ML25016A346

Unit Nos. 1 and 2 - Issuance of Amendment Nos. 320 and 320, Technical Specification 3.8 Containment Change and Addition of Surveillance Requirement, Temperature, and Pressure Values (EPID L-2024-LLA-0038)

ML25016A346
Person / Time
Site:	Surry
Issue date:	02/20/2025
From:	John Klos Plant Licensing Branch II
To:	Carr E Virginia Electric & Power Co (VEPCO)
Klos, J
References
EPID L-2024-LLA-0038
Download: ML25016A346 (1)

Text

February 20, 2025 Eric S. Carr President - Nuclear Operations and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 320 AND 320, TECHNICAL SPECIFICATION 3.8, CONTAINMENT, CHANGE AND ADDITION OF SURVEILLANCE REQUIREMENT, TEMPERATURE, AND PRESSURE VALUES (EPID L-2024-LLA-0038)

Dear Eric Carr:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 320 to Subsequent Renewed Facility Operating License No. DPR-32 and Amendment No. 320 to Subsequent Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2 (Surry) respectively. The amendments revise the Technical Specifications (TSs) in response to your application dated March 22, 2024, (Agencywide Documents Access and Management System (ADAMS), Accession No. ML24087A208).

The proposed changes revise the Surry TS 3.8, Containment, to: 1) change the TS 3.8.D. title from Internal Pressure to Containment Pressure, 2) add a minimum containment air partial pressure limit to the Limiting Condition for Operation (LCO) in TS 3.8.D., and 3) add a new TS 3.8.e., Containment Air Temperature, to provide an LCO for containment average air temperature. The proposed changes also revise TS 4.1, Operational Safety Review, incorporate a TS Surveillance Requirements for containment air partial pressure and containment average air temperature. The amendment request stated that the proposed changes are consistent with NUREG-1431, Revision 5, Standard Technical Specifications -

Westinghouse Plants, Section 3.6, Containment Systems, Volume 1 (ML21259A155).

A copy of the Safety Evaluation is also enclosed. The 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. 320 to DPR-32

2. Amendment No. 320 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. 320 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 March 22, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, 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. 320, 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 60 days.

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

Attachment:

Changes to License No. DPR-32 and the Technical Specifications Date of Issuance: February 20, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.02.20 10:50:46 -05'00' VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 320 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 March 22, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, 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. 320, 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 60 days.

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

Attachment:

Changes License No. DPR-37 and the Technical Specifications Date of Issuance: February 20, 2025 MICHAEL MARKLEY Digitally signed by MICHAEL MARKLEY Date: 2025.02.20 10:51:35 -05'00'

ATTACHMENT SURRY POWER STATION, UNIT NOS. 1 AND 2 TO LICENSE AMENDMENT NO. 320 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND TO LICENSE AMENDMENT NO. 320 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages License License License No. DPR-32, page 3

License No. DPR-32, page 3 License No. DPR-37, page 3

License No. DPR-37, page 3 TSs TSs 3.8-3 3.8-3 3.8-3a 3.8-3a 3.8-5 3.8-5 4.1-2 4.1-2 4.1-5b 4.1-5b Surry - Unit 1 Subsequent Renewed License No. DPR-32 Amendment No. 320

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. 320 are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

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

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

E. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 Surry - Unit 2 Subsequent Renewed License No. DPR-37 Amendment No. 320

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. 320 are hereby incorporated in this 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

c.

Isolate each affected penetration within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual valve or blind flange, or

d. Otherwise, place the unit in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

D.

Containment Pressure 1.

Containment air partial pressure shall be 10.1 psia and within the acceptable operation range as identified in Figure 3.8-1 whenever the Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively.

a.

With the containment air partial pressure outside the acceptable operation range, restore the air partial pressure to within acceptable limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

E.

Containment Temperature 1.

Containment average air temperature shall be 75°F and 125°F whenever the Reactor Coolant System temperature and pressure exceed 350°F and 450 psig, respectively.

a.

If containment average temperature is not within the limits, restore the containment average temperature to within the limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

Basis CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment will be restricted to those leakage paths and associated leak rates assumed in the accident analysis. These restrictions, in conjunction with the allowed leakage, will limit the site boundary radiation dose to the applicable limits of 10 CFR 50.67 or Regulatory Guide 1.183 during accident conditions.

TS 3.8-3 Amendment Nos. 320 and 320

The operability of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment.

The opening of manual or deactivated automatic containment isolation valves on an intermittent basis under administrative control includes the following considerations:

(1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and TS 3.8-3a Amendment Nos. 320 and 320

If the containment air partial pressure rises to a point above the allowable value the reactor shall be brought to the HOT SHUTDOWN condition. If a LOCA occurs at the time the containment air partial pressure is at the maximum allowable value, the maximum containment pressure will be less than design pressure (45 psig), the containment will depressurize to 2.0 psig within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than 0.0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The radiological consequences analysis demonstrates acceptable results provided the containment pressure does not exceed 2.0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the Design Basis Accident.

If the containment air partial pressure cannot be maintained greater than or equal to the minimum pressure in Figure 3.8-1, the reactor shall be brought to the HOT SHUTDOWN condition. The shell and dome plate liner of the containment are capable of withstanding an internal pressure as low as 3 psia, and the bottom mat liner is capable of withstanding an internal pressure as low as 8 psia.

During a Design Basis Accident, with an initial containment average air temperature within the specified temperature limits, the resultant peak accident temperature is maintained below the containment design temperature. As a result, the ability of containment to perform its design function is ensured.

References UFSAR Section 4.2.2.4 Reactor Coolant Pump UFSAR Section 5.2 Containment Isolation UFSAR Section 5.2.1 Design Bases UFSAR Section 5.2.2 Isolation Design UFSAR Section 5.3.4 Containment Vacuum System TS 3.8-5 Amendment Nos. 320 and 320

TS 4.1-2 Amendment Nos. 320 and 320 H. If the RWST Water Chemistry exceeds 0.15 PPM for Cl and/or F, flushing of sensitized stainless steel piping as required by 4.1.E will be performed once the RWST Water Chemistry has been brought within specification limit of less than 0.15 PPM chlorides and/or fluorides. Samples will be taken periodically until the sample indicates the Cl and/or F and levels are below 0.15 PPM.

I.

Containment Pressure - Verify containment air partial pressure is within limits at the frequency specified in the Surveillance Frequency Control Program.

J.

Containment Air Temperature - Verify containment average air temperature is within limits at the frequency specified in the Surveillance Frequency Control Program.

BASIS Check Failures such as blown instrument fuses, defective indicators, and faulted amplifiers which result in upscale or downscale indication can be easily recognized by simple observation of the functioning of an instrument or system.

Furthermore, such failures are, in many cases, revealed by alarm or annunciator action, and a periodic check supplements this type of built-in surveillance.

Calibration Calibration shall be performed to ensure the presentation and acquisition of accurate information.

The nuclear flux (power level) channels shall be calibrated against a heat balance standard to account for errors induced by changing rod patterns and core physics parameters. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

TS 4.1-5b Amendment Nos. 320 and 320 Trending the results of this surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The SFCP 7 day Frequency considers the low probability of a gross fuel failure during this time.

Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and I-134, it is acceptable to include the minimum detectable activity for Kr-85 in this calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity.

CONTAINMENT AIR PARTIAL PRESSURE Verifying containment air partial pressure is within the LCO limits ensures containment operation remains within the limits assumed for the containment analyses. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

CONTAINMENT AVERAGE AIR TEMPERATURE Verifying containment average air temperature remains within the LCO limits ensures containment operation remains within the limits assumed for the containment analyses. To determine the containment average air temperature, a weighted average is calculated using measurements taken at locations within containment selected to provide a representative sample of the overall containment atmosphere. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 320 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 320 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 March 22, 2024 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24087A208), Virginia Electric and Power Company (the licensee) submitted a license amendment request (LAR) for changes to the Surry Power Station, Unit Nos. 1 and 2 (Surry) Technical Specifications (TSs).

The proposed changes revise Surrys TS 3.8, Containment, to: 1) change the TS 3.8.D., title from Internal Pressure to Containment Pressure, 2) add a minimum containment air partial pressure limit to the Limiting Condition for Operation (LCO) in TS 3.8.D, and 3) add a new TS 3.8.E., Containment Temperature, to provide an LCO for containment average air temperature. The proposed changes also revise TS 4.1, Operational Safety Review, to incorporate TS Surveillance Requirements (SRs) for containment air partial pressure and containment average air temperature.

In its \

letter dated March 22, 2024, the licensee stated that the proposed LCO and SRs are required to be included in the Surry TS pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Part 50.36(c), and the changes are consistent with Standard TS (STS) 3.6.4B, Containment Pressure (Subatmospheric), and 3.6.5C, Containment Air Temperature (Subatmospheric), contained in NUREG-1431, Revision 5, Volume 1, Standard Technical Specifications - Westinghouse Plants, (ML21259A155).

2.0 REGULATORY EVALUATION

2.1 System Description 2.1.1 Containment Pressure In Section 2.1.1, Containment Pressure, of Attachment 1 to the LAR, the licensee stated:

Containment air partial pressure is a process variable that is monitored and controlled. The containment air partial pressure is maintained as a function of Refueling Water Storage Tank (RWST) temperature and Service Water (SW) temperature as shown on Figure 2.1 (TS Figure 3.8-1 [Surry Technical Specification Curve for Containment Allowable Air Partial Pressure Indication vs. Service Water Temperature]) to ensure that, following a Design Basis Accident (DBA), the containment pressure will be less than 2.0 psig [pounds per square in gauge] within one (1) hour and less than 0.0 psig within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Controlling containment partial pressure within prescribed limits also prevents the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of an inadvertent actuation of the Containment Spray (CS) system.

Pursuant to TS 3.8.D, containment air partial pressure shall be maintained within the acceptable operation range as identified on Figure 3.8-1 whenever the Reactor Coolant System (RCS) temperature and pressure exceed 350 °F and 450 psig, respectively. The control of containment partial pressure within these limits is achieved by the Containment Vacuum and Leakage Monitoring System.

The relevance of RWST temperature to TS figure 3.8-1, as noted above, is not evident from the figure. However, the licensee states in Section 2.2, Current Technical Specifications Requirements, of Attachment 1 to the LAR that [t]he associated SPS [Surry] TS 3.8 Basis states:[that] The horizontal upper limit line in TS Figure 3.8-1 is based on MSLB [main steam line break] peak calculated pressure criteria.

Surrys Updated Final Safety Analysis Report (UFSAR), Revision 56 (ML24269A218)

Table 5.4-17, Key Parameters in the Containment Analysis, lists RWST temperature as 32-46.6 °F, which includes 1.6 °F uncertainty. Note a. of that Table states that Minimum RWST temperature of 32 °F is assumed for evaluation of the inadvertent CS actuation event.

Normal operating range for RWST temperature is 40-45 °F. This clarifies the relevance of RWST temperature to TS figure 3.8-1.

2.1.2 Containment Temperature In Section 2.1.2, Containment Temperature, of Attachment 1 to the LAR, the licensee stated:

Containment average air temperature is an initial condition used in the DBA analyses that establishes the containment environmental qualification operating envelope for both pressure and temperature. The limit for containment average air temperature ensures that operation is maintained within the assumptions

used in the DBA analyses for containment. Specifically, two temperatures affect the maximum allowable containment partial pressure: SW temperature and containment average air temperature (see Figure 2.1 [TS Figure 3.8-1]). Both temperatures affect the allowable containment partial pressure in the same way: the higher the temperature, the lower the upper limit for allowable partial pressure. This is because the higher any of these temperatures are at the start of a [Loss of Coolant Accident] LOCA, the higher the peak accident pressure will be. However, the existing analyses for operation with a core rated power of 2587 [MegaWatt thermal] MWth allow containment bulk air temperature to vary between 75 °F and 125 °F, with an upper limit on air partial pressure of 11.3 [pounds per square inch absolute] psia. The containment partial pressure LCO specifies a lower limit of not less than 10.1 psia. This lower limit is imposed to ensure the containment pressure does not decrease below the value of 8.0 psia during inadvertent initiation of CS as noted above.

Containment average air temperature is computed as a weighted average of twenty-four dry bulb platinum resistance temperature detectors (RTDs).

Containment partial pressure and average air temperature must be verified to be less than their upper limits and greater than their lower limits or the reactor must be placed in HOT SHUTDOWN.

2.1.3 Overview of Containment Design Section 5.1 of the SPS UFSAR states, in part that:

The containment system, together with the engineered safeguards (Chapter 6), is designed to limit radiation doses under conditions resulting from design-basis accident (Chapter 14) to less than or equal to the limits specified in 10 CFR 50.67 at the site boundary and beyond.

The steel-lined, reinforced-concrete containment structure, including foundations, access openings, and penetrations are designed and constructed to maintain full containment integrity when subjected to the temperatures, pressures, potential missiles resulting from the design-basis accident, and the earthquake conditions and tornados described in Chapter 2. Systems are provided to remove heat from the containment and to ensure against breaching containment integrity at the time of, or following, the design-basis accident, or any lesser accident.

The original containment concept includes provisions for routine operation at a reduced internal pressure in which the air partial pressure varies between about 9.0 and 10.3 psia, and for the return to subatmospheric pressure within 60 minutes after the design-basis accident through the use of multiple spray systems. This concept provides for positive termination of outleakage of fission products from the containment, since the containment is maintained at subatmospheric pressure after depressurization. The pressure following depressurization is maintained at less than 14.7 psia. The current concept for the design basis accident containment internal pressure reduction, consistent with alternate source term (AST) analysis, is discussed in Section 5.4.

The upper limit on air partial pressure ensures that the containment can be depressurized within the required time. The lower limit on air partial pressure ensures that containment structural limits will not be challenged by a drop in pressure caused by the inadvertent actuation of the

quench spray system. Containment air partial pressure is controlled as a function of refueling water storage tank temperature and service water temperature. The limits on containment air temperature preserve the initial conditions assumed in the accident analyses.

2.2 Description of Proposed Changes to the TS In Section 2.4 of Attachment 1 to the LAR, the licensee states:

TS 3.8.D will be retitled from Internal Pressure to Containment Pressure for consistency with STS 3.6.4 and the minimum containment pressure value of 10.1 psia will be added to the TS LCO.

New TS 3.8.E, Containment Temperature, will be included in TS 3.8 to provide an LCO for containment average temperature as follows:

E. Containment Temperature

1. Containment average air temperature shall be 75 °F and 125 °F whenever the Reactor Coolant System temperature and pressure exceed 350 °F and 450 psig, respectively.

a. If containment average temperature is not within the limits, restore the containment average temperature to within the limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and 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 />.

As noted in the STS, the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> provided to restore the containment average temperature to within the LCO limits is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.

Two new TS SRs, 4.I, Containment Pressure, and 4.J, Containment Temperature, are being added to TS 4.1, Operational Safety Review, as follows:

I.

Containment Pressure - Verify containment air partial pressure is within limits at the frequency specified in the Surveillance Frequency Control Program.

J. Containment Air Temperature - Verify containment average air temperature is within limits at the frequency specified in the Surveillance Frequency Control Program.

The initial surveillance frequencies to be included in the plant Surveillance Frequency Control Program (SFCP) for containment pressure and containment temperature will be 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, respectively. The 12-hour frequency for verification of containment air partial pressure is considered acceptable based on operating experience related to trending of containment pressure variations and pressure instrument drift during applicable REACTOR OPERATION conditions. Furthermore, the 12-hour frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment pressure condition. The 24-hour frequency for verification of containment average air temperature is considered acceptable based on observed slow rates of temperature

increase within containment as a result of environmental heat sources (due to the large volume of containment). Furthermore, the 24-hour frequency is considered adequate in view of other indications available in the control room, including alarms, to alert the operator to an abnormal containment temperature condition.

If subsequent changes to the surveillance frequencies are deemed appropriate, they would be evaluated in accordance with NEI 04-10, Risk-informed Method for Control of Surveillance Frequencies, Revision 1, [(ML071360456)] as required by TS 6.4.S, Surveillance Frequency Control Program (SFCP).

Conforming changes to the TS 3.8 and 4.1 Bases are also being made as indicated in Attachments 2 and 3 of the LAR.

2.3 Reason for the Proposed Change In Section 2.3, Attachment 1 of the LAR, the licensee states:

SPS TS Figure 3.8-1, Surry Technical Specification Curve for Containment Allowable Air Partial Pressure Indication vs. Service Water Temperature, (see Figure 2.1 above),

provides the limiting operating conditions for the reactor containment including limits on containment air partial pressure and containment average air temperature. TS 3.8.D provides the LCO for containment air partial pressure; however, the SPS TS do not currently include a SR for containment air partial pressure, nor do they include a TS LCO or SR for containment average air temperature for verifying containment partial air pressure and containment average air temperature remain within their limits as indicated on TS Figure 3.8-1.

10 CFR 50.36(c)(2)(ii), Criterion 2, requires an LCO be provided for (a) process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Since the containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA, it is required to have a TS LCO.

Likewise, 10 CFR 50.36(c)(3) requires SRs relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Consequently, SRs are necessary for containment air partial pressure and containment average air temperature to verify they are within their respective LCO limits to ensure that containment operation remains within the limits assumed for the containment analyses.

2.4 Regulatory Requirements and Guidance The General Design Criteria (GDC) included in Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, General Design Criteria for Nuclear Power Plants, became effective on May 21, 1971. The construction permits for Surry were issued prior to May 21, 1971; consequently, Surry is not subject to the current GDC requirements. Surrys Updated Final Safety Analysis Report (UFSAR), Revision 56 (ML24269A218). Section 1.4, Compliance with Criteria, of the UFSAR provides, in part, that Surry meets the intent of GDC of Appendix A to 10 CFR Part 50. The NRC staff reviewed Surrys UFSAR plant-specific design criteria and

found that Section 1.4.10, Containment, and Section 1.4.49, Containment Design Basis, is consistent with GDC 38, Containment heat removal, and GDC 50 Containment design basis.

Surry UFSAR, Section 1.4.10 states, in part, that:

Engineered safeguards, which consist of safety injection systems and containment depressurization systems, serve to cool the reactor core and return the containment to subatmospheric pressure and maintain it at subatmospheric pressure for as long as the situation requires.

Surry UFSAR, Section 1.4.49 states, in part, that:

The containment structure, including access openings and penetrations and any necessary containment heat removal systems, is designed to accommodate, without exceeding the design leakage rate, the pressures and temperatures resulting from the largest credible energy release following a LOCA, including a considerable margin for the effects of metal-water or other chemical reactions that can occur as a consequence of the failure of safety injection systems.

Regulations GDC 38, Containment heat removal, of Appendix A to 10 CFR Part 50 requires, in part, that a system to remove heat from the reactor containment shall be provided and that the system safety function shall be to reduce rapidly, consistent with the functioning of other associated systems, the containment pressure and temperature following any loss-of-coolant accident and maintain them at acceptably low levels.

GDC 50, Containment design basis, requires, in part, that the containment structure, including access openings, penetrations, and the containment heat removal system, shall be designed so that the containment structure and its internal compartments can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions after a loss-of-coolant accident.

The regulations in 10 CFR 50.36(c)(2) require that technical specifications include limiting conditions for operation (LCO). LCO are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications.

The regulations at 10 CFR 50.36(c)(2)(ii) state, in part, that technical specification LCO must be established for each item meeting one or more of the following criteria:

(A)

Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(B)

Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either

assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(C)

Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D)

Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The regulations at 10 CFR 50.36(c)(3) state:

Surveillance requirements [SRs] are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

3.0 TECHNICAL EVALUATION

3.1 Containment Pressure Containment air partial pressure is an initial condition used in the containment DBA analyses to establish the maximum peak containment internal pressure below the maximum design value of 45 psig for the containment. The licensee states, in Section 2.2, Attachment 1 of the LAR that maintaining containment pressure within the TS limits of TS 3.8.D ensures that following a DBA, the resultant peak containment accident pressure will be maintained below the containment design pressure. The NRC staff reviewed Surrys UFSAR (ML24269A218) Tables 5.4-10, Containment LOCA Analysis Initial Conditions, and 5.4-11, Containment LOCA Analysis Peak Pressure Results, and confirmed that using the containment pressure within TS Figure 3.8-1 limits the peak pressure calculated for containment LOCA analysis to 43.95 psig, which is below the containment design pressure of 45 psig.

The licensee states that these TS limits also prevent the containment pressure from exceeding the containment design negative pressure differential with respect to the outside atmosphere in the event of inadvertent actuation of the CS system. The NRC staff independently reviewed this and confirmed that the containment liner would continue to meet the performance criteria Surry UFSAR Section 5.3.4.3 for inadvertent operation of the CS system.

In Section 2.2, Attachment 1 of the LAR, the licensee stated that the LCO limits and TS Basis ensure the containment pressure does not exceed 2.0 psig for the interval from 1 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following a DBA, and less than 0.0 psig beyond 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to achieve acceptable radiological results. The NRC notes this is consistent with the approved Surry updated alternative source term analyses.

The licensee is proposing to (1) add a minimum containment air partial pressure limit, as given in TS figure 3.8-1, to the LCO in TS 3.8.D and (2) change Internal Pressure to Containment Pressure, which is editorial. As discussed above, the NRC staff confirmed that the licensees

proposed changes to TS 3.8.D will not challenge the containment design pressure, the containment design negative pressure differential, or radiological limits. Therefore, the staff finds the proposed changes to be acceptable.

3.2 Containment Temperature The containment structure serves to contain radioactive material and limit radiation doses under conditions resulting from a DBA. The containment average air temperature is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a LOCA.

Surry UFSAR (ML24269A218) Section 5.4.2.1.7, LOCA - Containment Pressure and Temperature Results, states that for the double-ended guillotine break, the initial conditions which result in the maximum depressurization time included an initial containment temperature of 125 °F; the initial conditions which result in the maximum time to approach the four-hour subatmospheric condition included an initial containment temperature of 75 °F. The NRC staff finds that the licensees proposed LCO in TS 3.8.D that states that Containment average air temperature shall be 75 °F and 125 °F whenever the RCS temperature and pressure exceed 350 °F and 450 psig, respectively, stated in Section 2.4, Attachment 1 of the LAR is consistent with TS figure 3.8-1 and the Surry licensing basis and is therefore acceptable.

3.3 Containment Pressure and Temperature Analysis The NRC staff reviewed the proposed containment pressure value of 10.1 psia in TS 3.8.D to determine if new analyses were required. Based on its independent review, the NRC staff finds the proposed value to be acceptable because it is bounded by the analyses of record for the current TSs.

The NRC staff independently reviewed the proposed containment average temperature values of 75 °F and 125 °F, in TS 3.8.E, to determine if further analyses is required. The NRC staff finds that the proposed values are bounded by the analyses of record for the current TSs. The NRC staff also reviewed the proposed completion times in TS 3.8.E and found them to be consistent with NUREG-1431, Revision 5, Standard Technical Specifications - Westinghouse Plants, Section 3.6, Containment Systems, Volume 1 (ML21259A155). Based on the above, the NRC staff finds the proposed changes to be acceptable.

The NRC staff finds that no new analysis is required for the proposed changes because the proposed TS changes are bounded by the current containment DBA licensing basis analysis, which is consistent with SPS plant design criteria and the regulations in 10 CFR 50, Appendix A, GDC 38 and 50. As such, the staff finds these changes to be acceptable.

3.4 Technical Specifications The NRC staff reviewed the proposed changes to TS 3.8 and the licensees justifications for the proposed changes. The NRC staff determined that the proposed title change to TS 3.8.D is editorial in nature, and is, therefore, acceptable.

The NRC staff determined that the new TS 3.8.E is more restrictive than current TS requirements. Current TSs do not have a requirement for containment average temperature.

The NRC staff determined that the changes are consistent with the licensing basis for Surry and meet the criteria for inclusion in TSs.

The NRC staff finds that the LCO and applicability statement and ACTIONS are acceptable because they meet the requirements of 10 CFR 50.36(c)(2). Therefore, the proposed changes to TS 3.8 are acceptable.

The NRC staff compared the new SRs (TS 4.1.I and 4.1.J of the proposed new TS 3.8.E) to the requirements for SRs in 10 CFR 50.36(c)(3). The staff found that the new SRs are more restrictive than the licensees current requirements because they are not currently included in the TSs. Based on this, the staff finds that the SRs are sufficient to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCO will be met, in accordance with in 10 CFR 50.36(c)(3).

3.5 TS Bases The NRC reviewed the proposed TS Bases for information only. The NRC does not approve TS Bases changes.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Commonwealth of Virginias State official was notified of the proposed issuance of the amendments on January 14, 2025. On January 14, 2025, the Commonwealth of Virginias 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 and change surveillance requirements. 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 June 11, 2024 (89 FR 49242), 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: Andrea Russell, NRR Date: February 20, 2025

ML25016A346 NRR-058 OFFICE DORL/LPL2-1/PM DORL/LPL2-1/LA DSS/STSB/BC NAME JKlos KZeleznock SMehta DATE 01/10/2025 01/27/2025 and 02/10/25 12/02/2024 OFFICE DSS/SNSB/BC DSS/SCPB/BC OGC - NLO NAME PSahd MValentin MWright DATE 09/22/2024 09/18/2024 02/07/2025 OFFICE DORL/LPL2-1/BC DORL/LPL2-1/PM NAME MMarkley JKlos DATE 02/20/2025 02/20/2025