ML20095K041
| ML20095K041 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 12/19/1995 |
| From: | Wiens L NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20095K045 | List: |
| References | |
| NUDOCS 9512280069 | |
| Download: ML20095K041 (8) | |
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UNITED STATES i
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NUCLEAR REGULATORY COMMISSION 4
WASHINGTON, D.C. 30666-0001 gs...../
DUKE POWER COMPANY DOCKET NO. 50-369 McGUIRE N4 CLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.160 License No. NPF-9 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility), Facility Operating License No. NPF-9 filed 4
by the Duke Power Company (licensee) dated September 1, 1995, as supplemented October 17 and Hovember 15, 1995, 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 4
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 Comission; 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 Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the comon 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.
9512280069 951219 PDR ADOCK 05000369 P
. 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 Facility Operating License No. NPF-9 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 160, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Pl an.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMISSION
/
N Leonard A. Wiens, Acting Director Project Directorate 11-2 i
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation l
Attachment:
Technical Specification Changes l
Date of Issuance:
December 19, 1995 l
4 4
l 5
i yang y-t UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30666 4 001
- )
DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 142 License No. NPF-17 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility), Facility Operating License No. NPF-17 filed by the Duke Power Company (licensee) dated September 1, 1995, as supplemented October 17 and November 15, 1995, 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; 8.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations 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 Facility Operating License No.
NPF-17 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.
142, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b
Leonard A. Wiens, Acting Director Project Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance:
December 19, 1995
ATTACHMENT TO LICENSE AMENDMENT NO. 160 FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND TO LICENSE AMENDMENT NO. 142 FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
Remove Paaes Insert Paaes 6-21a 6-21a 6-21b 6-21b 6-21c*
- overflow page - no changes i
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ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 4
j The naalytical methods used to determine the core operating limits shall be thos:, previously reviewed and approved by NRC in:
)
1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"
b ly 1985 (W Proprietary).
(Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux i
j Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
2.
WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," June 1983 (W Proprietary).
i j
(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed 5
Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (W(Z) j surveillance requirements for F. Methodology.)
I 3.
WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL j
USING BASH CODE," March 1987 (M Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
l 4.
BAW-10M8P, Rev.1, "B&W Loss-of-Coolant Accident Evaluation Model for Recircul# ting Steam Generator Plants," SER dated January 1991 (B&W i
Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
)
5.
DPC-NE-2011PA, " Duke Power Company Nuclear Design Methodology for Core l
l Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).
k (Methodology for Specifications 2.2.1 - Reactor Trip System i
Instrumentation Setpoints, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
6.
DPC-NE-3001PA, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," November 1991 (DPC Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coeffi-cient, 3.1.3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Insertion Linits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot i
j Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
)
7.
DPC-NF-2010PA, " Duke Power Company McGuire Nuclear Station Catawba fluclear Station Nuclear Physics Methodology for Reload Design," June 1985 (DPC Proprietary).
4 (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 3.9.1 - RCS and Refueling Canal Boron l
Concentration, and Specification 3/4.9.12 - Spent Fuel Pool Baron l
Concentration.)
i McGUIRE - UNITS 1 and 2 6-21a Amendment No. 160 (Unit 1)
Amendment No. 142 (Unit 2) 4
1 2
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 8.
DPC-NE-3002A, "fSAR Chapter 15 System Transient Analysis Methodology,"
November 1991.
(Methodology used in the system thermal-hydraulic analyses which determine i
the core operating limits) 4 9.
DPC-NE-3000P-A, " Thermal-Hydraulic Transient Analysis Methodology," August 1994.
(Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P,"
l November 1992.
s (Methodology for Specification 3.1.1.3 - Moderator Temperature j
Coefficient.)
- 11. DPC-NE-2004P-A, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC Proprietary).
h i
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 1
- Nuclear Enthalpy Rise Hot Channel Factor FoH(X,Y).)
12'.
DPC-NE-200lP-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW fuel," October 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation l
Setpoints.)
i
- 13. DPC-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,"
i February 1995 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, Specification 3.2.1 - Axial Flux Difference, and 3.2.3 -
Nuclear Enthalpy Rise Hot Channel Factor).
14.
BAW-10162P-A, TAC 03 Fuel Pin Thermal Analysis Computer Code, B&W Fuel Company, November 1989.
~~ ~
~(Mesdology ised'fcF Specification ~2.~2~.1 - Reactor Trip Sy: item Instrumentation setpoints).
[
15.
BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated February, 1994.
4 l
(Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).
McGUIRE - UNITS 1 and 2 6-21b Amendment No. 160 (Unit 1)
Amendment No. 142 (Unit 2) i i
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supple-ments thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.
McGUIRE - UNITS 1 and 2 6-21 c -
Amendment No. 160 (Unit 1)
Amendment No. 142 (Unit 2) l
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