Text

License Amendment Request to Remove Licensee Control (BFN TS-527, SQN-TS-20-08, and WBN-TS-20-016)
	ML20230A210
Person / Time
Site:	Browns Ferry, Watts Bar, Sequoyah
Issue date:	08/14/2020
From:	Jim Barstow; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
References
	CNL-20-041
	Download: ML20230A210 (54)
	v • d • e

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-20-041 August 14, 2020 10 CFR 50.90 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, 50-296, and 72-052 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327, 50-328, and 72-034 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390, 50-391, and 72-1048

Subject:

Tennessee Valley Authority License Amendment Request to Remove Table of Contents from Technical Specifications and Place Under Licensee Control (BFN TS-527, SQN-TS-20-08, and WBN-TS-20-016)

In accordance with Title10 of the Code of Federal Regulations (10 CFR) 50.90, "Application for amendment of license, construction permit, or early site permit," Tennessee Valley Authority (TVA) requests amendments to the Technical Specifications (TS) for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3; Sequoyah Nuclear Plant (SQN), Units 1 and 2; and Watts Bar Nuclear Plant (WBN), Units 1 and 2, to remove the table of contents (TOC) from the TS and place it under licensee control. Two additional ancillary administrative changes are made to the WBN, Unit 1 and SQN, Units 1 and 2 TS that resulted from a review of the TOC.

The Enclosure provides an evaluation of the proposed changes. Attachment 1 to the Enclosure provides a markup of the TS for the three sites. Attachment 2 to the Enclosure provides the re-typed pages to the WBN, Unit 1 and SQN, Units 1 and 2 TS.

U.S. Nuclear Regulatory Commission CNL-20-041 Page 2 August 14, 2020 TVA has determined that there are no significant hazards considerations associated with the proposed changes and that the change qualifies for a categorical exclusion from environmental review pursuant to 10 CFR 51.22(c)(9). Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter and enclosures to the Alabama State Department of Public Health and the Tennessee State Department of Environment and Conservation.

TVA requests approval of the proposed license amendment within one year from the date of this submittal with implementation within 30 days following NRC approval.

There are no new regulatory commitments contained in this submittal. If you have any questions concerning this submittal, please contact Gordon R. Williams, Senior Manager, Fleet Licensing (Acting), at (423) 751-2687.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 14th day of August 2020.

Sincerely, James Barstow Vice President, Nuclear Regulatory Affairs & Support Services Enclosure Evaluation of Proposed Change cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Senior Resident Inspector - Sequoyah Nuclear Plant NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant NRC Project Manager - Sequoyah Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant State Health Officer, Alabama State Department of Public Health Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation

ENCLOSURE EVALUATION OF THE PROPOSED CHANGE

Subject:

Tennessee Valley Authority License Amendment Request to Remove Table of Contents from Technical Specifications and Place Under Licensee Control (BFN TS-527, SQN-TS-20-08, and WBN-TS-20-016)

Contents 1.0

SUMMARY

DESCRIPTION ................................................................................... 2 2.0 DETAILED DESCRIPTION .................................................................................... 2

3.0 TECHNICAL EVALUATION

.................................................................................. 3

4.0 REGULATORY EVALUATION

.............................................................................. 3 4.1 Applicable Regulatory Requirements/Criteria ................................................ 3 4.3 No Significant Hazards Consideration ........................................................... 4 4.4 Conclusions ...................................................................................................... 5

5.0 ENVIRONMENTAL CONSIDERATION

................................................................. 5

6.0 REFERENCES

....................................................................................................... 6 ATTACHMENTS

1. Proposed Technical Specification Changes (Mark-Ups) for BFN Units 1, 2, and 3, SQN Units 1 and 2, and WBN Units 1 and 2

2. Re-typed Technical Specification Pages for WBN Unit 1 and SQN Units 1 and 2 CNL-20-041 E1-6

1.0

SUMMARY

DESCRIPTION This evaluation supports a request for the Nuclear Regulatory Commission (NRC) to amend Renewed Facility Operating License (RFOL) Nos. DPR-33, DPR-52, and DPR-68 for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3; RFOL Nos.

DPR-77 and DPR-79 for Sequoyah Nuclear Plant (SQN), Units 1 and 2; and Facility Operating License (FOL) Nos. NPF-90 and NPF-96 for Watts Bar Nuclear Plant (WBN), Units 1 and 2. The proposed change would revise the RFOLs and FOLs to remove the table of contents (TOC) from the technical specifications (TS). The TOC is not being eliminated, as it will continue to be published with the site-controlled version of the TS. However, following approval of this License Amendment Request (LAR), revisions to the TOC will be made administratively under licensee control, and without NRC review and approval. Additionally, as part of reviewing the TOC for WBN and SQN, two additional administrative changes were identified for WBN Unit 1 and SQN Units 1 and 2 TS.

2.0 DETAILED DESCRIPTION The proposed change would revise the RFOLs and FOLs to remove the TOC from the TS. For purposes of this LAR, the TOC includes other front matter, such as List of Tables, List of Figures, and List of Acronyms, as applicable, in the respective BFN, SQN, and WBN TS.

The TOC for the TS is not being eliminated. Rather, following approval of this LAR, responsibility for maintenance and issuance of updates to the TS TOC will transfer from the NRC to the Tennessee Valley Authority (TVA). The TOC will no longer be included in the NRC issued TS and as such will no longer be part of the TS (Appendix A to the RFOLs and Operating Licenses). A TOC for the TS will be maintained under TVA control, and TOC changes will be issued by TVA in conjunction with the implementation of future NRC-approved TS amendments.

Placing the TOC under licensee control eliminates the regulatory burden of submitting TOC pages for NRC review and allows timely administrative corrections and improvements to the TOC without NRC review and approval.

As part of reviewing the TOCs for this application, it was noted the WBN, Unit 1 TS Page 3.7-1 does not lead with the title 3.7 PLANT SYSTEMS (as shown by the Westinghouse Standard Technical Specifications (NUREG-1431)), and is revised.

This inconsistency has been present on this page since issuance of the License Amendment 19 (ML003691130). It was also noted that the SQN, Unit 1 and 2 TS Pages 3.2.2-1 do not contain parenthesis around the term FH(X,Y) where it is first defined (also as shown by NUREG-1431 for the analogous TS), and is revised. This inconsistency has been present since the issuance of License Amendments 334/327 for SQN, Units 1 and 2, respectively (ML15238B460).

CNL-20-041 E2-6

3.0 TECHNICAL EVALUATION

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

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.

Additionally, 10 CFR 50.36(c) states the "Technical specifications will include items in the following categories." Review of 10 CFR 50.36 indicates that a TOC was not listed in the regulation as one of the categories.

The TOC references the page number of each Specification in the TS and does not contain any technical information required by 10 CFR 50.36. Because the TOC does not include information required to be in the TS by 10 CFR 50.36, inclusion of a TOC within the TS is optional. Removal of the TOC from the TS is an administrative change and is acceptable.

Additionally, the "Writer's Guide for Plant-Specific Improved Technical Specifications (Reference 1) was reviewed for guidance. The writer's guide refers to the TOC as "Technical Specification Front Matter," which also includes the Title Page and List of Effective Pages. These portions of the TS are under licensee control, as proposed for the TOC.

The TS TOC will be maintained, revised, and distributed in accordance with TVA administrative procedures. Holders of controlled copies of the TS will continue to receive periodic updates of the TOC pages resulting from approved license amendments.

The addition of 3.7 PLANT SYSTEMS below the header on Page 3.7-1 of the WBN, Unit 1 TS (consistent with Reference 1) is an administrative change. Similarly, adding parenthesis around the term FH(X,Y) to Page 3.2.2-1 of the SQN, Units 1 and 2 TS (where it is first defined) is an administrative change.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The NRC's regulation at Title 10 of the Code of Federal Regulations (10 CFR),

Paragraph 50.36, contains requirements for what is to be included in a licensee's TS.

The Table of Contents, which contains only administrative information, is not required to be in the TS per the requirements of 10 CFR 50.36.

4.2 Precedent A license amendment was issued to the Exelon fleet to remove the TOC from the TS (Reference 2) on December 18, 2019. While this is the most recent precedent, there are numerous others that preceded this.

CNL-20-041 E3-6

4.3 No Significant Hazards Consideration In accordance with the requirements of 10 CFR 50.90, Tennessee Valley Authority (TVA) requests an amendment to remove the table of contents (TOC) from the Technical Specifications (TS) of Browns Ferry Nuclear Plant (BFN), Sequoyah Nuclear Plant (SQN), and Watts Bar Nuclear Plant (WBN) and place it under licensee control. Two associated minor formatting changes are also made to Page 3.7-1 of the WBN, Unit 1 TS to add 3.7 PLANT SYSTEMS below the header, and to Pages 3.2.2-1 of the SQN, Units 1 and 2 TS to add parenthesis around the term FH(X,Y) where it is first defined. TVA has evaluated the proposed amendment in accordance with 10 CFR 50.91 against the standards in 10 CFR 50.92 and has determined that the operation of the facility in accordance with the proposed amendment presents no significant hazards. TVAs evaluation against each of the criteria in 10 CFR 50.92 follows.

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

Response: No.

The proposed change is administrative and affects control of a document, the TOC, listing and providing the page number of the specifications in the plant TS.

Transferring control of changes to the TOC to the licensee does not affect the operation, physical configuration, or function of plant equipment or systems. It does not affect the initiators or assumptions of analyzed events; nor does it affect the mitigation of accidents or transient events. The change has no effect on the probability or consequences of an accident.

The formatting changes to the WBN and SQN TS are similarly administrative with no effect on the probability or consequences of an accident.

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 to place the TOC under licensee control is administrative and does not alter the plant configuration, require installation of new equipment, alter assumptions about previously analyzed accidents, or affect the operation or function of plant equipment or systems.

The formatting changes to the WBN and SQN TS are similarly administrative with no interface with previously analyzed accidents.

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

CNL-20-041 E4-6

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

Response: No.

The proposed change is administrative. The TOC is not required by regulation to be included in the operating license or the attached TS. Transfer of the TOC from NRC control to TVA control does not affect any safety assumptions or have the potential to reduce a margin of safety as described in the TS Bases. No change in the technical content of the TS specifications is involved.

The formatting changes to the WBN and SQN TS are similarly administrative with no effect on any safety margins.

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

Based on the above, TVA 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.

4.4 Conclusions In conclusion, and based on the considerations discussed above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by the removal of the TS TOC, and the WBN and SQN TS formatting changes; (2) the changes will be in compliance with NRC regulations; and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

TVA has determined that the proposed amendment does not change any requirements 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. The proposed amendment does not involve (i) a significant hazards consideration, or (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for a categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, TVA concludes pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

CNL-20-041 E5-6

6.0 REFERENCES

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

2. NRC Letter to 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 (ML19302E700)

CNL-20-041 E6-6

ATTACHMENT 1 Proposed Technical Specification Changes (Mark-Ups) for BFN Units 1, 2, and 3, SQN Units 1 and 2, and WBN Units 1 and 2 CNL-20-041

TABLE OF CONTENTS Section . Page No.

1.0 USE AND APPLICATION ................. 1.1-1 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 ................. 2.0-1 2.2 SL Violations ......... ;.2.0-1 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 3.1.1 SHUTDOWN MARGIN (SDM) ............................................. 3.1-1 3.1.2 Reactivity Anomalies .............................................. 3.1-5 3.1.3 Control Rod OPERABILITY .............................................. 3.1-7 3.1.4 Control Rod Scram Times .............................................. 3.1-12 3.1.5 Control Rod Scram Accumulators ........................................ 3.1-16 3.1.6' Rod Pattern Control ....................... ....................... 3.1-20 3.1.7 Standby Liquid Control (SLC) System.................................. 3.1-23.

3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves ..... 3.1-28 3.2 POWER DISTRIBUTION LIMITS ........................... 3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ................................... 3.2-1 3.2;2 MINIMUM CRITICAL POWER RATIO (MCPR) ..................... 3.2-3 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ..................... 3.2-5 3.3 INSTRUMENTATION ................................... 3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation .3.3-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation .................... 3.3-10 3.3.2.1 Control Rod Block Instrumentation ................................... 3.3-16 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation .................................. 3.3-22 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation ................. 3.3-24 3.3.3.2 Backup Control System ..................................3.3-28 BFN-UNIT 2 i Amendment No. 253

Section Page No.

3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation............................................................... 3.3-30 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ....................... 3.3-33 3.3.5.1 Emergency Core Cooling System (ECCS)

Instrumentation............................................................... 3.3-36 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation............................................................... 3.3-49 3.3.6.1 Primary Containment Isolation Instrumentation .3.3-53 3.3.6.2 Secondary Containment Isolation Instrumentation .3.3-62 3.3.7.1 Control Room Emergency Ventilation (CREV)

System Instrumentation .3.3-66 3.3.8.1 Loss of Power (LOP) Instrumentation .3.3-71 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring....................................................................... 3.3-76 3.4 REACTOR COOLANT SYSTEM (RCS) .3.4-1 3.4.1 Recirculation Loops Operating .3.4-1 3.4.2 Jet Pumps ......... 3.4-5 3.4.3 Safety/Relief Valves (S/RVs) .3.4-7 3.4.4 RCS Operational LEAKAGE .3.4-9 3.4.5- RCS Leakage Detection Instrumentation .3.4-12 3.4.6 RCS Specific Activity.3.4-15 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown ............... 3.4-18 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown .. 3.4-21 3.4.9 RCS Pressure and Temperature (P/T) Limits ................... 1 3.4-24 3.4.10 Reactor Steam Dome Pressure .3.4-30 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM .3.5-1 3:5.1 ECCS - Operating .3.5-1 3.5.2 ECCS - Shutdown .3.5-8 3.5.3 RCIC System .3.5-12 BFN-UNIT 2 ii Amendment No. 253

Section Pace No.

3.6 CONTAINMENT SYSTEMS ........................................... 3.6-1 3.6.1.1 Primary Containment ........................................... 3.6-1 3.6.1.2 Primary Containment Air Lock ........................................... 3.6-3 3.6.1.3' Primary Containment Isolation Valves (PCIVs) .................... 3.6-9 3.6.1.4 Drywell Air Temperature ........................................... 3.6-17 3.6.1.5 Reactor Building-to-Suppression Chamber Vacuum Breakers ...................................... 3.6-19 3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers ........... 3.6-22 3.6.2.1 Suppression Pool Average Temperature ............................. 3.6-24 3.6.2.2 Suppression Pool Water Level ...................................... 3.6-29 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling .. 3.6-31 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray ........................................ 3.6-34 3.6.2.5 Residual Heat Removal (RHR) Drywell Spray ..................... 3.6-36 3.6.2.6 Drywell-to-Suppression Chamber Differential Pressure ....... 3.6-38 3.6.3.1 Containment Atmosphere Dilution (CAD) System ................ 3.6-40 3.6.3.2 Primary Containment Oxygen Concentration ....................... 3.6-42 3.6.4.1 Secondary Containment ................ , 3.6-44 3.6.4.2 Secondary Containment Isolation Valves ?SCIVs) ............... 3.6-47 3.6.4.3 Standby Gas Treatment (SGT) System ................................ 3.6-51 3.7 PLANT SYSTEMS ...................................... 3.7-1 3.7.1 Residual Heat Removal Service Water (RHRSW)

Systerr and Ultinate Heat Sinfk (ul i) ........... 3.7-1 3.7.2 Emergency Equipment Cooling Water (EECW)

System and Ultimate Heat Sink (UHS) ........................... 3.7-7 3.7.3 Control Room Emergency Ventilation (CREV) System ........ 3.7-9 3.7.4 Control Room Air Conditioning (AC) System ....................... 3.7-13 3.7.5 Main Turbine Bypass System ......................................... 3.7-17 3.7.6 Spent Fuel Storage Pool Water Level .................................. 3.7-19 3.8 ELECTRICAL POWER SYSTEMS ......................................... 3.8-1 3.8.1 AC Sources - Operating ......................................... 3.8-1 3.8.2 AC Sources - Shutdown ......................................... 3.8-14 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air ........................... 3.8-19 3.8.4 DC Sources - Operating ......................................... 3.8-22 3.8.5 DC Sources - Shutdown ......................................... 3.8-26 3.8.6 Battery Cell Parameters ......................................... 3.8-29 3.8.7 Distribution Systems - Operating ......................................... 3.8-33 3.8.8 Distribution Systems - Shutdown ......................................... 3.8-39 BFN-UNIT 2 iii Amendment No.-26a 263 MAR 1 4 2000

Section Page No.

3.9 REFUELING OPERATIONS ............................................ 3.9-1 3.9.1 Refueling Equipment Interlocks ........................................... 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock................................ 3.9-3 3.9.3 Control Rod Position ............................................ 3.9-5 3.9.4 Control Rod Position Indication ............................................ 3.9-7 3.9.5 Control Rod OPERABILITY - Refueling ..................... t......... 3.9-10 3.9.6 Reactor Pressure Vessel (RPV) Water Level ...................... 3.9-12 3.9.7 Residual Heat Removal (RHR) - High Water Level ............. 3.9-14 3.9.8 Residual Heat Removal (RHR) - Low Water Level .............. 3.9-18 3.10 SPECIAL OPERATIONS ...................................... 3.10-1 3.10.1 Inservice Leak and Hydrostatic Testing Operation .............. 3.10-1 3.10.2 Reactor Mode Switch Interlock Testing ................................ 3.10-4 3.10.3 Single Control Rod Withdrawal - Hot Shutdown ................... 3.10-7 3.10.4 Single Control Rod Withdrawal - Cold Shutdown ....... :........ 3.10-10 3.10.5 Single Control Rod Drive (CRD) Removal - Refueling . 3.10-14 3.10.6 Multiple Control Rod Withdrawal - Refueling .3.10-17 3.10.7 Control Rod Testing - Operating. 3.10-20 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling .3.10-22 4.0 . DESIGN FEATURES. 4.0-1 4.1 Site Location .4.0-1 4.2 Reactor Core. 4.0-1 4.3 Fuel Storage .4.0-2 5.0 ADMINISTRATIVE CONTROLS. 5.0-1 5.1 Responsibility .5.0-1 5.2 Organization .5.0-2 5.3 Unit Staff Qualifications .5.0-6 5.4 Procedures. 5.0-7 5.5 Programs and Manuals . 5.0-8 5.6 Reporting Requirements .5.0-22 5.7 High Radiation Area .5.0-26 BFN-UNIT 2 iv Amendment No. 253

BROWNS FERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS (REQUIREMENTS)

LIST OF FIGURES Figure Paqe No.

Figure 3.1.7-1 ........ 3.1-27 Figure 3.4.1-1 ........ 3.4-4 Figure 3.4.9-1 ........ 3.4-29 Figure 3.7.1-1 ........ 3.7-6 BFN-UNIT 2 v Amendment No. 2537 263 MAR i 4 2Offl

BROWNS FERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS (REQUIREMENTS)

TABLE OF CONTENTS Section Page No.

1.0 USE AND APPLICATION .................. 1.1-1 1.1 Definitions......................1........................................................... 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 ................. 2.0-1 2.2 SL Violations ................. 2.0-1 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 3.1.1 SHUTDOWN MARGIN (SDM) ............................................. 3.1-1 3.1.2 Reactivity Anomalies ............................................... 3.1-5 3.1.3 Control Rod OPERABILITY ............................................... 3.1-7 3.1.4 Control Rod Scram Times ............................................... 3.1-12 3.1.5 Control Rod Scram Accumulators ........................................ 3.1-16 3.1.6 Rod Pattern Control ........................ ....................... 3.1-20 3.1.7 Standby Liquid Control (SLC) System .................................. 3.1-23 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves .....3.1-28 3.2 POWER DISTRIBUTION LIMITS .............................................. 3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) .................................. 3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) .................... 3.2-3 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) ..................... 3.2-5 (continued)

BFN-UNIT 3 Amendment No. 213 SEP 0 3 1998

BROWNS FERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS (REQUIREMENTS)

TABLE OF CONTENTS (continued)

Section Page No.

3.3 INSTRUMENTATION ....................................... 3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation .............. 3.3-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation .................... 3.3-10 3.3.2.1 Control Rod Block Instrumentation ..................................... 3.3-16 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation .................................... 3.3-22 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation . ................

3-24 3.3.3.2 Backup Control System ................................... 3.3-28 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrumentation .. 3.3-30 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation .. 3.3-33 3.3.5.1 Emergency Core Cooling System (ECCS)

Instrumentation .. 3.3-36 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation ...................................... 3.3-49 3.3.6.1 Primary Containment Isolation Instrumentation . ..................

3-53 3.3.6.2 Secondary Containment Isolation Instrumentation ................ 3.3-62 3.3.7.1 Control Room Emergency Ventilation (CREV)

System Instrumentation ........................... 3.3-66 3.3.8.1 Loss of Power (LOP) Instrumentation ........................... 3.3-71 3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring ................................ 3.3-76 3.4 REACTOR COOLANT SYSTEM (RCS) ................................ 3.4-1 3.4.1 Recirculation Loops Operating ................................ 3.4-1 3.4.2 Jet Pumps ................................ 3.4-5 3.4.3 Safety/Relief Valves (S/RVs) ................................ 3.4-7 3.4.4 RCS Operational LEAKAGE ................................ 3.4-9 3.4.5 RCS Leakage Detection Instrumentation ............................. 3.4-12 3.4.6 RCS Specific Activity................................ 3.4-15 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown .. 3.4-18 (continued)

BFN-UNIT 3 ii Amendment No. 213 ISF P e 1,9!e

BROWNS FERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS (REQUIREMENTS)

TABLE OF CONTENTS (continued)

Section Page No.

3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown................................................. 3.4-21 3.4.9 RCS Pressure and Temperature (PIT) Limits .3.4-24 3.4.10 Reactor Steam Dome Pressure .3.4-30 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)

SYSTEM .3.5-1 3.5.1 ECCS - Operating .................................... 3.5-1 3.5.2 ECCS - Shutdown .................................... 3.5-8 3.5.3 RCIC System .................................... 3.5-12 3.6 CONTAINMENT SYSTEMS .............. ...................... 3.6-1 3.6.1.1 Primary Containment .................................... 3.6-1 3.6.1.2 Primary Containment Air Lock.................................... 3.6-3 3.6.1.3 Primary Containment Isolation Valves (PCIVs) .................... 3.6-9 3.6.1.4 Drywell Air Temperature .............. ...................... 3.6-17 3.6.1.5 Reactor Building-to-Suppression Chamber Vacuum Breakers .3.6-19 3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers ........... 3.6-22 3.6.2.1 Suppression Pool Average Temperature ............................. 3.6-24 3.6.2.2 Suppression Pool Water Level ...................................... 3.6-29 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling .3.6-31 3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray .3.6-34 3.6.2.5 Residual Heat Removal (RHR) Drywell Spray ..................... 3.6-36 3.6.2.6 Drywell-to-Suppression Chamber Differential Pressure ....... 3.6-38 3.6.3.1 Containment Atmosphere Dilution (CAD) System ................ 3.6-40 3.6.3.2 Primary Containment Oxygen Concentration ....................... 3.6-42 3.6.4.1 Secondary Containment ........................................ 3.6-44 3.6.4.2 Secondary Containment Isolation Valves (SCIVs) ............... 3.6-47 3.6.4.3 Standby Gas Treatment (SGT) System ................................ 3.6-51 (continued)

BFN-UNIT 3 iii Amendment No. 213 SEP 0 3 1908

BROWNS FERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS (REQUIREMENTS)

TABLE OF CONTENTS (continued)

Section Page No.

3.7 PLANT SYSTEMS .......................................... 3.7-1 3.7.1 Residual Heat Removal Service Water (RHRSW)

System and Ultimate Heat Sink (UHS) . 3.7-1 3.7.2 Emergency Equipment Cooling Water (EECW)

System and Ultimate Heat Sink (UHS) ........................... 3.7-7 3.7.3 Control Room Emergency Ventilation (CREV) System ........ 3.7-9 3.7.4 Control Room Air Conditioning (AC) System ....................... 3.7-13 3.7.5 Main Turbine Bypass System .......................................... 3.7-17 3.7.6 Spent Fuel Storage Pool Water Level .................................. 3.7-19 3.8 ELECTRICAL POWER SYSTEMS . ............................3.8-1 3.8.1 AC Sources - Operating ................................... 3.8-1 3.8.2 AC Sources - Shutdown ................................. 3.8-14 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air . 3.8-19 3.8.4 DC Sources - Operating ................................. 3.8-22 3.8.5 DC Sources - Shutdown ................................. 3.8-26 3.8.6 Battery Cell Parameters ................................. 3.8-29 3.8.7 Distribution Systems - Operating . ...........................3.8-33 3.8.8 Distribution Systems - Shutdown . ...........................3.8-39 3.9 REFUELING OPERATIONS ................................. 3.9-1 3.9.1 Refueling Equipment Interlocks ................................. 3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock ................................ 3.9-3 3.9.3 Control Rod Position ................................. 3.9-5 3.9.4 Control Rod Position Indication ................................. 3.9-7 3.9.5 Control Rod OPERABILITY - Refueling ............................... 3.9-10 3.9.6 Reactor Pressure Vessel (RPV) Water Level ...................... 3.9-12 3.9.7 Residual Heat Removal (RHR) - High Water Level ............. 3.9-14 3.9.8 Residual Heat Removal (RHR) - Low Water Level .............. 3.9-18 (continued)

BFN-UNIT 3 iv Amendment No. 242-213, 223

BROWNS FERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS (REQUIREMENTS)

TABLE OF CONTENTS (continued)

Section Page No.

3.10 SPECIAL OPERATIONS .............................................. 3.10-1 3.10.1 Inservice Leak and Hydrostatic Testing Operation .............. 3.10-1 3.10.2 Reactor Mode Switch Interlock Testing ................................ 3.10-4 3.10.3 Single Control Rod Withdrawal - Hot Shutdown .................. 3.10-7 3.10.4 Single Control Rod Withdrawal - Cold Shutdown ................ 3.10-10 3.10.5 Single Control Rod Drive (CRD) Removal - Refueling ......... 3.10-14 3.10.6 Multiple Control Rod Withdrawal - Refueling ....................... 3.10-17 3.10.7 Control Rod Testing - Operating .......................................... 3.10-20 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling ................... 3.10-22 4.0 DESIGN FEATURES .............................................. 4.0-1 4.1 Site Location .............................................. 4.0-1 4.2 Reactor Core .............................................. 4.0-1 4.3 Fuel Storage .............................................. 4.0-2 5.0 ADMINISTRATIVE CONTROLS .............................................. 5.0-1 5.1 Responsibility .............................................. 5.0-1 5.2 Organization .............................................. 5.0-2 5.3 Unit Staff Qualifications .............................................. 5.0-6 5.4 Procedures .............................................. 5.0-7 5.5 Programs and Manuals .............................................. 5.0-8 5.6 Reporting Requirements .............................................. 5.0-22 5.7 High Radiation Area ................................................ 5.0-26 BFN-UNIT 3 v Amendment No. 213 SEP 0 3 12°8

BROWNS FERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS (REQUIREMENTS)

LIST OF TABLES Table Page No.

Table 1.1-1 ......... 1.1-8 I Table 3.1.4-1 . ......... 3.1-15 Table 3.3.1.1-1 ......... 3.3-7 Table 3.3.1.2-1 ......... 3.3-15 Table 3.3.2.1-1 ......... 3.3-21 Table 3.3.3.1-1 ......... 3.3-27 Table 3.3.5.1-1 ......... 3.3-43 Table 3.3.5.2-1 ......... 3.3-52 Table 3.3.6.1-1 ......... 3.3-59 Table 3.3.6.2-1 ......... 3.3-65 Table 3.3.7.1-1 ......... 3.3-70 Table 3.3.8.1-1 ......... 3.3-75 Table 3.8.6-1 ......... 3.8-32 BFN-UNIT 3 vi Amendment No. 213 SLE ?9'3

BROWNS FERRY NUCLEAR PLANT TECHNICAL SPECIFICATIONS (REQUIREMENTS)

LIST OF FIGURES Figure Page No.

Figure 3.1.7-1 ........ 3.1-27 Figure 3.4.1-1 ........ 3.4-4 Figure 3.4.9-1 ........ 3.4-29 Figure 3.7.1-1 ........ 3.7-6 I BFN-UNIT 3 vii Amendment No. 2 12, 213, 223

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TABLE OF CONTENTS Page 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 Core Reactivity .................................................................................................3.1.2-1 3.1.3 Moderator Temperature Coefficient (MTC) .......................................................3.1.3-1 3.1.4 Rod Group Alignment Limits .............................................................................3.1.4-1 3.1.5 Shutdown Bank Insertion Limits ........................................................................3.1.5-1 3.1.6 Control Bank Insertion Limits ............................................................................3.1.6-1 3.1.7 Rod Position Indication .....................................................................................3.1.7-1 3.1.8 PHYSICS TESTS Exceptions - MODE 2 ..........................................................3.1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (F Q (X,Y,Z)) ........................................................3.2.1-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F H (X,Y)) ......................................3.2.2-1 3.2.3 AXIAL FLUX DIFFERENCE (AFD) ...................................................................3.2.3-1 3.2.4 QUADRANT POWER TILT RATIO (QPTR) ......................................................3.2.4-1 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation .....................................................3.3.1-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS)

Instrumentation ...........................................................................................3.3.2-1 3.3.3 Post Accident Monitoring (PAM) Instrumentation ..............................................3.3.3-1 3.3.4 Remote Shutdown Monitoring Instrumentation .................................................3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ...........................................................................................3.3.5-1 3.3.6 Containment Ventilation Isolation Instrumentation ............................................3.3.6-1 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation ...........................................................................................3.3.7-1 3.3.8 Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation ...........................................................................................3.3.8-1 3.3.9 Boron Dilution Monitoring Instrumentation (BDMI) ............................................3.3.9-1 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 SEQUOYAH - UNIT 2 i Amendment 327

TABLE OF CONTENTS Page 3.4 REACTOR COOLANT SYSTEM (RCS) (continued) 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 Power Operated Relief Valves (PORVs) ...................................... 3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP) System ........................... 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 ......................................................................................3.4.16-1 3.4.17 Steam Generator (SG) Tube Integrity .............................................................3.4.17-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators ....................................................................................................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 Refueling Water Storage Tank (RWST) ............................................................3.5.4-1 3.5.5 Seal Injection Flow............................................................................................3.5.5-1 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 System ..............................................................................3.6.6-1 3.6.7 Shield Building ..................................................................................................3.6.7-1 3.6.8 Hydrogen Mitigation System (HMS) ..................................................................3.6.8-1 3.6.9 Vacuum Relief Valves .......................................................................................3.6.9-1 3.6.10 Emergency Gas Treatment System (EGTS)

Air Cleanup Subsystem.............................................................................3.6.10-1 3.6.11 Air Return System (ARS) ................................................................................3.6.11-1 3.6.12 Ice Bed ...........................................................................................................3.6.12-1 3.6.13 Ice Condenser Doors ......................................................................................3.6.13-1 3.6.14 Divider Barrier Integrity ...................................................................................3.6.14-1 3.6.15 Containment Recirculation Drains...................................................................3.6.15-1 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 Isolation Valves (MFIVs), Main Feedwater Regulating Valves (MFRVs) and MFRV Bypass Valves ..............................3.7.3-1 3.7.4 Atmospheric Relief Valves (ARVs)....................................................................3.7.4-1 3.7.5 Auxiliary Feedwater (AFW) System ..................................................................3.7.5-1 3.7.6 Condensate Storage Tank (CST)......................................................................3.7.6-1 3.7.7 Component Cooling Water System (CCS) ........................................................3.7.7-1 3.7.8 Essential Raw Cooling Water (ERCW) System.................................................3.7.8-1 SEQUOYAH - UNIT 2 ii Amendment 327

TABLE OF CONTENTS Page 3.7 PLANT SYSTEMS (continued) 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 Air-Conditioning System (CRACS) ........................................... 3.7.11-1 3.7.12 Auxiliary Building Gas Treatment System (ABGTS) ........................................ 3.7.12-1 3.7.13 Spent Fuel Pool Water Level ..........................................................................3.7.13-1 3.7.14 Spent Fuel Pool Boron Concentration .............................................................3.7.14-1 3.7.15 Spent Fuel Pool Storage .................................................................................3.7.15-1 3.7.16 Secondary Specific Activity .............................................................................3.7.16-1 3.7.17 Cask Pit Pool Boron Concentration .................................................................3.7.17-1 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 Unborated Water Source Isolation Valves ........................................................3.9.2-1 3.9.3 Nuclear Instrumentation ....................................................................................3.9.3-1 3.9.4 Containment Penetrations ................................................................................3.9.4-1 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level.................................................................................................3.9.5-1 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level.................................................................................................3.9.6-1 3.9.7 Refueling Cavity Water Level ............................................................................3.9.7-1 3.9.8 Decay Time ......................................................................................................3.9.8-1 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 5.6 Reporting Requirements ......................................................................................5.6-1 5.7 High Radiation Area.............................................................................................5.7-1 SEQUOYAH - UNIT 2 iii Amendment 327

FH(X,Y) 3.2.2

( )

3.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor F H (X,Y)

LCO 3.2.2 FH(X,Y) shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ A.1 Reduce allowable 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Required Actions A.3 THERMAL POWER from and A.5 must be RTP by RRH% multiplied completed whenever times the F H min margin.

Condition A is entered.

AND F H min margin < 0. A.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Neutron Flux - High trip setpoints by RRH%

multiplied times the F H min margin.

AND A.3 Perform SR 3.2.2.1. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.4 Reduce Overtemperature 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months T trip setpoint by TRH multiplied times the F H min margin.

AND SEQUOYAH - UNIT 2 3.2.2-1 Amendment 327

TABLE OF CONTENTS TABLE OF CONTENTS i LIST OF TABLES * * *

V LIST OF FIGURES vi LIST OF ACRONYMS . . . . . vii LIST OF EFFECTIVE PAGES Viii 1.0 USE AND APPLICATION l.P*l 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. I SLs . . . . . . 2.0-1 2.2 SL Violations . . .. 2.0-1 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

3. I. I SHUTDOWN MARGIN (SDM)-Tavg > 200°F . . 3.1-1 3 .1. 2 SHUTDOWN MARGIN (SOM) -T. g ~ 200°F . .

11 3.1-2 3 .1.3 Core Reactivity . . . . . . . 3 *. 1-3 3 .1.4 Moderator Temperature Coefficient (MTC) 3.1-5 3 .1. 5 Rod Group Alignment Limits .. 3.1-8 3 .1.6 Shutdown Bank Insertion Limits 3.1-12 3 .1. 7 Control Bank Insertion Limits 3.1-14 3 .1.8 Rod Position Indication . . . 3.1-17 3.1. 9 PHYSICS TESTS Exceptions-MOOE I 3.1-20 3 .1.10 PHYSICS TESTS Exceptions -MOOE 2 3.1-23 3.2 POWER DISTRIBUTION LIMITS . . . . . . . 3.2-1 3.2.1 Heat Flux Hot Channel Factor {Fa(Z)) 3.2-1 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor 3.2-6 3.2.3 AXIAL FLUX DIFFERENCE (AFO) ... . 3.2-9 3.2.4 QUADRANT POWER TILT RATIO (QPTR) . . . . 3.2-10 3.3 INSTRUMENiATION . . . . . . . . . . . . . . . 3.3-1 3.3.l Reactor Trip System (RTS} Instrumentation . . . 3.3-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS}

Instrumentation . . ....... . 3.3-24 3.3.3 Post Accident Monitoring (PAM} Instrumentation . . 3.3-41 3.3.4 Remote Shutdown System. . . . . . . . . . . . . . 3.3-46 3.3.5 Loss of Power (LOP) Diesel Generator (OG) Start Instrumentation . . .......... . 3.3-49 3.3.6 Containment Vent Isolation Instrumentation .. . 3.3-52 3.3.7 Control Room Emergency Ventilation System (CREVS)

Actuation Instrumentation . . ..... . 3.3-57 3.3.8 Auxiliary Building Gas Treatment System (ABGTS}

Actuation Instrumentation . . . . .. 3.3-61 (continued)

Watts Bar-Unit I

TABLE OF CONTENTS (continued) 3.4 REACTOR COOLANT SYSTEM (RCS) ..................................................... 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ................................................. 3.4-1 3.4.2 RCS Minimum Temperature for Criticality ..................................... 3.4-3 3.4.3 RCS Pressure and Temperature (P/T) Limits ................................ 3.4-5 3.4.4 RCS Loops MODES 1and 2 ..................................................... 3.4-7 3.4.5 RCS Loops MODE 3 .................................................................. 3.4-8 3.4.6 RCS Loops MODE 4 .................................................................. 3.4-11 3.4.7 RCS Loops MODE 5,Loops Filled ............................................ 3.4-14 3.4.8 RCS Loops MODE 5, LoopsNot Filled...................................... 3.4-16 3.4.9 Pressurizer ..................................................................................... 3.4-18 3.4.10 Pressurizer Safety Valves .............................................................. 3.4-20 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) .................... 3.4-22 3.4.12 Cold Overpressure Mitigation System (COMS) ............................. 3.4-25 3.4.13 RCS Operational LEAKAGE .......................................................... 3.4-30 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage................................ 3.4-32 3.4.15 RCS Leakage Detection Instrumentation....................................... 3.4-36 3.4.16 RCS Specific Activity........................................................................ 3.4-39 3.4.17 Steam Generator (SG) Tube Integrity .............................................. 3.4-43 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) ................................ 3.5-1 3.5.1 Accumulators ................................................................................. 3.5-1 3.5.2 ECCS Operating ........................................................................ 3.5-4 3.5.3 ECCS Shutdown ........................................................................ 3.5-7 3.5.4 Refueling Water Storage Tank (RWST)......................................... 3.5-9 3.5.5 Seal Injection Flow ......................................................................... 3.5-11 3.6 CONTAINMENT SYSTEMS........................................................................ 3.6-1 3.6.1 Containment................................................................................... 3.6-1 3.6.2 Containment Air Locks ................................................................... 3.6-3 3.6.3 Containment Isolation Valves ........................................................ 3.6-8 3.6.4 Containment Pressure ................................................................... 3.6-15 3.6.5 Containment Air Temperature........................................................ 3.6-16 3.6.6 Containment Spray System ........................................................... 3.6-18 3.6.7 Deleted........................................................................................... 3.6-20 3.6.8 Hydrogen Mitigation System (HMS)............................................... 3.6-22 3.6.9 Emergency Gas Treatment System (EGTS).................................. 3.6-24 3.6.10 Air Return System (ARS) ............................................................... 3.6-26 3.6.11 Ice Bed ........................................................................................... 3.6-28 3.6.12 Ice Condenser Doors ..................................................................... 3.6-31 3.6.13 Divider Barrier Integrity .................................................................. 3.6-35 3.6.14 Containment Recirculation Drains ................................................. 3.6-38 3.6.15 Shield Building ............................................................................... 3.6-40 3.7 PLANT SYSTEMS ............................................................................................. 3.7-1 3.7.1 Main Steam Safety Valves (MSSVs) ............................................. 3.7-1 3.7.2 Main Steam Isolation Valves (MSIVs)............................................ 3.7-5 (continued)

Watts Bar-Unit 1 ii Amendment

TABLE OF CONTENTS (continued) 3.7 PLANT SYSTEMS (continued) 3.7.3 Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves ............................................... 3.7-7 3.7.4 Atmospheric Dump Valves (ADVs) ................................................ 3.7-9 3.7.5 Auxiliary Feedwater (AFW) System ............................................... 3.7-11 3.7.6 Condensate Storage Tank (CST)................................................... 3.7-15 3.7.7 Component Cooling System (CCS) ............................................... 3.7-17 3.7.8 Essential Raw Cooling Water (ERCW) System ............................. 3.7-19 3.7.9 Ultimate Heat Sink (UHS) .............................................................. 3.7-21 3.7.10 Control Room Emergency Ventilation System (CREVS) ............... 3.7-22 3.7.11 Control Room Emergency Air Temperature Control System (CREATCS) ................................................................ 3.7-25 3.7.12 Auxiliary Building Gas Treatment System (ABGTS) ...................... 3.7-27 3.7.13 Fuel Storage Pool Water Level ...................................................... 3.7-29 3.7.14 Secondary Specific Activity ............................................................ 3.7-30 3.7.15 Spent Fuel Assembly Storage........................................................ 3.7-31 3.7.16 Component Cooling System (CCS) - Shutdown ........................ 3.7-33 3.7.17 Essential Raw Cooling Water (ERCW) System - Shutdown..........3.7-36 3.8 ELECTRICAL POWER SYSTEMS ............................................................. 3.8-1 3.8.1 AC Sources Operating ............................................................... 3.8-1 3.8.2 AC Sources Shutdown ............................................................... 3.8-18 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air...................................... 3.8-21 3.8.4 DC Sources Operating ............................................................... 3.8-24 3.8.5 DC Sources Shutdown............................................................... 3.8-30 3.8.6 Battery Parameters .......................................................... 3.8-33 3.8.7 Inverters Operating .................................................................... 3.8-37 3.8.8 Inverters Shutdown .................................................................... 3.8-39 3.8.9 Distribution Systems Operating.................................................. 3.8-41 3.8.10 Distribution Systems Shutdown ................................................. 3.8-43 3.9 REFUELING OPERATIONS ....................................................................... 3.9-1 3.9.1 Boron Concentration ...................................................................... 3.9-1 3.9.2 Unborated Water Source Isolation Valves ..................................... 3.9-2 3.9.3 Nuclear Instrumentation ................................................................. 3.9-4 3.9.4 Deleted ........................................................................................... 3.9-6 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation HighWater Level ............................................... 3.9-8 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation LowWater Level................................................ 3.9-10 3.9.7 Refueling Cavity Water Level......................................................... 3.9-12 3.9.8 Deleted ........................................................................................... 3.9-14 3.9.9 Spent Fuel Pool Boron Concentration............................................ 3.9-16 3.9.10 Decay Time .................................................................................... 3.9-17 4.0 DESIGN FEATURES .................................................................................. 4.0-1 4.1 Site ................................................................................................. 4.0-1 4.2 Reactor Core .................................................................................. 4.0-1 4.3 Fuel Storage................................................................................... 4.0-2 (continued)

Watts Bar-Unit 1 iii Amendment 6,

TABLE OF CONTENTS (continued) 5.0 ADMINISTRATIVE CONTROLS................................................................. 5.0-1 5.1 Responsibility .............................................................................................. 5.0-1 5.2 Organization................................................................................... 5.0-2 5.3 Unit Staff Qualifications.................................................................. 5.0-5 5.4 Training .......................................................................................... 5.0-6 5.5 Reviews and Audits........................................................................ 5.0-7 5.6 Technical Specifications (TS) Bases Control Program .................. 5.0-8 5.7 Procedures, Programs, and Manuals ............................................ 5.0-9 5.8 Safety Function Determination Program (SFDP)........................... 5.0-26 5.9 Reporting Requirements ................................................................ 5.0-27 5.10 Record Retention ........................................................................... 5.0-33 5.11 High Radiation Area ....................................................................... 5.0-34 Watts Bar-Unit 1 iv Amendment

LIST OF TABLES Table No. Title Page .................................................................................................... Page 1.1-1 MODES ....................................................................................................... 1.1-7 3.3.1-1 Reactor Trip System Instrumentation.......................................................... 3.3-15 3.3.2-1 Engineered Safety Features Actuation System Instrumentation ..................................................................................... 3.3-34 3.3.3-1 Post-Accident Monitoring Instrumentation .................................................. 3.3-44 3.3.5-1 LOP DG Start Instrumentation.................................................................... 3.3-51 3.3.6-1 Containment Vent Isolation Instrumentation ............................................... 3.3-5

3.3.7-1 CREVS Actuation Instrumentation .............................................................. 3.3-60 3.3.8-1 ABGTS Actuation Instrumentation .............................................................. 3.3-63 3.7.1-1 OPERABLE Main Steam Safety Valves versus Maximum Allowable Power................................................................... 3.7-3 3.7.1-2 Main Steam Safety Valve Lift Settings ........................................................ 3.7-4 3.8.1-1 Diesel Generator Test Schedule ................................................................. 3.8-17 Watts Bar-Unit 1 v Amendment 6,

LIST OF FIGURES Figure No. Title Page .................................................................................................... Page 2.1.1-1 Reactor Core Safety Limits ......................................................................... 2.0-2 3.4.16-1 Deleted ........................................................................................................ 3.4-42 4.3-1 Spent Fuel Pool Plan .................................................................................. 4.0-7 4.3-2 New Fuel Storage Rack Loading Pattern.................................................... 4.0-8 Watts Bar-Unit 1 vi Amendment

LI ST OF ACRONYMS Acronym Title ABGTS Auxiliary Building Gas Treatment System ACRP Auxiliary Control Room Panel ASME American Society of Mechanical Engineers AFD Axial Flux Difference AFW Auxiliary Feedwater System ARO All Rods Out ARFS Air Return Fan System ARV Atmospheric Relief Valve BOC Beginning of Cycle CAOC Constant Axial Offset Control ccs Component Cooling Water System CFR Code of Federal Regulations COLR Core Operating Limits Report CREVS Control Room Emergency Ventilation System css Containment Spray System CST Condensate Storage Tank DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPD Effective Full-Power Days EGTS Emergency Gas Treatment System EOC End of Cycle ERCW Essential Raw Cooling Water ESF Engineered Safety Feature ESFAS Engineered Safety Features Actuation System HEPA High Efficiency Particulate Air HVAC Heating, Ventilating, and Air-Conditioning LCO limiting Condition For Operation MFIV Main Feedwater Isolation Valve MFRV Main Feedwater Regulation Valve MSIV Main Steam Line Isolation Valve MSSV Main Steam Safety Valve MTC Moderator Temperature Coefficient NMS Neutron Monitoring System ODCM Offsite Dose Calculation Manual PCP Process Control Program PIV Pressure Isolation Valve PORV Power-Operated Relief Valve PTLR Pressure and Temperature limits Report QPTR Quadrant Power Tilt Ratio RAOC Relaxed Axial Offset Control RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RTP Rated Thermal Power RTS Reactor Trip System RWST Refueling Water Storage Tank SG Steam Generator SI Safety Injection SL Safety limit SR Surveillance Requirement UHS Ultimate Heat Sink Watts Bar-Unit 1 vii

3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs}

LCO 3.7.1 Five MSSVs per steam generator shall ce OPERABLE.

APPLICABILITY: MODES 1. 2. and 3.

ACTIONS

NOTE--------

Separate Condition entry is allowed for each MSSV.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more steam A.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months generators with one < 58 % RTP.

MSSV inoperable.

B. One or more steam B.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months generators with two or less than or equal to more MSSVs incperab'e. -he Maximum Allowable

RTP soeci`ied in Table 3.7.1-i for the number of OPERABLE MSSVs.

AND

NOTE---------

Only requireC in MODE 1 B.2 Reduce the Power Range Neutron Flux - High reactor trip setpoint 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to less than or equal to the Maximum Allowable % RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND OR C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or more steam generators with > 4 MSSVs inoperable. _____

Watts Bar-Unit 1 3.7-1 Amendment 19. 31 r -1I - . '

TABLE OF CONTENTS TABLE OF CONTENTS . i LIST OF TABLES vi LIST OF FIGURES . vii LIST OF ACRONYMS viii LIST OF EFFECTIVE PAGES .. x 1.0 USE AND APPLICATION ... 1.1-1 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 . 2.0-1 2.2 SL Violations . 2.0-1 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 3.1.1 SHUTDOWN MARGIN (SDM) - Tavg > 200°F ............... 3.1-1 3.1.2 SHUTDOWN MARGIN (SDM) - Tavg 200°F . 3.1-2 3.1.3 Core Reactivity . 3.1-3 3.1.4 Moderator Temperature Coefficient (MTC) . 3.1-5 3.1.5 Rod Group Alignment Limits .. 3.1-7 3.1.6 Shutdown Bank Insertion Limits 3.1-10 3.1.7 Control Bank Insertion Limits . 3.1-12 3.1.8 Rod Position Indication ... 3.1-15 3.1.9 PHYSICS TESTS Exceptions - MODE 1 .... 3.1-18 3.1.10 PHYSICS TESTS Exceptions - MODE 2 .... 3.1-20 3.2 POWER DISTRIBUTION LIMITS .. 3.2-1 3.2.1 Heat Flux Hot Channel Factor (FQ(Z)) .. 3.2-1 N

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FH ) ........ 3.2-6 3.2.3 AXIAL FLUX DIFFERENCE (AFD) ...... 3.2-8 3.2.4 QUADRANT POWER TILT RATIO (QPTR) 3.2-9 i (continued)

Watts Bar - Unit 2

TABLE OF CONTENTS (continued) 3.3 INSTRUMENTATION .. 3.3-1 3.3.1 Reactor Trip System (RTS) Instrumentation ... 3.3-1 3.3.2 Engineered Safety Feature Actuation System (ESFAS)

Instrumentation .. 3.3-24 3.3.3 Post Accident Monitoring (PAM) Instrumentation .. 3.3-42 3.3.4 Remote Shutdown System . 3.3-48 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation . 3.3-51 3.3.6 Containment Vent Isolation Instrumentation ............. 3.3-54 3.3.7 Control Room Emergency Ventilation System (CREVS)

Actuation Instrumentation .............. 3.3-59 3.3.8 Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation . 3.3-63 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits .. 3.4-1 3.4.2 RCS Minimum Temperature for Criticality ... 3.4-3 3.4.3 RCS Pressure and Temperature (P/T) Limits . 3.4-4 3.4.4 RCS Loops - MODES 1 and 2 ... 3.4-6 3.4.5 RCS Loops - MODE 3 . 3.4-7 3.4.6 RCS Loops - MODE 4 . 3.4-9 3.4.7 RCS Loops - MODE 5, Loops Filled .... 3.4-12 3.4.8 RCS Loops - MODE 5, Loops Not Filled . 3.4-14 3.4.9 Pressurizer 3.4-16 3.4.10 Pressurizer Safety Valves .. 3.4-18 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) 3.4-20 3.4.12 Cold Overpressure Mitigation System (COMS) .. 3.4-23 3.4.13 RCS Operational LEAKAGE .. 3.4-28 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage . 3.4-30 3.4.15 RCS Leakage Detection Instrumentation 3.4-33 3.4.16 RCS Specific Activity ............. 3.4-35 3.4.17 Steam Generator (SG) Tube Integrity .. 3.4-38 ii (continued)

Watts Bar - Unit 2

TABLE OF CONTENTS (continued) 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) .............. 3.5-1 3.5.1 Accumulators 3.5-1 3.5.2 ECCS - Operating ... 3.5-3 3.5.3 ECCS - Shutdown ... 3.5-6 3.5.4 Refueling Water Storage Tank (RWST) .. 3.5-8 3.5.5 Seal Injection Flow .. 3.5-10 3.6 CONTAINMENT SYSTEMS 3.6-1 3.6.1 Containment ..... 3.6-1 3.6.2 Containment Air Locks ... 3.6-2 3.6.3 Containment Isolation Valves 3.6-7 3.6.4 Containment Pressure .... 3.6-14 3.6.5 Containment Air Temperature ... 3.6-15 3.6.6 Containment Spray System ... 3.6-17 3.6.7 RESERVED FOR FUTURE ADDITION ... 3.6-19 3.6.8 Hydrogen Mitigation System (HMS) . 3.6-20 3.6.9 Emergency Gas Treatment System (EGTS) .. 3.6-22 3.6.10 Air Return System (ARS) 3.6-24 3.6.11 Ice Bed .. 3.6-25 3.6.12 Ice Condenser Doors .. 3.6-28 3.6.13 Divider Barrier Integrity ... 3.6-31 3.6.14 Containment Recirculation Drains 3.6-33 3.6.15 Shield Building . 3.6-35 3.7 PLANT SYSTEMS .............. 3.7-1 3.7.1 Main Steam Safety Valves (MSSVs) ........... 3.7-1 3.7.2 Main Steam Isolation Valves (MSIVs) .. 3.7-4 3.7.3 Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves ........... 3.7-6 3.7.4 Atmospheric Dump Valves (ADVs) ... 3.7-8 3.7.5 Auxiliary Feedwater (AFW) System .. 3.7-10 3.7.6 Condensate Storage Tank (CST) . 3.7-14 3.7.7 Component Cooling System (CCS) .. 3.7-16 3.7.8 Essential Raw Cooling Water (ERCW) System . 3.7-18 3.7.9 Ultimate Heat Sink (UHS) .. 3.7-20 iii (continued)

Watts Bar - Unit 2

TABLE OF CONTENTS 3.7 PLANT SYSTEMS (continued) 3.7.10 Control Room Emergency Ventilation System (CREVS) .. 3.7-21 3.7.11 Control Room Emergency Air Temperature Control System (CREATCS) 3.7-24 3.7.12 Auxiliary Building Gas Treatment System (ABGTS) .. 3.7-26 3.7.13 Fuel Storage Pool Water Level .. 3.7-28 3.7.14 Secondary Specific Activity ........... 3.7-29 3.7.15 Spent Fuel Assembly Storage ... 3.7-30 3.7.16 Component Cooling System (CCS) - Shutdown ........... 3.7-31 3.7.17 Essential Raw Cooling Water (ERCW) System (ERCW) - Shutdown ..... 3.7-34 3.8 ELECTRICAL POWER SYSTEMS ............... 3.8-1 3.8.1 AC Sources - Operating ....................... 3.8-1 3.8.2 AC Sources - Shutdown ....................... 3.8-15 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air .............. 3.8-18 3.8.4 DC Sources - Operating ................ 3.8-21 3.8.5 DC Sources - Shutdown ............... 3.8-26 3.8.6 Battery Parameters .................... 3.8-29 3.8.7 Inverters - Operating ............................ 3.8-33 3.8.8 Inverters - Shutdown ............................ 3.8-35 3.8.9 Distribution Systems - Operating ............. 3.8-37 3.8.10 Distribution Systems - Shutdown .. 3.8-39 3.9 REFUELING OPERATIONS .. 3.9-1 3.9.1 Boron Concentration ... 3.9-1 3.9.2 Unborated Water Source Isolation Valves .. 3.9-2 3.9.3 Nuclear Instrumentation ..... 3.9-3 3.9.4 RESERVED FOR FUTURE ADDITION ... 3.9-5 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation -

High Water Level .. 3.9-6 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation -

Low Water Level 3.9-8 3.9.7 Refueling Cavity Water Level 3.9-10 3.9.8 RESERVED FOR FUTURE ADDITION ........ 3.9-11 3.9.9 Spent Fuel Pool Boron Concentration .. 3.9-12 3.9.10 Decay Time ............................................................................................ 3.9-13 iv (continued)

Watts Bar - Unit 2

TABLE OF CONTENTS (continued) 4.0 DESIGN FEATURES 4.0-1 4.1 Site . 4.0-1 4.2 Reactor Core 4.0-1 4.3 Fuel Storage . 4.0-2 5.0 ADMINISTRATIVE CONTROLS .............. 5.0-1 5.1 Responsibility ..................... 5.0-1 5.2 Organization .................................... 5.0-2 5.3 Unit Staff Qualifications .................... 5.0-4 5.4 Training ............................................ 5.0-5 5.5 Reviews and Audits .......................... 5.0-6 5.6 Technical Specifications (TS) Bases Control Program ....................... 5.0-7 5.7 Procedures, Programs, and Manuals ..................... 5.0-8 5.8 Safety Function Determination Program (SFDP) ..................... 5.0-28 5.9 Reporting Requirements . 5.0-29 5.10 Record Retention ............................. 5.0-36 5.11 High Radiation Area ......................... 5.0-37 v (continued)

Watts Bar - Unit 2

LIST OF TABLES TABLE No. TITLE PAGE 1.1-1 MODES ................. 1.1-8 3.3.1-1 Reactor Trip System Instrumentation .... 3.3-15 3.3.2-1 Engineered Safety Feature Actuation System Instrumentation ..... 3.3-34 3.3.3-1 Post Accident Monitoring Instrumentation .... 3.3-45 3.3.5-1 LOP DG Start Instrumentation ....... 3.3-53 3.3.6-1 Containment Vent Isolation Instrumentation .... 3.3-58 3.3.7-1 CREVS Actuation Instrumentation ..... 3.3-62 3.3.8-1 ABGTS Actuation Instrumentation ..... 3.3-65 3.7.1-1 OPERABLE Main Steam Safety Valves Versus Maximum Allowable Power 3.7-3 3.7.1-2 Main Steam Safety Valve Lift Settings .. 3.7-3 3.8.1-1 Diesel Generator Test Schedule ........ 3.8-14 vi (continued)

Watts Bar - Unit 2 Amendment

LIST OF FIGURES FIGURE No. TITLE PAGE 2.1.1-1 Reactor Core Safety Limits . 2.0-2 4.3-1 Spent Fuel Storage Racks ......... 4.0-6 4.3-2 New Fuel Storage Rack Loading Pattern . 4.0-7 vii (continued)

Watts Bar - Unit 2 Amendment

LIST OF ACRONYMS (Page 1 of 2)

ACRONYM TITLE ABGTS Auxiliary Building Gas Treatment System ACRP Auxiliary Control Room Panel AFD Axial Flux Difference AFW Auxiliary Feedwater System ARFS Air Return Fan System ARO All Rods Out ARV Atmospheric Relief Valve ASME American Society of Mechanical Engineers BOC Beginning of Cycle CAOC Constant Axial Offset Control CCS Component Cooling Water System CFR Code of Federal Regulations COLR Core Operating Limits Report CREVS Control Room Emergency Ventilation System CSS Containment Spray System CST Condensate Storage Tank DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPD Effective Full-Power Days EGTS Emergency Gas Treatment System EOC End of Cycle ERCW Essential Raw Cooling Water ESF Engineered Safety Feature ESFAS Engineered Safety Features Actuation System HEPA High Efficiency Particulate Air HVAC Heating, Ventilating, and Air-Conditioning viii (continued)

Watts Bar - Unit 2

LIST OF ACRONYMS (Page 2 of 2)

ACRONYM TITLE LCO Limiting Condition For Operation MFIV Main Feedwater Isolation Valve MFRV Main Feedwater Regulation Valve MSIV Main Steam Line Isolation Valve MSSV Main Steam Safety Valve MTC Moderator Temperature Coefficient NMS Neutron Monitoring System ODCM Offsite Dose Calculation Manual PCP Process Control Program PDMS Power Distribution Monitoring System PIV Pressure Isolation Valve PORV Power-Operated Relief Valve PTLR Pressure and Temperature Limits Report QPTR Quadrant Power Tilt Ratio RAOC Relaxed Axial Offset Control RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RTP Rated Thermal Power RTS Reactor Trip System RWST Refueling Water Storage Tank SG Steam Generator SI Safety Injection SL Safety Limit SR Surveillance Requirement UHS Ultimate Heat Sink ix (continued)

Watts Bar - Unit 2

ATTACHMENT 2 Re-typed Technical Specification Pages for WBN Unit 1 and SQN Units 1 and 2 CNL-20-041

MSSVs 3.7.1 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs)

LCO 3.7.1 Five MSSVs per steam generator shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS

NOTE------------------------------------------------------------

Separate Condition entry is allowed for each MSSV.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more steam A.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months generators with one MSSV 58 % RTP.

inoperable.

B. One or more steam generators B.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months with two or more MSSVs less than or equal to the inoperable. Maximum Allowable % RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

AND

NOTE-----------------

Only required in MODE 1 B.2 Reduce the Power Range 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Neutron Flux - High reactor trip setpoint to less than or equal to the Maximum Allowable %

RTP specified in Table 3.7.1-1 for the number of OPERABLE MSSVs.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND OR C.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months One or more steam generators with 4 MSSVs inoperable.

Watts Bar-Unit 1 3.7-1 Amendment 19, 31, ___

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FH(X,Y) 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FH(X,Y))

LCO 3.2.2 FH(X,Y) shall be within the limits specified in the COLR.

APPLICABILITY: MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ A.1 Reduce allowable 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Required Actions A.3 THERMAL POWER from and A.5 must be RTP by RRH% multiplied completed whenever times the F H min margin.

Condition A is entered.

AND F H min margin < 0. A.2 Reduce Power Range 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Neutron Flux - High trip setpoints by RRH%

multiplied times the F H min margin.

AND A.3 Perform SR 3.2.2.1. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AND A.4 Reduce Overtemperature 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months T trip setpoint by TRH multiplied times the F H min margin.

AND SEQUOYAH - UNIT 2 3.2.2-1 Amendment 327, ___

v t e Letter Sequence Request
EPID:L-2020-LLA-0185, License Amendment Request to Remove Licensee Control (BFN TS-527, SQN-TS-20-08, and WBN-TS-20-016) (Approved, Closed)
Initiation Request Acceptance... Administration Questions Answers Results Approval Other:
MONTHYEAR2020-08-14 [Table View]

CNL-20-041, License Amendment Request to Remove Licensee Control (BFN TS-527, SQN-TS-20-08, and WBN-TS-20-016)

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Subject:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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CNL-20-041, License Amendment Request to Remove Licensee Control (BFN TS-527, SQN-TS-20-08, and WBN-TS-20-016)

Text

Subject:

Subject:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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