ML19357A195
| ML19357A195 | |
| Person / Time | |
|---|---|
| Site: | Point Beach, Seabrook, Turkey Point  |
| Issue date: | 02/10/2020 |
| From: | Justin Poole NRC/NRR/DORL/LPL1 |
| To: | Moul D Point Beach |
| Poole J, 415-2004 | |
| References | |
| EPID L-2019-LLA-0055 | |
| Download: ML19357A195 (43) | |
Text
UNITED STATES WASHINGTON, D.C. 20555-0001 February 10, 2020 Mr. Don Moul Vice President, Nuclear Division and Chief Nuclear Officer Florida Power & Light Company NextEra Energy Point Beach, LLC NextEra Energy Seabrook, LLC Mail Stop: NT3/JW 15430 Endeavor Drive Jupiter, FL 33478
SUBJECT:
POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2; SEABROOK STATION, UNIT NO. 1; AND TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 - ISSUANCE OF AMENDMENT NOS. 265, 268, 164, 290, AND 284 RE: REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-563 (EPID L-2019-LLA-0055)
Dear Mr. Moul:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the following enclosed amendments:
- Amendment Nos. 265 and 268 to Renewed Facility Operating License Nos. DPR-24 and DPR-27 for the Point Beach Nuclear Plant, Units 1 and 2 (Point Beach),
respectively
- Amendment No. 164 to Renewed Facility Operating License No. NPF-86 for the Seabrook Station, Unit No. 1 (Seabrook);
- Amendment Nos. 290 and 284 to Subsequent Renewed Facility Operating License Nos. DPR-31 and DPR-41 for the Turkey Point Nuclear Generating Unit Nos. 3 and 4 (Turkey Point), respectively.
In response to the application dated March 18, 2019 (L-2019-003), from NextEra Energy Resources/Florida Power & Light Company, the amendments revise the current instrumentation testing definitions of channel calibration, channel operational test, and trip actuating device operational test to permit determination of the appropriate frequency to perform the surveillance requirements based on the devices being tested in each step. The proposed changes are based on Technical Specifications Task Force Traveler, TSTF-563, Revision 0, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program,"
dated May 10, 2017, which was approved by the NRC on December 4, 2018.
D. Moul A copy of the Safety Evaluation is enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely,
/RA/
Justin C. Poole, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-266, 50-301, 50-443, 50-250, and 50-251
Enclosures:
- 1. Amendment No. 265 to DPR-24
- 2. Amendment No. 268 to DPR-27
- 3. Amendment No. 164 to NPF-86
- 4. Amendment No. 290 to DPR-31
- 5. Amendment No. 284 to DPR-41
- 6. Safety Evaluation cc: Listserv
NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 265 Renewed License No. DPR-24
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A The application for amendment filed by NextEra Energy Point Beach, LLC, (the licensee), dated March 18, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act}, and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.
Enclosure 1
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 4.B of the Renewed Facility Operating License No. DPR-24 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 265, are hereby incorporated in the renewed operating license. NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of issuance: February 10, 2020
WASHINGTON, D.C. 20555-0001 NEXTERA ENERGY POINT BEACH, LLC DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 268 Renewed License No. DPR-27
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by NextEra Energy Point Beach, LLC, (the licensee), dated March 18, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commi.ssion's 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 Commission's regulations and all applicable requirements have been satisfied.
Enclosure 2
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 4.8 of the Renewed Facility Operating License No. DPR-27 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 268, are hereby incorporated in the renewed operating license. NextEra Point Beach shall operate the facility in accordance with Technical Specifications.
- 3. This license amendment is effective as of the date of issuance and shall be implemented within 90 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of issuance: February 10, 2020
A TIACHMENT TO LICENSE AMENDMENTS POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 AMENDMENT NO. 265 RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AMENDMENT NO. 268 RENEWED FACILITY OPERATING LICENSE NO. DPR-27 DOCKET NOS. 50-266 AND 50-301 Replace the following page of Renewed Facility Operating License No. DPR-24 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3 3 Replace the following page of Renewed Facility Operating License No. DPR-27 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain a marginal line indicating the areas of change.
Remove Insert 1.1-1 1.1-1 1.1-2 1.1-2 1.1-6 1.1-6
D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach 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 or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.
- 4. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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 below:
A. Maximum Power Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 265, are hereby incorporated in the renewed operating license.
NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.
C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.
Renewed License No. DPR-24 Amendment No. 265
C. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach to receive, possess and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed source for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; D. Pursuant to the Act and 10 CFR Parts 30, 40 and 70, NextEra Energy Point Beach 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 or instrument calibration or associated with radioactive apparatus or components; and E. Pursuant to the Act and 10 CFR Parts 30 and 70, NextEra Energy Point Beach to possess such byproduct and special nuclear materials as may be produced by the operation of the facility, but not to separate such materials retained within the fuel cladding.
- 4. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; 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 below:
A. Maximum Power Levels NextEra Energy Point Beach is authorized to operate the facility at reactor core power levels not in excess of 1800 megawatts thermal.
B. Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 268, are hereby incorporated in the renewed operating license.
NextEra Energy Point Beach shall operate the facility in accordance with Technical Specifications.
C. Spent Fuel Pool Modification The licensee is authorized to modify the spent fuel storage pool to increase its storage capacity from 351 to 1502 assemblies as described in licensee's application dated March 21, 1978, as supplemented and amended. In the event that the on-site verification check for poison material in the poison assemblies discloses any missing boron plates, the NRC shall be notified and an on-site test on every poison assembly shall be performed.
Renewed License No. DPR-27 Amendment No. 268
Definitions 1.1 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.
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 (AFD) signals between 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. 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.
CHANNEL CHECK 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.
Point Beach 1.1-1 Unit 1 - Amendment No. 265 Unit 2 - Amendment No. 268
Definitions 1.1 1.1 Definitions CHANNEL OPERATIONAL A COT shall be the injection of a simulated or TEST (COT) 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 OPERATING LIMITS The COLR is the unit specific document that REPORT (COLR) provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.4. Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The determination of DOSE EQUIVALENT 1-131 shall be performed using thyroid dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion."
DOSE EQUIVALENT DOSE EQUIVALENT Xe-133 shall be that concentration of Xe-133 Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body Deep Dose Equivalent as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131 m, 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 111.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil" or the average gamma disintegration energies as provided in ICRP Publication 38, "Radionuclide Transformations," or similar source.
INSERVICE TESTING The INSERVICE TESTING PROGRAM is the licensee PROGRAM program that fulfills the requirements of 10 CFR 50.55a(f).
Point Beach 1.1-2 Unit 1 - Amendment No. 265 Unit 2 - Amendment No. 268
Definitions 1.1 1.1 Definitions TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip OPERATIONAL TEST actuating device and verifying the OPERABILITY of (TADOT) all devices in the channel required for trip actuating device OPERABILITY. 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.
Point Beach 1.1-6 Unit 1 - Amendment No. 265 Unit 2 - Amendment No. 268
NEXTERA ENERGY SEABROOK, LLC, ET AL.*
DOCKET NO. 50-443 SEABROOK STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 164 Renewed License No. NPF-86
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment filed by NextEra Energy Seabrook, LLC, et al.
(the licensee), dated March 18, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;
- 8. 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.
- NextEra Energy Seabrook, LLC, is authorized to act as agent for the: Hudson Light & Power Department, Massachusetts Municipal Wholesale Electric Company, and Taunton Municipal Lighting Plant and has exclusive responsibility and control over the physical construction, operation and maintenance of the facility.
Enclosure 3
- 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 Renewed Facility Operating License No. NPF-86 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 164, are incorporated into the Renewed Facility Operating License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: February 10, 2020
A TIACHMENT TO LICENSE AMENDMENT NO. 164 SEABROOK STATION, UNIT NO. 1 RENEWED FACILITY OPERATING LICENSE NO. NPF-86 DOCKET NO. 50-443 Replace the following page of Renewed Facility Operating License No. NPF-86 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove 3
Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain a marginal line indicating the area of change.
Remove Insert 1-1 1-1 1-2 1-2 1-7 1-7
(3) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR Part 70, to receive, possess, and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (4) NextEra Energy Seabrook, LLC, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (5) NextEra Energy Seabrook, LLC, 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 or instrument calibration or associated with radioactive apparatus or components; and (6) NextEra Energy Seabrook, LLC, 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 authorized herein.
(7) DELETED C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's 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 NextEra Energy Seabrook, LLC, is authorized to operate the facility at reactor core power levels not in excess of 3648 megawatts thermal (100% of rated power).
(2) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 164, are incorporated into the Renewed Facility Operating License No. NPF-86. NextEra Energy Seabrook, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.
ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.
ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.
ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy. The ANALOG CHANNEL OPERATIONAL TEST 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.
AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input.
The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions. 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.
CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
SEABROOK- UNIT 1 1-1 Amendment No. 164
DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrations required to be closed during accident conditions are either:
- 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or
- 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions.
- b. All equipment hatches are closed and sealed,
- c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,
- d. The containment leakage rates are in accordance with the Containment Leakage Rate Testing Program, and
- e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE AL TERATIONs shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) provides core operating limits for the current operating reload cycle. The cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.8.1.6. Plant operation within these operating limits is addressed in individual specifications.
DIGITAL CHANNEL OPERATIONAL TEST 1.11 A DIGITAL CHANNEL OPERATIONAL TEST shall consist of exercising the digital computer hardware using data base manipulation and/or injecting simulated process data to verify OPERABILITY of alarm and/or trip functions. The Digital Channel Operational Test definition is only applicable to the Radiation Monitoring Equipment. The DIGITAL CHANNEL OPERATIONAL TEST 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.
SEABROOK - UNIT 1 1-2 Amendment No. 49, 84,164
DEFINITIONS STAGGERED TEST BASIS 1.37 A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and
- b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
THERMAL POWER 1.38 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE OPERATIONAL TEST 1.39 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions.
The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy. The TRIP ACTUATING DEVICE OPERATIONAL TEST 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.
UNIDENTIFIED LEAKAGE 1.40 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.
UNRESTRICTED AREA 1.41 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.
1.42 (NOT USED)
VENTING 1.43 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
SEABROOK- UNIT 1 1-7 Amendment No. +,--9;--ee, 84-,
164
FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-250 TURKEY POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO SUBSEQUEN T RENEWED FACILITY OPERATING LICENSE Amendment No. 290 Subsequent Renewed License No. DPR-31
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Florida Power & Light Company (the licensee) dated March 18, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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 con.ducted in compliance with the Commission's 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 Commission's regulations and all applicable requirements have been satisfied.
Enclosure 4
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Subsequent Renewed Facility Operating License No. DPR-31 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 290, are hereby incorporated into this subsequent renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Subsequent Renewed Facility Operating License and Technical Specifications Date of Issuance: February 10, 2020
UNITED STATES WASHINGTON, D.C. 20555-0001 FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-251 TURKEY POINT NUCLEAR GENERATING UNIT NO. 4 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 284 Subsequent Renewed License No. DPR-41
- 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Florida Power & Light Company (the licensee) dated March 18, 2019, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.
Enclosure 5
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Subsequent Renewed Facility Operating License No. DPR-41 is hereby amended to read as follows:
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 284 are hereby incorporated into this subsequent renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Subsequent Renewed Facility Operating License and Technical Specifications Date of Issuance: February 10, 2020
ATTACHMENT TO LICENSE AMENDMENTS TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 AMENDME NT NO. 290 SUBSEQU ENT RENEWED FACILITY OPERATING LICENSE NO. DPR-31 AMENDME NT NO. 284 SUBSEQU ENT RENEWED FACILITY OPERATING LICENSE NO. DPR-41 DOCKET NOS. 50-250 AND 50-251 Replace the following page of Subsequent Renewed Facility Operating License No. DPR-31 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3 3 Replace the following page of Subsequent Renewed Facility Operating License No. DPR-41 with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Remove Insert 3 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain a marginal line indicating the areas of change.
Remove Insert 1-1 1-1 1-2 1-2 1-6 1-6
3 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 below:
A. Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2644 megawatts (thermal).
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 290, are hereby incorporated into this subsequent renewed license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this subsequent renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
C. Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on November 1, 2001, describes certain future inspection activities to be completed before the period of extended operation.
The licensee shall complete these activities no later than July 19, 2012.
The Final Safety Analysis Report supplement as revised on November 1, 2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50.71(e)(4), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
D. Fire Protection FPL shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment requests dated June 28, 2012 and October 17, 2018 (and supplements dated September 19, 2012; March 18, April 16, and May 15, 2013; January 7, April 4, June 6, July 18, September 12, November 5, and December 2, 2014; and February 18, 2015; October 24, and December 3, 2018; and January 31, 2019), and as approved in the safety evaluations dated May 28, 2015 and March 27, 2019. Except where NRC approval for changes or deviations is required by 10 CFR 50.48(c), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the Subsequent Renewed License No. DPR-31 Amendment No. 290
3 A. Maximum Power Level The applicant is authorized to operate the facility at reactor core power levels not in excess of 2644 megawatts (thermal).
B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 284, are hereby incorporated into this subsequent renewed operating license. The Environmental Protection Plan contained in Appendix B is hereby incorporated into this subsequent renewed license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
C. Final Safety Analysis Report The licensee's Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on November 1, 2001, describes certain future inspection activities to be completed before the period of extended operation.
The licensee shall complete these activities no later than April 10, 2013.
The Final Safety Analysis Report supplement as revised on November 1, 2001, described above, shall be included in the next scheduled update to the Final Safety Analysis Report required by 10 CFR 50. 71 (e )( 4 ), following the issuance of this renewed license. Until that update is complete, the licensee may make changes to the programs described in such supplement without prior Commission approval, provided that the licensee evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.
D. Fire Protection FPL shall implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), as specified in the licensee amendment requests dated June 28, 2012 and October 17, 2018 (and supplements dated September 19, 2012; March 18, April 16, and May 15, 2013; January 7, April 4, June 6, July 18, September 12, November 5, and December 2, 2014; and February 18, 2015; October 24, and December 3, 2018; and January 31, 2019), and as approved in the safety evaluations dated May 28, 2015 and March 27, 2019. Except where NRC approval for changes or deviations is required by 10 CFR 50.48( c ), and provided no other regulation, technical specification, license condition or requirement would require prior NRC approval, the licensee may make changes to the fire protection program without prior approval of the Commission if those changes satisfy the provisions set forth in 10 CFR 50.48(a) and 10 CFR 50.48(c), the change does not require a change to a technical specification or a license condition, and the criteria listed below are satisfied.
Risk-Informed Changes that May Be Made Without Prior NRC Approval A risk assessment of the change must demonstrate that the acceptance criteria below are met. The risk assessment approach, methods, and data shall be acceptable to the NRC and shall be appropriate for the nature and scope of the change being evaluated; be based on the as-built, as-operated, and maintained plant; and reflect the operating experience at the plant. Acceptable methods to assess the risk of the change may include methods that have been used in the Subsequent Renewed License No. DPR-41 Amendment No. 284
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.
ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.
ACTUATION LOGIC TEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATIO N LOGIC TEST shall include a continuity check. as a minimum, of output devices.
ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm. interlock and/or trip functions.
The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments. as necessary. of the alarm, interlock and/or Trip Setpoints such that the setpoints are within the required range and accuracy. The ANALOG CHANNEL OPERATIONAL TEST 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.
AXIAL FLUX DIFFERENCE 1.4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION 1.5 A CHANNEL CAUBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions. 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.
CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
TURKEY POINT - UNITS 3 & 4 1-1 AMENDMENT NOS. 290 AND 284
CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:
- a. All penetrations required to be closed during accident conditions are either:
- 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or
- 2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Specification 3.6.4.
- b. The equipment hatch is closed and sealed,
- c. Each air lock is in compliance with the requirements of Specification 3.6.1.3,
- d. The containment leakage rates are within the limits of Specification 3.6.1.2, and
- e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.
CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATIONS 1.9 CORE AL TERA TIONS shall be the movement of any fuel, sources, reactivity control components, or other components affecting reactivity within the reactor vessel with the vessel head removed and fuel in the vessel.
Suspension of CORE AL TERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS REPORT 1.10 The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with NRC approved methodology. Unit operation within these operating limits is addressed in individual specifications. The COLR is submitted to the NRC in accordance with the requirements of 6.9.1.7.
DIGITAL CHANNEL OPERATIONAL TEST 1.11 A DIGITAL CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock, and/or trip functions. The DIGITAL CHANNEL OPERATIONAL TEST 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.
TURKEY POINT - UNITS 3 & 4 1-2 AMENDMENT NOS. 290 AND 284
THERMAL POWER 1.28 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE OPERATIO NAL TEST 1.29 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required setpoint within the required accuracy. The TRIP ACTUATING DEVICE OPERATIO NAL TEST 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.
UNIDENTIFIED LEAKAGE 1.30 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.
UNRESTRICTED AREA 1.31 An UNRESTRICTED AREA shall mean an area, access to which is neither limited nor controlled by the licensee.
VENTING 1.32 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.
TURKEY POINT - UNITS 3 & 4 1-6 AMENDM ENT NOS. 290 AND 284
UNITED STATES WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMEN T NO. 265 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-24 AMENDMENT NO. 268 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-27 AMENDMEN T NO. 164 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-86 AMENDMENT NO. 290 TO SUBSEQUEN T RENEWED FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO. 284 TO SUBSEQUEN T RENEWED FACILITY OPERATING LICENSE NO. DPR-41 POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 SEABROOK STATION. UNIT NO. 1 TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 NEXTERA ENERGY RESOURCES /FLORIDA POWER & LIGHT COMPANY, ET AL.
DOCKET NOS. 50-266, 50-301, 50-443, 50-250, AND 50-251 1.0 INTRODUCTI ON By letter dated March 18, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19079A240), NextEra Energy Point Beach, LLC; NextEra Energy Seabrook, LLC; Florida Power & Light Company (NextEra/FPL), requested changes to the Technical Specifications {TSs) for Point Beach Nuclear Plant, Units 1 and 2 (Point Beach);
Seabrook Station, Unit No. 1 (Seabrook); and Turkey Point Generating Unit Nos. 3 and 4 (Turkey Point) (the facilities).
Specifically, NextEra/FPL requested revisions to the current instrumentation testing definitions of Channel Calibration, Channel Operational Test (COT), and Trip Actuating Device Operational Test {TADOT) in the Point Beach TSs, and Channel Calibration, Analog COT, Digital COT, and TADOT in the Seabrook and Turkey Point TSs, to permit determination of the appropriate frequency to perform the surveillance requirement (SR) based on the devices being tested in each step. The proposed changes are based on Technical Specifications Task Force (TSTF}
Traveler, TSTF-563, Revision 0, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program," dated May 10, 2017 (ADAMS Accession No. ML17130A819). The U.S. Nuclear Regulatory Commission (NRC or the Commission) issued a final safety evaluation (SE) approving TSTF-563, Revision 0, on December 4, 2018 (ADAMS Accession No. ML18333A144).
Enclosure 6
Surveillance Frequency Control Programs (SFCPs) were previously incorporated into the facilities TSs by the following amendments:
- Seabrook, Unit No. 1, License Amendment No. 141, dated July 24, 2014 (ADAMS Accession No. ML13212A069)
- Turkey Point, Unit Nos. 3 and 3, License Amendment Nos. 263 and 258, respectively, dated July 16, 2015 (ADAMS Accession No. ML15166A320)
- Point Beach, Units 1 and 2, License Amendment Nos. 253 and 257, respectively, dated July 28, 2015 (ADAMS Accession No. ML15195A201)
NextEra/FPL proposed variations to the changes described in TSTF-563. These variations are described in Section 2.2.1 of this SE and evaluated in Section 3.1 of this SE.
2.0 REGULATOR Y EVALUATION 2.1 Description of Surveillance Frequency Control Program and Instrument Testing The TSs require the surveillances for instrumentation channels to be performed within the specified frequency, using any series of sequential, overlapping, or total channel steps. A prior amendment revised the TSs to relocate all periodic surveillance frequencies to licensee control.
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 the channel, 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 make up 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 change extends 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 the appropriate frequency based on the component's long-term performance characteristics.
Allowing an appropriate surveillance frequency for performing a channel calibration on each component or group of components could reduce radiation dose associated with inplace 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 Currently, the channel calibration, COT (analog and digital for Seabrook and Turkey Point), and TADOT 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, COT (analog and digital for Seabrook and Turkey Point), and TADOT to indicate that the step must be performed within the most limiting frequency for the components included in that step.
The following paragraphs denote the changes to these three definitions. Additions are shown in italics, and deletions are shown in strikethrough:
Seabrook Station, Unit No. 1 ANALOG CHANNEL OPERATION AL TEST 1.3 An ANALOG CHANNEL OPERATION AL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILIT Y of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATION AL TEST shall include adjustments, as necessary, of the alarm, interlock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy. The ANALOG CHANNEL OPERATIONAL TEST 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.
CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions.
The CHANNEL CAL/BRA TION aAd 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 suoh that the entire ohannel is oalibrated.
DIGITAL CHANNEL OPERATION AL TEST 1.11 A DIGITAL CHANNEL OPERATION AL TEST shall consist of exercising the digital computer hardware using data base manipulation and/or injecting simulated process data to verify OPERABILIT Y of alarm and/or trip functions.
The Digital Channel Operation Test definition is only applicable to the Radiation Monitoring Equipment. The DIGITAL CHANNEL OPERATIONAL TEST 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.
TRIP ACTUATING DEVICE OPERATION AL TEST 1.39 A TRIP ACTUATING DEVICE OPERATION AL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILIT Y of alarm,
interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trip Actuating Device such that it actuates at the required Setpoint within the required accuracy. The TRIP AC TUA TING DEVICE OPERA TJONAL TEST 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.
Point Beach Nuclear Plant, Units 1 and 2 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. 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.
CHANNEL OPERATIONAL TEST (COT)
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.
TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)
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 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.
Turkey Point Nuclear Generating Units Nos. 3 and 4 ANALOG CHANNEL OPERATIONAL TEST 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify
OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or Trip Setpoints such that the setpoints are within the required range and accuracy. The ANALOG CHANNEL OPERATIONAL TEST 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.
CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions.
The CHANNEL CAL/BRA TION aoo 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 such that the entire channel is calibrated.
DIGITAL CHANNEL OPERATIONAL TEST 1.11 A DIGITAL CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock, and/or trip functions. The DIGITAL CHANNEL OPERATIONAL TEST 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.
TRIP ACTUATING DEVICE OPERATIONAL TEST 1.29 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or trip functions. The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include adjustment, as necessary, of the Trfp Actuating Device such that it actuates at the required setpoint within the required accuracy. The TRIP ACTUATIN G DEVICE OPERATIONAL TEST 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 change extends 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 the appropriate frequency based on the component's long-term performance characteristics.
The proposed changes in the definition for instrument testing would allow NextEra/FPL to control the frequency of associated components being tested in each step. The SR for the overall instrumentation channel remains unchanged. The proposed change has 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: The [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 is evaluated in accordance with the SFCP.
2.2.1 Variations from TSTF-563 The Seabrook, Point Beach, and Turkey Point TS requirements differ from the Standard Technical Specifications (STS) as follows:
(a) The Seabrook and Turkey Point TSs are different from Revision 4.0 of NUREG-1431, "Standard Technical Specifications, Westinghouse Plants," Volume 1, "Specifications,"
and Volume 2, "Bases," dated April 2012 (ADAMS Accession Nos. ML12100A222 and ML12100A228, respectively).
(b) The STS (NUREG-1431) on which TSTF-563 is based. The definition of Channel Calibration in the Seabrook and Turkey Point TS has wording differences from the STS (NUREG-143 1) definition. The Seabrook and Turkey Point TS Channel Calibration definition states: The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/ or trip functions and may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated." The STS separates these concepts into two sentences. The latter portion states: "The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps." In the proposed change, a period is placed after the phrase "trip functions" to be consistent with the standard. The phrases "The CHANNEL CALIBRATION" and "means of' are added, and the phrase "so that the entire channel is calibrated" is eliminated.
(c) The Seabrook and Turkey Point TSs are different from Revision 4 of the STS (NUREG-1431) on which TSTF-563 is based. Unlike the definitions in the Westinghouse STS, the Seabrook and Turkey Point definitions of TADOT do not include a statement that the tests may be performed by means of any series of sequential, overlapping, or total channel steps. This provision was added to the STS definitions of COT and TADOT by TSTF-205-A, Revision 3, "Revision of Channel Calibration, Channel Functional Test, and Related Definitions," which was approved by the NRC on January 13, 1999, and incorporated into Revision 2 of NUREG-1431. NextEra/FPL stated that the addition of this sentence to the T ADOT definition is necessary to adopt TSTF-563.
(d) The Seabrook and Turkey Point TSs are different from Revision 4 of the Westinghouse STS (NUREG-1431) on which TSTF-563 is based. Unlike the definitions in the Westinghouse STS, the Seabrook and Turkey Point definitions separate the COT definition into an "analog COT" and "digital COT." The Seabrook and Turkey Point divisions of the COT definition into definitions for analog COT and digital COT were added to the TSs by Seabrook Amendment No. 7, dated October 10, 1991 (ADAMS Accession No. ML011800188), and Turkey Point Amendment Nos. 140 and 135, dated April 23, 1991 (ADAMS Accession No. ML013380091). This difference does not affect
the applicability of the changes proposed in TSTF-563. As discussed above, the STS permits the COT to be performed by means of any series of sequential, overlapping, or total channel steps and was added to the STS definitions of COT by TSTF-205-A, Revision 3. NextEra/FPL stated that the addition of this sentence to the TADOT definition is necessary to adopt TSTF-563.
(e) TSTF-563 and the NRC's SE for TSTF-563 discuss the applicable regulatory requirements and guidance, including Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC). Seabrook was licensed to the 10 CFR Part 50, Appendix A GDC. Point Beach and Turkey Point were not licensed to the 10 CFR Part 50, Appendix A GDC, but were licensed to the applicable Atomic Energy Commission (AEC) preliminary general design criteria identified in the Federal Register notice published July 11, 1967 (32 FR 10213),
(ADAMS Accession No. ML043310029). The applicable 10 CFR Part 50, Appendix A GDC were compared to the proposed criteria, which are documented in the Point Beach and Turkey Point Updated Final Safety Analysis Reports (UFSARs).
2.3 Applicable Regulatory Requirements and Guidance The regulations in 10 CFR 50.36(a)(1) require each applicant for a license authorizing operation of a utilization facility to include the proposed TSs in the application.
The regulation at 10 CFR 50.36(b) requires, in part:
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 § 50.34. 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, 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 TS 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."
A prior amendment revised and relocated most periodic surveillance frequencies to licensee control. Changes to the relocated surveillance frequencies are made in accordance with the SFCP. The SFCP requires that changes to the relocated frequencies be made in accordance with NRC staff-approved topical report Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies, Industry Guidance Document."
NEI 04-10 describes an evaluation process and a multi-disciplinary plant decisionmaking panel that considers the detailed evaluation of proposed surveillance frequency revisions. The evaluations are based on operating experience, test history, manufacturer 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 licensee's probabilistic risk assessment (PRA) models, and for imposing corrective actions, if necessary, following implementatio n of a revised frequency.
The NRC staff's 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 (ADAMS Accession No. 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 staff's review includes consideration of whether the proposed changes are consistent with the applicable referenced 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 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ADAMS Accession No. ML100910006), 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.
RG 1.177, Revision 1, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:
Technical Specifications" (ADAMS Accession No. ML100910008), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.
RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (ADAMS Accession No. ML090410014),
describes an acceptable approach for determining the technical adequacy of PRAs.
The NRC staff's guidance for evaluating the technical basis for proposed risk-informed changes is provided in SRP Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance" (ADAMS Accession No. ML071700658). The NRC staff's guidance on evaluating PRA technical adequacy is provided in SRP Section 19.1, Revision 3, "Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load" (ADAMS Accession No. ML12193A107). More specific guidance related to risk-informed TS changes is provided in SRP Section 16.1, Revision 1, "Risk-Informed Decision Making:
Technical Specifications" (ADAMS Accession No. ML070380228), which includes changes to surveillance test intervals (STls) (i.e., surveillance frequencies) as part of risk-informed decisionmaking. Section 19.2 of the SRP references the same criteria as RG 1.177, Revision 1, and RG 1.174, Revision 2, 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-de pth 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 Commission's Safety Goal Policy Statement.
- The impact of the proposed change should be monitored using performance measurement strategies.
The NRG staff's guidance in NUREG-1431, Revision 4.0, Volumes 1 and 2.
3.0 TECHNICAL EVALUATION
Revising the frequency of a channel calibration, COT, and TADOT 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 shall 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, channel functional test, COT, or TADOT 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, 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 the tested devices. As required by the channel calibration definition, the test would still encompass all devices in the channel required for channel operability.
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 the plant licensing basis would not be invalidated, and other factors. The NRC staff finds these measures acceptable in determining SR extensions.
In addition, Step 16 of Section 4.0 of NEI 04-10 requires an Independent Decisionmaking 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 SR 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. NextEra/FPL is not proposing to change the methodology or the acceptance criteria for extending STls, and NextEra/FPL 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 change to determine an acceptable test frequency for individual steps within instrumentation channel surveillance tests is acceptable, because any extended STls 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 change only affects the frequency of performance
and does not affect the surveillance testing method or acceptance criteria. Therefore, the proposed change is consistent with the surveillance testing requirements of 10 CFR 50.36.
Probabilistic Risk Assessment Acceptability The guidance in RG 1.200 states that the quality of a licensee's 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 will have performed an assessment of the PRA models used to support the approved SFCP that uses NEI 04-1 O 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 NRG-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 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 revision 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 staff's review of the PRA modeling during the review of the application to implement an SFCP that uses NEI 04-10.
NextEra/FPL is using a PRA that was used to support its application that implemented an SFCP that uses NEI 04-10. The amendments will change the capability of NextEra/FPL 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 NextEra/FPL's application to implement an SFCP that uses NEI 04-10 is adequate.
The NRC staff determined that the proposed changes to the TSs meet the standards for TSs in 10 CFR 50.36(b ). The regulations at 10 CFR 50.36 require that TSs 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 TSs, as revised, continue to specify the appropriate SRs for tests and inspections to ensure the necessary quality of affected SSCs is maintained.
Additionally, the changes to the TSs were reviewed and found 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 10 CFR 50.36(b), 10 CFR 50.36(c)(3), and
- 10 CFR 50.36(c)(5) for the reasons discussed above, and thus provide reasonable assurance that adoption of these TSs will have the requisite requirements and controls to operate safely.
Therefore, the NRC staff concludes that the proposed TS changes are acceptable.
3.1 Variations From TSTF-563 NextEra/FPL described variations from TSTF-563 in Section 2.2 of the LAR. NextEra/FPL provided justification for the proposed variations and exceptions. The NRC staff reviewed the justifications and concluded the variations are acceptable because:
- The addition of the phrases "The CHANNEL CALIBRATION" and "means of' to the CHANNEL CALIBRATION definition are editorial in nature. Also, eliminating the phrase "so that the entire channel is calibrated" continues to require the entire channel to be calibrated because the original definition contains the phrase 'The CHANNEL CALIBRATION shall encompass the entire channel," and will remain in the proposed definition. These changes do no change the applicability of TSTF-563, and therefore, are acceptable.
- The addition of the sentences "The TRIP ACTUATING DEVICE OPERATIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps" and "The ANALOG CHANNEL OPERATIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps" are necessary to adopt TSTF-563. This provision was added to the STS in TSTF-205. The addition of this sentence allows the adoption of TSTF-563 and does not change the requirements of the T ADOT and analog COT. These changes do not change the applicability of TSTF-563, and therefore, are acceptable.
- Point Beach and Turkey Point were not licensed to the 10 CFR Part 50, Appendix A GDC, but were licensed to the applicable AEC general design criteria. TSTF-563 references 10 CFR Part 50, Appendix A, GDC 13 and GDC 21. Point Beach UFSAR Section 1.3.3 and Turkey Point UFSAR Section 1.3.4 provide a GDC comparison to the AEC proposed criteria. NextEra/FPL stated that Point Beach and Turkey Point will remain in compliance with applicable AEC design criteria. These differences do not affect the applicability of TSTF-563, and therefore, are acceptable.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the NRC staff notified officials from the States of Wisconsin, New Hampshire, Florida, and the Commonwealth of Massachusetts on January 22, 2020, of the proposed issuance of the amendments. Each official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change SRs.
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 has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding published in the Federal Register on September 10, 2019 (84 FR 47550). 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
Based on the aforementioned considerations, the NRC staff concluded 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 amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: T. Sweat Date: February 10, 2020
ML19357A195 *b memorandum **b email OFFICE NRR/DORL/LPL 1/PM NRR/DORL/LPL 1/LA NRR/DSS/STSB/BC*
NAME JPoole LRonewicz VCusumano DATE 1/23/2020 1/16/2020 11/27/2019 OFFICE OGC-NLO** NRR/DORL/LPL 1/BC NAME KGamin JDanna JPoole DATE 1/28/2020 02/07/2020 02/10/2020