ML20209E186

Proposed Tech Specs,Adding Reactor Core Safety Limit Figures to Section 5.6.5 of Ts,Clarifying That Overpower N-16 Setpoint Remains in TS & Reflecting NRC Approval of TRs Used to Determine Core Operating Limits Presented in COLR

ML20209E186
Person / Time
Site:	Comanche Peak
Issue date:	07/09/1999
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20209E183	List: ML20209E178 ML20209E186
References
NUDOCS 9907140170
Download: ML20209E186 (4)

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, . Reactor Core SLs l

B 2.1.I BASES (continued)

A[gopriate function)ng of the RPSia_nd the.. steam generator safety vabes ensure that foryariationsiin theTHER. MAL POWERT RCS Pres _sure; RCS1 average. temperature,1RCS. flow ratef and 41.that the reactor core..SLs1will be satisfied during ste.ady_ state' operation, normal operational transients,;andAOOsaLimits,on.proce.ss variables lare developed both to prote.ct thef reactor core _SLs and for compliance with the add $lonalfestrictions;on hot leg _enthalpy.and vessel exit quality.1The Reactor, Core: Safety Limit _ figures; proyided.in the. COLR,Treflect. these process variable.limitsj APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety ,

valves or automatic protection actions serve to prevent RCS heatup to  !

the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Allowable Values for the reactor trip 4 functions are specified in LCO 3.3.1, " Reactor Trip System (RTS) l Instrumentation." In MODES 3,4,5, and 6, Applicability is not required ,

since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT The following SL violation responses are applicable to the reactor core  :

VIOLATIONS SLs. If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

Per 10CFR50.30, if a Safety Limit is violated, operations must not be resumed until authorized by the Commission.

! REFERENCES 1. 10 CFR 50, Appendix A, GDC 10.

2. FSAR, Chapter 7.

3. Pc.ver Distribution Control Analysis and Overtemperature N-16 and Overpower N-16 Trip Setpoint Methodology," RXE COMANCHE PEAK - UNITS 1 AND 2 B 2.0-4 Amendment No. 64 990714017G 990709 PDR ADOCK 05000445 p PDR

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, 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:

1 l 1) Moderator temperature coefficient limits for Specification 3.1.3,

2) Shutdown Rod Insertion Limit for Specification 3.1.5,

3) Control Rod insertion Limits for Specification 3.1.6,

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 FoC(Z)

! allowances for Specification 3.2.1,

6) 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-16JriplSetpoint 'in Specification'3;3.11 1

' 10) R.eactor.. Coolant,8ystem pressurettemperaturopand flow In Specification.l3.4/1

11) ReactoLCorejafety Limit figures (Safety Limit 2.1.1) l

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those 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 - TOP! CAL REPORT," September 1974 (W Proprietary).

(continued)

COMANCHE PEAK - UNITS 1 AND 2 5.0-31 Amendment No. 64 l

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. .- Reporting Requirements 5.6

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L. 5.6 Repcrting Requirements 5.6.5 CORE' OPERATING LIMITS REPORT (continued)'

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 AXlAL OFFSET CONTROL Fa SURVEILLANCE TECHNICAL  ;

SPECIFICATION," February 1994 (W Proprietary).

6) WCAP-10079-P-A,"NOTRUMP, A NODAL TRANSIENT SMALL 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).

8) WCAP-11145-P-A," WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL GENERIC STUDY WITH THE NOTRUMP i

CODE", October 1986, (W Proprietary).

9) RXE-90 006-PfA," Power Distribution Control Analysis and Overtemperature N-16 and Overpower N-16 Trip Setpoint Methodology, " June;1994]

10) RXE-88-102-P!A,"TUE-1 Departure from Nucleate Boiling Correlation", Jcly;19921

11) RXE-88-102-P, Sup.1, "TUE-1 DNB Correlation - Supplement 1",

December 1990.

(continued) i l

! COMANCHE PEAK- UNITS 1 AND 2 5.0-32 Amendment No. 64

d li~ , .., R porting Requir ments 5.6 L

L 5.6. Reporting Requirements (continued)

-5.6.5 CORE OPERATING LIMITS REPORT (continued)

12) RXE-89-002M,"VIPRF-01 Core Thermal-Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications",

Septembed993;

13) RXE-91-001M," Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications", October;1993]

14) RXE-91-002$, " Reactivity Anomaly Events Methodology", Octobe(

199313

15) RXE-90-007%,"Large Break Loss of Coolant Accident Analysis Mothodology",6prlU993]

16) TXX-88306, " Steam Generator Tube Rupture Analysis", March 15, 1988.

17) RXE-91-005M, " Methodology for Reactor Core Response to l Steamline Break Events," february;1994. l

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-pia,"Small Break Loss of Coolant Accident Analysis Methodology," September _199.61

c. The core operating limits shall be 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.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued) i CO.MANCHE PEAK - UNITS 1 'AND 2 5.0-33 Amendment No. 64