ML13149A354
| ML13149A354 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/17/2013 |
| From: | Ellen Brown Plant Licensing Branch II |
| To: | Pierce C Southern Nuclear Operating Co |
| Brown E NRR/DORL/LPL2-1 | |
| References | |
| TAC ME9244, TAC ME9245 | |
| Download: ML13149A354 (21) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 17, 2013 Mr. C. R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc.
P. O. Box 12951 Bin - 038 Birmingham, AL 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING CHANGES TO NUCLEAR METHODOLOGY REFERENCES (TAC NOS. ME9244 AND ME9245)(NL-12-1226)
Dear Mr. Pierce:
The Nuclear Regulatory Commission has issued the enclosed Amendment No.191 to Renewed Facility Operating License No. NPF-2 and Amendment No.187 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively. This amendment is in response to your application dated August 14, 2012 as supplemented by letters dated February 28, April 19, and June 24,2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12227A884, ML13063A291, ML13112A154, and ML13175A352 respectively). This amendment specifically adds a reference to WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON" and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology" to Technical Specification (TS) 5.6.5, "Core Operating Limits Report [COLR]" (ADAMS Accession No. ML042250345 and ML072570352, respectively). Co"ectively, these reports are referred to as WCAP-16045-P-A and addendum.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance wi" be included in the Commission's biweekly Federal Register notice.
Sincerely, IRA!
Eva Brown, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosures:
- 1. Amendment No. 191 to NPF-2
- 2. Amendment No. 187 to NPF-8
- 3. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 191 Renewed License No. NPF-2
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Southern Nuclear Operating Company, Inc.
(Southern Nuclear), dated August 14, 2012, as supplemented by a letters dated February 28, April 19, and June 24, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ij) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Enclosure 1
- 2
- 2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-2 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 191 , are hereby incorporated in the renewed license.
Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
'~-
Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 17, 2013
ATTACHMENT TO LICENSE AMENDMENT NO. 191 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the License and Appendix 'A Technical Specifications (TSs) with the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert License Page License Page NPF-2, page 4 NPF-2, page 4 TS Page TS Page 5.6-4 5.6-4 5.6-5 5.6-5 5.6-6 5.6-6
-4 (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 191, are hereby incorporated in the renewed license.
Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated.
The removal of these conditions shall be made by an amendment to the renewed license supported by a favorable evaluation by the Commission.
- a. Southern Nuclear shall not operate the reactor in Operational Modes 1 and 2 with less than three reactor coolant pumps in operation.
- b. Deleted per Amendment 13
- c. Deleted per Amendment 2
- d. Deleted per Amendment 2
- e. Deleted per Amendment 152 Deleted per Amendment 2
- f. Deleted per Amendment 158
- g. Southern Nuclear shall maintain a secondary water chemistry monitoring program to inhibit steam generator tube degradation.
This program shall include:
- 1) Identification of a sampling schedule for the critical parameters and control points for these parameters;
- 2) Identification of the procedures used to quantify parameters that are critical to control points;
- 3) Identification of process sampling points;
- 4) A procedure for the recording and management of data; Farley - Unit 1 Renewed License No. NPF-2 Amendment No. 191
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," Aprif 1995 (W Proprietary).
(Methodology for LCO 3.2.1 - Heat Flux Hot Channel Factor and LCO 3.4:I-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 AT and Thermal Overtemperature AT Trip Functions," September 1986 (Westinghouse Proprietary)
(Methodology for Overpower AT and Thermal Overtemperature AT 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 i'iuclear Design System for Pressurized Water Reactor Cores," June 1988
NO TE-------------------------------------
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 NEXUS Nuclear Data Methodology," August 2007 (Methodology for LCO 3.9.1 - Boron Concentration.)
(continued)
Farley Units 1 and 2 5.6-4 Amendment No.19i (Unit 1)
Amendment No.187 (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 midcycle 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 cool down 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 f1uence 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. 191 (Unit 1)
Amendment No. 187 (Unit 2)
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.8 PAM Report When a report is required by Condition Bar 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. 191 (Unit 1)
Amendment No. 187 (Unit 2)
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 187 Renewed License No. NPF-8
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Southern Nuclear Operating Company, Inc.
(Southern Nuclear), dated August 14, 2012, as supplemented by a letters dated February 28, April 19, and June 24, 2013, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Enclosure 2
- 2
- 2. Accordingly, the license is amended by changes to the Technical Specifications, as indicated in the attachment to this license amendment; and paragraph 2. C.(2) of Renewed Facility Operating License No. NPF-8 is hereby amended to read as follows:
(2) Technical Specifications The Technical SpeCifications contained in Appendix A, as revised through Amendment No. 187 , are hereby incorporated in the renewed license.
Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
- 3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: July 17, 2013
ATTACHMENT TO LICENSE AMENDMENT NO. 187 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 DOCKET NO. 50-364 Replace the following pages of the License and Appendix A Technical Specifications (TSs) with the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
Remove Insert License Page License Page NPF-8, page 3 NPF-8, page 3 TS Page TS Page 5.6-4 5.6-4 5.6-5 5.6-5 5.6-6 5.6-6
-3 (2) Alabama Power Company, pursuant to Section 103 of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess but not operate the facility at the designated location in Houston County, Alabama in accordance with the procedures and limitations set forth in this renewed 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; and (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 may be produced by the operation of the facility.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at reactor core power levels not in excess of 2775 megawatts thermal.
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.1S7 , are hereby incorporated in the renewed license.
Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
Farley - Unit 2 Renewed License No. NPF-8 Amendment No. 187
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 0!1. 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.1T and Thermal Overtemperature.1T Trip Functions," September 1986 (Westinghouse Proprietary)
(Methodology for Overpower.1T and Thermal Overtemperature.1T 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 Nuclear Design System for Pressurized Water Reactor Cores," June 1988
N0 TE--------------------------------------
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 NEXUS Nuclear Data Methodology," August 2007 (Methodology for LCO 3.9.1 - Boron Concentration.)
(continued)
Farley Units 1 and 2 5.6-4 Amendment No. 191 (Unit 1)
Amendment No. 187 (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:I-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 midcycle 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. 191 (Unit 1)
Amendment No. 187 (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 (MC) Source Out of Service Report The NRC shall be notified if the MC source is out of service for greater than 10 days.
Farley Units 1 and 2 5.6-6 Amendment No. 191 (U nit 1)
Amendment No. 187 (Unit 2)
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 191 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-2 AND AMENDMENT NO. 187 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-8 SOUTHERN NUCLEAR OPERATING COMPANY, INC.
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-348 AND 50-364
1.0 INTRODUCTION
By letter dated August 14, 2012, as supplemented by letters dated February 28,2013, April 19, and June 24, 2013 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML12227A884, ML13063A291, ML13112A154 and ML13175A352, respectively).
This amendment specifically adds a reference to WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON" and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology" to Technical Specification (TS) 5.6.5, "Core Operating Limits Report [COLR)" (ADAMS Accession No. ML042250345 and ML072570352, respectively). Collectively, these reports are referred to as WCAP-16045-P-Aand addendum.
The supplement dated February 28, April 1, and June 24, 2013, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published in the Federal Register on October 9,2012 (77 FR 61440).
2.0 REGULATORY EVALUATION
In Section 50.36 to Title 10 to the Code of Federal Regulations (10 CFR), the Commission established its regulatory requirements related to the content of the TSs. Consistent with 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings (LSSS), and limiting control settings; (2) LCOs [Limiting Conditions for Operation]; (3) surveillance requirements; (4) design features; and (5) administrative controls.
Enclosure 3
- 2 Section 50.34. "Contents of Applications; Technical Information" of 10 CFR requires that safety analysis reports include analysis and evaluation of the design and performance of structures, systems, and components for the prevention of accidents and the mitigation of the consequences of accidents. As part of the core reload design process, licensees (or vendors) perform reload safety evaluations to ensure that their safety analyses remain bounding for the design cycle. To confirm that the analyses remain bounding, licensees confirm that key inputs to the safety analyses (such as the critical power ratio) are conservative with respect to the current design cycle. If key safety analysis parameters are not bounded, a reanalysis or reevaluation of the affected transients or accidents is performed to ensure that the applicable acceptance criteria are satisfied.
The guidance in NRC Generic Letter (GL) 88-16, "Removal of Cycle-Specific Parameter Limits from Technical Specifications," indicates that it is acceptable for licensees to control reactor physics parameter limits by specifying the calculation methodology. The GL indicates that such parameter limits may be removed from TS and placed in a cycle-specific core operating limit report (COLR). The COLR is defined in the TS, and the Reporting Requirements in the TS require that a COLR be submitted to the NRC each operating cycle, or each time the COLR is revised.
GL 88-16 also recommends that licensees include, in the TS, a list of references for the NRC-approved methodologies, which are used to generate the cycle-specific parameter operating limits.
WCAP-16045-P-A and addendum comprise a nuclear data methodology. As such, there are no directly applicable regulatory requirements. The safety evaluations (SEs) generically approving WCAP-16045-P-A and addendum cite 10 CFR 50.34 as their regulatory basis, insofar as applicants and licensees are required to provide safety analyses. Accurate nuclear analysis is required to ensure that all downstream safety analyses, which may be governed by more explicit regulatory requirements, are appropriate and applicable to a given facility or operating cycle.
These safety analyses are in turn used to establish the LCOs and LSSS contained in the TSs.
PHOENIX-P is described in WCAP-11596-P-A, "Qualification of the PHOENIX-P/ANC Nuclear DeSign System for Pressurized Water Reactor Cores," June 1988, and is the current nuclear desjgn code referenced in the TS 5.6.5. PHOENIX-P will be replaced with PARAGON and NEXUS.
3.0 TECHNICAL EVALUATION
The NRC staff considered the technical and regulatory bases for the generic approvals of the WCAP-16045-P-A and addendum in its present review, in order to establish that the nuclear code suite is appropriate for use at Farley and that the proposed changes to the TS maintained the affected TS consistent with the provisions in GL 88-16.
3.1 PARAGON and NEXUS Code Descriptions The PARAGON code, described in WCAP-16045-P-A, is a neutron transport code, intended to replace PHOENIX-P for use in providing nuclear input data for the Westinghouse proprietary Advanced Nodal Code (ANC), which is a core simulator code. NEXUS (WCAP-16045-P-A, Addendum 1-A) is an improvement to the PARAGON code system. The NRC staff SE approving
- 3 WCAP-16045-P-A, Addendum 1-A, characterizes the change between the two systems as follows:
The primary difference between the PARAGON/ANC code system and the NEXUS/ANC code system is the method of communicating the nuclear data generated by PARAGON to the ANC core simulator. In previous applications, the PARAGON/ANC code system required specific boron letdown cUlVes specified by the user to account for variations in the neutron spectrum as a result of changing boron concentration during the cycle... Westinghouse has proposed a more direct approach to accounting for the spectral changes by parameterizing the cross section output of PARAGON, such that cycle specific boron letdown cUlVes do not need to be provided in the analysis.
WCAP-16045-P-A and addendum are NRC-approved, based on extensive validation and verification. As described in the topical report and addendum, the data set used for validation and verification includes numerous criticality experiments and plant data. The plant data reflect a widely varied set of operating conditions that includes numerous fuel designs, lattice geometries, and burnable absorber loadings in use in the Westinghouse- and Combustion Engineering-designed nuclear steam supply systems. The verification was expanded to include comparison of NEXUS results for critical boron concentration to those obtained using the previous PARAGON methodology. The results showed excellent agreement between NEXUS predictions and available data.
3.2 Applicability to Farley, Units 1 and 2 The NRC staff reviewed information contained in the Updated Final Safety Analysis Report (UFSAR) to verify that the database discussed in Section 3.1 of this SE includes fuel with similar characteristics to that used at Farley. Chapter 4, "Reactor," of the UFSAR indicates that the units use 17x17 Westinghouse LOPAR and VANTAGE 5 fuel with standard, wet annular, and/or integral fuel burnable absorber. All of these design features are represented in the database described above; hence, the NRC staff determined that the NEXUS code system is acceptable for use at Farley.
The only condition or limitation established in the SEs approving PARAGON and NEXUS precludes their use for mixed oxide (MOX) cores. The NRC staff verified that Farley does not use MOX fuel by reviewing TS 4.2, "Reactor Core." TS 4.2.1 states as follows:
The reactor shall contain 157 fuel assemblies. Each assembly shall consist of a matrix of zirconium alloy, zircaloy-4, or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material.
The NRC staff determined, therefore, that a separate licensing action would be required in order to use MOX fuel, so the MOX limitation is inherently satisfied for Farley.
In conclusion, the NRC staff determined that WCAP-16045-P-A and addendum are NRC-approved methods that are applicable to the Farley units. The guidance in GL 88-16 indicates the use of NRC-approved methods to determine core operating limits. Furthermore, the applicability of the generic qualification establishes that nuclear design analyses performed for
-4 Farley using the methods described in WCAP-16045-P-A and addendum will be reasonably accurate, consistent with 10 CFR 50.34 requirements for safety analyses.
3.3 Additional Review Topics As originally proposed, the licensee would have only added reference to WCAP-16045-P-A, Addendum 1-A, without adding a reference to the base topical report. The NRC staff was unable to determine how the NEXUS methodology would be used in determining core operating limits without also using PARAGON. Based on a staff request for additional information expressing this concern, the licensee proposed, by letter dated April 19, 2013, to add reference also to WCAP-16045-P-A. The supplemental letter also included a NOTE, which clarified that WCAP-11596-P-A, the PHOENIX-P/ANC topical report, would no longer be used after implementation of PARAGON and NEXUS. Because the proposed revision added clarity and specificity regarding how the core operating limits would be determined using PARAGON and NEXUS, the NRC staff determined that the proposed supplemental change was acceptable.
Since the addendum to WCAP-16045-P-A brings about a change to the way boron letdown curves are calculated and input into the overall nuclear design method, the staff determined that it was necessary to verify that no changes were being made to the way that post-loss of coolant accident (LOCA) subcriticality and boric acid precipitation behavior were being analyzed. By letter dated February 28,2013, the licensee provided the following supplemental information with respect to the scope of the proposed change to the TS COLR
References:
The proposed change does not affect the inputs of method(s) for ensuring core subcriticality, both short and long-term post-LOCA, thereby precluding the potential for return to power following a large-break LOCA. Since neither the post-LOCA boron source concentration nor heat generation are impacted by the proposed change, the current emergency operating procedure timing for boric acid precipitation and the action time for switching to simultaneous injection will continue to remain valid. Core design specific parameters that are verified each cycle to be conservative with respect to the LOCA inputs, such as Fq, FdH [power peaking factors], and refueling boron concentration, will continue to be calculated using NRC-approved methods.
The NRC staff determined that the supplemental information shown above is acceptable, because it confirms that the proposed implementation of PARAGON and NEXUS will not affect the calculation of post-LOCA boron requirements or emergency procedures to mitigate post-LOCA boric acid precipitation analysis. The NRC staff concluded, based on this information, that the existing post-LOCA analyses for subcriticality and long-term cooling remain applicable.
Based on the following considerations, the NRC staff determined that the proposed TS changes required to replace the PHOENIX-P code system with the PARAGON/NEXUS code system are acceptable: (1) PARAGON and NEXUS are NRC-approved methods, applicable to the Farley units; (2) the proposed TS revisions specifically delineate which topical reports will be used; and (3) the new methodology does not require an update to post-LOCA long-term core cooling or subcriticality analyses. The NRC staff determined, based on these considerations, that the proposed TS revision is consistent with the requirements at 10 CFR 50.34 and the guidance contained in GL 88-16.
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4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the State of Alabama official was notified of the proposed issuance of the amendments. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment relates to changes in recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(10). The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding 77 FR 61440. Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
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 Contributor: Benjamin Parks, NRR Date: July 17, 2013
July 17, 2013 Mr. C. R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc.
P. O. Box 12951 Bin - 038 Birmingham, AL 35201-1295
SUBJECT:
JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING CHANGES TO NUCLEAR METHODOLOGY REFERENCES (TAC NOS. ME9244 AND ME9245)(NL-12-1226)
Dear Mr. Pierce:
The Nuclear Regulatory Commission has issued the enclosed Amendment No.191 to Renewed Facility Operating License No. NPF-2 and Amendment No.187 to Renewed Facility Operating License No. NPF-8 for the Joseph M. Farley Nuclear Plant, Units 1 and 2, respectively. This amendment is in response to your application dated August 14, 2012 as supplemented by letters dated February 28, April 19, and June 24, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12227A884, ML13063A291, ML13112A154, and ML13175A352 respectively). This amendment specifically adds a reference to WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON" and WCAP-16045-P-A, Addendum 1-A, "Qualification of the NEXUS Nuclear Data Methodology" to Technical Specification (TS) 5.6.5, "Core Operating Limits Report [COLR]" (ADAMS Accession No. ML042250345 and ML072570352, respectively). Collectively, these reports are referred to as WCAP-16045-P-A and addendum.
A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.
Sincerely, IRA!
Eva Brown, Senior Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosures:
- 1. Amendment No. 191 to NPF-2
- 2. Amendment No. 187 to NPF-8
- 3. Safety Evaluation cc w/encls: Distribution via Listser DISTRIBUTION:
Public RidsNrrDssSrxb Resource LPL2-1 R/F RidsNrrPMFarley Resource RidsAcrsAcnw_MailCTR Resourc RidsNrrLASFigueroa Resource RidsNrrDirsltsb Resource RidsRgn2MailCenter Resource RidsNrrDorlDpr Resource RidsNrrDorlLpl2-1 Resource RidsOgcMailCenter Resource ADAMS Accession No.: ML13149A354 OFFICE NRR/LPL2-1/PM NRR/LPL2-lILA OGC - NLO NRR/LPL2-lIBC NRR/LPL2-1/PM
- NAME EBrown SFigueroa AGhosh RPascarelli EBrown DATE 6/4/13 6/27/13 06/14/13 07/17/13 07/17/13 OFFICIAL RECORD COpy