ML23191A059

R. E. Ginna Nuclear Power Plant - Issuance of Amendment No. 155 Adoption of TSTF-577, Revised Frequencies for Steam Generator Tube Inspections, Revision 1

ML23191A059
Person / Time
Site:	Ginna
Issue date:	07/21/2023
From:	Joel Wiebe NRC/NRR/DORL/LPL3
To:	Rhoades D Constellation Energy Generation, Constellation Nuclear
References
EPID L-2022-LLA-0115
Download: ML23191A059 (1)

Text

July 21, 2023 Mr. David P. Rhoades Senior Vice President Constellation Energy Generation, LLC President and Chief Nuclear Officer Constellation Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

R. E. GINNA NUCLEAR POWER PLANT ISSUANCE OF AMENDMENT NO. 155 RE: ADOPTION OF TSTF-577, REVISED FREQUENCIES FOR STEAM GENERATOR TUBE INSPECTIONS, REVISION 1 (EPID L-2022-LLA-0115)

Dear Mr. Rhoades:

The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 155 to Renewed Facility Operating License No. DPR-18 for the R. E. Ginna Nuclear Power Plant. This amendment is in response to your application dated August 10, 2022, as supplemented by letters dated December 7, 2022, and May 22, 2023. The amendments for other plants included in your letter dated August 10, 2022, will be issued under separate cover.

The amendment revises the Steam Generator (SG) Program and the Steam Generator (SG)

Tube Inspection Report technical specifications based on Technical Specifications Task Force (TSTF) Traveler TSTF 577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (TSTF 577), and the associated NRC staff safety evaluation of TSTF 577.

D. Rhoades A copy of our related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Joel S. Wiebe, Senior Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-244

Enclosures:

1. Amendment No. 155 to DPR-18

2. Safety Evaluation

3. Notices and Environmental Findings cc: Listserv

CONSTELLATION ENERGY GENERATION, LLC DOCKET NO. 50-244 R. E. GINNA NUCLEAR POWER PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No.155 Renewed License No. DPR-18

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment filed by Constellation Energy Generation, LLC (Constellation, the licensee) dated August 10, 2022, as supplemented by letters dated December 7, 2022, and May 22, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-18 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 155, are hereby incorporated in the renewed license. Constellation Energy Generation, LLC shall operate the facility in accordance with the Technical Specifications.

Enclosure 1

3. This license amendment is effective as of the date of its issuance and shall be implemented within 120 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Jeffrey Jeffrey A. A. Whited Date: 2023.07.21 Whited 09:45:56 -04'00' Jeffrey A. Whited, Chief Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: July 21, 2023

ATTACHMENT TO LICENSE AMENDMENT NO. 155 R. E. GINNA NUCLEAR POWER PLANT RENEWED FACILITY OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Replace the following page of the Facility Operating License with the attached revised page.

The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert Page 4 Page 4 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert Page 3.4.17-1 Page 3.4.17-1 Page 3.4.17-2 Page 3.4.17-2 Page 5.5-5 Page 5.5-5 Page 5.5-6 Page 5.5-6 Page 5.6-5 Page 5.6-5 Page 5.6-6

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 155 are hereby incorporated in the renewed license.

Constellation Energy Generation, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Fire Protection Constellation Energy Generation, LLC shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensees amendment request dated March 28, 2013, supplemented by letters dated December 17, 2013; January 29, 2014; February 28, 2014; September 5, 2014; September 24, 2014; December 4, 2014; March 18, 2015; June 11, 2015; August 7, 2015; and as approved in the safety evaluation report dated November 23, 2015. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.

(a) Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the peer-reviewed fire PRA model, methods that have been approved by NRC through a plant-specific license amendment or NRC approval of generic methods specifically for use in NFPA 805 risk assessments, or methods that have been demonstrated to bound the risk impact.

1. Prior NRC review and approval is not required for changes that clearly result in a decrease in risk. The proposed change must also be consistent with the defense in-depth philosophy and must maintain sufficient safety margins. The change may be implemented following completion of the plant change evaluation.

R. E. Ginna Nuclear Power Plant Amendment No. 155

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

- NOTE -

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube affected tube(s) is plugging criteria and not maintained until the next plugged in accordance refueling outage or SG tube with the Steam Generator inspection.

Program.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following Generator Program. the next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

R.E. Ginna Nuclear Power Plant 3.4.17-1 Amendment 155

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the Steam In accordance with Generator Program. the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube plugging criteria is plugged in accordance with MODE 4 following a the Steam Generator Program. SG tube inspection R.E. Ginna Nuclear Power Plant 3.4.17-2 Amendment 155

Programs and Manuals 5.5 5.5.8 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:

a. 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.

b. 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.

1. 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.

2. 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 each SG. Leakage is not to exceed 1 gpm per SG.

R.E. Ginna Nuclear Power Plant 5.5-5 Amendment 155

Programs and Manuals 5.5

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. 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.

d. 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, 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 d.1, d.2, and d.3 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.

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

2. After the first refueling outage following SG installation, inspect 100% of the tubes in each SG at least every 96 effective full power months, which defines the inspection period.

3. If crack indications are found in any SG tube, 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. 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.

e. Provisions for monitoring operational primary to secondary LEAKAGE R.E. Ginna Nuclear Power Plant 5.5-6 Amendment 155

Reporting Requirements 5.6

2. As an alternative to the use of WCAP-14040-A Section 3.2 methodology, the existing Ginna specific LTOP Setpoint Methodology submitted to the NRC in the letter to Guy S.

Vissing (NRC) from Robert C. Mecredy (RG&E), "Application for Amendment to Facility Operating License, Revision to Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) Administrative Controls Requirements,"

Attachment VI, Section 3.2, dated September 29, 1997 and approved in letter to Robert C. Mecredy (RG&E) from S.

Singh Bajwa (NRC), "R.E. Ginna - Acceptance for Referencing of Pressure Temperature Limits Report,"

Revision 2 (TAC No. M96529), dated November 28, 1997, may be utilized.

d. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for revisions or supplement thereto.

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.8, 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.

R.E. Ginna Nuclear Power Plant 5.6-5 Amendment 155

Reporting Requirements 5.6

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; and

f. The results of any SG secondary side inspections.

R.E. Ginna Nuclear Power Plant 5.6-6 Amendment 155

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 155 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 CONSTELLATION ENERGY GENERATION, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Application Safety Evaluation Date August 10, 2022, (ML22222A068) July 21, 2023 December 7, 2022, (ML22342B230) Principal Contributors to Safety May 22, 2023, (ML23143A136) Evaluation Clinton Ashley

1.0 PROPOSED CHANGE

S By letter dated August 10, 2022, as supplemented by letters dated December 7, 2022, and May 22, 2023, Constellation Energy Generation, LLC (the licensee) requested changes to the technical specifications (TSs) for R. E. Ginna Nuclear Power Plant (Ginna), by license amendment request (application). In its August 10, 2022, letter, the licensee requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) process the proposed amendment under the Consolidated Line-Item Improvement Process (CLIIP). The proposed changes would revise the Steam Generator (SG) Tube Integrity, the Steam Generator (SG)

Program, and the Steam Generator Tube Inspection Report TSs, based on Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (TSTF-577) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21060B434), and the associated NRC staff safety evaluation (SE) of TSTF-577 (ML21098A188). This request was removed from the CLIIP process because of the complexity caused by not completing the required 100 percent SG tube inspections before the date of the request.

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.

Ginna has one unit and the SGs have Alloy 690 thermally treated (Alloy 690TT) tubes.

Enclosure 2

1.1 Proposed TS Changes to Adopt TSTF-577 The licensee proposed changes that would revise Ginna TS 3.4.17, Steam Generator (SG)

Tube Integrity, TS 5.5.8, Steam Generator (SG) Program, and TS 5.6.7, Steam Generator Tube Inspection Report. Specifically, the licensee proposed the following changes to adopt TSTF-577:

TS 3.4.17, Steam Generator (SG) Tube Integrity:

TS 3.4.17 would be revised by replacing tube repair criteria with tube plugging criteria in several locations.

TS 5.5.8, Steam Generator (SG) Program:

The introductory paragraph to TS 5.5.8 would be revised by replacing steam generator with SG in a couple instances.

The last sentence in the introductory paragraph to TS 5.5.8 ended shall include the following provisions. It was changed to delete the word provisions.

TS 5.5.8.b.1 would be revised by replacing steam generator with SG in one instance.

TS 5.5.8.b.1 would be revised to correct a misplaced closing parenthesis and associated punctuation (commas). It now reads, in part, All in-service 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.

TS 5.5.8.c would be revised to rename the title from Provisions for SG tube repair criteria to Provisions for SG tube plugging criteria.

TS 5.5.8.d would be revised to replace the words tube repair criteria with the words tube plugging criteria and to replace the words assessment of degradation with the words degradation assessment.

TS 5.5.8.d.1 would be revised by replacing the words SG replacement with the words SG installation.

TS 5.5.8.d.2 would be revised by deleting the requirement to inspect 100 percent of the tubes at sequential periods (144, 108, 72, and thereafter 60 effective full power months (EFPM)) and by adding the requirement to inspect 100 percent of the tubes every 96 EFPM. The 72 EFPM limit or three refueling outages (whichever is less) between inspections would be replaced with the 96 EFPM inspection period.

TS 5.5.8.d.2 would be revised by deleting the requirement to inspect 50 percent of the tubes at the inspection nearest the midpoint of the period and the remaining 50 percent of the tubes nearest the end of the period.

TS 5.5.8.d.3 would be revised by replacing the words for each SG with the words for each affected and potentially affected SG.

TS 5.5.8.d.3 would be revised by replacing the words shall not exceed 24 effective full power months or one refueling outage (whichever is less) with the words shall be at the next refueling outage.

TS 5.6.7 Steam Generator (SG) 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 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.

1.2 Additional TS Changes - Variations The licensee identified several variations:

The Ginna TSs have different numbering than Standard Technical Specifications (STS) on which TSTF-577 was based. Specifically, Steam Generator (SG) Tube Integrity is numbered as TS 3.4.17 in Ginna TSs rather than TS 3.4.20 as identified in TSTF-577.

Similarly, the Steam Generator (SG) Program, is numbered as TS 5.5.8 in Ginna TSs rather than TS 5.5.9 as identified in TSTF-577. The licensee concluded that these

numbering differences are administrative and do not affect the applicability of TSTF-577 to the Ginna TSs.

The Ginna TS 5.5.8.b.1 is revised to correct the placement of hyphens (e.g.,

primary-to-secondary). The licensee concluded that these corrections are administrative and do not affect the applicability of TSTF-577 to the Ginna TSs.

The Ginna TS 5.5.8.b.2 contains a requirement that differs from the STS on which TSTF-577 was based, but was NRC-approved by Amendment 100 dated March 1, 2007 (ML070370623). Specifically, Ginna TS 5.5.8.b.2 provides a plant-specific accident induced leakage criterion. The NRC-approved plant-specific criterion differs from the STS and is being retained. The licensee concluded that retaining the approved plant-specific leakage criterion does not affect the applicability of TSTF-577 to the Ginna TSs.

Ginna TS 5.5.8.d.2 currently states: , with the exception that each SG is to be inspected during the fourth refueling outage in G1R44 following inspections completed in refueling outage G1R40. The licensee inspected 100 percent of the SG tubes during the Spring 2023 refueling outage (G1R44). In its May 22, 2023, letter, the licensee concluded that the TS exception wording may be removed because the requirement has been completed and the exception is no longer needed.

The NRC staff identified one additional variation:

The NRC staff noted that TS 5.5.8.b is revised by replacing steam generator with SG in one instance.

2.0 REGULATORY EVALUATION

The regulations in Title 10 of the Code of Federal Regulations (10 CFR) paragraph 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. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in

[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 NRC staffs guidance for the review of TSs is in 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). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared Standard Technical Specifications (STSs) for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREG-14311, as modified by NRC-approved travelers.

1 U.S. Nuclear Regulatory Commission, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 5, September 2021 (ADAMS Accession Nos. ML21259A155 and ML21259A159, respectively).

TSTF-577 revised the STSs related to SG tube inspections and SG tube inspection reporting requirements. The NRC-approved TSTF-577, under the CLIIP on April 14, 2021, (ML21099A086).

3.0 TECHNICAL EVALUATION

3.1 Proposed TS Changes to Adopt TSTF-577 Section 2.2 of the TSTF-577 Traveler SE identified in part that at the time TSTF-577 was approved, the STS SG related requirements were established by TSTF-449, Revision 4, Steam Generator Tube Integrity (ML051090200). TSTF-449, Revision 4, was approved on May 2, 2005 (ML051160106), was adopted by all operating pressurized-water reactor (PWR) plants, and was incorporated into the STS, Revision 4. Subsequent to the publication of the Revision 4 STS, the NRC staff approved STS changes to the SG program, reporting, and tube integrity specifications in TSTF-510, Revision 2, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection (ML110610350), on October 19, 2011 (ML112101604). TSTF-510, Revision 2, has been adopted by most of the applicable plants.

Section 2.2 of the TSTF-577 SE identified, in part that although multiple versions of marked-up STS were provided in the traveler (TSTF-577), the NRC staff focused on reviewing the version in the enclosure titled, Changes to the Technical Specifications Based on TSTF-510 (also referred to as the Markup Based on TSTF-510). TSTF-577 included mark-ups based on the TSTF-510 changes but the TSTF-510 changes were not reevaluated in the SE of TSTF-577 since the NRC staff had already approved TSTF-510, Revision 2.

Ginna TSs are based on TSTF-449. Therefore, as part of this SE, the NRC staff also considered the version of marked-up STS provided in the traveler (TSTF-577) enclosure titled, Changes to the Technical Specifications Based on TSTF-449 (also referred to as the Markup Based on TSTF-449). The TSTF-577 Markup Based on TSTF-449 includes mark-ups that incorporate applicable changes attributable to NRC-approved TSTF-510. Therefore, the TSTF-510 related changes associated with TSTF-577 markups based on TSTF-449, were not reevaluated in this SE since the NRC staff had already approved TSTF-510, Revision 2.

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 Ginna is a 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 Ginna TSs in Section 1.1 of this SE are consistent with NRC-approved 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 Ginna TS described 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 Variations The variations are described in Section 1.2 of this SE. The NRC staff finds the variations are acceptable because the variations do not substantively alter TS requirements and do not affect the applicability of TSTF-577 to the Ginna TSs.

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 CONCLUSION

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

NOTICES AND ENVIRONMENTAL FINDINGS RELATED TO AMENDMENT NO. 155 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-18 CONSTELLATION ENERGY GENERATION, LLC R. E. GINNA NUCLEAR POWER PLANT DOCKET NO. 50-244 Application Safety Evaluation Date August 10, 2022, (ML22222A068) July 21, 2023 December 7, 2022, (ML22342B230) Principal Contributors to Safety May 22, 2023, (ML23143A136) Evaluation Clinton Ashley

1.0 INTRODUCTION

By letter dated August 10, 2022, as supplemented by letters dated December 7, 2022, and May 22, 2023, Constellation Energy Generation, LLC (the licensee) requested changes to the technical specifications (TSs) for R. E. Ginna Nuclear Power Plant (Ginna), by license amendment request (application). In its August 10, 2022, letter, the licensee requested that the U.S. Nuclear Regulatory Commission (NRC, the Commission) process the proposed amendment under the Consolidated Line-Item Improvement Process (CLIIP). The proposed changes would revise the Steam Generator (SG) Tube Integrity, the Steam Generator (SG)

Program, and the Steam Generator Tube Inspection Report TSs based on Technical Specifications Task Force (TSTF) Traveler TSTF-577, Revision 1, Revised Frequencies for Steam Generator Tube Inspections (TSTF-577) (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21060B434), and the associated NRC staff safety evaluation (SE) of TSTF-577 (ML21098A188). This request was removed from the CLIIP process because of the complexity caused by not completing the required 100 percent SG tube inspections before the date of the request.

The supplements dated December 7, 2022, and May 22, 2023, 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 October 4, 2022 (87 FR 60216).

2.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on June 6, 2023. The State official had no comments.

Enclosure 3

3.0 ENVIRONMENTAL CONSIDERATION

The amendment relates, in part, to changes in recordkeeping, reporting, or administrative procedures or requirements. The amendment also relates, in part, to changing requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves 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 amendment involves no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on October 4, 2022 (87 FR 60216).

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

Date of issuance: July 21, 2023

ML23191A059 NRR-058 OFFICE NRR/DORL/LPL3/PM NRR/DORL/LPL3/LA NRR/DSS/STSB/BC NAME JWiebe SRohrer VCusumano DATE 7/7/2023 7/11/23 6/6/2023 OFFICE NRR/DORL/LPL3/BC NRR/DORL/LPL3/PM NAME JWhited JWiebe DATE 7/21/2023 7/21/2023