ML20073A066

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Proposed Tech Specs Re Core Operating Limits Repts
ML20073A066
Person / Time
Site: Catawba, McGuire, Mcguire  Duke Energy icon.png
Issue date: 09/08/1994
From:
DUKE POWER CO.
To:
Shared Package
ML20073A061 List:
References
NUDOCS 9409200106
Download: ML20073A066 (7)
v • d • e


Text

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-kOMitilSTRATIVECONTROLS  !

-[0RE OPERATING' LIMITS REPORT  !

6.9.1.9 Core operating limits .shall be established and documented in the CORE f OPERATING LIMITS REPORT before each reload cycle or any remaining part of a- l reload cycle for the following: "

-1. Moderator Temperature Coefficient BOL and E0L limits and 300 ppm _ [

surveillance limit for Specification 3/4.1.1.3, -

2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, ]

-3. Control Bank Insertion Limits for Specification 3/4.1.3.6, l

4. Axial Flux Difference limits, target band *, and APL""* for j Specification 3/4.2.1,  !
5. Heat Flux Hot Channel Factor, F""', K(Z), W(Z)**, APL""** and W(Z),t** For l Specification 3/4.2.2, and
6. ** F""****, and Power Nuclear Enthalpy Factor Multiplier, MFRise 3 ,"", Hot limitsChannel for Spec Factor, FI,*ification b"/4.2.3., or
7. Overtemperature and Overpower Delta T setpoint parameter values for Specification 2.2.1.
8. Boric Acid Storage System and Refueling Water Storage Tank volume and boron concentration limits for Specifications 3/4.1.2.5 and 3/4.1.2.6.
9. Accumulator-and Refueling Water Storage Tank boron concentration limits

.for Specification 3/4.5.1 and 3/4.5.5.

The analytical methods used to determine the coro _ operating limits shall- be

those.previously reviewed and approved by NRC in:

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

July 1985 (H Proprietary). Ok /I (Methodology for Specifications 3.1.1.3 - Moderator Temperature M nab II3 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 . DN Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

  • Reference 5 is not applicable to target band and APL" .

References 4 and 5 are not applicable to W(Z), APL"", and W(Z),t.

"* Reference 1isnotapplicabletoFI,. -

  • "- Reference 5 is not applicable to F[and MF3 McGUIRE . UNITS 1 and 2 6-21 Amendment No., 143 (Unit 1) 9409200106 940908 PDR ADOCK 05000369 P .PDR
10. Reactor Coolant System and refueling canal boron concentration ,

limits for Specification 3/4.9.1.

11. Spent fuel pool boron concentration limits for Specification 3/4.9.12.

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g ADMINISTRAT4VE CONTR0lS 1

CORE OPERATING LIMITS REPORT

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 FQ 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 Recirculating Steam Generator Plants," September 1989 (B&W Proprietary).

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

5. DPC-NE-20llP, " 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, 3.1.3.5 - 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-3001P, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," March 1991 (DPC Proprietary).

I (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 ' actor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

7. DPC-NE-2010P, " Duke Power Company McGuire Nuclear Station Catawba Nuclear Station Nuclear Physics Methodology for Reload Design," April 1984 (DPC Proprietary).

(Methodology for Specification 3.1.1.3 - Moderator Temperature CoefficientpfEClfkAT/0MJ.9.1- (RCS M,b P4fdCUAJG CAM 4L. 00W ggg) -

AM9 SAscjpyf/0Ul}k.4.rt.,-59WRML Po0L 8dhvJ Ca]WWM0

d. DPC-NE-3002, "FSAR Ch' apter 15 System Transient Analysis Methodology,"

August 1991.

(Methedology used in the system thermal-hydraulic analyses which determine the cora operating limits)

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, 'O l'cGUIRE - UNITS 1 and 2 6-21a' Amendment No. 143 (Unit 1)  !

Amendment No. 125 (Unit 2) l

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ATTACHMENT iib Technical Specification Markups Catawba t i

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ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (continued)

10. Accumulator and Refueling Water Storage Tank boron concentration limits for Specifications 3/4.5.1 and 3/4.5.4.

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,"

4Y July 1985 (M Proprietary). ,

(Methodology for Specifications 3.1.1.3 - Moderator Temperature ATTAdh Coefficient, 3.1.3.5 - Shutdown Bank Insertion Litait, 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 F oMethodology.)

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

(Methodology for Specification 3.2.2 - Heat Flux ilot 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-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, 3.1.3.5 - 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-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

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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.)

CATAWBA - UNITS 1 & 2 6-19 Amendment No.115 (Unit 1)

Amendment No.109 (Unit 2)

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11. Reactor Coolant System and refueling canal boron concentration limits for Specirication 3/4.9.1 12 Standby Makeup Pump water supply boron concentration limit of Specification 4.7.13.3 i

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T ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) (

7. DPC-NF-2010P-A, " Duke Power Company McGuire Nuclear Station Catawba i Nuclear Station Nuclear Physics Methodology for Reload Design," June 1985 l (DPC Proprietary). j (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefi1cient& SPCc1fi<PriorJ 1.1.13,3.Sermaby' Hgcop Pu/19 W A W 5 j g g jf g gg,eff CoM \tMfRJrn0Al AM .SMcme.anor).3,9. RCS AM f6faEM2 6 C#^]AL 0%
8. DPC-NE-3002A, "[SAR Chapter 15 System kransient Analysis Methodology," M M M M)\\i November 1991. I I

(Methodology used in the system thermal-hydraulic analyses which i determine the core operating limits) l

9. DPC-NE-3000P-A, Rev.1, " Thermal-Hydraulic Transient Analysis Methodology," November 1991.

(Modeling used in the system thermal-hydraulic analyses)

10. DPC-NE-1004A, " 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 Trip System Instrumentation Setpoints, 3.2.1 - Axial Flux Difference (AFD), and 3.P. 3 - Nuclear Enthalpy Rise Hot Channel Factor FSi (X,Y).)

12. DPC-NE-2001P-A, Rev.1, " Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel," October 1990 (DPC Proprietary).

i (Methodology 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 in accordance with 10 CFR 50.4.

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CATAWBA - 1; NITS 1 & 2 6-19a Amendment No. 120 (Unit 1)

Amendment No. 114 (Unit 2)

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