ML20093L286

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Proposed Tech Specs Listings of Core Operating Limit Methodologies
ML20093L286
Person / Time
Site: Mcguire, Catawba, McGuire  Duke Energy icon.png
Issue date: 10/17/1995
From:
DUKE POWER CO.
To:
Shared Package
ML20093L266 List:
References
NUDOCS 9510250284
Download: ML20093L286 (4)
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Text

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  • 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 l Recirculating Steam Generator Plants," SER dated January 1991 (B&W l 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.)

6.

DPC-NE-3001PA, " Multidimensional Physics Parameter Methodology," NovemberReactor Transients 1991 (DPC Proprietaryand

. Safety) Analysis l

(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 DR 951017 ADOCK 05000369 PDR

7 7 l

ADMINISTRATIVE CONTROLS l

~ CORE' OPERATING LIMITS REPORT ,

3

8. DPC-NE-3002A, "FSAR Chapter 15 System Transient Analysis Methodology,"

November 1991. l l

(Methodology used in the s l the core operating limits)ystem themal-hydraulic analyses which determine l

9. DPC-NE-3000P-A, " Thermal-Hydraulic Transient Analysis Methodology," August- l 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 l Proprietary) . l (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).)  !

l

12. DPC-NE-2001P-A, Rev. 1, " Fuel Mechanical Reload Analysis Methodology for I 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).

14. BAW-10162P-A, TAC 03 Fuel Pin Themal Analysis Computer Code, B&W Fuel Company, November 1989.

(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.

(Used for S Setpoints) pecification 2.2.1, Reactor Trip System Instrumentation I

J l

l McGUIRE - UNITS 1 AND 2 6-21b Amendment No. Unit 1 Amendment No. Unit 2

I 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

1. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," l 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  !

i 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 I 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 l Proprietary)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

l l

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, l 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, c 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 ( 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)

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