Text

Units, Units 1 and 2 - Proposed License Amendment Request Administrative Changes to Subsequent Renewed Operating Licenses and Technical Specifications
	ML22171A013
Person / Time
Site:	Surry
Issue date:	06/20/2022
From:	Lawrence D; Virginia Electric & Power Co (VEPCO)
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	22-035
	Download: ML22171A013 (56)
	v • d • e

VIRGINIA ELEcrRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 June 20, 2022 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 PROPOSED LICENSE AMENDMENT REQUEST Serial No.:

NRA/GDM:

Docket Nos.:

License Nos.:

ADMINISTRATIVE CHANGES TO SUBSEQUENT RENEWED OPERATING LICENSES AND TECHNICAL SPECIFICATIONS22-035 RO 50-280 50-281 DPR-32 DPR-37 Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) proposes a change to the Surry Power Station (SPS) Units 1 and 2 Subsequent Renewed Operating Licenses (Ols), DPR-32 and 37, respectively, and the associated Technical Specifications (TS) to delete expired License Conditions (LCs) and TS that were previously included in the Surry Units 1 and 2 Ols and TS to support one-time plant modifications or the initial implementation of plant programs.

Administrative changes and correction of editorial errors are also being addressed. The proposed change is administrative in nature and will therefore not result in any changes to plant design or operation. Associated TS Basis changes are also included for the NRC's information. provides a description and assessment of the proposed change, and Attachments 2 and 3 provide marked-up and typed proposed Ols, TS and Basis pages, respectively.

Dominion Energy Virginia has evaluated the proposed administrative amendment request 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.

Due to the administrative nature of the proposed LAR, we have also determined that operation with the proposed change will not result in any significant increase in the amount of effluents that may be released off-site or any 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 proposed change has been reviewed and approved by the Facility Safety Review Committee. Dominion Energy Virginia requests approval of the proposed change by June 30, 2023, with a 30-day implementation period.

Serial No.22-035 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, Douglas C.

rence Vice President - Nuclear Engineering and Fleet Support Commitments contained in this letter: None Attachments:

1. Description and Assessment

2. Proposed Subsequent Renewed Operating Licenses, Technical Specifications, and Basis Pages (Marked-up)

3. Proposed Subsequent Renewed Operating Licenses, Technical Specifications, and Basis Pages (Typed)

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. Douglas C. Lawrence, 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 2.o-Ht day of 5 U-wz...,

My Commission Expires: _ 12--+f-~..... '/"""':1'--ii------

CRAIG D SLY Notary Public Commonwealth of Virginia Reg. # 7518653

,-i.f My Commission Expires December 31, 20_

2022.

cc:

U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 NRC Senior Resident Inspector Surry Power Station Mr. L. John Klos NRC Project Manager - Surry U.S. Nuclear Regulatory Commission One White Flint North Mail Stop 09 E-3 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 State Health Commissioner Virginia Department of Health James Madison Building - 7th floor 109 Governor Street Suite 730 Richmond, VA 23219 Serial No.22-035 Docket Nos. 50-280/281 Page 3 of 3 DESCRIPTION AND ASSESSMENT Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2 Serial No.22-035 Docket Nos. 50-280/281

DESCRIPTION AND ASSESSMENT Serial No.22-035 Docket Nos. 50-280/281

1.0 INTRODUCTION

Pursuant to 10 CFR 50.90, Virginia Electric and Power Company (Dominion Energy Virginia) proposes a change to the Surry Power Station (SPS) Units 1 and 2 Subsequent Renewed Operating Licenses (Ols), DPR-32 and 37, respectively, and the associated Technical Specifications (TS) to delete expired License Conditions (LCs) and TS that were previously included in the Surry Units 1 and 2 Ols and TS to support one-time plant modifications or the initial implementation of plant programs.

Administrative changes and correction of editorial errors are also being addressed by the proposed change.

The proposed change is administrative in nature and will therefore not result in changes to plant design or operation.

Associated TS Basis changes are also included for the NRC's information.

2.0 DETAILED DESCRIPTION OF THE PROPOSED CHANGE The proposed change to the SPS Units 1 and 2 Ols and accompanying TS: 1) deletes expired, obsolete information and requirements associated with previously completed plant modifications and plant program implementation requirements, and 2) makes administrative changes and corrects minor editorial errors in various sections of the TS.

The affected LCs and TS, and their associated license amendments or reference documents, as applicable, are provided in Table 1. A description of, and the reason for, each proposed change is provided in Section 2.

TABLE 1 Proposed Administrative Changes to the SPS Units 1 and 2 Subsequent Renewed Operating Licenses DPR-32 and 37 and Technical Specifications LC/TS Reason for Change Proposed Change Reference Documents Delete expired LC 3.R Completion of Main Control Delete expired LC3.R footnotes associated TS 1.D Room (MCR) and Emergency with:

Amendments TS 3.23.C.2.a.1 Switchgear Room (ESGR) 258/257 TS 3.23.C.2.b.1 Chilled Water Piping TS 1.D Replacement Project TS 3.23.C.2.a.1 TS 3.23.C.2.b.1 Page 1 of 11

TABLE 1 Serial No.22-035 Docket Nos. 50-280/281 Proposed Administrative Changes to the SPS Units 1 and 2 Subsequent Renewed Operating Licenses DPR-32 and 37 and Technical Specifications LC/TS Reason for Change Proposed Change Reference Documents Completion of first performance of MCR/ESGR Amendments LC3.S Envelope Habitability Program Delete expired LC 3.S 260/260 Testing, Assessment and Measurement Measurement Uncertainty Recapture (MUR) Power Amendments LC3.T Uprate License Condition 3. T Delete expired LC 3. T 269/268 commitments have been completed TS 3. 1 Pages:

Incorrect Unit 2 Amendment Correct Unit 2 License 3.1-13 No. 250 included on TS Amendment No. on Amendments 3.1 -14a Pages bottom of affected 267/266 3.1-14b associated TS pages Incorrect reference to TS Correct referenced TS Amendments TS 3.10 3.23.C in TS 3.10.A.13 and 14 section numbers 260/260 and TS 3.10.B.6 and 7 Completion of Unit 2 Reserve Delete expired Station Service Transformer footnotes and Basis Amendments TS 3.16 (RSST) C and Associated 5-discussion associated 293/293 and KV Cable Replacement with completed 297/297 modifications TS 4.9.A Incorrect reference to TS Correct referenced TS Amendments Section 3.7.E section to TS 3.7.D 228/228 TS 6.1.2.2.d Operations organizational Revise position title N/A position title change NRG Environmental TS 6.7.B Typographical error Correct typo Qualification Order dated October 24, 1980 Page 2 of 11

Serial No.22-035 Docket Nos. 50-280/281 2.1 Deletion of Expired License Conditions, Technical Specifications, and Associated Basis Text 2.1.1 LC 3.R, TS 1.D OPERABLE Definition Footnote, and TS 3.23.A.2.a.1 and b.1 Footnotes - Replacement of MCR/ESGR Air-Conditioning System (ACS)

Chilled Water Piping SPS Units 1 and 2 Amendments 258 and 257 (Reference 6.1) implemented a temporary LC and TS to facilitate the replacement of the MCR/ESGR ACS chilled water piping. Specifically, these amendments added LC 3.R to each unit's Operating License (OL) and 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 allowed outage time (AOT) extensions four times in a 24-month time span to permit replacement of degraded MCR/ESGR ACS chilled water piping. The amendments also revised the definition of OPERABLE in TS 1. D specific to the MCR/ESGR ACS air handling units on the operating chilled water loop by including a footnote to permit either the 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 associated design change package was closed out in June 2009. Furthermore, the 24-month time span permitted for use of the four AOT extensions began when the first AOT extension was entered on February 4, 2008, and therefore expired on February 4, 2010.

Consequently, LC 3.R and the footnotes associated with TS 1.0, 3.23.A.2.a.1 and 3.23.A.2.b.1 have expired and are therefore being deleted. (It should be noted that the TS 3.23 footnote refers to TS 3.23.C.2.a.1 and b.1.

TS 3.23.C was subsequently renumbered as TS 3.23.A by Amendments 260/260 as noted in Item 2.2.1 below.)

Related text in the TS 3.23 Basis is also being deleted as it refers to the work completed under the expired temporary TS AOTs.

2.1.2 LC 3.S - Control Room Envelope (CRE) Habitability Program SPS Units 1 and 2 Amendments 260/260 (Reference 6.2) incorporated OL and TS changes pursuant to Technical Specification Task Force (TSTF) Traveler TSTF-448, Revision 3, "Control Room Habitability." One of the changes was the incorporation of LC 3.S into the OLs which specified schedular requirements associated with: 1) the first performance of Surveillance Requirement (SR) 4.18, 2) the periodic assessment of MCR/ESGR envelope habitability specified in TS 6.4.R.3.b, and 3) the periodic measurement of MCR/ESGR envelope pressure as specified in TS 6.4.R.4. The first performance of these three requirements associated with CRE habitability surveillance, assessment, and measurement were completed as indicated in Table 2.1.

Since the first performance of the TS 4.18 surveillance, TS 6.4.R.3.b periodic assessment, and TS 6.4.R.4 periodic measurement of the CRE have been completed in accordance with the LC 3.S schedular requirements, LC 3.S is no longer required and is being deleted.

Page 3 of 11

Serial No.22-035 Docket Nos. 50-280/281 TABLE 2.1 -COMPLETION OF LICENSE CONDITION 3.5(1), (2), AND (3)

(Control Room Envelope Habitability Requirements)

LC#

LC Due Date Completion Date 3.S(1) 07/18/11 10/15/10 3.S(2) 06/24/09 06/18/09 3.S(3) 12/04/08 12/01/08 2.1.3 LC 3.T - Measurement Uncertainty Recapture (MUR)

SPS Units 1 and 2 Amendments 269/268 (Reference 6.3) implemented an MUR power uprate from 2546 to 2587 MWt. The license amendments included new LC 3.T that required the completion of sixteen commitments, fourteen of which were required to be completed prior to exceeding 2546 MWt as noted below.

TABLE 2.2-MUR License Condition Commitments and Due Dates COMMITMENT SCHEDULED COMPLETION DATE

1. VEPCO will perform the final acceptance of the Surry 1 /2 Prior to operating above uncertainty analysis to ensure the results are bounded by the 2546 MWt (98.4% RP) statements contained in the LAR

2. Technical Requirements Manual (TRM) will be revised to Prior to operating above include UFM administrative controls 2546 MWt (98.4% RP)

3. Revise procedures, programs, and documents for the new Prior to operating above UFM (including transducer replacement) 2546 MWt (98.4% RP)

4. Appropriate personnel will receive training on the UFM and Prior to operating above affected procedures 2546 MWt (98.4% RP)

5. The FAC CHECWORKS SFA models will be updated to Prior to operating above reflect the MUR PU conditions 2546 MWt (98.4% RP)

6. Simulator changes and validation will be completed Prior to operating above 2546 MWt (98.4% RP)

7. Revise existing plant operating procedures related to Prior to operating above temporary operation above full steady-state licensed power 2546 MWt (98.4% RP) levels Page 4 of 11

Serial No.22-035 Docket Nos. 50-280/281 TABLE 2.2-MUR License Condition Commitments and Due Dates

8. Process UFSAR changes in accordance with 10 CFR 50.59 In accordance with 10 CFR 50.71(e)

9. UFM commissioning and calibration will be completed Prior to operating above 2546 MWt (98.4% RP)

10. Confirm flow normalization factors Prior to operating above 2546 Mwt (98.4% RP)

11. Rescaling and calibration of main turbine first stage pressure Prior to operating above input to AMSAC 2546 Mwt (98.4% RP)

12. Determine EQ service life for excore detectors Prior to operating above 2546 Mwt (98.4% RP)

Unit 1: fall 201 O

13. The excore neutron detectors are scheduled to be replaced Refueling Outage Unit 2: spring 2011 Refueling Outage

14. Revise EOP setpoints Prior to operating above 2546 Mwt (98.4% RP)

15. The UFM feedwater flow and temperature data will be Prior to operating above compared to the feedwater flow venturis output and the 2546 MWt (98.4% RP) feedwater RTD output

16. For the applicable UFSAR Chapter 14 events, Surry 1/2 will re-analyze the transient consistent with VEPCO's NRC-Prior to operating above approved reload design methodology in VEP-FRD-42, Rev.

2546 Mwt (98.4% RP) 2.1-A.

As required by LC 3.T, Commitments 1 through 7, 9 through 12, and 14 through 16 were completed for each unit prior to exceeding 2546 MWt, which occurred on December 8, 2010, for SPS Unit 1 and on June 30, 2011, for SPS Unit 2.

The remaining two items, Commitments 8 and 13, required the UFSAR to be updated in accordance with 10 CFR 50.71 (e), and the Unit 1 and Unit 2 excore neutron detectors to be replaced by the completion of the fall 2010 and spring 2011 refueling outages (RFOs), respectively.

Commitment 8 was completed when the SPS Units 1 and 2 UFSAR was updated in Revision 43 dated September 27, 2011 (Serial No.11-459)

[ADAMS Accession No. ML11287A165] to reflect the MUR power uprate. Commitment 13 was met upon the replacement of the SPS Units 1 and 2 excore neutron detectors on November 30, 2010, during the Unit 1 fall 2010 RFO and on May 27, 2011, during the Unit 2 spring 2011 RFO, respectively. Consequently, LC 3.T is no longer required and is being deleted.

Page 5 of 11

Serial No.22-035 Docket Nos. 50-280/281 2.1.4 TS 3.16, "Emergency Power System," - Temporary Extended Allowed Outage Times (AOTsl for Replacement of RSST C and its associated 5-KV Cable SPS Units 1 and 2 Amendments 293/293 (Reference 6.4) implemented a temporary, one-time, 21-day AOT in a footnote to TS 3.16, Action 8.2, to permit replacement of the Unit 2 RSST C. Amendments 297/297 (Reference 6.5) revised 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 one-time, 14-day AOT was exited on May 24, 2020. Consequently, the temporary, one-time AOT footnote to TS 3.16, Action 8.2, has expired and can be deleted. Related text in the TS 3.16 Basis is also being deleted as it discusses the alternate dependable power source that would be used while RSST C was out of service, as well as the additional actions to be taken prior to entering the temporary, one-time AOT permitted by the footnote.

2.2 Administrative Changes and Corrections 2.2.1 TS 3.10, "Refueling" As noted above, SPS Units 1 and 2 Amendments 260/260 (Reference 6.2) incorporated TS changes pursuant to TSTF-448, Revision 3, "Control Room Habitability."

The amendments included a revision to then TS 3.23, "Main Control Room and Emergency Switchgear Room Ventilation and Air Conditioning System."

The amendments separated TS 3.23 into two separate TS sections: TS 3.21, "Main Control Room and Emergency Switchgear Room (MCR/ESGR) Emergency Ventilation System (EVS)," and TS 3.23, "Main Control Room and Emergency Switchgear Room Air Conditioning System." As part of that change, existing TS 3.23.A and B, which addressed the MCR/ESGR EVS, were revised and relocated to new TS section 3.21, while TS 3.23.C, which only addressed the MCR/ESGR Air Conditioning System (ACS) was renumbered as TS 3.23.A.

However, it was not recognized at the time those changes were made that TS 3.10 included references to TS 3.23.C, which had been subsequently renumbered as TS 3.23.A by Amendments 260/260.

Consequently, the proposed change revises TS 3.10.A.13 and 14 and TS 3.10.B.6 and 7 to reference TS 3.23.A instead of TS 3.23.C, which no longer exists.

2.2.2 TS 4.9, "Radioactive Gas Storage Monitoring System" SPS Units 1 and 2 Amendments 228/228 (Reference 6.6) implemented changes to the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS) instrumentation systems. The amendments revised the surveillance Page 6 of 11

Serial No.22-035 Docket Nos. 50-280/281 test intervals, AOTs, bypass times, and limiting conditions for operation (LCO) for those systems. As part of those changes, TS 3.7.A was deleted and the remaining sections were renumbered. This change resulted in TS 3.7.E, which was associated with the explosive gas monitoring instrumentation

channel, being renumbered as TS 3.7.D. However, it was not identified at the time the change was made that TS 3.7.E was referenced in TS 4.9, "Radioactive Gas Storage Monitoring System."

Consequently, the reference to TS 3.7.E in TS 4.9.A was not changed to reference the revised TS number 3.7.D.

Therefore, the proposed change revises TS 4.9.A to reference TS 3.7.D rather than TS 3.7.E.

2.2.3 TS 6.1.2, "Organization" TS 6.1.2 in the Administrative Controls section of the TS includes the Unit Staff requirements in TS 6.1.2.2. TS 6.1.2.2.d currently states, "The Supervisor Nuclear Shift Operations shall hold an active Senior Reactor Operator License for Surry Power Station." The title "Supervisor Nuclear Shift Operations" was subsequently changed to "Superintendent Nuclear Shift Operations."

The proposed change revises TS 6.1.2.2.d to reflect the current position title.

The position responsibilities and qualification requirements are unchanged.

2.2.4 TS 6.7, "Environmental Qualification" TS 6.7 was incorporated into the SPS Units 1 and 2 TS pursuant to an NRC Order dated October 24, 1980, "Orders for Modification of Licenses Concerning Environmental Qualification of Safety-Related Electrical Equipment" (Reference 6.7). When TS 6.7.B was included in the TS as required by the Order, an editorial error was inadvertently introduced.

Specifically, the first line of TS 6.7.B includes the words, "By on later than... " as opposed to the correct wording, "By no later than... " The proposed change corrects the TS wording discrepancy.

3.0 TECHNICAL EVALUATION

The proposed change to the SPS Units 1 and 2 Subsequent Renewed Ols and TS is administrative in nature and does not affect how plant equipment is operated or maintained.

The change deletes expired LCs and TS footnotes associated with completed plant modifications and programmatic implementation requirements, makes administrative changes to correct errors introduced in previous license amendments, and updates a personnel title. No changes to the physical plant or analytical methods are described, and there are no impacts on the Updated Final Safety Analysis Report (UFSAR) accident analyses.

Page 7 of 11

4.0 REGULATORY EVALUATION

4.1 Determination of No Significant Hazards Consideration Serial No.22-035 Docket Nos. 50-280/281 Virginia Electric and Power Company (Dominion Energy Virginia) has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed administrative change to the Surry Power Station (SPS) Units 1 and 2 Subsequent Renewed Operating Licenses (SROLs) DPR-32 and DPR-37 and the Technical Specifications (TS). The proposed change deletes expired License Conditions (LCs) and TS footnotes associated with completed plant modifications and programmatic implementation requirements, makes administrative changes to correct minor editorial errors introduced in previous license amendments, and updates a personnel title. In accordance with the criteria set forth in 10 CFR 50.92, Dominion Energy Virginia has performed an analysis of the proposed change and concluded that it does not represent a significant hazards consideration.

The following discussion is provided in support of this conclusion:

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

Response: No The proposed change to delete expired, temporary LCs and TS, correct minor TS administrative errors and update a personnel title does not impact the condition or performance of any plant structure, system, or component.

The proposed administrative change does not affect the initiators of any previously analyzed event or the assumed mitigation of accident or transient events. As a result, the proposed change to the SPS Units 1 and 2 SROLs and TS does not involve an increase in the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated since neither accident probabilities nor consequences are being affected by the proposed administrative change.

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

Response: No The proposed change to delete expired, temporary LCs and TS, correct minor TS editorial errors and update a personnel title is administrative in nature, and therefore does not involve any changes in station operation or physical modifications to the plant. In addition, no changes are being made in the methods used to respond to plant transients that have been previously analyzed. No changes are being made to plant parameters within which the plant is normally operated or in the setpoints that initiate protective or mitigative actions, and no new failure modes are being introduced. Therefore, the proposed administrative change to the Surry SROLs and Page 8 of 11

Serial No.22-035 Docket Nos. 50-280/281 TS does not create the possibility of a new or different kind of accident or malfunction of equipment important to safety from any previously evaluated.

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

Response: No The proposed change to delete expired, temporary LCs and TS, correct minor TS editorial errors and update a personnel title is administrative in nature and therefore does not affect the acceptance criteria for any analyzed event, nor is there a change to any safety limit. The proposed change does not affect any structures, systems or components or their capability to perform their intended functions and does not alter the way safety limits, limiting safety system settings or limiting conditions for operation are determined.

Neither the safety analyses nor the safety analyses acceptance criteria are affected by the proposed change. Therefore, the proposed administrative change to the Surry SROLs and TS does not involve a reduction in a margin of safety.

4.2 Conclusion In conclusion, based on the considerations discussed above, (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.

4.3 Precedents Numerous licensees have submitted LARs to make administrative and editorial changes including the deletion of expired, obsolete and/or completed requirements contained in their Ols and TS. A sample of these LARs is provided below:

By letter dated March 31, 2021, DTE Electric Company submitted an LAR for Fermi 2 Power Plant to revise its TS to remove obsolete information, make minor corrections, and make miscellaneous editorial changes.

(ADAMS Accession No. ML21090A194)

By letter dated June 12, 2018, Energy Northwest submitted an LAR for Columbia Generating Station to clean-up its OL and TS.

(ADAMS Accession No. ML18163A351)

By letter dated February 23, 2017, Exelon Generation Company, LLC, submitted an administrative LAR for the Braidwood and Byron Power Stations' Units 1 and 2 to remove time, cycle, or modification-related items from their respective Ols and TS Page 9 of 11

Serial No.22-035 Docket Nos. 50-280/281 and to make editorial and formatting changes.

(ADAMS Accession No.

ML17055A631)

By letter dated February 12, 2016, NextEra Energy Point Beach, LLC, submitted an administrative LAR for Point Beach Nuclear Plant that removed LCs that had been completed and were no longer in effect. (ADAMS Accession No. ML16043A217)

By letter dated July 11, 2014, Exelon Generation Company, LLC, submitted an administrative LAR for Peach Bottom Atomic Power Station, Units 2 and 3, which deleted an LC associated with a one-time commitment. (ADAMS Accession No. ML141928143) 5.0 ENVIRONMENTAL ASSESSMENT This amendment request meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) as follows:

(i) The amendment involves no significant hazards consideration.

As described above, the proposed administrative change does not involve a significant hazards consideration.

(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 administrative change to the SPS Units 1 and 2 SROLs and TS does not involve the installation of any new equipment or the modification of any equipment that may affect the types or amounts of effluents that may be released offsite.

Plant operation is not affected in any manner by this proposed administrative change. Therefore, there is no significant change in the types or 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 administrative change does not involve plant physical changes or changes in the method of plant operation.

Therefore, there is no significant increase in individual or cumulative occupational radiation exposure.

Based on the above assessment, Dominion Energy Virginia concludes that the proposed administrative change meets the criteria specified in 10 CFR 51.22 for a categorical exclusion from the requirements of 10 CFR 51.22 relative to requiring a specific environmental assessment or impact statement by the Commission.

Page 10 of 11

6.0 REFERENCES

Serial No.22-035 Docket Nos. 50-280/281 6.1 Letter from US NRC to Virginia Electric and Power Company dated January 23, 2008, "

Subject:

Surry Power Station, Unit Nos. 1 and 2, Issuance of Amendments Regarding Replacement of Main Control Room and Emergency Switchgear Room Air-Conditioning System Chilled Water Piping (TAC Nos.

MD4622 and MD4623)," [ADAMS Accession No. ML073480287].

6.2 Letter from US NRC to Virginia Electric and Power Company dated July 7, 2008,

Subject:

Surry Power Station, Unit Nos. 1 and 2, Issuance of Amendments Regarding Control Room Habitability (TAC Nos. MD6139 and MD6140),"

[ADAMS Accession No. ML081750690].

6.3 Letter from US NRC to Virginia Electric and Power Company dated September 24, 2010, "

Subject:

Surry Power Station, Unit Nos. 1 and 2, Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (TAC Nos. ME3293 and ME3294)," [ADAMS Accession No. ML101750002].

6.4 Letter from US NRC to Virginia Electric and Power Company dated October 5, 2018, "

Subject:

Surry Power Station, Unit Nos. 1 and 2, Issuance of Amendments Revising Technical Specifications Section 3.16, 'Emergency Power System,' for a Temporary 21-Day Allowed Outage Time (EPID L-2017-LLA-0380)," [ADAMS Accession No. ML18261A099].

6.5 Letter from US NRC to Virginia Electric and Power Company dated March 16, 2020, "

Subject:

Surry Power Station, Units 1 and 2, Issuance of Amendments Nos. 297 and 297, to Revise Technical Specifications 3.16, 'Emergency Power System,' to Allow a One-Time 14-Day Allowed Outage Time for Replacement of 5-KV Cables Associated with Reserve Station Service Transformer C (EPID No.

L-2019-LLA-0244)," [ADAMS Accession No. ML20058F966].

6.6 Letter from US NRC to Virginia Electric and Power Company dated August 31, 2001, "

Subject:

Surry Units 1 and 2 - Issuance of Amendments Re: Changes to Surveillance Test Intervals and Allowed Outage Times for Instrumentation Systems (TAC Nos. MA9355 And MA9356)," [ADAMS Accession No.

ML012480506].

6.7 Letter from US NRC to Virginia Electric and Power Company dated October 24,

1980, "Orders for Modification of Licenses Concerning Environmental Qualification of Safety-Related Electrical Equipment," [ADAMS Accession No. ML012490039).

Page 11 of 11 Serial No.22-035 Docket Nos. 50-280/281 PROPOSED SUBSEQUENT RENEWED OPERATING LICENSES, TECHNICAL SPECIFICATIONS, AND BASIS PAGES (MARKED-UP}

Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2

~

Deleted by Amendment XXX 05 04 21 (3 ) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders P& disoussed in the footnote to Teehnical Specifications a.2a.c.2.a.1 and 3.23.C.2.b.1, the use of temporary 45 day and 14 day allowed outage time e3Etonsions to permit replaoement of the Main Control Room and Emergenoy Switchgear Room Air Conditioning System chilled *Nater piping shall be in accoFdance 'Nith the basis, risk evaluation, equipment unavailability restrictions, and compensatory actions Deleted by pro>Jided in the licensee's submittal dated February 26, 2007 (Serial No. 07 0109) and

._A_m_e_nd_m_e_nt_x_x_x_. in the associated supplemental transmittals, as appro>Jed by tho NRG Safety Ewluation.

Upon implementation of Amendment No. 260 adopting TSTF 448, Revision 3, the determination of Main Control RoomlEmer-genoy Switchgear Room (MCRiESGR) envelope unfiltered air inlealmge as required by TS SR 4.18 in aocoFdance *Nith TS 6.4.R.8.a, the assessment of MCRiESGR envelope haeitability as required by Specification 6.4.R.3.b, and the measurement of MCRlESGR envelope pressure as required by Speoification 6.4.R.4, shall be oonsidered met. Following implementation:

(1) The first performanoo of SR 4.18, in acoordance *with Specification 6.4.R.3.a, shall be within the specified frequenoy of 6 yearo plus the 18 month allowanoe of SR 4.0.2, as measured from January 18; 2004, tho date of the most reoent suooossful traoer gas test, as stated in the April 22, 2004 letter response to Generic Letter 2003 01, or within the no3Et 18 months if tho time period since the most recent successful tracer gas test is greater than 6 years.

(2) The first performanoe of the pmiodic assessment of MCRlESGR envelope habitability, Specifioation 6.4.R.8.b, shall be within a years, plus the 9 month allowanoe of SR 4.0.2, as measured from January 18, 2004, the date of the most reoent suocessful traoer gas test, as stated in the April 22, 2004 letter response to Generic Letter 2003 01, or within the no3Et 9 months if the time period since the most reoent sueoessful traoer gas test is greater than 3 years.

(8) The first performance of the periodic measurement of MGRlESGR en*,*elope pressure, Specification 6.4.R.4, shall be within 18 months, plus tl'le 138 days allmved by SR 4.0.2, as mea&ured fFom January 19, 2007, tl'le date of the most resent successful pressure measurement test, OF within 138 days if not performed previously.

Surry - Uriif f Subsequent Renewed License No. DPR-32

05 04 21 T.

Deleted by Amendment XXX SCHEDULED COMMITMENT CQMPLETION DATE

1. VEPGQ 'Nill 13eFfeFm tl=le JiAal aeee13taAee ef PFieF te e13eFatiA9 aee111e Sl:IFP/ 1 1:1AeeFlainty aRalysis te ens1:1re the 2646 MWt (98.4% RP).

Fes1:llts aFe ee1:1Aefeef ey U=ie statemeAts eeRtaiAeef iA the LAR (AttaehmeRt 6, SeetieR U.9.4.1).

2. +eehnieal Reei1:1iFements MaR1:ml (+RM) will PFieF te e13eratin9 aeo*.. e ee re*.. iseel to ineh:1ele YFM aaministratitff0 2646 MWt (98.4% RP).

eeRtFels (AttaehmeRt 1 SeetieA a.o).

a. Re111ise 13Foeeef1:1res, 13Fe9rams, aRef PFieF te 013eratiR§ aee111e elee1:1meRts feF the RS'N YFM (iRel1:1efiA9 2646 MWt (98.4% RP).

tFaRsEl1:1eeF re13laeemeRt) (AttaehmeAt 5, Seetions 1.1, 1.1.D.1, 1.1.H, '111.1, '/11.2.A, anef '/11.4).

4. Af3f3F8f3Fiate f39FS9Anel,,,ill rneei 11e tFaiRiR§ PFieF ta e13eratiR§ aeeve eR the YFM aRef affeeteel 13FoeeE11:1res 2646 MWt (98.4% RP).

(Attaehment 6, SeetioAs 1.1.9.1.1, Vll.2.A, Vll.2.9, anEI VII.a).

6. +Re FAG GHEGWQRKS SF-A meElels 1Nill PFieF ta e13eFatiR9 aee11e ee 1:1J3elatea te reJleet Um MblR PY 2646 M'.'1/t (98.4 % RP).

eeRelitieAs (Attaet:imeAt 6, SeetieR IV.1.e.iii).

6. Sim1:1lateF ehan9es ana valiefatien will 13e PFieF ta e13eFatin9 aeeve eem13leteel (Attaet:imeRt 4, SeetieR Vll.2.G).

2646 M'A<t (98.4% RP).

7. Re\\1ise e*istiA§ 13laRt e13eratin9 13Feeeef1:1Fes PFieF ta e13eFatin9 al3e11e Felateef te tem13eraP,i ef3eFatieR al3e¥e f1:1II 2546 M\\'\\<t (98.4% RP).

steaely state lieenses f3SWeF le11els (Attaehment 5, Seetien '111.4).

8. PFSeess blFSAR ehaA9es in aeeeFElaRee IA aeeeFelaAee with with 1 0 GFR 50.69 (AttaohmeRt 1, Seetien 10 GFR 50.71 (o)

&-G}-:

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

05 04 21 T. (Continued)

SCHEDULED COMPLETION COMMl+MEN=i=

BA=FE

9. Yl=M eemmissiening ana ealisration 1.,.,m ee PFieF te e13erating aeeve com13lctca (Attaohmcnt 5, 2546 M'Nt (98.4% RP).

Scetion 1.1.9.2.1).

10. GonfiFm flew noFmalii!ation faeteFS PFioF to 013eratin§ aeeve (Attaehment 5, Seetien U.G).

2546 M'.\\!t (98.4% RP).

11. Resealing ana calieFatien of main tuFeine PFioF to 013eFating abo\\*e fiFSt stage 13Fess1:1Fe in13ut te AMSP,G 2546 MWt (98.4% RP).

(Attaohment 5, Seotiens 11.2.28, Vll.2.B,

'Jlll.2, ana VIII.a).

12. 9eteFmine EQ seFYiee life feF C*SOFC PFieF te 9(3CFating aeO\\le deteeteFs (Attaohment 5, Seotions I11.2.A 2546 M'Nt (98.4% RP).

anePJ.1.G).

1 a. +t:ie e:lEeere neutFen aeteetoFS aFe l:Jnit 1: fall 201 0 scheduled to be Fe13laced (Attaehment 5, Reh:1elin§ Gutage.

Section V.1.G).

Unit 2: s13ring 2011 Refueling Gt:1ta§e.

14. Re*1ise EGP sot13eints (Attaohment 5 PFioF te e13eratin§ abe¥e Seetien Vll.2.A).

2546 M1/4tt (98.4% RP).

15. +AO UFM feeelwateF flo1.,, and tem13eFatt:1Fe PFioF to 0(3emtin§ aeo*1e aata will be eom(3aFOa to the feeeiwateF flo1.*.i 2546 MWt (98.4% RP).

ventuFis out(3ut and the feeelwateF R+9 01:1t131:!t (Attaehment 5, Scotian 1.1.9.2.1 ).

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

05 04 21 T. (Continued)

SCHEDULED COMMITMENT COMPLETIQN DATE 1 e. FeF the a1313lieaele IJFSAR Ct=ia13teF 14 PFieF te e13erating aee*1e events, S1:::1Fiy 1 1*.1ill Fe analYz!e tl=ie transient 2646 MWt (98.4% RP).

eensistent with liEPCQ's NRG a1313Fovea Felead design methodology in VEP FRO 42, Rev. 2.1 A Yi-If NRG Feview is deemes neeessaiy l3UFSl:tant to U:10 Feei1:::1iFOments ef 1 0 CFR 60.69, tt:ie aeeident analyses will ee suemitted to U::10 NRG foF F011iew 13FioF to 0130Fation at the u13Fate 13eweF level. These sommitments a1313ly to U=te folloi.*.1ing SuFFy 1 YFSAR Cha13teF 14 QNBR analyses tl=iat

'NCFe analYz!Cd at 2§46 M\\6i.(t eonsistent *Nith U=ie Statistieal DNBR E~iah,1ation Methedelegy in VEP NE 2 A:

Seetion 14.2.7 E*eessi*~*e l=leat Remo*~ial d1:::1e te Feedi,*.iateF System Malfunetions (Full PoweF FeedwateF Tem13eratme Red1:::1etion ease only);

Seetien 14.2.8 E*6essi*1e bead lnornase lnsidont;

Seetien 14.2.9 boss of ReaetoF Coelant Flow; and

Scotian 14.2.10 boss of E~Etemal EleetFieal bead 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 (PAA) 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 (AN0-2) 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 PAA Standard RA-Sa-2009; as specified in License Amendment No. 301 dated December 8, 2020.

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

Deleted by Amendment XXX (3) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders 05 04 21 As discussed in the footnote to Technical Specifications 3.23.C.2.a.1 and 3.23.C.2.b.1, the use of temporary 45 day and 14 day allowed outage time

..------ --, extensions to permit replacement of the Main Control Room and Emergency Switchgear Room Air Conditioning System chilled 'Nater piping shall be in accordance with the basis, risk CP.<aluation, equipment unavailability restrictions, and Deleted by Amendment XXX

-r----'

compensatory actions pF0'1ided in the licensee's submittal dated February 26, 2007 (Serial No. 07 0109) and in the associated supplemental transmittals, as approved by the NRG Safety E*mluation.

Upon implementation of Amendment No. 260 adopting TSTF 448, Revision 3, the determination of Main Control Room/Emergency S*Nitehgear Room (MCR/ESGR) en>1elope unfiltered air inleakage as required by TS SR 4.18 in accordance with TS 6.4.R.3.a, the assessment of MCR/ESGR envelope habitability as required by Specification 6.4.R.3.b, and the measurement of MCR/ESGR envelope pressure as required by Speoifieation 6.4.R.4, shall be considered met. Following implementation:

(1) The first performance of SR 4.18, in accordance with Specification 6.4.R.3.a, shall be within the specified frnquenoy of 6 years plus.the 18 month all01Nance of SR 4.0.2, as measured from January 18, 2004, the date of the most recent successful tracer gas test, as stated in the April 22, 2004 letter response to Generic Letter 2003 01, or 'Nithin the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(2) The first performance of the pOFiodic assessment of MCR/-ESGR envelope habitability, Specification 6.4.R.3.b, shall be wilhin 3 years, plus the 9 month allowance of SR 4.0.2, as measured from January 18, 2004, the date of the most recent successful tracer gas test, as stated in the April 22, 2004 letter response to GenOFic Letter 2003 01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(3) The first performance of the periodie measurement of MCR/ESGR envelope pressure, Specification 6.4. R.4, shall be within 18 months, plus the 138 days allowed by SR 4.0.2, as measured from January 19, 2007, the date of the most recent successful pressure measurement test, or 'Nithin 138 days if not performed previously.

Surry - Unit 2 Subsequent Renewed License No. DPR-37

05 04 21 T.

Deleted by Amendment XXX SCHEDULED COMMITMENT COMPLETION QA-l=E

1. VEPGG *.-..ill 13eFfeFm tl=ie fiRal aeee13taRee ef PFieF te e13eratiR§ aee*.<<e 8l::IFP/ 2 l::IReeFtaiRPy aRalysis te eRSl::IFe tl=ie 2646 MWt (98.4% RP).

Fes1::1lts aFe ee1::1Reteel ey U:ie statemeRts ceRtaiReel in the LAR (Attacl=imeRt 5, Seetian 1.1.D.4.1).

2. +eehnieal Reet1::1iFements Mamial (+RM) will PFiaF ta a13eratin§ aea*.re ee Fe1.<<iseel ta iRcl1::1ele IJFM aaministrati*.<<e 2646 M'JVt (98.4% RP).

cantFals (Attaehment 1 Seetien a.0).

a. Re1t<ise 13Faeeel1::1Fes, 13Fa§rams, anEI PFieF ta a13eFatin§ aaeve eiae1::1meRts feF the Re*,¥ IJFM (inel1::1EliR§ 2546 M'Nt (98.4% RP).

transEl1::1ceF Fe13laeement) (Attachment 5, Seetions 1.1, 1.1.D.1, 1.1.H, Vll.1, Vll.2.A, anEI l.!11.4).

4. P,13pFOl3Fiate 13eFSanRel \\*till Feeeive trainin§ PFieF te e13eratiR§ aae*.<<e en the IJFM anet affeeteEI 13reeeEl1::1Fes 2546 MV'.'t (98.4% RP).

(Attaehment 5, Seetions 1.1.D.1.1, Vll.2.A, Vll.2.D, anEI lJll.a).

5. +he FP,G GHEGVi.<QRKS SFA maelels 1Nill PFioF te e13eratin§ aaeve se 1::113ElateEI te Fefleet tl=ie MIJR PIJ 2646 M'A't (98.4% RP).

eenelitions (Attachment 5, Seetien IV.1.e.iii).

6. Sim1::1latoF ehan§es anEI \\<aliElatien,.,,,m ae PFieF te e13eFatiR§ aeeve cem13leteel (Attael=imeRt 4, Seetien \\tll.2.G).

2546 M\\6Jt (98.4% RP).

7. Revise e*istiR§ 13lant e13eratiA§ 13rnceEl1::1Fes PFiSF te 9f39FatiR§ aee*.<<e Felateel te tem13eraPy e13eFatien aee*.*e f1::1II 2546 MWt (98.4% RP).

steaEly state lieenses 13eweF le*.<<els (Attael=iment 5, Seetion Vll.4).

8. PFoeess IJFSAR el=ian§es in aeeoraanee In aeeoFelanee witl=i with 1 O GFR 50.59 (Attael=iment 1, 1 0 GFR 50.71 (e)

Seetion a.Q).

Surry - Unit 2 Subsequent Renewed License No. DPR-37

05 04 21 T. (Continued)

SCHEDULED COMMITMENT COMPbE+ION DA+E g_ UFM eommissionin§ aAEI eali13mtion will 13e PFioF to 01:3erat:in§ al3o¥e I

-* I A I -t n I'\\ -t\\

n,:-Aa....,~,~~ AO/ nn,

~-* *r

- \\'.. *-

LI*-* J * --

- -.,.... \\---....,..,.

10. Gonfimi flo*N noFmalii!ation faeters PFiOF to Oj3efat:iR§ aBOJ/0 (Attaehment 5, Seetion U.G).

2546 MWt (98.4% RP).

11. Resealin§ and ealieFation of main tuFeine PFiOF to Oj3eFatin§ aeo11e first sta§e 1:3FessuFe in13ut to AMSAG 2546 M\\¥t (98.4% RP).

(Attaohment 6, Sections 11.2.28, Vll.2.B, Vlll.2, and VIII.a).

12. getemiine E:Q seFViee life leF e*eoFO PFioF to 01:3erat:in§ aeo>ii<<e deteetora (Attashment 5, Seotions 111.2.A 2546 MWt (98.4% RP).

ana V.1.G).

1 a. +he e*eeFe neutFGn aeteetoFS arn Unit 1 : fall 201 O Refueling seheduled to so Fe13laood (Attaehment 5, Guta§e.

Seetion V.1.G).

Unit 2: Sj3FiA§ 2EH 1 Rofuelin§ Guta§e.

14. Revise EQP sot1:3oints (Attaohment 5 PFiOF to O1:3erat:in§ ae0\\'0 Seetion Vll.2.A).

2546 MWt (98.4% RP).

15. +he UFM feeawateF flo*N and tem1:3erat:uFe PFioF to 01:3erat:in§ above aata 'Nill ee GOFRj3aFea to the feeB'A'ffiOF flow 2546 M'.'\\'t (98.4% RP).

ventuFis out131:1t ans the feeetwateF R+9 out13ut (Attaehment 5, Seetion l.1.9.2J ).

Surry - Unit 2 Subsequent Renewed License No. DPR-37

06 04 21 T. (Continued)

SCHEDULED COMMITMENT GOMPLE=l'.ION DATE

16. FoF the a1313lieaele YFSAR Gtm~eF 14 0-'t'ents, Sl;lFPj PFiaF to 013emtin§ aBO'le 2 *Nill Fe analy;rn the tFansient eonsistent *Nith 2546 MWt (98.4% RP).

VEPGG's NRG a1313ro,.*ea reloaa aesign methoaology in VEP FRD 42, Rev. 2.1 A.

If NRG Fei.tiew is aeemea neeessai:;* 13ursuant te the FOEJuirements ef 19 GFR 59.59, the aeeident analyses *Nill ee suemittea to the NRG foF rei.tiew 13FioF to 013emtien at the u13Fate 13eweF l0-'1el. +!=Iese eommitments apply te the folle>Ning SUFPj 2 l:JFSAR Gl=lapteF 14 DNBR analyses that were anal~ed at 2546 M1Aft eonsistent witt:i the Statistieal E}NBR Ewl1;1ation Metl=loaolo§y iA VEP NE 2 A:

Seetion 14.2.7 E>Eeessive Heat Reme*.*al Elue to Feed*.vateF System Malfunetiens (Full Poi.*,ieF FeeawateF

+em13eraturn Reduetion ease only);

Seetion 14.2.8 E>messi*.*e Loaa lnerease lneident;

Seetion 14.2.9 Loss of ReaetoF Coolant Flow; ana

Seetion H.2.19 boss of e~eFnal EleetFieal Loaa 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 (AN0-2) 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 2 Subsequent Renewed License No. DPR-37

5.

REACTOR CRITICAL TS 1.0-2 01 23 08 When the neutron chain reaction is self-sustaining and keff = 1.0.

6.

POWER OPERATION When the reactor is critical and the neutron flux power range 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, train, component, or device shall be operable or have operability when it is capable of performing its specified function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and""'emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support 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 has been tested periodically in accordance with Section 4 and meets its performance requirements.

E.

PROTECTIVE INSTRUMENTATION LOGIC I.

ANALOG CHANNEL An arrangement of components and modules as required to generate a single protective action digital signal when required by a unit condition. An analog channel loses its identity when single action signals are combined.

"' For the purpose of perfofffling Technical Specification fCEJ.ttirecl sHn*eillances that render aa emergency cliesel geaerator iHOJJernble, the clefiaitioa of OPERABLE for the air hat1dHng units on the operating chi.Ilea *uatcr loop is moclified k> reEJ.uire the Hormal or emergency eleetrieal po1uer soHrce to be capable of performing its related support funetioa. This footftote shall only apply to TS 3.23.C.2.a.l, TS 3.23.C.2.b.1, a.Hd TS 3.23.C.2.e.1 duriHg the 45 day allO\\ved outage time exteasions to permit replacement of the Maia CoatroJ Room a.ad Emergeaey Switchgear Room,'\\ir Conditioaiag System chilled water pipiHg.

Amendment Nos. 258 a.Rd 257

C.

RCS Operational LEAKAGE Applicability TS 3.1-13 11-05-09 The following specifications ate applicable to RCS operational LEAKAGE whenever Tavg (average RCS temperature) exceeds 200°F (200 degrees Fahrenheit).

Specifications

1.

RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c.

10 gpm identified LEAKAGE, and

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

2.a. If RCS operational LEAKAGE is not within the limits of 3.1.C.l for reasons other than pressure boundary LEAKAGE Qr primary to secondary LEAKAGE, reduce LEAKAGE to within the specified limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. If the LEAKAGE is not reduced to within the specified limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the unit shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.

If RCS pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within the limit specified in 3.1.C. l.d, the unit shall be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30hours.

~

Amendment Nos. 267 and~

TS 3.l-14a 11-05-09 This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE SAFETY ANALYSES - Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage* can affect the probability 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 requirement to limit primary to secondary LEAKAGE through any oneSG to less than orequal to 150 gallons 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 steam 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. Releases continue through the unaffected steam generators until the Residual Heat Removal System is placed in service.

The MSLB is less limiting for site radiation releases than the SGTR. The safety analysis for the MSLB accident assumes I gpm total primary to secondary 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 l O CFR 50.36( c)(2)(ii).

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

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

LEAKAGE of this type is unacceptable as the leak it~elf could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals apd gaskets is not pressure boundary LEAKAGE.

~

Amendment Nos. 267 and~

b. Unidentified LEAKAGE TS 3.1-14b 11-05-09 One gallon per minute (gpm) of unidentified LEAK.AGE is allowed as a reasonable minimwn 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 normai 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 sha11 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 Tavg exceeds 200°F, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

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

LCO 3. l.C.5 measures leakage through each individual pressure isolation valve (PIV) and can impact this LCO. Of the two PIVs in series.in each isolated line, leakage measured through one PN does not result in RCS LEAKAGE when the other is leak.tight. 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. 267 and~

TS 3.10-4 10 29 09

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.

A~

13. Three chillers shall be OPERABLE in acco*r,.n,...__.-- with the pow supply requirements of Specification 3.23.G.

1th one of the required O RABLE chillers inoperable or not powered as required by Specification 3.23.G. l, return the inoperable chiller to OPERABLE status within 7 days or comply with Specification 3.1 0.C. With two of the required OPERABLE chillers inoperable or not powered as required by Specification 3.23.G. l, comply with Specification 3.10.C.

~

14. Eight air handling units (AHUs) shall be OPERABLE in accordance with the operability requirements of Specification 3.23.. With two AHUs inoperable on the shutdown unit, ensure that one AHU is O ERABLE in each unit's main control room and emergency switchgear room, an restore an inoperable AHU to OPERABLE status within 7 days, or comply wit Specification 3.10.C. With more than two AHUs inoperable, comply with Spec1 1cation 3.10.C.

Amendment Nos. 266 emd 265 t

TS 3.10-5 10 29 09 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 continuously monitored 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

6. ::m::i:::~:n ::::: b:e 0

*spe:: A d~::: :~:* the pow~

pply requirements of Specification 3.23.. With one of the required O~ Ri~LE chillers inoperable or not powered as required by Specification 3.23.G.1, return the inoperable chiller to OPERABLE status within 7 days or comply with Specification 3.1 0.C. With two of the required OPERABLE chillers inoperable or not powered as required by Specification 3.23.G.l, comply with Specification 3. IO.C.

~

Amendment Nos. 266 aAd 265

TS 3.10-6

~

070708

7.

Eight air handling units (AHUs) shall be OP~

BLE in accordance with the operability requirements of Specification 3.23.G. With two AHUs inoperable on either unit, ensure that one AHU is OPERABLE in each unit's 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 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 Building shall cease and irradiated fuel shall be placed in a safe position, work shall be initiated 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 interlocks and safety features, provide assurance that an accident, which would result in a hazard to public health and safety, will not occur during unit REFUELING OPERA TIO NS or irradiated fuel movement in the Fuel Building. When no change is being made in core geometry, one neutron detector is sufficient to monitor the core and permits 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 pressurization 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 components into and out of the containment. During REFUELING OPERATIONS, the equipment hatch must be capable of being closed.

Amendment Nos. 260 and 260

TS 3.16-3 03 16 20

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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . If specification A-4 is not satisfied within seven (7) days, the unit shall be brought to COLD SHUTDOWN.f'19 I

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

b.

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

(*) To foeilite.te the reple.eeffteftt of the ReseFVe 8te.tiott Serviee Transfoffiler C 5 KV ee.bles to transformer bus F during the Spring 2020 UHit 2 refuelittg outage, the use of a tefflf)Oflll)',

oHe time, 14 day allowed oetage time (AOT) is permitted for the t1tta*1ailability of a primary source. P-rior to entry into Bftd dl:lring the 14 day AOT, the foUo*.ving e.etions shall he takee:

Withie 30 days prior to enteriflg the temporary 14 day A.OT, fueetioHality of Hie Alternate A.C (.'\\AC) System (i.e., the supplemeHtal power source) shall be verified.

DttriHg the 14 day AOT, the functionality of the AA.C System shall he checked once per shift. If the /\\iA£ System beeomes noH funetiomtl at a.Hy time deriHg the 14 day AOT, it shall he restored to fuHetione.l status within 24 hottrs, or the tmit sha1l he brottght to HOT SHUTDOWN 1tvithin the next 6 hotirs e:Hd COLD SHUTDOWN within the follo'+ving 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Amendment Nos.297 aed 297

TS 3.16-7 03 16 20 TS action statement 3.16.B. l.a.2 provides an allowance to avoid unnecessary testing of an OPERABLE EDG(s). If it can be determined that the cause 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 the applicable required action(s) would be entered. Once the failure is repaired, the common cause failure 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 confirmed not to exist on the remaining EDG(s),

performance of the operability test within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides 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 requirement for the OPERABLE EDG(s), the corrective action program will continue to evaluate the common cause possibility, including the other unit's EDG or the shared EDG. This continued evaluation, however, is no longer under the 24-hour constraint imposed by the action statement.

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

Reserve Station Service TflHlsfo11B0f (RSST) C is the primary offsHe powef souree fof the lH tlftd 2J Emergency Buses Yia transfer bus F. To facilitate the replaccmeRt of the RSST C 5KV cables to tfaRsformer bus F duriBg the Spring 2020 Ueit 2 refueling outage, TeehRieal SpecificatioR 3.16.B.2 is modified by a footRote permittiRg the use of a temporary, oRe time, 14 day allowea 00tage time (,i\\.OT). The 14 Elay 1\\.0T *.vill permit URit 1 to coRtiRue to operate for 14 days. While RSST C is liRBYailable to facilitste replacement of the RSST C 5KV cables, traRsfer hus F will be powered from the elepenattble alternste souree (i.e, bsekfeed throttgh the Unit 2 MaiB Step ttp Transformer/

StatioR Service Trllftsformer 2C). The baekfeeel power supply will allow tnmsfer bus F to perform. its Hormal ftmetieH while the RSST C 5KV cables are beiHg replaeed. Prier te entry into the 14 day AOT, the followiRg aetioRs shall be takeR:

-h WithiR 30 days prior to eRtering the temporary 14 dey AOT, functioRality of the A:ltemate AC (AAC) System (i.e., the sttpplemeRtal powef source) shall be verified.

~ DuriHg the 14 aay A:OT, the fuRctioRality of the tu'*.. C System shall be checked eRee per shift. If the A.. AC System beeomes ROB fuRetioaal at any tittle auriag the 14 day

,AcOT, it shall he restored te funetioaal statas withia 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or the uRit shall be brought to HOT SHUTDOWN withiR the Re'l!:t 6 houfs aRd COLD SHUTDOWN withiR the followiRg 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Amendment Nos.297 aRel 297

TS 3.16-7a 03 16 20 The '<'efifieation of functionality of the /'.A.C System pFiof to eetefiHg the tempontry 14 day AOT 1Nill be based OH the pFe'lious satisfactory quarterly test. The oeee pef shift fueetionality check *Nill be pcffonned durieg shiftly opemtof roueds.

In adelition to *rerifying and checking fueetiomtlity of the AA.C System priof to and durieg the temponu-y 14 day,A.QT, the following aetioas *.viii also be takea:

..J--

J,VeathCf eoaditioas will be moaitored tmel pFephmaed maiatcaaace *.viii aot be sehedulcel if se'ICfC weathCf cOHditioas are anticipated.

a The system load dispatcher,,,,.m be contacted oace pCF day to easure ao significant grid pCFturbations (high griel loadiag unable to v1ithstand a siftgle eoetiageney of Hae or geaCFatioa outage) ftfe eKpected deriag the temporary 14 day AOT.

Compoaeet testieg or maiateaaaee of safety systems aed importaet aon safety equipment iH the offsite powCf systems that ean iacrease the ltloolihood of a plant transient (ueit trip) or LOOP will be a*1oieled. In addition, HO cliseretionary switehyftfd mainteeaeee will be pefformed.

TS required systems, sebsystems, trains, eomponet1ts, aed de*rices that depeed on ff=le FCmaieing powCf sources will be verified to be operable aHd positi*;e measures will be prm1ided to preelade sebsequeet testing or mainteHaA:ce aeti¥ities OR these s~*stems, subsystems, trains, compoaeA:ts, and deYices.

OpCFatioa Of maintenaRce of plaet eqeipme1tt whea its reduedant equipment or train is out of ser¥iee 'Nill be eontrolled in accordaRee wiff=l procedure OP SU 601, "Protected Equipmeat." The Uait 1 steam dri"t1en Auxiliary Feedwater Pttmp will be eoatroHed as "Proteeteel Equipment" cluriag the temporary 14 day AOT.

The stattts of the l'n\\C diesel generator, EDGs, RSST A aad RSST B will be monitored oaee pCF shift.

Amendment Nos.297 and 297

2. Air Handling Units (AHUs)

TS 3.23-2 01 23 08

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

1. If either any single Unit 1 AHU or two Unit I AHUs on the same chilled water loop (l-VS-AC-1 and l-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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

2. If two Unit 1 AHUs on different chilled water loops and in different air conditioning zones (l-VS-AC-1 and l-VS-AC-6 or 1-VS-AC-2 and l-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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3. If two Unit 1 AHUs in the same air conditioning zone (l-VS-AC-1 and 1-VS-AC-2 or l-VS-AC-6 and l-VS-AC-7) become inoperable, restore operability of at least one Unit 1 AHU in each air conditioning zone (l-VS-AC-1 or 1-VS-AC-2 and 1-VS-AC-6 or l-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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4. If more than two Unit 1 AHUs become inoperable, restore operability of at least one Unit 1 AHU in each air conditioning zone (l-VS-AC-1 or l-VS-AC-2 and 1-VS-AC-6 or l-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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

The req1:1iremeAts of TS 3.23.C.2.a.l !IAd TS 3.23.C.2.b.l may be temporarily suspeAded aeeoftiiAg to the limirotions Roted in items 1, 2, 3 and 4 below for the purpose of repl11eemeRt of the MaiA CoAtrol Room (MCR) aad Emergeaey Switchgear Room (ESGR) Air Coaditioaing System (ACS) chilled water pipi11g. The !tllowed out11ge time e1Ete11sio11s speeifietl ia items 1, 2, 3 !lfld 4 of this footnote sh11U aot eiEeeed 24 months hegiaAiAg with entry iAto the fiFSt temporary eiEtension. Each eiEtensioA shall be limited to a oae time use *which eAds when the affeetetl MGR and ESGR ACS eompoaeats are retumed to OPERABLE status. Concurrent ese of more than oee allowed outage time 0,{teftsion is not pennitted.

I) The time period to aeeofflfftodate replaeement of the ehillea water loop C fliping iA the MCR aed the BSGR shall aot eiEeeed 45 days.

2) The time periotl to accommodate replacement of the ehilletl water loofl A pifliRg i11 the MCR aed the ESGR shall 11ot eiEeeed 45 days.

3) The time period to neeommedate replaeement of the chilletl water piping iA the Meehaaieal Bt;jttiflmeRt Reem. #3 (MER 3) assoeiated with chiller 1 VS E 4A shall not eiEeectl 14 tlays.

4) The time fleriotl to neeofflm.odate replacement of the chilled 'Nater piping iA the MER 3 assoeiatetl with chiller 1 VS E 4C sh!tll eot ex.eecd 14 d11ys.

Amendment Nos. 258 tmd 257

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 TS 4.9-1 04 29 11 A. The concentration of oxygen in the waste gas holdup system shall be determined to be within the limits of Specification 3.11.A by continuously monitoring the waste gases in the waste gas holdup system with th~

n monitor required to be OPERABLE by Table 3.7-5(a) of Specification 3.7.E.

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 activity > 2200 µCi/gm dose equivalent Xe-133, the waste gas decay tanks shall be sampled once per day.

Amendment Nos. 273 and 272

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

TS 6.1-2 09 15 05

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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.

Superintendent

d. The operations manager shall hold r have previously held) a Senior Reactor Operator License for Surry er Station or a similar design Pressurized Water Reactor plant. The 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 of ANSI 3.1 ( 12/79 Draft) for comparable positions.

Exceptions to this requirement are specified in the QA Program.

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. 214 ttftd 243

6.7 Environmental Qualifications TS 6.7-1 10 24 80 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 IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental Qualification 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 oo later than December I, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related 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 current as equipment is replaced, further tested, or otherwise further qualified.

Serial No.22-035 Docket Nos. 50-280/281 PROPOSED SUBSEQUENT RENEWED OPERA TING LICENSES, TECHNICAL SPECIFICATIONS, AND BASIS PAGES (TYPED)

Virginia Electric and Power Company (Dominion Energy Virginia)

Surry Power Station Units 1 and 2 (3 ) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders R. Deleted by Amendment XXX.

s. Deleted by Amendment XXX.

T. Deleted by Amendment XXX.

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 1 -

Subsequent Renewed License No. DPR-32 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 NRG 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.

Enclosure:

FOR THE UNITED STATES NUCLEAR REGULATORY COMMISSION Signed by Veil, Andrea on 05/04/21 Andrea Veil, Director Office of Nuclear Reactor Regulation Appendix A - Technical Specifications for Surry Power Station, Units 1 and 2 Date of Issuance: May 4, 2021 Surry - Unit 1 Subsequent Renewed License No. DPR-32 (3) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders R. Deleted by Amendment XXX.

S. Deleted by Amendment XXX.

T. Deleted by Amendment XXX.

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 (AN0-2) 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 2 Subsequent Renewed License No. DPR-37 Amendment XXX 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 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 NRG 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 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 keff = 1.0.

6.

POWER OPERATION When the reactor is critical and the neutron flux power range 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, train, component, or device shall be operable or have operability when it is capable of performing its specified function(s). Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support 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 has been tested periodically in accordance with Section 4 and meets its performance requirements.

E.

PROTECTIVE INSTRUMENTATION LOGIC

1.

ANALOG CHANNEL An arrangement of components and modules as required to generate a single protective action digital signal when required by a unit condition. An analog channel loses its identity when single action signals are combined.

Amendment Nos. XXX and XXX

C.

RCS Operational LEAKAGE Applicability TS 3.1-13 11-05-09 The following specifications are applicable to RCS operational LEAKAGE whenever Tavg (average RCS temperature) exceeds 200°F (200 degrees Fahrenheit).

Specifications

1.

RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE,

b. 1 gpm unidentified LEAKAGE,

c. 10 gpm identified LEAKAGE, and

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

2.a. If RCS operational LEAKAGE is not within the limits of 3.1.C.l for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE, reduce LEAKAGE to within the specified limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b. If the LEAKAGE is not reduced to within the specified limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the unit shall be brought to HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.

If RCS pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within the limit specified in 3.1.C.l.d, the unit shall be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Amendment Nos. 267 and 266

TS 3.l-14a 11-05-09 This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

APPLICABLE SAFETY ANALYSES - Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA; the amount of leakage can affect the probability 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 requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons 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 steam 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. Releases continue through the unaffected steam generators until the Residual Heat Removal System is placed in service.

The MSLB is less limiting for site radiation releases than the SGTR. The safety analysis for the MSLB accident assumes 1 gpm total primary to secondary 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 CPR 50.36(c)(2)(ii).

LIMITING CONDffiONS 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 could 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. 267 and 266

b. Unidentified LEAKAGE TS 3.1-14b 11-05-09 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 leak.off (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 Tavg exceeds 200°F, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In COLD SHUTDOWN and REFUELING SHUTDOWN, 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 individual 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. 267 and 266

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 requirements of Specification 3.23.A. With two AHUs inoperable on the shutdown unit, ensure that one AHU is OPERABLE in each unit's 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. XXX and XXX

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 continuously monitored 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. XXX and XXX

TS 3.10-6

7.

Eight air handling units (AHUs) shall be OPERABLE in accordance with the operability requirements of Specification 3.23.A. With two AHUs inoperable on either unit, ensure that one AHU is OPERABLE in each unit's 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 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 Building shall cease and irradiated fuel shall be placed in a safe position, work shall be initiated 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 interlocks and safety features, provide assurance that an accident, which would result in a hazard to public health and safety, will not occur during unit REFUELING OPERATIONS or irradiated fuel movement in the Fuel Building. When no change is being made in core geometry, one neutron detector is sufficient to monitor the core and permits 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 pressurization 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 components into and out of the containment. During REFUELING OPERATIONS, the equipment hatch must be capable of being closed.

Amendment Nos. XXX and XXX

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 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provided the other battery and battery chargers remain OPERABLE with one battery charger carrying the DC load of the failed battery's supply system. If the battery is not returned to OPERABLE status within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, the reactor shall be placed in HOT SHUTDOWN. If the battery is not restored to OPERABLE status within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , 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 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , and

b.

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

Amendment Nos. XXX and XXX

TS 3.16-7 TS action statement 3.16.B.l.a.2 provides an allowance to avoid unnecessary testing of an OPERABLE EDG(s). If it can be determined that the cause 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 the applicable required action(s) would be entered. Once the failure is repaired, the common cause failure 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 confirmed not to exist on the remaining EDG(s),

performance of the operability test within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months provides 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 requirement for the OPERABLE EDG(s), the corrective action program will continue to evaluate the common cause possibility, including the other unit's EDG or the shared EDG. This continued evaluation, however, is no longer under the 24-hour constraint imposed by the action statement.

According to Generic Letter 84-15 (Ref. 6), 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is reasonable 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. XXX and XXX

TS 3.23-2

2. Air Handling Units (AHUs)

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

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

2. If two Unit 1 AHUs on different chilled water loops and in different air conditioning zones (l-VS-AC-1 and l-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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3. If two Unit 1 AHUs in the same air conditioning zone (l-VS-AC-1 and 1-VS-AC-2 or 1-VS-AC-6 and l-VS-AC-7) become inoperable, restore operability of at least one Unit 1 AHU in each air conditioning zone (l-VS-AC-1 or l-VS-AC-2 and 1-VS-AC-6 or l-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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

4. If more than two Unit 1 AHUs become inoperable, restore operability of at least one Unit 1 AHU in each air conditioning zone (l-VS-AC-1 or 1-VS-AC-2 and 1-VS-AC-6 or l-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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Amendment Nos. XXX and XXX

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 TS 4.9-1 A The concentration of oxygen in the waste gas holdup system shall be determined to be within the limits of Specification 3.1 LA by continuously monitoring the waste gases in the waste gas holdup system with the oxygen monitor required to be OPERABLE by Table 3.7-S(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 activity> 2200 µCi/gm dose equivalent Xe-133, the waste gas decay tanks shall be sampled once per day.

Amendment Nos. XXX and XXX

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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 of ANSI 3.1 (12/79 Draft) for comparable positions.

Exceptions to this requirement are specified in the QA Program.

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. XXX and XXX

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 IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental Qualification 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-related 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 current as equipment is replaced, further tested, or otherwise further qualified.

Amendment Nos. XXX and XXX

v t e Letter Sequence Request
EPID:L-2022-LLA-0089, TSTF-448 (Approved, Closed)
Initiation Request, Request Acceptance Supplement Administration Questions Answers 27 October 2022 Results Approval, Approval Other:
MONTHYEAR2022-07-25 [Table View]

ML22171A013

Text

1.0 INTRODUCTION

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

6.0 REFERENCES

Subject:

Subject:

Subject:

Subject:

Subject:

Subject:

Enclosure:

Enclosure:

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