ML24103A204

Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition

ML24103A204
Person / Time
Site:	Calvert Cliffs, Peach Bottom, Nine Mile Point, Byron, Braidwood, Limerick, Ginna, Clinton, FitzPatrick, LaSalle
Issue date:	04/12/2024
From:	David Helker Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
Download: ML24103A204 (1)

Text

200 Exelon Way Kennett Square, PA 19348 www.constellation.com

10 CFR 50.90

April 12, 2024

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Braidwood Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50- 457

Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50- 455

Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318

Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

James A. FitzPatrick Nuclear Power Plant Renewed Facility Operating License No. DPR-59 NRC Docket Nos. 50-333

LaSalle County Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374

Limerick Generating Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License No. DPR-63 and NPF-69 NRC Docket No. 50-220 and 50- 410

Peach Bottom Atomic Power Station, Units 2 and 3 Subsequent Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

U.S. Nuclear Regulatory Commission Application to Revise TS to Adopt TSTF-591 April 12, 2024 Page 2

R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

Subject:

Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and revise 10 CFR 50.69 License Condition

Pursuant to 10 CFR 50.90, Constellation Energy Generation, LLC (CEG) is submitting a request for an amendment to the Technical Specifications (TS) for Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Clinton Power Station, Unit 1, James A. Fitzpatrick Nuclear Power Plant, LaSalle County Station, Units 1 and 2, Limerick Generating Station, Units 1 and 2, Nine Mile Point Nuclear Station, Units 1 and 2, Peach Bottom Atomic Power Station, Units 2 and 3, and R.E. Ginna Nuclear Power Plant.

CEG requests adoption of TSTF-591-A, "Revise Risk Informed Completion Time (RICT) Program" Revision 0, which is an approved change to the Standard Technical Specifications (STS), into the above licensees Technical Specifications (TS).

TSTF-591-A revises the TS Section 5.5 Programs and Manuals, "Risk Informed Completion Time Program," to reference Regulatory Guide (RG) 1.200, Revision 3, instead of Revision 2, and to make other changes. A new report is added to TS Section 5.6, "Reporting Requirements," to inform the NRC of newly developed methods used to calculate a RICT. Additionally, CEG is requesting the removal of certain stations Facility Operating License condition sections associated with the implementation of the 10 CFR 50.69 and Risk-Informed Completion Time programs.

provides a description and assessment of the proposed changes. provides the existing TS pages marked up to show the proposed changes.

The proposed changes do not affect the TS Bases.

CEG requests that the amendments be reviewed under the Consolidated Line Item Improvement Process (CLIIP). Approval of the proposed amendments is requested within 6 months of completion of the NRCs acceptance review. Once approved, the amendments shall be implemented within 90 days.

CEG has concluded that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92, "Issuance of amendment."

The proposed changes have been reviewed by the stations Plant Operations Review Committee in accordance with the requirements of the CEG Quality Assurance Program.

There are no regulatory commitments contained in this application. Furthermore, this application contains no changes to each licensees Probabilistic Risk Analysis (PRA) that would be considered a Newly Developed Method, in accordance with the Pressurized Water Reactor Owners Group (PWROG), PWROG-19027-NP, Revision 2, Newly Developed U.S. Nuclear Regulatory Commission Application to Revise TS to Adopt TSTF-591 April 12, 2024 Page 3

Method Requirements and Peer Review, issued July 2020, as endorsed in Regulatory Guide 1.200 Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3, dated December 30, 2022.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated Illinois, Pennsylvania, Maryland, and New York officials.

If you have any questions or require additional information, please contact Steve Flickinger, Licensing and Regulatory Affairs, at 267-533-5302.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 12th day of April 2024.

Respectfully,

David P. Helker Sr. Manager, Licensing Constellation Energy Generation, LLC

Attachments: 1 - Evaluation of Proposed Changes 2 - Facility Operating License and Technical Specification Mark-Up Pages U.S. Nuclear Regulatory Commission Application to Revise TS to Adopt TSTF-591 April 12, 2024 Page 4

cc: (w/ Attachments)

Regional Administrator - NRC Region I Regional Administrator - NRC Region III NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station NRC Senior Resident Inspector - Calvert Cliffs Nuclear Power Plant NRC Senior Resident Inspector - Clinton Power Station NRC Senior Resident Inspector - James A. FitzPatrick Nuclear Power Plant NRC Senior Resident Inspector - LaSalle County Station NRC Senior Resident Inspector - Limerick Generating Station NRC Senior Resident Inspector - Nine Mile Point Nuclear Station NRC Senior Resident Inspector - Peach Bottom Atomic Power Station NRC Senior Resident Inspector - R. E. Ginna Nuclear Power Plant NRC Project Manager, NRR - Braidwood Station NRC Project Manager, NRR - Byron Station NRC Project Manager, NRR - Calvert Cliffs Nuclear Power Plant NRC Project Manager, NRR - Clinton Power Station NRC Project Manager, NRR - James A. FitzPatrick Nuclear Power Plant NRC Project Manager, NRR - LaSalle County Station NRC Project Manager, NRR - Limerick Generating Station NRC Project Manager, NRR - Nine Mile Point Nuclear Station NRC Project Manager, NRR - Peach Bottom Atomic Power Station NRC Project Manager, NRR - R. E. Ginna Nuclear Power Plant Illinois Emergency Management Agency - Division of Nuclear Safety Director, Bureau of Radiation Protection, PA Department of Environmental Protection S. Seaman, State of Maryland A. L. Peterson, NYSERDA

ATTACHMENT 1

Evaluation of Proposed Changes

License Amendment Request

Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Clinton Power Station, Unit 1, James A. Fitzpatrick Nuclear Power Plant, LaSalle County Station, Units 1 and 2, Limerick Generating Station, Units 1 and 2, Nine Mile Point Nuclear Station, Units 1 and 2, Peach Bottom Atomic Power Station, Units 2 and 3, and R.E. Ginna Nuclear Power Plant.

Docket Nos.

STN 50-456 and STN 50-457, STN 50- 454 and STN 50-455, 50- 317 and 50 -318, 50- 461, 50-333, 50-373 and 50- 374, 50- 352 and 50-353, 50 -220 and 50-410, 50- 277 and 50- 278, and 50-244

Subject:

Application to Revise Technical Specifications to Adopt TSTF-591-A,

Revise Risk Informed Completion Time (RICT) Program Revision 0 and revise 10 CFR 50.69 License Condition.

1.0 DESCRIPTION

2.0 ASSESSMENT

2.1 Applicability of Safety Evaluation

2.2 Variations

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis

3.2 Conclusion

5.0 REFERENCES

Constellation License Amendment Request Adoption of TSTF-591-A Rev 0 and modifications to FOL Pages

1.0 DESCRIPTION

Constellation Energy Generation, LLC (CEG) requests adoption of TSTF-591-A Revision 0, "Revise Risk Informed Completion Time (RICT) Program," (Reference 1) which is an approved change to the Standard Technical Specifications (STS), into the Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Clinton Power Station, Unit 1, James A. FitzPatrick Nuclear Power Plant, LaSalle County Station, Units 1 and 2, Limerick Generating Station, Units 1 and 2, Nine Mile Point Nuclear Station, Units 1 and 2, Peach Bottom Atomic Power Station, Units 2 and 3, and R.E. Ginna Nuclear Power Plant Technical Specifications (TS).

TSTF-591-A revises the TS Section 5.5 Programs and Manuals, "Risk Informed Completion Time Program," to reference Regulatory Guide (RG) 1.200, Revision 3 (Reference 2), instead of Revision 2, and to make other changes. A new report is added to TS Section 5.6, "Reporting Requirements," to inform the NRC of newly developed methods used to calculate a RICT.

2.0 ASSESSMENT

2.1 Applicability of Safety Evaluation

CEG has reviewed the safety evaluation for TSTF-591-A provided to the Technical Specifications Task Force in a letter dated September 21, 2023. This review included the NRC staffs evaluation, as well as the information provided in TSTF-591. CEG has concluded that the justifications presented in TSTF-591-A and the safety evaluation prepared by the NRC staff are applicable to Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Clinton Power Station, Unit 1, James A. FitzPatrick Nuclear Power Plant, LaSalle County Station, Units 1 and 2, Limerick Generating Station, Units 1 and 2, Nine Mile Point Nuclear Station, Units 1 and 2, Peach Bottom Atomic Power Station, Units 2 and 3, and R.E. Ginna Nuclear Power Plant TS and justify this amendment for incorporation of the changes to the aforementioned plants TS.

This application contains no changes to each licensees Probabilistic Risk Analysis (PRA) that would be considered a Newly Developed Method, in accordance with the Pressurized Water Reactor Owners Group (PWROG) topical report, PWROG-19027-NP, Revision 2, Newly Developed Method Requirements and Peer Review, issued July 2020 (Refence 3), as endorsed in Regulatory Guide 1.200 Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3.

2.2 Variations

CEG is proposing the following variations from the TS changes described in TSTF-591-A or the applicable parts of the NRC staffs safety evaluation.

CEG proposes changes to the Facility Operating License (FOL) pages for Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, LaSalle County Station, Units 1 and 2, Nine Mile Point Nuclear Station, Unit 2, Peach Bottom Atomic Power Station, Units 2 and 3, and R.E. Ginna Nuclear Power Plant. The change to the FOL pages removes the paragraph under the 10 CFR 50.69 Risk-informed categorizations and treatment of structures, systems and components for nuclear power plants that illustrates

Page l 1 Constellation License Amendment Request Adoption of TSTF-591-A Rev 0 and modifications to FOL Pages

programmatic implementation conditions of 10 CFR 50.69. The paragraph includes reference to RG 1.200, Revision 2. Adopting TSTF-591-A for the RICT program and not removing reference to RG 1.200 Revision 2 for the 50.69 program would create a conflict in guidance for the Probabilistic Risk Assessment (PRA) model which both programs use. Additionally, since each above station has implemented the 10 CFR 50.69 program, these implementation conditions are no longer relevant. Limerick Generating Station Units 1 and 2 have recently eliminated this paragraph in license amendments 261 and 223, respectively, in the NRC SE (ML23094A171), and Limerick Generating Station TS Appendix C SE (ML23089A124).

The LaSalle County Station, Units 1 and 2, Nine Mile Point Nuclear Station, Unit 2, R.E. Ginna Nuclear Power Plant, and Peach Bottom Atomic Power Station Units 2 and 3 licenses contain a condition that serves the same purpose as Paragraph e in TSTF-505, Revision 2, "Provide Risk -

Informed Extended Completion Times - RITSTF Initiative 4b." CEG proposes to remove the license condition since, because each above station has implemented the Risk-Informed Completion Time Program, these implementation conditions are no longer relevant.

Some CEG plants TS utilize different numbering and titles than the STS on which TSTF-591-A was based. These differences are administrative and do not affect the applicability of TSTF-591-A. The following table describes the different numbering from TSTF-591-A s new STS Section 5.6. These section numbers are reaffirmed in the inserts in Attachment 2.

NUREG TSTF-BRW/BYR/ CPS JAF LAS LIM NMP1 PBAPS/NMP2 591 CAL 1430, 1431, 5.6.8 5.6.10 1432 1433 5.6.6 5.6.8 5.6.7 6.9.3 5.6.9 1434 5.6.7 5.6.6 Custom TS N/A 6.6.8

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Analysis

CEG requests adoption of TSTF-591-A, Revise Risk Informed Completion Time (RICT)

Program, which is an approved change to the Standard Technical Specifications (STS),

into the Braidwood Station, Units 1 and 2, Byron Station, Units 1 and 2, Calvert Cliffs Nuclear Power Plant, Units 1 and 2, Clinton Power Station, Unit 1, James A. FitzPatrick Nuclear Power Plant, LaSalle County Station, Units 1 and 2, Limerick Generating Station, Units 1 and 2, Nine Mile Point Nuclear Station, Units 1 and 2, Peach Bottom Atomic Power Station, Units 2 and 3, and R.E. Ginna Nuclear Power Plant Technical Specifications (TS). TSTF-591-A revises the TS Section 5.5 Programs and Manuals, "Risk Informed Completion Time Program," to reference Regulatory Guide (RG) 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

Revision 3, instead of Revision 2, and to make other changes. A new report is added to TS Section 5.6, "Reporting Requirements," to inform the NRC of newly developed methods used to calculate a RICT. Additionally, CEG is requesting the removal of

Page l 2 Constellation License Amendment Request Adoption of TSTF-591-A Rev 0 and modifications to FOL Pages

certain stations Facility Operating License condition sections associated with the implementation of the 10 CFR 50.69 and Risk-Informed Completion Time programs.

CEG has evaluated whether a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No

The proposed change updates the standard for maintaining and updating PRA models used to calculate a RICT from NRC-approved Regulatory Guide 1.200, Revision 2, to NRC-approved Regulatory Guide 1.200, Revision 3. A new report is added to inform the NRC when a newly developed method is used.

The proposed change does not involve a significant increase in the probability of an accident previously evaluated because the change involves no change to the plant or its modes of operation. The proposed change does not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended Risk Informed Completion Time are no different from those during the existing Completion Time. The submittal of information-only reports has no effect on the initiators or consequences of any accidents previously evaluated. The requested change to remove certain stations Facility Operating License condition sections associated with the implementation of the 10 CFR 50.69 and Risk-Informed Completion Time programs has no effect on the initiators or consequences of any accidents previously evaluated.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No The proposed change updates the standard for maintaining and updating PRA models used to calculate a RICT from NRC-approved Regulatory Guide 1.200, Revision 2, to NRC-approved Regulatory Guide 1.200, Revision 3. A new report is added to inform the NRC when a newly developed method is used.

Additionally, the proposed change removes certain stations Facility Operating License condition sections associated with the implementation of the 10 CFR 50.69 and Risk-Informed Completion Time programs.

The proposed change does not change a design function or method of operation of the plant. The proposed change does not involve a physical alteration of the plant (no new or different kind of equipment will be installed).

Page l 3 Constellation License Amendment Request Adoption of TSTF-591-A Rev 0 and modifications to FOL Pages

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

Response: No

The proposed change updates the standard for maintaining and updating PRA models used to calculate a RICT from NRC-approved Regulatory Guide 1.200, Revision 2, to NRC-approved Regulatory Guide 1.200, Revision 3. A new report is added to inform the NRC when a newly developed method is used.

Additionally, the proposed change removes certain stations Facility Operating License condition sections associated with the implementation of the 10 CFR 50.69 and Risk-Informed Completion Time programs.

The proposed change supports the extension of Completion Times provided risk is assessed and managed in accordance with the NRC-approved RICT Program. The proposed change does not alter any design basis or safety limits.

The proposed change affects the standard used to maintain the PRA models used in the RICT Program by changing from one NRC-approved standard to a later NRC-approved version and requiring submittal of an information-only report. The RICT Program will continue to assure that adequate margins of safety are maintained. The removal of certain stations Facility Operating License conditions does not alter any design basis or safety limits.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, CEG concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c),

and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 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 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.

4.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20 or would change an inspection or surveillance requirement. However, the proposed amendments do not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed

Page l 4 Constellation License Amendment Request Adoption of TSTF-591-A Rev 0 and modifications to FOL Pages

amendments meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendments.

5.0 REFERENCES

1. Technical Specification Task Force Traveler 591-A Revise Risk-Informed Completion Time (RICT) Program, Revision 0, March 22, 2022 (ML22081A224).

2. Regulatory Guide 1.200 Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3, dated December 30, 2022 (ML20238B871).

3. Pressurized Water Reactor Owners Group (PWROG), PWROG-19027-NP, Revision 2, Newly Developed Method Requirements and Peer Review, issued July 2020 (ML20213C660).

Page l 5 ATTACHMENT 2

Facility Operating License and Technical Specification Mark-Ups

Facility Operating License and Technical Specification Mark-U ps

Braidwood Station, Units 1 and 2

Docket Nos.

STN 50-45 6 and STN 50-45 7

(c) The flux thimble tube corrective actions, inspections, and replacements identified in the SER, Commitment No. 24, for Braidwood Units 1 and 2, shall be implemented in accordance with the schedule in the Commitment. Periodic eddy current testing/inspections of all flux thimble tubes shall be performed at least every two refueling outages, and the data shall be trende d and retained in auditable form. A flux thimble tube shall not remain in service for more than two (2) operating fuel cycles without successful completion of eddy current testing for that thimble tube.

(14) Adoption of 10 CFR 50.69, Risk-informed categorization an d treatment of structures, systems, and components for nuclear power plants

Constellation Energy Generation, LLC 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) models to evaluate risk associated with internal events, including internal flooding, a nd internal fire; 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, Class 3, and non-Code class SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e.,

seismic margin analysis (SMA) to evaluate seismic risk, 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 the license amendment No. 198, dated October 22, 2018.

The licensee will complete the updated implementation items listed in Attachment 1 of Exelon letter to NRC dated September 13, 2018, prior to implementation of 10 CFR 50.69. A ll issues identified in the attachment will be addressed and any associated changes will be made, focused scope peer reviews wil l be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a chang e to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Renewed License No. NPF-72 Amendment No. 224 (c) The flux thimble tube corrective actions, inspections, and replacements identified in the SER, Commitment No. 24, for Braidwood Units 1 and 2, shall be implemented in accordance with the schedule in the Commitment. Periodic eddy current testing/inspections of all flux thimble tubes shall be performed at least every two refueling outages {RFO], and the data shall be trended and retained in auditable form. A flux thimble tube sha ll not remain in service for more than two (2) operating fuel cycl es without successful completion of eddy current testing for that thimble tube.

(d) Deleted

(13) Adoption of 10 CFR 50.69, Risk-informed categorization an d treatment of structures, systems, and components for nuclear power plants

Constellation Energy Generation, LLC is approved to implement 1 0 CFR 50.69 using the processes for categorization of Risk-Informed S afety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; 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 Clas s 2, Class 3, and non-Code class SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e.,

seismic margin analysis (SMA) to evaluate seismic risk, 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 the license amendment No. 198, dated October 22, 2018.

The licensee will complete the updated implementation items listed in Attachment 1 of Exelon letter to NRC dated September 13, 2018, prior to implementation of 10 CFR 50.69. Al l issues identified in the attachment will be addressed and any associated changes will be made, focused scope peer reviews wil l be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Renewed License No. NPF-77 Amendment No. 229

Programs and Manuals 5.5 5.5 Programs and Manuals

5.5.19 Surveillance Frequency Control Program

This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The Provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

5.5.20 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODE 1 and 2;

c. When a RICT is being used, any change to the plant configuration change, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1.For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2.For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3.Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

BRAIDWOOD UNITS 1 & 2 5.5 24 Amendment 206 Programs and Manuals 5.5 5.5 Programs and Manuals

5.5.20 Risk Informed Completion Time Program (continued)

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1.Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2.Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

BRAIDWOOD UNITS 1 & 2 5.5 25 Amendment 206 Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.9 Steam Generator (SG) Tube Inspection Report (continued)

h. For Unit 2, the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report,

i. For Unit 2, the calculated accident induced leakage rate from the portion of the tubes below 14.01 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 3.11 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined, and

j. For Unit 2, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

BRAIDWOOD UNITS 1 & 2 5.6 7 Amendment 172 Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA m odel, and seismic hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, R evision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A report shall be submitted in accordance with Specification 5.6.10 before a newly developed method is us ed to calculate a RICT.

Insert 2:

5.6.10 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Facility Operating License and Technical Specification Mark-U ps

Byron Station, Units 1 and 2

Docket Nos.

STN 50-45 4 and STN 50-455

(23) License Renewal License Conditions

(a) The information in the UFSAR supplement, submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, and as supplemented by the Commitments applicable to Byron Unit 1 in Appendix A of the Safety Evaluation Report Related to the License Renewal of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (SER) dated July 2015, is collectively the 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, the licensee may make changes to the programs and activities applicable to Byron Unit 1 describe d in this Supplement provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwis e complies with the requirements in that section.

(b) This License Renewal UFSAR Supplement, as revised per License Condition 23(a) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation.

1. The licensee shall implement those new programs and enhancements to existing programs no later than April 30, 2024.

2. The licensee shall complete those activities as noted in th e Commitments applicable to Byron Unit 1 in this Supplement no later than April 30, 2024 or the end of the last refueling outage prior to the period of extended operation, whichever occurs later.

3. The licensee shall notify the NRC in writing within 30 days after having accomplished item (b)1 above and include the status of those activities that have been or remain to be completed in item (b)2 above.

(24) Adoption of 10 CFR 50.69, Risk-informed categorization an d treatment of structures, systems, and components for nuclear power plants

Constellation Energy Generation, LLC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed S afety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using:

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, a nd internal fire; 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

Renewed License No. NPF-37 Amendment No. 226

Class 2, Class 3, and non-Code class SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e.,

seismic margin analysis (SMA) to evaluate seismic risk, 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 the license amendment No. 204, dated October 22, 2018.

The licensee will complete the updated implementation items listed in Attachment 1 of Exelon letter to NRC dated September 13, 2018, prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

D. The facility requires no exemptions from the requirements of 10 CFR Part 50.

E. Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualifications, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 1, which contain Safeguards Information protected under 10 CFR 73.21, is entitled:

ion Plan, and Byron Nuclear Power Station Security Plan, Training and Qualificat Safeguards Contingency Plan, Revision 3, submitted by letter dated May 17, 2006.

Constellation Energy Generation, LLC shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The CSP was approved by License Amendment No. 175 and modified by License Amendment No. 191.

F. Deleted

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan

Renewed License No. NPF-37 Amendment No. 226

(c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders

(12) License Renewal License Conditions

(a) The information in the UFSAR supplement, submitted pursuant to 10 CFR 54.21(d), as revised during the license renewal application review process, and as supplemented by the Commitments applicable to Byron Unit 2 in Appendix A of the Safety Evaluation Report Related to the License Renewal of Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (SER) dated July 2015, is collectively the 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, the licensee may make changes to the programs and activities applicable to Byron Unit 2 describe d in this Supplement provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwis e complies with the requirements in that section.

(b) This License Renewal UFSAR Supplement, as revised per License Condition 12(a) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation.

1. The licensee shall implement those new programs and enhancements to existing programs no later than May 6, 2026.

2. The licensee shall complete those activities as noted in th e Commitments applicable to Byron Unit 2 in this Supplement no later than May 6, 2026, or the end of the last refueling outage prior to the period of extended operation, whichever occurs later.

3. The licensee shall notify the NRC in writing within 30 days after having accomplished item (b)1 above and include the status of those activities that have been or remain to be completed in item (b)2 above.

(13) Adoption of 10 CFR 50.69, Risk-informed categorization an d treatment of structures, systems, and components for nuclear power plants

Constellation Energy Generation, LLC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed S afety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using:

Renewed License No. NPF-66 Amendment No. 226

Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, a nd internal fire; 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, Class 3, and non-Code class SSCs and their associated supports; and the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e.,

seismic margin analysis (SMA) to evaluate seismic risk, 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 the license amendment No. 204, dated October 22, 2018.

The licensee will complete the updated implementation items listed in Attachment 1 of Exelon letter to NRC dated September 13, 2018, prior to implementation of 10 CFR 50.69. A ll issues identified in the attachment will be addressed and any associated changes will be made, focused scope peer reviews wil l be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a chang e to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

D. The facility requires no exemptions from the requirements of 10 CFR Part 50.

An exemption was previously granted pursuant to 10 CFR 70.24. The exemption was granted with NRC materials license No. SNM-1916, issued Mar ch 4, 1985, and relieved the licensee from the requirement of having a criticality alarm system. Therefore, the licensee is exempted from the criticali ty alarm system provision of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.

E. The licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the licensees Fire Protection Report and the licensees letters dated September 23, 1986, October 23, 19 86, November 3, 1986, December 12 and 15, 1986, and January 21, 1987, and as approved in the SER dated February 1982 through Supplement No. 8, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Renewed License No. NPF-66 Amendment No. 226 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.19 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program 5.5.20 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODE 1 and 2;

c. When a RICT is being used, any change to the plant configuration change, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1.For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

BYRON UNITS 1 & 2 5.5 23 Amendment 231 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.20 Risk Informed Completion Time Program (continued)

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

BYRON UNITS 1 & 2 5.5 24 Amendment 231 Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.9 Steam Generator (SG) Tube Inspection Report (continued)

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;

g. For Unit 2, the operational primary to secondary leakage rate observed (greater than three gallons per day) in each steam generator (if it is not practical to assign the leakage to an individual steam generator, the entire primary to secondary leakage should be conservatively assumed to be from one steam generator) during the cycle preceding the inspection which is the subject of the report; h. For Unit 2, the calculated accident induced leakage rate from the portion of the tubes below 14.01 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 3.11 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and

i. For Unit 2, the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

BYRON UNITS 1 & 2 5.6 7 Amendment 231 Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA model, seismic hazards using penalty f actors, and configuration-specific tornado missile hazards using penalty factors.

Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A report shall be submitted in accordance with Specification 5.6.10 before a newly developed method is us ed to calculate a RICT.

Insert 2:

5.6.10 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Technical Specification Mark-Ups

Calvert Cliffs Nuclear Power Plant, Units 1 and 2

Docket Nos.

50- 317 and 50-318 Programs and Manuals 5.5

5.5 Programs and Manuals

inleakage, and assessing the CRE boundary as required by paragraphs c and d respectively.

5.5.18 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09, Revision 0-A, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODE 1, and 2;

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09, Revision 0-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d. If the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

CALVERT CLIFFS - UNIT 1 5.5-19 Amendment No. 326 CALVERT CLIFFS - UNIT 2 Amendment No. 304 Programs and Manuals 5.5

5.5 Programs and Manuals

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the completion times must be PRA methods used to support Amendment Nos. 326/304, or other methods approved by the NRC for generic use. Any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

5.5.19 Surveillance Frequency Control Program

This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, Risk-Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies, Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

CALVERT CLIFFS - UNIT 1 5.5-20 Amendment No. 326 CALVERT CLIFFS - UNIT 2 Amendment No. 304 Reporting Requirements 5.6

5.6 Reporting Requirements

thermal mechanical limits, core thermal hydraulic limits, ECCS limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Not Used

5.6.7 Post-Accident Monitoring Report

When a report is required by Condition B or F of LCO 3.3.10, "Post Accident Monitoring Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.8 Tendon Surveillance Report

Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

5.6.9 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 5.5.9, Steam Generat or (SG)

Program. The re port shall include:

CALVERT CLIFFS - UNIT 1 5.6-6 Amendment No. 346 CALVERT CLIFFS - UNIT 2 Amendment No. 324 ReportingRequirements 

5.6

5.6Reporting Requirements 

a. Thescope ofinspections performed oneach SG;

b. Thenondestructive examination techniques utilized for tubeswith increased degradation susceptibility;

c. Foreach degradation mechanism found:

1. Thenondestructive examination techniques utilized;

2. Thelocation, orientation (if linear), measured size(if available), and voltage response for each indication.For tube wear atsupport structures lessthan 20percent through wall, only the total numberofindications needs tobereported;

3. A description ofthe condition monitoring assessmentand results, including the margin tothe tubeintegrity performance criteria and comparison withthe margin predicted toexist atthe inspectionbythe previous forward looking tube integrityassessment; and

4. Thenumber  oftubes plugged during the inspection outage.

d. Ananalysis summary ofthe tube integrity conditions predictedtoexist atthe next scheduled inspection (theforward looking tube integrity assessment) relativetothe applicable performance criteria, includingthe analysis methodology, inputs, and results;

e. Thenumber and percentage oftubes plugged todate, and theeffective plugging percentage ineach SG; and

f. The results ofany SGsecondary side inspections.

CALVERTCLIFFS  UNIT1 5.67Amendment  

CALVERTCLIFFS  UNIT2 AmendmentNo.

Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA model, seismic hazards us ing penalty f actors, and tornado missile hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A report shall be submitted in accordance with Specification 5.6.10 before a newly developed method is us ed to calculate a RICT.

Insert 2:

5.6.10 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Technical Specification Mark-Ups

Clinton Power Station, Unit 1

Docket No.

50- 461 Programs and Manuals 5.5

5.5 Programs and Manuals (continued)

5.5.17 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODE 1 and 2;

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

CLINTON 5.0-17 Amendment No. 238 Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR),

2. LCO 3.2.2, Minimum Critical Power Ratio (MCPR) and MCPR99.9%,

3. LCO 3.2.3, Linear Heat Generation Rate (LHGR),

4. LCO 3.3.1.1, RPS Instrumentation (SR 3.3.1.1.14),

5. LCO 3.3.1.3, Oscillation Power Range Monitor (OPRM)

Instrumentation, and

6. LCO 3.7.6, Main Turbine Bypass System, (cycle dependent thermal power limits for an inoperable Main Turbine Bypass System).

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in

(1) General Electric Standard Application for Reactor Fuel (GESTAR), NEDE-24011-P-A, or

(2) NEDO-32465, "BWR Owners' Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications."

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

CLINTON 5.0-19 Amendment No. 238 Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA m odel, and seismic hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, R evision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A report shall be submitted in accordance with Specification 5.6.6 before a newly developed method is used to calculate a RICT.

Insert 2:

5.6.6 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Facility Operating License and Technical Specification Mark-U ps

R.E. Ginna Nuclear Power Plant

Docket No.

50-24 4 (17) Adoption of Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -RITSTF Initiative 4b

Constellation Energy Generation, LLC is approved to implement TSTF-505, Revision 2, modifying the Technical Specification requirements related to Completion Times (CT) for Required Actions to provide the option to calculate a longer, risk-informed CT (RICT). The methodology for using the new Risk-Informed Completion Time Program is described in NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, which was approved by the NRC on May 17, 2007.

Constellation Energy Generation, LLC will complete the implementation items listed in Attachment 6 of Exelon Letter to the NRC dated May 20, 2021, prior to implementation of the RICT Program. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to the implementation of the RICT Program.

(18) Deleted

(19) Constellation Energy Generation, LLC shall provide to the Director of the Office of Nuclear Reactor Regulation or the Director of the Office of Nuclear Material Safety and Safeguards, as applicable, a copy of any application, at the time it is filed, to transfer (excluding grants of security interests or liens) from Constellation Energy Generation, LLC to its direct or indirect parent, or to any other affiliate company, facilities for the production, transmission, or distribution of electric energy having a depreciated book value exceeding ten percent (10%) of Constellation Energy Generation, LLCs consolidated net utility plant, as recorded on Constellation Energy Generation, LLCs books of account.

(20) Constellation Energy Generation, LLC 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) models to evaluate risk associated with internal events, including internal flooding, and internal fire; 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 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in Exelon's submittal letter dated May 20, 2021, and all its subsequent associated supplements as specified in License Amendment No. 151 dated June 22, 2022.

R. E. Ginna Nuclear Power Plant Amendment No. 151 Programs and Manuals 5.5

e. The quantitative limits on unfiltered air inleakage into the CRE.

These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability and determining CRE unfiltered inleakage as required by paragraph c.

5.5.17 Surveillance Frequency Control Program

This program provides controls for the Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Controlled Program shall be made in accordance with NEI 04-10, Risk-Informed Method for Control of Surveillance Frequency, Revision 1.

c. The provisions of Surveillance Requirement 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

5.5.18 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, Risk-Managed Technical Specifications (RMTS) Guidelines.

The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODES 1 and 2;

R.E. Ginna Nuclear Power Plant 5.5-13 Amendment 150

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required If the plant configuration change would lower plant risk and would result in a longer RICT.

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support License Amendment No. 150, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

R.E. Ginna Nuclear Power Plant 5.5-14 Amendment 150

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

Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA model, seismic hazards us ing penalty f actors, and configuration-specific tornado missile hazards using penalty factors.

Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A report shall be submitted in accordance with Specification 5.6.8 before a newly developed method is us ed to calculate a RICT.

Insert 2:

5.6.8 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Technical Specification Mark-Ups

James A. Fitzpatrick Nuclear Power Plant

Docket No.

50-333 Programs and Manuals 5.5

5.5 Programs and Manuals (continued)

5.5.16 Risk Informed Completion Time Program (continued)

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete priortoexceeding theCompletion Time, the RICT shallaccount forthe increased possibility of common cause failure (CCF) by either:

1. Numerically accounting forthe increased possibilityofCCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that performthe function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support License Amendment No., or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

JAFNPP 5.5-16 Amendment 353

Insert 1:

e. A RICT calculation must include the following hazard groups and means of assessing the hazard: internal flood and internal events using a PRA model, internal fires using a PRA model, and seismic hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities."

g. A report shall be submitted in accordance with Specification 5.6.8 before a newly developed method is used to calculate a RICT.

Insert 2:

5.6.8 Risk Informed Completion Time (RICT) Program Upgrade Report

A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to using those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Facility Operating License and Technical Specification Mark-U ps

LaSalle County Station, Units 1 and 2

Docket Nos.

50-3 73 and 50-3 74 Renewed License No. NPF-11

(47) Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants

Constellation Energy Generation, LLC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed S afety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) mo dels to evaluate risk associated with internal events, including int ernal flooding, and internal fire; 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; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in the EGC submittal letter dated January 31, 2020, and all its subsequent associated supplements, as specifi ed in License Amendment No. 249 dated May 27, 2021.

The licensee will complete the implementation items listed in Table APLA-01.2 in Attachment 1 of EGC letter to NRC dated October 29, 2020, prior to implementation of 10 CFR 50.69 progr am. All issues identified will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA Standard (ASME/ANS RA-Sa-200 9, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Amendment No. 254 Renewed License No. NPF-11

(48) Adoption of Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -RITSTF Initiative 4b

Constellation Energy Generation, LLC is approved to implement TSTF-505, Revision 2, modifying the Technical Specification requirements related to Completion Times (CT) for Required Acti ons to provide the option to calculate a longer, risk-informed CT (RIC T). The methodology for using the new Risk-Informed Completion Time Program is described in NEI 06-09-A, Risk-Informed Technical Specifica tions Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, which was approved by the NRC on May 17, 2007.

The licensee will complete the implementation item listed in At tachment 5 of Exelon letter to the NRC dated January 31, 2020, prior to implementation of the RICT Program. All issues identified in t he attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that ar e PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, a s endorsed by RG 1.200, Revision 2), and any findings will be res olved and reflected in the PRA of record prior to implementation of the R ICT Program.

Am. 102 03/16/95 D. The facility requires exemptions from certain requirements of 10 CFR Part 50, 10 CFR Part 70, and 10 CFR Part 73. These include:

(a) Exemptions from certain requirements of Appendices G, H and J and 10 CFR Part 73 are described in the Safety Evaluation Report an d Supplement No. 1, No. 2, No. 3 to the Safety Evaluation Report.

(b) DELETED

(c) DELETED

(d) DELETED

Am. 226 (e) DELETED 11/16/17

Amendment No. 254 Renewed License No. NPF-18

3. The licensee shall notify the NRC in writing within 30 days after having accomplished item (b)1 above and include the status of those activities that have been or remain to be completed in item (b)2 above.

(36) Adoption of 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants

Constellation Energy Generation, LLC is approved to implement 10 CFR 50.69 using the processes for categorization of Risk-Informed S afety Class (RISC)-1, RISC-2, RISC-3, and RISC-4 Structures, Systems, and Components (SSCs) using: Probabilistic Risk Assessment (PRA) m odels to evaluate risk associated with internal events, including int ernal flooding, and internal fire; the shutdown safety assessment pro cess 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; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach as described in the EGC submittal letter dated January 31, 2020, and all its subsequent associated supplements, as specifi ed in License Amendment No. 235 dated May 27, 2021.

The licensee will complete the implementation items listed in Table APLA-01.2 in Attachment 1 of EGC letter to NRC dated October 29, 2020, prior to implementation of 10 CFR 50.69 progr am. All issues identified will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA Standard (ASME/ANS RA-Sa-200 9, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Amendment No. 240 Renewed License No. NPF-18

(37) Adoption of Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -RITSTF Initiative 4b

Constellation Energy Generation, LLC is approved to implement TSTF-505, Revision 2, modifying the Technical Specification requirements related to Completion Times (CT) for Required Acti ons to provide the option to calculate a longer, risk-informed CT (RIC T). The methodology for using the new Risk-Informed Completion Time Program is described in NEI 06-09-A, Risk-Informed Technical Specifica tions Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, which was approved by the NRC on May 17, 2007.

The licensee will complete the implementation item listed in At tachment 5 of Exelon letter to the NRC dated January 31, 2020, prior to implementation of the RICT Program. All issues identified in t he attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that ar e PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, a s endorsed by RG 1.200, Revision 2), and any findings will be res olved and reflected in the PRA of record prior to implementation of the R ICT Program.

Amendment No. 240 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.16 Surveillance Frequency Control Program (continued)

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

5.5.17 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODES 1 and 2;

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned change, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d. For emergent conditions, if the extent of conditions evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

(continued)

LaSalle 1 and 2 5.5-16 Amendment No. 251/237 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.17 Risk Informed Completion Time Program (continued)

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

LaSalle 1 and 2 5.5-17 Amendment No. 251/237 Reporting Requirements 5.6

5.6 Reporting Requirements

5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report

When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

LaSalle 1 and 2 5.6-5 Amendment No. 177/163 Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA m odel, and seismic hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, R evision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A report shall be submitted in accordance with Specification 5.6.7 before a newly developed method is used to calculate a RICT.

Insert 2:

5.6.7 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Technical Specification Mark-Ups

Limerick Generating Station, Units 1 and 2

Docket Nos.

50- 352 and 50-353 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

c. The program shall, as allowed by 10 CFR 50.55a, meet Subsection ISTA, "General Requirements," and Subsection ISTD, "Preservice and lnservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants," in lieu of Section XI of the ASME B&PV Code ISI requirements for snubbers, or meet authorized alternatives pursuant to 10 CFR 50.55a.

d. The 120- month program updates shall be made in accordance with 10 CFR 50.55a subject to the limitations and conditions listed therein.

l. Explosive Gas Monitoring Program

This program provides controls for potentially explosive gas mixtures contained downstream of the off-gas recombiners.

The program shall include:

a. The limit for the concentration of hydrogen downstream of the offgas recombiners and a surveillance program to ensure the limit is maintained. This limit shall be appropriate to the systems design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the Explosive Gas Monitoring Program surveillance frequencies.

m. Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days.

b. A RICT may only be utilized in OPERATIONAL CONDITIONs 1 and 2.

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the ACTION allowed outage time (i.e.,

not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

LIMERICK - UNIT 1 6-14e Amendment No. 223, 228, 240 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the ACTION allowed outage time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the completion times must be PRA methods approved for use with this program in Amendment No. 240, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

LIMERICK - UNIT 1 6-14f Amendment No. 240 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b. MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c. The MINIMUM CRITICAL POWER RATIO (MCPR) and MCPR(99.9%) Specification 3.2.3,for

d. The MCPR(P) and MCPR(F) adjustment factors for specification 3.2.3,

e. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

f. The power biased Rod Block Monitor setpointsand the Rod Block Monitor MCPR OPERABILITY limits of of Specification 3.3.6 Specification 3.1.4.3,

g. The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,

h. The Oscillation Power Range Monitor (OPRM) period based detection algorithm (PBDA) setpoints for Specification 2.2.1,

i. The minimum required number of operable main turbine bypass valves for Specification 3.7.8 and the TURBINE BYPASS SYSTEM RESPONSE TIME for Specification 4.7.8.c.

6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a. NEDE-24011-P-A "General Electric Standard Application forFuel" (Latest approved revision),* Reactor

b. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, " August 1996.

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.13 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

a. Limiting Condition for Operation Section 3.4.6, "RCS

b. Surveillance Requirement Section 4.4.6, "RCS Pressure/TemperaturePressure/Temperature Limits" Limits" The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a. BWROG-TP-11-022-A, Revision 1 (SIR-05-044), "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated August 2013.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplements thereto.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

• For Cycle 8, specific documents were approved in the Safety Evaluation dated (5/4/98) to support License Amendment No. (127).

LIMERICK - UNIT 1 6-18a Amendment No. 37,66,77,127, 142, 177, 200, 236, 253 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

c. The program shall, as allowed by 10 CFR 50.55a, meet Subsection ISTA, "General Requirements," and Subsection ISTD, "Preservice and lnservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants," in lieu of Section XI of the ASME B&PV Code ISI requirements for snubbers, or meet authorized alternatives pursuant to 10 CFR 50.55a.

d. The 120- month program updates shall be made in accordance with 10 CFR 50.55a subject to the limitations and conditions listed therein.

l. Explosive Gas Monitoring Program

This program provides controls for potentially explosive gas mixtures contained downstream of the off-gas recombiners.

The program shall include:

a. The limit for the concentration of hydrogen downstream of the offgas recombiners and a surveillance program to ensure the limit is maintained. This limit shall be appropriate to the systems design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the Explosive Gas Monitoring Program surveillance frequencies.

m. Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days.

b. A RICT may only be utilized in OPERATIONAL CONDITIONs 1 and 2.

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the ACTION allowed outage time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

LIMERICK - UNIT 2 6-14e Amendment No. 184, 191, 203 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the ACTION allowed outage time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the completion times must be PRA methods approved for use with this program in Amendment No.203, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

LIMERICK - UNIT 2 6-14f Amendment No. 203

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b. MAPFAC(P) and MAPFAC(F) factors for Specification 3.2.1,

c. The MINIMUM CRITICAL POWER RATIO (MCPR) and MCPR(99.9%) for Specification 3.2.3,

d. The MCPR(P) and MCPR(F) adjustment factor for specification 3.2.3,

e. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

f. The power biased Rod Block Monitor setpoints3.3.6 and the Rod Block Monitor MCPR OPERABILITY limits of of Specification Specification 3.1.4.3.

g. The Reactor Coolant System Recirculation Flow upscale trip setpoint and allowable value for Specification 3.3.6,

h. The Oscillation Poalgorithm (PBDA) setpoints for Specification 2.2.1wer Range Monitor (OPRM) period based detection,

i. The minimum required number of operable main turbine bypass valves for Specification 3.7.8 and the TURBINE BYPASS SYSTEM RESPONSE TIME for Specification 4.7.8.c.

6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a. NEDE-24011-P-A "General Fuel" (Latest approved revision),Electric Standard Application for Reactor

b. NEDO-32465-A, "Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications, "

August 1996.

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 6.9.1.13 RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

a. Limiting Condition for Operation Section 3.4.6, "RCS

b. Surveillance Requirement Section 4.4.6, "RCS Pressure/TemperaturePressure/Temperature Limits" Limits" The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

a. BWROG-TP-11-022-A, Revision 1 (SIR-05-044), Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," dated August 2013.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplements thereto.

SPECIAL REPORTS

6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

LIMERICK - UNIT 2 6-18a Amendment No. 139, 161,199, 215 4, 38,48, 104, Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA m odel, and seismic hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, R evision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A report shall be submitted in accordance with Specification 6.9.3 before a newly developed method is us ed to calculate a RICT.

Insert 2:

6.9.3 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Facility Operating License and Technical Specification Mark-U ps

Nine Mile Point Nuclear Station, Units 1 and 2

Docket Nos.

50-2 20 and 50-4 10 6.5.9 Surveillance Frequency Control Program

This program provides controls for the Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Controlled Program shall be made in accordance with NEI 04-10, Risk-Informed Method for Control of Surveillance Frequency, Revision 1.

c. The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

6.5.10 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines."

The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in the Power Operating Condition;

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

AMENDMENT NO. 222, 250 355b

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support License Amendment No. 250, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

AMENDMENT NO. 250 355c

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, U.S. Supplement, (NRC approved version specified in the COLR).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as shutdown margin (SDM), transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

6.6.6 Special Reports

Special reports shall be submitted within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:

a. - f. (Deleted)

g. Sealed Source Leakage In Excess Of Limits, Specification 3.6.5.2 (Three months).

h. Accident Monitoring Instrumentation Report, Specification 3.6.11.a (Table 3.6.11-2, Action 3 or 4) (Within 14 days following the event).

6.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and inservice leakage and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in th e PTLR for the following:

1. Limiting Condition for Operation Section 3.2.1, Reactor Vessel Heatup and Cooldown Rates.

2. Limiting Condition for Operation Section 3.2.2, Minimum Reactor Vessel Temperature for Pressurization.

3. Surveillance Requirement Section 4.2.2, Minimum Reactor Vessel Temperature for Pressurization.

AMENDMENT NO. 142, 157, 162, 181, 184, 204 358

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

1. SIR-05-044-A, Pressure-Temperature Limits Report Methodology for Boiling Water Reactors, Revision 0, April 2007.

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

AMENDMENT NO. 204 358a

Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA model, seismic hazards using penalty factors, and configuration-specific straight wind and tornado wind pressure / tornado missile hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-In formed Activities."

g. A report shall be submitted in accordance with Specification 5.6.9 before a newly developed method is used to calculate a RICT.

Insert 2:

6.6.8 Risk Informed Completion Time (RICT) Program Upgrade Report

A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

(25) Within 14 days of the closing of the transaction approved on November 16, 2021, Constellation Energy Generation, LLC shall submit to the NRC the Nuclear Operating Services Agreement reflecting the terms set forth in the application dated February 25, 2021. Section 7.1 of the Nuclear Operating Services Agreement may not be modified in any material respect related to financial arrangements that would adversely impact the ability of the licensee to fund safety-related activities authorized by the license without the prior written consent of the Director of the Office of Nuclear Reactor Regulation.

(26) Deleted.

(27) Deleted

(28) Deleted.

(29) Adoption of Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extension Completion Times - RITSTF Initiative 4b"

Constellation Energy Generation, LLC is approved to implement TSTF-505, Revision 2, modifying the Technical Specification requirements related to Completion Times (CT) for Required Actions to provide the option to calculate a longer, risk-informed CT (RICT). The methodology for using the new Risk-Informed Completion Time Program is described in NEI 06-09-A, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines,"

Revision 0, which was approved by the NRC on May 17, 2007.

Constellation Energy Generation, LLCwill complete the implementation items listed in Attachment 6 of Exelon Letter to the NRC dated October 31, 2019, prior to implementation of the RICT Program. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to the implementation of the RICT Program.

Renewed License No. NPF-69 Amendment 140, 144, 183, 186, 189

(30) Constellation Energy Generation, LLC 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) models to evaluate risk associated with internal events, including internal flooding, and internal fire; 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 and non-Class SSCs and their associated supports; the results of the non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards updated using the external hazard screening significance process identified in ASME/ANS PRA Standard RA-Sa-2009 for other external hazards except seismic; and the alternative seismic approach described in Exelon's submittal letter dated December 26, 2019, and all its subsequent associated supplements as specified in License Amendment No. 183 dated January 29, 2021.

Constellation Energy Generation, LLC will complete the items listed in Attachment 7 of Exelon letter to NRC dated December 26, 2019, prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

D. The facility requires exemptions from certain requirements of 10 CFR Part 50 and 10 CFR Part 70.

i) An exemption from the critically alarm requirements of 10 CFR Part 70.24 was granted in the Special Nuclear Materials License No. SNM-1895 dated November 27, 1985. This exemption is described in Section 9.1 of Supplement 4 to the SER. This previously granted exemption is continued continued in this operating license.

ii) Exemptions to certain requirements of Appendix J to 10 CFR Part 50 are described in Supplements 3, 4, and 5 to the SER. These include (a) (this item left intentionally blank); (b) an exemption from the requirement of Option B of Appendix J, exempting main steam isolation valve measured leakage from the combined leakage rate limit of 0.6 La. (Section 6.2.6 of SSER 5)*; (c) an exemption from Option B of Appendix J, exempting the

• The parenthetical notation following the discussion of each exemption denotes the section of the Safety Evaluation Report (SER) and/or its supplements wherein the safety evaluation of the exemption is discussed.

Renewed License No. NPF-69 Amendment 140, 144, 183, 186. 189 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.15 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines."

The program shall include the following:

a. The RICT may not exceed 30 days;

b. A RICT may only be utilized in MODE 1, 2;

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods used to support License Amendment No. 186, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

NMP2 5.5-14 Amendment 186 Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report

When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and system leakage and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Limiting Condition for Operation 3.4.11, RCS Pressure and Temperature (P/T) Limits.

2. Surveillance Requirements 3.4.11.1 through 3.4.11.9

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents.

1. NEDC-33178P-A, Revision 1, General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, dated June 2009. The licensee will calculate the fluence for determining the adjusted reference temperature using either; (1) values determined using an NRC-approved, RG-1.190-adherent method, or (2) a fluence estimate, which the licensee has verified as conservative, using an NRC-approved, RG 1.190-adherent method.

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

5.6.8 OPRM Report

When a report is required by Required Action F.3 of TS 3.3.1.1, RPS Instrumentation, a report shall be submitted within the following 90 days.

The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to operable status.

NMP2 5.6-4 Amendment 91, 92, 145, 151 Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA m odel, and seismic hazards using penalty factors. Changes to these means of assessing the hazard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, R evision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A report shall be submitted in accordance with Specification 5.6.9 before a newly developed method is used to calculate a RICT.

Insert 2:

5.6.9 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.

Facility Operating License and Technical Specification Mark-U ps

Peach Bottom Atomic Power Station, Units 2 and 3

Docket Nos.

50-2 77 and 50-2 78 (16) Maximum Extended Load Line Limit Analysis Plus (MELLLA+) S pecial Consideration

The licensee shall not operate the facility within the MELLLA+ operating domain with a feedwater heater out of service resulting in more than a 10°F reduction in feedwater temperature below the design feedwater temperature.

(17) Adoption of 10 CFR 50.69, Risk-informed Categorization an d Treatment of Structures, Systems, and Components for Nuclear Power Plants

In support of implementing License Amendment No. 321 permittin g the adoption of the provisions of 10 CFR 50.69 for Renewed Facility Operating License No. DPR-44 for Peach Bottom Unit 2, the license is amen ded to add the following license condition:

(a) The licensee is approved to implement 10 CFR 50.69 using th e processes for categorization of Risk-Informed Safety Class (RIS C)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including intern al flooding, and internal fire; 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 the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, 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 Unit 2 License Amendment No. 321 dated October 25, 2018.

The licensee will complete the implementation items listed in Attachment 2 of Exelons letter to the NRC dated June 6, 2018, prior to implementation of 10 CFR 50.69. All issues identified in th e attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

Page 13 Subsequent Renewed License No. DPR-44 Amendment No. 340 the licensee may make changes to the programs, activities, and commitments described in the Subsequent License Renewal UFSAR Supplement, provided the licensee evaluates such changes pursua nt to the criteria set forth in 10 CFR 50.59, Changes, Tests, and Experiments, and otherwise complies with the requirements in that section.

(b) The Subsequent License Renewal UFSAR Supplement, as defined in subsequent renewed license condition (19)(a) above, describes programs to be implemented and activities to be completed prior to the subsequent period of extended operation, which is the perio d following the August 8, 2033, expiration of the initial renewed license.

1. Constellation Energy Generation, LLC shall implement those new programs and enhancements to existing programs no later than 6 months before the subsequent period of extended operation.

2. Constellation Energy Generation, LLC shall complete those activities by the 6-month date prior to the subsequent period o f extended operation or by the end of the last refueling outage before the subsequent period of extended operation, whichever occurs later.

3. Constellation Energy Generation, LLC shall notify the NRC in writing within 30 days after having accomplished item (b)1 above and include the status of those activities that have been or remain to be completed in item (b)2 above.

(20) PRA Model Updates to Support Implementation of the Risk In formed Completion Time (RICT) Program

Constellation Energy Generation, LLC is approved to implement TSTF-505, Revision 2, modifying the Technical Specification requirements related to Completion Times (CT) for Required Actions to provide the optio n to calculate a longer, risk-informed CT. The methodology for using the new Risk-Informed Completion Time (RICT) Program is described in NEI 06-09-A, Risk-Informed Technical Specifications Initiativ e 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, which was approved by the NRC on May 17, 2007.

Constellation Energy Generation, LLC will complete the implemen tation items listed in Attachment 6 of Exelon letter to the NRC dated May 29, 2020, prior to implementation of the RICT Program. All issues identified in the attachment will be addressed and any associated changes wil l be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-200 9, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the RICT Program.

Page 15 Subsequent Renewed License No. DPR-44 Amendment No. 340 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.15 Battery Monitoring and Maintenance Program (continued)

1. Actions to restore battery cells with float voltage

< 2.13 V;

2. Actions to determine whether the float voltage of the remaining battery cells is > 2.13 V when the float voltage of a battery cell has been found to be

< 2.13 V;

3. Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates;

4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal v oltage; and

5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

5.5.16 Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06- 09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days.

b. A RICT may only be utilized in MODEs 1 and 2.

c. When a RICT is being used, any change to the plant configuration, as defined in NEI 06- 09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

PBAPS UNIT 2 5.0- 18c Amendment No. 338 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.16. Risk Informed Completion Time Pr ogram (continued)

d. For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exc eeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (R MAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specif ied in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Co mpletion Times must be PRA methods approved for use with this program in Amendment No. 338, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

PBAPS UNIT 2 5.0- 18d Amendment No. 338 Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall out line the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant S ystem (RCS) PRESSURE AND TEMPERATURE LIMI TS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

i) Limiting Co nditions for Operation Section 3.4.9, RCS Pressure and Temperature (P/T) Limits

ii) Surveillance Requirements Section 3.4.9, RCS Pressure and Temperature (P/T) Limits

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

i) NEDC-33178P-A, GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, Revision 1, June 2009

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

5.6.8 OPRM Report

When an OPRM report is required by CONDITION I of LCO 3.3.1.1, "RPS Instrumentation," the repo rt shall be submitted within the following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPE RABLE status.

PBAPS UNIT 2 5.0- 22 Amendment No. 305

2. Level 1 performance criteria.

3. The methodology for establishing the limit curves used for the Level 1 and Level 2 performance.

(e) The results of the power ascension testing to verify the continued structural integrity of the steam dryer shall be submitted to the NRC staff in a report in accordance with 10 CFR 50.4. The report shall include a final load definition and stress report of the steam dryer, including the results of a complete re-analysis using the end-to-end B/Us from Peach Bottom Unit 2 benchmarking at EPU conditions.

The report shall be submitted within 90 days of the completion of EPU power ascension testing for Peach Bottom Unit 3.

(f) During the first two scheduled refueling outages after reaching EPU conditions, a visual inspection shall be conducted of the steam dryer as described in the inspection guidelines contained in WCAP-17635-P.

(g) The results of the visual inspections of the steam dryer shall be submitted to the NRC staff in a report in accordance with 10 CFR 50.4. The report shall be submitted within 90 days following startup from each of the first two respective refueling outages.

(h) Within 6 months following completion of the second refueling outage, after the implementation of the EPU, the licensee shall submit a long-term steam dryer inspection plan based on industry operating experience along with the baseline inspection results.

The license condition described above shall expire: (1) upon satisfaction of the requirements in paragraphs (f) and (g), provided that a visual inspection of the steam dryer does not reveal any new unacceptable flaw(s) or unacceptable flaw growth that is due to fatigue, and; (2) upon satisfaction of the requirements specified in paragraph (h).

(16) Maximum Extended Load Line Limit Analysis Plus (MELLLA+) Special Consideration

The licensee shall not operate the facility within the MELLLA+ operating domain with a feedwater heater out of service resulting in more than a 10°F reduction in feedwater temperature below the design feedwater temperature.

(17) Adoption of 10 CFR 50.69, "Risk-informed Categorization and Treatment of Structures, Systems, and Components for Nuclear Power Plants"

In support of implementing License Amendment No. 324 permitting the adoption of the provisions of 10 CFR 50.69 for Renewed Facility Operating License No. DPR-56 for Peach Bottom Unit 3, the license is amended to add the following license condition:

Page 12 Subsequent Renewed License No. DPR-56 Amendment No. 336 (a) The licensee is approved to implement 10 CFR 50.69 using th e processes for categorization of Risk-Informed Safety Class (RIS C)-1, RISC-2, RISC-3, and RISC-4 structures, systems, and components (SSCs) using: Probabilistic Risk Assessment (PRA) models to evaluate risk associated with internal events, including internal flooding, and internal fire; 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 the results of non-PRA evaluations that are based on the IPEEE Screening Assessment for External Hazards, i.e., seismic margin analysis (SMA) to evaluate seismic risk, 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 Unit 3 License Amendment No. 324 dated October 25, 2018.

The licensee will complete the implementation items listed in Attachment 2 of Exelons letter to the NRC dated June 6, 2018, prior to implementation of 10 CFR 50.69. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the 10 CFR 50.69 categorization process.

Prior NRC approval, under 10 CFR 50.90, is required for a change to the categorization process specified above (e.g., change from a seismic margins approach to a seismic probabilistic risk assessment approach).

(18) This subsequent renewed license is subject to the following conditions for the protection of the environment:

(a) To the extent matters related to thermal discharges are tre ated therein, operation of Peach Bottom Atomic Power Station, Unit No. 3, will be governed by NPDES Permit No. PA 0009733, as now in effe ct and as hereafter amended. Questions pertaining to conformance thereto shall be referred to and shall be determined by the NPD ES Permit issuing or enforcement authority, as appropriate.

(b) In the event of any modification of the NPDES Permit related to thermal discharges or the establishment (or amendment) of alternative effluent limitations established pursuant to Section 316 of the Federal Water Pollution Control Act, the licensee shall inf orm the NRC and analyze any associated changes in or to the Station, it s components, its operation or in the discharge of effluents therefrom.

If such change would entail any modification to this license, or any Technical Specifications which are part of this license, or req uire NRC approval pursuant to 10 CFR 50.59 or involve an environment al

Page 13 Subsequent Renewed License No. DPR-56 Amendment No. 343

2. Constellation Energy Generation, LLC 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.

3. Constellation Energy Generation, LLC shall notify the NRC in writing within 30 days after having accomplished item (b)1 above and include the status of those activities that have been or remain to be completed in item (b)2 above.

(20) PRA Model Updates to Support Implementation of the Risk Informed Completion Time (RICT) Program

Constellation Energy Generation, LLC is approved to implement TSTF-505, Revision 2, modifying the Technical Specification requirements related to Completion Times (CT) for Required Actions to provide the option to calculate a longer, risk-informed CT. The methodology for using the new Risk-Informed Completion Time (RICT) Program is described in NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, which was approved by the NRC on May 17, 2007.

Constellation Energy Generation, LLC will complete the implementation items listed in Attachment 6 of Exelon letter to the NRC dated May 29, 2020, prior to implementation of the RICT Program. All issues identified in the attachment will be addressed and any associated changes will be made, focused-scope peer reviews will be performed on changes that are PRA upgrades as defined in the PRA standard (ASME/ANS RA-Sa-2009, as endorsed by RG 1.200, Revision 2), and any findings will be resolved and reflected in the PRA of record prior to implementation of the RICT Program.

3. This subsequent renewed license is effective as of the date of issuance and shall expire at midnight on July 2, 2034.

FOR THE UNITED STATES NUCLEAR REGULATORY COMMISSION

/RA/

Ho K. Nieh, Director Office of Nuclear Reactor Regulation

Attachments:

Appendix A - Technical Specifications Peach Bottom Atomic Power Station Unit No. 3 Appendix B - Environmental Protection Plan

Date of Issuance: March 5, 2020

Page 15 Subsequent Renewed License No. DPR-56 Amendment No. 343 Order CLI-22-04 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.15 Battery Monitoring and Maintenance Program (continued)

1. Actions to restore battery cells with float voltage

< 2.13 V;

2. Actions to determine whether the float voltage of the remaining battery cells is > 2.13 V when the float voltage of a battery cell has been f ound to be < 2.13 V;

3. Actions to equalize and test batt ery cells that had been discovered with electrolyte level below the top of the plates;

4. Limits on average electrolyte temperature, battery connection resistance, and battery terminal voltage; and

5. A requirement to obtain specific gravity readings of all cells at each discharge test, consistent with manufacturer recommendations.

5.5.16. Risk Informed Completion Time Program

This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06- 09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS)

Guidelines." The program shall include the following:

a. The RICT may not exceed 30 days.

b. A RICT may only be utilized in MODEs 1 and 2.

c. When a RICT is being used, any change t o the plant configuration, as defined in NEI 06- 09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent condition s, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration chan ge would lower plant risk and would result in a longer RICT.

PBAPS UNIT 3 5.0- 18c Amendment No. 341 Programs and Manuals 5.5

5.5 Programs and Manuals

5.5.16. Risk Informed Completion Time Program (continued)

d. For emergent conditions, if the exten t of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically a ccounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e. The risk assessment approaches and methods shall be acceptable to the NRC. The plant PRA shall be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant, as specified in Regulatory Guide 1.200, Revision 2. Methods to assess the risk from extending the Completion Times must be PRA methods approved for use with this program in Amendment No. 341, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

PBAPS UNIT 3 5.0- 18d Amendment No. 341 Reporting Requirements 5.6

5.6 Reporting Requirements (continued)

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report

When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

i) Limiting Conditions for Operation Section 3.4.9, RCS Pressure and Temperature (P/T) Limits

ii) Surveillance Requirements Section 3.4.9, RCS Pressure and Temperature (P/T) Limits

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

i) NEDC-33178P-A, GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves, Revision 1, June 2009

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

5.6.8 OPRM Report

When an OPRM report is required by CONDITION I of LCO 3.3.1.1, "RPS Instrumentation," the report shall be submitted within the following 90 days. The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to OPERABILE status.

PBAPS UNIT 3 5.0- 22 Amendment No. 309 Insert 1:

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA model, seismic hazards using penalty factors, and tornado missile hazards using penalty factors. Changes to these means of assessing the haz ard groups require prior NRC approval.

f. The PRA models used to calculate a RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, R evision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed A ctivities."

g. A r eport shall be submitted in accordance with Specification 5.6.9 before a newly developed method is us ed to calculate a RICT.

Insert 2:

5.6.9 Risk Informed Completion Time (RICT) Program Upgrade Report

A r eport describing newly deve loped methods and their i mplementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a. The PRA models upgraded to include newly developed methods;

b. A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c. Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d. All changes to key assumptions related to newly developed methods or their implementations.