ML20034E752
| ML20034E752 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 02/17/1993 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20034E749 | List: |
| References | |
| NUDOCS 9303010303 | |
| Download: ML20034E752 (1) | |
Text
M i% - C I
ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT
[
l h
4.
BAW-10168P, Rev. 1, "B&W Loss-of-Coolant Accident Evaluation Model for J
e Recirculating Steam Generator Plants," September,1989 (B&W Proprietary).
l 4
i W
t (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)
l-
=
5.
OPC-NE-2011P, " Duke Power Company Nuclear Design Methodology for Core
@j Operating Limits of Westinghouse Reactors," March, 1990 (DPC Proprietary).
y e
6 d
i g
(Methodologv for Specification 3.1.3.5 - Shutdown Rod Insertion Limits, V
3.1.3.5 - control Bank Insertion Limits, 3.2.1 - Axial Flux Difference, o
)
3.2.2 Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise g
(
Hot. Channel Factor.)
o e
b 7
6.
DPC-NE-3001P, " Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology," March, 1991 (DPC Proprietary).
e
.F (Methodology for Specification 3.1.1.3 - Moderator Temperature Coeffi-f
}
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.)
3--
O e
~j i
7.
DPC-NE-2010P, " Duke Power Company McGurie Nuclear Station Catawba Nuclear J
-j, Station Nuclear Physics Methodology for Reload Design," April,1984 j
(DPC. Proprietary).
I (
(Methodology for Specification 3.1.1.3 - Moderator Temperature g
Coefficient.)
O g
8.
DPC-NE-3002, "FSAR Chapter 15 System Transient Analysis Methodology,"
h August 1991.
5 3
(Methodology used in the system thermal-hydraulic analyses which determine d
the core operating limits)
~~',
3 c' I-
~2
~ d 9.
DPC-NE-3000, Rev. 1, " Thermal-Hydraulic Transient Analysis Methodology,"
o-May 1989.
d odeling used in the sy ++- thermal-hydraulic analyses)
O h The core operating limits shall be determined so that all applicable limits 3'f 7-e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS (limits, nuclear limits such as shutdown margin, and transient and accident s
- jn analysis limits) of the safety analysis are met.
o z
( 't The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or i
e u
3 -4' supplements thereto, shall be provided upon issuance, for each reload cycle, p ~~7. g to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
,o 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.
i McGUIRE - UNITS 1 and 2 6-21a Amendment No. 1 Unit 1) l Amendment No.
(Unit 2) i 1
9303010303 930217 PDR ADOCK 05000369 i
P PDR