ML23136B139

Issuance of Amendment No. 313 Administrative Changes to Subsequent Renewed Operating Licenses and Technical Specifications

ML23136B139
Person / Time
Site:	Surry
Issue date:	06/29/2023
From:	John Klos Plant Licensing Branch II
To:	Stoddard D Virginia Electric & Power Co (VEPCO)
Klos, J
References
EPID L-2022-LLA-0091
Download: ML23136B139 (32)

Text

June 29, 2023

Mr. Daniel G. Stoddard Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.

Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENT NOS. 313 AND 313 RE: ADMINISTRA TIVE CHANGES TO SUBSEQUENT RENEWED OPERATING LICENSES AND TECHNICAL SPECIFICATIONS (EPID L-2022-LLA-0091)

Dear Mr. Stoddard:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 313 to Subsequent Renewed Facility Operating License No. DPR-32 and Amendment No. 313 to Subsequent Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively. The amendments revise the technical specifications (TS) and delete expired license conditions in response to your application dated June 20, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22171A013), as supplemented by letter dated November 18, 2022 (ML22322A198).

The proposed amendments would revise the Surry subsequent renewed facility operating license and TS to make a number of editorial changes and corrections, including removal of the TS and license condition associated with a one-time plant modification.

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

Sincerely,

/RA/

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

Enclosures:

1. Amendment No. 313 to DPR-32

2. Amendment No. 313 to DPR-37

3. Safety Evaluation cc: Listserv VIRGINIA ELECTRIC AND POWER COMPANY

DOCKET NO. 50-280

SURRY POWER STATION, UNIT NO. 1

AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE

Amendment No. 313 Subsequent Renewed License No. DPR-32

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

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated June 20, 2022, as supplemented by letter dated November 18, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;

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

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-32 is hereby amended to read as follows:

(B) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 313, are hereby incorporated in the subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Subsequent Renewed Facility Operating License No. DPR-32 and Technical Specifications

Date of Issuance June 29, 2023

VIRGINIA ELECTRIC AND POWER COMPANY

DOCKET NO. 50-281

SURRY POWER STATION, UNIT NO. 2

AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE

Amendment No. 313 Subsequent Renewed License No. DPR-37

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

A. The application for amendment by Virginia Electric and Power Company (the licensee) dated June 20, 2022, as supplemented by letter dated November 18, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I;

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

C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-37 is hereby amended to read as follows:

(B) Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 313, are hereby incorporated in the subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Michael T. Markley, Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes Subsequent Renewed Facility Operating License No. DPR-37 and Technical Specifications

Date of Issuance June 29, 2023 ATTACHMENT

SURRY POWER STATION, UNIT NOS. 1 AND 2

TO LICENSE AMENDMENT NO. 313

SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32

DOCKET NO. 50-280

AND

TO LICENSE AMENDMENT NO. 313

SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37

DOCKET NO. 50-281

Replace the following pages of the Licenses and the Appendix A Technical Specifications (TS) 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. DPR-32, page 3 License No. DPR-32, page 3 License No. DPR-32, page 6 License No. DPR-32, page 6 License No. DPR-32, page 7 License No. DPR-32, page 7 License No. DPR-32, page 8 ------

License No. DPR-32, page 9 ------

License No. DPR-32, page 10 ------

License No. DPR-32, page 11 ------

License No. DPR-37, page 3 License No. DPR-37, page 3 License No. DPR-37, page 6 License No. DPR-37, page 6 License No. DPR-37, page 7 License No. DPR-37, page 7 License No. DPR-37, page 8 ------

License No. DPR-37, page 9 ------

License No. DPR-37, page 10 ------

License No. DPR-37, page 11 ------

Remove Pages Insert Pages

TS TS 1.0-2 1.0-2 3.10-4 3.10-4 3.10-5 3.10-5 3.10-6 3.10-6

Remove Pages Insert Pages

TS TS 3.16-3 3.16-3 3.16-7 3.16-7 3.16-7a ------

3.16-7b ------

3.23-2 3.23-2 3.23-6 ------

4.9-1 4.9-1 6.1-2 6.1-2 6.7-1 6.7-1

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

A. Maximum Power Level

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

B. Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 313 are hereby incorporated in the subsequent renewed license.

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

C. Reports

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

D. Records

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

E. Deleted by Amendment 65

F. Deleted by Amendment 71

G. Deleted by Amendment 227

H. Deleted by Amendment 227

Surry - Unit 1 Subsequent Renewed License No. DPR-32 Amendment No. 313 (3 ) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders

R. Deleted by Amendment 313.

S. Deleted by Amendment 313.

T. Deleted by Amendment 313.

U. Deleted by Amendment No. 289 V. The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; the Appendix R program to evaluate fire risk; a modified version of the Electric Power Research Institute (EPRI) 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, Tier 1 approach to assess seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Class 3 SSCs and their associated supports; and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in License Amendment No. 301 dated December 8, 2020.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from an Appendix R program fire risk evaluation to a fire probabilistic risk assessment approach.)

Surry - Unit 1Surry - Unit 1 Subsequent Renewed License No. DPR-32 Amendment No. 313 W. Subsequent Renewed License Conditions

(1) The information in the Updated Final Safety Analysis Report (UFSAR) supplement submitted pursuant to 10 CFR 54.21(d), as revised during the subsequent license renewal application review process, and Virginia Electric and Power Company commitments as listed in Appendix A of the Safety Evaluation Report Related to the Subsequent License Renewal of Surry Power Station, Units 1 and 2, dated March 2020, are collectively the Subsequent License Renewal UFSAR Supplement. This Supplement is henceforth part of the UFSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, Virginia Electric and Power Company may make changes to the programs, activities, and commitments described in the Subsequent License Renewal UFSAR Supplement, provided Virginia Electric and Power Company evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, Changes, Tests, and Experiments, and otherwise complies with the requirements in that section.

(2) The Subsequent License Renewal UFSAR Supplement, as defined in subsequent renewed license condition (W)(1) above, describes programs to be implemented and activities to be completed prior to the subsequent period of extended operation, which is the period following the May 25, 2032, expiration of the initial renewed license.

a. Virginia Electric and Power Company shall implement those new programs and enhancements to existing programs no later than 6 months before the subsequent period of extended operation.

b. Virginia Electric and Power Company shall complete those activities by the 6-month date prior to the subsequent period of extended operation or by the end of the last refueling outage before the subsequent period of extended operation, whichever occurs later.

c. Virginia Electric and Power Company shall notify the NRC in writing within 30 days after having accomplished item (2)a above and include the status of those activities that have been or remain to be completed in item (2)b above.

4. This subsequent renewed license is effective as of the date of issuance and shall expire at midnight on May 25, 2052.

FOR THE UNITED STATES NUCLEAR REGULATORY COMMISSION

Signed by Veil, Andrea on 05/04/21 Andrea Veil, Director Office of Nuclear Reactor Regulation

Enclosure:

Appendix A - Technical Specifications for Surry Power Stat ion, Units 1 and 2 Date of Issuance: May 4, 2021

Surry - Unit 1Surry - Unit 1 Subsequent Renewed License No. DPR-32

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

A. Maximum Power Level

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

B. Technical Specifications

The Technical Specifications contained in Appendix A, as revised through Amendment No. 313 are hereby incorporated in this subsequent renewed license.

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

C. Reports

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

D. Records

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

E. Deleted by Amendment 54

F. Deleted by Amendment 59 and Amendment 65

G. Deleted by Amendment 227

H. Deleted by Amendment 227

Surry - Unit 2 Subsequent Renewed License No. DPR-37 Amendment No. 313 (3) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders

R. Deleted by Amendment 313.

S. Deleted by Amendment 313.

T. Deleted by Amendment 313.

U. Deleted by Amendment 289 V. The licensee is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed Safety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) model to evaluate risk associated with internal events, including internal flooding; the Appendix R program to evaluate fire risk; a modified version of the Electric Power Research Institute (EPRI) 3002012988, Alternative Approaches for Addressing Seismic Risk in 10 CFR 50.69 Risk-Informed Categorization, Tier 1 approach to assess seismic risk; the shutdown safety assessment process to assess shutdown risk; the Arkansas Nuclear One, Unit 2 (ANO-2) passive categorization method to assess passive component risk for Class 2 and Clas s 3 SSCs and their associated supports; and a screening of other external hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009; as specified in License Amendment No. 301 dated December 8, 2020 Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from an Appendix R program fire risk evaluation to a fire probabilistic risk assessment approach).

Surry - Unit 2 Subsequent Renewed License No. DPR-37 Amendment No. 313 W. Subsequent Renewed License Conditions (1) The information in the Updated Final Safety Analysis Report (UFSAR) supplement submitted pursuant to 10 CFR 54.21(d), as revised during the subsequent license renewal application review process, and Virginia Electric and Power Company commitments as listed in Appendix A of the Safety Evaluation Report Related to the Subsequent License Renewal of Surry Power Station, Units 1 and 2, dated March 2020, are collectively the Subsequent License Renewal UFSAR Supplement. This Supplement is henceforth part of the UFSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, Virginia Electric and Power Company may make changes to the programs, activities, and commitments described in the Subsequent License Renewal UFSAR Supplement, provided Virginia Electric and Power Company evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, Changes, Tests, and Experiments, and otherwise complies with the requirements in that section.

(2) The Subsequent License Renewal UFSAR Supplement, as defined in subsequent renewed license condition (W)(1) above, describes programs to be implemented and activities to be completed prior to the subsequent period of extended operation, which is the period following the January 29, 2033, expiration of the initial renewed license.

a. Virginia Electric and Power Company shall implement those new programs and enhancements to existing programs no later than 6 months before the subsequent period of extended operation.

b. Virginia Electric and Power Company shall complete those activities by the 6-month date prior to the subsequent period of extended oper ation or by the end of the last refueling outage before the subsequent period of extended operation, whichever occurs later.

c. Virginia Electric and Power Company shall notify the NRC in writing within 30days after having accomplished item (2)a above and include the status of those activities that have been or remain to be completed in item (2)b above.

4. This subsequent renewed license is effective as of the date of issuance and shall expireat midnight on January 29, 2053.

FOR THE UNITED STATES NUCLEAR REGULATORY COMMISSION

Signed by Veil, Andrea on 05/04/21 Andrea Veil, Director Office of Nuclear Reactor Regulation

Enclosure:

Appendix A - Technical Specifications for Surry Power Station, Units 1 and 2 Date of Issuance: May 4, 2021 Surry - Unit 2 Subsequent Renewed License No. DPR-37 TS 1.0-2

5. REACTOR CRITICAL

When the neutron chain reaction is self-sustaining and k eff = 1.0.

6. POWER OPERATION

When the reactor is critical and the ne utron flux power ra nge instrumentation indicates greater than 2% of rated power.

7. REFUELING OPERATION

Any operation involving movement of core components when the vessel head is unbolted or removed.

D. OPERABLE

A system, subsystem, trai n, component, or device shall be operable or have operability when it is capabl e of performing its specif ied function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cool ing or seal water, lubrication or other auxiliary equipm ent that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related s upport function(s). The system or component shall be considered to have this capability when: (1) it satisfies the limiting conditions for operation defined in Section 3, and (2) it h as been tested periodically in accordance with Section 4 and meets its performance requirements.

E. PROTECTIVE INSTRUMENTATION LOGIC

1. ANALOG CHANNEL

An arrangement of compon ents and modules as requ ired to generate a single protective action dig ital signal when re quired by a unit co ndition. An analog channel loses its identity when single action signals are combined.

Amendment Nos. 313 and 313 TS 3.1-14a 11-05-09

This LCO deals with protecti on of the reactor coolant pr essure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assump tions from being exceed ed. The conseque nces of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE SAFETY ANALYSES - Except for primary to seco ndary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leak age can affect the pr obability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary to secondary LEAKAGE from all steam generators (SGs) is 1 gpm or increases to 1 gpm as a result of accident induced conditions. The LCO requiremen t to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gall ons per day is significantly less than the conditions assumed in the safety analysis.

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a main steam line break (MSLB ) accident. Other accidents or transients involve secondary steam release to the atmosphere, such as a st eam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The UFSAR (Ref. 2) analysis for SGTR assumes the contaminated secondary fluid is released via power operated relief valves or safety valves. The source term in the primary system coolant is transported to the affected (ruptured) steam generator by the break flow. The affected steam generator discharges steam to the environment for 30 minutes until the generator is manually isolated. The 1 gpm primary to secondary LEAKAGE transports the source term to the unaffected steam generators. Releas es continue through the unaffected steam genera tors until the Residual Heat Removal System is placed in service.

The MSLB is less limiting for site radiation releases than the SG TR. The safety analysis for the MSLB accident assumes 1 gpm total primary to second ary LEAKAGE, including 500 gpd leakage into the faulted generator. The dose consequences resulting from the MSLB and the SGTR accidents are within the limits defined in the plant licensing basis.

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LIMITING CONDITIONS FOR OPERATION - RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration.

LEAKAGE of this type is unacceptable as the leak itself c ould cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Amendment Nos. 313 and 3 13 TS 3.1-14b 11-05-09

b. Unidentified LEAKAGE

One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c. Identified LEAKAGE

Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.

d. Primary to Secondary LEAKAGE through Any One SG

The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 3). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day. The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

APPLICABILITY - In REACTOR OPERATION conditions where T avg exceeds 200°F, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In COLD SHUTDOWN and REFUELING SH UTDOWN, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.1.C.5 measures leakage through each indivi dual pressure isolation valve (PIV) and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leaktight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

Amendment Nos. 313 and 313 TS 3.10-4

10. A spent fuel cask or heavy loads exceeding 110 percent of the weight of a fuel assembly (not including fuel handling tool ) shall not be moved over spent fuel, and only one spent fuel assembly will be handled at one time over the reactor or the spent fuel pit.

This restriction does not apply to the movement of the transfer canal door.

11. Two Main Control Room/Emergency Switchgear Room (MCR/ESGR)

Emergency Ventilation System (EVS) trains shall be OPERABLE.

a. With one required train inoperable for reasons other than an inoperable MCR/ESGR envelope boundary, restore the inoperable train to OPERABLE status within 7 days. If the inoperable train is not returned to OPERABLE status within 7 days, comply with Specification 3.10.C.

b. If two required trains are inoperable or one or more required trains are inoperable due to an inoperable MCR/ESGR envelope boundary, comply with Specification 3.10.C.

12. Manual actuation of the MCR/ESGR Envelope Isolation Actuation Instrumentation shall be OPERABLE as specified in TS 3.7.F.

13. Three chillers shall be OPERABLE in accordance with the power supply requirements of Specification 3.23.A. With one of the required OPERABLE chillers inoperable or not powered as required by Specification 3.23.A.1, return the inoperable chiller to OPERABLE status within 7 days or comply with Specification 3.10.C. With two of the required OPERABLE chillers inoperable or not powered as required by Specification 3.23.A.1, comply with Specification 3.10.C.

14. Eight air handling units (AHUs) shall be OPERABLE in accordance with the operability requirem ents of Specification 3.23.A. With two AHUs inoperable on the shutdown unit, ensure that one AHU is OPERABLE in each units main control room and emergency switchgear room, and restore an inoperable AHU to OPERABLE status within 7 days, or comply with Specification 3.10.C. With more than two AHUs inoperable, comply with Specification 3.10.C.

Amendment Nos. 313 and 313 TS 3.10-5

B. During irradiated fuel movement in the Fuel Building the following conditions are satisfied:

1. The fuel pit bridge area monitor and the ventilation vent stack 2 particulate and gas monitors shall be OPERABLE and co ntinuously monitore d to identify the occurrence of a fuel handling accident.

2. A spent fuel cask or heavy loads exceeding 110 percent of the weight of a fuel assembly (not including fuel handling tool ) shall not be moved over spent fuel, and only one spent fuel assembly will be handled at one time over the reactor or the spent fuel pit.

This restriction does not apply to the movement of the transfer canal door.

3. A spent fuel cask shall not be moved into the Fuel Building unless the Cask Impact Pads are in place on the bottom of the spent fuel pool.

4. Two MCR/ESGR EVS trains shall be OPERABLE.

a. With one required train inoperable for reasons other than an inoperable MCR/ESGR envelope boundary, restore the inoperable train to OPERABLE status within 7 days. If the inoperable train is not returned to OPERABLE status within 7 days, comply with Specification 3.10.C.

b. If two required trains are inoperable or one or more required trains are inoperable due to an inoperable MCR/ESGR envelope boundary, comply with Specification 3.10.C.

5. Manual actuation of the MCR/ESGR Envelope Isolation Actuation Instrumentation shall be OPERABLE as specified in TS 3.7.F.

6. Three chillers shall be OPERABLE in accordance with the power supply requirements of Specification 3.23.A. With one of the required OPERABLE chillers inoperable or not powered as required by Specification 3.23.A.1, return the inoperable chiller to OPERABLE status within 7 days or comply with Specification 3.10.C. With two of the required OPERABLE chillers inoperable or not powered as required by Specification 3.23.A.1, comply with Specification 3.10.C.

Amendment Nos. 313 and 313 TS 3.10-6

7. Eight air handling units (AHUs) shall be OPERABLE in accordance with the operability requirem ents of Specification 3.23.A. With two AHUs inoperable on either unit, ensure that one AHU is OPERABLE in each units main control room and emergency switchgear room, and restore an inoperable AHU to OPERABLE status within 7 days, or comply with Specification 3.10.C. Wi th more than two AHUs inoperable on a unit, comply with Specification 3.10.C.

C. If any one of the specified limiting conditions for refueling is not met, REFUELING OPERATIONS or irradiated fuel movement in the Fuel Build ing shall cease and irradiated fuel shall be placed in a safe posi tion, work shall be in itiated to correct the conditions so that the specified limit is met, and no operations which increase the reactivity of the core shall be made.

D. After initial fuel loading and after each core refueling operation and prior to reactor operation at greater than 75% of rated power, the movable incore detector system shall be utilized to verify proper power distribution.

E. The requirements of 3.0.1 are not applicable.

Basis

Detailed instructions, the above specified precautions, and the design of the fuel handling equipment, which incorporates built-in interlo cks and safety featur es, provide assurance that an accident, which would result in a hazard to public health and safety, will not occur during unit REFUELING OPERATIONS or irra diated fuel move ment in the Fuel Building. When no change is being made in core geome try, one neutron detector is sufficient to monitor the core and pe rmits maintenance of the out-of-function instrumentation. Continuous monitoring of radiation levels and neutron flux provides immediate indication of an unsafe condition.

Potential escape paths for fission product radioactivity within containment are required to be closed or capable of closure to prevent the release to the environment. However, since there is no potential for significant containment pressuri zation during refueling, the Appendix J leakage criteria and tests are not applicable.

The containment equipment access hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and co mponents into and out of the containment. During REFUELING OPER ATIONS, the equipment hatch must be capable of being closed.

Amendment Nos. 313 and 313 TS 3.16-3

2. If a primary source is not available, the unit may be operated for seven (7) days provided the dependable alternate source can be OPERABLE within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. If specification A-4 is not satisfied within seven (7) days, the unit shall be brought to COLD SHUTDOWN.

3. One battery may be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provided the other battery and battery chargers remain OPERABLE with one battery charger carrying the DC load of the failed batterys supply system. If the batte ry is not returned to OPERABLE status within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the reactor shall be placed in HOT SHUTDOWN. If the battery is not restored to OPERABLE stat us within an additi onal 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in COLD SHUTDOWN.

4. One buried fuel oil storage tank may be inoperable for 7 days for tank inspection and related repair, provided the following actions are taken:

a. prior to removing the tank from service, verify that 50,000 gallons of replacement fuel oil is available offsite and transportation is available to deliver that volume of fuel oil within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, and

b. prior to removing the tank from service and at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, verify that the remaining buried fuel oil storage tank contains 17,500 gallons, and

c. prior to removing the tank from service and at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, verify that the above ground fuel oil storage tank contains 50,000 gallons.

Amendment Nos. 313 and 313 TS 3.16-7

TS action statement 3.16.B.1.a.2 provides an allowance to avoid unnecessary testing of an OPERABLE EDG(s). If it can be determined that the cau se of an inoperable EDG does not exist on the OPERABLE EDG(s), operability testing does not have to be performed. If the cause of the inoperability exists on the other EDG(s), then the other EDG(s) would be declared inoperable upon discovery, and th e applicable require d action(s) would be entered. Once the failu re is repaired, the common cause fa ilure no longer exists and the operability testing requirement for the OPERABLE EDG(s) is satisfied. If the cause of the initial inoperable EDG cannot be confirme d not to exist on the remaining EDG(s),

performance of the oper ability test within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prov ides assurance of continued operability of those EDG(s).

In the event the inoperable EDG is restored to OPERABLE status prior to completing the operability testing requirem ent for the OPERABLE EDG( s), the corrective action program will continue to evaluate the common cause possibility, including the other units EDG or the shared EDG. This continued evaluation, howeve r, is no long er under the 24-hour constraint imposed by the action statement.

According to Generic Letter 84-15 (Ref. 6), 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is r easonable to confirm that the OPERABLE EDG(s) is not affected by the same problem as the inoperable EDG.

References

(1) UFSAR Section 8.5 Emergency Power System

(2) UFSAR Section 9.3 Residual Heat Removal System

(3) UFSAR Section 9.4 Component Cooling System

(4) UFSAR Section 10.3.2 Auxiliary Steam System

(5) UFSAR Section 10.3.5 Condensate and Feedwater System

(6) Generic Letter 84-15, Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability, dated July 2, 1984

Amendment Nos. 313 and 313 TS 3.23-2

2. Air Handling Units (AHUs)

a. Unit 1 air handling units, 1-VS-AC-1, 1-VS-AC-2, 1-VS-AC-6, and 1-VS-AC-7, must be OPERABLE whenever Unit 1 is above COLD SHUTDOWN.

1. If either any single Unit 1 AHU or two Unit 1 AHUs on the same chilled water loop (1-VS-AC-1 and 1-VS-AC-7 or 1-VS-AC-2 and 1-VS-AC-6) become inoperable, restore operability of the one inoperable AHU or two inoperable AHUs within seven (7) days or bring Unit 1 to HOT SHUTDOWN within the next six (6) hours and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

2. If two Unit 1 AHUs on different chilled water loops and in different air conditioning zones (1-VS-AC-1 and 1-VS-AC-6 or 1-VS-AC-2 and 1-VS-AC-7) become inoperable, restore operability of the two inoperable AHUs within seven (7) days or bring Unit 1 to HOT SHUTDOWN within the next six (6) hours and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

3. If two Unit 1 AHUs in the same air conditioning zone (1-VS-AC-1 and 1-VS-AC-2 or 1-VS-AC-6 and 1-VS-AC-7) become inoperable, restore operability of at least one Unit 1 AHU in each air conditioning zone (1-VS-AC-1 or 1-VS-AC-2 and 1-VS-AC-6 or 1-VS-AC-7) within one (1) hour or bring Unit 1 to HOT SHUTDOWN within the next six (6) hours and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4. If more than two Unit 1 AHUs become inoperable, restore operability of at least one Unit 1 AHU in each air conditioning zone (1-VS-AC-1 or 1-VS-AC-2 and 1-VS-AC-6 or 1-VS-AC-7) within one (1) hour or bring Unit 1 to HOT SHUTDOWN within the next six (6) hours and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. Unit 2 air handling units, 2-VS-AC-8, 2-VS-AC-9, 2-VS-AC-6, and 2-VS-AC-7 must be OPERABLE whenever Unit 2 is above COLD SHUTDOWN.

1. If either any single Unit 2 AHU or two Unit 2 AHUs on the same chilled water loop (2-VS-AC-7 and 2-VS-AC-9 or 2-VS-AC-6 and 2-VS-AC-8) become inoperable, restore operability of the one inoperable AHU or two inoperable AHUs within seven (7) days or bring Unit 2 to HOT SHUTDOWN within the next six (6) hours and be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Amendment Nos. 313 and 313 TS 4.9-1

4.9 RADIOACTIVE GAS STORAGE MONITORING SYSTEM

Applicability

Applies to the periodic monitoring of radioactive gas storage.

Objective

To ascertain that waste gas is stored in accordance with Specification 3.11.

Specification

A. The concentration of oxygen in the waste gas holdup system shall be determined to be within the limits of Speci fication 3.11.A by cont inuously monitori ng the waste gases in the waste gas holdup system with the oxygen monitor required to be OPERABLE by Table 3.7-5(a) of Specification 3.7.D.

B. The quantity of radioactive material contained in each gas storage tank shall be determined to be within the limits of Specification 3.11.B at the frequency specified in the Surveillance Frequency Control Program when the specific activity of the primary reactor coolant is 2200 µCi/gm dose equivalent Xe-133. Under the conditions which result in a specific activ ity > 2200 µCi/gm dose equivalent Xe-133, the waste gas decay tanks shall be sampled once per day.

Amendment Nos. 313 and 313 TS 6.1-2

2. Unit Staff The unit staff organization shall include the following:

a. Each on-duty shift shall be composed of at least the minimum shift crew composition for each unit as shown in Table 6.1-1.

b. A radiation protection technician shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the position.

c. All core alterations shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator limited to fuel handling who has no other concurrent responsibilities during this operation.

d. The operations manager shall hold (or have previously held) a Senior Reactor Operator License for Surry Power Station or a similar design Pressurized Water Reactor plant. The Superintendent Nuclear Shift Operations shall hold an active Senior Reactor Operator License for Surry Power Station.

e. Procedures will be established to insure that NRC policy statement guidelines regarding working hours established for employees are followed. In addition, procedures will provide for documentation of authorized deviations from those guidelines and that the documentation is available for NRC review.

6.1.3 Unit Staff Qualifications

1. Each member of the unit staff shall meet or exceed the minimum qualifications referenced for comparable positions as specified in the Nuclear Facility Quality Assurance Program Description. Incumbents in the positions of Shift Manager, Unit Supervisor (SRO), Control Room Operator (RO), and the individual providing advisory technical support to the unit operations shift crew, shall meet or exceed the requirements of 10 CFR 55.59(c) and 55.31(a)(4).

2. For the purpose of 10 CFR 55.4, a licensed Senior Reactor Operator and a licensed Reactor Operator are those individuals who, in addition to meeting the requirements of TS 6.1.3.1 perform the functions described in 10 CFR 50.54(m).

Amendment Nos. 313 and 313 TS 6.7-1

6.7 Environmental Qualifications

A. By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of: Division of Operating Reactors Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors (DOR Guidelines); or, NUREG-0588 Interim Staff Position on Environmental Qua lification of Safety-Related Electrical Equipment, December 1979. Copies of these documents are attached to Order for Modification of Licence Nos. DPR-32 and DPR-37 dated October 24, 1980.

B. By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-r elated electrical equipment in sufficient details to document the degree of compliance with the DOR Guidelines or NUREG-0588.

Thereafter, such records should be updated and maintained curre nt as equipment is replaced, further tested, or otherwise further qualified.

Amendment Nos. 313 and 313 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION

RELATED TO

AMENDMENT NO. 313 TO SUBSEQUENT RENEWED

FACILITY OPERATING LICENSE NO. DPR-32

AND

AMENDMENT NO. 313 TO SUBSEQUENT RENEWED

FACILITY OPERATING LICENSE NO. DPR-37

VIRGINIA ELECTRIC AND POWER COMPANY

SURRY POWER STATION, UNIT NOS. 1 AND 2

DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated June 20, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML ML22171A013), as supplemented by letter dated November 18, 2022 (ML22322A198), Virginia Electric and Power Company (the licensee) submitted a request to revise the technical specifications (TS) and delete expired license conditions for Surry Power Station, Units 1 and 2 (Surry, SPS). The pr oposed amendments would revise the Surry subsequent renewed facility operating license (SRFOL) and TS to make a number of editorial changes and corrections, including removal of the TS and license condition (LC) associated with a one-time plant modification.

The supplement dated November 18, 2022, 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 on September 6, 2022 (87 FR 54553).

2.0 REGULATORY EVALUATION

2.1 Description of Proposed Changes

In its letter dated June 20, 2022, the licensee stated that the proposed changes are administrative in nature and will therefore not result in changes to the plant design or operation. The licensee also provided proposed changes to the TS Basis for NRC information.

Enclosure 3

2.1.1 Proposed TS Changes

The specific changes the licensee proposed to the TS are as follows:

1. Removes expired footnote in TS 1.0.D., DEFINITIONS, applied to the definition of OPERABLE.

2. In TS 3.10, REFUELING, revises reference from obsolete TS 3.23.C to TS 3.23.A in TS 3.10.A.13, TS 3.10.A.14, TS 3.10.B.6, and TS 3.10.B.7.

3. In TS 3.16, EMERGENCY POWER SYSTEM, removes expired footnote and applicable additional actions relating to a temporary extended allowed outage time (AOT) for replacements of the Reserve Station Service Transformer (RSST) C and its associated 5-kV Cable.

4. In TS 3.23, MAIN CONTROL ROOM AND EMERGENCY SWITCHGEAR AIR CONDITIONING SYSTEM, deletes footnote with temporary TS requirements and associated Basis discussion.

5. In TS 4.9, RADIOACTIVE GAS STORAGE MONITORING SYSTEM, revises reference from obsolete TS 3.7.E to TS 3.7.D in TS 4.9.A.

6. In TS 6.1.2.2.d, Revises Supervisor Nuclear Shift Operations to Superintendent Nuclear Shift Operations to reflect the current position title, and

7. In TS 6.7, Environmental Evaluation, corrects the typo By on later than to By no later than.

2.1.2 Proposed Operating License Changes

The licensee proposed the following LC changes to the SRFOL:

1. Remove LC 3.R. LC 3.S, and LC 3.T.

2.2 Regulatory Requirements

The U.S. Nuclear Regulatory Commission (NRC or the Commission) regulatory requirements related to the content of the TS are contained in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." The regulations in 10 CFR 50.36(b), each license authorizing operation of a production or utilization facility will include TS. The categories of items required to be in the TS are provided in 10 CFR 50.36(c).

The regulations in 10 CFR 50.50, "Issuance of licenses and construction permits," state, in part, that "the Commission will issue a license... in such form and containing such conditions and limitations including technical specifications, as it deems appropriate and necessary."

The regulations in 10 CFR 50.54, "Conditions of licenses," specify, paragraphs and applicable requirements, that are conditions in every nuclear power reactor operating license.

Such conditions were placed on the operating license for Surry, requiring the licensee to address outstanding licensing issues to facilitate issuance of the original operating license. The original issuance of the Surry operating license was subject to several additional requirements (license conditions). These additional requirements were incorporated in the license, as required, as a result of ongoing reviews via the license amendment process pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit." In some cases, subsequent license amendments removed certain license conditions when they were satisfied or are no longer applicable. However, the majority of the license conditions have been left intact, including several that have expired by their terms.

The licensee is proposing changes to the Surry SR FOL to remove or revise those license conditions that no longer apply or have been modified. This proposed SRFOL "cleanup" license amendment request also includes minor editorial changes for consistency. Revision of the SRFOL is being proposed to retain only those license conditions that remain pertinent to current Surry operations.

3.0 TECHNICAL EVALUATION

3.1 Proposed SRFOL and TS Changes

3.1.1 SRFOL and TS Changes Related to Replacement of Main Control Room/Emergency Switchgear Room (MCR/ESGR) Air Conditi oning System (ACS) Chilled Water Piping

By letter dated January 23, 2008 (ML073480287), the NRC approved license Amendment Nos. 258/257 for Surry, which implemented a temporary LC and TS to facilitate the replacement of the MCR/ESGR ACS chilled water piping. These amendments added LC 3.R to each units OL as well as a footnote to TS 3.23.C.2.a.1 and 3.23.C.2.b.1, permitting the use of temporary 45-day and 14-day AOT extensions four times in a 24-month time span to permit replacement of degraded MCR/ESGR ACS chilled water piping. These amendments also revised the definition of OPERABLE in TS 1.0.D specific to the MCR/ESGR ACS air handling units on the operating chilled water loop by adding a footnote to permit either a normal or emergency electrical power source to be capable of performing its related support function for the purpose of performing TS-required surveillances that would render an emergency diesel generator inoperable. The chilled water piping replacement work was completed in 2008 and the 24-month time span permitted for use of the four AOT extensions expired on February 4, 2010. The NRC staff concludes that removal of the expired LC 3.R and the footnotes associated with TS 1.0.D, 3.23.C.2.a.1, 3.23.C.2.b.1, and the basis on page TS 3.23-6 will not result in any change to the technical or operating requirements at Surry and are, therefore, acceptable.

3.1.2 SRFOL Changes Related to Control Room Envelope Habitability Program

By letter dated July 7, 2008 (ML081750690), the NRC approved license Amendment Nos. 260/260 for Surry, revising OL and TS requirements related to MCR/ESGR envelope habitability consistent with the NRC approved Revision 3 of Technical Specification Task Force Standard Technical Specifications Change Traveler 448, Control Room Habitability (ML062210095). The NRC staff have found that LC 3.S(1), 3.S(2), and 3.S(3) have all been completed (October 15, 2010; June 18, 2009; and December 1, 2008, respectively) per the LC 3.S implementation timeline and removal of LC 3.S is, therefore, acceptable.

3.1.3 SRFOL Changes Related to Measurement Uncertainty Recapture (MUR)

By letter dated September 24, 2010 (ML101750002), the NRC approved license Amendment Nos. 269/268 for Surry, which changed TS associated with an implementation of a MUR power uprate from 2546 to 2587 megawatts thermal (MWt). LC 3.T provided sixteen new requirements, fourteen of which had to be completed prior to operating above 2546 MWt. All sixteen requirements have been completed as of September 27, 2011. The NRC staff concludes that removal of LC 3.T is editorial in nature, and is, therefore, acceptable.

3.1.4 TS Changes Related to Temporary Extended AOTs for Replacement of RSST C and its associated 5-kV Cable

By letter dated October 5, 2018 (ML18261A099), the NRC approved license Amendment Nos. 293/293 to Surry, implementing a temporary, one-time, 21-day AOT in a footnote to TS 3.16, Action B.2, to permit replacement of the Unit 2 RSST C. By letter dated March 16, 2020 (ML20058F966), the NRC approved license Amendment Nos. 297/297, revising the footnote included in amendments 293/293 to permit a temporary, one-time, 14-day AOT for replacement of the 5-kV cabling associated with Unit 2 RSST C consistent with the specified footnote conditions. RSST C was replaced during the fall 2018 Unit 2 RFO, and its associated cabling was replaced during the spring 2020 Unit 2 RFO. The NRC staff concludes that removal of the one-time AOT footnote in the TS 3.16, Action B.2 and footnote has been satisfied and is, therefore, acceptable.

3.1.5 Editorial TS Changes

The licensee proposed the following editorial changes in the TS:

1. Amendments 228/228 renumbered TS 3.7.E to TS 3.7.D. However, this was not updated in TS 4.9, Radioactive Gas Storage Monitoring System, section TS 4.9.A, which reference TS 3.7.E. This change will correct the reference from TS 3.7.E to TS 3.7.D in this applicable section.

2. Amendments 260/260 renumbered TS 3.23.C to TS 3.23.A. However, this was not updated in TS 3.10, Refueling, sections TS 3.10.A.13; TS 3.10.A.14; TS 3.10.B.6; and TS 3.10.B.7, which all reference TS 3.23.C. This change will correct the reference from TS 3.23.C to TS 3.23.A in these applicable sections.

3. TS 6.1.2, Organization includes the Unit Staff requirements in TS 6.1.2.2. TS 6.1.2.2.d currently refers to the title of Supervisor Nuclear Shift Operations. The amendments

would update the position title to be Superintendent Nuclear Shift Operations. The title is changed and not the role or qualifications specified by the TS.

4. This editorial change will correct the typo By on later than with By no later than in TS 6.7.B.

The NRC staff reviewed these four editorial changes and agree that correction of items one through five above are editorial in nature, and are, therefore, acceptable.

3.3 NRC Staff Conclusion

The NRC staff finds that the proposed changes to: TS 1.0.D; TS 3.10; TS 3.16; TS 3.23.C.2.a.1; TS 3.23.C.2.b.1; TS 4.9.A; TS 6.1.2.2.d; and TS 6.7.B and LC 3.R; 3.S; and 3.Tare editorial in nature and do not change the technical requirements.

The NRC staff concludes that with the deletion of the expired footnotes and basis in TS 1.0.D; TS 3.23.C.2.a.1; TS 3.23.C.2.b.1; TS 3.16; LCs 3.R; 3.S; and 3.T, the changes in TS 3.10, TS 4.9.A; TS 6.1.2.2.d; and TS 6.7.B, Surry Units 1 and 2, continue to meet the regulatory requirements of 10 CFR 50.36 and are, therefore, acceptable. The NRC staff considered the proposed TS Basis for information, but pursuant to 10 CFR 50.36(a)(1), the bases shall not become part of the technical specifications.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Commonwealth of Virginia official was notified of the proposed issuance of the amendments on May 15, 2023. On May 15, 2023, the State official confirmed that the Commonwealth of Virginia had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increa se in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration published in the Federal Register on September 6, 2022, 87 FR 54553, 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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: ZTurner, NRR JKlos, NRR

Date: June 29, 2023

ML23136B139 OFFICE DORL/LPL2-1/PM DORL/LPL2-1/PM DORL/LPL2-1/LA DSS/STSB/BC NAME ZTurner JKlos KGoldstein VCusumano DATE 05/15/2023 05/15/2023 05/17/2023 05/22/2023 OFFICE OGC DORL/LPL2-1/BC DORL/LPL2-1/PM NAME BVaisey MMarkley JKlos DATE 06/07/2023 06/29/2023 06/29/2023