NL-13-1232, Revision to License Amendment Request to Allow the Nexus Methodology in the Preparation of the Core Operating Limits Report
| ML13175A352 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/24/2013 |
| From: | Pierce C Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-13-1232 | |
| Download: ML13175A352 (8) | |
Text
Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7872 Fax 205. 992.7601 SOUTHERN'\\
June 24, 2013 COMPANY Docket Nos.: 50-348 50-364 U. S. Nuclear Regulatory Commission NL-13-1232 ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Revision to License Amendment Request to Allow the NEXUS Methodology in The Preparation of the Core Operating Limits Report Ladies and Gentlemen:
By letter dated August 14, 2012 (ML12227A884), and as supplemented by letters dated February 28,2013 (ML13063A291), and April 19, 2013 (ML13112A154),
Southern Nuclear Operating Company (SNC) requested a revision to the Technical Specification (TS) Section 5.6.5, "Core Operating Limits Report (COLR)," to reference and allow the use of Westinghouse WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology," to determine core operating limits. The SNC TS markups contained a Note stating that the NEXUS methodology shall be used commencing Unit 1 Cycle 26 and Unit 2 Cycle 24. Due to uncertainty regarding the NEXUS methodology implementation schedule, SNC requests to revise the Note so that the NEXUS methodology shall be used commencing Unit 1 Cycle 27. The implementation cycle for Unit 2 remains unchanged. contains the revised markup of the proposed TS. Enclosure 2 contains the revised clean typed TS. Changing the commencement cycle for the NEXUS methodology does not change "Significant Hazards Consideration" conclusions in the original August 14, 2012 submittal. As such, the "Significant Hazards Consideration" given in the August 14, 2012 submittal remains valid.
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.
U. S. Nuclear Regulatory Commission NL-13-1232 Page 2 Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.
Respectfully submitted, C. R. Pierce Regulatory Affairs Director Sworn to and subscribed before me this M day of ~,2013.
~~~~~
Notary Public My commission expires: U-6? - ('3 CRP/RMJ
Enclosures:
- 1. Revised Technical Specifications Marked-up Pages
- 2. Revised Technical Specification Clean Typed Pages cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. A. Lynch, Vice President - Farley Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. B. J. Adams, Vice President - Fleet Operations RTYPE: CFA04.054 U. S. Nuclear Regulatorv Commission Mr. V. M. McCree, Regional Administrator Ms. E. A. Brown, NRR Project Manager - Farley Mr. P. K. Niebaum, Senior Resident - Farley Mr. J. R. Sowa, Senior Resident - Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer
Joseph M. Farley Nuclear Plant Revision to License Amendment Request to Allow the NEXUS Methodology in The Preparation of the Core Operati ng Limits Report Revised Technical Specifications Marked-up Pages
5.6 Reporting Requirements 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 3a.
WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998 ('J:J... Proprietary).
3b.
WCAP-12610-P-A, "Vantage+ Fuel Assembly Reference Core Report," April 1995 0!:!... Proprietary).
(Methodology for LCO 3.2.1 - Heat Flux Hot Channel Factor and LCO 3.4.1-RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
3c.
WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)" M.E. Nissley, et aI., January 2005 (Proprietary).
- 4.
WCAP-8745-P-A, "Design Bases for the Thermal Overpower LlT and Thermal Overtemperature LlT Trip Functions," September 1986 (Westinghouse Proprietary)
(Methodology for Overpower LlT and Thermal Overtemperature LlT Trip Functions)
- 5.
WCAP-14750-P-A Revision 1, "RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop PWRs. (Westinghouse Proprietary)
(Methodology for minimum RCS flow determination using the elbow tap measurement.)
6~CAP-11596-P-A, "Qualification of the Phoenix-P/ANC Nuclear Design System for Pressurized Water Reactor Cores," June 1988 (Methodology for LCO 3.9.1 - Boron Concentration.)
- 7.
WCAP-11397-P-A "Revised Thermal Design Procedure," April 1989 (Methodology for LCO 2.1.1-Reactor Core Safety Limits, LCO 3.4.1 RCS Pressure, Temperature and Flow Departure from Nucleate BOiling Limits.)
NOTE -----------------------------------------------
Commencing Unit 1 Cycle 27 and Unit 2 Cycle 24, methods 6b and 6c shall be used in lieu of method 6a.
6b. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code (contin ued)
PARAGON," August 2004 6c. WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS mendment No. +74 (Unit 1)
,--_N_u_c_le_a_rD_a_t_a_M_e_th_o_d_ol_09;::.:Y,-,'_' A_U-::9_u_st_2_0_07___________-',mendment No. +e+ (Unit 2)
Joseph M. Farley Nuclear Plant Revision to License Amendment Request to Allow the NEXUS Methodology in The Preparation of the Core Operating Limits Report Revised Technical Specifications Clean Typed Pages
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued) 3a.
WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998 (W Proprietary).
3b.
WCAP-12610-P-A, "Vantage+ Fuel Assembly Reference Core Report," April 1995 (W Proprietary).
(Methodology for LCO 3.2.1 - Heat Flux Hot Channel Factor and LCO 3.4.1-RCS Pressure, Temperature and Flow Departure from Nucleate Boiling Limits.)
3c.
WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)" M.E. Nissley, et aI., January 2005 (Proprietary).
- 4.
WCAP-8745-P-A, "Design Bases for the Thermal Overpower ~T and Thermal Overtemperature ~T Trip Functions," September 1986 (Westinghouse Proprietary)
(Methodology for Overpower ~T and Thermal Overtemperature ~T Trip Functions)
- 5.
WCAP-14750-P-A Revision 1, "RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop PWRs. (Westinghouse Proprietary)
(Methodology for minimum RCS flow determination using the elbow tap measurement.)
6a.
WCAP-11596-P-A, "Qualification of the Phoenix-P/ANC l\\Juclear Design System for Pressurized Water Reactor Cores," June 1988
N 0 T E ---------------------------------------
Commencing Unit 1 Cycle 27 and Unit 2 Cycle 24, methods 6b and 6c shall be used in lieu of method 6a.
6b.
WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004 6c.
WCAP-16045-P-A, Addendum 1-A, "Qualification of the I\\IEXUS Nuclear Data Methodology," August 2007 (Methodology for LCO 3.9.1 - Boron Concentration.)
(continued)
Farley Units 1 and 2 5.6-4 Amendment No.
(Unit 1)
Amendment No.
(Unit 2)
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
- 7.
WCAP-11397-P-A "Revised Thermal Design Procedure," April 1989 (Methodology for LCO 2.1.1-Reactor Core Safety Limits, LCO 3.4.1 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 mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
5.6.6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
- a.
The reactor coolant system pressure and temperature limits, including heatup and cooldown rates, shall be established and documented in the PTLR for LCO 3.4.3.
- b.
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 NRC letters dated March 31, 1998 and April 3, 1998.
- c.
The PTLR shall be provided to the NRC upon issuance for each reactor 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 shall be reported within 30 days.
Reports on EDG failures shall include a description of the failures, underlying causes, and corrective actions taken per the Emergency Diesel Generator Reliability Monitoring Program.
(continued)
Farley Units 1 and 2 5.6-5 Amendment No.
(Unit 1)
Amendment No.
(Unit 2)
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.8 PAM Report When a report is required by Condition B or F 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 (SG) 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.
Active degradation mechanisms found,
- c.
Nondestructive examination techniques utilized for each degradation mechanism,
- d.
Location, orientation (if linear), and measured sizes (if available) of service induced indications,
- e.
Number of tubes plugged during the inspection outage for each active degradation mechanism,
- f.
Total number and percentage of tubes plugged to date, and
- g.
The results of condition monitoring, including the results of tube pulls and in-situ testing.
5.6.11 Alternate AC (AAC) Source Out of Service Report The NRC shall be notified if the AAC source is out of service for greater than 10 days.
Farley Units 1 and 2 5.6-6 Amendment No.
(Unit 1)
Amendment No.
(Unit 2)