorough understanding of the licensees methods and to verify that values used in the dose assessment code were in line with values provided in the LAR dated May 12, 2025. The NRC staff finds that the EAB, LPZ, and CR radiological doses for the FHA meet the applicable accident dose criteria and provide reasonable assurance of adequate protection and are, therefore, acceptable.

3.6 MSLB Accident The postulated MSLB accident assumes a double-ended break of one main steam line outside the primary containment. This leads to an uncontrolled release of steam from the steam system.

The resultant depressurization of the steam system causes the main steam isolation valves to close and, if the plant is operating at power when the event is initiated, causes a reactor scram.

For the MSLB DBA radiological consequence analysis, a loss of offsite power (LOOP) occurs coincident with the reactor trip. Following a reactor trip and turbine trip, the radioactivity is released to the environment from the break point on the faulted steam generator (SG). Because the LOOP renders the main condenser unavailable, the plant is cooled down by releasing steam to the environment.

The radiological consequences of an MSLB outside containment bounds the consequences of a break inside containment. Therefore, only the MSLB outside of containment is considered regarding the radiological consequences. The affected SG, hereafter referred to as the faulted SG, rapidly depressurizes, and releases its initial contents to the environment. The MSLB accident is described in Vogtle, Units 1 and 2, UFSAR Section 15.1.5, Steam System Piping Failure (ML24297A648). RG 1.183, Rev. 1, Appendix F, identifies acceptable radiological analysis assumptions for a PWR MSLB.

As stated above, the steam release from a rupture of a main steam line would result in an initial increase in steam flow, which decreases during the accident as the steam pressure decreases.

The increased energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. If the most reactive rod cluster control assembly is stuck in its fully withdrawn position after the reactor trip, there is an increased possibility that the core will become critical and return to power. The core is ultimately shut down by the boric acid delivered by the SI (safety injection) system.

3.6.1 Source Term Appendix F of RG 1.183, Rev. 1, identifies acceptable radiological analysis assumptions for a PWR MSLB accident. RG 1.183, Appendix E, Regulatory Position E-2, states that:

If no or minimal fuel breach1 is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by the technical specifications (TS). Two cases of iodine spiking should be assumed:

1 Minimal fuel breach is defined for use in this appendix as an amount of damage that will yield reactor coolant system activity concentration levels less than the maximum technical specification limits. The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining [dose equivalent] DE I-131, only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

SNCs evaluation indicates that no fuel damage would occur resulting from a MSLB accident.

Therefore, the licensee considered the two radioiodine spiking cases described in RG 1.183, Rev. 1. The first case is referred to as a pre-accident iodine spike and assumes that a reactor transient has occurred prior to the postulated MSLB that has raised the primary coolant iodine concentration to the maximum value permitted by the TS for a spiking condition. For Vogtle, Units 1 and 2, the maximum iodine concentration allowed by TS 3.4.16 as the result of an iodine spike is 60 micro curies per gram (Ci/gm) of dose equivalent I-131 (DEI).

The second case assumes that the primary system transient associated with the MSLB causes an iodine spike in the primary system. This case is referred to as a concurrent iodine spike. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant increases to a value 500 times greater than the release rate corresponding to the iodine concentration at the equilibrium value specified in Vogtle, Units 1 and 2, TS. For Vogtle, Units 1 and 2, the RCS TS 3.4.16.2 limit for equilibrium or normal operation is 1.0 Ci/gm DEI. The duration of the concurrent iodine spike is assumed to be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months in accordance with RG 1.183, Rev. 1. No fuel damage is postulated with an MSLB. All of the above MSLB Source Term values are consistent with the CLB.

3.6.2 Release Transport SNC followed the guidance as described in RG 1.183, Rev. 1, Appendix F, Regulatory Position F-5 in all aspects of the transport analysis for the MSLB. For additional conservatism, the licensee assumes a total primary-to-secondary leak rate equal to 1 gpm (1,440 gallons per day (gpd)), which is higher than the TS 3.4.13 d. limits primary-to-secondary leakage to 150 gpd through any one SG, which is a total of 600 gpd from all four SGs. SNC modeled the assumed primary-to-secondary leakage of 0.35 gpm into the faulted SG and 0.65 gpm total into the remaining three intact SGs. These are the same values as the CLB.

RG 1.183, Rev. 1, Appendix F, Regulatory Position F-6.4, states:

The primary-to-secondary leakage should be assumed to continue until the primary system pressure is less than the secondary system pressure, or until the temperature in the bulk of the primary system is less than 100 degrees Celsius (212 degrees Fahrenheit). The primary-to-secondary leak rate at later stages of the transient may be reduced if justified by plant-specific design and engineering analyses. The release of radioactivity from unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.

In accordance with RG 1.183, Rev. 1, SNC assumes that primary-to-secondary leakage from the intact SGs continues until the RCS reaches 212°F, which is 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after the MSLB, at which time shutdown cooling is initiated. Primary-to-secondary leakage in the faulted SG also ceases when residual heat removal (RHR) is placed in service 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after initiation of the event. This is consistent with the CLB. In accordance with RG 1.183, Rev. 1, the licensee assumes that all noble gas radionuclides released from the primary system are released to the environment without reduction or mitigation. Following the guidance from RG 1.183, Rev. 1, Appendix E, SNC assumes that all the primary-to-secondary leakage into the faulted SG will flow directly from the RCS to the environment with no partitioning. For the unaffected SGs that are used for plant cooldown, the licensee assumes a partition factor of 100 is applied to the iodine nuclides. The iodine releases to the environment from the unaffected SGs are assumed to be 97 percent elemental, and 3 percent organic, which is consistent with Regulatory Position F-5 in RG 1.183, Rev. 1, Appendix F.

Primary-to-Secondary leakage (consistent with CLB) is assumed to be 0.35 gpm to the faulted SG, and 0.65 gpm (total) going to the intact SGs. The postulated MSLB event causes the associated faulted SG to blow dry, releasing activity directly to the environment through the broken main steam line. Activity from three intact SGs released to the environment via steaming until the RCS is placed on RHR cooling (assumed at 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ).

The single change to MSLB accident analysis release transport is a change in the intact steam generator release. The increase in the periodic and total pounds mass (lbm) release by the intact SGs was increased, which would result in a more conservative result (increased dose at receptor locations). The NRC staff finds this change acceptable. The remainder of the assumptions are consistent with the CLB 3.6.3 CR Habitability All inputs, assumptions, and initial conditions discussed in Section 3.3, Control Room Habitability for all DBAs, above are applicable to the MSLB DBA. SNC evaluated CR habitability for the CR envelope assuming that the CREFS automatically transfers to the isolation and pressurization mode of operation upon SI or high radiation in the CR signal. The licensee stated the time for CRI is 11.3 seconds after an MSLB and subsequent SI signal, and CR pressurization mode initiation occurs at 99.3 seconds, 88 seconds after a valid CR isolation signal. The initial conditions, inputs, and assumptions for CR habitability for the MSLB accident have not changed from the CLB and are, therefore, acceptable.

3.6.4 Conclusion SNC evaluated the radiological consequences resulting from the postulated MSLB in both the pre-accident and the concurrent iodine spike cases and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67 and accident specific dose criteria specified in SRP Section 15.0.1. The NRC staff finds that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions associated with the MSLB dose consequence analysis are summarized within the LAR as Table 4: MSLB Accident Inputs and Assumptions, and SNCs calculated dose results are provided in Table 3.8 -

Calculated MSLB Accident Radiological Consequences. The NRC staff performed independent review of all inputs, assumptions and initial conditions in the SNC dose assessment files, and performed independent confirmatory calculations using RADTRAD version 5.1.1, as necessary, to ensure a thorough understanding of the licensee's methods and to verify that values used in the dose assessment code were in line with values provided in the LAR dated May 12, 2025.

The NRC staff finds that the EAB, LPZ, and CR radiological doses for the MSLB in both the pre-accident and the concurrent iodine spike cases meet the applicable accident dose criteria and provide reasonable assurance of adequate protection and are, therefore, acceptable.

3.7 SGTR Accident The SGTR accident assumes an instantaneous and complete severance of a single SG tube.

The postulated break allows RCS to leak to the secondary side of the ruptured SG. The radioactivity from the leaking SG tube mixes with the shell-side water in the affected SG. The release location from the faulted SG is assumed conservatively to originate from the main steamline room closest to the CR. The SGTR assumes a concurrent LOOP to maximize the release to the environment. Section 15.6.3, Steam Generator Tube Failure, of the Vogtle, Units 1 and 2, UFSAR describes the CLB DBA.

After the initiation of the accident, the faulted SG is manually isolated in 20 minutes. This is accomplished by isolating steam flow from and stopping feedwater flow to the faulted SG. With the isolation of the faulted SG (FSG), the associated atmospheric relief valve (ARV) fails open.

With the ARV failed open, release RCS to the environment via the faulted SG continues until isolation of the ARV via block valve closure occurs. Manual operator action is assumed to locally close the block valve associated with the failed open ARV at 30 minutes after the FSG is isolated. ARV isolation occurs 50 minutes after initiation of the event. For the SGTR DBA, radiological consequence analysis, mass transfer from the primary to the secondary in the faulted SG continues until the break flow is terminated. Break flow from the primary side of the SG to the secondary side is terminated in 102 minutes. The time for the break flow termination via manual operation of the block valve has changed from the CLB value of 92 minutes to 102 minutes in this application. The time for atmospheric release from faulted SG ended has changed from the CLB value of 36 minutes to 50 minutes in this application. These are conservative changes which result in higher calculated dose. The NRC staff verified that the change is properly accounted for in the dose analysis and is, therefore, acceptable.

Leakage into the intact SGs continues with activity released to the environment through steaming until the RCS is cooled to Cold Shutdown conditions (200°F) and placed on RHR after 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . The only change from CLB assumptions in the SGTR analysis are those associated with the ARV closure time and SG steam release rates. All the changed values are conservative resulting in higher calculated doses at the three receptor locations.

3.7.1 Source Term Appendix E of RG 1.183, Rev. 1, identifies acceptable radiological analysis assumptions for a SGTR accident. Appendix E, Regulatory Position E-2, states:

If no or minimal fuel breach1 is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by technical specifications (TS). Two cases of iodine spiking should be assumed:

A reactor transient has occurred before the postulated SGTR and has raised the primary coolant iodine concentration to the maximum value permitted at full-power operations by the TS (typically 60 microcuries per gram (Ci/g) dose equivalent (DE) iodine (I)-131). This is the pre-accident iodine spike case.

The primary system transient associated with the SGTR causes an iodine spike in the primary system. The increase in primary coolant iodine concentration is estimated using a spiking model that assumes that the iodine release rate from the fuel rods to the primary coolant (expressed in curies per unit time) increases to a value 335 times greater than the release rate corresponding to the iodine concentration at the equilibrium value specified in the TS (typically 1.0 Ci/g DE I-131). This is the concurrent iodine spike case. A concurrent iodine spike need not be considered if fuel damage is postulated. The assumed iodine spike duration should be 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Shorter spike durations may be considered on a case-by-case basis if it can be shown that the activity released by the 8-hour spike exceeds that available for release from the fuel pins assumed to have defects.

1 Minimal fuel breach is defined for use in this appendix as an amount of damage that will yield reactor coolant system activity concentration levels less than the maximum technical specification limits. The activity assumed in the analysis should be based on the activity associated with the projected fuel damage or the maximum technical specification values, whichever maximizes the radiological consequences. In determining DE I-131, only the radioiodine associated with normal operations or iodine spikes should be included. Activity from projected fuel damage should not be included.

SNCs evaluation indicates that no fuel damage would occur resulting from a SGTR accident.

Therefore, consistent with RG 1.183, Rev. 1, the licensee performed the SGTR accident analyses for two radioiodine spiking cases. The first case is referred to as a pre-accident iodine spike and assumes that a reactor transient has occurred prior to the postulated SGTR that has raised the primary coolant iodine concentration to the maximum value permitted by the TS for a spiking condition. For Vogtle, Units 1 and 2, the maximum iodine concentration allowed by TS Section 3.4.16, resulting from an iodine spike, is 60 Ci/gm DEI.

The second case assumes that the primary system transient associated with the SGTR causes an iodine spike in the primary system. This case is referred to as a concurrent iodine spike.

Initially, the plant is assumed to be operating with the RCS iodine activity at the TS limit for normal operation. For Vogtle, Units 1 and 2, the TS Section 3.4.16 limit for normal operation is 1.0 Ci/gm DEI. The increase in primary coolant iodine concentration for the concurrent iodine spike case is estimated using a spiking model that assumes that as a result of the accident, iodine is released from the fuel rods to the primary coolant at a rate that is 335 times greater than the iodine equilibrium release rate. The concurrent iodine spike duration is assumed to be a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . SNC assumes that the activity released from the iodine spiking mixes instantaneously and homogeneously throughout the RCS. The CLB value of iodine release rate of 500 times greater than the release rate corresponding to the iodine concentration at equilibrium, was revised to align with RG 1.183, Rev. 1, Regulatory Position E-2.2 of a value of 335 times greater than the release rate corresponding to the iodine concentration at equilibrium.

In accordance with RG 1.183, Rev. 1., Appendix E, Regulatory Position E-5, SNC assumes the speciation for iodine release from the SGs is 97 percent elemental and 3 percent organic. In addition, SNC included the radiological dose contribution from the release of secondary coolant iodine activity at the limit of 0.1 Ci/gm DEI in TS Section 3.7.16. In both the pre-accident iodine spike and the concurrent spike, the RCS activity includes equilibrium noble gas concentration of 280 Ci/g DE Xe-133 and assumption of 1 percent of the fuel rods having cladding defects.

Although a LOOP is assumed, the licensee modeled continued feedwater system flows into the SG, until failed SG isolation, as a source of radioiodine in this analysis for conservatism. All the initial conditions, inputs, and assumptions associated with the source term during an SGTR are consistent with the CLB and are, therefore, acceptable.

3.7.2 Release Transport SNC followed the guidance as described in RG 1.183, Rev. 1, Appendix E, in all aspects of the transport analysis for the SGTR. For additional conservatism, SNC assumes a total primary-to-secondary leak rate equal to 1 gpm (1,440 gpd), which is higher than the TS 3.4.13 d. primary-to-secondary leakage limit of 150 gpd through any one SG. The licensee modeled the assumed primary-to-secondary leakage of 0.3 gpm into the faulted SG and 0.7 gpm into the three intact SGs.

RG 1.183, Rev. 1, Appendix E, Regulatory Position E-6.2, states:

The density used in converting volumetric leak rates (e.g., in gallons per minute) to mass leak rates (e.g., in pounds mass per hour) should be consistent with the basis of surveillance tests used to show compliance with leak rates in the TS.

These tests are typically based on cool liquids. Facility instrumentation used to determine leakage is typically located on lines containing cool liquids. In most cases, the density should be assumed to be 1.0 gram per cubic centimeter (62.4 pounds mass per cubic foot).

SNC assumes that the release of radioactivity from the faulted SG continues for 50 minutes at which time the failed open ARV for the faulted SG is manually blocked, and the unaffected SGs release continues for 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months at which time shutdown cooling is initiated, and steam releases from the SGs have been terminated. At this point in the accident sequence, steaming is no longer required for cooingl down and releases from the intact SGs are terminated. The application included changes from the CLB to the intact steam generator steam release rate, the ruptured steam generator steam release rate, and the ruptured tube break flow rate. Sensitivity studies performed by the NRC staff conclude that these changes have negligible effect on the final TEDE values at the three receptor locations and are therefore acceptable. The analysis models an exposure duration of 30 days for the LPZ and occupants of the CR, and the worst 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the EAB.

SNC assumes that the source term resulting from the radionuclides in the RCS, including the contribution from iodine spiking, is transported to the ruptured SG by the break flow. A portion of the break flow is assumed to flash to steam based upon the thermodynamic conditions in the RCS relative to the secondary system. The licensee assumes that the flashed portion of the break flow will ascend through the bulk water in the SG, enter the steam space of the affected SG, and be immediately available for release to the environment with no credit taken for scrubbing. Although RG 1.183, Rev. 1, allows the use of the methodologies described in NUREG-0409, Iodine Behavior in a PWR Cooling System Following a Postulated Steam Generator Tube Rupture Accident, January 1978 (ML19269F014), to determine the amount of scrubbing credit applied to the flashed portion of the break flow, SNC did not credit scrubbing of the activity in the flashed break flow in the ruptured SG.

In accordance with RG 1.183, Rev. 1, the licensee assumes that all noble gas radionuclides released from the RCS are released through the SGs to the environment without reduction or mitigation. In the ruptured SG, SNC assumes the iodine in the flashed portion of the break flow is immediately available for release without reduction or mitigation. The break and leakage flow that does not flash mixes uniformly with the SG liquid mass and activity is released to the environment in direct proportion to the steaming rate and the partition coefficient, in accordance with RG 1.183, Rev. 1, Appendix E, Regulatory Position E-6.5.4 a SG partition coefficient for iodine nuclides of 100 is assumed.

3.7.3 CR Habitability The initial conditions, inputs, and assumptions for CR habitability for an SGTR accident are unchanged from the CLB.

3.7.4 Conclusion SNC evaluated the radiological consequences resulting from the postulated SGTR in both the pre-accident and the concurrent iodine spike cases and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67 and accident specific dose criteria specified in SRP Section 15.0.1. The NRC staff finds that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions associated with the SGTR dose consequence analysis are summarized within the LAR as Table 5: SGTR Accident Inputs and Assumptions, and SNCs calculated dose results are provided in Table 3.9 - Calculated SGTR Accident Radiological Consequences. The NRC staff performed independent review of all inputs, assumptions and initial conditions in the SNC dose assessment files, and performed independent confirmatory calculations using RADTRAD version 5.1.1, as necessary, to ensure a thorough understanding of the licensees methods and to verify that values used in the dose assessment code were in line with values provided in the LAR dated May 12, 2025. The NRC staff finds that the EAB, LPZ, and CR radiological doses for the SGTR in both the pre-accident and the concurrent iodine spike cases meet the applicable accident dose criteria and provide reasonable assurance of adequate protection and are, therefore, acceptable.

3.8 CREA Vogtle, Units 1 and 2, UFSAR Section 15.4.8, Spectrum of Rod Cluster Control Assembly Ejection Accidents (ML24297A648), describes the CREA as the mechanical failure of a control rod drive mechanism (CRDM) pressure housing resulting in the ejection of a rod cluster control assembly and drive shaft. The consequence of this mechanical failure is a rapid positive reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.

3.8.1 Source Term Following the applicable guidance, SNC evaluated two separate release scenarios for the CREA. In the first case, the failed fuel resulting from the CREA is released in its entirety into the containment via the ruptured CRDM housing, is mixed in the free volume of the containment, and then released to the environment at the containment TS leak rate, plus 5 percent for conservatism, for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at half that value for the remaining 29 days. For the second case, the radiological consequence from a CREA is evaluated assuming that the RCS boundary remains intact and that fission products are released to the environment from the secondary system. In this case, fission products from the damaged fuel are assumed to be released to the RCS and transported to the secondary system through primary-to-secondary leakage in the SGs. RG 1.183, Rev. 1, Appendix H, Regulatory Position H-1, states, in part:

Regulatory Position 3 of this guide provides assumptions acceptable to the NRC staff regarding core inventory. The fission product release from the breached fuel to the coolant is based on Regulatory Position 3.2 of this guide and the estimate of the number of fuel rods breached. In addition to the combined fission product inventory (steady-state gap plus transient release), the release attributed to fuel melting is based on the fraction of the fuel that reaches or exceeds the initiation temperature for fuel melting, and on the assumption that 100 percent of the noble gases and 50 percent of the iodines contained in that fraction are released to the reactor coolant.

In accordance with RG 1.183, Rev. 1, Appendix H, Regulatory Position H-3, 100 percent of the released activity is assumed to be released instantaneously and mixed homogeneously throughout the containment atmosphere for the ruptured CRDM housing, and 100 percent of the released activity is assumed to be released instantaneously and completely dissolved in the RCS and available for release to the secondary containment in the secondary side release scenario. SNC assumed that 100 percent of the noble gases and 25 percent of the iodine isotopes within the melting rods are available for release from the containment pathway and 100 percent of the noble gases and 50 percent of the iodine isotopes within the melting rods are available for release from the RCS through the secondary system pathway. In addition, the licensee included the radiological dose contribution from the release of secondary coolant iodine activity at the TS 3.7.16 limit of 0.1 Ci/gm DEI. The alkali metals in the secondary coolant are assumed to be 10 percent of those in the RCS corresponding to 1 percent failed fuel. The value of 10 percent of the alkali metals in the secondary coolant is based upon the ratio of DEI in the RCS, 1.0 Ci/gm DEI, vs the secondary concentration of 0.1 Ci/gm DEI. This source term section has no changes from the CLB.

3.8.1.1 Gap Fractions for CREA For adaptation of RG 1.183, Rev. 1, the LAR includes updates of the gap release fractions in the CREA analysis. As this change from the CLB aligns with Table 4, PWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap, in Rev. 1. The NRC finds these assumptions acceptable because they are consistent with the NRC guidance.

3.8.2 Transport from Containment SNC assumes that the activity released to the containment through the rupture in the reactor vessel head mixes instantaneously throughout the containment with no credit assumed for removal of elemental iodine or noble gas in the containment due to containment sprays or for natural deposition of elemental iodine. The licensee is taking credit for natural deposition of aerosols in containment and a removal rate of 3.012x10-2 per hour. SNC assumes that all containment leakage is 0.20 percent per day, which equals the CLB TS limit, for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and half of that value, 0.10 percent per day, thereafter. The licensee assumes that the iodine released to the containment from the fuel consists of 95 percent particulate, 4.85 percent elemental, and 0.15 percent organic per RG 1.183, Rev. 1, Appendix H, Regulatory Position H-4. No credit is taken for iodine or particulate removal by containment sprays.

The only change to the CLB in CREA Containment Transport is the change in containment leak rate. The CLB assumes 0.201 leakage from time 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 0.105 percent from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days. The assumptions used in this LAR align directly with the TS leak rate of 0.20 percent leakage from time 0 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and 0.10 percent from 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 30 days. This change is less conservative but is acceptable due to the fact these values align directly with the licensees TSs and are consistent with the NRC-approved analytical methodology. All other assumptions are consistent with the CLB and all above containment transport assumptions are acceptable.

3.8.3 Transport from Secondary System In accordance with RG 1.183, Rev. 1, Appendix H, Regulatory Position H-7, SNC evaluated the transport of activity from the RCS to the SGs secondary side assuming a total primary-to-secondary leak rate equal to 1 gpm (1440 gallons per day (gpd) from all four SGs, which is higher than the TS 3.4.13 d. total allowable leak rate of 150 gpd through any one SG (600 gpd for all SGs), to account for any accident induced leakage.

In accordance with RG 1.183, Rev. 1, SNC assumes that all noble gas radionuclides released from the RCS are released to the environment without reduction or mitigation. Following the guidance from RG 1.183, Rev. 1, Appendix E, Regulatory Position E-6.5 the licensee models that all of the primary-to-secondary leakage in the SGs mix with secondary water without flashing. For iodine, because the SG tubes remain covered for the duration of the CREA, the partition coefficient of 100 was utilized in accordance with RG 1.183, Rev. 1. The retention of particulate radionuclides in the SG is limited by the moisture carryover from the SG. The licensee modeled the transport of particulates and iodine using a maximum moisture carryover value, which is 0.32 percent in its submittal. SNC assumed for the secondary side release that the chemical form of iodine released from the SGs to the environment is 97 percent elemental and 3 percent organic. These Secondary System Transport assumptions are consistent with the CLB.

The only change to the Secondary System Transport assumptions is in the volumetric release of secondary steam. The new values captured in Table 6: Control Rod Ejection Accident Inputs and Assumptions. The LAR provide additional margin to the calculations, and are, therefore, acceptable.

3.8.4 CR Habitability for Control Rod Ejection Accident (CREA)

All inputs, assumptions, and initial conditions discussed above in Section 3.3 are applicable to the CREA DBA. There is no change from the CLB to the CR habitability for the CREA in this application.

3.8.5 Control Rod Ejection Accident (CREA) Conclusion SNC evaluated the radiological consequences resulting from the postulated CREA as prescribed in RG 1.183, Rev. 1, Appendix H, which has the licensee consider two distinct release paths, and concluded that the radiological consequences at the EAB, LPZ, and CR are within the dose guidelines provided in 10 CFR 50.67 and accident specific dose criteria specified in SRP Section 15.0.1. The NRC staff finds that SNC used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions associated with the CREA dose consequence analysis are summarized within the LAR as Table 6: Control Rod Ejection Accident Inputs and Assumptions, and SNCs calculated dose results are provided in Table 3.10 - Calculated Control Rod Ejection Accident Radiological Consequences. The NRC staff performed independent review of all inputs, assumptions and initial conditions in the SNC dose assessment files, and performed independent confirmatory calculations using RADTRAD version 5.1.1, as necessary, to ensure a thorough understanding of the licensee's methods and to verify that values used in the dose assessment code were in line with values provided in the LAR dated May 12, 2025. The NRC staff finds that the EAB, LPZ, and CR radiological doses for the CREA from both the containment leakage pathway and the release via secondary system pathway meet the applicable accident dose criteria and provide reasonable assurance of adequate protection and are, therefore, acceptable.

3.9 LRA The LRA considers the instantaneous seizure of a reactor coolant pump (RCP) rotor, which causes a rapid reduction in the flow through the affected RCS loop. The sudden decrease in core coolant flow causes a reactor trip. SNCs evaluation indicates that fuel cladding damage will occur because of this accident. Activity from the fuel cladding damage is transported to the secondary side due to primary-to-secondary side leakage. Radioactivity is released to the outside atmosphere from the secondary coolant system via steaming until cold shutdown conditions are established in the RCS. Following reactor trip and based on a coincident assumption of LOOP, the condenser is unavailable, and reactor cooldown is achieved using steam releases from the SGs until initiation of shutdown cooling. For conservatism, the licensee assumes a total primary-to-secondary leak rate equal to 1 gpm from all four SGs, which is higher than the TS total allowable leak rate of 150 gpd through any one SG. Section 15.3.3, Reactor Coolant Pump Shaft Seizure (Locked Rotor), of the Vogtle, Units 1 and 2, UFSAR describes the CLB DBA.

3.9.1 Source Term SNC assumed that the instantaneous seizure of the RCP rotor associated with the LRA results in a small percentage of fuel clad damage. As in the Vogtle, Units 1 and 2, CLB, radiological dose analysis for this event assumes 5 percent fuel clad damage with no fuel melt predicted.

Therefore, the source term available for release is associated with this fraction of damaged fuel cladding and the fraction of core activity existing in the gap. A radial peaking factor of 1.7 was applied to the fission product inventory of the damaged rods. The activity released from the fuel is assumed to be released instantaneously and homogeneously through the RCS.

Following the guidance in RG 1.183, Rev. 1, Appendix G, Regulatory Position G-4, SNC assumes that the chemical form of radioiodine released from the fuel to the reactor coolant consists of 95 percent CsI, 4.85 percent elemental iodine, and 0.15 percent organic iodide, and that the iodine releases from the SGs to the environment is 97 percent elemental iodine and 3 percent organic iodine.

In addition, the licensee included the radiological dose contribution from the release of secondary coolant iodine activity at the TS 3.7.16 limit of 0.1 Ci/gm DEI. The alkali metals in the secondary coolant are assumed to be 10 percent of those in the RCS, based on the ratio of the secondary to RCS DE I-131 concentrations, corresponding to 1 percent failed fuel. SNC assumes a LOOP to maximize the release to the environment; however, continued feedwater flow is modeled into the SG as a source of radioiodine in this analysis for conservatism. This source term has not changed from the CLB.

3.9.1.2 Gap Release Fractions For adaptation of Rev.1 of RG 1.183, the LAR includes updates of the gap release fractions in the LRA analysis. As this change from the CLB aligns with Table 4. PWR Steady-State Fission Product Release Fractions Residing in the Fuel Rod Plenum and Gap, in Rev. 1, these assumptions are acceptable because they are consistent with the NRC guidance.

3.9.2 Release Transport The activity that originates in the RCS is released to the secondary coolant via the primary-to-secondary coolant leakage. SNC assumes a design basis leak rate of 1 gpm from all four SGs.

This equates to a total of 1,440 gpd, which is greater than the maximum allowable operational leakage of 150 gpd for any one SG imposed in TS 3.4.13 d. A LOOP is assumed to occur concurrently with the reactor trip, which results in releases to the environment associated with the secondary coolant steaming from the SGs.

Because of the release dynamic of the activity from the SGs, RG 1.183, Rev. 1, allows for a reduction in the amount of activity released to the environment based on partitioning of nuclides between the liquid and gas states of water for this release path. For iodine, because the SG tubes remain covered for the duration of the LRA, the partition coefficient of 100 was taken directly from the suggested guidance. The retention of particulate radionuclides in the SG is limited by the moisture carryover from the SG which is 0.32 percent at Vogtle, Units 1 and 2.

Because of its volatility, 100 percent of the noble gases are assumed to be released directly to the environment. All remaining isotopes are transported to the SGs at the rate of 1 gpm. The release continues for 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months by which time the RCS is placed on RHR system cooling.

In its application, the licensee increased the volumetric flow rate of the secondary side release.

This change increases the total amount of radionuclides and adds margin to the accident analysis. This is a conservative change to the accident analysis and is, therefore, acceptable.

Transient Fission Gas Release Transient Fission Gas Release (TFGR) is defined as the additional fission product release that may occur during a transient from pellet fracturing and grain boundary separation. Release fraction Tables 2 and 4 of RG 1.183, Rev. 1, provide the release fractions for MHA LOCA and non-LOCA respectively. The analysis of the Vogtle, Units 1 and 2, gap release fractions is within the enrichment and rod burnup limitations of RG 1.183, Rev. 1. Five reactivity insertion events were discussed including CREA, FHA, LRA, MSLB, and SGTR.

TFGR is not considered for MSLB or SGTR as the Vogtle, Units 1 and 2, design and licensing basis does not predict fuel failures for these events. TFGR is also not considered for the FHA, as it is a low-temperature event. For the CREA, the TFGR was calculated using the guidance in RG 1.183, Rev. 1, Section 3.2 (i.e., the TFGR equations in Equation 1 and 2 of RG 1.183 Rev. 1)..

For an LRA, to determine a bounding radiological consequence, the licensee assumed that the cladding fails for all rods which experience departure from nucleate boiling. The Vogtle, Units 1 and 2, licensing basis assumes that 5 percent of the fuel rods experience cladding failure due to an LRA. The calculated TFGR was based on initial core inventory, the steady-state gap fractions, the failed fuel fraction (5 percent), and the radial peaking factor (RPF). The Vogtle, Units 1 and 2, licensing basis assumes that all failed fuel was operating with an RPF of 1.7, which is bounding of the Core Operating Limit Reports.

3.9.3 CR Habitability All inputs, assumptions, and initial conditions discussed above in Section 3.3 of this SE are applicable to the LRA DBA. SNC evaluated CR habitability for the CRE assuming the CREFS automatically transfers to the isolation and pressurization mode of operation upon SI or high radiation in the CR signal. The licensee determined that CRI would occur at 608 seconds after an LRA. This would be 600 seconds to generate the isolation signal, and 8 seconds for CR isolation. CR pressurization would be achieved at 698 seconds. These values have not changed from the CLB 3.9.4 Conclusion SNC evaluated the radiological consequences resulting from the postulated LRA utilizing Rev.1 of RG 1.183 and concluded that the radiological consequences at the EAB, LPZ, and CR are within the radiation dose reference values provided in 10 CFR 50.67 and the accident specific dose criteria specified in SRP Section 15.0.1. The NRC staff finds that the licensee used analysis assumptions and inputs consistent with applicable regulatory guidance identified in Section 2.0 of this SE. The assumptions associated with the LRA dose consequence analysis are summarized within the LAR as Table 7: Locked Rotor Accident Inputs and Assumptions, and SNCs calculated dose results are provided in Table 3.11 - Calculated Locked Rotor Accident Radiological Consequences. The NRC staff performed independent review of all inputs, assumptions and initial conditions in the SNC dose assessment files, and performed independent confirmatory calculations using RADTRAD version 5.1.1, as necessary, to ensure a thorough understanding of the licensee's methods and to verify that values used in the dose assessment code were in line with values provided in the LAR dated May 12, 2025. The NRC staff finds that the EAB, LPZ, and CR radiological doses for the LRA meet the applicable accident dose criteria and provide reasonable assurance of adequate protection and are, therefore, acceptable.

TABLES Table 1 Total Effective Dose Equivalent per Accident in roentgen equivalent man (rem)

Accident EAB1 LPZ2 SRP 15.0.1 and RG 1.183 Rev. 1 Acceptance Criteria Control Room 10 CFR 50.67 and GDC 19 Limit Loss of Coolant Accident 4.4 7.1 25 3.0 5

Fuel Handling Accident in Spent Fuel Pool 0.4 0.3 6.3 1.5 5

Main Steam Line Break Pre-incident Spike Concurrent Spike

<0.1 0.1

<0.1 0.2 25 2.5

<0.1 0.2 5

5 Steam Generator Tube Rupture Pre-incident Spike Concurrent Spike 1.8 1.1 1.0 0.6 25 2.5 1.5 0.6 5

5 Control Rod Ejection Accident Containment release Secondary release 3.7 0.6 4.5 0.6 6.3 6.3 1.4 1.9 5

5 Locked Rotor Accident

<0.1

<0.1 2.5 0.3 5

3.10 Environmental Qualification In Section 3.11 of the LAR dated May 12, 2025, SNC indicated that it intends to continue using the methodology in RG 1.89, Rev. 1 (ML003740271), Environmental Qualification [EQ] of Certain Electric Equipment Important to Safety for Nuclear Power Plants, for radiation environmental qualification of electrical equipment. RG 1.89, Rev. 1, assumes a release to containment and into recirculating sump fluid based on the AST provided in TID-14844, for the qualification of electrical equipment in these areas or exposed to radiation dose from the containment or recirculating sump fluid. TID-14844 provided the original accident source term assumptions to be used in reactor siting analysis and has traditionally been used for the accident source term in determining the accident dose for the EQ of most electrical equipment in currently operating reactors licensed under Part 50. The use of the TID-14844 AST has been found to be acceptable for continued use in EQ regardless of the licensees adopting the MHA LOCA AST specified in RG 1.183, Rev. 0, to demonstrate compliance with 10 CFR 50.67.

Resolution of the continued use of the TID-14844 accident source to demonstrate compliance with 10 CFR 50.49 can be found under the Resolution of Generic Safety Issues: Issue 187: The Potential Impact of Postulated Cesium Concentration on Equipment Qualification.

The AST presented in RG 1.183, Rev. 1, adopts an improved MHA LOCA AST, which bounds higher fuel enrichments and burnup levels. However, the adjustments to the AST are not attributed to either increased enrichments or burnup levels. Rather, differences among the source terms are due predominantly to improved understanding of the physics of core meltdown accidents. Heat losses from the degrading reactor core prolong the process of in-vessel release of radionuclides. Improved understanding of the chemistries of tellurium and cesium under reactor accidents changes the predicted behavior characteristics of these radioactive elements relative to what was assumed in the derivation of the RG 1.183, Rev. 0, source term.

While RG 1.183, Rev. 1, provides updated and acceptable guidance for demonstrating compliance with 10 CFR 50.67 with an AST, the MHA LOCA source terms described in all the methodologies (TID-14844, RG 1.183, Rev. 0, and RG 1.183, Rev. 1) all assume a substantial core damage accident. This substantial core damage to define the AST clearly bounds the 10 CFR 50.49 requirement to assess the most severe design basis accident. DBAs do not result in substantial core damage, because plant safety systems prevent significant core melting from occurring.

The NRC staff finds that assuming a substantial core damage source term continues to be conservative and provides assurance that the doses calculated for EQ of electrical equipment are adequately considered. Therefore, the AST of TID-14844 adequately addresses the requirement that the maximum dose to electrical equipment is considered with the RG 1.89, Rev. 1, assumption that radioactive material would be distributed uniformly throughout containment and the reactor coolant sump for large PWR dry containments. As a result, the NRC staff concludes that there would be no discernible risk reduction associated with requiring the use of the updated source term.

Based on the above, the NRC staff finds that the approach proposed by SNC to continue using RG 1.89, Rev. 1, methodology with the TID-14844 AST for radiological EQ of electrical equipment is sufficient to address the most severe DBA requirement in 10 CFR 50.49 and the requirements of EQ in GDC 4.

3.11 AST Methodology Technical Conclusion The NRC staff has reviewed the RG 1.183, Rev. 1, AST implementation proposed by SNC for Vogtle, Units 1 and 2. In performing its review, the NRC staff relied upon information placed on the docket by SNC and NRC staff review of all inputs, assumptions and initial conditions contained in the licensee dose calculations.

As described above, the NRC staff reviewed the assumptions, inputs, and methods used by SNC to assess the radiological impacts of the proposed adaptation of RG 1.183, Rev. 1. The NRC staff finds that the licensee used analysis methods and assumptions consistent with the guidance of RG 1.183, Rev. 1, with differences discussed, and accepted earlier in this SE. The NRC staff finds the methods and assumptions used by the licensee to be in compliance with applicable requirements. The NRC staff finds with reasonable assurance that SNCs estimates of the TEDE doses due to DBAs will comply with the requirements of 10 CFR 50.67 and meet the guidance of RG 1.183, Rev. 1.

The NRC staff finds reasonable assurance that Vogtle, Units 1 and 2, will continue to provide sufficient safety margins with adequate defense in depth to address design-basis accidents and to compensate for uncertainties in accident progression and analysis assumptions and parameters. Based on the above, the NRC staff concludes that the licensees request to modify its licensing basis to adopt RG 1.183, Rev. 1, and proposed implementation is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Georgia State officials were notified on February 16, 2026, of the proposed issuance of the amendments. On February 25, 2026, the State officials informed the NRC that the State of Georgia has no comments.

5.0 DISPOSITION OF PUBLIC COMMENTS On July 8, 2025, the NRC staff published a Notice of Consideration of Issuance of Amendments to Facility Operating Licenses, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing, in the Federal Register associated with the proposed amendment request (90 FR 30110). In accordance with the requirements in 10 CFR 50.91, the notice provided a 30-day period for public comment on the proposed no significant hazards consideration (NSHC) determination. A public comment was received regarding the proposed amendment (ML25217A461). The issue discussed in the public comment did not pertain to the proposed NSHC determination. The public comments pertained to carbon dioxide emissions and climate change. This issue in the public comment in not within the scope of the proposed NSHC.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration published in the Federal Register on July 8, 2025 (90 FR 30110). 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 amendment.

7.0 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.

Date: March 26, 2026

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 230 AND 212, REGARDING ALTERNATIVE SOURCE TERM USING REGULATORY GUIDE 1.183, REVISION 1 (EPID L-2025-LLA-0080) DATED MARCH 26, 2026 DISTRIBUTION:

PUBLIC RidsNrrLAKZeleznock Resource RidsNrrDorlLpl2-1 Resource RidsNrrDraArcb Resource RidsNrrDssSnsb Resource RidsNrrDssStsb Resource RidsNrrDssScpb Resource RidsNrrDexEltb Resource RidsNrrDexExhb Resource RidsNrrDnrlNcsg Resource RidsNrrDssSfnb Resource RidsACRS_MailCTR Resource RidsNrrPMVogtle Resource RidsRgn2MailCenter Resource ADAMS Accession Nos.:

ML26082A259 (Package)

ML26062A818 (Letter) 20260303-50033 (eConcurrence case)

via eConcurrence NRR-058

v t e Letter Sequence Approval
EPID:L-2025-LLA-0080, (Approved, Closed)
Initiation Acceptance Administration Questions Answers Results Approval, Approval, Approval, Approval, Approval Other: ML25321A270, ML25344A421
MONTHYEAR2025-07-11 [Table View]

Person / Time
ML26062A818
Site:	Vogtle (NPF-068, NPF-081)
Issue date:	03/26/2026
From:	John Lamb Plant Licensing Branch II
To:	Coleman J Southern Nuclear Operating Co
References
EPID L-2025-LLA-0080
Download: ML26062A818 (0)
v • d • e

ML26062A818

Text

SUBJECT:

Dear Ms. Coleman:

Enclosures:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

6.0 ENVIRONMENTAL CONSIDERATION

7.0 CONCLUSION

SUBJECT:

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