L-24-063, License Amendment Request to Remove the Table of Contents from the Technical Specifications

License Amendment Request to Remove the Table of Contents from the Technical Specifications

ML24191A432
Person / Time
Site:	Davis Besse
Issue date:	07/08/2024
From:	Tony Brown Vistra Operations Company
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
L-24-063
Download: ML24191A432 (1)

Text

Davis-Besse Nuclear Power Station Terry J. Brown Site Vice President 5501 N. State Route 2 Oak Harbor, Ohio 43449 419.321.7676 L-24-063 July 8, 2024 10 CFR 50.90

A TIN : Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 License Amendment Request to Remove the Table of Contents from the Technical Specifications

In accordance with 10 CFR 50.90, Vistra Operations Company LLC hereby requests an amendment to Facility Operating License No. NPF-3 to revise the Technical Specifications (I'S) for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS). The proposed amendment would remove the table of contents (TOC) from the TS and place it under licensee control.

provides the evaluation of the proposed change. Attachment 2 provides the existing TS pages marked up to show the proposed change.

The proposed amendment does not involve a significant hazard consideration as determined per 10 CFR 50.92, "Issuance of amendment." Also, pursuant to 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," Section (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

• Approval of the proposed amendment is requested by December 1, 2024. Once approved, the amendment shall be implemented within 90 days.

There are no regulatory commitments contained in this submittal. If there are any questions, or if additional information is required, please contact Jack Hicks, Senior Manager, Fleet Licensing, at (254) 897-6725 or jack.hicks @luminant.com.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July I'.'"~ 2024.

655'- SIERRA DRIVE IRVING TFXAS 75039 0 214 812 4600 VISTRACORP.COM Davis-Besse Nuclear Power Station, Unit No. 1 L-24-063 Page 2

Attachments:

1. Evaluation of the Proposed Change

2. Proposed Technical Specification Changes (Mark-Up)

cc: NRC Region III Administrator NRC Resident Inspector NRR Project Manager Executive Director, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board

Attachment 1 L-24-063

Evaluation of the Proposed Change Page 1 of 6

Subject:

Removal of the Table of Contents from the Technical Specifications 1.0

SUMMARY

DESCRIPTION

2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration Analysis 4.4 Conclusions

6.0 REFERENCES

L-24-063 Page 2 of 6

1.0

SUMMARY

DESCRIPTION

This evaluation supports the Vistra Operations Company LLC request to amend Facility Operating License No. NPF-3 for Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS).

The proposed changes would revise the Technical Specifications (TS) to remove the table of contents (TOC) from the TS and place it under licensee control.

2.0 DETAILED DESCRIPTION

The proposed change would remove the TOC from the TS and place it under Vistra Operations Company LLC (licensee) control. No TS pages will be issued as a result of this proposed change. The TOC for the TS is not being eliminated. Following approval of this license amendment request (LAR), the responsibility for maintenance and issuance of updates to the TS TOC will transfer from the Nuclear Regulatory Commission (NRC) to Vistra Operations Company LLC. The TOC will no longer be included in the TS and as such will no longer be part of the TS (Appendix A to the Facility Operating License).

3.0 TECHNICAL EVALUATION

The TOC does not meet the criteria specified in 10 CFR 50.36 requiring its inclusion within a plants TS. 10 CFR 50.36(b) states, in part:

Each license authorizing operation of a production or utilization facility of a type described in § 50.21 or § 50.22 will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to § 50.34.

The result of the Vistra Operations Company LLC review indicates that the TS TOC was not listed in the regulation as one of the categories listed in 10 CFR 50.36(c).

10 CFR 50.36(a) indicates that a license application may provide other information associated with the TS and gives an example of this information (the TS bases), but also clearly indicates that the bases are not a part of the TS. 10 CFR 50.36(a)(1) states:

Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

Additionally, the Writers Guide for Plant-Specific Improved Technical Specifications (Reference 1) was reviewed for guidance. The writers guide refers to the TOC as Technical Specification Front Matter, which also includes the Title Page and List of Effective Pages. The writers guide describes the TS content as starting with Chapter 1, Use and Application, and does not include the TOC as part of the content of the TS.

L-24-063 Page 3 of 6

A TOC for the TS will be maintained under Vistra Operations Company LLC control. The TOC will be issued by Vistra Operations Company LLC in conjunction with the implementation of NRC-approved TS amendments that alter the TOC.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria

10 CFR 50.36(a)(1) states:

Each applicant for a license authorizing operation of a production or utilization facility shall include in his application proposed technical specifications in accordance with the requirements of this section. A summary statement of the bases or reasons for such specifications, other than those covering administrative controls, shall also be included in the application, but shall not become part of the technical specifications.

Like the TS bases, the TOC does not meet the criteria specified in 10 CFR 50.36 for inclusion within a plants TS.

Vistra Operations Company LLC has evaluated this change against the applicable regulatory requirements as described above. Based on this, there is reasonable assurance that the health and safety of the public, following approval of this change, will remain unaffected.

4.2 Precedent

A similar request was approved for Perry Nuclear Power Plant, Unit 1 (Reference 2) on November 17, 2023. Amendment 201 was issued to remove the TOC from the TS.

A similar request was approved for Wolf Creek Generating Station, Unit 1 (Reference 3) on August 16, 2022. Amendment 233 was issued to remove the TOC from the TS.

A similar request was approved for Exelon Generation Company, LLC units (Reference 4) on December 18, 2019. The amendments issued to remove the TOC from the TS for each unit are as follows:

- Amendments 204 and 204 for Braidwood Station, Units 1 and 2, respectively;

- Amendments 210 and 210 for Byron Station, Units 1 and 2, respectively;

- Amendments 264 and 257 for Dresden Nuclear Power Station, Units 2 and 3, respectively;

- Amendment 330 for James A. FitzPatrick Nuclear Power Plant;

- Amendment 135 for R. E. Ginna Nuclear Power Plant;

- Amendments 240 and 226 for LaSalle County Station, Units 1 and 2, respectively;

- Amendments 238 and 201 for Limerick Generating Station, Units 1 and 2, respectively;

- Amendments 240 and 178 for Nine Mile Point Nuclear Station, Units 1 and 2, respectively; L-24-063 Page 4 of 6

- Amendments 330 and 333 for Peach Bottom Atomic Power Station, Units 2 and 3, respectively;

- Amendments 277 and 272 for Quad Cities Nuclear Power Station, Units 1 and 2, respectively.

4.3 No Significant Hazards Consideration Analysis

In accordance with 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit, Vistra Operations Company LLC requests an amendment to Facility Operating License No. NPF-3 to revise the Technical Specifications (TS) for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS). The proposed amendment would remove the table of contents (TOC) from the TS and place it under licensee control.

Vistra Operations Company LLC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, 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 amendment is administrative and affects control of a document, the TOC, listing the specifications in the plant TS. Transferring control from the Nuclear Regulatory Commission (NRC) to Vistra Operations Company LLC does not affect the operation, physical configuration, or function of plant equipment or systems. The proposed amendment does not impact the initiators or assumptions of analyzed events; nor does it impact the mitigation of accidents or transient events.

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 amendment does not involve a physical alteration of the plant (that is, no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. The proposed amendment does not alter any assumptions made in the safety analysis.

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

L-24-063 Page 5 of 6

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

Response: No

The proposed amendment is administrative. The TOC is not required by regulation to be in the TS. Removal does not impact any safety assumptions or have the potential to reduce a margin of safety. The proposed amendment involves a transfer of control of the TOC from the NRC to Vistra Operations Company LLC. No change in the technical content of the TS is involved.

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

Based on the above, Vistra Operations Company LLC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.5 Conclusions

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.

5.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(10). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Technical Specifications Task Force, TSTF-GG-05-01, Writers Guide for Plant-Specific Improved Technical Specifications, Revision 1, dated August 2010.

2. Letter from Scott P. Wall (U.S. NRC) to Mr. Rod L. Penfield (Energy Harbor Nuclear Corp.), Perry Nuclear Power Plant, Unit No. 1 - Issuance of Amendment No. 201 Regarding Removal of Table of Contents from the Technical Specifications (EPID L-2023-LLA-0111), dated November 17, 2023 (ADAMS Accession No. ML23292A297).

L-24-063 Page 6 of 6

3. Letter from Samson S. Lee (U.S. NRC) to Mr. Cleveland Reasoner (Wolf Creek Nuclear Operating Corporation), Wolf Creek Generating Station, Unit 1 - Issuance of Amendment No. 233 RE: Removal of Table of Contents from the Technical Specifications (EPID L-2022-LLA-0008), dated August 16, 2022 (ADAMS Accession No. ML22199A294).

4. Letter from Blake Purnell (U.S. NRC) to Mr. Bryan C. Hanson (Exelon Generation Company, LLC), Braidwood Station, Units 1 and 2; Byron Station, Unit Nos. 1 and 2; Dresden Nuclear Power Station, Units 2 and 3; James A. FitzPatrick Nuclear Power Plant; R. E. Ginna 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 Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments to Remove Table of Contents from Technical Specifications (EPID L-2019-LLA-0125), dated December 18, 2019 (ADAMS Accession No. ML19302E700).

Attachment 2 L-24-063

Proposed Technical Specification Changes (Mark-Up)

(4 pages follow)

TABLE OF CONTENTS Page Number

1.0 USE AND APPLICATION 1.1 Definitions..............................................................................................................1.1-1 1.2 Logical Connectors................................................................................................1.2-1 1.3 Completion Times.................................................................................................1.3-1 1.4 Frequency.............................................................................................................1.4-1

2.0 SAFETY LIMITS (SLs)................................................................................................ 2.0-1 2.1 SLs 2.2 SL Violations

3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY........................... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY.......................................... 3.0-4

3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)...........................................................................3.1.1-1 3.1.2 Reactivity Balance..............................................................................................3.1.2-1 3.1.3 Moderator Temperature Coefficient (MTC)........................................................ 3.1.3-1 3.1.4 CONTROL ROD Group Alignment Limits.......................................................... 3.1.4-1 3.1.5 Safety Rod Insertion Limits................................................................................ 3.1.5-1 3.1.6 AXIAL POWER SHAPING ROD (APSR) Alignment Limits................................ 3.1.6-1 3.1.7 Position Indicator Channels................................................................................ 3.1.7-1 3.1.8 PHYSICS TESTS Exceptions - MODE 1........................................................... 3.1.8-1 3.1.9 PHYSICS TESTS Exceptions - MODE 2........................................................... 3.1.9-1

3.2 POWER DISTRIBUTION LIMITS 3.2.1 Regulating Rod Insertion Limits......................................................................... 3.2.1-1 3.2.2 AXIAL POWER SHAPING ROD (APSR) Insertion Limits.................................. 3.2.2-1 3.2.3 AXIAL POWER IMBALANCE Operating Limits.................................................. 3.2.3-1 3.2.4 QUADRANT POWER TILT (QPT)..................................................................... 3.2.4-1 3.2.5 Power Peaking Factors...................................................................................... 3.2.5-1

3.3 INSTRUMENTATION 3.3.1 Reactor Protection System (RPS) Instrumentation............................................ 3.3.1-1 3.3.2 Reactor Protection System (RPS) Manual Reactor Trip.................................... 3.3.2-1 3.3.3 Reactor Protection System (RPS) - Reactor Trip Module (RTM)....................... 3.3.3-1 3.3.4 CONTROL ROD Drive (CRD) Trip Devices....................................................... 3.3.4-1 3.3.5 Safety Features Actuation System (SFAS) Instrumentation.............................. 3.3.5-1 3.3.6 Safety Features Actuation System (SFAS) Manual Initiation............................. 3.3.6-1 3.3.7 Safety Features Actuation System (SFAS) Automatic Actuation Logic.............................................................................................3.3.7-1 3.3.8 Emergency Diesel Generator (EDG) Loss of Power Start (LOPS).................... 3.3.8-1 3.3.9 Source Range Neutron Flux............................................................................... 3.3.9-1 3.3.10 Intermediate Range Neutron Flux.................................................................... 3.3.10-1 3.3.11 Steam and Feedwater Rupture Control System (SFRCS)

Instrumentation...........................................................................................3.3.11-1 3.3.12 Steam and Feedwater Rupture Control System (SFRCS)

Manual Initiation.........................................................................................3.3.12-1 3.3.13 Steam and Feedwater Rupture Control System (SFRCS)

Actuation....................................................................................................3.3.13-1

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3.3 INSTRUMENTATION (continued)

3.3.14 Fuel Handling Exhaust - High Radiation.......................................................... 3.3.14-1 3.3.15 Station Vent Normal Range Radiation Monitoring............................................ 3.3.15-1 3.3.16 Anticipatory Reactor Trip System (ARTS) Instrumentation.............................. 3.3.16-1 3.3.17 Post Accident Monitoring (PAM) Instrumentation............................................. 3.3.17-1 3.3.18 Remote Shutdown System............................................................................... 3.3.18-1

3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits.....................................................................................3.4.1-1 3.4.2 RCS Minimum Temperature for Criticality.......................................................... 3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits.................................................... 3.4.3-1 3.4.4 RCS Loops - MODES 1 and 2............................................................................ 3.4.4-1 3.4.5 RCS Loops - MODE 3........................................................................................ 3.4.5-1 3.4.6 RCS Loops - MODE 4........................................................................................ 3.4.6-1 3.4.7 RCS Loops - MODE 5, Loops Filled................................................................... 3.4.7-1 3.4.8 RCS Loops - MODE 5, Loops Not Filled............................................................ 3.4.8-1 3.4.9 Pressurizer.........................................................................................................3.4.9-1 3.4.10 Pressurizer Safety Valves................................................................................ 3.4.10-1 3.4.11 Pressurizer Pilot Operated Relief Valve (PORV)............................................. 3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP)........................................ 3.4.12-1 3.4.13 RCS Operational LEAKAGE............................................................................ 3.4.13-1 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage.................................................. 3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation......................................................... 3.4.15-1 3.4.16 RCS Specific Activity........................................................................................ 3.4.16-6 3.4.17 Steam Generator (SG) Tube Integrity.............................................................. 3.4.17-1

3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Core Flooding Tanks (CFTs).............................................................................. 3.5.1-1 3.5.2 ECCS - Operating..............................................................................................3.5.2-1 3.5.3 ECCS - Shutdown..............................................................................................3.5.3-1 3.5.4 Borated Water Storage Tank (BWST)................................................................ 3.5.4-1

3.6 CONTAINMENT SYSTEMS 3.6.1 Containment.......................................................................................................3.6.1-1 3.6.2 Containment Air Locks....................................................................................... 3.6.2-1 3.6.3 Containment Isolation Valves............................................................................. 3.6.3-1 3.6.4 Containment Pressure........................................................................................3.6.4-1 3.6.5 Containment Air Temperature............................................................................ 3.6.5-1 3.6.6 Containment Spray and Air Cooling Systems.................................................... 3.6.6-1 3.6.7 Trisodium Phosphate Dodecahydrate (TSP) Storage........................................ 3.6.7-1

3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs).................................................................. 3.7.1-1 3.7.2 Main Steam Isolation Valves (MSIVs)................................................................ 3.7.2-1 3.7.3 Main Feedwater Stop Valves (MFSVs), Main Feedwater Control Valves (MFCVs), and associated Startup Feedwater Control Valves (SFCVs)...........................................................................................3.7.3-1

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3.7 PLANT SYSTEMS (continued)

3.7.4 Turbine Stop Valves (TSVs)............................................................................... 3.7.4-1 3.7.5 Emergency Feedwater (EFW)............................................................................3.7.5-1 3.7.6 Condensate Storage Tanks (CSTs)................................................................... 3.7.6-1 3.7.7 Component Cooling Water (CCW) System........................................................ 3.7.7-1 3.7.8 Service Water System (SWS)............................................................................ 3.7.8-1 3.7.9 Ultimate Heat Sink (UHS)................................................................................... 3.7.9-1 3.7.10 Control Room Emergency Ventilation System (CREVS)................................. 3.7.10-1 3.7.11 Control Room Emergency Air Temperature Control System (CREATCS)....... 3.7.11-1 3.7.12 Station Emergency Ventilation System (EVS).................................................. 3.7.12-1 3.7.13 Spent Fuel Pool Area Emergency Ventilation System (EVS)........................... 3.7.13-1 3.7.14 Spent Fuel Pool Water Level............................................................................ 3.7.14-1 3.7.15 Spent Fuel Pool Boron Concentration.............................................................. 3.7.15-1 3.7.16 Spent Fuel Pool Storage.................................................................................. 3.7.16-1 3.7.17 Secondary Specific Activity.............................................................................. 3.7.17-1 3.7.18 Steam Generator Level.................................................................................... 3.7.18-1

3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating..................................................................................... 3.8.1-1 3.8.2 AC Sources - Shutdown..................................................................................... 3.8.2-1 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air.......................................................... 3.8.3-1 3.8.4 DC Sources - Operating..................................................................................... 3.8.4-1 3.8.5 DC Sources - Shutdown..................................................................................... 3.8.5-1 3.8.6 Battery Parameters............................................................................................3.8.6-1 3.8.7 Inverters - Operating.......................................................................................... 3.8.7-1 3.8.8 Inverters - Shutdown.......................................................................................... 3.8.8-1 3.8.9 Distribution Systems - Operating........................................................................ 3.8.9-1 3.8.10 Distribution Systems - Shutdown..................................................................... 3.8.10-1

3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration..........................................................................................3.9.1-1 3.9.2 Nuclear Instrumentation.....................................................................................3.9.2-1 3.9.3 Decay Time........................................................................................................3.9.3-1 3.9.4 Decay Heat Removal (DHR) and Coolant Circulation - High Water Level........ 3.9.4-1 3.9.5 Decay Heat Removal (DHR) and Coolant Circulation - Low Water Level......... 3.9.5-1 3.9.6 Refueling Canal Water Level.............................................................................. 3.9.6-1

4.0 DESIGN FEATURES..................................................................................................4.0-1 4.1 Site Location 4.2 Reactor Core 4.3 Fuel Storage

5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility........................................................................................................5.1-1 5.2 Organization..........................................................................................................5.2-1 5.3 Unit Staff Qualifications......................................................................................... 5.3-1 5.4 Procedures............................................................................................................5.4-1 5.5 Programs and Manuals......................................................................................... 5.5-1

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5.0 ADMINISTRATIVE CONTROLS (continued)

5.6 Reporting Requirements.......................................................................................5.6-1 5.7 High Radiation Area.............................................................................................. 5.7-1

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