ML20093L286
| ML20093L286 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 10/17/1995 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20093L266 | List: |
| References | |
| NUDOCS 9510250284 | |
| Download: ML20093L286 (4) | |
Text
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t ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT Th'e 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) 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).
1 (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
4.
BAW-10168P, Rev.1, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," SER dated January 1991 (B&W Proprietary).
(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, 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.)
DPC-NE-3001PA, " Multidimensional Reactor Transients and Safety) Analysis 6.
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 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.)
7.
DPC-NF-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 - Spent Fuel Pool Boron Concentration.)
McGUIRE - UNITS 1 AND 2 6-21a Amendment No.
(Unit 1) 9510250284 951017 DR ADOCK 05000369 PDR
7 7
ADMINISTRATIVE CONTROLS l
~ CORE' OPERATING LIMITS REPORT 3
8.
DPC-NE-3002A, "FSAR Chapter 15 System Transient Analysis Methodology,"
November 1991.
(Methodology used in the s l
the core operating limits)ystem themal-hydraulic analyses which determine 9.
DPC-NE-3000P-A, " Thermal-Hydraulic Transient Analysis Methodology," August-1994.
t (Modeling used in the system thermal-hydraulic analyses)
- 10. DPC-NE-1004A, " Nuclear Design Methodology Using CASMO-3/ SIMULATE-3P,"
j November 1992.
(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)
- 11. DPC-NE-2004P-A, " Duke Power Company McGuire and Catawba Nuclear Stations l
Core Themal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation 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-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,"
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).
(Methodology used for Specification 2.2.1 - Reactor Trip System
)
Instrumentation setpoints).
i
- 15. BAW-10183P, Fuel Rod Gas Pressure Critericn, B&W Fuel Company, May 1994.
Setpoints) pecification 2.2.1, Reactor Trip System Instrumentation (Used for S J
l McGUIRE - UNITS 1 AND 2 6-21b Amendment No.
Unit 1 Amendment No.
Unit 2
ADMINISTRATIVE CONTROLS CORE OP.ERATING LIMITS REPORT (Continued) lb.
- Accumulator and Refueling Water Storage Tank boron concentration limits for Specifications 3/4.5.1 and 3.4.5.4.
- 11. Reactor Coolant System and refueling canal boron concentration limits for Specification 3/4.9.1.
J
- 12. Standby Makeup Pump water supply boron concentration limits of Specification 4.7.13.3.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in:
l 1.
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,"
July 1985 (W Proprietary).
(Methodology for Specification 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).
J (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).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
4.
BAW-10168P Rev.1, "B&W Loss-of-Coolant Accident Evaluation Model for i
Recirculating Steam Generator Plants," SER dated January 1991 (B&W Proprietary)
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
CATAWBA - UNITS 1 & 2 6-19 Amendment No.
(Unit 1)
Amendment No.
(Unit 2)
gNINISTRATIVECONTROLS
(
CORE 0PERATING LIMITS REPORT (Continued) 5.
DPC-NE-2011P-A, " Duke Power Company Nuclear Design Methodology for Core Operating Limits of Westinghouse Reactors," March,1990 (DPC Proprietary).
(Methodology for Specifications 2.2.1 - Reactor Trip System Instrumentation Setpoints, 3.3.3.5 - 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 j
Hot Channel Factor.)
6.
DPC-NE-3001P-A, " 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 Limi's, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot c
Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)
7.
DPC-NF-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
(
i Coefficient, Specification 4.7.13.3 - Standby Makeup Pump Water Supply Boron Concentration, and Specification 3.9.1 - RCS and Refueling Canal Bnron Concentration, and Specification 3.9.12 - Spent Fuel Pool Boron Concentration.)
8.
.DPC-NE-3002A, "FSAR Chapter 15 System Transient Analysis Methodology,"
November 1991.
(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," November 1931.
(Modeling used in the system thermal-hydraulic analyses)
I i
CATAWBA - UNITS 1 & 2 6-19a Amendment No.
(Unit 1)
Amendment No.
(Unit 2)
-