ML22257A051
| ML22257A051 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 11/04/2022 |
| From: | Kimberly Green Plant Licensing Branch II |
| To: | Jim Barstow Tennessee Valley Authority |
| Green K | |
| References | |
| EPID L-2021-LLA-0224 | |
| Download: ML22257A051 (32) | |
Text
November 4, 2022 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. 156 AND 64 REGARDING ADOPTION OF TECHNICAL SPECIFICATION TASK FORCE (TSTF) TRAVELER TSTF-205-A, REVISION 3, REVISION OF CHANNEL CALIBRATION, CHANNEL FUNCTIONAL TEST, AND RELATED DEFINITIONS, AND TSTF-563-A, REVISE INSTRUMENT TESTING DEFINITIONS TO INCORPORATE THE SURVEILLANCE FREQUENCY CONTROL PROGRAM (EPID L-2021-LLA-0224)
Dear Mr. Barstow:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 156 to Facility Operating License No. NPF-90 and Amendment No. 64 to Facility Operating License No. NPF-96 for the Watts Bar Nuclear Plant, Units 1 and 2, respectively. These amendments are in response to your application dated December 9, 2021.
The amendments revise a few instrument testing and calibration definitions in the Watts Bar Nuclear Plant, Units 1 and 2, technical specifications (TSs), and incorporate the surveillance frequency control program into a few of these definitions. The amendments are based on TS Task Force (TSTF) Traveler TSTF-205-A, Revision 3, Revision of Channel Calibration, Channel Functional Test, and Related Definitions, and TSTF-563-A, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program.
A copy of the related safety evaluation is also enclosed. A Notice of issuance will be included in the Commissions monthly 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. 156 to NPF-90
- 2. Amendment No. 64 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. 156 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 December 9, 2021, 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.
J. Barstow 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. 156 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.
FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: November 4, 2022 David J.
Wrona Digitally signed by David J. Wrona Date: 2022.11.04 10:37:00 -04'00'
ATTACHMENT TO AMENDMENT NO. 156 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 pages are identified by amendment number and contain marginal lines indicating the area of change.
Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Pages Insert Pages 1.1-1 1.1-1 1.1-2 1.1-2 1.1-4 1.1-4 1.1-6 1.1-6
Facility License No. NPF-90 Amendment No. 156 (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. 156 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.
Definitions 1.1 Watts Bar-Unit 1 1.1-1 (continued)
Amendment 156 1.0 USE AND APPLICATION 1.1 Definitions
NOTE---------------------------------------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.
AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux signals between (AFD) the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.
The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
Definitions 1.1 1.1 Definitions (continued)
Watts Bar-Unit 1 1.1-2 Amendment 146, 156 CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)
CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)
DOSE EQUIVALENT I-131 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
CORE ALTERATION shall be the movement of any fuel, sources, or other reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
The COLR is the unit specific document that provides cycle specific parameter limits for the initial and current reload cycle.
These cycle specific parameter limits shall be determined for the initial and each reload cycle in accordance with Specification 5.9.5. Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using thyroid dose conversion factors from Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977.
DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present.
If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."
(continued)
Definitions 1.1 1.1 Definitions (continued)
Watts Bar-Unit 1 1.1-4 LEAKAGE 3.
(continued)
LEAKAGE through a steam generator to the Secondary System (primary-to-secondary LEAKAGE);
b.
Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; c.
Pressure Boundary LEAKAGE LEAKAGE (except primary-to-secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
MASTER RELAY TEST MODE OPERABLE-OPERABILITY PDMS PHYSICS TESTS A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means RIany series of sequential, overlapping, or total steps.
A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
The Power Distributio Mo n
nitoring System (PDMS) is a real-time three dimensional core monitoring system. The system utilizes existing core instrumentation data and an on-line neutronics code to provide surveillance of core thermal limits.
PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
Amendment 65, 82, 156
Definitions 1.1 1.1 Definitions (continued)
Watts Bar-Unit 1 1.1-6 Amendment 156 SHUTDOWN MARGIN (SDM) a.
All rod cluster control assemblies (RCCAs) are fully (continued) inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST STAGGERED TEST BASIS THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)
A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means RIany series of sequential, overlapping, or total steps.
A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means
RI any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-391 WATTS BAR NUCLEAR PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 64 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 December 9, 2021, 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. 64 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.
FOR THE NUCLEAR REGULATORY COMMISSION David J. Wrona, Chief Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Operating License and Technical Specifications Date of Issuance: November 4, 2022 David J.
Wrona Digitally signed by David J. Wrona Date: 2022.11.04 10:40:00 -04'00'
ATTACHMENT TO AMENDMENT NO. 64 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 pages are identified by amendment number and contain marginal lines indicating the area of change.
Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Remove Pages Insert Pages 1.1-1 1.1-1 1.1-2 1.1-2 1.1-4 1.1-4 1.1-6 1.1-6 1.1-7 1.1-7 Unit 2 Facility Operating License No. NPF-96 Amendment No. 64 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 3459 megawatts thermal.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 64 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)
FULL SPECTRUM LOCA Methodology shall be implemented when 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:
Definitions 1.1 (continued)
Watts Bar - Unit 2 1.1-1 Amendment64 1.0 USE AND APPLICATION 1.1 Definitions
NOTE------------------------------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term Definition ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)
CHANNEL CALIBRATION ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.
AFD shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.
Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping RUtotal channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
Definitions 1.1 1.1 Definitions (continued)
(continued)
Watts Bar - Unit 2 1.1-2 Amendment 52, 64 CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)
CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)
DOSE EQUIVALENT I-131 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of all devices in the channel required for channel OPERABIL,TY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for the channel OPERABILITY such that the setpoints are within the necessary range and accuracy.
The COT may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
CORE ALTERATION shall be the movement of any fuel, sources, or other reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
The COLR is the unit specific document that provides cycle specific parameter limits for the initial and current reload cycle. These cycle specific parameter limits shall be determined for the initial and each reload cycle in accordance with Specification 5.9.5. Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. The determination of DOSE EQUIVALENT I-131 shall be performed using thyroid dose conversion factors from Table E-7 of Regulatory Guide 1.109, Rev. 1, NRC, 1977.
Definitions 1.1 1.1 Definitions (continued)
(continued)
Watts Bar - Unit 2 1.1-4 Amendment 64 LEAKAGE LEAKAGE shall be:
a.
Identified LEAKAGE 1.
LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.
LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or 3.
Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary-to-secondary LEAKAGE);
b.
Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; c.
Pressure Boundary LEAKAGE LEAKAGE (except primary-to-secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
MASTER RELAY TEST MODE A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means RIany series of sequential, overlapping, or total steps.
A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
Definitions 1.1 1.1 Definitions (continued)
(continued)
Watts Bar - Unit 2 1.1-6 QUADRANT POWER TILT RATIO (QPTR)
RATED THERMAL POWER (RTP)
REACTOR TRIP SYSTEM (RTS) RESPONSE TIME QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3459 MWt.
The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC, or the components have been evaluated in accordance with an NRC approved methodology.
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
a.
All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and b.
In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
Amendment 42, 47, 64
Definitions 1.1 1.1 Definitions (continued)
Watts Bar - Unit 2 1.1-7 Amendment 64 STAGGERED TEST BASIS THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)
A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NOS. 156 AND 64 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 December 9, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21344A027), the Tennessee Valley Authority (TVA, the licensee), submitted a license amendment request (LAR) for changes to the Watts Bar Nuclear Plant (Watts Bar), Units 1 and 2, Technical Specifications (TSs).
The proposed amendments would revise the Watts Bar, Units 1 and 2, TS definitions for Actuation Logic Test, Channel Calibration, Channel Operational Test (COT), Master Relay Test, Slave Relay Test, and Trip Actuating Device Operational Test (TADOT). These proposed changes are based on Technical Specification Task Force (TSTF) Traveler TSTF-205-A, Revision 3, Revision of Channel Calibration, Channel Functional Test, and Related Definitions dated December 23, 1998 (ML040570179). The U.S. Nuclear Regulatory Commission (NRC or the Commission) approved TSTF-205, Revision 3, on January 13, 1999 (ML20199E634).
The proposed amendments would also revise the definitions of Channel Calibration, COT, and TADOT to permit determination of the appropriate frequency to perform the Surveillance Requirement (SR) based on the devices being tested in each step. These proposed changes are based on TSTF Traveler TSTF-563, Revision 0, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program, dated May 10, 2017 (ML17130A819).
The NRC issued a final safety evaluation (SE) approving TSTF-563, Revision 0, on December 4, 2019 (ML18333A144).
A Surveillance Frequency Control Program (SFCP) was incorporated into the Watts Bar, Units 1 and 2, TSs in Amendment Nos. 132 and 36, respectively, dated February 28, 2020 (ML20028F733).
2.0 REGULATORY EVALUATION
2.1 Description of SFCP and Instrument Testing The Watts Bar, Units 1 and 2, TSs require the surveillances for instrumentation channels be performed within the specified frequency, using any series of sequential, overlapping, or total channel steps. Prior amendments revised the TSs to relocate all periodic surveillance frequencies to licensee control (ML20028F733). Changes to the relocated surveillance frequencies are made in accordance with the TS program referred to as the SFCP. The SFCP allows a new surveillance frequency to be determined for instrumentation channels, but that frequency must consider all components in the channel and applies to the entire channel.
A typical instrument channel consists of many different components, such as sensors, rack modules, and indicators. These components have different short-term and long-term performance (drift) characteristics, resulting in the potential for different calibration frequency requirements. Under the current TSs, the most limiting component calibration frequency for the channel must be chosen when a revised frequency is considered under the SFCP. As a result, all components that makeup a channel must be calibrated at a frequency equal to the channel component with the shortest (i.e., most frequent) surveillance frequency.
Some channel components, such as pressure transmitters, are very stable with respect to drift and could support a substantially longer calibration frequency than the other components in the channel. Currently, the SRs in many plants are performed in steps (e.g., a pressure sensor or transmitter is calibrated during a refueling outage and the rack signal conditioning modules are calibrated while operating at power). The proposed changes extend this concept to permit the surveillance frequency of each step to be determined under the SFCP based on the component(s) surveilled in the step instead of all components in the channel. This will allow each component to be tested at a frequency that is based on the components long-term performance characteristics rather than testing all components in a channel based on the component with the shortest surveillance frequency.
Allowing a surveillance frequency based on the components long-term performance characteristics for performing a channel calibration on each component or group of components could reduce radiation dose associated with in-place calibration of sensors, reduce wear on equipment, reduce unnecessary burden on plant staff, and reduce opportunities for calibration errors.
2.2 Proposed Changes to the Technical Specifications (TSTF-205 and TSTF-563)
Currently, the channel calibration and channel functional test may be performed by any series of sequential, overlapping, or total channel steps. The proposed changes to the TSs would revise the definitions of channel calibration and channel functional test to indicate that the step must be performed within the most limiting frequency for the components included in that step by adding the phrase, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step at the end of the last sentence of each definition.
The following paragraph denotes the changes to the ACTUATION LOGIC TEST definition related to TSTF-205-A. Changes are shown in bold italics:
An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output.
The following paragraph denotes the changes to the CHANNEL CALIBRATION definition related to TSTF-205-A. Deletions are shown in strikethrough text and additions are shown in bold italics:
A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel so output such that it responds within the required necessary range and accuracy to known input values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include all devices in the channel required for channel OPERABILITY.
and The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.
The following paragraph denotes the changes to the CHANNEL CALIBRATION definition related to TSTF-563. Changes are shown in bold italics:
The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
The following paragraph denotes the changes to the CHANNEL OPERATIONAL TEST (COT) definition related to TSTF-205-A. Deletions are shown in strikethrough text and additions are shown in bold italics:
A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so required for channel OPERABILITY such that the setpoints are within the required necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps The following paragraph denotes the changes to the COT definition related to TSTF-563.
Changes are shown in bold italics:
The COT may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
The following paragraph denotes the changes to the MASTER RELAY TEST definition related to TSTF-205-A. Deletions are shown in strikethrough text and additions are shown in bold italics:
A MASTER RELAY TEST shall consist of energizing each master relay all master relays in the channel required for OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
The following paragraph denotes the changes to the SLAVE RELAY TEST definition related to TSTF-205-A. Deletions are shown in strikethrough text and additions are shown in bold italics:
A SLAVE RELAY TEST shall consist of energizing each slave relay all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.
The following paragraph denotes the changes to the TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT) definition related to TSTF-205-A. Deletions are shown in strikethrough text and additions are shown in bold italics:
A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of the required alarm, interlock, display, and trip functions all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the required necessary accuracy.
The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps The following paragraph denotes the changes to the TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT) definition related to TSTF-563. Changes are shown in bold italics:
The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
The various instrumentation functions in the TSs require surveillances to verify the correct functioning of the instrument channel. The proposed changes extend the definition of instrumentation channel components to permit the surveillance frequency of each step to be determined under the SFCP based on the component(s) surveilled in the step instead of all components in the channel. This will allow each component to be tested at a frequency based on the components long-term performance characteristics.
The proposed changes in the definition for instrument testing would also allow the licensee to control the frequency of associated components being tested in each step. The SR for the overall instrumentation channel remains unchanged. The proposed changes have no effect on the design, fabrication, use, or methods of testing the instrumentation channels, and will not affect the ability of the instrumentation to perform the functions assumed in the safety analysis.
These instrumentation testing definitions state that, [t]he [test type] may be performed by means of any series of sequential, overlapping, or total channel steps. The surveillance frequency of these subsets would be established based on the characteristics of the components in the step rather than the most limiting component characteristics in the entire channel. Each of these steps would be evaluated in accordance with the SFCP.
The licensee is not proposing any variations from the TS changes described in TSTF-563 and TSTF-205-A, but noted differences in the Watts Bar TS Bases in relation to TSTF-205-A.
Specifically, Watts Bar has several different COT and TADOT SR numbers from the Standard Technical Specification (STS) SR numbers; STS has several COT and TADOT SRs that either Watts Bar does not have, or pertain to instrument functions that are not applicable to Watts Bar; or Watts Bar has several COT and TADOT SRs that apply to instrument functions not required by the STS.
2.3 Applicable Regulatory Requirements and Guidance Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36(a)(1) requires, in part, each applicant for a license authorizing operation of a utilization facility to include in the application the proposed TSs.
The regulation at 10 CFR 50.36(b) requires:
Each license authorizing operation of a utilization facility 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 [10 CFR] 50.34 [Contents of applications; technical information]. The Commission may include such additional technical specifications as the Commission finds appropriate.
The categories of items required to be in the TSs are provided in 10 CFR 50.36(c). One such category is SRs, which are defined in 10 CFR 50.36(c)(3) as 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 limiting conditions for operation will be met.
The regulation at 10 CFR 50.36(c)(5) requires TSs to include administrative controls, which are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.
Prior Watts Bar amendments replaced specific surveillance frequencies with references to a SFCP required by TS 5.7.2.23 (ML20028F733). That TS requires the licensee to establish, implement, and maintain an SFCP to ensure that TS SRs are performed at intervals listed in, and controlled by, the SFCP. Technical Specification 5.7.2.23 also requires that changes to surveillance frequencies listed in the SFCP be made in accordance with NRC staff-approved topical report Nuclear Energy Institute (NEI) 04-10, Revision 1, Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document, dated April 2007 (ML071360456).
Topical report NEI 04-10, Revision 1, describes an evaluation process and a multi-disciplinary plant decision-making panel that considers the detailed evaluation of proposed surveillance frequency revisions. The evaluations are based on operating experience, test history, manufacturers recommendations, codes and standards, and other deterministic factors, in conjunction with risk insights. The evaluation considers all components being tested by the SR.
Process elements are included for determining the cumulative risk impact of the changes, updating the licensees probabilistic risk assessment (PRA) models, and for imposing corrective actions, if necessary, following implementation of a revised frequency.
The NRC staffs guidance for the review of TSs is in Chapter 16.0, Technical Specifications, of NUREG-0800, Revision 3, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition (SRP), dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared STS for each of the LWR nuclear designs. Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with the applicable reference STS (i.e., the current STS), as modified by NRC-approved Travelers. In addition, the guidance states that comparing the change to previous STS can help clarify the TS intent.
Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (ML17317A256), describes an acceptable risk-informed approach for assessing the nature and impact of proposed permanent licensing basis changes by considering engineering issues and applying risk insights. This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.
Regulatory Guide 1.177, Revision 2, Plant-Specific, Risk-Informed Decision-Making: Technical Specifications, dated January 2021 (ML20164A034), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.
Regulatory Guide 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated December 2020 (ML20238B871), describes an acceptable approach for determining the technical adequacy of PRAs.
The NRC staffs guidance for evaluating the technical basis for proposed risk-informed changes is provided in SRP, Chapter 19, Section 19.2, Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance, dated June 2007 (ML071700658). The NRC staffs guidance for evaluating PRA technical adequacy is provided in SRP, Chapter 19, Section 19.1, Revision 3, Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load, dated September 2012 (ML12193A107). More specific review guidance related to risk-informed TS changes is provided in SRP, Chapter 16, Section 16.1, Revision 1, Risk-Informed Decision Making: Technical Specifications, dated March 2007 (ML070380228),
which includes changes to surveillance test intervals (STIs) (i.e., surveillance frequencies) as part of risk-informed decision-making. Section 19.2 of the SRP references the same criteria as RG 1.177, Revision 1, and RG 1.174, Revision 3, and states that a risk-informed application should be evaluated to ensure that the proposed changes meet the following key principles:
The proposed change meets the current regulations unless it explicitly relates to a requested exemption or rule change.
The proposed change is consistent with the defense-in-depth philosophy.
The proposed change maintains sufficient safety margins.
When proposed changes result in an increase in risk associated with core damage frequency or large early release frequency, the increase(s) should be small and consistent with the intent of the Commissions Safety Goal Policy Statement.
The impact of the proposed change should be monitored using performance measurement strategies.
The NRCs, Standard Technical Specifications, Westinghouse Plants, NUREG-1431, Volume 1, Specifications, and Volume 2, Bases, Revision 5 (ML21259A155 and ML21259A159, respectively) provide STS for Westinghouse-designed light-water reactors.
3.0 TECHNICAL EVALUATION
The NRC staff evaluated the licensees LAR to determine if the proposed changes are consistent with the regulations, guidance, and plant-specific design and licensing basis information discussed in Section 2.3 of this safety evaluation.
3.1 Adoption of TSTF-205 Traveler TSTF-205-A, Revision 3, revised the definitions for Channel Calibration, Channel Functional Test, and related definitions in the improved STSs to remove potential ambiguity in what constitutes an acceptable test.
Section 1.1 of the Watts Bar TSs includes definitions for instrumentation testing requirements. In accordance with 10 CFR 50.36(c)(3), Surveillance requirements, SRs are requirements related to test, calibration, or inspection needed to assure that the necessary quality of systems and components is maintained, facility operation is within limits, and that the TS limiting condition for operation is met. The definitions of Channel Calibration, Channel Functional Test, and Actuation Logic Test establish requirements for conducting testing including what the test involves, the scope of components that the test encompasses, and instructions on how the test is to be performed. The terms defined in Section 1.1 are referenced throughout instrumentation TS SRs to help assure consistent performance of SRs.
In its application, the licensee proposed to adopt TSTF-205-A, Revision 3, TS changes to the definitions of Actuation Logic Test, Channel Calibration, COT, Master Relay Test, Slave Relay Test, and TADOT to eliminate current ambiguity and potential misinterpretations of testing requirements.
The current definitions for instrumentation Channel Calibration, COT, and TADOT use the phrases the entire channel including the sensor and alarm and/or trip functions, and including alarm and/or trip functions and channel failure trips to describe those instrument channel devices required to be included for specified tests. The current definitions for instrumentation Master Relay and Slave Relay Tests use phrases each slave relay and each master relay.
There is ambiguity in whether the list is inclusive of all devices that must be tested or whether the list is representative of devices to be tested. Thus, the licensee proposed to adopt changes from TSTF-205-A, Revision 3, which would replace the string of required instrument channel devices in the definitions discussed above with all devices in the channel required for channel OPERABILITY.
The revised COT and TADOT definitions do not address the method for conducting testing of all required channel devices. The NRC staff position is that a successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable COT and TADOT of a relay. This is acceptable because all the other required contacts of the relay are verified by other TS and non-TS tests at least once per refueling interval with applicable extensions.
This NRC staff position is incorporated into the TSTF-205-A, Revision 3, Bases for COT and TADOT. The licensee proposed to adopt the approved TSTF-205-A, Revision 3, Bases to clarify testing requirements by modifying the Bases of applicable surveillances to provide acceptable methods of testing.
In addition, the current TS definition for Channel Calibration specifies that testing may be performed by means of any series of sequential, overlapping or total channel steps so that the entire channel is calibrated. The proposed TS changes delete the phrase such that the entire channel is calibrated from the definition of Channel Calibration to eliminate a verbatim conflict between the definition and the TSTF-205-A, Revision 3, Bases, which state a successful test to be the verification of the change of state of a single contact of the relay.
The NRC staff reviewed the changes proposed by the licensee and finds them acceptable because they are consistent with the STSs, do not result in any substantive change in operating requirements, and are consistent with 10 CFR 50.36(c)(3). These changes will provide for a consistent application of the definitions, tests, and calibrations.
3.2 Adoption of TSTF-563 Revising the frequency of a channel calibration and channel functional test instrument channel under the SFCP requires assurance that component performance characteristics, such as drift between each test, will not result in undetected instrument errors that exceed the assumptions of the safety analysis and supporting instrument loop uncertainty calculations. These requirements are consistent with the methodology described in NEI 04-10, which is required by the SFCP. The SFCP does not permit changes to the TS Allowable Values or Nominal Trip Setpoints; but allows only the surveillance frequency to be changed when determined permissible by NEI 04-10. Therefore, prior to extending the test intervals for an instrument channel component or components associated with a given calibration step, the component performance characteristics must be evaluated to verify the Allowable Value or Nominal Trip Setpoint will still be valid and to establish a firm technical basis supporting the extension. In addition, each change must be reviewed by the licensee to ensure the applicable uncertainty allowances are conservative (bounding) (e.g., sensor drift, rack drift, indicator drift).
Documentation to support the changes will be retained per the guidance in NEI 04-10.
Five key safety principles that must be evaluated before changing any surveillance frequency are identified in Section 3.0 of NEI 04-10. Principle 3 requires confirmation of the maintenance of safety margins, which, in this case, includes performance of deterministic evaluations to verify preservation of instrumentation trip setpoint and indication safety margins.
The evaluation methodology specified in NEI 04-10 also requires consideration of common-cause failure effects and monitoring of the instrument channel component performance following the frequency change to ensure channel performance is consistent with the analysis to support an extended frequency.
The method of evaluating a proposed surveillance frequency change is not dependent on the number of components in the channel. Each step needs to be evaluated to determine the acceptable surveillance frequency for that step. The proposed change to permit changing the surveillance frequency of channel component(s) does not affect the test method or evaluation method. The requirement to perform a channel calibration or channel functional test on the entire channel is not changed.
For example, an evaluation in accordance with NEI 04-10 may determine that a field sensor (e.g., a transmitter) should be calibrated every 48 months, that the rack modules should be calibrated every 30 months, and the indicators should be calibrated every 24 months. Under the current TS requirements, all devices in the channel must be calibrated every 24 months.
However, under the proposed change, sensors, rack modules, and indicators would be calibrated at the appropriate frequency for each of the tested devices. As required by the channel calibration definition, the test would still encompass all devices in the channel required for channel operability.
Per TS 5.7.2.23, the NEI 04-10 methodology is used to evaluate surveillance frequency changes to determine if such SR extensions could be applied. Process elements are used to determine the cumulative risk impact of changes, update the PRA, and impose corrective actions, if needed, following implementation. Several steps are required by NEI 04-10, Step 7, to be evaluated prior to determining the acceptability of changes. These steps include history of surveillance tests, industry and plant specific history, impact on defense in depth, vendor recommendations, required test frequencies for the applicable codes and standards, ensuring that plant licensing basis would not be invalidated and other factors. The NRC staff finds these measures acceptable in determining the SR extensions.
In addition, Step 16 of Section 4.0 of NEI 04-10 requires an Independent Decision-Making Panel (IDP) to review the cumulative impact of all STI changes over a period of time. This is also required by RGs 1.174 and 1.177. The IDP is composed of the site Maintenance Rule Expert Panel, Surveillance Test Coordinator, and Subject Matter Expert, who is a cognizant system manager or component engineer. Based on the above information, the NRC staff finds that the setpoint changes will be tracked in an acceptable manner.
Licensees with an SFCP may currently revise the surveillance frequency of instrumentation channels. The testing of these channels may be performed by means of any series, sequential, overlapping, or total channel steps. However, all required components in the instrumentation channel must be tested in order for the entire channel to be considered Operable.
The NRC staff notes that industry practice is to perform instrument channel surveillances, such as channel calibrations and channel functional tests, using separate procedures based on the location of the components. Each of these procedures may be considered a step. The results of all these procedures are used to satisfy the SRs using the existing allowance to perform it by means of any series of sequential, overlapping, or total channel steps. The proposed changes would allow for determining an acceptable surveillance frequency for each step.
The NRC staff notes that the NEI 04-10 methodology includes the determination of whether the structure, system, and components (SSCs) affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out. The methodology adjusts the failure probability of the impacted SSCs based on the proposed change to the surveillance frequency. Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria and truncation levels are addressed by the requirements for PRA technical adequacy, consistent with the guidance contained in RG 1.200, and by sensitivity studies identified in NEI 04-10. The licensee is not proposing to change the methodology, or the acceptance criteria for extending STIs, and the licensee will need to evaluate changes in the frequency for performing each of the steps in the instrumentation surveillance test per the methodology in NEI 04-10.
Therefore, the NRC staff concludes that the proposed changes to determine an acceptable test frequency for individual steps within instrumentation channel surveillance tests are acceptable because any extended STIs will be developed within the established constraints of the SFCP and NEI 04-10.
The regulatory requirements in 10 CFR 50.36 are not specific regarding the frequency of performing surveillance tests. The proposed changes only affect the frequency of performance and do not affect the surveillance testing method or acceptance criteria. Therefore, the proposed changes are consistent with the surveillance testing requirements of 10 CFR 50.36.
PRA Acceptability The guidance in RG 1.200 states that the quality of a licensees PRA should be commensurate with the safety significance of the proposed TS change and the role the PRA plays in justifying the change. That is, the greater the change in risk or the greater the uncertainty in that risk as a result of the requested TS change, or both, the more rigor that should go into ensuring the quality of the PRA.
The NRC staff previously performed an assessment of the PRA models used to support the approved SFCP that uses NEI 04-10, using the guidance of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of SSCs, using plant-specific data and models. Capability Category II of the NRC-endorsed PRA standard is the target capability level for supporting requirements for the internal events PRA for this application. Any identified deficiencies to those requirements are assessed further by the licensee to determine any impacts to proposed decreases to surveillance frequencies, including the use of sensitivity studies, where appropriate, in accordance with NEI 04-10.
The SFCP permits revising of the surveillance frequency for instrumentation channels. The NRC staff evaluated whether NEI 04-10 can be applied to subsets in an instrument channel when the SFCP currently specifies a surveillance interval that is applied to the entire channel. The NRC staff notes that the current channel surveillance may be performed by means of any series of sequential, overlapping, or total channel steps. In practice, this means that a channel is divided into subsets and each subset is tested separately. Therefore, the current instrument channel testing is already composed of a sequence of individual tests.
The instrument function may be modeled in the PRA differently depending on the site and the function (e.g., channel may be modeled individually, subsets may be modeled, or the channel function may be modeled as a single entity). There are different steps through the evaluation methodology in NEI 04-10 that could be used based on the different PRA modeling approaches.
The appropriate modeling of these different approaches is included in the NRC staffs review of the PRA modeling during the review of the application to implement an SFCP that uses NEI 04-10.
The PRA in use at Watts Bar is the same as that was used to support the application that implemented an SFCP that uses NEI 04-10. The amendments will change the capability of the licensee to change the surveillance frequency of an entire channel to now change the frequency of each subset of the channel. The NRC staff finds that changes to the surveillance frequency caused by defining and using individual, testable component subsets can be appropriately evaluated with the current SFCP and the current PRAs. The NRC staff finds that the risk-informed methodology review and the PRA acceptability review that were performed during the review of the licensees application to implement an SFCP that uses NEI 04-10 is adequate and still applicable.
The NRC staff determined that the proposed changes to the TSs meet the requirements for TSs in 10 CFR 50.36(b). The regulations at 10 CFR 50.36 require that TS include items in specified categories, including SRs. The proposed changes modify the definitions applicable to instrumentation channel components but do not alter the technical approach that was approved by the NRC in NEI 04-10, and the TS, as revised, continue to specify the appropriate SRs for tests and inspections to ensure the necessary quality of affected SSCs is maintained.
Additionally, the NRC staff finds the proposed TS changes to be technically clear and consistent with customary terminology and format in accordance with SRP Chapter 16.0. The NRC staff reviewed the proposed changes against the regulations and concludes that the changes continue to meet the requirements of Sections 50.36(b), 50.36(c)(3), and 50.36(c)(5) of 10 CFR, for the reasons discussed above, and thus provide reasonable assurance that the revised TSs provide the requisite requirements and controls for the facility to operate safely. Therefore, the NRC staff concludes that the proposed TS changes are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the Tennessee State official was notified of the proposed issuance of the amendment on August 9, 2022. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. 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 February 22, 2022 (87 FR 9653), 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). 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 Commission's 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 Contributor: T. Sweat, NRR Date: November 4, 2022
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