ML20016A278
| ML20016A278 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 06/22/2020 |
| From: | Kimberly Green Plant Licensing Branch II |
| To: | Jim Barstow Tennessee Valley Authority |
| Green K | |
| References | |
| EPID L-2019-LLA-0120 | |
| Download: ML20016A278 (46) | |
Text
June 22, 2020 Mr. James Barstow Vice President, Nuclear Regulatory Affairs and Support Services Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 135 AND 39 REGARDING MISCELLANEOUS ADMINISTRATIVE CHANGES TO THE TECHNICAL SPECIFICATIONS (EPID L-2019-LLA-0120)
Dear Mr. Barstow:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 135 to Facility Operating License No. NPF-90 and Amendment No. 39 to Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant (WBN), Units 1 and 2, respectively. These amendments are in response to your application dated June 7, 2019, as supplemented by letters dated October 9, 2019, and April 14, 2020. The amendments revise the WBN Technical Specifications by making several administrative changes.
A copy of the related safety evaluation is also enclosed. Notice of issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Kimberly J. Green, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-390 and 50-391
Enclosures:
- 1. Amendment No. 135 to NPF-90
- 2. Amendment No. 39 to NPF-96
- 3. Safety Evaluation cc: Listserv TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-390 WATTS BAR NUCLEAR PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 135 License No. NPF-90 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated June 7, 2019, as supplemented by letters dated October 9, 2019, and April 14, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-90 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 135 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: June 22, 2020 Undine S.
Shoop Digitally signed by Undine S.
Shoop Date: 2020.06.22 14:24:17
-04'00'
ATTACHMENT TO AMENDMENT NO. 135 WATTS BAR NUCLEAR PLANT, UNIT 1 FACILITY OPERATING LICENSE NO. NPF-90 DOCKET NO. 50-390 Replace page 3 of Facility Operating License No. NPF-90 with the attached revised page 3.
The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Pages Insert Pages ii ii iii iii iv iv v
v vi vi 3.0-4 3.0-4 3.0-5 3.0-6 3.0-7 3.5-5 3.5-5 3.6-18 3.6-18 3.6-22 3.6-22 3.6-23 3.6-23 3.6-33 3.6-33 3.6-34 3.6-34 3.7-20 3.7-20 4.0-1 4.0-1 4.0-5 4.0-5 4.0-6 4.0-6 5.0-24 5.0-24
Facility License No. NPF-90 Amendment No. 135
(4)
TVA, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required, any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis, instrument calibration, or other activity associated with radioactive apparatus or components; and (5)
TVA, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.
(1)
Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 3459 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 135 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Safety Parameter Display System (SPDS) (Section 18.2 of SER Supplements 5 and 15)
Prior to startup following the first refueling outage, TVA shall accomplish the necessary activities, provide acceptable responses, and implement all proposed corrective actions related to having the Watts Bar Unit 1 SPDS operational.
(4)
Vehicle Bomb Control Program (Section 13.6.9 of SSER 20)
During the period of the exemption granted in paragraph 2.D.(3) of this license, in implementing the power ascension phase of the approved initial test program, TVA shall not exceed 50% power until the requirements of 10 CFR 73.55(c)(7) and (8) are fully implemented. TVA shall submit a letter under oath or affirmation when the requirements of 73.55(c)(7) and (8) have been fully implemented.
TABLE OF CONTENTS (continued)
(continued)
Watts Bar-Unit 1 ii Amendment 135 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 1 and 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, Loops Not 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
TABLE OF CONTENTS (continued)
(continued)
Watts Bar-Unit 1 iii Amendment 6, 135 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 High Water Level............................................... 3.9-8 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation Low Water 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
TABLE OF CONTENTS (continued)
Watts Bar-Unit 1 iv Amendment 135 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
LIST OF TABLES Watts Bar-Unit 1 v
Amendment 6, 135 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............................................... 3.3-55 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
LIST OF FIGURES Watts Bar-Unit 1 vi Amendment 135 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
SR Applicability 3.0 (continued)
Watts Bar-Unit 1 3.0-4 Amendment 42, 114, 121, 135 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.
Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a "once per..." basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
SR Applicability 3.0 Watts Bar-Unit 1 3.0-5 Amendment 55, 135 3.0 SR APPLICABILITY SR 3.0.3 When the Surveillance is performed within the delay period and the Surveillance (continued) is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.
This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
SURVEILLANCE REQUIREMENTS SR 3.5.2.1 SR 3.5.2.2 SR 3.5.2.3 SR 3.5.2.4 SR 3.5.2.5 Watts Bar-Unit 1 SURVEILLANCE Verify the following valves are in the listed position with power to the valve operator removed.
Number FCV-63-1 FCV-63-22 Position Open Open Function RHR Supply SIS Discharge Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
Verify ECCS piping is full of water.
Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.
Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
3.5-5 ECCS - Operating 3.5.2 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the lnservice Testing Program In accordance with the Surveillance Frequency Control Program (continued)
Amendment 43, 132, 135
Containment Spray System 3.6.6 Watts Bar-Unit 1 3.6-18 Amendment 93, 135 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 Two containment spray trains and two residual heat removal (RHR) spray trains shall be OPERABLE.
NOTE-----------------------------------------------
The RHR spray train is not required in MODE 4.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One containment spray train inoperable.
A.1 Restore containment spray train to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B.
One RHR spray train inoperable.
B.1 Restore RHR spray train to OPERABLE status.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C.
Required Action and associated Completion Time not met.
C.1 Be in MODE 3.
AND C.2 Be in MODE 5.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 84 hours
HMS 3.6.8 Watts Bar-Unit 1 3.6-22 Amendment 10, 135 3.6 CONTAINMENT SYSTEMS 3.6.8 Hydrogen Mitigation System (HMS)
LCO 3.6.8 Two HMS trains shall be OPERABLE.
APPLICABILITY:
MODES 1 and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One HMS train inoperable.
A.1 Restore HMS train to OPERABLE status.
OR A.2 Perform SR 3.6.8.1 on the OPERABLE train.
7 days Once per 7 days B.
One containment region with no OPERABLE hydrogen ignitor.
B.1 Restore one hydrogen ignitor in the affected containment region to OPERABLE status.
7 days C.
Required Action and associated Completion Time not met.
C.1 Be in MODE 3.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
SURVEILLANCE REQUIREMENTS SR 3.6.8.1 SR 3.6.8.2 SR 3.6.8.3 SURVEILLANCE Energize each HMS train power supply breaker and verify E33 ignitors are energized in each train.
Verify at least one hydrogen igniter is OPERABLE in each containment region.
Energize each hydrogen igniter and verify temperature is E1700°F.
HMS 3.6.8 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Watts Bar-Unit 1 3.6-23 Amendment 10, 132, 135
Ice Condenser Doors 3.6.12 SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.3 SR 3.6.12.4 Watts Bar-Unit 1 SURVEILLANCE FREQUENCY Verify, by visual inspection, each inlet door is not impaired by ice, frost, or debris.
3 months during first year after receipt of license AND In accordance with the Surveillance Frequency Control Program Verify torque required to cause each inlet door to begin to open is :S 675 in-lb.
3.6-33 3 months during first year after receipt of license AND In accordance with the Surveillance Frequency Control Program (continued)
Amendment 3, 132, 135
SURVEILLANCE REQUIREMENTS (Continued)
SR 3.6.12.5 SR 3.6.12.6 Watts Bar-Unit 1 SURVEILLANCE Perform a torque test on a sampling ofO 50% of the inlet doors.
Verify for each intermediate deck door:
a.
No visual evidence of structural deterioration; b.
Free movement of the vent assemblies; and c.
Free movement of the door.
3.6-34 Ice Condenser Doors 3.6.12 FREQUENCY 3 months during first year after receipt of license In accordance with the Surveillance Frequency Control Program 3 months during first year after receipt of license AND In accordance with the Surveillance Frequency Control Program
( continued)
Amendment 3, 132, 135
ACTIONS (continued)
CONDITION REQUIRED ACTION B.
Required Action and B.1 Be in MODE 3.
associated Completion Time of Condition A not AND
- met.
B.2 Be in MODE 5.
SURVEILLANCE REQUIREMENTS SR 3.7.8.1 SURVEILLANCE
NOTE--------
lsolation of ERCW flow to individual components does not render the ERCW inoperable.
Verify each ERCW manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.
ERCW 3.7.8 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)
Watts Bar-Unit 1 3.7-20 Amendment 69, 132, 135
Design Features 4.0 (continued)
Watts Bar Unit 1 4.0-1 Amendment 8, 40, 48, 67, 77, 86, 107, 127, 135 4.0 DESIGN FEATURES 4.1 Site The Watts Bar Nuclear Plant is located on a tract of approximately 1770 acres in Rhea County on the west bank of the Tennessee River at river mile 528. The site is approximately 1-1/4 miles south of the Watts Bar Dam. The 1770 acre reservation is owned by the United States and is in the custody of TVA. The exclusion area is determined by a circle of radius 1200 meters centered on a point 20 feet from the north wall of the turbine building along the building centerline. The distance to the low population zone is a radius of 3 miles.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy, ZIRLO, or Optimized ZIRLOTM clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.
A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. For Unit 1, Watts Bar is authorized to place a maximum of 1792 Tritium Producing Burnable Absorber Rods into the reactor in an operating cycle.
4.2.2 Control Rod Assemblies The reactor core shall contain 57 control rod assemblies. The control material shall be either silver-indium-cadmium or boron carbide with silver indium cadmium tips as approved by the NRC.
Design Features 4.0 (continued)
Watts Bar Unit 1 4.0-5 Amendment 6, 135 Page Intentionally Left Blank
Design Features 4.0 (continued)
Watts Bar Unit 1 4.0-6 Amendment 6, 135 Page Intentionally Left Blank
Procedures, Programs, and Manuals 5.7 (continued)
Watts Bar-Unit 1 5.0-24 Amendment 5, 63, 135 5.7 Procedures, Programs, and Manuals 5.7.2.18 Safety Function Determination Program (SFDP) (continued)
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
a.
A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or b.
A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or c.
A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 15.0 psig.
The maximum allowable containment leakage rate, La, at Pa, is 0.25% of the primary containment air weight per day.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 39 License No. NPF-96
- 1.
The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Tennessee Valley Authority (TVA, the licensee) dated June 7, 2019, as supplemented by letters dated October 9, 2019, and April 14, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.
- 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-96 is hereby amended to read as follows:
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 39 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3.
This license amendment is effective as of the date of its issuance and shall be implemented within 30 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Undine Shoop, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Facility Operating License and Technical Specifications Date of Issuance: June 22, 2020 Undine S.
Shoop Digitally signed by Undine S.
Shoop Date: 2020.06.22 14:26:39
-04'00'
ATTACHMENT TO AMENDMENT NO. 39 WATTS BAR NUCLEAR PLANT, UNIT 2 FACILITY OPERATING LICENSE NO. NPF-96 DOCKET NO. 50-391 Replace page 3 of Facility Operating License No. NPF-96 with the attached revised page 3.
The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Pages Insert Pages vi vi vii vii 3.0-4 3.0-4 3.0-5 3.0-6 3.0-7 3.0-8 3.0-9 3.0-10 3.0-11 3.7-17 3.7-17 3.9-8 3.9-8 4.0-1 4.0-1 4.0-4 4.0-4 4.0-5 4.0-5 5.0-25 5.0-25 5.0-25a Unit 2 Facility Operating License No. NPF-96 Amendment No. 39 C.
The license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act, and to the rules, regulations, and orders of the Commission now or hereafter in effect, and is subject to the additional conditions specified or incorporated below.
(1)
Maximum Power Level TVA is authorized to operate the facility at reactor core power levels not in excess of 3411 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 39 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. TVA shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
TVA shall implement permanent modifications to prevent overtopping of the embankments of the Fort Loudon Dam due to the Probable Maximum Flood by June 30, 2018.
(4)
PAD4TCD may be used to establish core operating limits until the WBN Unit 2 steam generators are replaced with steam generators equivalent to the existing steam generators at WBN Unit 1.
(5)
By December 31, 2019, the licensee shall report to the NRC that the actions to resolve the issues identified in Bulletin 2012-01, Design Vulnerability in Electrical Power System, have been implemented.
(6)
The licensee shall maintain in effect the provisions of the physical security plan, security personnel training and qualification plan, and safeguards contingency plan, and all amendments made pursuant to the authority of 10 CFR 50.90 and 50.54(p).
(7)
TVA shall fully implement and maintain in effect all provisions of the Commission approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The TVA approved CSP was discussed in NUREG-0847, Supplement 28, as amended by changes approved in License Amendment No. 7.
(8)
TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report for the facility, as described in NUREG-0847, Supplement 29, subject to the following provision:
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 Watts Bar - Unit 2 vi (continued)
Amendment 39
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 Watts Bar - Unit 2 (continued)
Amendment 39
SR Applicability 3.0 3.0 SR APPLICABILITY (continued)
Watts Bar - Unit 2 3.0-4 (continued)
Amendment No. 3, 12, 39 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.
Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a "once per..."
basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
SR Applicability 3.0 3.0 SR APPLICABILITY (continued)
Watts Bar - Unit 2 3.0-5 Amendment No. 3, 39 SR 3.0.3 (continued)
When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3.
When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.
This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
SURVEILLANCE REQUIREMENTS SR 3.7.7.1 SR 3.7.7.2 SR 3.7.7.3 SR 3.7.7.4 SR 3.7.7.5 Watts Bar - Unit 2 SURVEILLANCE Verify that the alternate feeder breaker to the C-S pump is open.
NOTE------
lsolation of CCS flow to individual components does not render the CCS inoperable.
Verify each CCS manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.
Verify each CCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.
NOTE----------------------
Verification of CCS pump 1 B-B automatic start on Unit 2 SI is not required when CCS pump 1 B-B is supporting CCS Train B OPERABILITY.
Verify each CCS pump starts automatically on an actual or simulated actuation signal.
---NOTE--------------------
Only required to be met when CCS pump 1 B-B is supporting CCS Train B OPERABILITY.
Verify CCS pump 1 B-B is aligned to CCS Train B and is in operation.
3.7-17 ccs 3.7.7 FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Amendment 36, 39
RHR and Coolant Circulation - Low Water Level 3.9.6 Watts Bar - Unit 2 3.9-8 Amendment 39 3.9 REFUELING OPERATIONS 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level LCO 3.9.6 Two RHR loops shall be OPERABLE, and one RHR loop shall be in operation.
APPLICABILITY:
MODE 6 with the water level < 23 ft above the top of reactor vessel flange.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Less than the required number of RHR loops OPERABLE.
A.1 Initiate action to restore required RHR loops to OPERABLE status.
Immediately OR A.2 Initiate action to establish 23 ft of water above the top of reactor vessel flange.
Immediately (continued)
Design Features 4.0 Watts Bar - Unit 2 4.0-1 (continued)
Amendment 27, 30, 39 4.0 DESIGN FEATURES 4.1 Site The Watts Bar Nuclear Plant is located on a tract of approximately 1770 acres in Rhea County on the west bank of the Tennessee River at river mile 528. The site is approximately 1-1/4 miles south of the Watts Bar Dam. The 1770 acre reservation is owned by the United States and is in the custody of TVA. The exclusion area is determined by a circle of radius 1200 meters centered on a point 20 feet from the north wall of the turbine building along the building centerline. The distance to the low population zone is a radius of 3 miles.
4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of ZIRLO or Optimized ZIRLO' clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.
Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. For Unit 2, Watts Bar is authorized to place a maximum of 1792 Tritium Producing Burnable Absorber Rods into the reactor in an operating cycle.
4.2.2 Control Rod Assemblies The reactor core shall contain 57 control rod assemblies. The control material shall be silver indium cadmium as approved by the NRC.
Design Features 4.0 (continued)
Watts Bar - Unit 2 4.0-4 Amendment 39 Page Intentionally Left Blank
Design Features 4.0 (continued)
Watts Bar - Unit 2 4.0-5 Amendment 39 A
Page Intentionally Left Blank
Procedures, Programs, and Manuals 5.7 5.7 Procedures, Programs, and Manuals (continued)
Watts Bar - Unit 2 5.0-25 Amendment 11, 39 5.7.2.18 Safety Function Determination Program (SFDP) (continued)
A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:
a.
A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or b.
A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or c.
A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.
The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
5.7.2.19 Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995.
For containment leakage rate testing purposes, a value of 15.0 psig, which is equivalent to the maximum allowable internal containment pressure, is utilized for Pa to bound the peak calculated containment internal pressure for the design basis loss of coolant accident.
The maximum allowable containment leakage rate, La, at Pa, is 0.25%
of the primary containment air weight per day.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 135 AND 39 TO FACILITY OPERATING LICENSE NOS. NPF-90 AND NPF-96 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391
1.0 INTRODUCTION
By application dated June 7, 2019, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19158A398), as supplemented by letters dated October 9, 2019 (ADAMS Accession No. ML19283G098), and April 14, 2020 (ADAMS Accession No. ML20105A343), the Tennessee Valley Authority (TVA, the licensee) submitted a license amendment request (LAR) to revise Watts Bar Nuclear Plant, Unit 1 and 2, Technical Specifications (TSs). The proposed changes would make several administrative changes, including: elimination of historical one-time license amendments; replacement of TS Figures 4.1-1 and 4.1-2; revision of Unit 2 TS 3.7.7 and 3.9.6 to be consistent with Unit 1 TSs; and correction of the Table of Contents.
The supplemental letters dated October 9, 2019, and April 14, 2020, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or Commission) staffs original proposed no significant hazards consideration determination as published in the Federal Register on August 13, 2019 (84 FR 40099).
2.0 REGULATORY EVALUATION
2.1 Description of Proposed Changes In the LAR, the licensee proposed removing historical one-time license amendments from the following TSs:
WBN, Unit 1 3.0, 3.5.2, 3.6.6, 3.6.8, 3.6.12, 3.7.8, 5.7.2.19 WBN, Unit 2 3.0, 5.7.2.19 The licensee proposed to replace TS Figures 4.1-1 and 4.1-2 with text.
The licensee also proposed to revise TSs 3.7.7, 3.9.6, and various footers in TSs for consistency between WBN, Units 1 and 2.
Lastly, the licensee proposed to revise the Table of Contents (TOC) based on license amendments that caused the changes to the corresponding TSs but did not recognize the associated TOC effects.
The NRC staffs evaluation of the specific TS changes described in the license amendment request is provided in Section 3.0 below.
2.2 Applicable Regulatory Requirements and Guidance Title 10 of the Code of Federal Regulations (10 CFR) 50.36, Technical Specifications, establishes the regulatory requirements related to the content of TSs. Paragraph 50.36(a)(1) requires an application for an operating license to include proposed TSs. 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 TSs.
Pursuant to 10 CFR 50.36, TSs for operating reactors are required to include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirement (SRs);
(4) design features; and (5) administrative controls. In accordance with 10 CFR 50.36(c)(2),
LCOs are the lowest functional capability or performance level of equipment required for safe operation of the facility. When LCOs are not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the LCO can be met. In accordance with 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.
Pursuant to 10 CFR 50.57(a)(3), Issuance of operating license, the Commission may issue an operating license upon finding, in part: There is reasonable assurance (i) that the activities authorized by the operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the regulations in this chapter Volumes 1 and 2 of NUREG-1431, Revision 4 (ADAMS Accession Nos. ML12100A222 and ML12100A228, respectively), contain the standard technical specifications (STS) and bases for Westinghouse Plants. The NRC staff has prepared STSs for each of the light-water reactor nuclear designs. These TSs have been found to be acceptable by the NRC staff for reactors similar to those at WBN, Units 1 and 2. These TSs may be used as a guide to the extent that they match the systems and equipment at WBN.
3.0 TECHNICAL EVALUATION
The NRC staff used the information provided by the licensee in its June 7, 2019, amendment request, as supplemented by letters dated October 9, 2019, and April 14, 2020, to review the proposed changes to the TSs. All the proposed changes in the licensees application propose administrative or editorial changes. None of the proposed changes revise either units design bases. Therefore, the NRC staff finds these changes acceptable for the reasons explained in each section below.
3.1 Evaluation of One-Time License Amendment Removal 3.1.1 WBN, Unit 1 Proposed Changes The following table describes the WBN, Unit 1 one-time TSs that are proposed for removal.
Technical Specification Information Being Removed Reason/TS Expiration Date TS 3.0, Surveillance Requirement (SR)
Applicability Delete Table SR 3.0.2-1 and the text invoking it in SR 3.0.2 November 30, 2017 TS 3.5.2, ECCS
[Emergency Core Cooling System] -Operating Delete Note in Frequency column of SR 3.5.2.3 Fall 2003 refueling outage TS 3.6.6, Containment Spray System Delete footnote and reference to it for Completion Time of Condition A June 30, 2013 TS 3.6.8, Hydrogen Mitigating System (HMS)
Delete the Note and references to it for LCO 3.6.8, Condition A and Condition B Delete the Note and reference to it for SR 3.6.8.1 The next WBN Unit 1 entry into MODE 3 after June 9, 1998.
TS 3.6.12, Ice Condenser Doors Delete Note in Frequency column of SRs 3.6.12.3, SR 3.6.12.4, and SR 3.6.12.5 October 21, 1996 TS 3.7.8, Essential Raw Cooling Water (ERCW)
System Delete Condition C and associated footnote July 31, 2008 TS 5.7.2.19, Containment Leakage Rate Testing Program Delete second paragraph regarding conducting the 10-year Type A test Fall 2012 The NRC staff finds that these changes are editorial in nature and are acceptable because:
These items are no longer relevant to the current TSs because the dates for each of the one-time license amendments for WBN, Unit 1 have expired.
The proposed changes simplify the TSs and eliminate a potential source of confusion to plant operators.
3.1.2 WBN, Unit 2 Proposed Changes The following table describes the WBN, Unit 2 one-time TSs that are proposed for removal.
Technical Specification Information Being Removed TS Expiration Date TS 3.0, Surveillance Requirement (SR)
Applicability Delete Table SR 3.0.2-1 (6 total pages) and the text invoking it in SR 3.0.2.
November 30, 2017 TS 5.7.2.19, Containment Leakage Rate Testing Program Delete Table 5.7.2-1 (and page) and the text invoking it in TS 5.7.2.19.
December 31, 2017 The NRC staff finds that these changes are editorial in nature and are acceptable because:
These items are no longer relevant to the current TSs because the dates for each of the one-time license amendments for WBN, Unit 2 have expired.
The proposed changes simplify the TSs and eliminate a potential source of confusion to plant operators.
3.2 Evaluation of Replacement of TS Figures 4.1-1 and 4.1-2 with Text The licensee proposed to revise TS Section 4.0, Design Features, by deleting the text in TS 4.1.1 and 4.1.2, which refers to Figures 4.1-1 and 4.1-2, and adding a description of the site location in Section 4.1, Site. Figures 4.1-1, Site and Exclusion Area Boundary, and 4.1-2, Low Population Zone, will be removed from the WBN, Units 1 and 2 TSs and replaced with Page Intentionally Left Blank. The proposed replacement text is shown below:
The Watts Bar Nuclear Plant is located on a tract of approximately 1770 acres in Rhea County on the west bank of the Tennessee River at river mile 528. The site is approximately 1-1/4 miles south of the Watts Bar Dam. The 1770 acre reservation is owned by the United States and is in the custody of TVA. The exclusion area is determined by a circle of radius 1200 meters centered on a point 20 feet from the north wall of the turbine building along the building centerline. The distance to the low population zone is a radius of 3 miles.
The NRC staff finds that these changes are administrative in nature and are acceptable because:
The proposed replacement text is derived from Chapter 2, Site Characteristics, of the dual-unit WBN Updated Final Safety Analysis Report (UFSAR) (ADAMS Accession No. ML19176A153). Additionally, the information on these figures is depicted in UFSAR Figures 2.1-4B, Site Boundary/Exclusion Area Boundary, and 2.1-3, Site Location 0-10 Miles.
The changes will make the WBN, Units 1 and 2 TSs more like the STS and the TSs of other TVA nuclear sites, which do not include similar figures.
3.3 Evaluation of Consistency Changes The licensee proposed the following miscellaneous changes to WBN, Unit 2 TSs to gain consistency between WBN Units 1 and 2:
TS 3.7.7, Component Cooling System (CCS), - Insert Unit 2 in Note to SR 3.7.7.4.
TS 3.9.6, Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level,- Change AND to OR as connector between Required Action A.1 and A.2.
Various - Miscellaneous footer changes 3.3.1 Surveillance Requirement 3.7.7.4 The SR 3.7.7.4 Note for WBN, Unit 2 presently states:
Verification of CCS pump 1B-B automatic start on SI [safety injection] is not required when CCS Pump 1B-B is supporting CCS Train B OPERABILITY.
The proposed SR 3.7.7.4 Note for WBN, Unit 2 is (proposed addition is in bold):
Verification of CCS pump 1B-B automatic start on Unit 2 SI is not required when CCS Pump 1B-B is supporting CCS Train B OPERABILITY.
This proposed change is editorial in nature. The NRC staff finds that this change is acceptable because the insertion of Unit 2 before SI is done for consistency with the Unit 1 TSs and was already reflected in the Unit 2 TS SR 3.7.7.4 Bases.
3.3.2 Technical Specification 3.9.6 WBN, Unit 2 LCO 3.9.6 REQUIRED ACTION A presently states:
A.1 Initiate action to restore required RHR loops to OPERABLE status.
AND A.2 Initiate action to establish 23 ft of water above the top of reactor vessel.
The proposed WBN, Unit 2 LCO 3.9.6 REQUIRED ACTION A is (proposed change is in bold):
A.1 Initiate action to restore required RHR loops to OPERABLE status.
OR A.2 Initiate action to establish 23 ft of water above the top of reactor vessel.
Required Actions A.1 and A.2 are connected by an OR logic statement in the WBN, Unit 1 TSs and STS. The licensee provided the following statement from the plants original TS Bases for WBN Unit 2:
If less than the required number of RHR loops are OPERABLE, actions shall be immediately initiated and continued until the RHR loop is restored to OPERABLE status and to operation or until 23 ft of water level is established above the reactor vessel flange. When the water level is 23 ft above the reactor vessel flange, the Applicability changes to that of LCO 3.9.5, and only one RHR loop is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions.
Based on a review of information from the original design basis of the RHR in Section 5.5.7 of the WBN Final Safety Analysis Report (ADAMS Accession No. ML19176A139) and from the statement from the plants TS bases for LCO 3.9.6 given above, the NRC staff determined that either action will restore compliance with the LCO. Per 10 CFR 50.36(c)(2), When a limiting condition for operation is not met, the licensee shall shutdown the reactor or follow any remedial action specified by the technical specifications until the condition can be met. Therefore, the NRC staff finds that the change of the logical connector AND to OR is acceptable.
3.3.3 Miscellaneous Footer Changes Where a TS page is otherwise affected by this LAR, the footers have been revised to the correct format. The NRC staff finds that these changes are considered editorial in nature and are, therefore, acceptable.
3.4 Evaluation of Changes to Table of Contents The licensee proposed to make several changes to the TS TOC based on license amendments that caused the changes to the corresponding TSs but did not recognize the associated effects on the TOC.
3.4.1 WBN, Unit 1 Changes to TS TOC The following table describes the proposed changes to the WBN, Unit 1 TS TOC.
TS Table of Contents Item Proposed Change Pages iii, v, 4.0-5, 4.0-6, Delete handwritten note at the bottom of each page, -to be implemented no later than completion of the reracking modification or prior to the movement of fuel assemblies into the spent fuel pool for the Cycle 1 refueling outage 3.4.17 Add TS 3.4.17, Steam Generator (SG) Tube Integrity, with page 3.4-43 3.6.7 TS 3.6.7 Replaced Hydrogen Recombiners with Deleted 3.7.16 Add TS 3.7.16, Component Cooling System (CCS) - Shutdown, with page 3.7-33 3.7.17 Add TS 3.7.17, Essential Raw Cooling Water (ERCW) System - Shutdown with page 3.7-36 TS Table of Contents Item Proposed Change 3.8.6 Change title from Battery Cell Parameters, to Battery Parameters 3.9.4 Replace Containment Penetrations with Deleted 3.9.8 Replace Reactor Building Purge Air Cleanup Units with Deleted 3.9.10 Add TS 3.9.10, Decay Time, with Page number 3.9-17 4.2 Add TS 4.2, Reactor Core, with Page number 4.0-1 4.3 Add TS 4.3, Fuel Storage with Page number 4.0-2 5.8 Revise Page number from 5.0-29 to 5.0-26 5.9 Revise Page number from 5.0-30 to 5.0-27 5.10 Revise Page number from 5.0-36 to 5.0-33 5.11 Revise Page number from 5.0-37 to 5.0-34 Table 3.3.4-1 Delete Table and Page number 3.3-48 Table 3.3.6-1 Revise Page number from 3.3-56, to 3.3-55 Table 3.3.8-1 Revise Page number from 3.3-64, to 3.3-63 Table 3.7.1-1 Revise Title to OPERABLE Main Steam Safety Valves versus Maximum Allowable Power Table 3.7.15-1 Delete Table with Page number 3.7-32 Table 3.8.6-1 Delete Table and Page number 3.8-36 Table 5.6.2.12-1 Delete Table with Page number 5.0-20 Table 5.7.2.12-2 Delete Table with Page number 5.0-21 Figure 3.4.16-1 Replace Reactor Coolant DOSE EQUIVALENT I-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER with Deleted Figure 4.1-1 Delete Figure with Page number 4.0-4 Figure 4.1-2 Delete Figure with Page number 4.0-5 Figure 4.3-1 Add Figure 4.3-1, Spent Fuel Pool Plan and Page number 4.0-7 Figure 4.3-2 Add Figure 4.3-2, New Fuel Storage Rack Loading Pattern and Page number 4.0-8 Section 50.36 of 10 CFR does not require that a Table of Contents be controlled as part of TSs.
In the STS, a TOC is shown; however, the STS are considered guidance. Because the licensee has submitted these changes in this LAR, and based on the information above, the NRC staff finds that these WBN Unit 1 TS Table of Contents changes are administrative in nature and are, therefore, acceptable.
3.4.2 WBN, Unit 2 Changes to TS TOC The following table describes the proposed changes to the WBN, Unit 2 TS TOC.
TS Table of Contents Item Proposed Change Table 3.3.4-1 Delete Table with Page number 3.3-48 Table 3.8.6-1 Delete Table with Page number 3.8-32 Figure 4.1-1 Delete Figure with Page number 4.0-4 Figure 4.1-2 Delete Figure with Page number 4.0-5 Figure 4.3-3 Delete Figure with Page number 4.0-8 TS Table of Contents Item Proposed Change Figure 4.3-4 Delete Figure with Page number 4.0-9 As discussed above in Section 3.4.1, the NRC staff finds that the proposed changes to the WBN, Unit 2 TS TOC are administrative in nature and are, therefore, acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendments on January 15, 2020. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change SRs.
The amendments also make editorial, corrective or other minor revisions. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission previously issued a proposed finding that the amendment involves no significant hazards consideration, published in the Federal Register on August 13, 2019 (84 FR 40099),
and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and (c)(10)(v).
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the 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.
Principal Contributors:
L. Wheeler, NRR P. Snyder, NRR Date: June 22, 2020
- by e-mail OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DSS/STSB/BC*
NAME KGreen BAbeywickrama VCusumano (w/edits)
DATE 03/27/2020 01/17/2020 12/13/2019 OFFICE OGC* - NLO NRR/DORL/LPL2-2/BC*
NRR/DORL/LPL2-2/PM NAME STurk UShoop KGreen DATE 05/28/2020 06/22/2020 06/22/2020