ML23242A229

Issuance of Amendment Nos. 315 and 315, Regarding Emergency Procedure Relocation of the Technical Support Center

ML23242A229
Person / Time
Site:	Surry
Issue date:	11/07/2023
From:	John Klos NRC/NRR/DORL/LPL2-1
To:	Carr E Virginia Electric & Power Co (VEPCO)
References
EPID L-2022-LLA-0167
Download: ML23242A229 (36)

Text

November 7, 2023 Mr. Eric S. Carr Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.

Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENT NOS. 315 AND 315, REGARDING REVISED EMERGENCY PLAN FOR RELOCATION OF THE TECHNICAL SUPPORT CENTER (EPID L-2022-LLA-0167)

Dear Mr. Carr:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 315 to Subsequent Renewed Facility Operating License No. DPR-32 and Amendment No. 315 to Subsequent Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively. The amendments revise the license and Emergency Plan in response to your application dated November 18, 2022, as supplemented by letters dated May 31 and August 7, 2023.

These amendments approve relocation of the Technical Support Center (TSC) from its current location adjacent to the Main Control Room (MCR) to a building outside the protected area that was used previously as the Local Emergency Operations Facility (LEOF).

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commissions monthly Federal Register notice. If you have any questions, please contact me at john.klos@nrc.gov, or 301-415-5136.

Sincerely,

/RA/

John Klos, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281

Enclosures:

1. Amendment No. 315 to DPR-32

2. Amendment No. 315 to DPR-37

3. Safety Evaluation cc: Listserv

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 315 Subsequent Renewed License No. DPR-32

1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated November 18, 2022, as supplemented by letters dated May 31 and August 7, 2023. complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, by Amendment No. 315, Subsequent Renewed Facility Operating License No. DPR-32 is hereby amended to authorize revision to the Emergency Plan for the Surry Power Station, Unit 1, as set forth in Virginia Electric and Power Company (Dominion Energy Virginia) application dated November 18, 2022, as supplemented by letters dated May 31 and August 7, 2023, and evaluated in the NRC staffs safety evaluation dated November 7, 2023.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-32 Date of Issuance: November 7, 2023 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2023.11.07 07:30:04 -05'00'

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 315 Subsequent Renewed License No. DPR-37

1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated November 18, 2022, as supplemented by letters dated May 31 and August 7, 2023. complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2.

Accordingly, by Amendment No. 315, Subsequent Renewed Facility Operating License No. DPR-32 is hereby amended to authorize revision to the Emergency Plan for the Surry Power Station, Unit 1, as set forth in Virginia Electric and Power Company (Dominion Energy Virginia) application dated November 18, 2022, as supplemented by letters dated May 31, and August 7, 2023, and evaluated in the NRC staffs safety evaluation dated November 7, 2023.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes License No. DPR-37 Date of Issuance November 7, 2023 Michael T.

Markley Digitally signed by Michael T. Markley Date: 2023.11.07 07:30:42 -05'00'

ATTACHMENT SURRY POWER STATION, UNIT NOS. 1 AND 2 LICENSE AMENDMENT NO. 315 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND LICENSE AMENDMENT NO. 315 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages License License No. DPR-32, page 3 License No. DPR-37, page 3 TSs N/A Insert Pages License License No. DPR-32, page 3 License No. DPR-37, page 3 TSs N/A Surry - Unit 1 Subsequent Renewed License No. DPR-32 Amendment No. 315

3. This subsequent renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 315 are hereby incorporated in the subsequent renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facility operating records in accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 Surry - Unit 2 Subsequent Renewed License No. DPR-37 Amendment No. 315

3. This subsequent renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power Levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 315 are hereby incorporated in this subsequent renewed license.

The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facility operating records in accordance with the Requirements of the Technical Specifications.

E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment 65 G. Deleted by Amendment 227 H. Deleted by Amendment 227

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 315 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 315 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated November 18, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22322A182), as supplemented by letters dated May 31 (ML23167B007) and August 7, 2023 (ML23220A148), Virginia Electric and Power Company (licensee, Dominion) submitted a license amendment request (LAR) to revise the license and Emergency Plan for Surry Power Station (Surry), Units 1 and 2. The proposed change would revise the Surry Emergency Plan (SEP) to allow the relocation of the Technical Support Center (TSC) from its current location adjacent to the Main Control Room (MCR) to a building that was used previously as the Local Emergency Operations Facility (LEOF). The proposed new TSC on the former LEOF is adjacent to the Surry training building on the north side of the Owner Controlled Area but outside of the Protected Area (PA). This proposed change also removes reference to the MCR as an alternate location for the TSC. The licensee states that the alternate TSC location is not the same as the alternative facility required by Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix E.IV.E.8.d, which remains unchanged by the proposed amendment.

The licensee states that the reason for the proposed SEP changes is that the current TSC building has been selected as the centralized location for the new non-safety-related control platform installation due to its close proximity to the MCR and the existing Surry plant computer

system. To support this new installation activity, Dominion proposes relocating of the current TSC to the building that was formerly used for the LEOF.

This proposed amendment would maintain the current licensing basis (CLB), including the previously approved Alternative Source Term (AST) has been applied in calculating radiological doses and habitability at the relocated TSC. This proposed revision to the SEP was evaluated by the licensee to be a reduction in effectiveness as defined in 10 CFR 50.54(q)(1)(iv), and the proposed changes were, therefore, submitted to the NRC for approval prior to implementation by the licensee as required under 10 CFR 50.54(q). The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on January 24, 2023, 88 FR 4216, and there has been no public comment on such finding.

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulations and guidance during its evaluation of Dominions proposed TSC relocation SEP changes.

2.1 Regulations Section 50.47(b)(8) of Title 10 of the Code of Federal Regulations (10 CFR) requires that adequate emergency facilities and equipment to support the emergency response are provided and maintained.

The regulations in 10 CFR 50.67, Accident source term, require, in part, in § 50.67(b)(2) that the applicants analysis must demonstrate 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) [roentgen equivalent man] 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; and (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) TEDE for the duration of the accident.

The regulations in 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 19 (GDC 19), Control room, require, in part, that:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary

instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

Additionally, 10 CFR Part 50, Appendix A, Criterion 19, states:

[H]olders of operating licenses using an alternative source term under § 50.67, shall meet the requirements of this criterion, except that with regard to control room access and occupancy, adequate radiation protection shall be provided to ensure that radiation exposures shall not exceed 0.05 Sv [Sievert] (5 rem) total effective dose equivalent (TEDE) as defined in § 50.2 for the duration of the accident.

Section IV.E.8.a(i) of 10 CFR Part 50, Appendix E, Emergency Planning and Preparedness for Production and Utilization Facilities, requires a licensee onsite TSC and an emergency operations facility (EOF) from which effective direction can be given and effective control can be exercised during an emergency.

Section IV.E.9.c of 10 CFR Part 50, Appendix E, requires that provision for communications among the nuclear power reactor control room, the onsite TSC, and the EOF; and among the nuclear facility, the principal State and local emergency operations centers, and the field assessment teams. Such communications systems shall be tested annually.

Section IV.E.9.d of 10 CFR Part 50, Appendix E, requires that provisions for communications by the licensee with NRC Headquarters and the appropriate NRC Regional Office Operations Center from the nuclear power reactor control room, the onsite TSC, and the EOF. Such communications shall be tested monthly.

2.2 Regulatory Guidance NUREG-0696 Functional Criteria for Emergency Response Facilities February,1981, Final Report (ML051390358), provides guidance for the design and implementation of emergency response facilities and criteria that the NRC staff will use in evaluating whether an applicant/licensee meets the requirements of 10 CFR 50, Appendix E.IV.E.8 and Appendix A, GDC 19. Section 2.9, Technical Data and Data System, states, in part, that the TSC displays shall include, but not be limited to, alphanumeric and/or graphical representations of plant system variables, in-plant radiological variables, meteorological information, and offsite radiological information. Section 2.8, Instrumentation, Data System Equipment, and Power Supplies, states that the design of TSC data system equipment shall incorporate human factors engineering with consideration for both operating and maintenance personnel.

NUREG-0737, Supplement No. 1 (ML102560009), Clarification of [Three Mile Island]

TMI Action Plan Requirements, January 1983, Section 8.2, Technical Support Center, provides guidance regarding acceptable means for meeting the fundamental TSC requirements, and Section 8.2.1, Requirements, Section b. states that the TSC will be located within the site PA so as to facilitate necessary interaction with control room, OSC

[operations support center], EOF and other personnel involved with the emergency.

Section 8.2.1.c. states that the TSC will be sufficient to accommodate and support NRC

and licensee pre-designated personnel, equipment and documentation in the center.

Section k. states that the TSC will be designed taking into account good human factors engineering principles.

NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition, Section 2.3.4, Short-Term Diffusion Estimates for Accidental Atmospheric Releases, Revision 3, (ML070730398) provides guidance to the NRC staff for use in the review of license applications concerning atmospheric dispersion models, meteorological data used for models, and the derivation of diffusion parameters.

NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Section 6.4, Control Room Habitability System, Revision 3, (ML070550069) provides guidance to the NRC staff for use in the review of license applications related to control room ventilation systems and control building layout and structures.

NUREG-1923, Safety Evaluation Report for an Early Site Permit (ESP) at the Vogtle Electric Generating Plant (VEGP) ESP Site, Section 13.3.3.2.8, Emergency Facilities and Equipment (10 CFR 50.47(b)(8); NUREG-0654/FEMA-REP-1, planning standard H), (ML092290650), provides the staffs review of the emergency response facilities and the equipment that will be used for accident assessment and monitoring functions following the declaration of an emergency.

Regulatory Guide (RG) 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors, Revision 6, June 2021 (ML21111A090), states that NUREG-0654/FEMA-REP-1, Revision 1, November 1980 (ML040420012), as amended in March 2002 (ML021050240). NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, Revision 1 (dated November 1980, ML040420012),

as amended in March 2002 (ML021050240) provides specific acceptance criteria for complying with the standards set forth in 10 CFR 50.47. Specifically,Section II.H of NUREG-0654/ FEMA-REP-1, Revision 1, includes guidance for TSCs, that each licensee shall establish a Technical Support Center and onsite operations support center (assembly area) in accordance with NUREG-0696, Revision 1.

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

RG 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, July 2000 (ML003716792) provides guidance to licensees of operating power reactors on acceptable applications of AST; the scope, nature, and documentation of associated analyses and evaluations; consideration of impacts on analyzed risk; and content of submittals. RG 1.183 establishes an acceptable AST and identifies the significant attributes of other ASTs that may be found acceptable by the NRC staff. RG 1.183 also identifies acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

RG 1.140, Design, Inspection, and Testing Criteria for Air Filtration and Adsorption Units of Normal Atmosphere Cleanup Systems in Light-Water-Cooled Nuclear Power Plants, Revision 3, August 2016, (ML16070A277), provides guidance to licensees and describes a method acceptable to the NRC regarding the design, inspection, and testing of normal atmosphere cleanup systems for controlling releases of airborne radioactive materials to the environment during normal operations, including anticipated operational occurrences.

RG 1.219, Guidance on Making Changes to Emergency Plans for Nuclear Power Reactors, Revision 1, July 2016 (ML16061A104) provides guidance to licensees making changes to their emergency plans with specific clarification on the meaning of reduction in effectiveness as stated in 10 CFR 50.54(q), the process for evaluating proposed changes to emergency plans, and a method for evaluating proposed changes to emergency plans.

RG 1.23, Meteorological Monitoring Programs for Nuclear Power Plants, Revision 1, March 2007, (ML070350028), provides guidance to licensees and applicants concerning criteria for an onsite meteorological measurements program.

3.0 TECHNICAL EVALUATION

The NRC staff considered the regulatory requirements and guidance documents in Section 2.0 of this safety evaluation (SE). In particular, the NRC staff reviewed the proposed TSC relocation against the guidance provided in Section 2, Technical Support Center, of NUREG-0696 and the plant-specific licensing history implementing NUREG-0737, Supplement 1.

The current SEP is Revision 71, dated October 22, 2020. In its submittal dated November 18, 2022, the licensee provided Attachment 2, Marked-Up Surry Emergency Plan Pages, hereafter referred to as the marked-up SEP (MU-SEP) pages that show the specific wording changes and revisions to the current SEP to reflect the proposed TSC relocation implementation. In MU-SEP Section 7.0, Emergency Facilities and Equipment, the SEP states that the TSC was designed to meet the intent of the guidance in NUREG-0696 and the clarification in NUREG-0737, Supplement 1.

3.1 TSC Function Section 2.1 of NUREG-0696 states that the TSC will provide the following functions:

Provide plant management and technical support to plant operations personnel during emergency conditions.

Relieve the reactor operators of peripheral duties and communications not directly related to reactor system manipulations.

Prevent congestion in the control room.

Perform EOF functions for the Alert Emergency class [Alert] and for the Site Area Emergency class [SAE] and General Emergency class [GE] until the EOF is functional.

In the LAR, Section 3.1.1, Function, the licensee states, in part, that Revision 71 of the SEP maintained the proposed new TSC support functions specified in NUREG-0696 and would continue to provide the following functions: to provide management and technical support to the plant operations personnel during emergencies, relieve the MCR of peripheral duties, prevent congestion in the MCR, and continue to perform EOF functions until the Corporate Emergency Response Center (CERC) is staffed. In addition, Revision 71 of the SEP, Figure 5.2, shows that the TSC will be activated for Alert, SAE, and GE. In the LAR, Attachment 2, The MU-SEP for page 1.5 defines the TSC as the central control center for the onsite emergency response organization after shift augmentation and MU-SEP Section 7.1.3, Technical Support Center, states that emergency response personnel will assemble at the primary TSC unless otherwise instructed by the SEM (Site Emergency Manager).

As part of its evaluation, the NRC staff physically walked down the current and proposed Surry TSC facilities. As discussed in the summary (ML23173A060) for the audit that was completed on June 14, 2023, the NRC verified independently the physical size, layout, and capabilities of the TSC to provide plant management and technical support to plant operations during an emergency.

Based on its review of the licensees LAR submittal, as supplemented, and the NRC staffs walkdown audit of the proposed TSC, the NRC staff concludes that the functional design capabilities of the proposed TSC adequately conform to the guidance in Section 2.1 of NUREG-0696. Therefore, the NRC staff finds that the proposed TSC meets the applicable requirements of 10 CFR 50.47(b)(8) and Subsection IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to TSC functions and functional capabilities.

3.1.1 Removal of Reference to MCR as an Alternate Location

Section 2.6, Habitability, of NUREG-0696 states, in part, that, if the TSC becomes uninhabitable, the TSC plant management function shall be transferred to the control room.

Dominion proposes to remove reference to the MCR as an alternate location for the TSC. The licensee states that NUREG-0696 does not require licensees to establish a backup or an alternate TSC. The alternate location for the TSC referenced in the current SEP serves as a pre-planned compensatory measure for a loss of TSC functionality. This alternate TSC location is not the same as the Alternative Facility required by 10 CFR 50, Appendix E.IV.E.8.d which remains unchanged by this proposed amendment. This pre-planned compensatory measure is currently described in Emergency Plan Implementing Procedure (EPIP)-3.02, "Activation of Technical Support Center," and will be maintained in the EPIP.

The NRC staff determined that this change is administrative in nature and is, therefore, acceptable. The NRC also notes that this change does not impact the emergency plan. The guidance in NUREG-0696 does not require licensees to establish a backup or an alternate TSC.

However, the licensee has a pre-planned compensatory measure for a loss of TSC functionality currently described in EPIP and the license will continue to maintain it in the EPIP.

Based on its review of the licensees LAR submittals and the site-walkdown audit of the proposed TSC, NRC staff has determined that the proposed removal of reference to the MCR as an alternate location for the TSC from the SEP adequately addresses the applicable guidance of NUREG 0737, Supplement 1, Section 8.2.1, and NUREG-0696, Section 2.6.

Therefore, the staff finds that the proposed change meets applicable requirements of 10 CFR 50.47(b)(8) and 10 CFR 50, Appendix E, paragraph IV.E.8.a(i).

3.2 TSC Location 3.2.1 Distance between TSC and MCR Locations Section 2.2 of NUREG-0696 states, in part, that:

The onsite TSC is to provide facilities near the MCR for detailed analyses of plant conditions during abnormal conditions or emergencies by trained and competent technical staff.

To accomplish this, the TSC shall be as close as possible to the MCR, preferably located within the same building. The walking time from the TSC to the control room shall not exceed 2 minutes. This close location will facilitate face-to-face interaction between control room personnel and the senior plant manager working in the TSC. This proximity will provide access to information in the control room that is not available in the TSC data system.

The guidance in Section 8.2.1, Requirements, Item b, of NUREG-0737, Supplement 1, (ML102560009) states that the TSC will be located within the site PA so as to facilitate necessary interaction with MCR, OSC, EOF and other personnel involved with the emergency.

In a previous precedent provided in NUREG-1923, the applicant proposed a common TSC for Units 1 through 4. to be located in the lower level of an administration building. The NRC staff evaluated various factors in determining the appropriateness and acceptability of providing flexibility relating to the 2-minute walking time between the TSC and control room in the guidance document, including the advances in communication technologies since NUREG-0696 was published in 1981. The NRC staff found that from a support and functional standpoint, the applicant's proposed TSC location was acceptable, subject to a demonstration of adequacy during the full participation exercise.

In LAR Section 3.1.1, Function, the licensee states that relocation of the TSC would continue to provide management and technical support to the plant operations personnel during emergencies. LAR Section 3.1.2, Location, also states that the plant data needed for emergency response provided to the MCR via the Plant Computer System (PCS) is available on TSC workstations which eliminates the need for TSC personnel to walk to the MCR to obtain plant data. The PCS provides plant monitoring, data acquisition, and critical plant data in the form of real-time status displays for the purpose of making a rapid evaluation of the reactor plants safety status. The PCS monitors are strategically located in the MCR, TSC, and CERC.

In addition, MU-SEP Section 7.1.3, Technical Support Center, states that the TSC contains controlled copies of selected manuals, procedures, and drawings.

Based on its review of the licensees LAR and walkdown of the proposed TSC during the audit, the LARs description of the proposed TSC PCS plant condition data and plant emergency information real-time status displays that will be located and available to personnel in the proposed TSC, the NRC staff determined that TSC personnel will not need to travel to the MCR to obtain additional plant condition data or plant status data during plant abnormal or emergency conditions. NSIR to look.

In LAR Section 3.1.2, the licensee states that While the proposed location of the new facility does not allow for direct face-to-face communications between the Shift Manager/SEM in the

MCR and the SEM in the TSC, adequate communications capability in the form of dedicated phone lines and use of inter-facility communicator positions ensures continued and effective communication is maintained. The guidance of NUREG-0696, Section 2.2, states, in part, that During recent events at nuclear power plants, telephone communications between facilities were ineffective in providing all of the necessary management interaction and technical information exchange. This ineffectiveness demonstrates the need for face-to-face communications between the TSC and control room personnel. The LAR states that the effectiveness of the proposed TSC communication and data capability is demonstrated during SEP drills and exercises using the Surry Simulator, MCR, and the existing TSC. In addition, the LAR also states that these processes have been the subject of inspection and have not resulted in observation of a performance deficiency. MU-SEP, Section 7.1.3, Technical Support Center, states that the dedicated phone line communications have been established with the MCR to keep TSC personnel knowledgeable on current operating evolutions and to provide consultation and recommendations to the MCR staff.

In LAR Section 3.1.2 the licensee states that the proposed TSC will relocate the TSC outside the Surry PA boundary, that this new location will be greater than a two-minute walk from the MCR, and that this new TSC location does not allow for direct face-to-face communications between the SEM in the MCR and the senior plant manager(s) working in the TSC. In LAR Section 3.1.3, Staffing and Training, the licensee states that the new TSC location will be approximately eight minutes walking distance from the MCR.

The licensee also states:

In LAR Section 3.1.2, Location, Item c, that the plant data needed for emergency response provided to the MCR via the PCS is available on TSC workstations, which eliminates the need for TSC personnel to walk to the MCR to obtain data. LAR Section 3.1.6, Habitability, Item c.4, Protective Equipment, also states that Improvements in voice and data communications capabilities eliminates the need for direct face-to-face communications.

The NRC staff has determined, based on its review of the licensees LAR and the staffs walkthrough audit of the proposed TSC, that the PSC real-time status displays of plant monitoring, critical plant data, and data acquisition, and the communication capabilities between the proposed TSC and MCR, demonstrate that:

1. The listed TSC PSC real-time plant data acquisition capabilities and the TSC-to-MCR communication effectiveness have reduced the need for direct face-to-face communications between the TSC and MCR personnel during abnormal conditions or emergency response events as described in NUREG-0696 guidance descriptions.

2. The LAR described TSC continuous and effective voice and data communications capabilities, and PCS real-time plant data provided to the TSC managers during a Surry emergency event addresses the intent of the TSC close proximity to the MCR guidance of NUREG-0696, Section 2.2.

3. Efficiency from advances in computer, data, and communications technology provides balance for the longer 8 minute walk time from the proposed TSC to the MCR such that relocation of the TSC to the new proposed location outside the Surry PA, collectively, does not adversely impact the intent of the TSC-to-MCR listed 2 minute walk time and face-to-face interaction guidance of NUREG-0696 or the listed guidance of

NUREG-0737, Supplement 1, describing that the TSC should be located within the site PA. This is consistent with the staffs analysis in in NUREG-1923.Per NUREG add to reg eval NSIR to add how an exceedance from 2 minutes is acceptable.

Based on its review of the licensees LAR, as supplemented, and the NRC staffs walkdown audit of the proposed TSC, the staff has determined that given the PCS real-time plant status capabilities and the proposed TCS listed communications capabilities, as noted above, the proposed TSC relocation design information adequately addresses the intent of the applicable location guidance of Section 2.2 of NUREG-0696 and Section 8.2 of Supplement 1 to NUREG-0737. Therefore, the staff finds that the proposed TSC relocation meets the applicable location requirements of 10 CFR 50.47(b)(8) and the applicable location requirements contained in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50.

3.2.2 Safe and Timely Movement of Personnel Between the TSC and the MCR Section 2.2 of NUREG-0696 states, in part, that:

Provisions shall be made for the safe and timely movement of personnel between the TSC and the control room under emergency conditions.

These provisions shall include consideration of the effects of direct radiation and airborne radioactivity from in-plant sources on personnel traveling between the two facilities.

There should be no major security barriers between these two facilities other than access control stations for the TSC and control room.

The LAR states that proposed TSC design has eliminated the need for TSC personnel to walk to the MCR to access data and the need for TSC and MCR personnel direct face-to-face communications during site emergencies (refer to SE Section 3.2.1, Distance between TSC and MCR Locations). In addition, the licensee also states in LAR Section 3.1.6, Habitability, Item c.4, Protective Equipment that:

Improvements in voice and data communications capabilities eliminates the need for direct face-to-face communications. Therefore, protective clothing to support personnel travel between the TSC and MCR is not necessary.

Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown audit of the proposed TSC, the NRC staff has determined that, given the proposed TSCs PCS real-time plant status displays available to the TSC personnel and the TSC voice and data communications capabilities sufficiently address the intent of the necessary management interaction and technical information exchange guidance during nuclear power plant events of Section 2.2 of NUREG-0696 and Section 8.2, of Supplement 1 to NUREG-0737. Therefore, the staff finds that the new location of the proposed TSC outside of the Surry PA would continue to meet the applicable requirements of 10 CFR 50.47(b)(8) and Subsection IV.E.8.a(i) of Appendix E to 10 CFR Part 50.

3.3 Staffing and Training Section 2.3 of NUREG-0696 states, in part, that:

Upon activation of the TSC, designated personnel shall report directly to the TSC and achieve full functional operation within 30 minutes.

The licensee designated TSC staff shall consist of sufficient technical, engineering, and senior designated licensee officials to provide the needed support to the control room during emergency conditions.

For the TSC to function effectively, TSC staff personnel must be aware of their responsibilities during an accident. The licensee shall, therefore, develop training programs for these personnel.

The TSC staff shall participate in TSC activation drills that shall be conducted periodically in accordance with the licensees emergency plan.

Operating procedures and staff training in the use of data systems and instrumentation shall contain guidance on the limitations of instrument readings including whether the information can be relied upon following such events as accidents resulting from earthquakes or the release of radiation.

Section 8.2.1, Requirements, of NUREG-0737, Supplement 1, states, in part, that:

When activated, the TSC is staffed by pre-designated technical, engineering, senior management, and other licensee personnel, and five pre-designated NRC personnel.

During periods of activation, the TSC will operate uninterrupted to provide plant management and technical support to plant operations personnel, and to relieve the reactor operators of peripheral duties and communications not directly related to reactor system manipulations.

The TSC will perform EOF functions for the Alert, SAE, and GE Emergency classifications until the EOF is functional.

Staffed by sufficient technical, engineering, and senior designated licensee officials to provide needed support, and be fully operational within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after activation.

The staff notes that the guidance of NUREG-0737, Supplement 1, Section 8.2.1.j, published in January 1983, states that the TSC should be fully operational within approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (60 minutes) after activation (versus 30 minutes as stated in NUREG-0696, published February 1981).

In LAR Section 3.1.3, Staffing and Training, the licensee states that the proposed SEP revisions to support the TSC relocation maintain the existing TSC staffing levels, TSC Emergency Response Organization (ERO) training, and the 60-minute augmented response times for the TSC responders, and continues to provide staffing of senior management,

engineering, and technical support positions, and a TSC staff training program. In addition, the response times are currently demonstrated in drills and exercises. As such, the licensee is not proposing any changes to ERO staffing as currently described in the SEP and the licensees submittal further states that the training of the ERO will continue to be maintained as currently described in the SEP.

Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown audit of the proposed TSC, the NRC staff has determined that the staffing and training of the proposed TSC, including ERO response times and drills and exercises, remains unchanged from that currently described in the SEP. Based on the above, the NRC staff concludes that the staffing and training descriptions for the proposed TSC conform to the guidance in Section 2.3 of NUREG-0696 and Section 8.2 of NUREG-0737, Supplement 1, and, therefore, meet the applicable requirements of 10 CFR 50.47(b)(8) and paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to the staffing and training of TSC personnel.

3.4 Size Section 2.4 of NUREG-0696 states that the TSC working space shall be sized for a minimum of 25 persons, including 20 persons designated by the licensee and 5 NRC personnel, with enough space to allow a minimum of 75 square feet/person (1,875 square feet).

In LAR Section 3.1.4, Size, the licensee states that the proposed TSC change will increase the size of the TSC from 3,000 square feet to approximately 3,400 square feet and that this assures a minimum of 75 square feet of working space per person for the 20 TSC ERO positions identified in the current revision of the SEP. The proposed TSC:

Size is larger than the current TSC.

Replicates the layout of the current TSC facility.

Consist of a TSC Operations Floor and separate rooms for Operations Support, Technical Support, Dose Assessment, and NRC personnel.

Includes a breakroom and bathroom to support long-term operation of the facility.

Provides library space with space for the storage of plant records and historical data.

Provides adequate space to support maintenance of TSC data, communications systems, and equipment.

Includes an equipment room for housing the TSC support systems and equipment to include Local Area Network (LAN) and TSC communications network switches.

The NRC staff observed the above proposed TSC physical design details during its walkdown audit, in addition to other proposed TSC design details.

Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown audit of the proposed TSC, the NRC staff has determined that the proposed TSC will be of sufficient size to accommodate and support Surry TSC ERO personnel, NRC personnel, TSC equipment, and documentation. As such, the NRC staff concludes that the proposed TSC

design descriptions conform to the guidance of Section 2.4 of NUREG-0696 and the applicable guidance of NUREG-0737, Supplement 1, Section 8.2.1. Therefore, staff finds that the proposed TSC size would continue to meet the applicable requirements of 10 CFR 50.47(b)(8) and the applicable requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to adequate sizing of the proposed TSC.

3.5 Structure Section 2.5 of NUREG-0696 provides the following structural guidance for the TSC:

The TSC complex must be able to withstand the most adverse conditions reasonably expected during the design life of the plant including adequate capabilities for:

(1) earthquakes, (2) high winds (other than tornadoes), and (3) floods.

Winds and floods with a 100-year-recurrence frequency are acceptable as a design-basis.

Existing buildings may be used to house the TSC complex if they satisfy the above minimum criteria.

In addition, the guidance of NUREG-0737, Supplement 1, Section 8.2.1.d, states that the TSC will be structurally built in accordance with the Uniform Building Code (UBC).

In LAR Section 3.1.5, Structure, the licensee states that the proposed new location for the TSC is in the building that formerly housed the LEOF. In its letter dated August 7, 2023, the licensee stated that the LEOF was engineered and designed in accordance with the 1978 Building Officials Code Administrators International (BOCA) Code which was the UBC used by the State of Virginia at the time the LEOF was designed. In its letter dated August 7, 2023, the licensee stated that the LEOF was constructed in accordance with the 1981 BOCA Code, and even though the LEOF was not designed as a Class 1 structure, any building constructed in accordance with the BOCA Code is considered a well-engineered structure with adequate capability to withstand earthquakes.

The proposed TSC (former LEOF building) is constructed of 12-inch-thick reinforced concrete exterior walls, roof, and a 24-inch-thick mat/slab. The LAR states that a review of this design determined that due to the thickness and reinforcement of the walls and slab floor, the proposed TSC structure exceeds the requirements from the UBC during that timeframe and can withstand the applicable loading. The LAR also states that the proposed TSC building, as designed, will withstand the 100-year wind speeds. Relative to flooding, the maximum 100-year flood in the vicinity of Surry is 19 feet Mean Sea Level. The proposed TSC building has a finished floor elevation of 33 feet.

Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown audit of the proposed TSC, the NRC staff has determined that the LAR structure design descriptions and the physical structure of the proposed TSC conform to the guidance in Section 2.5 of NUREG-0696 and the applicable guidance of Section 8.2.1 of NUREG-0737, Supplement

1. Therefore, NRC staff finds that proposed TSC structure meets the applicable requirements of

10 CFR 50.47(b)(8) and Subsection IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to the structural integrity of the proposed TSC.

3.6 Habitability 3.6.1 TSC Personnel Radiological Protection Equipment Section 2.6, Habitability, of NUREG-0696 states, in part, that:

Equipment that protects personnel shall be provided in the TSC for the staff who must travel between the TSC and the control room, or the EOF, under adverse radiological conditions.

Protective equipment also shall be provided to allow TSC personnel to continue to function during the presence of low-level airborne radioactivity or radioactive surface contamination.

Sufficient potassium iodide [Kl] shall be provided for use by TSC and control room personnel.

If the TSC becomes uninhabitable, the TSC plant management function shall be transferred to the control room.

In LAR Section 3.1.6, Habitability, the licensee states that the proposed TSC building consists of 12-inch-thick reinforced concrete exterior walls and roof, and a 24-inch-thick mat/slab, and that an existing hatch between the TSC penthouse and the ground level of the proposed TSC (occupied space) will be permanently sealed with a 12-inch concrete plug to prevent radiation exposure to the occupied TSC space below.

The LAR states that proposed TSC design has eliminated the need for TSC personnel to walk to the MCR to access data and the need for TSC and MCR personnel direct face-to-face communications during site emergencies (refer to Section 3.2.1 of this SE, Distance between TSC and MCR Locations). In addition, LAR Section 3.1.6.c.4, Protective Equipment, states that due to the proposed TSCs design improvements in voice and data communications capabilities, the need for direct face-to-face communications is eliminated. In its letter dated Augst 7, 2023, the licensees response to RAI 3 concludes that:

With advances in communication technology and the use of improved plant computer displays, the distance between the emergency response facilities is no longer material as inter-facility travel is no longer necessary or performed. As a result, there is no longer a need to maintain protective clothing within the TSC for this purpose.

In the unlikely event that radiation levels rendered the TSC uninhabitable, the TSC functions would be transferred to the MCR and CERC, and TSC personnel would relocate to the near-site facility or alternative facility. Thus, the need for providing protective equipment in the TSC is obviated.

Protective equipment can be dispatched to the TSC on an as needed basis rather than maintaining these items in the TSC.

The MU-SEP Appendix Table titled Emergency Kits [Health Physics] HP Area, Control Room, OSC, TSC, documents that TSC personnel radiological protection equipment will not be included in the proposed TSC.

The SEP implementing procedures require potassium iodide to be maintained in the TSC and will continue to be required in the proposed TSC. In addition, thyroid-blocking agents will continue to be maintained onsite for use as needed.

The licensee has provided adequate information for not having protective equipment (i.e PCs) maintained in the TSC.

Based on its review of the licensees submittal, as supplemented, the NRC staff has determined that the proposed TSC design to eliminate the need for TSC ERO personnel to travel to the MCR to obtain additional plant data during an emergency addresses the intent of the NUREG-0696, Section 2.6 guidance to provide protective equipment to protect TSC personnel who must travel between the TSC and the MCR under adverse radiological conditions.

Therefore, the NRC staff finds that the proposed TSC meets the applicable requirements of 10 CFR 50.47(b)(8) and Subsection IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to TSC functions and functional capabilities.

3.6.1.1 TSC Radiation Monitoring Systems Section 2.6 of NUREG-0696 states, in part, that:

Radiation monitoring systems shall be provided in the TSC. These monitoring systems may be composed of installed monitors or portable monitoring equipment dedicated to the TSC.

These systems shall continuously indicate radiation dose rates and airborne radioactivity concentrations inside the TSC while it is in use during an emergency.

These monitoring systems shall include local alarms with trip levels set to provide early warning to TSC personnel of adverse conditions that may affect the habitability of the TSC.

Detectors shall be able to distinguish the presence or absence of radioiodine at concentrations as low as 10-7 microcuries/cc.

In LAR Section 3.1.6, the licensee states that the proposed TSC will be provided with:

One (1) Victoreen radiation monitor to detect airborne radioactivity which will include particulate, iodine, and noble gas detectors, and will be able to distinguish the presence or absence of radioiodine at concentrations as low as 10E-7 microcuries/cc. This monitor will continuously sample the atmosphere from locations throughout the TSC and provide an audible alarm to alert TSC personnel of adverse conditions.

Two (2) Mirion DRM-2 general area radiation monitors will be wall mounted at separate locations around the TSC operation floor and will provide an audible alarm to alert TSC personnel of adverse conditions.

These radiation monitors will provide continuous indication of the dose rate and airborne radioactivity in the TSC during an emergency and will alert personnel of adverse radiological conditions.

Based on its review of the licensees submittal, as supplemented, the NRC staff has determined that the proposed TSC design meets the intent of the NUREG-0696, Section 2.6 guidance to provide radiation monitoring, and indication systems (with alarms and detectors) to protect TSC personnel. Therefore, the NRC staff finds that the proposed TSC meets the applicable requirements of 10 CFR 50.47(b)(8) and Subsection IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to TSC functions and functional capabilities.

3.6.1.2 Atmospheric Dispersion The licensees LAR involves the TSCs relocation and a /Q value recalculation for the demonstration of radiological dose for the proposed new TSCs habitability. In its letter dated May 31, 2023, the licensee included the meteorological data and the associated inputs and assumptions used in the atmospheric dispersion analysis. In Attachment 3 of the that supplement the licensee identified the following release pathways to the TSC for the atmospheric dispersion analysis: U1 Containment, U2 Containment, Auxiliary Building East Louvers, Auxiliary Building West Louvers, and Vent #2. The licensee developed new /Q values for each release pathway for the TSC receptor.

3.6.1.2.1 Meteorology Data In its licensees letter dated May 31, 2023, the licensee provided information regarding the atmospheric dispersion analysis which contained hourly onsite meteorological data for a 5-year period from calendar years 2009 through 2013. The meteorological data were formatted for the ARCON96 atmospheric dispersion code (NUREG/CR-6331, Revision 1, Atmospheric Relative Concentrations in Building Wakes, ADAMS Accession No. ML17213A187) in order to calculate updated /Q values for the TSC. This format contained hourly data on wind speed, wind direction, and atmospheric stability class taken from the 10.6-m and 46.1-m levels of the meteorological tower.

The NRC staff previously completed a detailed review related to the acceptability and representativeness of the onsite hourly meteorological data for the issued alternative source term LAR (ML18075A021). Based on this review, the NRC staff considers the onsite meteorological dataset from calendar years 2009 through 2013 suitable for use in making calculations for the atmospheric dispersion analyses used to support this LAR.

3.6.1.2.2 TSC Atmospheric Dispersion Analysis In support of this LAR, the licensee used the computer code ARCON96 to estimate TSC /Q values for potential accidental releases of radioactive material. RG 1.194 endorses the ARCON96 model for determining /Q values to be used in the design-basis evaluations of control room radiological habitability, and the staff finds the licensees use of ARCON96 acceptable for estimating /Q values for the TSC for potential accidental releases.

The ARCON96 Code estimates /Q values for various time-averaged periods ranging from 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 30 days. The meteorological input to ARCON96 consists of hourly values of wind speed, wind direction, and atmospheric stability class. The /Q values calculated through

ARCON96 are based on the theoretical assumption that material released to the atmosphere will be normally distributed (Gaussian) about the plume centerline. A straight-line trajectory is assumed between the release points and receptors. The diffusion coefficients account for enhanced dispersion under low wind speed conditions and in building wakes.

In its letter dated May 31, 2023, the licensee provides Attachment 1 that includes Table 5.1, Summary of Changes for TSC Relocation /Q Calculation. The table includes the meteorological data period, the lower and upper height measurements, the units of wind speed used in the analysis, describes the source input values of vertical velocity, stack flow, stack radius, and diffusion coefficients. Table 5.1 also includes the release type, release height, building area for each release pathway as well as the distances to receptor, intake height, elevation difference, and direction to source for each release pathway. Finally, Table 5.1 lists the /Q values from the ARCON96 output for the 0-2, 2-8, 8-24, 24-96, and 96-720, hour time-intervals for each of the release pathways.

The NRC staff confirmed independently the licensees atmospheric dispersion estimates by running the ARCON96 computer model and obtaining similar results. Both the NRC staff and licensee used a ground-level release assumption for each of the release pathway-receptor combinations as well as the previously discussed source-receptor distances, directions, heights, and area values. Based on the results of its confirmatory analysis, the NRC staff finds the licensees TSC /Q values acceptable for use in the radiological dose assessments for the LAR.

3.6.1.2.3 Conclusion The NRC staff reviewed the guidance, assumptions, and methodology used by the licensee to assess the /Q values associated with postulated releases from the potential release pathways.

The NRC staff found that the licensee used methods consistent with RG 1.194. The licensee also used onsite meteorological data that complied with the guidance of RG 1.23. The inputs and assumptions used to calculate the TSC /Q values were also consistent with the guidance of RG 1.194. Based on the above, NRC staff finds the licensees proposed /Q values acceptable for use in calculating the radiological dose assessments associated with the LAR.

Therefore, the NRC staff finds that the proposed TSC meets the applicable requirements of 10 CFR 50.47(b)(8) and Subsection IV.E.8.a(i) of Appendix E to 10 CFR Part 50 with regard to TSC functions and functional capabilities.

3.6.2 TSC Dose Consequence Analysis In its letter dated May 31, 2023, the licensee documents changes in the atmospheric dispersion factors, to model the 30-day TEDE to emergency response occupants in the relocated TSC from a design-basis dose consequence loss-of-coolant accident (LOCA) and to model all necessary modifications to design-basis input parameters. The use of the design-basis dose consequence LOCA dose consequence analysis is appropriate since the assumed magnitude of the release for this accident bounds all other design-basis dose consequence analyses.

The NRC staff notes that the design-basis dose consequence LOCA analysis is intended to be based upon a major accident or possible event, resulting in dose consequences not exceeded by those from any accident considered credible. Historically, this accident analysis, which is performed to show compliance with the dose criteria specified in 10 CFR 50.67, is referred to as the maximum hypothetical accident. The requirements of 10 CFR 50.46 ensure that the emergency core cooling system will prevent significant core damage during a design-basis LOCA. Notwithstanding the requirements of 10 CFR 50.46, the maximum hypothetical accident

for dose consequence determinations deterministically assumes a substantial core melt with an appreciable release of fission products into the containment. Therefore, the maximum hypothetical accident is a conservative surrogate to enable a deterministic evaluation of the response of a facilitys engineered safety features (ESFs). All design-basis dose consequence accident analyses are performed in an intentionally conservative manner in order to compensate for known uncertainties in accident progression, activity product transport, and atmospheric dispersion.

Section 2.6, Habitability, of NUREG-0696 states, in part, that:

[The TSC] shall have the same radiological habitability as the control room under accident conditions.

TSC personnel shall be protected from radiological hazards, including direct radiation and airborne radioactivity from in-plant sources under accident conditions, to the same degree as control room personnel. Applicable criteria are specified in 10 CFR, Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, GDC 19, Control Room, SRP 6.4, and NUREG-0737, Item II.B.2.

The TSC ventilation system shall function in a manner comparable to the control room ventilation system.

NUREG-0737, Supplement 1, Section 8.2.1.f, states that the TSC will be:

Provided with radiological protection and monitoring equipment necessary to assure that radiation exposure to any person working in the TSC would not exceed 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident.

3.6.2.1 Source Term and Transport The licensee incorporated the source term used in the CLB LOCA accident dose consequence analysis in license amendment nos. 306 and 306, for Surry, Units 1 and 2, which updated the loss-of-coolant accident alternate source term dose analysis (ML21253A063). The licensee followed all aspects of the guidance outlined in RG 1.183, Revision 0, Regulatory Position 3, regarding the reactor core inventory, release fractions, and timing for the evaluation of its dose consequence LOCA. The radioactivity released into the containment is assumed to terminate at the end of the early in-vessel phase that occurs 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the onset of a LOCA.

In its letter dated November 18, 2022, Attachment 1 page 17 of 36, the licensee states that the LOCA dose consequence calculation analysis includes dose contributions from the following potential radioactive material release pathways.

Containment leakage Emergency core cooling leakage Leakage to the Reactor Water Storage Tank Containment direct shine Cloud shine

TSC ventilation system filter shine.

A detailed discussion of the licensees CLB LOCA source term and transport assumptions can be found in two Surry license amendment requests (ML003704270 and ML20338A542).

3.6.2.2 Infiltration of Airborne Radioactivity Using the CLB source term and transport assumptions for the dose consequence LOCA analysis, the licensee in its letter dated May 31, 2023, Table 4-1: Summary of TSC Dose Results, further identified and evaluated the TEDE contributions to TSC occupants from the following constituents:

Infiltration of airborne radioactivity into the TSC.

Direct dose from the accumulation of radioactive material on the TSC filter system.

Cloudshine - Dose from the external cloud of airborne radioactive materials surrounding the TSC.

Containment Direct Shine - Direct dose from the source term in containment.

Containment Skyshine - Dose from radiation scatter from atmosphere back to ground.

In the licensees letter dated November 18, 2022, Attachment 1, the licensee states that to limit the infiltration of airborne radioactivity, the TSC emergency ventilation system is designed to provide a filtered makeup air flow rate to maintain the TSC habitability envelope at a positive pressure (Attachment 1, page 13 of 36). In its letter dated May 31, 2023, the licensee also states the filtration system air flow rates in Table 5-2,Summary of Changes for TSC Relocation LOCA Dose.

In its letter dated May 31, 2023, Attachment 1, Table 5-2, Summary of Changes for TSC Relocation LOCA Dose, the licensee states that the makeup flow rate is specified as a minimum of 670 cubic feet per minute (cfm) to a maximum of 1,100 cfm operational flow rate before isolation. The licensee also states that for the TSC mixing volume of 34,900 cubic feet, the after isolation, minimum pressurization makeup flow rate of 603 cfm will result in more than 0.5 volume changes per hour In addition, the TSC ventilation system includes two-door vestibules to eliminate the intake of unfiltered in-leakage. The volume change stated above by the licensee, per SRP 6.4, allows for the least restrictive frequency for periodic verification.

Additionally, the licensees design uses a two-door vestibule to eliminate unfiltered in-leakage as is allowed by footnote 4 in SRP 6.4, footnote 4, (ML070550069) which states for a pressurized control room that infiltration is normally 5 L/s (10 cfm) infiltration [which] is assumed for conservatism. This flow could be reduced or eliminated if the applicant provides assurance that backflow (primarily as a result of ingress and egress) will not occur. This may mean installing two-door vestibules or equivalent. (NOTE: Unfiltered in-leakage is usually the summation of the contribution from opening and closing doors associated with activities in accordance with the plant emergency plans and procedures. Normally, 10 cfm is used for this unfiltered in-leakage.)

In its letter dated May 31, 2023, Attachment 1, the licensee states that the TSC ventilation filtered makeup pressurization system includes a filter bank consisting of a high-efficiency particulate air (HEPA) filter in series with a high-efficiency gas adsorption (HEGA) filter with a nominal, 1000 cfm, flow rate. This filter removes particulate radioactive air contaminants and remaining iodine compounds. RG 1.140 (ML16070A277) provides guidance regarding filter

efficiencies of 99, 95, and 95 percent for particulates, elemental iodine, and organic iodide, respectively. In its letter dated May 31, 2023, the licensee assigns an efficiency of 95 percent for particulates (Table 5-2, ML23167B007) which the NRC staff has reviewed as adequate.

In the licensees letter dated May 31, 2023, Attachments 1 and 2, the licensee states that the licensees dose consequence analysis provides two dose calculations, one which assumes that the TSC filtered makeup pressurization system will be activated at 30 seconds, immediately after a Safety Injection (SI) signal, and a second analysis which isolates the TSC 60 minutes after accident initiation. The licensee states that the TSC will not be provided automatic isolation of the ventilation system upon an SI signal, therefore the licensee relied on the analysis which provided manual isolation at 60 min.

3.6.2.3 Direct Shine Dose from the TSC Filtration System In its letter dated May 31, 2023, the licensee used acceptable assumptions to evaluate the dose to TSC occupants from radioactive materials retained in the TSC filtration system. The licensee maximized the calculated dose from the TSC filters by assuming that the HEPA and charcoal filters were 99 percent efficient, which maximized the dose contribution from filter shine.

The licensee performed a detailed evaluation of the filter dose contribution using the Monte Carlo N Particle (MCNP) Transport Code to assess dose from TSC filters. The results of the evaluation using MCNP Transport Code were used to determine the shine dose to the TSC personnel from the nuclide buildup on the TSC Heating Ventilation Air Conditioning filters and is documented in the supplement dated May 31, 2023, Table 4-1: Summary of TSC Dose Results.

3.6.2.4 Direct Shine Dose from the External Cloud of Airborne Radioactive Materials, Containment Direct Shine, and Containment Skyshine In the licensees letter dated May 31, 2023, the licensee states that the increase in the distance to the proposed TSC from the previous location resulted in higher total TSC dose primarily due to higher results in the dose contributions due to containment leakage and containment skyshine. These two values represent over 60 percent of the calculated dose, and each of the other contributors represent less than 10 percent of TEDE limit. Other increases can be attributed to modeling calculations and design of the new TSC. The staff reviewed the methods and assumptions used by in these calculations to be in compliance with applicable requirements. The staff compared the doses estimated by in the LAR to the applicable acceptance criteria and to the results estimated by the staff in its confirmatory calculations. The NRC staffs independent confirmatory calculations of the shine components utilized Microshield v13 and obtained results within regulatory limits and similar to results provided by the licensee.

The increase in the dose contribution due to the containment leakage, is predominantly due to the newly modeled /Q values for each release pathway for the TSC receptor and isolation time.

The value for containment leakage rose from 0.1350 rem TEDE to 0.5866 rem TEDE over the duration of the accident (ML23167B007, Table 4-1: Summary of TSC Dose Results).

The increase in the dose contribution due to the containment skyshine, is due to a number of changes in the relocation analysis. These changes include but are not limited to lack of credit for TSC wall or roof shielding, and use of fewer time edits for containment inventory. The value for containment skyshine rose from 0.3810 rem TEDE to 1.6028 rem TEDE over the duration of the accident (ML23167B007, Table 4-1: Summary of TSC Dose Results).

Summary of TSC Dose Results Data (see ML23167B007):

LOCA Dose Component TSC Dose [rem TEDE]

TSC Relocation

[CLB]

AST TSC Dose SI-Based Isolation 60-min Isolation Dose Containment Leakage 0.07 0.5866 0.135 ECCS Leakage 0.2055 0.2691 0.111 RWST Leakage 0.0162 0.0173 0.21 Containment Direct Shine 0.005 0.005 0.047 Containment Skvshine 1.6028 1.6028 0.381 Cloudshine 0.0361 0.0361 0.195 Filter Shine 0.1 0.1 0.16 Containment Shine through Main Steam Line Penetrations 0.005 SI Piping under Main Steam Valve House & OS Pump House 0.194 Hydrogen Re-combiner Vault 0.004 Total 2.04 2.62 1.442 Sources indicated with a result of-were deemed negligible with respect to the new TSC location due to source/receptor geometry.

3.6.2.5 TSC Dose Consequence Analysis Conclusion The results of the licensees calculation indicate that the TSC personnel will be protected from radiological hazards, including direct radiation and airborne radioactivity from in-plant sources under accident conditions, to the same degree as MCR personnel (radiation exposures shall not exceed 5 rem TEDE for the duration of the postulated accident). Also requirements contained in Clarification of TMI Action Plan Requirements NUREG-0737 (ML051400209) were considered in this SE.

The staff used RADTRAD version 5.0.3 and MicroShield version 13 to perform independent confirmatory calculations. These confirmatory calculations utilizing the LOCA analysis initial conditions, inputs and assumptions provided in the LAR and CLB documents and obtained results similar to those provided in the application and within acceptance criteria. Based on its

review of the licensees dose calculation and the associated atmospheric dispersion estimates, the NRC staff has determined there is reasonable assurance that the proposed TSC will provide its occupants an adequate level of radiological protection under design-basis accident conditions.

3.6.2.6 TSC Ventilation System Section 2.6 of NUREG-0696 states, in part, that:

The TSC ventilation system shall function in a manner comparable to the control room ventilation system. The TSC ventilation system need not be seismic Category I qualified, redundant, instrumented in the control room, or automatically activated to fulfill its role.

A TSC ventilation system that includes HEPA, charcoal filters is needed as a minimum. The capacity of the installed TSC ventilation filter system shall be independent of these thyroid-blocking provisions.

In addition, the guidance of NUREG-0737, Supplement 1, Section 8.2.1.e, states, in part, that the TSC should be:

Environmentally controlled to provide room air temperature, humidity, and cleanliness appropriate for personnel and equipment.

In Attachment 1 to its letter dated November 18, 2022, the licensee states that the ventilation system for the proposed TSC has a filter bank that consists of a HEPA filter in series with a HEGA filter with a nominal flow rate of 1000 cubic feet per minute (cfm). The HEPA filter removes particulate radioactive air contaminants and the HEGA removes other pollutants such as iodine compounds. The ventilation system design is sized to provide heating and cooling that maintains facility temperature at approximately 75°F dry bulb(+/-30F) during the summer and 72°F dry bulb (+/-3°F) during the winter. Additionally, there is an alarm function which provides an alert to the TSC staff of an Emergency HVAC system component failure.

3.6.2.7 TSC Habitability Conclusion Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown audit of the proposed TSC, the NRC staff has determined that the proposed TSC ERO personnel will be adequately protected from radiological hazards including direct radiation and airborne radioactivity from in-plant sources under accident conditions and that the proposed TSC design sufficiently addresses the radiological protection guidance for TSC ERO personnel, similar to the MCR, as identified in Section 2.6 of NUREG-0696 and Section 8.2 of NUREG-0737, Supplement 1, and the control room habitability requirements of GDC 19. The licensee also performed a dose calculation and verified that the proposed TSC conforms to the guidance outlined in Section 2.6 of NUREG-0696 and Section 8.2 of NUREG-0737, Supplement 1, including assurance that the dose to the proposed TSC occupants is limited to less than five rem TEDE for the 30-day accident mitigation period.

As such, the NRC staff concludes that the habitability design of the proposed TSC conforms to the habitability guidance in Section 2.6 of NUREG-0696 and Section 8.2 of NUREG-0737, Supplement 1.Based on the above and the NRCs confirmatory dose analysis, the NRC staff finds that Surry proposed TSC habitability design meets the applicable requirements of 10 CFR

50.47(b)(8) and the applicable requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50, with regard to habitability for TSC ERO personnel.

3.7 Communications Guidance in NUREG-0696, Section 2.7, specifies that the TSC will be the primary onsite communications center during an emergency at a nuclear power plant and that it needs to include reliable primary and backup communications to the MCR, OSC, EOF, and NRC, as well as State and local operations centers. The TSC voice communications facilities shall include means for reliable primary and backup communication. In addition, the guidance of NUREG-0737, Supplement 1, Section 8.2.1.g, specifies that the TSC should be provided with reliable voice and data communications with the MCR and EOF and reliable voice communications with the OSC, NRC Operations Center and State and local operations centers.

In LAR Section 3.1.7, Communications, the licensee states that the proposed TSC will replicate the communications capabilities provided in the existing TSC and will continue to provide communications with the MCR, OSC, onsite personnel, mobile monitoring teams, CERC, Offsite Response Organizations, and the NRC. Further in the MU-SEP, Section 7.1.4, Corporate Emergency Response Center, the licensee states that the CERC is the consolidated EOF for Surry.

The communications capabilities of the proposed TSC will continue to include those communications capabilities currently in use to support engineering assessment activities, including damage control team planning and preparation. The proposed TSC communications capabilities include dedicated voice communications to the MCR, OSC, CERC, Virginia Emergency Operations Center, Primary Remote Assembly Area, the Security Shift Supervisor, and the Radiation Protection Supervisor. The proposed TSC communications also include the new Dominion Energy Emergency Notification System (DEENS), Station Private Branch Telephone Exchange, commercial lines, public address intercom, a station radio system, and NRC lines.

The SEP, Section 7.2, Communications Systems, states that the station communications system is designed to provide redundant means to communicate with all essential areas of the station associated with Surry and to essential locations remote from the station during normal operation and under accident conditions. A failure of one communication system will not affect the operation of other communication systems at the Surry. The communication systems have diverse power supplies. In addition, the onsite communication systems normally will be in use or will be periodically tested. Therefore, equipment failure will not go unnoticed.

Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown of the proposed TSC, the NRC staff has determined that the proposed TSC has appropriate communications capabilities to support TSC functions and NRC response activities. As such, the NRC staff concludes that the proposed TSC communication design meets the guidance of NUREG-0696, Section 2.7, and NUREG-0737, Supplement 1, Section 8.2.1. Therefore, the staff finds that the proposed TSC conforms to 10 CFR 50.47(b)(8) and the applicable requirements in Subsections IV.E.8.a(i), IV.E.9.c, and IV.E.9.d of Appendix E to 10 CFR Part 50, with regard to reliable TSC voice and data communications capabilities to the NRC and other Surry emergency response facilities and is acceptable.

3.8 Instrumentation, Data System Equipment, and Power Supplies Section 2.8 of NUREG-0696 states, in part, that:

Equipment shall be provided to gather, store, and display data needed in the TSC to analyze plant conditions. The data system equipment shall perform these functions independent of actions in the MCR and without degrading or interfering with MCR and plant functions.

The total TSC data system reliability shall be designed to achieve an operational unavailability goal of 0.01 during all plant operating conditions above cold shutdown.

The TSC electrical equipment load shall not degrade the capability or reliability of any safety-related power source. Circuit transients or power supply failures and fluctuations shall not cause a loss of any stored data vital to the TSC functions.

Sufficient alternate or backup power sources shall be provided to maintain continuity of TSC functions and to immediately resume data acquisition, storage, and display of TSC data if loss of the primary TSC power sources occurs.

In LAR Section 3.1.8, TSC Power Supplies, the licensee states that the proposed TSC electrical loads will be powered from utility power through an automatic transfer switch (ATS) which distributes the utility power. This utility power is the proposed TSC normal power supply.

If normal utility power is lost the ATS will automatically start backup power from a TSC dedicated 200 kW/250 KVA diesel generator and repower the TSC electrical distribution system.

During the time that the dedicated TSC diesel generator is starting and has yet to reach full speed and frequency, a 50 KVA uninterruptible power supply (UPS) will provide power to TSC critical loads through its batteries. In addition, emergency lighting for the proposed TSC Operations Floor and the NRC Communications Room are powered by the 50 KVA UPS. The 50 KVA UPS provides 15 minutes of power to critical TSC loads during the time the TSC power distribution system is transitioning to the backup diesel generator.

The sites PCS data capabilities will remain unchanged by the proposed TSC relocation. The proposed TSC will use PCS workstations that will be connected to the Surry LAN secure connections. These workstation connections will be connected through LAN switches inside the new proposed TSC which will be powered from the proposed TSC normal utility power and TSC diesel generator backup and UPS power supply system. LAR Section 3.1.9 further states that the PCS design provides system reliability to achieve an operational unavailability goal of 0.01 during all plant operating conditions above cold shutdown and that this proposed TSC PSC connectivity is functionally equivalent to the current TSC with respect to the data provided, means of access, method of presentation, and system reliability.

Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown of the proposed TSC, the NRC staff has determined that the proposed TSC provides for independent and reliable TSC instrumentation, data system equipment, and power supplies. As such, the NRC staff concludes that the TSC instrumentation, data systems, and power supply design conform to the guidance in Section 2.8 of NUREG-0696 and, therefore, meets the standard of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50, with regard to providing reliable equipment to gather, store, and display data needed in the TSC and is acceptable.

3.9 Technical Data and Data Systems Section 2.9 of NUREG-0696 states, in part:

The TSC technical data system shall receive, store, process, and display information acquired from different areas of the plant as needed to perform the TSC function. The data available for display in the TSC must enable the plant management, engineering, and technical personnel assigned there to aid the control room operators in handling emergency conditions.

The data set available to the TSC data system must be complete enough to permit accurate assessment of the accident without interference from the control room emergency operation.

There is to be data storage and recall provided for the TSC data set.

In addition, the guidance of NUREG-0737, Supplement 1, Section 8.2.1, specifies that the TSC should:

Be capable of reliable data collection, storage, analysis, display, and communication sufficient to determine site and regional status, determine changes in status, forecast status and take appropriate actions.

Make available in the TSC specific plant and meteorological variables as specified in Regulatory Guide 1.97, Rev. 2. Principally those data must be available that would enable evaluating incident sequence, determining mitigating actions, evaluating damage, and determining plant status during recovery operations.

In LAR Section 3.1.9, Technical Data, Data Systems, and Data System Equipment [Support Center]SC Power Supplies, the licensee states that the current PCS design and data capabilities remains unchanged by the proposed TSC relocation. The proposed TSC receives plant parameter data from the PCS. The PCS provides plant monitoring, data acquisition, and critical plant data in the form of real-time status displays for the purpose of making a rapid evaluation of the reactor plants safety status via the PCS workstation monitors which will be located in the proposed TSC. Meteorological information and onsite radiation monitor data is displayed through the PCS. In addition, meteorological data and effluent monitoring data is provided from the PCS directly to the Meteorological Information and Dose Assessment System for use in making offsite dose projections.

Offsite radiological data obtained from the field teams is relayed to the TSC by radio and/or telephone.

Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown of the proposed TSC, the staff has determined that the TSC technical data and data systems meets the guidance in Section 2.9 of NUREG-0696 and the applicable guidance of Section 8.2.1. of NUREG-0737, Supplement 1. Therefore, the staff finds that the proposed TSC design meets the applicable requirements of 10 CFR 50.47(b)(8) and Subsection IV.E.8.a(i) of Appendix E to 10 CFR Part 50, with regard to data receipt, storage, processing, and display.

3.10 Records Availability and Management The guidance in Section 2.10 of NUREG-0696 states, in part, that the TSC shall have a complete and up-to-date repository of plant records and procedures at the disposal of TSC personnel to aid in their technical analysis and evaluation of emergency conditions. In particular, up-to-date as-built drawings of the plant systems are needed in the TSC to diagnose sensor data, evaluate data inconsistencies, and identify and counteract fault plant system elements. In addition, the guidance of NUREG-0737, Supplement 1, Section 8.2.1.i, specifies that the TSC should be provided with accurate, complete, and current plant records (drawings, schematic diagrams, etc.) essential for evaluation of the plant under accident conditions.

In LAR Section 3.1.10, Records Availability and Management, the licensee states that the proposed TSC location will maintain the current TSC records availability which includes a complete set of controlled drawings, technical manuals, and other plant records. The proposed TSC will contain controlled copies of selected manuals, procedures, drawings, and other documents as designated by the Surry Nuclear Records Department directives.

Based on its review of the licensees submittal, as supplemented, and the NRC staffs walkdown of the proposed TSC, the NRC staff finds that the proposed TSC design descriptions for the proposed TSC records availability and management conforms to the guidance in Section 2.10 of NUREG-0696 and the applicable guidance of Section 8.2.1 of NUREG-0737, Supplement 1, and therefore, meet the standard of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50, with regard to TSC records availability and management.

3.11 Human Factors In the licensees letter dated, November 18, 2022, Section 3.1.9, Technical Data, Data Systems, and Data System Equipment SC Power Supplies, (ML22322A182) the licensee states that human factors engineering was considered in the design of the TSC data system for both maintenance and operating personnel.

In the licensees submittal dated August 7, 2023, (ML23220A148), the licensee also states that the PCS includes graphical and alphanumeric displays of variables and information related to plant systems, in-plant radiological information, as well as meteorological and offsite radiological information.

Per the LARs audit plan (ML23117A095), a site-walkdown was completed by staff to qualify that comprehensive human factors were applied to the licensees proposed relocation of the TSC related to size, habitability, communications, the technical data system, and records availability and management. The staffs audit summary (ML23173A060) states that staff did not identify any deficiencies related to the licensees proposed TSC relocation and appropriate human factors engineering for operating and maintenance personnel.

3.11.1 Human Factors Technical Conclusion Based on its review of the licensees LAR submittals and the site-walkdown audit of the proposed TSC, NRC staff has determined that the proposed TSC relocation related to human factors design engineering adequately addresses the applicable guidance of NUREG-0737, Supplement 1, Section 8.2.1, and NUREG-0696, Section 2.9. Therefore, the staff finds that the proposed TSC relocation meets the applicable human factors location requirements of 10 CFR 50.47(b)(8) and the applicable location requirements contained in 10 CFR 50, Appendix E, paragraph IV.E.8.a(i).

3.12 Technical Evaluation Summary As discussed above, the NRC staff finds that the proposed relocation of the Surry TSC continues to meet the applicable planning standard requirements of 10 CFR 50.47(b)(8) and the requirements in paragraph IV.E.8.a(i) of Appendix E to 10 CFR Part 50. Given the proposed TSC PCS plant data real-time data acquisition and display capabilities, communication capabilities, layout, and increased size and capabilities of the new facility, the NRC staff finds that the proposed TSC will continue to conform to the TSC guidance of NUREG-0696 and NUREG-0737, Supplement 1. Therefore, the NRC staff finds reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

Therefore, the NRC staff concludes that the licensees proposed relocation of the Surry TSC, as detailed in the licensees amendment and supplements, is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth of Virginia official was notified of the proposed issuance of the amendments on August 24, 2023. On August 24, 2023, the State official confirmed that the Commonwealth of Virginia had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements 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 amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on January 24, 2023, 88 FR 4216, and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Ken Mott, NSIR Michael Norris, NSIR Sean Meighan, NRR Jason White, NRR John Klos, NRR Date: November 7, 2023

ML23242A229

• by email NRR-106 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA* NSIR/DPR/RLB/A(BC)* NRR/DRA/ARCB/BC*

NAME JKlos KGoldstein NDiFrancesco KHsueh DATE 8/21/2023 8/31/2023 8/22/2023 8/21/2023 OFFICE NRR/DEX/EXHB/(A)BC* NRR/DRO/IOLB/(A)BC* OGC*

DORL/LPL2-1/BC NAME NTiruneh KMartin STurk MMarkley DATE 8/15/2023 9/6/2023 9/21/2023 11/7/2023 OFFICE NRR/DORL/D NRR/D NRR/DORL/LPL2-1/PM NAME BPham (VCusumano for)

AVeil (MKing for)

JKlos DATE 10/19/2023 11/06/2023 11/07/2023