ML20217P720
| ML20217P720 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 04/08/1998 |
| From: | Berkow H NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20217P724 | List: |
| References | |
| NUDOCS 9804100164 | |
| Download: ML20217P720 (10) | |
Text
g UNITED STATES g
j NUCLEAR REGULATORY COMMISSION e
WASHINGTON, D.C. =as -t
]
%, * *,***,/
DUKE ENERGY CORPORATION DOCKET NO. 50-369 McGUIRE NUCLEAR STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.178 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 by the Duke Energy Corporation (licensee) dated December 17,1997, complies with the standards
)
and requirements of the Atomic Energy Act of 1954, as amended (the Act), and i
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; 1
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the i
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.
i 9804100164 980408 I
PDR ADOCK 05000369 i
P PDR i
L 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-g is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.178, are hereby incorporated into this license. The licensee 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 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION H
N. Berk i ctor Project Directorate ll Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date ofissuance: April 8,1998 i
A
ATTACHMENT TO LICENSE AMENDMENT NO. ' 178 FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO.50-36g i
i l
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 Insert i
6-21 6-21 6-22 6-22 1
u 1
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 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 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).-
(Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control)
J and 3.2.2 - Heat Flux Hot Channel Factor (W(Z) surveillance requirements for Fo Methodology.)
3.
WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
4.
BAW-10168P-A, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," Rev.1 SER dated January 1991; Rev. 2, SER dated August 22,1996; i
Rev. 3, SER dated June 15,1994. (B&W Proprietary).
j (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
5.
DPC-NE-2011PA, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).
(Methodology for Specification 2.2.1 - Reactor Trip System Instrumentation Setpoints, j
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.)
i 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 Coefficient, 3.1.3.5 -
Shutdown Rod insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux l
Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot l
Channel Factor.)
7.
DPC-NE-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nucle ar Physics Methodology for Reload Design," June 1985 (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3/4.9.12 - Spent Fuel Pool Boron Concentration.)
McGUIRE - UNIT 1 6-21 Amendment No.178
[T ADMINISTRATIVE CONTROLS l1i'
'~
CORE OPERATING LIMITS REPORT 8.
DPC-NE-3002A, through Rev 2, "FSAR Chapter 15 System Transient Analysis Methodology," SER dated April 26,1996.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits)
' 9.
DPC-NE-3000P-A, Rev.1, " Thermal-Hydraulic Transient Analysis Methodology," SER dated December 27,1995.
i (Modeling used in the system thermal-hydraulic analyses) 10.
DPC-NE-1004A, Rev.1, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P,"
SER dated April 26,1996.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
11.
DPC-NE-2004P-A, Rev.1, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," SER dated February 20,1997 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor FAH(X,Y).)
12.
DPC-NE-2001P-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 Setpoints.)
13.
DPC-NE-2005P-A, Rev.1, " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7,1996 (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.
DPC-NE-2008P-A, " Fuel Mechanical Reload Analysis Methodology Using TACO 3," SER dated April 3,1995 (DPC Proprietary).
(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).
15.
BAW-10183P-A, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated July,1995.
J (Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).
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 LIMFTS REPORT, including any mid-cycle revisions or supplements 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.
L McGUlRE - UNIT 1 6-22 Amendment No.178
fmtg k 4
UNITED STATES
'g j
NUCLEAR REGULATORY COMMISSION wAsMmoroN. o.c. sonesian l
DUKE ENERGY CORPORATION DOCKET NO. 50-370 McGUlRE NUCI ::AR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 160 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 Energy Corporation (licensee) dated December 17,1997, 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 hesith and safety of %e public, and (ii) that such activities will be conducted in compliance with t%
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 i
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 satisfed.
I
E
! - 2.
Accordingly, the license is hereby amende'.i 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:
(2) Technical Specifications I
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.
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 REGUI.ATORY COMMISSION j
nf 1
rt N. Berkow,l Director l
He 1
Project Directorate 11-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date ofissuance: April 8, 1998 l-l
ATTACHMENT TO LICENSE AMENDMENT NO.160 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 verticallines indicating the areas of change.
Bamo.ya insert 6-21 6-21 l
6 22 6-22
t ADMINISTRATIVE CONTROLS l
CORE OPERATING LIMITS REPORT The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 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 Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot i
Channel Factor.)
2.
' WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ '
SURVEILLANCE TECHNICAL SPECIFICATION", June 1983 (W Proprietary).
l (Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (W(Z) surveillance requirements for Fo Methodology.)
3.
WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987 (W Proprietary).
l (Methodology for Specificat'on 3.2.2 - Heat Flux Hot Channel Factor.)
4.
BAW-10168P-A,'"B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam l
Generator Plants," Rev.1, SER dated January 1991; Rev. 2, SER dated August 22,1996; Rev. 3, SER dated June 15,1994 (B&W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
l S.
DPC-NE 2011PA,
- Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).
j (Methodology for Specification 2.2.1 - Reactor Trip Oy:; tem Instrumentation Setpoints, 3.1.3.5 - Shutdown Rod insertion Limits, 3.1.3.6 - Control Bank Inscrtiori Limits,3.2.1 -
Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Entha'py Rise Hot Channel Factor.)
l B.
DPC-NE-3001PA, " Multidimensional Reactor Transients and Safety Analysis Physics i-Parameter Methodology," November 1991 (DPC Proprietary).
(Methodology for Specification 3.1.1.3 - Moderator Temper 9ture Coefficient, 3.1.3.5 -
Shutdown Rod insertion Limits, 3.1.3.6 - Control Bank Insertion Limits, 3.2.1 - Axial Flux i
Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
7.
DPC-NE-2010A, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient, Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3/4.9.12 - 6 pent Fuel Pool Boron Concentration.)
McGUIRE - UNIT 2 6-21 Amendment No.160
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORI 8.
DPC-NE-3002A, Through Rev. 2, "FSAR Chapter 15 System Transient Analysis Methodology," SER dated April 23,1996.
(Methodology used in the system thermal-hydraulic analyses which determine the core operating limits) 9.
DPC-NE-3000P-A, Rev.1, " Thermal-Hydraulic Transient Analysis Methodology," SER dated December 27,1995.
(Modeling used in the system thermal-hydraulic unalyses) 10.
DPC-NE-1004A, Rev.1, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P,"
~- ~
SER dated April 26,1996.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
11.
DPC-NE-2004P-A, Rev.1, " Duke Power Company McGuire and Catawba Nuclear Stations Core Thermal 'Aydraulic Methodology using VIPRE-01," SER dated February 20,1997 (DPC Propriccary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumenta-tion Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor FAH(X,Y).)
12.
DPC-NE-2001P-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 Setpoints.)
13.
DPC-NE-2005P-A, Rev.1, " Thermal Hydraulic Statistical Core Design Methodology," SER dated November 7,1996 (DPC Proprietary).
(Methodology for Specif; cation 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.
DPC-NE-2008P-A, " Fuel Mechanical Reload Analysis Methodology Using TACO 3," SER dated April 3,1995 (DPC Proprietary).
(Methodology used for Specification 2.2.1 - Reactor Trip System instru-mentation setpoints).
15.
BAW-10183P-A, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated July,1995.
(Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).
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 supplements thereto, sha'J be provided upon iswance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
McGUIRE - UNIT 2 6-22 Amendment No.160