ML20073A066
| ML20073A066 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, Catawba, McGuire  |
| Issue date: | 09/08/1994 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20073A061 | List: |
| References | |
| NUDOCS 9409200106 | |
| Download: ML20073A066 (7) | |
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 Nuclear Enthalpy Rise Hot Channel Factor, FI,*ification b"/4.2.3., or F""****,
and Power 6.
Factor Multiplier, MF,"",
limits for Spec 3
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:
l 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 MF 3
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.
a
R
]
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)
~
)
'O l'cGUIRE - UNITS 1 and 2 6-21a' Amendment No.
143 (Unit 1)
Amendment No.
125 (Unit 2)
k l
I I
l l
'[
ATTACHMENT iib Technical Specification Markups Catawba t
i
)
l
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 Methodology.)
o 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
~
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)
7-.
- 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
o a
1 6
I 1
I 4
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 (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.
2 DPC-NE-3002A, "[SAR Chapter 15 System kransient Analysis Methodology," M M M M)\\\\
8.
i November 1991.
I I
(Methodology used in the system thermal-hydraulic analyses which i
determine the core operating limits) 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)