ML20029E948
| ML20029E948 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 04/20/2020 |
| From: | Ellen Brown Plant Licensing Branch II |
| To: | Moul D Florida Power & Light Co |
| Brown E | |
| References | |
| EPID L-2019-LLA-0126 | |
| Download: ML20029E948 (23) | |
Text
April 20, 2020 Mr. Don Moul Vice President, Nuclear Division and Chief Nuclear Officer Florida Power & Light Company Mail Stop: NT3/JW 15430 Endeavor Drive Jupiter, FL 33478
SUBJECT:
TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 - ISSUANCE OF AMENDMENT NOS. 292 AND 285 CONCERNING MODIFICATION OF REACTOR TRIP SYSTEM TURBINE TRIP INSTRUMENTATION REQUIREMENTS (EPID L-2019-LLA-0126)
Dear Mr. Moul:
The U.S. Nuclear Regulatory Commission (Commission) has issued the enclosed Amendment No. 292 to Subsequent Renewed Facility Operating License No. DPR-31 and Amendment No. 285 to Subsequent Renewed Facility Operating License No. DPR-41 for Turkey Point Nuclear Generating Unit Nos. 3 and 4, respectively. The amendments consist of changes to the technical specifications (TS) in response to your application dated June 13, 2019.
The amendments revise the TS reactor trip system turbine trip instrumentation requirements to align with the reactor trip 10 percent pressure permissive interlock. Additionally, the amendments resolve two non-conservative conditions associated with a potential loss of instrument function during TS testing.
A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions biweekly Federal Register notice.
Sincerely,
/RA/
Eva A. Brown, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-250 and 50-251
Enclosures:
- 1. Amendment No. 292 to DPR-31
- 2. Amendment No. 285 to DPR-41
- 3. Safety Evaluation cc: Listserv
FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-250 TURKEY POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 292 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 June 13, 2019, 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 Title 10 of the Code of Federal Regulations (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 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. 292, 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.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Undine Shoop, Chief Plant Licensing Branch II-2 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: April 20, 2020
FLORIDA POWER & LIGHT COMPANY DOCKET NO. 50-251 TURKEY POINT NUCLEAR GENERATING UNIT NO. 4 AMENDMENT TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE Amendment No. 285 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 June 13, 2019, 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 Title 10 of the Code of Federal Regulations (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 3.B 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. 285, 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.
- 3.
This license amendment is effective as of its date of issuance and shall be implemented within 90 days.
FOR THE NUCLEAR REGULATORY COMMISSION
/RA/
Undine Shoop, Chief Plant Licensing Branch II-2 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: April 20, 2020
ATTACHMENT TO LICENSE AMENDMENT NOS. 292 AND 285 TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NOS. DPR-31 AND DPR-41 DOCKET NOS. 50-250 AND 50-251 Replace page 3 of Subsequent Renewed Facility Operating License No. DPR-31 with the attached page 3. The revised page is identified by amendment number and contains a marginal line indicating the area of change.
Replace page 3 of Subsequent Renewed Facility Operating License No. DPR-41 with the attached page 3. The revised page is identified by amendment number and contains a marginal line 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 areas of change.
Remove Insert 3/4 3-3 3/4 3-3 3/4 3-7 3/4 3-7 3/4 3-9 3/4 3-9 3/4 3-11 3/4 3-11 3/4 3-22A 3/4 3-22A
3 Subsequent Renewed License No. DPR-31 Amendment No. 292 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. 292, 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 licensees 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
3 Subsequent Renewed License No. DPR-41 Amendment No. 285 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. 285, 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 licensees 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
TABLE 3.3-1 (Continued)
REACTOR TRIP SYSTEM INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO.
OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION
- 11. Steam Generator Water Level--Low-Low 3/stm. gen.
2/stm. gen.
2/stm. gen.
1, 2 6
- 12. Steam Generator Water Level--
Low Coincident With Steam/
Feedwater Flow Mismatch 2 stm. gen.
level and 2 stm./feed-water flow mismatch in each stm. gen.
1 stm. gen.
level coin-cident with 1 stm./feed-water flow mismatch in same stm.
gen.
1 stm. gen.
level and 2 stm./feed-water flow mismatch in same stm. gen.
or 2 stm. gen.
level and 1 stm./feedwater flow mismatch in same stm.
gen.
1, 2 6
- 13. Undervoltage--4.16 KV Busses A and B (Above P-7) 2/bus 1/bus on both busses 2/bus 1
12
- 14. Underfrequency-Trip of Reactor Coolant Pump Breaker(s) Open (Above P-7) 2/bus 1 to trip RCPs***
2/bus 1
11
- 15. Turbine Trip (Above P-7) a.
Emergency Trip Header Pressure b.
Turbine Stop Valve Closure 3
2 2
2 2
2 1 (Above P-7) 1 (Above P-7) 12 12 TURKEY POINT - UNITS 3 & 4 3/4 3-3 AMENDMENT NOS. 292 A N D 285
TURKEY POINT - UNITS 3 & 4 3/4 3-7 AMENDMENT NOS. 292 AND 285 TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued)
ACTION 11 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 12 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ACTUATION LOGIC TEST provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 13 - With the number of OPERABLE channels one less than the Total number of channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION ANALOG CHANNEL OPERATIONAL TEST TRIP ACTUATING DEVICE OPERATIONAL TEST ACTUATION LOGIC TEST MODES FOR WHICH SURVEILLANCE IS REQUIRED
- 12. Steam Generator Water Level--Low Coincident with Steam/Feedwater Flow Mismatch SFCP SFCP(a), (b)
SFCP(a), (b)
N.A.
N.A.
1, 2
- 13. Undervoltage - 4.16 kV Busses A and B N.A.
SFCP N.A.
N.A.
N.A.
1
- 14. Underfrequency - Trip of Reactor Coolant Pump Breakers(s) Open N.A.
SFCP N.A.
N.A.
N.A.
1
- 15. Turbine Trip a.
Emergency Trip Header Pressure b.
Turbine Stop Valve Closure N.A.
N.A.
SFCP(a), (b)
SFCP N.A.
N.A.
Prior to P-7 (10)
Prior to P-7 (10)
N.A.
N.A.
1****
1****
- 16. Safety Injection Input from ESF N.A.
N.A.
N.A.
SFCP N.A.
1, 2
- 17. Reactor Trip System Interlocks a.
Intermediate Range Neutron Flux, P-6 b.
Low Power Reactor Trips Block, P-7 (includes P-10 input and Turbine Inlet Pressure) c.
Power Range Neutron Flux, P-8 N.A.
N.A.
N.A.
SFCP(4)
SFCP(4)
SFCP(4)
N.A.
N.A.
N.A.
N.A.
N.A.
2**
1 1
TURKEY POINT - UNITS 3 & 4 3/4 3-9 AMENDMENT NOS. 292 285 AND
TURKEY POINT - UNITS 3 & 4 3/4 3-11 AMENDMENT NOS. 292 AND 285 TABLE 4.3-1 (Continued)
TABLE NOTATIONS When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.
Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
Above P-7 (Low Power Reactor Trips Block Interlock) Setpoint.
(a)
If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.
(b)
The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTS) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTS are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (field settings) to confirm channel performance. The NTS and methodologies used to determine the as-found and the as-left tolerances are specified in UFSAR Section 7.2.
(1)
If not performed in previous 31 days.
(2)
Comparison of calorimetric to excore power level indication above 15% of RATED THERMAL POWER (RTP). Adjust excore channel gains consistent with calorimetric power level if the absolute difference is greater than 2%. Below 70% RTP, downward adjustments of NIS excore channel gains to match a lower calorimetric power level are not required. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(3)
Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(4)
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5)
This table Notation number is not used.
(6)
Incore-Excore Calibration, above 75% of RATED THERMAL POWER (RTP). If the quarterly surveillance requirement coincides with sustained operation between 30% and 75% of RTP, calibration shall be performed at this lower power level. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.
(7)
Each train shall be tested in accordance with the Surveillance Frequency Control Program.
(8)
DELETED (9)
Quarterly surveillance in MODES 3*, 4*, and 5* shall also include verification that permissive P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window. Quarterly surveillance shall include verification of the High Flux at Shutdown Alarm Setpoint of 1/2 decade above the existing count rate.
(10)
Required whenever Unit has been in MODE 3 if not performed within previous 31 days. Setpoint verification is not applicable.
(11)
The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the OPERABILITY of the undervoltage and shunt trip attachment of the Reactor Trip Breakers.
TURKEY POINT - UNITS 3 & 4 3/4 3-22A AMENDMENT NOS. 292 AND 285 TABLE 3.3-2 (Continued)
TABLE NOTATION (Continued)
ACTION 24A -
With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, within 7 days restore the inoperable channel to OPERABLE status or place the Control Room Emergency Ventilation System in the recirculation mode.
ACTION 24B -
With the number of OPERABLE channels two less than the Minimum Channels OPERABLE requirement, either:
1.
Immediately place the Control Room Emergency Ventilation System in the recirculation mode with BOTH Control Room emergency recirculation fans operating, OR 2.
- a. Immediately place the Control Room Emergency Ventilation System in the recirculation mode with ONE Control Room emergency recirculating fan operating, AND b.
Restore at least one inoperable channel to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. If this ACTION applies to both Units simultaneously, then be in at least HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ACTION 25 -
With number of OPERABLE channels one less than the Total number of channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in accordance with the Risk Informed Completion Time Program.
ACTION 26 -
With one channel inoperable, operation may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST or TRIP ACTUATING DEVICE OPERATIONAL TEST provided the inoperable channel is place in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in accordance with the Risk Informed Completion Time Program.
ACTION 27 -
With one channel inoperable, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or in accordance with the Risk Informed Completion Time Program, or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION AMENDMENT NO. 292 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO. 285 TO SUBSEQUENT RENEWED FACILITY OPERATING LICENSE NO. DPR-41 FLORIDA POWER & LIGHT COMPANY TURKEY POINT NUCLEAR GENERATING UNIT NOS. 3 AND 4 DOCKET NOS. 50-250 AND 50-251
1.0 INTRODUCTION
By letter dated June 13, 2019 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML19170A094), Florida Power & Light Company (the licensee) requested changes to the Technical Specifications (TSs) for Turkey Point Nuclear Generating Units 3 and 4 (Turkey Point), which are contained in Appendix A of Subsequent Renewed Facility Operating License Nos. DPR-31 and DPR-41, respectively. The licensee proposed to revise the TS reactor trip system (RTS) turbine trip instrumentation requirements to align with the reactor trip 10 percent pressure permissive interlock (P-7). Additionally, the request supports resolution of two non-conservative conditions associated with a potential loss of instrument function during TS-required testing.
2.0 REGULATORY EVALUATION
2.1 Regulatory Requirements The regulation under paragraph 50.36(b) to Title 10 of the Code of Federal Regulations (10 CFR) requires that:
[e]ach license authorizing operation of a... utilization facility... 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. The Commission may include such additional technical specifications as the Commission finds appropriate.
The categories of items that the TSs must include are listed in 10 CFR 50.36(c). In accordance with 10 CFR 50.36(c)(2), TSs must include limiting conditions for operation (LCOs), which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When LCOs are not met, the licensee must shut down the reactor or follow any remedial action permitted by the TSs until the condition can be met.
Section 50.36(c)(3) of 10 CFR defines surveillance requirements (SRs) 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 Turkey Point construction permits were issued in 1967 based, in part, on the review of principal design criteria for water-cooled nuclear power plants. On February 20, 1971, the Atomic Energy Commission published an amendment to 10 CFR Part 50, Licensing of Production and Utilization Facilities, which added Appendix A, General Design Criteria for Nuclear Power Plants (GDC). The regulation became effective on May 21, 1971, and is not applicable to plants with construction permits issued prior to May 21, 1971. Therefore, Turkey Point Unit Nos. 3 and 4 are not required to comply with current Appendix A to 10 CFR Part 50.
Section 3.1.2 of the Turkey Point Updated Final Safety Analysis Report (UFSAR) defines the principal design criteria to which the plant was licensed.
The 1967 proposed GDC 19 states that protection systems shall be designed for high functional reliability and inservice testability necessary to avoid undue risk to the health and safety of the public. This license amendment request (LAR) affects when surveillance testing of the RTS is required to be conducted to provide assurance of system operability during required plant conditions. The 1967 proposed GDC 20 states that redundancy and independence designed into protection systems shall be sufficient to assure that no single failure on removal from service of any component or channel of such a system will result in loss of the protection function. The proposed amendments address these criteria by changing the authorization conditions for placing inoperable instrument channels out of service for channels that use a two-of-three coincidence logic for safety function actuation.
2.2.
Regulatory Guidance The NRC staff reviewed Volume 1, Specifications, and Volume 2, Bases, of NUREG-1431, Standard Technical Specifications, Westinghouse Plants, Revision 4.0 (STS) (ADAMS Accession Nos. ML12100A222 and ML12100A228, respectively), to verify that the proposed changes do not alter the format and structure of the TSs in a way that could affect the specified safety functions of the affected TSs structure, systems, and components.
2.3.
System Description
The RTS is designed to trip the reactor to prevent core damage caused by departure from nucleate boiling and to preserve reactor coolant system (RCS) integrity during limiting fault conditions. The RTS monitors neutron flux; RCS pressure, temperature and flow; pressurizer water level; steam generator water level; feedwater flow; and turbine generator operational status. The RTS provides the primary trip functions for the over-power T (change in temperature), over-temperature T, and nuclear overpower reactor trips.
The RTS provides additional reactor trip functions such as the high and the low pressurizer pressure trips, safety injection actuation, turbine trip, etc., which serve as a backup to the primary trip functions for specific accident conditions and mechanical failures. Certain power range reactor trip channels are automatically bypassed at low power levels where they are not required to assure safety. Nuclear source range and intermediate range trips provide protection at subcritical or low power operation and are manually bypassed at higher power levels.
The engineered safeguards feature actuation system (ESFAS) senses process variables in the reactor coolant, steam, reactor containment, and auxiliary systems, and actuates the engineered safety features (ESFs) to ensure that predetermined limits will not be exceeded.
The ESFAS will actuate the safety injection, containment isolation, emergency containment cooling, and/or the containment spray system, depending upon the type and severity of the plant condition.
The RTS and ESFAS actuation logic circuitry uses permissive interlocks to enable protection system functions that are necessary to assure safety. One of these permissive interlocks is the low power reactor trips block, or the P-7 interlock, which is set at approximately 10 percent rated thermal power (RTP). The following reactor trips are enabled only when reactor power level is above the P-7 permissive setting:
Pressurizer low-pressure reactor trip Pressurizer high-level reactor trip 4 kilovolt (kV) bus A and B under-voltage reactor trip Reactor coolant pump (RCP) under-frequency reactor trip RCP open breaker (two-out-of-three pumps) reactor trip Reactor coolant loop low-flow (two-out-of-three loops) reactor trip Turbine trip Upon decreasing reactor power, the above-listed trips are automatically blocked when reactor power is below the P-7 permissive setting.
The reactor trip on turbine trip protective feature anticipates the loss of heat removal capability of the secondary system following a turbine trip to minimize the pressure and temperature transient on the reactor. The reactor trip on turbine trip is actuated when two-out-of-three emergency trip header oil pressure sensors indicate pressure below a preset value established by safety analyses or two-out-of-two turbine stop valve position sensors indicate both valves being closed. Following a reactor trip on turbine trip, sensible heat stored in the reactor coolant is removed without actuating the steam generator code safety valves by means of controlled steam bypass to the condenser and by feedwater injection to the steam generators in order to reach RCS no-load conditions.
For Turkey Point, the reactor trip on turbine trip is enabled when reactor power level is above the P-7 permissive setting (approximately 10 percent RTP). Conversely, the reactor trip on turbine trip is automatically blocked when power level is below the P-7 reset setting where intermediate range neutron flux permissives enable low power protective features and where the reactor control system and the steam dump system provide reactor protection at zero power conditions.
The reactor trip on turbine trip function is only credited in the excess feedwater flow and reduction in feedwater enthalpy (Section 14.1.7 of the UFSAR) accident analyses. The sequence of events (Table 14.1.7-1 of the UFSAR) and transient responses (Figures 14.1.7-1
- 14.1.7-3 of the UFSAR) showed that a reactor trip on turbine trip occurs 42.9 seconds into the transient, while the reactor power is far above 10 percent RTP, as permitted by P-7. In the event of a turbine trip from full power without an immediate reactor trip, protection circuits associated with coolant conditions directly tied to core limits will actuate to prevent core damage. Therefore, reactor trip on turbine trip circuitry should satisfy the applicable protection system standards for separation, redundancy, testability, and reliability.
2.4 Licensees Proposed Changes As discussed in Section 2.2 of the enclosure to the LAR, the licensee proposed the following changes to the TSs.
TS Change 1 The proposed change would modify the applicable modes of TS 3.3.1, Table 3.3-1, Reactor Trip System Instrumentation, to require operability of the emergency trip header pressure and turbine stop valve closure instrument channels (designated Functional Unit (FU) 15.a and FU 15.b, respectively) from Mode 1 to Mode 1 at or above the 10 percent RTP (P-7) permissive power level.
TS Change 2 The proposed change would modify the SRs of TS 3.3.1, Table 4.3-1, Reactor Trip System Instrumentation Surveillance Requirements, for performing trip actuating device operational test (TADOT) of the FU 15.a and FU 15.b channels from startup to require performance prior to reaching the P-7 permissive power level.
TS Change 3 The proposed change would modify the modes for which a surveillance is required column in TS 3.3.1, Table 4.3-1, by adding a new table notation denoted by four asterisks (****) to specify Mode 1 above the P-7 permissive.
TS Change 4 The proposed change would delete Note 1 from the FU 15.a and FU 15.b channels in TS 3.3.1, Table 4.3-1, and add to Note 10 a statement requiring TADOT of the FU 15.a and FU 15.b channels whenever the unit has been in Mode 3 if the TADOT was not performed within the previous 31 days.
TS Change 5 The proposed change to Action 13 of TS 3.3.1, Table 3.3-1, would remove the authorization to place an inoperable FU-5, Over-temperature-T; FU-6, Overpower T; or FU-9, Pressurizer Level-High, instrument channel in bypass status for the purpose of allowing performance of a Digital Channel Operational Test (DCOT).
TS Change 6 The proposed change to Action 25, Table 3.3-2, Engineered Safety Features Actuation System Instrumentation, of TS 3.3.2 would remove the authorization to place an inoperable FU-1.f, Safety Injection - Steam Line Flow-High Coincident with Tavg-Low, or FU-4.d, Steam Line Isolation - Steam Line Flow-High Coincident with Tavg-Low, instrument channel in bypass status to allow performance of a subsequent required DCOT.
3.0 TECHNICAL EVALUATION
The NRC staffs review focused on determining whether the proposed changes to the TSs provide acceptable remedial actions in accordance with 10 CFR 50.36 in conditions where the turbine trip RTS safety functions are no longer bypassed during TS SR testing. Additionally, the NRC staff reviewed whether the proposed changes maintain the specified safety functions for the affected structures, systems, and components, as designed, consistent with the applicable 1967 proposed GDC.
3.1 Proposed Changes to the TS RTS Instrumentation TS Change 1 In Section 2.2.1 of the enclosure to the LAR, the licensee indicated that the proposed change is intended to modify the applicable modes of TS Table 3.3-1 to require operability of the FU 15.a, Emergency Trip Header Pressure, and FU 15.b, Turbine Stop Valve Closure, instrument channels from Mode 1 to Mode 1 above the P-7 permissive power level in Table 3.3-1.
The NRC staff reviewed the applicable 1967 proposed GDC, including their description of the design intended in the Turkey Point UFSAR, and compared the proposed TS changes to TS 3.3.1, Table 3.3-1, with the corresponding functions of 16.a and 16.b in TS Table 3.3.1-1 of NUREG-1431. As discussed previously, the turbine trip - reactor trip signal is intended to protect the RCS from a thermal transient (overpressure or overtemperature) when the energy removal from the coolant abruptly diminishes. The NRC staff finds that the proposed TS changes are consistent with these design requirements and NUREG-1431 TS format, since these turbine trip functions are only needed to be operable in Mode 1 at power levels above the power range neutron flux interlock.
The NRC staff finds these changes acceptable because the trip functions FU 15.a and FU 15.b are disabled and are not required to be operable for conditions where power levels are at or below the P-7 permissive, and because the operability requirements for these functions are retained for power levels above the P-7 permissive. Therefore, the ability to perform the intended safety function in the required operating modes is maintained.
TS Change 2 In Section 2.2.2 of the enclosure to the LAR, the licensee indicated that the proposed change modifies the SRs of TS 3.3.1, Table 4.3-1, for performing trip actuating device operational test (TADOT) of the FU 15.a and FU 15.b channels from startup to require performance prior to the P-7 permissive power level.
The NRC staff reviewed the applicable 1967 proposed GDC, including their description of the design intended in the Turkey Point UFSAR, and compared the proposed TS changes to TS 3.3.1, Table 4.3-1, with the corresponding SR in NUREG-1431. As discussed previously, the turbine trip - reactor trip signal is intended to protect the RCS from a thermal transient (overpressure or overtemperature) when the energy removal from the coolant abruptly diminishes. The NRC staff finds that the proposed TS changes are consistent with these design requirements since requiring the performance of TADOT prior to the power range neutron flux interlock power level, if not performed within the previous 31 days will ensure the interlock is above the P-7 permissive.
The NRC staff finds these changes acceptable because the trip functions FU 15.a and FU 15.b are disabled and are not required to be operable for conditions where power levels are below the P-7 permissive and because the operability requirements for these functions as established by performance of the TADOT will be retained for power levels above the P-7 permissive.
Therefore, the ability to perform the intended safety function in the required operating modes is maintained and the LCO can be determined to be met.
TS Change 3 In Section 2.2.2 of the enclosure to the LAR, the licensee indicated that the proposed change modifies the surveillance required modes with a new table notation denoted by four asterisks
(****) to specify Mode 1 above the P-7 permissive.
This modified note is consistent with the changes being made to Table 4.3-1 to exclude operability requirements for functions FU 15.a, and FU 15.b. The NRC staff finds these changes acceptable because the trip functions FU 15.a and FU 15.b are disabled and are not required to be operable for conditions where power levels are at or below the P-7 permissive and because the operability requirements for these functions as established by performance of the TADOT will be retained for power levels above the P-7 permissive. Therefore, the ability to perform the intended safety function in the required operating modes is maintained and the LCO can be determined to be met.
TS Change 4 In Section 2.2.2 of the enclosure to the LAR, the licensee indicated that the proposed change deletes TS Table 4.3-1, Note 1, from the FU 15.a and FU 15.b channels and adds to Note 10 a statement requiring TADOT of the FU 15.a and FU 15.b channels whenever the unit has been in Mode 3 if the TADOT was not performed within the previous 31 days.
The NRC staff reviewed the applicable 1967 proposed GDC and their description of design and intent in the Turkey Point UFSAR and compared the proposed TS changes to TS 3.3.1, Table 4.3-1, with the corresponding SR 3.3.1.15 of NUREG-1431. As discussed previously, the turbine trip - reactor trip signal is intended to protect the RCS from a thermal transient (overpressure or overtemperature) when the energy removal from the coolant abruptly diminishes. The NRC staff finds that the proposed TS changes are consistent with these design requirements since requiring the performance of TADOT prior to exceeding the power range neutron flux interlock power level, if not performed within the previous 31 days, will ensure the interlock is above the P-7 permissive. Additionally, the NRC staff finds that the revised Turkey Point SRs for performing TADOT are consistent with STS 3.3.1.15, which contains a similarly formatted note.
The NRC staff finds that the revised notes in TS Table 4.3-1 are acceptable because the trip functions FU 15.a and FU 15.b are disabled and are not required to be operable for conditions where power levels are at or below the P-7 permissive and because the operability requirements for these functions, as established by performance of the TADOT, will be established prior to power levels exceeding the P-7 permissive. Therefore, the ability to perform the intended safety function in the required operating modes is maintained, and the LCOs can be determined to be met.
3.2 Proposed Changes to Remove Non-Conservative TS Actions TS Change 5 In Section 2.2.3 of the enclosure to the LAR, the licensee indicated that the proposed change to Action 13 of TS 3.3.1, Table 3.3-1, would remove the authorization to place an inoperable FU-5, Over-temperature-T; FU-6, Overpower T; or FU-9, Pressurizer Level-High, instrument channel in bypass status for the purpose of allowing performance of a digital channel operational test (DCOT). To accomplish this, the licensee proposes to delete the following sentence from Action 13:
[f]or subsequent required DIGITAL CHANNEL OPERATIONAL TESTS the inoperable channel may be placed in bypass status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The functions that this change affects - FU-5, FU-6, and FU-9, are functions that are composed of three channels with a two-of-three channel coincidence logic. Because of this, when one of the three channels is placed in a bypass state, the resulting coincidence logic is reduced to two-of-two, which could result in a loss of function during DCOT performance. The licensee established interim measures to prevent bypassing of FU-5, FU-6, or FU-9 channels during DCOT in accordance with NRC Administrative Letter 98-10, Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety (ADAMS Accession No. ML031110108). The licensee states that these interim measures will remain in effect until this amendments are implemented.
The NRC staff finds that an amendment authorizing the removal of the TS authorization to place FU-5, FU-6, and FU-9 channels in bypass for DCOT supersedes the interim measures that are currently in place with respect to those channels. The TS change that removes this bypass authorization is acceptable because the licensee will no longer be able to perform DCOT testing while a channel of functions FU-5, FU-6, or FU-9 is inoperable. This eliminates the possibility of incurring a loss of function during performance of these tests. Therefore, the ability of these reactor trip instrument channels to perform their intended safety functions in the required operating modes is maintained.
TS Change 6 In Section 2.2.4 of the enclosure to the LAR, the licensee proposes to change Action 25 of TS 3.3.2, Table 3.3-2, by removing the authorization to place an inoperable FU-1.f or FU-4.d instrument channel in bypass status to allow performance of a subsequent required DCOT. To accomplish this, the licensee proposes to delete the following sentence from Action 25:
For subsequent required DIGITAL CHANNEL OPERATIONAL TESTS the inoperable channel may be placed in bypass status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
The affected functions, FU-1.f and FU-4.d, are composed of one channel per RCS loop and thus three total channels with a two-of-three (RCS loops) channel coincidence logic. Because of this, if one of the three channels is placed in a bypass state, the resulting coincidence logic is reduced to two-of-two coincidence logic, which could result in a loss of function during DCOT performance. The licensee established interim measures to prevent bypassing of FU-1.f, or FU-4.d channels during DCOT in accordance with NRC Administrative Letter 98-10, and the licensee states that these interim measures will remain in effect until these amendments are implemented.
The NRC staff finds that an amendment authorizing the removal of the sentence that allows the license to place FU-1.f and FU-4.d channels in bypass for DCOT supersedes the interim measures that are currently in place. The TS change that removes this bypass authorization is acceptable because the licensee will no longer be able to perform DCOT testing while a channel of functions FU-1.f or FU-4.d is inoperable. This eliminates the possibility of incurring a loss of function during performance of these tests. The NRC staff also finds that removal of TS authorization to place FU-1.f and FU-4.d channels in bypass for DCOT will supersede the interim measures that are currently in place. Therefore, the ability of these reactor trip instrument channels to perform their intended safety function in the required operating modes is maintained.
3.3 NRC Technical Conclusion The NRC staff determines that the proposed changes to the RTS and ESFAS actuation instrumentation maintain the necessary SRs relating to the plant protection system reliability and testability requirements. Furthermore, the NRC staff finds that redundancy and independence characteristics of the RTS and ESFAS protection systems remain sufficient to assure that no single failure will result in loss of the protection function and that removal from service of any component or channel of the RTS or ESFAS does not result in loss of the required minimum redundancy. The revised TSs, therefore, remain compliant with the criteria of the 1967 proposed GDC 19 and 20. Additionally, the NRC staff finds that the revised TSs provide acceptable actions for turbine trip functions in situations where specified LCOs are not met. Therefore, the NRC staff finds that the proposed changes to the TSs continue to ensure that RTS satisfactorily prevents core damage caused by departure from nucleate boiling, preserves RCS integrity during limiting fault conditions, and that the associated ESF actuations will continue to ensure that predetermined limits will not be exceeded.
The NRC staff finds that the proposed changes to the TS LCOs meet the requirements of 10 CFR 50.36(c)(2) because the TSs will continue to list the lowest functional capability or performance levels of equipment required for safe operation of the facility, and if the LCOs are not met, appropriate remedial actions are defined in the TSs until the condition can be met.
Additionally, the NRC staff finds that the proposed changes to the TS SRs will assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the applicable TS LCOs will be met.
4.0 STATE CONSULTATION
In accordance with the Commissions regulations, the NRC staff notified the State of Florida official (Ms. Cynthia Becker, M.P.H., Chief of the Bureau of Radiation Control, Florida Department of Health) on January 29, 2020, of the proposed issuance of these amendments.
The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change requirements with respect to the use of facility components located within the restricted area as defined in 10 CFR Part 20 and changes 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, which was published in the Federal Register on July 30, 2019 (84 FR 36968), that the amendments involve no significant hazards consideration, 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 aforementioned considerations, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: R. Stattel S. Peng M. Hamm Date: April 20, 2020
- by memorandum
- by e-mail OFFICE DORL/LPL2-2/PM DORL/LPL2-2/LA DEX/EICB/BC*
DSS/SNSB/BC(A)**
NAME EBrown BAbeywickrama MWaters JBorromeo DATE 2/6/2020 2/3/2020 12/16/2019 1/21/2020 OFFICE DSS/STSB/BC**
OGC - NLO w/edits** DORL/LPL2-2/BC DORL/LPL2-2/PM NAME VCusumano MYoung UShoop EBrown DATE 1/24/2020 4/15/2020 4/18/2020 4/20/2020