ML20211Q498
| ML20211Q498 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/10/1999 |
| From: | TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20211Q495 | List: |
| References | |
| NUDOCS 9909150081 | |
| Download: ML20211Q498 (4) | |
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1 to TXX-99212 Technical Specification Pages:
5.0-32 5.0-33 5.0-34 I
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i 9909150081 990910 PDR ADOCK 050*****
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5.6 i
5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) a.
Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following
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1)
Moderator temperature coefficient limits for Specification 3.1.3, I
2)
Shutdown Rod Insertion Limit for Specification 3.1.5, 1
3)
Control Rod Insertion Limits for Specification 3.1.6, 1
4)
AXIAL FLUX DIFFERENCE Limits and target band for Specification 3.2.3, 5)
Heat Flux Hot Channel Factor, K(Z), W(Z), FoRTP, and the Fo (Z) c allowances for Specification 3.2.1,
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Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification 3.2.2.
7)
SHUTDOWN MARGIN values in Specifications 3.1.1,3.1.4,3.1.5, 3.1.6 and 3.1.8.
8)
Refueling Boron Concentration limits in Specification 3.9.1.
9)
Overtemperature N-16 Trip Setpoint in Specification 3.3.1.
i 10)
Reactor Coolant System pressure, temperature, and flow in Specification 3.4.1.
11)
Reactor Core Safety Limit figures (Safety Limit 2.1.1) b.
The analytical methods used to determino the core operating limits shall be those previously reviewed and approved by the NRC, spac5cclly thosc.
When an initial assumed power level of 102 percent of rated power is specified in a previously approved method,101 percent of rated power may be_used only when feedwater flow measurement (used as input for reactor thermal power measurement) is provided by the leading edge flowmeter (LEFM/) as described in document number 20 listed below. - When feedwater i
flow measurements from the LEFM/ are not available, the originally approved j
initial power. level of 102 percent of rated thermal power shall be used.
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Future revisions of approved analytical methods listed in this technical specification that currently assume 102 percent of rated power shall include (continued)
COMANCIIE PEAK - UNITS 1 AND 2 5.0-32
R: porting R:quiram:nts 5.6 l
5.6 Reporting Requirements l
5.6.5 CORE OPERATING LIMITS REPORT (continued) l the condition given above allowing use of 101 percent of rated power in safety analysis methodology when the LEFM/ is used for feedwater flow measurement.
The approved analytical methods are described in the following documents:
1)
WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary).
2)
WCAP-8385," POWER DISTRIBUTION CONTROL AND LOAD FOLLOWING PROCEDURES - TOPICAL REPORT," September 1974 (W Proprietary).
3)
T. M. Anderson To K. Kniel(Chief of Core Performance Branch, NRC)
January 31,1980--
Attachment:
Operation and Safety Analysis Aspects of an Improved Load Follow Package.
4)
NUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981. Branch Technical Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981.
5)
WCAP-10216-P-A, Revision 1 A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL Fo SURVEILLANCE TECHNICAL SPECIFICATION," February 1994 (W Proprietary).
6)
WCAP-10079-P-A,"NOTRUMP, A NODAL TRANSIENT SMALL i
BREAK AND GENERAL NETWORK CODE," August 1985, (W Proprietary).
7)
WCAP-10054-P-A," WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE", August 1985, (W Proprietary).
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8)
WCAP-11145-P-A," WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL GENERIC STUDY WITH THE NOTRUMP CODE", October 1986, (W Proprietary).
9)
RXE-90-006-P-A," Power Distribution Control Analysis and j
Overtemperature N-16 and Overpower N-16 Trip Setpoint Methodology, " June 1994.
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COMANCHE PEAK - UNITS 1 AND 2 5.0-33 L
R porting R:quir:m:nts 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (continued) 10)
RXE-88-102-P-A,"TUE-1 Departure from Nucleate Boiling Correlation", September 1993.
11)
RXE-88-102-P, Sup.1, "TUE-1 DNB Correlation - Supplement 1",
December 1990.
12)
RXE-89-002-A,"VIPRE-01 Core Thermal-Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications",
June 1989.
13)
RXE-91-001-A," Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications", October 1993.
14)
RXE-91-002-A," Reactivity Anomaly Events Methodology",
October 1993.
15)
RXE-90-007-A,"Large Break Loss of Coolant Accident Analysis Methodology", April 1993, 16)
TXX-88306," Steam Generator Tube Rupture Analysis", March 15, 1988.
17)
RXE-91-005-A, " Methodology for Reactor Core Response to Steamline Break Events," February 1994.
18)
RXE-94-001-A," Safety Analysis of Postulated inadvertent Boron Dilution Event in Modes 3,4, and 5," February 1994.
19)
RXE-95-001-P-A,"Small Break Loss of Coolant Accident Analysis Methodology," September 1996.
l 20)
Caldon, Inc. Engineering Report-80P, " Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM/ Sye*S m," Revision 0, March 1997, c.
The core operating limits sha, i determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
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The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.
(continued) j COMANCHE PEAK - UNITS 1 AND 2 5.0-34 i