ML23317A207
| ML23317A207 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  (NPF-068, NPF-081) |
| Issue date: | 12/22/2023 |
| From: | John Lamb Plant Licensing Branch II |
| To: | Coleman J Southern Nuclear Operating Co |
| Shared Package | |
| ML23317A200 | List: |
| References | |
| EPID L-2023-LLA-0053 | |
| Download: ML23317A207 (32) | |
Text
December 22, 2023 Ms. Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.
3535 Colonnade Parkway Birmingham, AL 35243
SUBJECT:
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 - ISSUANCE OF AMENDMENT NOS. 223 AND 206, REGARDING REVISION TO TECHNICAL SPECIFICATIONS TO ADOPT TSTF-339-A, RELOCATE TECHNICAL SPECIFICATION PARAMETERS TO THE COLR [CORE OPERATING LIMITS REPORT] CONSISTENT WITH WCAP-14483 (EPID L-2023-LLA-0053)
Dear Ms. Coleman:
The U.S. Nuclear Regulatory Commission (NRC) has issued the enclosed Amendment No.
223 to Renewed Facility Operating License NPF-68 and Amendment No. 206 to Renewed Facility Operating License NPF-81 for the Vogtle Electric Generating Plant (Vogtle), Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated April 11, 2023, as supplemented by letters dated August 11 and October 4, 2023.
The amendments would revise TS 2.1.1, Reactor Coolant Safety Limits, (TS) 3.3.1, Reactor Trip System (RTS) Instrumentation, TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, and TS 5.6.5, Core Operating Limits Report (COLR), to adopt most of the TS and COLR changes described in Appendix A and Appendix B of Westinghouse topical report WCAP-14483-A, to relocate several cycle-specific parameter limits from the TS to the COLR. The amendments follow the guidance of technical specification task force (TSTF) change traveler TSTF-339-A, Revision 2. Along with the parameter relocations, the amendments also modify the Vogtle, Units 1 and 2, TS 5.6.5, to include WCAP-8745-P-A and WCAP-11397-P-A, and to revise the TS applicability for the WCAP-9272-P-A, in the list of the NRC approved methodologies used to develop the cycle-specific COLR. In addition, the amendments revise an error to the TS 3.3.1 depiction of an equation.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's monthly Federal Register notice.
If you have questions, you can contact me at 301-415-3100 or John.Lamb@nrc.gov.
Sincerely,
/RA/
John G. Lamb, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-424 and 50-425
Enclosures:
- 1. Amendment No. 223 to NPF-68
- 2. Amendment No. 206 to NPF-81
- 3. Safety Evaluation for Vogtle cc: Listserv SOUTHERN NUCLEAR OPERATING COMPANY, INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-424 VOGTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 223 Renewed License No. NPF-68
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Vogtle Electric Generating Plant, Unit 1 (the facility) Renewed Facility Operating License No. NPF-68 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated April 11, 2023, as supplemented by letters dated August 11 and October 4, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
J. Coleman 2.
Accordingly, the license is hereby amended by page 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-68 is hereby amended to read as follows:
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 223, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear 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 prior to the beginning of the fuel cycle in the fall 2024.
FOR THE NUCLEAR REGULATORY COMMISSION Shawn Williams, Acting Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. NPF-68 and the Technical Specifications Date of Issuance: December 22, 2023 Shawn A.
Williams Digitally signed by Shawn A. Williams Date: 2023.12.22 09:45:44 -05'00'
SOUTHERN NUCLEAR OPERATING COMPANY, INC.
GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-425 VOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 206 Renewed License No. NPF-81
- 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Vogtle Electric Generating Plant, Unit 2 (the facility) Renewed Facility Operating License No. NPF-81 filed by the Southern Nuclear Operating Company, Inc. (the licensee), acting for itself, Georgia Power Company Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the owners), dated April 11, 2023, as supplemented by letter dated August 11 and October 4, 2023, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as 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 set forth in 10 CFR Chapter I; 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.
2.
Accordingly, the license is hereby amended by page 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-81 is hereby amended to read as follows:
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 206, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear 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 prior to the beginning of the fuel cycle in spring 2025.
FOR THE NUCLEAR REGULATORY COMMISSION Shawn Williams, Acting Chief Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. NPF-81 and the Technical Specifications Date of Issuance: December 22, 2023 Shawn A. Williams Digitally signed by Shawn A.
Williams Date: 2023.12.22 09:46:30 -05'00'
ATTACHMENT VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 TO LICENSE AMENDMENT NO. 223 RENEWED FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 AND TO LICENSE AMENDMENT NO. 206 RENEWED FACILITY OPERATING LICENSE NO. NPF-81 DOCKET NO. 50-425 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Pages Insert Pages License License License No. NPF-68, page 4 License No. NPF-68, page 4 License No. NPF-81, page 3 License No. NPF-81, page 3 TSs TSs 2.0-1 2.0-1 2.0-2 3.3.1-20 3.3.1-20 3.3.1-21 3.3.1-21 3.3.1-22 3.3.1-22 3.4.1-1 3.4.1-1 3.4.1-2 3.4.1-2 5.6-3 5.6-3 5.6-4 5.6-4 5.6-5 5.6-5 5.6-6 5.6-6 5.6-7 5.6-7 Renewed Operating License NPF-68 Amendment No. 223 (1)
Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.
(2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 223, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3)
Southern Nuclear Operating Company shall be capable of establishing containment hydrogen monitoring within 90 minutes of initiating safety injection following a loss of coolant accident.
(4)
Deleted (5)
Deleted (6)
Deleted (7)
Deleted (8)
Deleted (9)
Deleted (10)
Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:
(a)
Fire fighting response strategy with the following elements:
- 1.
Pre-defined coordinated fire response strategy and guidance
- 2.
Assessment of mutual aid fire fighting assets
- 3.
Designated staging areas for equipment and materials
- 4.
Command and control
- 5.
Training and response personnel (b)
Operations to mitigate fuel damage considering the following:
- 1.
Protection and use of personnel assets
- 2.
Communications
- 3.
Minimizing fire spread
- 4.
Procedures for Implementing integrated fire response strategy
- 5.
Identification of readily-available pre-staged equipment
- 6.
Training on integrated fire response strategy Renewed Operating License NPF-81 Amendment No. 206 (2)
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia, pursuant to the Act and 10 CFR Part 50, to possess but not operate the facility at the designated location in Burke County, Georgia, in accordance with the procedures and limitations set forth in this license; (3)
Southern Nuclear, 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)
Southern Nuclear, 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)
Southern Nuclear, 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; (6)
Southern Nuclear, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as my be produced by the operation of the facility authorized herein.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter 1 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 Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 3625.6 megawatts thermal (100 percent power) in accordance with the conditions specified herein.
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 206 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
The Surveillance requirements (SRs) contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment No. 74. The SRs listed below shall be
Vogtle Units 1 and 2 2.0-1 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2) 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded:
2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained greater than or equal to the 95/95 DNBR criterion for the DNB criterion correlations and methodologies specified in Specification 5.6.5.
2.1.1.2 The peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 58°F per 10,000 MWD/MTU of burnup.
2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained 2735 psig.
2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2 If SL 2.1.2 is violated:
2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.
RTS Instrumentation 3.3.1 Vogtle Units 1 and 2 3.3.1-20 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2)
Table 3.3.1-1 (page 7 of 9)
Reactor Trip System Instrumentation Note 1: Overtemperature Delta-T The Allowable Value of each input to the Overtemperature Delta-T function as defined by the equation below shall not exceed its as-left value by more than the following:
(1) 0.5% T span for the T channel (2) 0.5% T span for the Tavg channel (3) 0.5% T span for the pressurizer pressure channel (4) 0.5% T span for the f1(AFD) channel
(AFD) f1 P}
P
{
K3
)
p
(
T s}
6
+
{1 1
T s}
5
+
{1 s}
+
{1 K2 K1 s}
3
+
{1 1
s}
2
+
{1 s}
1
+
{1 0
T T
100 4
Where:
T measured loop specific RCS differential temperature, degrees F T0 indicated loop specific RCS differential at RTP, degrees F 1+1s lead-lag compensator on measured differential temperature 1+2s 1, 2 time constants utilized in lead-lag compensator for differential temperature: 1 =
- seconds, 2 =
- seconds 1
1+3s lag compensator on measured differential temperature 3
time constant utilized in lag compensator for differential temperature,
- seconds K1 fundamental setpoint, *% RTP K2 modifier for temperature, = *% RTP per degree F 1+4s 1+5s lead-lag compensator on dynamic temperature compensation 4, 5 time constants utilized in lead-lag compensator for temperature compensation: 4
- seconds, 5
- seconds T
measured loop specific RCS average temperature, degrees F 1
1+6s lag compensator on measured average temperature 6
time constant utilized in lag compensator for average temperature,
- seconds T
indicated loop specific RCS average temperature at RTP,
- degrees F K3 modifier for pressure, = *% RTP per psig P
measured RCS pressurizer pressure, psig P
reference pressure,
- psig s
Laplace transform variable, inverse seconds
RTS Instrumentation 3.3.1 Vogtle Units 1 and 2 3.3.1-21 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2)
Table 3.3.1-1 (page 8 of 9)
Reactor Trip System Instrumentation Note 1: Overtemperature Delta-T (continued) f1(AFD) modifier for Axial Flux Difference (AFD):
- 1.
for AFD between *% and *%, = 0% RTP
- 2.
for each % AFD is below *%, the trip setpoint shall be reduced by *% RTP
- 3.
for each % AFD is above *%, the trip setpoint shall be reduced by *% RTP (p) The compensated temperature difference T
s}
6
+
1
{
1 T
s}
5
+
1
{
s}
+
1
{
4
shall be no more negative than
- degrees F.
The values denoted with [*] are specified in the COLR.
Note 2: Overpower Delta-T The Allowable Value of each input to the Overpower Delta-T function as defined by the equation below shall not exceed its as-left value by more than the following:
(1) 0.5% T span for the T channel (2) 0.5% T span for the Tavg channel Where:
T measured loop specific RCS differential temperature, degrees F T0 indicated loop specific RCS differential at RTP, degrees F 1+1s lead-lag compensator on measured differential temperature 1+2s 1, 2 time constants utilized in lead-lag compensator for differential temperature: 1 =
- seconds, 2 =
- seconds 1
1+3s lag compensator on measured differential temperature 3
time constant utilized in lag compensator for differential temperature,
- seconds K4 fundamental setpoint, *% RTP K5 modifier for temperature change: *% RTP per degree F for increasing temperature, *% RTP per degree F for decreasing temperature 7s 1+7s rate-lag compensator on dynamic temperature compensation 7
time constant utilized in rate-lag compensator for temperature compensation,
- seconds T
measured loop specific RCS average temperature, degrees F 1
1+6s lag compensator on measured average temperature
RTS Instrumentation 3.3.1 Vogtle Units 1 and 2 3.3.1-22 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2)
Table 3.3.1-1 (page 9 of 9)
Reactor Trip System Instrumentation Note 2: Overpower Delta-T (continued) 6 time constant utilized in lag compensator for average temperature,
indicated loop specific RCS average temperature at RTP,
- degrees F s
Laplace transform variable, inverse seconds f2(AFD) modifier for Axial Flux Difference (AFD), =
- The values denoted with [*] are specified in the COLR.
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Vogtle Units 1 and 2 3.4.1-1 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in the COLR.
APPLICABILITY:
MODE 1.
NOTE---------------------------------------------------
Pressurizer pressure limit does not apply during:
- a.
THERMAL POWER ramp > 5% RTP per minute; or
- b.
THERMAL POWER step > 10% RTP.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more RCS DNB parameters not within limits.
A.1 Restore RCS DNB parameter(s) to within limit.
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B.
RCS total flow rate degraded.
B.1.
Perform SR 3.4.1.4.
7 days C.
Required Action and associated Completion Time not met.
C.1 Be in MODE 2.
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />
RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Vogtle Units 1 and 2 3.4.1-2 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2)
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within the limit specified in the COLR.
In accordance with the Surveillance Frequency Control Program SR 3.4.1.2 Verify RCS average temperature is within the limit specified in the COLR.
In accordance with the Surveillance Frequency Control Program SR 3.4.1.3 Monitor RCS total flow rate for degradation.
In accordance with the Surveillance Frequency Control Program SR 3.4.1.4
NOTE-------------------------------
Not required to be performed until 7 days after 90% RTP.
Verify by precision heat balance that RCS total flow rate is within the limit specified in the COLR.
In accordance with the Surveillance Frequency Control Program
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
Vogtle Units 1 and 2 5.6-3 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2) 5.6.5 Core Operating Limits Report (COLR)
- a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:
SL 2.1.1 Reactor Core Safety Limits LCO 3.1.1 "SHUTDOWN MARGIN" LCO 3.1.3 "Moderator Temperature Coefficient" LCO 3.1.5 "Shutdown Bank Insertion Limits" LCO 3.1.6 "Control Bank Insertion Limits" LCO 3.2.1 "Heat Flux Hot Channel Factor" LCO 3.2.2 "Nuclear Enthalpy Rise Hot Channel Factor" LCO 3.2.3 "Axial Flux Difference" LCO 3.3.1 Reactor Trip System (RTS) Instrumentation LCO 3.4.1 Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) LImits LCO 3.9.1 "Boron Concentration"
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary). (Methodology for Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, and Nuclear Enthalpy Rise Hot Channel Factor, Reactor Trip System Instrumentation, and Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits.)
WCAP-10216-P-A, Revision 1A, "RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," February, 1994 (W Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FQ Methodology).)
WCAP-10266-P-A, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987.
(W Proprietary) (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FQ Methodology).)
WCAP-13749-P-A, Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement, March 1997.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
Vogtle Units 1 and 2 5.6-4 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2) 5.6.5 Core Operating Limits Report (COLR) (continued)
WCAP-16045-P-A, Qualification of the Two-Dimensional Transport Code PARAGON, August 2004 (Methodology for Moderator Temperature Coefficient.)
WCAP-16045-P-A, Addendum 1-A, Qualification of the NEXUS Nuclear Data Methodology, August 2007 (Methodology for Moderator Temperature Coefficient.)
WCAP-12610-P-A, VANTAGE+ Fuel Assembly Reference Core Report, April 1995 (Westinghouse Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FQ Methodology).)
WCAP-12610-P-A & CENPD-404-P-A, Addendum 1-A, Optimized ZIRLOTM, July 2006 (Westinghouse Proprietary). (Methodology for Axial Flux Difference (Relaxed Axial Offset Control) and Heat Flux Hot Channel Factor (W(Z) surveillance requirements for FQ Methodology).)
WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, September 1986 (W Proprietary). (Methodology for Reactor Trip System Instrumentation.)
WCAP-11397-P-A, "Revised Thermal Design Procedure," April 1989 (W Proprietary). (Methodology for Reactor Core Safety Limits and RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
Vogtle Units 1 and 2 5.6-5 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2) 5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a.
RCS pressure and temperature limits for heatup, cooldown, operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:
LCO 3.4.3 "RCS Pressure and Temperature (P/T) Limits"
- b.
The power operated relief valve lift settings required to support the Cold Overpressure Protection Systems (COPS) and the COPS arming temperature shall be established and documented in the PTLR for the following:
LCO 3.4.12 "Cold Overpressure Protection Systems"
- c.
The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1.
WCAP-14040-A, Rev. 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves.
- 2.
WCAP-16142-P, Rev. 1, Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Vogtle Units 1 and 2.
- 3.
The PTLR will contain the complete identification for each of the TS reference Topical Reports used to prepare the PTLR (i.e.,
report number, title, revision, date, and any supplements).
- d.
The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.
5.6.7 EDG Failure Report If an individual emergency diesel generator (EDG) experiences four or more valid failures in the last 25 demands, these failures and any nonvalid failures experienced by that EDG in that time period shall be reported within 30 days.
Reports on EDG failures shall include the information recommended in Regulatory Guide 1.9, Revision 3, Regulatory Position C.4, or existing Regulatory Guide 1.108 reporting requirement.
Reporting Requirements 5.6 5.6 Reporting Requirements (continued)
Vogtle Units 1 and 2 5.6-6 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2) 5.6.8 PAM Report When a Report is required by Condition G or J of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.
5.6.9 Deleted.
5.6.10 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:
- a.
The scope of inspections performed on each SG:
- b.
The nondestructive examination techniques utilized for tubes with increased degradation susceptibility;
- c.
For each degradation mechanism found:
- 1. The nondestructive examination techniques utilized;
- 2. The location, orientation (if linear), measured size (if available), and voltage response for each indication. For tube wear at support structures less than 20 percent through-wall, only the total number of indications needs to be reported;
- 3. A description of the condition monitoring assessment and results, including the margin to the tube integrity performance criteria and comparison with the margin predicted to exist at the inspection by the previous forward-looking tube integrity assessment; and
- 4. The number of tubes plugged during the inspection outage.
- d.
An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including the analysis methodology, inputs, and results;
Reporting Requirements 5.6 5.6 Reporting Requirements Vogtle Units 1 and 2 5.6-7 Amendment No. 223 (Unit 1)
Amendment No. 206 (Unit 2) 5.6.10 Steam Generator Tube Inspection Report (continued)
- e.
The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG;
- f. The results of any SG secondary side inspections;
- g.
The primary to secondary LEAKAGE rate observed in each SG (if it is not practical to assign the LEAKAGE to an individual SG, the entire primary to secondary LEAKAGE should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report;
- h.
The calculated accident induced leakage rate from the portion of the tubes below 15.2 inches from the top of the tubesheet for the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.48 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined; and
- i.
the results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discover and corrective action shall be provided.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 AMENDMENT NO. 223 TO RENEWED FACILITY OPERATING LICENSE NPF-68 AMENDMENT NO. 206 TO RENEWED FACILITY OPERATING LICENSE NPF-81 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
DOCKET NOS. 50-424, AND 50-425
1.0 INTRODUCTION
By letter dated April 11, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23101A159), as supplemented by letter dated August 11 (ML23223A223) and October 4, 2023 (ML23277A311), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) for changes to the technical specifications (TSs) for Vogtle Electric Generating Plant (Vogtle), Units 1 and 2.
The proposed amendments would revise TS 2.1.1, Reactor Coolant Safety Limits, TS 3.3.1, Reactor Trip System (RTS) Instrumentation, TS 3.4.1, Reactor Coolant System (RCS)
Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, and TS 5.6.5, Core Operating Limits Report (COLR), to adopt most of the TS and COLR changes described in Appendix A and Appendix B of Westinghouse topical report WCAP-14483-A, Generic Methodology for Expanded Core Operating Limits Report, January 19, 1999 (ML020430092),
to relocate several cycle-specific parameter limits from the TS to the COLR. The amendments follow the guidance of technical specification task force (TSTF) change traveler TSTF-339-A, Revision 2, Relocate TS parameters to the COLR consistent with WCAP-14483 (ML003723269). Along with the parameter relocations, the amendments also modify the Vogtle, Units 1 and 2, TS 5.6.5, to include WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Overtemperature T Trip Function, dated April 17, 1986 (non-public ML073521507) and WCAP-11397-P-A, Revised Thermal Design Procedure, dated April 30, 1989 (non-public ML080630437), and to revise the TS applicability for the WCAP-9272-P-A (ML19284E585), in the list of the U.S. Nuclear Regulatory Commission (NRC) approved methodologies used to develop the cycle-specific COLR. In addition, the amendments revise an error to the TS 3.3.1 depiction of an equation.
The supplements dated August 11 and October 4, 2023, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commissions original proposed no significant hazards consideration determination as published the Federal Register on June 13, 2023 (88 FR 38551).
1.1 Proposed Changes to Adopt TSTF-339 The licensees (SNC) proposed changes are consistent with NRC staff approved TSTF-339 in that certain TS requirements may be relocated to the COLR. Specifically, the licensee proposed the following changes to adopt TSTF-339:
Changes to TS 2.1.1, Reactor Core SLs [Safety Limits]:
o Relocate TS Figure 2.1.1-1, Reactor Core Safety Limits, to the COLR; o Replace TS 2.1.1 wording that states SLs specified in Figure 2.1.1-1 with limits specified in the COLR; and the following SLs shall not be exceeded:
o Add new SLs TS 2.1.1.1 and TS 2.1.1.2 for departure from nucleate boiling ratio (DNBR) and peak fuel centerline temperature limits, respectively.
Changes to TS 3.3.1, Reactor Trip System Instrumentation:
o Relocate Table 3.3.1-1 Note 1, Overtemperature Delta-T, and Note 2, Overpower Delta-T, trip parameter values from the TS to the COLR.
Changes to TS 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, o Relocate reactor coolant system (RCS) DNB limits from the TS Limiting Conditions for Operation (LCO) and from Surveillance Requirements (SRs) to the COLR.
1.2 Additional Proposed TS Changes In addition to the proposed changes that are consistent with TSTF-339 and described in Section 1.1 of this SE, SNC proposed the following variations.
1.2.1 TS 2.1.1, Reactor Core SLs The licensee, SNC, proposed to revise TS 2.1.1 consistent with TSTF-339, except that the 95/95 DNB criterion correlations are to be determined per the methodologies identified in the Vogtle COLR TS 5.6.5 rather than including the specific value in the Safety Limits Specification.
1.2.2 TS 3.3.1, Reactor Trip System (RTS) Instrumentation The licensee, SNC, stated that TSTF-339 defines Overtemperature Delta-T and Overpower Delta-T functions and terms differently in comparison to the NRC-approved Vogtle TS 3.3.1 Note 1 and Note 2 for Table 3.3.1-1. However, both include terms with values that may vary by
fuel cycle. Thus, the licensee proposed to move cycle-dependent terms to the COLR consistent with the TSTF.
1.2.3 TS 3.3.1, Table 3.3.1-1, Function 7, Overpower T, Note 2 The licensee, SNC, proposed an administrative change to Table 3.3.1-1, Function 7, Overpower T, Note 2 to correct a mis-identified term. Specifically, the denominator of the modifier for the K6 term in the equation for the Overpower T function is revised to include the Laplace transform variable s as a multiplier for the 6 time constant utilized in the lag compensator for average temperature.
1.2.4 TS 3.4.1, RCS Pressure, Temperature, and Flow DNB Limits The licensee, SNC, proposed to relocate the limit for RCS total flow from TS 3.4.1 LCO to the COLR. In addition, the licensee proposed to reformat TS 3.4.1 (LCO and SRs) to identify that the specified parameters shall be within the limits specified in the COLR.
1.2.5 TS 5.6.5, Core Operating Limits Report To reflect the changes described in Sections 1.1 and 1.2 of this SE, SNC proposed to revise TS 5.6.5.a, which references the individual specifications that address core operating limits, and TS 5.6.5.b, which references the analytical methods (Topical Reports) used to determine the core operating limits.
2.0 REGULATORY EVALUATION
2.1 Background
Guidance on the relocation of cycle-specific TS parameters to the COLR is provided to all power reactor licensees and applicants in NRC Generic Letter (GL) 88-16, Removal of Cycle-Specific Parameter Limits from Technical Specifications, (ML031200485). In GL 88-16, the NRC staff stated that license amendments are generally required every refueling outage to update the cycle-specific parameter limits in the TSs. However, methodologies for licensees to determine cycle-specific parameter limits have been reviewed and approved by the NRC staff. As a result, the NRC staffs review of the proposed changes to the TSs to update these parameter limits is primarily limited to the confirmation that the updated limits were calculated by the approved methodology and consistent with the plant-specific safety analysis. The COLR was created to place the NRC-approved methodologies in the TSs and allow licensees to use these methodologies to update the parameters without requiring a change to the TSs.
The NRC staff approved Westinghouse Topical Report WCAP-14483-A as an acceptable method to relocate certain TS requirements to the COLR consistent with GL 88-16. The WCAP-14483-A addresses the relocation of the (1) reactor core safety limits figure (with corresponding revision to the safety limits), (2) Overtemperature T (Delta-T) and Overpower T (Delta-T) setpoint parameter values, and (3) departure from nucleate boiling (DNB) parameter limits for RCS average temperature, RCS flow rate, and pressurizer pressure to the COLR. The NRC staff's SE for WCAP-14483-A (and incorporated into WCAP-14483-A) had no conditions specified on the use of WCAP-14483-A.
In TSTF-339, Revision 2, the NRC staff approved the incorporation of the TS changes identified in WCAP-14483-A into NUREG-1431, Volume 1, Revision 2, Standard Technical
Specifications, Westinghouse Plants, June 2001 (ML011840223). The latest revision of NUREG-1431, Volume 1, is Revision 5, dated September 2021 (ML21259A155). There are no substantive changes between Revision 2 and Revision 5 of NUREG-1431 related to the TSTF-339 impacted cycle-specific specifications.
2.2 Regulatory Requirements and Guidance The NRC staff considered the following regulatory requirements and guidance during its review of the proposed changes.
Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Domestic Licensing of Production and Utilization Facilities, establishes the fundamental regulatory requirements for production and utilization facilities. Section 50.36 of 10 CFR, Technical specifications, provides the content and information that must be included in a facility's TSs. TSs are required to include (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCO; (3) SRs; (4) design features; and (5) administrative controls.
The regulation 10 CFR 50.36(c)(1)(i)(A) states:
Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. Operation must not be resumed until authorized by the Commission.
The licensee shall retain the record of the results of each review until the Commission terminates the license for the reactor, except for nuclear power reactors licensed under § 50.21(b) or § 50.22 of this part. For these reactors, the licensee shall notify the Commission as required by § 50.72 and submit a Licensee Event Report to the Commission as required by § 50.73. Licensees in these cases shall retain the records of the review for a period of three years following issuance of a Licensee Event Report.
The regulation 10 CFR 50.36(c)(1)(ii)(A) states:
Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.
Where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. If, during operation, it is determined that the automatic safety system does not function as required, the licensee shall take appropriate action, which may include shutting down the reactor. The licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. The licensee shall retain the record of the results of each review until the Commission terminates the license for the reactor except for nuclear power reactors licensed under § 50.21(b) or § 50.22 of this part. For these reactors, the licensee shall notify the Commission as required by § 50.72 and submit a Licensee Event Report to the Commission as required by § 50.73. Licensees in these cases shall
retain the records of the review for a period of three years following issuance of a Licensee Event Report.
The regulation 10 CFR 50.36(c)(2)(i) states:
Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met. When a limiting condition for operation of any process step in the system of a fuel reprocessing plant is not met, the licensee shall shut down that part of the operation or follow any remedial action permitted by the technical specifications until the condition can be met. In the case of a nuclear reactor not licensed under § 50.21(b) or § 50.22 of this part or fuel reprocessing plant, the licensee shall notify the Commission, review the matter, and record the results of the review, including the cause of the condition and the basis for corrective action taken to preclude recurrence. The licensee shall retain the record of the results of each review until the Commission terminates the license for the nuclear reactor or the fuel reprocessing plant. In the case of nuclear power reactors licensed under § 50.21(b) or § 50.22, the licensee shall notify the Commission if required by § 50.72 and shall submit a Licensee Event Report to the Commission as required by § 50.73. In this case, licensees shall retain records associated with preparation of a Licensee Event Report for a period of three years following issuance of the report. For events which do not require a Licensee Event Report, the licensee shall retain each record as required by the technical specifications.
The regulation 10 CFR 50.36(c)(3) states:
Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.
The regulation 10 CFR 50.36(c)(5) states:
Administrative controls 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. Each licensee shall submit any reports to the Commission pursuant to approved technical specifications as specified in § 50.4.
The NRC staffs guidance for the review of TSs is in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor]
Edition (SRP), Chapter 16.0, Technical Specifications, Revision 3, dated March 2010 (ML100351425). As described therein, as part of the regulatory standardization effort, the NRC staff has prepared standard technical specifications (STSs) for each of the LWR nuclear designs.
Accordingly, the NRC staffs review includes consideration of whether the proposed changes are consistent with NUREG-1431, as modified by NRC-approved travelers.
The NRC GL 88-16 provides guidance to the licensees for the removal of cycle-dependent variables from the TSs, provided that these values are included in a COLR and are determined with NRC-approved methodologies referenced in the TSs.
The NRC-approved WCAP-8745-P-A describes the bases for the Overpower and Overtemperature T trip functions in Westinghouse reactors, and the analytical methods used to derive the limiting safety system settings for the trips.
NRC-approved WCAP-11397-P-A describes a revised thermal design procedure for predicting the departure from nucleate boiling ratio (DNBR) design limit in Westinghouse pressurized water reactors (PWRs).
3.0 TECHNICAL EVALUATION
3.1 Proposed Changes to Adopt TSTF-339 TSTF-339 relocates STS parameters to the COLR consistent with WCAP-14483-A. TSTF-339 and WCAP-14483-A are only applicable to Westinghouse plant designs. Specifically, TSTF-339 changes NUREG-1431, the STS for Westinghouse designed plants. In accordance with SRP Chapter 16.0, the NRC staff determined that the NUREG-1431 STS changes approved in TSTF-339 are applicable to Vogtle, Units 1 and 2, because Vogtle, Units 1 and 2, are Westinghouse designed plants.
In the NRC staffs approval of Topical Report WCAP-14483-A, the NRC staff concluded that the justification provided to support the proposed TS changes required to expand COLRs associated with Westinghouse plants was acceptable. Specifically, the TS changes included revising STS 2.1.1 to relocate Figure 2.1.1-1 to the COLR and specifying requirements regarding safety limits (DNBR and fuel centerline temperature). Figure 2.1.1-1 provides acceptable operation ranges for three RCS parameters (THERMAL POWER, RCS Highest Loop Average Temperature, and Pressurizer Pressure.) Additional changes to the STS Table 3.3.1-1 and STS 3.4.1 to relocate parameter values to the COLR were also found to be acceptable.
The NRC staff compared SNCs proposed TS changes in Section 1.1 of this SE against the STS changes approved in TSTF-339. The NRC staff finds that SNCs proposed changes to the Vogtle, Units 1 and 2, TSs described in Section 1.1 of this SE are consistent with those the NRC staff found acceptable in TSTF-339, except for identified variations. These variations are described in Section 1.2 of this SE and evaluated in Section 3.2 of this SE.
3.2 Additional Proposed Changes 3.2.1 TS 2.1.1, Reactor Core SLs The licensee, SNC, proposed to revise TS 2.1.1 consistent with TSTF-339, except that the 95/95 DNB criterion correlations are to be determined per the methodologies identified in the Vogtle, Units 1 and 2, COLR TS 5.6.5 rather than including the specific value in the Safety Limits Specification.
The NRC staff determined that SNCs proposed changes ensure that the licensee will continue to maintain DNB criteria in accordance with NRC-approved methodologies which are explicitly referenced for use at Vogtle, Units 1 and 2, TS 5.6.5 prior to implementation. While the change
moves the cycle-dependent parameters and acceptance criteria values from the TSs to the licensee-controlled COLR, that change does not impact any current operational practices or accident analyses of record currently utilized at Vogtle, Units 1 and 2. In addition, the staff notes that SNCs proposed changes to TS 2.1.1 are consistent with changes made in similar license amendments issued for Comanche Peak (ML053180521), Farley (ML013400451), and North Anna (ML052990145).
The NRC staff concludes that the proposed changes to TS 2.1.1 will use the NRC-approved methodology used to derive cycle-specific parameters (THERMAL POWER, RCS highest loop average temperature, and pressurizer pressure) contained in WCAP-9272-P-A and referenced in TS 5.6.5;; therefore, based on the discussion above, the NRC staff finds the proposed TS 2.2.1 changes acceptable.
3.2.2 TS 3.3.1, Reactor Trip System (RTS) Instrumentation The licensee, SNC, stated that TSTF-339 defines Overtemperature Delta-T and Overpower Delta-T functions and terms differently in comparison to the NRC-approved Vogtle, Units 1 and 2, TS 3.3.1 Note 1 and Note 2 for Table 3.3.1-1. As an example, the Vogtle, Units 1 and 2, Overtemperature Delta-T TS refers to axial flux difference as f1 (AFD) where the TSTF-339 refers to f1(I). The NRC staff find these kinds of naming differences acceptable because they are not substantive.
The NRC staff did seek clarification on a proposed change identified in Attachment 1 to the LAR. Specifically, TS 3.3.1, Table 3.3.1-1, Note 1: Overtemperature Delta-T, contains a value for note (p) that the mark-up replaced with an asterisk (*). The mark-up indicates that the values denoted with an asterisk (*) are specified in the COLR. Note (p) specifies that the compensated temperature difference shall be no more negative than 3 degrees Fahrenheit. The 3 degrees value that the LAR mark-up replaced with an asterisk (*), indicates that it should be considered a cycle-specific value. Note (p) is Vogtle, Units 1 and 2, specific, and therefore, was not assessed in TSTF-339.
In a letter dated July 12, 2023 (ML23193A783), the NRC staff issued a request for additional information on treating the value for note (p) as cycle specific. In the licensees response letter, dated August 11, 2023 (ML23223A223, SNC RAI #2 Response), the licensee explained that the Vogtle TS 3.3.1 Over Temperature Delta-T footnote with the 3 degrees value was proposed by a LAR dated October 30, 2001 (ML020160407), and was based on a cycle-specific design analysis. The associated NRC staffs SE dated August 9, 2002 (ML022270088), recognized that the 3 degrees value was based on a cycle-specific design analysis. Based on the information described above, the NRC staff finds that the 3 degrees value in the footnote (note (p)) is cycle-specific, and therefore, appropriate for relocation to the COLR.
The NRC staff finds that Vogtle, Units 1 and 2, Overtemperature Delta-T and Overpower Delta-T functions and terms continue to remain acceptable, because it is currently NRC-approved for Vogtle, Units 1 and 2, and SNC is not proposing any substantive changes to these functions and terms.
3.2.3 TS 3.3.1, Table 3.3.1-1, Function 7, Overpower T, Note 2 The licensee, SNC proposed an administrative change to Table 3.3.1-1, Function 7, Overpower T, Note 2 to correct a mis-identified term. Specifically, the denominator of the modifier for the K6 term in the equation for the Overpower T function is revised to include the Laplace
transform variable s (inverse seconds) as a multiplier for the 6 time constant (seconds) utilized in the lag compensator for average temperature. The NRC staff finds that correcting the mis-identified term as described above is necessary for unit (seconds) cancellation and is consistent with the Vogtle, Units 1 and 2, TS 3.3.1, Table 3.3.1-1, Function 7, Overpower T, Note 2 definition of the lag compensator for average temperature, and therefore, is acceptable.
3.2.4 TS 3.4.1, RCS Pressure, Temperature, and Flow DNB Limits The licensee, SNC, proposed to relocate the limit for RCS total flow from TS 3.4.1 LCO to the COLR. In addition, the licensee proposed to reformat TS 3.4.1 (LCO and SRs) to identify that the specified parameters shall be within the limits specified in the COLR.
Relocate the limit for RCS Total Flow from TS 3.4.1 LCO to the COLR This proposed change is a departure from the TSTF-339, Rev. 2, which retained the requirement both in TS 3.4.1 and the COLR. The rationale for the decision to retain the provision in the TS is described in the NRC staffs SE approving WCAP-14483-A (ML020430092). The SE approving WCAP-14483-A describes that the rationale for retaining the RCS flow rate in TS 3.4.1 was that RCS flow rates were varied based on the steam generator tube plugging levels in the 1990s. The NRC staff wished to ensure that maximum tube plugging levels would be assumed to ensure conservatism in the accident analyses. In the years that have followed, licensees have replaced steam generators and utilized improved chemistry controls that have resulted in much lower levels of tube plugging in the early 2020s than was present in the late 1990s. Furthermore, the NRC has also incorporated performance-based monitoring programs into the Administrative sections of STSs. Two of these applicable programs have been adopted at Vogtle Units 1 and 2: TS 5.5.9 Steam Generator Program and TS 5.6.10 Steam Generator Tube Inspection Report. TS 5.6.10.d states that, An analysis summary of the tube integrity conditions predicted to exist at the next scheduled inspection (the forward-looking tube integrity assessment) relative to the applicable performance criteria, including analysis methodology, inputs, and results. This provision makes it possible for SNC to confirm that tube plugging remains appropriately conservative with respect to the accident analysis on the same cycle-specific basis as the rest of the COLR parameters. The staff determined that this meets the intent of the GL 88-16 and WCAP-14483-A, and therefore, the RCS total flow parameter can be moved to the COLR without need to maintain a maximum tube plugging value in TS 3.4.1, as tube plugging assumptions and changes are addressed by TS 5.6.10.
The NRC staff also questioned whether the RCS total flow value is cycle-specific since the applicant had not made any recent changes to the technical specification value. On July 12, 2023, the NRC staff issued a request for additional information (ML23193A783) seeking clarification on the cycle-specific value for RCS total flow. SNC responded on October 4, 2023 (ML23277A311) and provided information that the cycle-specific value is changing from cycle to cycle based on their precision heat balance methodology at an administratively conservative higher value. SNC also provided additional precedents for consideration where the NRC had approved complete relocation to the COLR for RCS total flow at Arkansas Nuclear One (ML053130314), Calvert Cliffs (ML052720276), Ginna (ML052720231), Palisades (ML042160408), and San Onofre (ML083470091).
Based on the additional information provided by SNC, the NRC staff determined that the RCS total flow parameter meets the GL-88-16 requirement as a cycle specific COLR parameter.
Therefore, the NRC staff finds that the RCS Total Flow parameter is appropriate to remove from TS LCO 3.4.1 and fully relocate to the COLR.
Reformat TS 3.4.1 Regarding SNCs proposed TS 3.4.1 reformatting changes, the licensee identified that repeating the same parameters in the LCO and identifying that the values must be greater than or less than the limits (in the LCO and SRs) is an unnecessary detail. Specifically, SNCs reformatted approach simply requires that RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in the COLR.
In a letter dated July 12, 2023 (ML23193A783), the NRC staff issued a request for additional information on reformatting TS 3.4.1. In SNCs response letter, dated August 11, 2023 (ML23223A223; SNC RAI #3 Response and SNC RAI #4 Response), the licensee explained that the Vogtle, Units 1 and 2, TS 3.4.1 proposed changes were written to be consistent with (1) wording used in other Vogtle, Units 1 and 2, TSs that refer to limits in the COLR; (2) wording used for SNCs other PWR plants; and (3) industry guidance (ML12046A089; sent to NRC for information only) on writing specifications that refer to limits in the COLR. In addition, SNC clarified that items in the COLR, to include relocated inequality expressions and numerical limits, remain subject to the 10 CFR 50.59 change process. SNC further stated, that a change to any of the inequality symbols would occur in association with a change in methodology for determining the limits. Such a change in methodology would require prior NRC approval in accordance with 10 CFR 50.59 if modified after relocation to the COLR.
The NRC staff reviewed SNC's proposed changes and concludes that the affected parameters of pressurizer pressure, RCS average temperature and RCS total flow rate are appropriately controlled in the COLR.. The NRC staff finds SNCs proposed reformatting changes are acceptable, because it does not result in substantive changes to the Vogtle, Units 1 and 2, TS 3.4.1 requirements. The NRC staff also determined that these changes are consistent with guidance provided in GL 88-16 and WCAP-14483-A.
3.2.5 TS 5.6.5, Core Operating Limits Report To reflect the changes described in Section 1.0 of this SE, SNC proposed to revise Vogtle, Units 1 and 2, TS 5.6.5.a, which references the individual specifications that address core operating limits, and Vogtle, Units 1 and 2, TS 5.6.5.b, which references the analytical methods (Topical Reports) used to determine the core operating limits. Specifically, because of implementation of TSTF-339 to relocate the core limit Figure 2.1.1-1, the Overtemperature Delta-T (OTT) and Overpower Delta-T (OPT) setpoint parameter values under TS 3.3.1, and the DNB parameters under TS 3.4.1, to the COLR, Vogtle, Units 1 and 2, TS 5.6.5.a is revised to include Vogle, Units 1 and 2, TSs 2.1.1, 3.3.1, and 3.4.1, respectively. In addition, two Topical Reports, which are the accepted methodology used to calculate these cycle-specific parameters are added to TS 5.6.5.b and one topical report description is revised. The added reports are:
WCAP-8745-P-A, Design Bases for the Thermal Overpower T and Thermal Overtemperature T Trip Functions, September 1986 (Methodology for Reactor Trip System Instrumentation.)
WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989 (Methodology for Reactor Core Safety Limits and RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
The revised report description is (added text is underlined):
WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985, to include: (Methodology for Moderator Temperature Coefficient, Shutdown Bank Insertion Limit, Control Bank Insertion Limits, Nuclear Enthalpy Rise Hot Channel Factor, Reactor Trip System Instrumentation, and Reactor Coolant System Pressure, Temperature, and Flow Departure from Nucleate Boiling Limits.)
The NRC staff finds the changes to Vogtle, Units 1 and 2, TS 5.6.5.a are acceptable because the changes include references to the Vogtle, Units 1 and 2, individual TSs that address core operating limits. In addition, the NRC staff finds the changes to Vogtle, Units 1 and 2, TS 5.6.5.b are acceptable because the changes include references to the NRC-approved analytical methods used to determine Vogtle, Units 1 and 2, core operating limits. The NRC staff also determined that these changes are consistent with NUREG-1431 guidance.
3.3 Technical Summary and Conclusion NRC GL 88-16 allows licensees to remove cycle-dependent parameters from TSs provided that the values of these parameters are included in a COLR and are determined with an NRC-approved methodology which is referenced in the TS. These parameters are moved from TS to the COLR to avoid the need for frequent revision of TS to change the value of those operating limits which cannot be specified to reasonably bound several operating cycles without significant loss of operating flexibility.
Consistent with GL 88-16 guidance, the NRC approved WCAP-14483-A as an acceptable method to further relocate to the COLR: (1) reactor core safety limits figure in TS 2.1.1, replacing it with more specific requirements regarding safety limits (i.e., DNBR and fuel centerline temperature); (2) values of the cycle-specific parameters in the Overtemperature T and Overpower T setpoint functions in Table 3.3.1-1; and (3) the limit values of the RCS DNB parameters for RCS average temperature, RCS flow rate, and pressurizer pressure in TS 3.4.1.
TSTF-339 was approved by the NRC to implement the expanded COLR methodology of WCAP-11483-A. In TSTF-339, the NRC staff approved the incorporation of the TS changes identified in WCAP-14483-A into NUREG-1431.
The NRC staff evaluated SNCs proposed changes to relocate items from the Vogtle, Units 1 and 2, TS to the Vogtle, Units 1 and 2, COLR. Based on the evaluation provided in Section 3 of this SE, the NRC staff determined that the licensees proposed revisions to the Vogtle, Units 1 and 2, TSs are acceptable. The NRC staff also determined that these proposed revisions are consistent with NRC guidance contained in GL 88-16 and are a specific application, with variation, of the generic methodology for expanded core operating limits report approved by the NRC staff in WCAP-14483-A and incorporated into the STS by TSTF-339.
Therefore, because the proposed revisions to the Vogtle, Units 1 and 2, TSs are consistent with guidance, and the NRC staff determined that the proposed variations from guidance are acceptable, the NRC staff finds that the proposed revisions to the Vogtle, Units 1 and 2, TSs are acceptable, and the requirements of 10 CFR 50.36 will continue to be met.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Georgia State officials were notified on October 4, and November 17, 2023, of the proposed issuance of the amendments. On November 17, 2023, the State officials informed the NRC that the State of Georgia has no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. 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 (88 FR 38551, dated June 13, 2023), and there has been no public comment on such finding. Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors: Clint Ashley, Charles Peabody, and William Roggenbrodt.
Date: December 22, 2023
ML23317A207 OFFICE NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/PM NRR/DORL/LPL2-1/LA NRR/DSS/STSB/BC NAME JLamb KGoldstein SMehta (A)
PSahd DATE 11/17/2023 11/27/2023 11/17/2023 11/17/2023 OFFICE NRR/DEX/EICB/BC OGC - NLO NRR/DORL/LPL2-1/BC NRR/DORL/LPL2-1/BC NAME RStattel Aghosh-Naber MMarkley (SWilliams for) JLamb DATE 11/20/2023 12/13/2023 12/22/2023 12/22/2023