ML24005A004

Issuance of Amendment Nos. 296 and 279, Regarding Revised Emergency Plan for Relocation of the Technical Support Center

ML24005A004
Person / Time
Site:	North Anna
Issue date:	03/18/2024
From:	Geoffrey Miller Plant Licensing Branch II
To:	Carr E Southern Nuclear Operating Co
Miller G
References
EPID L 2023 LLA 0006
Download: ML24005A004 (39)

Text

March 18, 2024 Mr. Eric S. Carr, President

- Nuclear Operations and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.

Glen Allen, VA 23060-6711

SUBJECT:

NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENT NOS. 296 AND 279, REGARDING REVISED EMERGENCY PLAN FOR RELOCATION OF THE TECHNICAL SUPPORT CENTER (EPID L-2023-LLA-0006)

Dear Eric Carr:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 296 to Renewed Facility Operating License No. NPF-4 and Amendment No. 279 to Renewed Facility Operating License No. NPF-7 for the North Anna Power Station, Unit Nos. 1 and 2, respectively. The amendments revise the license and Emergency Plan in response to your application dated January 13, 2023, as supplemented by letters dated June 27 and December 11, 2023, and February 9, 2024.

These amendments revise the North Anna Power Station (NAPS) Emergency Plan to allow the relocation of the Technical Support Center (TSC) from its current location adjacent to the Main Control Room (MCR) to the building previously used 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 ed.miller@nrc.gov, or 301-415-2481.

Sincerely,

/RA/

G. Edward Miller, Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-338 and 50-339

Enclosures:

1. Amendment No. 296 to NPF-4

2. Amendment No. 279 to NPF-7

3. Safety Evaluation cc: Listserv

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 296 Renewed License No. NPF-4

1.

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

A.

The application for amendment by Virginia Electric and Power Company et al.,

(the licensee) dated January 13, 2023, as supplemented by letters dated June 27 and December 11, 2023, and February 9, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; 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. 296, Renewed Facility Operating License No. NPF-4 is hereby amended to authorize revision to the Emergency Plan for the North Anna Power Station, Unit 1, as set forth in Virginia Electric and Power Company (Dominion Energy Virginia) application dated January 13, 2023, as supplemented by letters dated June 27 and December 11, 2023, and February 9, 2024, and evaluated in the NRC staffs safety evaluation dated MONTH DAY, 2024.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Andrea D. Veil, Director Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-4 Date of Issuance: March 18, 2024 Michael F.

King Digitally signed by Michael F. King Date: 2024.03.18 13:51:20 -04'00'

VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 279 Renewed License No. NPF-7

1.

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

A.

The application for amendment by Virginia Electric and Power Company et al.,

(the licensee) dated January 13, 2023, as supplemented by letters dated June 27 and December 11, 2023, and February 9, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; 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. 279, Renewed Facility Operating License No. NPF-4 is hereby amended to authorize revision to the Emergency Plan for the North Anna Power Station, Unit 1, as set forth in Virginia Electric and Power Company (Dominion Energy Virginia) application dated January 13, 2023, as supplemented by letters dated June 27, and December 11, 2023, and February 9, 2024, and evaluated in the NRC staffs safety evaluation dated MONTH DAY, 2024.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Andrea D. Veil, Director Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-7 Date of Issuance: March 18, 2024 Michael F.

King Digitally signed by Michael F. King Date: 2024.03.18 13:51:42 -04'00'

ATTACHMENT NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 LICENSE AMENDMENT NO. 296 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 AND LICENSE AMENDMENT NO. 279 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 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 Insert Pages License License License No. NPF-4, page 3 License No. NPF-4, page 3 License No. NPF-7, page 3 License No. NPF-7, page 3 TSs TSs N/A N/A

NORTH ANNA - UNIT 1 Renewed License NPF-4 Amendment No. 296 (2) Pursuant to the Act and 10 CFR Part 70, VEPCO to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material, without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or component; and (5) Pursuant to the Act and 10 CFR Parts 30 and 70, VEPCO to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

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

(1) Maximum Power Level VEPCO is authorized to operate the North Anna Power Station, Unit No. 1, at reactor core power levels not in excess of 2940 megawatts (thermal).

(2) Technical Specifications Technical Specifications contained in Appendix A, as revised through Amendment No. 296 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

NORTH ANNA - UNIT 2 Renewed License NPF-7 Amendment No. 279 (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to receive possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material, without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or component; and (5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, VEPCO to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations as set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level VEPCO is authorized to operate the facility at steady state reactor core power levels not in excess of 2940 megawatts (thermal).

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 279 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the insurance of the condition or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission:

a. If VEPCO plans to remove or to make significant changes in the normal operation of equipment that controls the amount of radioactivity in effluents from the North Anna Power Station, the

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 296 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-4 AND AMENDMENT NO. 279 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-338 AND 50-339

1.0 INTRODUCTION

By letter dated January 13, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23013A195, as supplemented by letters dated June 27, 2023 (ADAMS Accession No. ML23192A215) and December 11, 2023 (ML23346A097), and February 9, 2024 (ML24043A144), Virginia Electric and Power Company (licensee, Dominion) submitted a license amendment request (LAR) to revise the license and Emergency Plan for North Anna Power Station (North Anna), Units 1 and 2. The proposed change would revise the North Anna Emergency Plan (NAEP) 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). This building is physically connected to the Simulator Building and is on the grounds of the North Anna Training Center.

The licensee states that the reason for the proposed NAEP 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 North Anna 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) and an increased fuel enrichment up to 5 weight-percent Uranium 235 (U-235), which has been applied in calculating radiological doses and habitability at the relocated TSC. This proposed revision to the NAEP was determined 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 March 21, 2023, 88 FR 17038, and there has been no public comment on such finding. The supplements dated June 27 and December 11, 2023, and February 9, 2024, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register.

2.0 REGULATORY EVALUATION

The NRC staff considered the following regulatory criterion and guidance during its evaluation of Dominions proposed TSC relocation LAR and NAEP changes.

2.1 Regulations The regulations in Title 10 of the Code of Federal Regulations (10 CFR) 10 CFR 50.36, Technical specifications, contain the requirements for the content of TSs. The regulations in 10 CFR 50.36(b) require TSs to be derived from the analyses and evaluations included in the safety analysis report and amendments thereto. The regulation in 10 CFR 50.36(c)(4), Design features, requires that TSs include design features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety, and are not described in paragraphs (c)(1), (2), and (3) of 10 CFR 50.36.

Section 50.47(b)(8) of 10 CFR Part 50, requires adequate emergency facilities and equipment to support the emergency response are provided and maintained.

The regulations in 10 CFR 50.67, Accident source term, requires, in part, 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 dated February,1981, Revision 1 (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 Part 50, Appendix E.IV.E.8.

NUREG-0696, 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-0696, Section 2.9, Technical Data and Data System, states, in part, that the TSC displays shall include alphanumeric and/or graphical representations of plant

system variables, in-plant radiological variables, meteorological information, and offsite radiological information.

NUREG-0737, Supplement No. 1 (ML102560009), Clarification of TMI Action Plan Requirements, dated January 1983, Section 8.2, Technical Support Center, provides guidance for licensees and the NRC staff regarding acceptable means for meeting the fundamental TSC requirements, and Section 8.2.1.b, Requirements, states that the TSC will be located within the site PA (Protected Area) 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 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.

Regulatory Guide (RG) 1.101, Emergency 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 alternative source terms (ASTs); 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 Absorption 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 concerning criteria for an onsite meteorological measurements program.

Regulatory Issue Summary (RIS) 2005-02, Clarifying the Process for Making Emergency Plan Changes, Revision 1 (ML100340545), dated April 19, 2011, provides guidance to licensees making changes to their emergency plans.

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 NAEP is Revision 52, dated October 30, 2020. In its submittal dated January 13, 2023, the licensee provided Attachment 2, Marked-Up NAPS Emergency Plan Pages, hereafter referred to as the marked-up NAEP, that show the specific wording changes and revisions to the current NAEP to reflect the proposed TSC relocation implementation. In marked up NAPS Emergency Plan Section 7.0, Emergency Facilities and Equipment, the NAEP 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 Technical Support Center (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 the proposed change will provide plant management and technical support to plant operations personnel during emergency conditions; relieve reactor operators of peripheral duties and communications not directly related to reactor system manipulations; prevent congestion in the control room; and perform TSC functions as described in NUREG-0696, NUREG-0737 Supplement 1, and Revision 52 of the NAEP.

The marked up NAEP defines the TSC as the central control center for the onsite emergency response organization after shift augmentation and marked up NAEP 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 NAPS TSC facilities. As discussed in the summary (ML23289A207) for the audit that was completed on October 24, 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.

3.2 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.

The NRC staff also evaluated various factors in determining the appropriateness, acceptability, and whether to provide flexibility to the licensee, relating to a -minute walking time between the relocated TSC and the control room. In its review, the NRC staff found that the advances in communication technologies since NUREG-0696 was published in 1981 compensates for the increase in travel time at North Anna in excess of the two-minute travel time.

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 Corporate Emergency Response Center (CERC). In addition, marked up NAEP 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.

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/Station Emergency Manager in the MCR and the Station Emergency Manager 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 emergency plan drills and exercises using the North Anna 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. Section 7.1.3 of the marked up NAEP 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 North Anna PA boundary, that this new location will be greater than a 2-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. During the

staffs walkthrough audit of the proposed TSC, the staffs walk time between the MCR and the proposed TSC was approximately 8 minutes. In LAR Section 3.1.3, Staffing and Training, the licensee states that the travel time walking from the Technical Services Building to the current TSC is approximately 6 minutes, whereas the travel time from the Technical Services Building to the proposed TSC location is 5-6 minutes and does not require traversing through security facilities as currently required.

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 obviates the need for TSC personnel to be physically present in the MCR to obtain data.

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 North Anna 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 approximately 8 minute walk time from the proposed TSC to the MCR such that relocation of the TSC to the new proposed location outside the North Anna 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.

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., Protective Equipment, states, in part, 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 North Anna 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 stated that the proposed NAEP revisions to support the TSC relocation maintains the existing TSC emergency response organization (ERO) staffing levels, TSC ERO augmentation response times, and TSC ERO training such that there is no impact on the timeliness of TSC activation or transfer of responsibilities from the MCR to the TSC. As such, the licensee is not proposing any changes to ERO staffing as currently described in the NAEP and the training of the ERO will continue to be maintained as currently described in the NAEP. The licensee further states that the new location also provides for improved off-hours staffing capability as responders would report directly to the TSC without having to pass through security access controls for entry into the PA.

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 NAPS Emergency Plan. 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 stated that the proposed TSC provides approximately 1,900 square feet of operating space for TSC personnel which assures a minimum of 75 square feet of working space for 25 persons. The proposed TSC:

Replicates the layout of the current TSC facility.

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

Will include 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 North Anna 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 its response to RAI 1 of its letter dated December 11, 2023, the licensee stated that, The proposed TSC will be located in the building that formerly housed the LEOF. The LEOF was constructed in accordance with the 1981 Building Officials Code Administration (BOCA) code which was the applicable Commonwealth of Virginia building code at the time of construction. Earthquake loads were addressed in accordance with Section 916 of the 1981 BOCA Code. Consequently, the North Anna LEOF is considered a well-engineered structure with adequate capability to withstand earthquakes.

In Section 3.1.5, Structure, of its letter dated January 13, 2023, the licensee stated, in part, that, The proposed LEOF [proposed TSC] structure is made of a 12-inch thick reinforced concrete roof, 12-inch exterior masonry walls with horizontal joint reinforcement, and a 5-inch slab on grade. A review of this design determined that, due to the thickness and reinforcement of the walls and slab floor, the structure exceeds the UBC requirements in place at the time of LEOF construction and can withstand the applicable loading. This as-designed building will withstand the 100-year wind speeds as described in the UFSAR (Reference 22).

In its response to RAI 2 of its letter dated December 11, 2023, the licensee stated that, The maximum 100-year flood in the vicinity of North Anna Power Station (NAPS) is 254.2 feet as shown on FEMAs [Federal Emergency Management Agency's] 100-year National Flood Hazard Layer (NFHL) Viewer map. Key data are as follows:

The proposed TSC floor level is 270 feet.

The NAPS Units 1 and 2 Updated Final Safety Analysis Report (UFSAR) identifies the maximum flood level for the North Anna site as 267.3 feet.

The FEMA 100-year flood level is 254.2 feet.

Therefore, the TSC floor level is located above the maximum flood level for the NAPS site as well as the FEMA 100-year flood level.

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, the 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 stated, in part, that, The construction of this building [proposed TSC] consists of 12-inch thick reinforced concrete roof, 12-inch thick exterior masonry walls with horizontal joint reinforcement, and a 5-inch slab on grade. The TSC ventilation system filter bank will be located in the penthouse of the new facility in an unoccupied space that is separated from ground level by a 12-inch thick concrete slab. An existing hatch between the penthouse and the ground level of the new TSC (occupied space) will be permanently sealed with a 12-inch concrete plug to prevent radiation exposure to the occupied TSC space below.

As described in Section 3.1.6 and Sections 3.2.1.and 3.2.2 of this SE, the licensee stated 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.

The licensee further stated that protective clothing to support TSC personnel will be maintained in the TSC emergency kit. Additionally, in its response to RAI 3 in its letter dated December 11, 2023, the licensee stated that, In accordance with the revised NAPS emergency plan implementing procedures, potassium iodide tablets will be maintained in the proposed 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, and maintaining protective clothing and potassium iodine addresses the intent of NUREG-0696, Section 2.6 guidance to provide personnel radioactive protection 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[µCi]/cc [cubic centimeter].

In LAR Section 3.1.6, the licensee stated that the proposed TSC will be provided with radiation monitors, as follows:

The RMS consists of a Mirion radiation monitor to detect airborne radioactivity, and two Mirion DRM-2 general area radiation monitors. The Mirian radiation monitor will include a particulate, an iodine, and a noble gas detector and will be able to distinguish the presence or absence of radioiodines at concentrations as low as 10 -7 ci/cc. The monitor will be located in the Dose Assessment Room and will continuously sample the facility atmosphere from locations throughout the TSC and provide an audible alarm to alert TSC personnel of adverse conditions. The two Mirion DRM-2 general area radiation monitors will be wall mounted at separate locations on 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 a request for review and approval of a revision to the NAPS Units 1 and 2 Emergency Plan which would move the TSC to a new location outside of the Protected Area (PA). The relocation of the TSC requires a recalculation of the /Q values used in the radiological dose calculation that demonstrates the habitability of the new TSC. In its letter dated June 27, 2023 (Accession No. ML23192A215), the licensee provided meteorological data and the inputs and assumptions used in the atmospheric dispersion analysis.

In its letter dated June 27, 2023, the licensee identified the following release pathways to the TSC for the atmospheric dispersion analysis: Unit 1 Containment, Vent Stack A, Vent Stack B, Unit 1 Blowout Panel, Unit 1 RWST (Refueling Water Storage Tank), Unit 2 Equipment Hatch.

The licensee developed new /Q values for each release pathway.

3.6.1.2.1 Meteorology Data In its letter dated June 27, 2023, the licensee provided information regarding the atmospheric dispersion analysis performed for the relocated TSC. The enclosure of the supplement contained hourly onsite meteorological data used in the analysis. 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 lower wind instrument height of 10.04m (meters) and upper instrument height of 48.43m levels of the meteorological tower.

The NRC staff conducted a detailed review related to the acceptability and representativeness of the onsite hourly meteorological dataset and found it to be consistent with the guidance outlined in RG 1.23. Based on the above, the NRC staff considers the onsite meteorological dataset 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 the LAR, the licensee used the computer code ARCON96 to estimate /Q values for the TSC 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.

of the June 27, 2023, supplement to the LAR includes Table 5-1, Summary of Changes for TSC Relocation X/Q Calculation. The table includes the meteorological data period, the lower and upper height measurements, and units of wind speed used in the analysis.

Table 5-1 also describes the source input values of vertical velocity, stack flow, stack radius, and diffusion coefficients. Table 5-1 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 the licensees atmospheric dispersion estimates by running the ARCON96 computer model and obtaining similar results. Both the 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 independent 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 TSC Atmospheric Dispersion Analysis 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 staff found that the licensee used methods consistent with regulatory guidance identified in Section 2.0 of this safety evaluation. The licensee used onsite meteorological data that complied with the guidance of RG 1.23. The inputs and assumptions used to calculate the Technical Support Center /Q values were also consistent with the guidance of RG 1.194.

Therefore, on the basis of this review of the atmospheric dispersion analysis, 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 June 27, 2023 (ML23192A215), the licensee provided additional information related to its loss-of-coolant accident (LOCA) dose analysis and a change to eliminate the CO2 fire suppression system from the High Efficiency Gas Adsorption (HEGA) system. The licensee used the design-basis LOCA dose consequence analysis 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 LOCA dose consequence analysis is expected to be based the limiting major accident or event, resulting in dose consequences that would not be exceeded by the doses from any other credible accident. For NAPS, an updated accident source term is used in its design basis radiological dose analysis to meet the dose criteria in 10 CFR 50.67. 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 to compensate for known uncertainties in accident progression, activity product transport, and atmospheric dispersion.

Additionally, this LAR included a change in the radionuclide inventory and dose conversion files reflect a licensing basis change from SCALE 4.4a to SCALE 6.2.3 for determination of the core source term.1 In its letter dated June 27, 2023, the licensee stated, in part, that, The updated core source term incorporated a wider range of enrichments and a higher core average burnup within the constraints of the rated thermal power plus calorimetric uncertainty. Having been part of the up-conversion of RADTRAD-NAI models, these files may also be used as a starting point for subsequent RADTRAD-NAI Version 1.3 analyses.

The NRC staff notes that the use of a higher batch average burnup of 59.8 GWd/MTU (gigawatt days/metric ton unit of uranium) in the analysis is more limiting than the previous analysis of record (AOR) relative to allowable plant burnup limits. As discussed in Section 3.6.2.1.1, plant operation would be required to remain within the authorized limit of 62 GWd/MTU unless an increase is approved by a subsequent LAR. Likewise, the proposed increased enrichment of 4.2-5.0 weight-percent enrichment of U-235 is above the AST analysis input value of 4.4-4.55 weight-percent U-235 enrichment but is within the limit of 5.0 weight-percent U-235 approved in amendment Nos. 279/262 dated July 27, 2018 (ML18180A197), and in current TS 4.3,1, Criticality.

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 provided an updated core source term for use in this LOCA dose analysis. 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 1 The Licensee is responsible for reflecting the use of SCALE 6.2.3 in the next appropriate UFSAR update in accordance with 10 CFR 50.71

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 June 27, 2023, Attachment 1, and summarized in Table 4-1, Summary of TSC Dose Results, the licensee stated that the LOCA dose consequence calculation analysis includes dose contributions from the following potential radioactive material release pathways.

Containment leakage Emergency core cooling leakage Refueling Water Storage Tank back leakage.

Containment direct shine Cloud shine TSC ventilation system filter shine.

The licensees current licensing basis LOCA source term analysis was approved via license amendments Nos. 240 and 221 for North Anna Power Station Units 1 and 2 issued on June 15, 2005 (ML051590510), and was later analyzed for higher enrichment in amendment Nos.

279/262 on July 27, 2018 (ML18180A197).

3.6.2.1.1 TSC Maximum hypothetical accident (MHA)-LOCA Analysis Inputs: Burnup, Enrichment, and Core Inventory In Attachment 1 of its letter dated June 27, 2023, the licensee provided a summary of the MHA-LOCA dose calculation. Table 5-2, Summary of Changes for TSC Relocation LOCA Dose, compared the inputs and assumptions in the updated MHA-LOCA dose analysis and the previous AOR analysis. As detailed in the table, the licensee made changes to several fuel-related parameters that are important for the dose calculation in terms of fuel burnup, enrichment, and updated core inventory. These particular changes are detailed in Table 5-2 of to the supplement and are assessed below.

Concerning burnup, the AOR for the end of cycle (EOC) core average burnup was assumed to be 38.9 GWd/MTU) and the batch average burnup was assumed to be up to 51.4 GWd/MTU.

The proposed LAR would increase these analyses values to 42.0 GWd/MTU for core average burnup and up to 59.8 GWd/MTU for batch average burnup.

The NRC staff notes that a batch average burnup of 59.8 GWd/MTU would be expected to lead to a rod-average burnup of over 62 GWd/MTU, and thus would likely exceed the burnup limits of applicability for several current analysis methods referenced in North Anna Technical Specification (TS) (Unit 1 ML052990145, Unit 2 ML052990147, Section 5.6.5, Core Operating Limits Report (COLR), item 5.6.5.b). This batch average burnup analysis provides a limiting assumption for TSC MHA-LOCA dose analysis, but does not authorize plant operation to reach rod-average burnups over 62 GWd/MTU; rather, that value is a limiting hypothetical assumption in the dose analysis. Other controls would prevent actual operation from resulting in such burnup levels (e.g., the burnup limit in an NRC-approved cladding material topical report referenced in the COLR). Moreover, the NRC would have to approve any proposed increase in burnup limits before a plant could operate above 62 GWd/MTU. Regarding the impact of increased burnup levels on the dose analysis, burnup is expected to have little effect on the MHA-LOCA source term release fractions (ML20126G376), however the radionuclide inventory would be greater at higher burnup levels. Since higher burnup levels would lead to a larger radionuclide inventory, assuming a higher burnup, provides a more conservative dose analysis

for evaluating the TSC relocation. Accordingly, the NRC staffs finds the assumed burnup levels assumed for the TSC MHA-LOCA dose analysis to be a more conservative prediction of radionuclide inventory than the AOR and are, therefore, acceptable.

Similarly, Table 5-2 also states that the uranium U-235 enrichment assumed for the updated TSC MHA-LOCA dose analysis ranges from 4.2-5.0 weight-percent (wt.%). This is a change from the existing AOR, which states that an enrichment range of 4.4-4.55 wt% was used in the MHA-LOCA analysis for AST implementation. As noted above. NAPS is currently licensed for fuel with up to 5.0 wt.% U-235. As such, the updated analysis is more representative of the design basis and more conservative than the previous analysis. Increases in enrichment levels typically result in increased dose, so it is important for the analysis to consider the upper range of enrichments permitted; accordingly, the updated enrichment range is more appropriate than the enrichment levels assumed in the previous analysis. Based on the above, the NRC staff finds the assumption of increased fuel enrichments in the updated calculation to be acceptable.

In Table 5-2, the licensee stated that 110 isotopes were considered in the TSC MHA-LOCA core inventory. This is an increase from the 94 isotopes considered in the AOR core inventory for implementation of the AST. Specifically, in the updated analysis, 20 isotopes were added to consideration in the core inventory and 4 isotopes (silver (Ag)-110m, cesium (Cs)-135, krypton (Kr)-89, and xenon (Xe)-137)) were removed. Cs-135 has a half-life of 2.3 million years, so it is essentially stable, and would thus have a negligible impact on dose. On the other hand, Kr-89 and Xe-137 do not have a significant importance to the dose analysis vs. their decay products.

Because the radioactive decay products of these isotopes (Kr-89 and Xe-137) are considered in the licensees analysis, the NRC staff finds that the effects of Kr-89 and Xe-137 were effectively considered in the analysis. Lastly, Ag-110m is understood to generally not be a significant dose contributor and as such it is not listed as an element that should be considered in design basis analyses in RG 1.183, Section 3.4, Radionuclide Composition, Table 5, Radionuclide Groups, Therefore, the isotopes that were removed are not expected to be major contributors to the overall dose and their omission is consistent with other plants NRC-approved MHA-LOCA dose analyses (for example, see Table 11.1-1 of the Salem Generating Station Updated Final Safety Analysis Report (ML22298A056) or Table 14.A.1-1 of the D.C. Cook Unit 1 Updated Final Safety Analysis Report, (ML22340A131). In sum, the NRC staff finds the removal of Ag-110m, Cs-135, Kr-89, and Xe-137 to be acceptable because (1) they are not expected to have a significant impact on the radiological consequences, (2) any significant decay products are considered, as appropriate, and (3) there is precedent for exclusion of these radionuclides.

Further, the NRC staff finds the additional radionuclides considered in the updated core inventory used in the TSC MHA-LOCA analysis to be acceptable because the consideration of these radionuclides adds conservatism to the dose analysis.

The NRC staffs review of the updated core inventory only applies to the TSC MHA-LOCA dose analysis and did not consider other analyses (e.g., non-LOCA dose analyses and demonstrating adherence with the exclusionary boundary, low population zone, and control room dose limits).

Accordingly, the NRC staffs approval of the updated core inventory in this amendment is only applicable to the TSC MHA-LOCA analysis and is not applicable to any other design basis accident analysis.

3.6.2.1.2 Technical Summary of Burnup, Enrichment, and Core Inventory changes to MHA-LOCA Analysis As discussed in the evaluation above, the NRC staff determined that the fuel burnup and core inventory assumptions utilized in the North Anna TSC MHA-LOCA dose calculation are acceptable and that the proposed design meets the guidance of RG 1.183, Revision 0.

Therefore, the staff finds that the dose consequence analysis for the proposed TSC meets the criteria in 10 CFR 50.67 and the applicable requirements of Appendix A, Criterion 19, to 10 CFR Part 50.

3.6.2.2 Infiltration of Airborne Radioactivity Using the updated core inventory and current licensing basis transport assumptions for the LOCA dose consequence analysis, the licensee in its letter dated June 27, 2023, Table 4-1:

Summary of TSC Dose Results, identified and evaluated the TEDE contributions to TSC occupants from the following constituents:

Containment Leakage ECCS Leakage - Emergency Core Cooling System leakage RWST Leakage - Refueling Water Storage Tank leakage Containment Direct Shine - Direct dose from the source term in containment.

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

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

Filter Shine - Direct shine dose from the TSC filtration system.

In its letter dated June 27, 2023, Attachment 1, Table 5-2, Summary of Changes for TSC Relocation LOCA Dose, the licensee provided the initial conditions, inputs and assumptions used to determine TSC TEDE dose. The licensee states, in part, 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.

In Table 5-2, the licensee provided information associated with TSC filtration system performance values. The makeup flow rate is specified as 405 cubic feet per minute (cfm) to a maximum of 1,100 cfm unfiltered operational flow rate before isolation. For the TSC volume of 40,800 ft3, the after isolation, minimum pressurization makeup flow rate of 365 cfm will result in more than 0.5 volume changes per hour. The volume change stated above by the licensee, per Standard Review Plan 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; this 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.

In its letter dated June 27, 2023, Attachment 1, the licensee stated 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 flow rate of 1,000 cfm. The HEPA filter removes particulate radioactive air contaminants and the HEGA filter removes remaining iodine compounds. RG 1.140 (ML16070A277) provides

guidance regarding acceptable filter efficiencies of 99, 95, and 95 percent for particulates, elemental iodine, and organic iodide, respectively. In its letter dated June 27, 2023, the licensee assigns an efficiency of 98 percent for particulates in Table 5.2, which is equal to the current licensing basis (CLB) value.

In its letter dated June 27, 2023, the licensee provides two dose consequence analyses, one of which assumes that the TSC filtered makeup pressurization system will be activated 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 a SI signal and, therefore, it performed the analysis of manual isolation at 60 min.

3.6.2.3 Direct Shine Dose from the TSC Filtration System In its letter dated June 27, 2023, the licensee evaluates 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 radionuclide buildup on the TSC HVAC filters and is documented in the supplement dated June 27, 2023, Table 4-1, Summary of TSC Dose Results.

The NRC staff reviewed the initial conditions, inputs, and assumptions used in the MCNP Transport Code to assess dose from TSC filters and determined them to be appropriate and sufficient to meet regulatory requirements.

3.6.2.4 Direct Shine Dose from the External Cloud of Airborne Radioactive Materials, Containment Direct Shine, and Containment Skyshine In its letter dated June 27,2023, the licensee stated that the increase in the distance to the proposed TSC from the previous location resulted in reduced accident dose values for each of the contributors. Table 4-1 indicates the total dose in the current TSC with AST is 2.33 rem TEDE and is reduced to 0.81 rem TEDE for 60-minute isolation, and 0.38 rem TEDE in the SI-based isolation of the TSC relocation analysis. The NRC staff reviewed the initial conditions, inputs and assumptions for the shine calculations associated with this LAR and determined them to be appropriate and sufficient to meet regulatory requirements.

In its letter dated June 27, 2023 (ML23192A215), the licensee noted that due to the increase in TSC distance from accident release points, calculated accident dose consequences at the TSC were considerably lower.

The NRC staff reviewed the calculations used to determine the direct shine dose from the external cloud of airborne radioactive materials, containment direct shine, and containment skyshine and determined that the initial conditions, inputs, and assumptions used in the calculations associated with this LAR are reasonable and appropriate. The NRC staff compared the dose estimates provided in the LAR to the applicable acceptance criteria and to the results estimated by the NRC staff in its independent confirmatory calculations, including those for

shine components utilizing MicroShield, Version 13. The NRC staffs estimated results are similar to those provided by the licensee, and these results are all within regulatory limits.

3.6.2.4.1 Summary of TSC Dose Results TSC Dose [rem TEDE]

LOCA Dose Component TSC Relocation

[CLB] AST TSC Dose SI-Based Isolation 60-min Isolation Dose Containment Leakage 0.0262 0.425 0.8 ECCS Leakage 0.0114 0.0491 0.12 RWST Leakage 0.0007 0.0007 0.0057 Containment Direct Shine 0.0019 0.0019 0.047 Containment Skyshine 0.266 0.266 0.381 Cloudshine 0.0368 0.0368 0.6165 Filter Shine 0.0311 0.0311 0.16 Containment Shine through Main Steam Line Penetrations 0.005 SI Piping under Main Steam Valve House and Quench Spray (QS) Pump House 0.194 Hydrogen Re-combiner Vault 0.004 Total 0.3741 0.8106 2.3332 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 Elimination of CO2 Fire Suppression System In its letter dated June 27, 2023, the licensee stated, in part, that a carbon dioxide (CO2) fire suppression system would be installed for the HEGA filter installed in the relocated TSCs ventilation system. Following the submittal of the LAR, it was determined the CO2 fire suppression system is not required due to the availability of other fire protection measures and equipment being installed. Consequently, the CO2 fire suppression system has been removed from the relocated TSC design modification. Further, in Attachment 4 to its letter dated June 27, 2023, the licensee stated, in part that, that a CO2 fire suppression system is not necessary for the relocated TSC since the TSC: 1) is not required for safe shutdown, 2) contains other fire protection measures, and 3) is unlikely to experience a HEGA filter fire.

Additionally, as the new TSC does not contain any safe shutdown equipment, it can be evacuated if the space becomes unfit for human habitation. In the unlikely event TSC occupants are exposed to radioactive contaminants during a HEGA filter fire (or after it has been extinguished), the area can be evacuated without risk to safe shutdown of the plant.

The licensee concluded that a designated CO2 fire suppression system for the TSC HEGA activated carbon filter is not required. The probability of ignition of the HEGA filter is minimal based on the insignificant radiological decay heat load and sufficient emergency operation

airflow. In the unlikely event auto-ignition of the charcoal filter were to occur, heat and/or cross-zone smoke detection initiates an alarm and deenergizes the emergency fan or air handling unit and closes all motor operated dampers to isolate the filter fire from the remainder of the TSC. A filter fire is unlikely to spread into the TSC general area due to the installed pre-action sprinkler system, and the existing separation between the Penthouse and the remainder of the TSC.

The NRC staff confirmed that the proposed changes to the NAPS Emergency Plan state that the Control Room is the alternate location for the TSC if the TSC became uninhabitable.

Additionally, the NAPS emergency plan implementing procedure (EPIP) EPIP-3.02, Activation of Technical Support Center, Revision 36, effective August 5, 2014 (ML14230A109) provides guidance if the TSC has to be evacuated.

The NRC staff also reviewed NUREG-0696 and determined that guidance is not specified for the use of different fire suppression systems in the emergency response facilities. The NRC staff reviewed the proposed change to the fire suppression system and determined that it is acceptable. Further, the NRC staff concludes that in the event of a LOCA, the absence of the CO2 fire suppression system would not alter the capability of the relocated TSC from meeting the dose consequence analysis assumptions.

3.6.2.6 TSC Dose Consequence Analysis Conclusion The results of the licensees updated analysis for the relocated TSC indicate that TSC personnel will be protected from radiological hazards, including direct radiation and airborne radioactivity from in-plant sources (summarized in the table in Section 3.6.2.4 above) under accident conditions, are within regulatory limits (radiation exposures shall not exceed 5 rem TEDE for the duration of the postulated accident).

The NRC staff used RADTRAD version 5.0.3 and MicroShield Version 13 to perform independent confirmatory calculations of the licensees analysis. These confirmatory calculations utilized the LOCA analysis initial conditions, inputs and assumptions provided in the LAR and CLB documents. The staffs calculations obtained results similar to those provided in the application, all of which were within acceptance criteria and regulatory limits for North Anna.

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

3.7 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, Operations Support Center (OSC), onsite personnel, mobile monitoring teams, CERC, Offsite Response Organizations, and the NRC.

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 NAEP, 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 at North Anna, Units 1 and 2, and essential locations remote from the station, during normal operation and under accident conditions. Communication systems vital to NAPS Units 1 and 2 operation and safety are designed so that failure of one component would not impair the reliability of the total communications system. This is accomplished within the Station by using diverse systems and designated personnel.

The current TSC has access to the sound powered telephone system. Because the sound powered telephone system is not capable of communicating over long distances, and the TSC will be located outside the PA, the TSC will no longer rely on a sound powered telephone system for back-up emergency communications. The licensee states that elimination of the sound powered telephone system in the TSC is acceptable because sound powered phones are typically used by operations for startup, shutdown, testing, and maintenance, as well as fire response. They are not typically used for TSC emergency response activities. Other communications links such as public address, private branch exchange (PBX), radios, etc., are used by the TSC to communicate with personnel in the field. The sound powered phone system was considered a back-up system to be used in the event one of the other systems was unavailable. As such, the NRC staff finds that the absence of a sound powered telephone system in the relocated TSC is acceptable.

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 NRC 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 North Anna 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 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 300kW/375KVA diesel generator and will 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 Plant Computer System (PCS) data capabilities will remain unchanged by the proposed TSC relocation. The proposed TSC will use PCS workstations that will be connected to the North Anna 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, Technical Data, Data Systems, and Data System Equipment [Support Center]SC Power Supplies, 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. Therefore, the NRC staff concludes that the relocated TSCs instrumentation, data system equipment, and power supply systems meet the requirements of 10 CFR 50.47(b)(8) and paragraph IV.E.8.a(i) of Appendix E to 10 FR Part 50, with regard to providing reliable equipment to gather, store, and display data needed in the TSC and are 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 are displayed through the PCS. In addition, meteorological data and effluent monitoring data are provided from the PCS directly to the Meteorological Information and Dose Assessment System for use in making offsite dose projections.

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 meet 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 North Anna 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 design descriptions for the proposed TSC records availability and management conform to the guidance in Section 2.10 of NUREG-0696 and applicable guidance in Section 8.2.1 of NUREG-0737, Supplement 1. Therefore, the NRC staff concludes that the licensees instrumentation, data system equipment, and power supplies meet the 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 TSC records availability and management.

3.11 Human Factors The guidance in Section 2.9, Technical Data and Systems, of NUREG-0696 states, in part, that the TSC displays shall include alphanumeric and/or graphical representations of plant systems variables, in-plant radiological variables, meteorological information, and offsite radiological information. The NRC staff conducted an audit as documented in the summary dated October 23, 2023 (ML23289A207). As part of the NRC staffs review, a Request for Additional Information (RAI) was generated on November 13, 2023 (ML23318A117). The licensee stated in its supplemental RAI response of December 11, 2023 (ML23346A097) that meteorological information and on-site radiological information is displayed through the PCS and is transferred to MIDAS which is available in the current TSC. In the proposed TSC, the licensee states that it will maintain the same capability as the current TSC; the current TSC includes the display of plant systems variables, in-plant radiological variables, meteorological information, and offsite radiological information.

The guidance in Section 8.2.1, Technical Support Center, of NUREG-0737, states, in part, that the TSC facility should be sufficient to accommodate NRC and licensee personnel, equipment and documentation, and designed considering good human factors engineering principles. In the licensees letter dated January 13, 2023, Section 3.1.9, Technical Data, Data Systems, and Data System Equipment, (ML23013A195) the licensee states that human factors engineering was considered in the design of the TSC technical data system for both operating personnel and maintenance personnel. In addition, the licensee states, in sub-paragraph C of Section 3.1.9, that a detailed human factors evaluation of the proposed TSC was performed in accordance with NUREG-0711, Human Factors Engineering Program Review Model, and that the licensee determined it to be acceptable.

In its review, the NRC staff evaluated the following human factors elements, as discussed in the referenced SE sections; Staffing and Training (Section 3.3), Size (Section 3.4), Habitability (Section 3.6), Communications (Section 3.7), and Instrumentation (Section 3.8). These were found to meet the guidance of NUREG-0696 and NUREG-0737 based on the licensees docketed information, which addressed those human factor elements. Therefore, the proposed change to the new TSC location supports the overall application of good human factors engineering principles for the proposed TSC, in accordance with 10 CFR 50.47(b)(8) and paragraph IV.E.8.a(i) of Appendix E of 10 CFR Part 50.

Per the LARs audit plan of August 5, 2023 (ML23223A040), a site-walkdown was completed by the NRC staff to examine the licensees non-docketed information, to gain a better understanding of the LAR, to verify information, and to identify information that may require docketing to support the basis of the NRCs staff licensing decision. The NRC staffs site audit did not identify any findings adverse to the issuance of the requested amendment.

3.11.1 Human Factors Technical 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 good human factors engineering principles were applied in the design of the proposed TSC, including staffing, training, size, habitability, communications, instrumentation, and the technical data system. Based on the above, the NRC staff concludes that the human factors engineering descriptions for the proposed TSC meet the guidance in Section 2.9 of NUREG-0696 and Section 8.2.1 of NUREG-0737, and, therefore, conform with 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 ensuring good human factors engineering principles are used in the design of the proposed TSC.

3.13 Technical Evaluation Summary As discussed above, the NRC staff finds that the proposed relocation of the NAPS TSC meets the applicable 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 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 at North Anna Units 1 and 2. Therefore, the NRC staff concludes that the licensees proposed relocation of the North Anna 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 and, on January 4, 2024, 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, as published in the Federal Register on March 21, 2023 (88 FR 17038), 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: M. Norris, NSIR K. Mott, NSIR S. Meighan, NRR J. White, NRR J. Messina, NRR Date: March 18, 2024

ML24005A004

• Via SE Input 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 1/3/2024 01/17/2024 12/19/2023 11/6/2023 OFFICE NRR/DEX/EXHB/BC NRR/DRO/IOLB/(A)BC* NRR/DSS/SFNB/BC*

OGC (NLO)

NAME BHayes BGreen SKrepel STurk DATE 1/3/2024 1/4/2024 9/8/2023 3/2/2024 OFFICE DORL/LPL2-1/BC NRR/DORL/D NRR/D NRR/DORL/LPL2-1/PM NAME MMarkley BPham (JHeisserer for) AVeil (MKing for)

GEMiller DATE 3/1/2024 3/6/2024 3/18/2024 3/18/2024