ML24219A237

Issuance of Amendment Nos. 319 and 319, Adoption of TSTF-577, Rev. 1, Revised Frequencies for Steam Generator Tube Inspection,

ML24219A237
Person / Time
Site:	Surry
Issue date:	08/23/2024
From:	John Klos Plant Licensing Branch II
To:	Carr E Dominion Innsbrook Technical Center
References
EPID L-2024-LLA-0016
Download: ML24219A237 (22)

Text

August 23, 2024 Eric S. Carr President - Nuclear Operations and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENT NOS. 319 AND 319, REGARDING ADOPTION OF TSTF-577, REV. 1, REVISED FREQUENCIES FOR STEAM GENERATOR TUBE INSPECTIONS, (EPID L-2024-LLA-0016)

Dear Eric Carr:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 319 to Subsequent Renewed Facility Operating License No. DPR-32 and Amendment No. 319 to Subsequent Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively. The amendments revise the technical specifications in response to your application dated February 19, 2024, as supplemented by letter dated May 9, 2024.

These amendments revise the Steam Generator (SG) Program and the Steam Generator Tube Inspection Report Technical Specifications (TS) based on TS Task Force Traveler (TSTF) TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (Agencywide Documents Access and Management System Accession No. ML21060B434), and the associated NRC staff safety evaluation (SE) of that TSTF (ML21098A188).

A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's monthly Federal Register notice.

Sincerely,

/RA/

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

Enclosures:

1. Amendment No. 319 to DPR-32

2. Amendment No. 319 to DPR-37

3. Safety Evaluation cc: Listserv VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 319 Subsequent Renewed License No. DPR-32

1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated February 19, 2024, as supplemented by letter dated May 9, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

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

(B)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 319, are hereby incorporated in the subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

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

Attachment:

Changes to License No. DPR-32 and the Technical Specifications Date of Issuance: August 23, 2024 SHAWN WILLIAMS Digitally signed by SHAWN WILLIAMS Date: 2024.08.23 14:57:36 -04'00' VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 319 Subsequent Renewed License No. DPR-37

1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated February 19, 2024, as supplemented by letter dated May 9, 2024, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

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

(B)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 319, are hereby incorporated in the subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 60 days.

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

Attachment:

Changes License No. DPR-37 and the Technical Specifications Date of Issuance: August 23, 2024 SHAWN WILLIAMS Digitally signed by SHAWN WILLIAMS Date: 2024.08.23 14:58:10 -04'00'

ATTACHMENT SURRY POWER STATION, UNIT NOS. 1 AND 2 TO LICENSE AMENDMENT NO. 319 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND TO LICENSE AMENDMENT NO. 319 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

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

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

License No. DPR-37, page 3 TSs TSs 6.4-11 6.4-11 6.4-12 6.4-12 6.4-13 6.4-13 6.4-13a na 6.6-3 6.6-3 6.6-3a 6.6-3a Surry - Unit 1 Subsequent Renewed License No. DPR-32 Amendment No. 319

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

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

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 319 are hereby incorporated in the subsequent renewed license.

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

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

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

E. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227 Surry - Unit 2 Subsequent Renewed License No. DPR-37 Amendment No. 319 3.

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

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

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 319 are hereby incorporated in this subsequent renewed license.

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

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

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

E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment 65 G. Deleted by Amendment 227 H. Deleted by Amendment 227

Q. Steam Generator (SG) Program An SG Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following:

1. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

2. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

a.

Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down),

all anticipated transients included in the design specification, and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

b. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Leakage is not to exceed 1 gpm for all SG.

TS 6.4-11 Amendment Nos. 319 and 319

c.

The operational LEAKAGE performance criterion is specified in TS 3.1.C and 4.13, RCS Operational LEAKAGE.

3. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube plugging criteria may be applied as an alternative to the 40% depth-based criteria:

a.

Tubes with service-induced flaws located greater than 17.89 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 17.89 inches below the top of the tubesheet shall be plugged upon detection.

4. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet except for any portions of the tube that are exempt from inspection by alternate repair criteria, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of 4.a, 4.b, and 4.c below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

a.

Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.

TS 6.4-12 Amendment Nos. 319 and 319

TS 6.4-13 Amendment Nos. 319 and 319 b.

After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 54 effective full power months, which defines the inspection period. If none of the SG tubes have ever experienced cracking other than in regions that are exempt from inspection by alternate repair criteria and the SG inspection was performed with enhanced probes, the inspection period may be extended to 72 effective full power months. Enhanced probes have a capability to detect flaws of any type equivalent to or better than array probe technology. The enhanced probes shall be used from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet except any portions of the tube that are exempt from inspection by alternate repair criteria. If there are regions where enhanced probes cannot be used, the tube inspection techniques shall be capable of detecting all forms of existing and potential degradation in that region.*

c.

If crack indications are found in the portions of the SG tube excluding any region that is exempt from inspection by alternate repair criteria, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall be at the next refueling outage but may be deferred to the following refueling outage if the 100% inspection of all SGs was performed with enhanced probes as described in paragraph 6.4.Q.4.b. If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

5. Provisions for monitoring operational primary to secondary LEAKAGE.

As approved by Amendment Nos. 299 and 299, the inspection of Surry Unit 2 SG B may be deferred, on a one-time basis, from the Surry Unit 2 spring 2020 refueling outage (S2R29) to the Surry Unit 2 2021 fall refueling outage (S2R30).

TS 6.6-3

b. Deleted

3. Steam Generator Tube Inspection Report A report shall be submitted within 180 days after Tavg exceeds 200°F following completion of an inspection performed in accordance with the Specification 6.4.Q, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG;

b. The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;

c. For each degradation mechanism found:

1. The nondestructive examination techniques utilized;

2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;

3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and

4. The number of tubes plugged during the inspection outage.

d. An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;

e. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;

f. The results of any SG secondary side inspections; Amendment Nos. 319 and 319

g. The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report;

h. The calculated accident induced LEAKAGE rate from the portion of the tubes below 17.89 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced LEAKAGE rate from the most limiting accident is less than 1.80 times the maximum operational primary to secondary LEAKAGE rate, the report should describe how it was determined; and

i. The results of the monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

TS 6.6-3a Amendment Nos. 319 and 319

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 319 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 319 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281

1.0 INTRODUCTION

By letter dated February 19, 2024, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML24051A178), as supplemented by letter dated May 9, 2024 (ML24130A203), Virginia Electric and Power Company (the licensee) submitted a license amendment request (LAR) for changes to the Surry Power Station, Unit Nos. 1 and 2 (Surry 1&2), technical specifications (TS).

The supplement dated May 9, 2024, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on April 16, 2024 (89 FR 26944).

The proposed changes would revise the Steam Generator (SG) Program and the Steam Generator Tube Inspection Report TS based on TS Task Force Traveler (TSTF) TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (ML21060B434), and the associated NRC staff safety evaluation (SE) of TSTF-577 (ML21098A188).

The tubes within an SG function as an integral part of the reactor coolant pressure boundary and, in addition, isolate fission products in the primary coolant from the secondary coolant and the environment. SG tube integrity means that the tubes are capable of performing this safety function in accordance with the plant design and licensing basis. Surry 1&2 SGs have Alloy 600 thermally treated (Alloy 600TT) tubes.

1.1 Proposed TS Changes to Adopt TSTF-577 In accordance with NRC staff-approved TSTF-577, the licensee proposed changes that would revise Surry TS 6.4.Q, Steam Generator (SG) Program, and TS 6.6.A.3, Steam Generator Tube Inspection Report. Specifically, the licensee proposed the following changes to adopt TSTF-577:

TS 6.4.Q, Steam Generator (SG) Program:

TS 6.4.Q introductory paragraph and paragraph 2.a would be revised by replacing steam generator with SG in a few instances.

TS 6.4.Q.4 would be revised by adding a phrase regarding portions of the tube that are exempt from inspection by alternate repair criteria.

TS 6.4.Q.4.b would be revised by deleting the requirement to base inspection frequency on the more restrictive metric between either the effective full power months (EFPM) or refueling outage and to use just the EFPM metric.

TS 6.4.Q.4.b would be revised by changing the requirement to inspect 100 percent of the tubes at periods of 120, 96, and 72 EFPM to 54 EFPM. A 72 EFPM inspection period would be permitted if SG tubing has never experienced cracking (not including regions exempt from inspection by alternate repair criteria) and the SG inspection was performed with enhanced probes. A description of the enhanced probe inspection would be added.

TS 6.4.Q.4.b would be revised by deleting the allowance to extend the inspection period by 3 months and by deleting the discussion of prorating inspections.

TS 6.4.Q.4.c would be revised by replacing shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections) with shall be at the next refueling outage.

TS 6.4.Q.4.c would be revised by adding a phrase regarding portions of the tube that are exempt from inspection by alternate repair criteria. An additional phrase would be added that permits deferring SG inspections after cracking indications are found if the 100 percent inspection was performed with enhanced probes.

TS 6.6.A.3, Steam Generator Tube Inspection Report:

Existing reporting requirement b. would be renumbered as c. and be revised by editorial and punctuation changes.

New reporting requirement b. would be added to require the nondestructive examination (NDE) techniques utilized for tubes with increased degradation susceptibility be reported.

Existing reporting requirement c. would be renumbered as c.1. and be revised by editorial and punctuation changes.

Existing reporting requirement d. would be renumbered as c.2. and be revised to note that the location, orientation (if linear), measured size (if available), and voltage response do not need to be reported for tube wear indications at support structures that are less than 20 percent through-wall. However, the total number of tube wear indications at support structures that are less than 20 percent through-wall would be reported.

New reporting requirement d. would be added to require an analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection relative to the applicable performance criteria, including the analysis methodology, inputs, and results.

Existing reporting requirement e. would be renumbered as c.4. and be revised by editorial and punctuation changes.

Existing reporting requirement f. would be renumbered as e. and be revised by editorial and punctuation changes.

New reporting requirement f. would be added to require the results of any SG secondary side inspections be reported.

Existing reporting requirement g. would be renumbered as c.3. and be revised to add the requirements to report a description of the condition monitoring assessment, the margin to the tube integrity performance criteria, and a comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment. In addition, the requirement to report the results of tube pulls and in-situ testing would be deleted.

Existing reporting requirements h., i., and j. would be renumbered to g., h., and i., and be revised by editorial and punctuation changes.

1.2 Additional Proposed TS Changes In addition to the changes proposed consistent with the traveler discussed in Section 1.1, the licensee proposed the following variations.

1.2.1 Editorial Variation The licensee identified one editorial variation.

1) Surry TSs have different numbering than standard technical specifications (STSs) on which TSTF-577 was based. Specifically, the Steam Generator (SG) Program is numbered 6.4.Q in Surry TSs rather than 5.5.9 as stated in the TSTF. In addition, the Steam Generator Tube Inspection Report is numbered 6.6.A.3 in Surry TSs rather than 5.6.7 as stated in the TSTF.

1.2.2 Other Variations The licensee identified three additional variations.

1) Surry TSs contain a requirement that differs from the STSs on which TSTF-577 was based. Specifically, Surry TS 6.4.Q Steam Generator (SG) Program, paragraph 2.b, states, Leakage is not to exceed 1 gpm for all SG rather than, Leakage is not to exceed [1 gpm] per SG (emphasis added).

2) Surry TSs currently contain a provision for alternate tube plugging criteria. The licensee noted that the description of the alternate tube plugging criteria in the proposed change is equivalent to the description in the current TSs.

3) Surry TS 6.4.Q.3 currently states, in part, The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria (emphasis added).

However, TSTF-577 states: The following alternate tube plugging [or repair] criteria may be applied as an alternative to the 40% depth based criteria (emphasis added).

The word shall will be changed to "may" in Surry TS 6.4.Q.3, which aligns with the wording in TSTF-577.

2.0 REGULATORY EVALUATION

The General Design Criteria (GDC) included in Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, General Design Criteria for Nuclear Power Plants, became effective on May 21, 1971. The construction permits for Surry were issued prior to May 21, 1971; consequently, Surry is not subject to the current GDC requirements. Surrys Updated Final Safety Analysis Report (UFSAR), Revision 55 (ML23275A077) Section 1.4 provides, in part, that Surry was designed to meet the intent of the GDC.

Regulations The regulations at 10 CFR 50.4, Technical Specification, Section 5.0, Administrative Controls, requires that an SG Program be established and implemented to ensure that SG tube integrity is maintained. Programs established by the licensee, including the SG Program, are listed in the administrative controls section of the TS to operate the facility in a safe manner.

The regulations in 10 CFR 50.36(c)(5), Administrative controls, state that [a]dministrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

Guidance NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425) which provides prepared STSs for each of the LWR nuclear designs.

NUREG-1431, Revision 5.0, Standard Technical Specifications Westinghouse Plants, Volume 1, Specifications, September 2021 (ML21259A155) and Volume 2, Bases,

September 2021 (ML21259A159). The NRC staff uses NUREG-1431 to serve as a reference for ensuring the licensee appropriately incorporates the affected set of parameters into this LAR.

The format of the standard technical specifications (STS) addresses the categories required by 10 CFR 50.36, Technical specifications, and consists of six sections covering the areas of:

definitions, safety limits and limiting safety system settings, LCOs, SRs, design features, and administrative controls. NRC recognizes that Surry has custom TS and the request for plant-specific applicability.

3.0 TECHNICAL EVALUATION

3.1 Proposed TS changes to Adopt TSTF-577 The NRC staff compared the licensees proposed TS changes in Section 1.1 of this SE against the changes approved in TSTF-577. In accordance with SRP Chapter 16.0, the NRC staff determined that the STS changes approved in TSTF-577 are applicable because Surry is a pressurized water reactor (PWR) design plant, and the NRC staff approved the TSTF-577 changes for PWR designs. The NRC staff finds that the licensees proposed changes to the Surry TSs in Section 1.1 of this SE are consistent with those found acceptable in TSTF-577.

In the SE of TSTF-577, the NRC staff concluded that the TSTF-577 changes to STS 5.5.9, Steam Generator (SG) Program, and STS 5.6.7, Steam Generator Tube Inspection Report, were acceptable because, as discussed in Section 3.0 of that SE, they continued to ensure SG tube integrity and, therefore, protected the public health and safety. In particular, the structural integrity performance criterion and accident-induced leakage performance criterion (explained in STS 5.5.9.b, items 1 and 2, respectively) will continue to be met with the proposed revised SG inspection intervals (maximum allowable time between SG inspections) and inspection periods (maximum allowable time between 100 percent of SG tubes inspections). Additionally, the proposed changes to the reporting requirements will provide more detailed and consistent information to the NRC. Therefore, the NRC staff found that the proposed changes to the SG program and inspection reporting requirements were acceptable because they continued to meet the requirements of 10 CFR 50.36(c)(5) by providing administrative controls necessary to assure operation of the facility in a safe manner. For these same reasons, the NRC staff concludes that the corresponding proposed changes to the Surry TSs in Section 1.1 of this SE continue to meet the requirements of 10 CFR 50.36(c)(5).

3.2 Additional Proposed TS Changes 3.2.1 Editorial Variation There is one editorial variation described in Section 1.2.1 of this SE. The NRC staff finds that this variation (i.e., different TS numbering) is acceptable because the variation does not substantively alter TS requirements.

3.2.2 Other Variations There are three variations identified in Section 1.2.2 of this SE.

For the first variation, the licensee explained that the Surry SG Program TSs currently contain a requirement that differs from the STS on which TSTF-577 was based. Specifically, the last sentence of Surry TS 6.4.Q.2.b states, Leakage is not to exceed 1 gpm for all SGs rather than

Leakage is not to exceed 1 gpm per SG (emphasis added). The NRC staff notes that the Surry TS 6.4.Q.2.b leakage rate requirements are more restrictive than the STS. As part of the request to adopt TSTF-577, the licensee did not propose any changes to these leakage rate requirements. Therefore, the NRC staff considers the variation acceptable.

For the second variation, the licensee explained that the Surry SG Program TSs currently contain a provision for an alternate tube plugging criteria. The current Surry TS that addresses alternate tube plugging criteria (i.e., TS 6.4.Q.3) reflects NRC approved changes contained in Amendment Nos. 277 and 278 to Renewed Facility Operating License No. DPR-32 and Amendment Nos. 277 and 278 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Units Nos.1 and 2, respectively (ML12109A270 and ML13018A086). As part of the request to adopt TSTF-577, the licensee did not propose any changes to these criteria.

Therefore, the NRC staff considers the variation acceptable.

For the third variation, the licensee noted that the Surry SG Program TSs currently contain a requirement that differs from the STS on which TSTF-577 was based. Specifically, Surry TS 6.4.Q.3 currently states, in part: The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria (emphasis added). However, TSTF-577 states: The following alternate tube plugging [or repair] criteria may be applied as an alternative to the 40% depth based criteria (emphasis added). In the application, the licensee explained that alternate tube plugging criteria are an alternative to the 40% tube plugging criterion, therefore, the word may is the proper word to use. Changing, the word shall to may in Surry Unit Nos. 1 and 2 TS 6.4.Q.3 also aligns with the STS wording contained in TSTF-577. The NRC staff finds the change from shall to may acceptable based on the explanation provided by the licensee and because it is consistent with STS SG Program guidance.

3.3 TS Change Consistency The NRC staff reviewed the proposed TS changes for technical clarity and consistency with the existing requirements for customary terminology and formatting. The NRC staff finds that the proposed changes are consistent with Chapter 16.0 of the SRP and are therefore acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Commonwealth of Virginias State official was notified of the proposed issuance of the amendments on August 1, 2024. On August 1, 2024, the Commonwealth of Virginias official confirmed that the Commonwealth of Virginia had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration as published in the Federal Register on April 16, 2024 (89 FR 26944), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no

environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

6.0 CONCLUSION

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

Principal Contributors: Clint Ashely, NRR Date: August 23, 2024

ML24219A237 OFFICE DORL/LPL2-1/PM DORL/LPL2-1/LA DSS/STSB/BC NAME JKlos KGoldstein SMehta DATE 8/1/2024 8/22/2024 6/12/2024 OFFICE OGC DORL/LPL2-1/BC DORL/LPL2-1/PM NAME IMurphy MMarkley (SWilliams for)

JKlos DATE 8/13/2024 8/23/2024 8/23/2024