ML20097G391

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Proposed Tech Specs Re Listing of Core Operating Limit Methodologies for Editorial Correction
ML20097G391
Person / Time
Site: Mcguire, McGuire  Duke Energy icon.png
Issue date: 02/13/1996
From:
DUKE POWER CO.
To:
Shared Package
ML20097G386 List:
References
NUDOCS 9602210197
Download: ML20097G391 (6)
v • d • e


Text

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CORE OPERATING LIMITS REPORT The analytical methods us'ad to determine the core operating limits shall be those previously reviewed and approved by NRC in:

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

l 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 (M Proprietary).

t (Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed i Axial Offset Control) and 3.2.2 - Heat Flux Hot Channel Factor (W(Z) i surveillance requirements for F Methodology.)

3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE," March 1987 (M Proprietary).

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

. I

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.)

l S. DPC-NE-2011PA, " Duke Power Company Nuclear Design Methodology for Core l 3

Operating Limits of Westinghouse Reactors," March 1990 (DPC Proprietary).

(Methodology for Specifications 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, 1 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-4 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 l Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

)

7. DPC-NF-201 lA, " Duke Power Company McGuire Nuclear Station Catawba Nuclear . .. .. - ...

Station Nu ' ear Physics Methodology for Reload Design," June 1985-(.DPG-Frgrut=A j (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. 160 (Unit 1)

Amendment No. 142 (Unit 2) 9602210197 960213 DR ADOCK0500g9 y -_._._,m ,,m

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 whien determine the core operating limits)

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,"

November 1992.

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)

11. DPC-NE-2004P-A, " Duke Fower Company McGuire and Catawha Nuclear Stations Core Thermal-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.)

l 13. DPC-2005P-A, " Thermal Hydraulic Statistical Core Design Methodology,"

February 1995 (DPC Proprietary).

l  :

e (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). l i 14. BAW-10162P-A, TAC 03 Fuel Pin Thermal Analysis Computer Code, B&W Fuel Company, November 1989.

4

_ . _ _ _ .(Methodology used for Specification 2.2.'1 - Reactor Trip System Instrumentation setpoints).

15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved l

j by SER dated February, 1994.

l (Used for Specification 2.2.1, Reactor Trip System Instrumentation Setpoints).

l

McGUIRE - UNITS 1 and 2 6-21b Amendment No. 160 (Unit 1)

Amendment No. 142 (Unit 2)

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued)

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 '

i 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 FoMethodology.)
3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL j USING BASH CODE", March 1987, (W Proprietary).

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

4. BAW-10168PA, Rev. 1, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," January,1991 (B&W Proprietary).

4

^

(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). 1 (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 i' 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).

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 l Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) 1

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.)

4 l

McGUIRE - UNIT 1 6-21 Amendment No.

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 the core operating limits)

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 CASM0-3/ SIMULATE-3P,"

November, 1992.

(dethodology for Specification 3.1.1.3-Moderator Temperature Coefficient.)

11. DPC-NE-2004P-A, " Duke Fower Company McGuire and Catawba Nuclear Stations Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (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 FoH(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).

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

(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).

15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated February 1994.

(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 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 - UNIT 1 6-22 Amendment No.

_ _ = _ _ __ - - _. ._. .

j..  !

l .

i .

l ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) ,

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 i Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET sCONTROL 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 F oMethodology.)

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-10168PA, Rev.1, "B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," 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 i Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot l 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 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 i Coefficient, Specification 3.9.1 - RCS and Refueling Canal Boron Concentration, and Specification 3/4.9.12 - Spent Fuel Pool Boron Concentration.)

McGUIRE - UNIT 2 6-21 Amendment No.

. -~

ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT

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

November 1991.

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

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,"

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 Core Thermal-Hydraulic Methodology using VIPRE-01," December 1991 (DPC Proprietary) .

(Methodology for Specifications 2.2.1 - Reactor Tri tion Setpoints, 3.2.1 - Axial Flux Difference (AFD)p,System Instrumenta-and 3.2.3 - Nuclear Enthalpy Rise Hut 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).

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

(Methodology used for Specification 2.2.1 - Reactor Trip System Instru-mentation setpoints).

15. BAW-10183P, Fuel Rod Gas Pressure Criterion, B&W Fuel Company, as approved by SER dated February 1994.

(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 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 - UNIT 2 6-22 Amendment No.

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