ML20274A329

Proposed License Amendment Request: Revise Reactor Core Safety Limit to Reflect WCAP-17642-P-A, Revision 1

ML20274A329
Person / Time
Site:	Surry
Issue date:	09/30/2020
From:	Mark D. Sartain Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
20-341
Download: ML20274A329 (18)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 September 30, 2020 10 CFR 50.90 U. S. Nuclear Regulatory Commission Serial No.: 20-341 Attention: Document Control Desk NRA/GDM: RO Washington, DC 20555-0001 Docket Nos.: 50-280 50-281 License Nos.: DPR-32 DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST REVISE REACTOR CORE SAFETY LIMIT TO REFLECT WCAP-17642-P-A, REVISION 1 Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) requests amend.ments to Surry Power Station (Surry) Units 1 and 2 Facility Operating License Numbers DPR-32 and DPR-37, respectively, in the form of a change to the Technical Specifications (TS). The proposed change revises the "Safety Limit, Reactor Core" (SL) 2.1.A.1.b to reflect the peak fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5)." Attachment 1 provides discussion and evaluation of the proposed change. The marked-up and proposed pages for the TS are provided in Attachments 2 and 3, respectively.

Dominion Energy Virginia has evaluated the proposed amendment and has determined it does not involve a significant hazards consideration as defined in 10 CFR 50.92. The basis for this determination is included in Attachment 1. We have also determined operation with the proposed change will not result in a significant increase in the amount of effluents that may be released offsite or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment is eligible for categorical exclusion from an environmental assessment as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is needed in connection with the approval of the proposed change.

The license amendment request has been reviewed and approved by the Facility Safety Review Committee. Dominion Energy Virginia requests approval of the proposed change by September 30, 2021 with a 90-day implementation period.

Serial No.20-341 Docket Nos. 50-280/281 Page 2 of 3 Should you have any questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.

Respectfully, Mark D. Sartain Vice President - Nuclear Engineering and Fleet Support Commitments contained in this letter: None Attachments:

1. Discussion of Change

2. Marked-up Technical Specifications Page

3. Proposed Technical Specifications Page COMMONWEAL TH OF VIRGINIA )

)

COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by Mr. Mark D. Sartain, who is Vice President - Nuclear Engineering and Fleet Support, of Virginia Electric and Power Company. He has affirmed before me that he is duly authorized to execute and file the foregoing document in behalf of that company, and that the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this 3D-fh day of 5'.-!f~ f:ur , 2020.

My Commission Expires: tJ.) 31 /w

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CRAIG Notary Commonwealt Notary Public Reg.# 7 My Commission Expire

Serial No.20-341 Docket Nos. 50-280/281 Page 3 of 3 cc: U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, GA 30303-1257 State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street Suite 730 Richmond, VA 23219 Mr. Vaughn Thomas NRC Project Manager - Surry U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 04 F-12 11555 Rockville Pike Rockville, MD 20852-2738 Mr. G. Edward Miller NRC Senior Project Manager - North Anna U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Surry Power Station

Serial No.20-341 Docket Nos. 50-280/281 Enclosure 1 DISCUSSION OF CHANGE Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 1 of 10 DISCUSSION OF CHANGE 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) requests amendments to Surry Power Station (Surry) Units 1 and 2 Facility Operating License Numbers DPR-32 and DPR-37, respectively, in the form of a change to the Technical Specifications (TS). The proposed change revises the "Safety Limit, Reactor Core" (SL) 2.1.A.1.b to reflect the peak fuel centerline melt temperature specified in WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PADS)."

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation Surry Units 1 and 2 must ensure acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs), consistent with the Surry Units 1 and 2 licensing bases. To accomplish this, Surry Units 1 and 2 TS 2.1, "Safety Limit, Reactor Core," ensures Departure from Nucleate Boiling (DNB) does not occur and the fuel centerline temperature remains below the fuel melting temperature. The proposed amendment revises the fuel centerline melting temperature specified in SL 2.1.A.1.b but does not alter the SL associated with the DNB ratio.

Fuel centerline melting occurs when the local linear heat rate (LHR), or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the fuel pellet following centerline melting could cause excessive cladding stress leading to failure of the cladding and uncontrolled release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak LHR below the level at which fuel centerline melting occurs.

The proper functioning of the Reactor Protection System and steam generator safety valves prevents violation of the Reactor Core SLs.

2.2 Current Technical Specification Requirement SL 2.1.A.1.b defines the burnup-dependent fuel temperature below which the fuel centerline temperature must be maintained. SL 2.1.A.1.b applies whenever the reactor is critical. TS 2.1.B requires the unit be placed in HOT SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in the event the Safety Limit is exceeded.

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 2 of 10 2.3 Reason for the Proposed Change Plant-specific safety analyses are performed to ensure compliance with the Safety Limit is maintained. Westinghouse Performance Analysis and Design Model (PAD5) methodology (Reference 1) defined the fuel pellet melting limit that is included in the PAD5 methodology based on available fuel pellet material properties. The NRC staff reviewed and approved the Westinghouse methodology and concluded the melting limits defined in Reference 1 are acceptable.

The proposed amendment will be implemented to maintain consistency between the value in Safety Limit 2.1.A.1.b and the criteria used when performing confirmatory safety analyses that rely on the NRC approved methodology in Reference 1.

2.4 Description of Proposed Changes The proposed change revises the peak fuel centerline temperature specified in SL 2.1.A.1.b but does not alter the Required Action that must be taken following a violation of the limit. The following changes are proposed to the Surry Units 1 and 2 TS.

The current version of SL 2.1.A.1.b reads:

"The peak fuel centerline temperature shall be maintained <5080°F, decreasing by 58°F per 10,000 MWDIMTU of burnup."

The revised version of SL 2.1.A.1.b would read:

"The peak fuel centerline temperature shall be maintained <5080°F, decreasing by 9°F per 10,000 MWDIMTU of burnup."

I A mark-up of the proposed change to TS Section' 2.1 is provided in Attachment

\\ 2. A clean copy of the proposed change is provided in Attachment 3.

3.0 TECHNICAL EVALUATION

The principal design tool used by Westinghouse for evaluating fuel rod performance is the Performance Analysis and Design (PAD) code (Reference 1). This computer program iteratively calculates the interrelated effects of fuel and cladding deformations including fuel densification, fuel swelling, fuel relocation, fuel rod temperatures, fill and fission gas release (FGR), and rod internal pressure (RIP) as a function of time and linear power. PAD evaluates the power history of a fuel rod

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 3 of 10 as a series of steady-state power levels with instantaneous jumps from one power level to another. The length of the fuel rod is divided into several axial segments, and each segment is assumed to operate at a constant set of conditions over its length. Fuel densification and swelling, cladding stresses and strains, temperatures, burnup and fission gas releases are calculated separately for each axial segment, and the effects are integrated to obtain the overall fission gas release and resulting internal pressure for each time step. The coolant temperature rise along the fuel rod is calculated based on the flow rate and axial power distribution, and the cladding surface temperature is determined with consideration of corrosion effects and the possibility of local boiling.

Model updates incorporated into the PAD5 code address all of the fuel and cladding performance models required for high burnup fuel design. Key fuel performance updates to the PAD5 models include fuel thermal conductivity degradation (TCD) with burnup, enhanced high burnup athermal fission gas release (pellet rim effects),

and enhanced high burnup fission gas bubble swelling. Cladding creep and growth models are also updated to reflect high burnup cladding performance. In addition to high burnup analysis capability, a key driver for the implementation of the PAD5 models in fuel design is to address regulatory concerns associated with fuel thermal conductivity degradation with burnup.

The PAD5 models are the latest evolutions of the Westinghouse PAD code (Reference 1). As part of the Reference 1 development, the burnup-dependent term of the fuel melting limits in PAD5 was updated based on journal:-published fuel material data. Additional validation performed in Section 2.1 of Appendix A of Reference 1 shows that the PAD5 code in conjunction with the new fuel melt limit accurately predicts fuel melt based on comparisons to experimental observations. Section 3. 7 .12 of the NRC Safety Evaluation Report in Reference 1 concluded that the fuel melting limits in PAD5 are acceptable.

The peak fuel centerline temperature SL is independent of the PAD5 methodology (Reference 2). The current licensing basis safety analyses use the existing SL 2.1.A.1.b for fuel melt as an acceptance criterion as required by the current methodology. Thus, Dominion Energy Virginia will continue to meet the existing SL when using its current licensing basis safety analyses even with the implementation of the proposed SL. Since the existing SL for peak fuel centerline temperature is more restrictive than the proposed limit, the current licensing basis safety analyses remain conservative with respect to the proposed SL.

A comprehensive description of all PAD5 models, NRC Requests for Additional Information, and the subsequent NRC Safety Evaluation are documented in Reference 1. The NRC Safety Evaluation Limitations and Conditions are discussed in Section 3.1 of this amendment request. As described in Section 3.1, the proposed SL will only be applicable for analyses performed with the method described in Reference 1.

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 4 of 10 3.1 Limits of Applicability The proposed amendment will only be used in applicable safety analyses that are performed with the approved fuel performance methods in Reference 1.

The Limitations and Conditions from the NRC Safety Evaluation in Reference 1 pertinent to this amendment request are detailed below along with details of how each is satisfied.

• The NRG staff limits the applicability of the PAD5 code and methodology to the cladding, fuel, and reactor parameters listed in Section 4. 1 of the Safety Evaluation in Reference 1.

Response: Dominion Energy Virginia will apply PADS within the limits specified in Section 4.1 of Reference 1 for cladding, fuel, and reactor parameters to be used at Surry Units 1 and 2. Because these PADS inputs depend on the reload design, these parameters are validated on a cycle-specific basis.

• The application of PAD5 should at no time exceed the fuel melting temperature as calculated by PAD5 due to the lack of properties for molten fuel in PAD5 and other properties such as thermal conductivity and fission gas release.

Response: Dominion Energy Virginia will limit the peak fuel centerline temperature per this amendment request.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria

• Section 182a of the Atomic Energy Act requires applicants for nuclear power plant operating licenses to include TS as part of the license. The TS ensure the operational capability of structures, systems, and components that are required to protect the health and safety of the public.

• 10 CFR 50.90 requires NRC approval for any modification to, addition to, or deletion from the plant Technical Specifications. Therefore, this activity requires NRC approval prior to making the proposed plant-specific changes included in this license amendment request.

• 10 CFR 50.36 requires that the TS include items in the following specific categories: (1) safety limits, limiting safety systems settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements per 10 CFR 50.36(c)(3); (4) design features; and (5)

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 5 of 10 administrative controls. This amendment application is related to the first category above since a change to the peak fuel centerline melt temperature Safety Limit is proposed.

• 10 CFR 50, Appendix A, GDC 10, Reactor Design, requires that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. The GDC included in Appendix A to 10 CFR 50 did not become effective until May 21, 1971. The Construction Permits for Surry Units 1 and 2 were issued prior to May 21, 1971; consequently, Surry Units 1 and 2 were not subject to current GDC requirements (SECY-92-223, dated September 18, 1992). However, during the initial plant licensing of Surry Units 1 and 2, it was demonstrated that the design of the Surry reactor core met the regulatory requirements in place at that time. The draft GDC published in 1967 included Criterion 6, Reactor Core Design (Category A), which was the precursor to the current GDC 10, and included the requirement that, "The reactor core shall be designed to function throughout its design lifetime without exceeding acceptable fuel damage limits which have been stipulated and justified.

The requirements of both the current GDC-10 and the draft GDC-6 are met by the restrictions of SL 2.1.A.1.b that prevent overheating of the fuel and cladding by maintaining the steady state peak temperature below the level at which fuel centerline melting occurs. The change to Specification 2.1.A.1.b "Safety Limit, Reactor Core" changes the limit to be consistent with the limit approved in Westinghouse Topical Report WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PADS)," November 2017.

4.2 No Significant Hazards Consideration The proposed amendment would revise the Surry Units 1 and 2, TS Section 2.1, "Safety Limit, Reactor Core" associated with peak fuel centerline temperature.

As required by 10 CFR 50.91 (a)(1 ), Dominion Energy Virginia has performed an evaluation to determine whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," Part 50.92(c), and has determined that the proposed amendment does not involve a significant hazards consideration, as discussed below:

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 6 of 10

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No There are no design changes associated with the proposed amendment.

All design, material, and construction standards that were applicable prior to this amendment request will continue to be applicable.

The proposed amendment will not affect accident initiators or precursors or alter the design, conditions, and configuration of the facility, or the manner in which the plant is operated and maintained, with respect to such initiators or precursors.

Compliance with Safety Limit 2.1.A.1.b is required to confirm that fuel cladding failure does not occur as a result of fuel centerline melting. The fuel centerline melt temperature limit is established to preclude centerline melting. The proposed change has been reviewed by the NRG and found to be appropriately conservative with respect to the fuel material properties in WCAP 17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PAD5)."

Accident analysis acceptance criteria will continue to be met with the proposed amendment. The proposed amendment will not affect the source term, containment isolation, or radiological release assumptions used in evaluating the radiological consequences of any accident previously evaluated. The proposed amendment will not alter any assumptions or change any mitigation actions in the radiological consequence evaluations in the Updated Final Safety Analysis Report (UFSAR). Consequently, the applicable radiological dose acceptance criteria will continue to be met.

Therefore, it is concluded that the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No There are no proposed design changes nor are there any changes in the method by which any safety-related plant structures, systems, and components perform their specified safety functions. The proposed amendment will not affect the normal method of plant operation or change

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 7 of 10 any operating parameters. No equipment performance requirements will be affected. The proposed amendment will not alter any assumptions made in the safety analyses.

The proposed amendment revises Safety Limit 2.1.A.1.b; however, the change does not involve a physical modification of the plant.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures will be introduced as a result of this amendment.

There will be no adverse effect or challenges imposed on any safety-related system as a result of this amendment.

Therefore, it is concluded that the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The revised Safety Limit 2.1.A.1.b has been calculated based on the NRG-approved methods which ensure that the plant operates in compliance with all the applicable regulatory criteria.

There will be no effect on those plant systems necessary to perform protection functions.

No instrument setpoints or system response times are affected and none of the acceptance criteria for any accident analysis will be changed.

Consequently, the proposed amendment will have no impact on the radiological consequences of a design basis accident.

Therefore, it is concluded that the proposed amendment does not involve a significant reduction in a margin of safety.

Therefore, Dominion Energy Virginia concludes that the proposed amendment does not involve a significant hazards consideration, under the standards set forth in 10 CFR 50.92(c), "Issuance of Amendment," and accordingly, a finding of "no significant hazards consideration" is justified.

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 8 of 10 4.3 Conclusion In summary, in accordance with Title 10 of the Code of Federal Regulations (CFR) 50.90, Dominion Energy Virginia requests NRC review and approval of the change to Technical Specification Safety Limit 2.1.A.1.b for Surry Units 1 and 2.

Based on the above discussions, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 Environmental Considerations The proposed amendment would revise the Surry Units 1 and 2, TS Section 2.1 "Safety Limit, Reactor Core" associated with peak fuel centerline temperature.

A review of the anticipated construction and operational effects of the requested amendment has determined the requested amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), in that:

(i) There is no significant hazards consideration.

As documented in Section 4.2, No Significant Hazards Consideration Determination, of this license amendment request, an evaluation was completed to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment." The Significant Hazards Consideration determined that (1) the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) the proposed amendment does not create the 1possibility of a new or different kind of accident from any accident previously evaluated; and (3) the proposed amendment does not involve a significant reduction in a margin of safety. Therefore, it is concluded that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed changes are unrelated to any aspects of plant construction or operation that would introduce any changes to effluent types (e.g., effluents

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 9 of 10 containing chemicals or biocides, sanitary system effluents, and other effluents) or affect any plant radiological or non-radiological effluent release quantities.

The proposed amendment does not adversely impact any functions associated with containing, controlling, channeling, monitoring, or processing radioactive or non-radioactive materials, nor do they diminish the functionality of any design or operational features that are credited with controlling the release of effluents during plant operation. The types and quantities of expected plant effluents are not changed. No effluent relea$e path is associated with this amendment.

Neither radioactive nor non-radioactive material effluents are affected by this activity. Furthermore, the proposed amendment does not diminish the functionality of any design or operational features that are credited with controlling the release of effluents during plant operation. Therefore, it is concluded that the proposed amendment does not involve a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment does not affect plant radiation zones described in UFSAR Section 11 and controls under 10 CFR Part 20 preclude a significant increase in occupational radiation exposure. The proposed amendment does not adversely impact radiologically controlled zones. Plant radiation zones, radiation controls established to satisfy 10 CFR Part 20 requirements, and expected amounts and types of radioactive materials are not affected by the proposed amendment. Therefore, individual and cumulative radiation exposures are not significantly affected by this change. Therefore, the proposed amendment does not involve a significant increase in individual or cumulative occupational radiation exposure.

Based on the above review of the proposed amendment, it has been determined that anticipated construction and operational effects of the proposed amendment do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant \increase in the amounts of any effluents that may, be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b ), an environmental impact statement or environmental assessment of the proposed amendment is not required.

Serial No.20-341 Docket Nos. 50-280/281 Attachment 1 Page 10 of 10 6.0 References

1. Westinghouse Topical Report WCAP-17642-P-A, Revision 1, "Westinghouse Performance Analysis and Design Model (PADS)," November 2017.

2. Letter from the USNRC to Florida Power and Light Company dated August 15, 2019, "

Subject:

Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Issuance of Amendment Nos. 288 and 282 Regarding Revised Reactor Core Safety Limit to Reflect Topical Report WCAP-17642-P-A." (ADAMS Accession No. ML19031C891)

Serial No.20-341 Docket Nos. 50-280/281 Attachment 2 MARKED-UP TECHNICAL SPECIFICATIONS PAGE Virginia Electric and Power Company

{Dominion Energy Virginia)

Surry Power Station Units 1 and 2

TS 2.1-l O& 12 14 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of THERMAL POWER, Reactor Coolant System pressure, coolant temperature and coolant flow \Vhcn a reactor is critical.

Obiectiye To maintain the integrity of the fuel cladding.

Specification A. The combination of reactor THERMAL POWER level, pressurizer pressure. and Reactor Coolant System (RCS) highest loop average tempernture shall not:

I. Exceed lhc limiL,; spc,cified in the CORE OPERATING LIMITS REPORT when full flow from three reactor coolant pumps exists, and the following Safety Limits shall not be exceeded:

a. The design limit for departure from nucleate boiling rnlio (DNBR) shall be maimaincd.:;:. 1.27 for transients analyzed using the Statistical DNBR Evaluation Methodology and the WRB-1 DNB correlation. For transients analyzed using the deterministic methodology, the DNBR shall be maintained greater than or equal lo the applicable DNB correlation limit (~ 1.17 for \VRB-1, ?_ 130 for W-3,.:;:. 1.14 for ABB-NV).

+

b. The peak fuel centerline temperature shall be maintained< 5080"F, decreasing b y ~ per I0,000 rvtWD/MTl! of humup.

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2. The reactor THERMAL POWER level shall not exceed 118!;{, of rated power.

Amendment Nos. +/-8J. and~

Serial No.20-341 Docket Nos. 50-280/281 Attachment 3 PROPOSED TECHNICAL SPECIFICATIONS PAGE Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

TS 2.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of THERMAL POWER, Reactor Coolant System pressure, coolant temperature and coolant flow when a reactor is critical.

Objective To maintain the integrity of the fuel cladding.

Specification A. The combination of reactor THERMAL POWER level, pressurizer pressure, and Reactor Coolant System (RCS) highest loop average temperature shall not:

1. Exceed the limits specified in the CORE OPERATING LIMITS REPORT when full flow from three reactor coolant pumps exists, and the following Safety Limits shall not be exceeded:

a. The design limit for departure from nucleate boiling ratio (DNBR) shall be maintained,:::: 1.27 for transients analyzed using the Statistical DNBR Evaluation Methodology and the WRB-1 DNB correlation. For transients analyzed using the deterministic methodology, the DNBR shall be maintained greater than or equal to the applicable DNB correlation limit (2:: 1.17 for WRB-1, ,:::: 1.30 for W-3, 2:_ 1.14 for ABB-NV).

b. The peak fuel centerline temperature shall be maintained< 5080°F, decreasing by 9°F per 10,000 MWD/MTU of burnup.

2. The reactor THERMAL POWER level shall not exceed 118 % of rated power.

Amendment Nos.