ML20078K753

Proposed Tech Specs Re Heat Flux Hot Channel Factor

ML20078K753
Person / Time
Site:	Comanche Peak
Issue date:	11/18/1994
From:	TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:
Shared Package
ML20078K750	List: ML20078K746 ML20078K753
References
NUDOCS 9411250159
Download: ML20078K753 (5)

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l ATTACHMENT 3 to TXX-94242 AFFECTED TECHNICAL SPECIFICATION PAGES (NUREG 1468)

(Pages 3/4 2-5, 3/4 2-6, 6-20 & 6-21) 9411250159 DR 941118 ADOCK 05000445 PDR

Attachment 3 to TXX 94242

. Page 1 of 4 POWER DISTRIBUTION LINITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2. F,(Z) shall be evaluated to determine if it is within its limit by:

a.

Using the movable incore detectors to obtain a power distribution '

map at any THERMAL POWER greater than 55 of RATED THERMAL POWER.

b.

Determining the computed heat flux hot channel factor, F,' (Z), as follows:

by 31 to account for ma,n(ufacturing tolerances and further i the value by 5% to account for measurement uncertainties.

c.

Verifying that F * (Z), obtained in Specification 4.2.2.2b. above, satisfies the relationship in Specification 3.2.2.  :

d. The F*

relat$on(ship at the time of the target flux determination:Z) obtai F,"(Z) $ F x K(Z) for P > 0.5 P x W(I)

F,"(Z) $ x K(Z)

F. for P s 0.5 U.5 x W(Z) where F *(Z) is obtained F (Z inasSpecification is the F,1tstI, K(Z) is the normalized a function of core 4.2.2.2b.

g ,P above is the fraction of RATED THEMAL , and W(Z) ts the cycle depen-dont function that accounts for power distribution transients encountered during normal operation. F RTP, K Z and W(Z) are specifted in the CORE OPERATING LIMITS kPORT (as)per Specification 6.g.l.4.

e.

Measuring F,(Z) according to the following schedule:

1. Upon achieving equilibrium condition after exceeding by 20% or more of RATED THERMAL last determined *, or POWER, the THERMAL POWER at which F,(Z) was ,

2. At least once per 31 Effective Full Power Days, whichever occurs '

first.

• Power level may be increased untti the THERMAL POWER for extended operation has been achieved.

. COMANCHE PEAK - UNITS 1 AND 2 3/4 2-5 Unit 1 - Amendment No. 1, 6

Attach 2ent 3 to TXX 94242 P' age 2 of 4 POWER DISTRIRITION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

f. With measurements indicating maximum , Fa*(Z)g over Z Ml has increased since the previous determination of oF *(Z) either of an allewam.e. \ the following actions shall be taken:

& 2 % as 1) Increase Fa*(Z) by and verify that this value satisfies the 3p.u mel m relationship in Specification 4.2.2.2d, or ye c cL.R.

s 2) F ac(Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum,Fa*(Z) is not increasing, over Z K(I)a -

g. With being the relationships specified in Specification 4.2.2.2d above not satisfied:

1) Calculate the percent that Fa (Z) exceeds its limits by the fol-lowing expression:

/[maximus F1f 7) r WI21 '

I C x K(Z) -l x 100 for P > 0.5 i koverZ P / s

'[ maximum FJf71 x WfI) }

C x K(Z) -l

• x 100 for P s 0.5 , and i koverZ 0.5 I s

2) The following action shall be taken:

t Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in the CORE OPERATING LIMITS REPORT by 15 AFD for each percent Fa (Z) exceeds its limits as determined in Specification 4.2.2.2g.1. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified limits.

COMANCHE PEAK - UNITS 1 AND 2 3/4 2-6 Unit 1 - W ndment No. 1, 6

Attachment 3 to TXX-94242

- Page 3 of 4 ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORTS (Continued) i shall be submitted on a monthly basis to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REo0RT 6.9.1.6a Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT (COLR) before each reload cycle or any remaining part of a reload cycle for the following:

)

1). Moderator temperature coefficient BOL and E0L limits and 300 ppe sur-veillance limit for Specification 3/4.1.1.3,  :

2). Shutdown Rod Insertion Limit for Specification 3/4.1.3.5, '

3). Control Rod Insertion Limits for Specification 3/4.1.3.5, 4). AXIAL FLUX DIFFERENCE Limits and target band for Specification a a R e_ _ 3/4.2.1 -

pgcq g 5). Heat tionFlux Not Channel Factor, K(Z), W(Z), en[Fq RTP for Specifica-3/4.2.2,

~

gm 3 6). Nuclear Enthalpy Rise Hot Channel Factor Limit and the Power Factor Multiplier for Specification 3/4.2.3.

, 6.9.1.6b limits are for Units 1 and 2, unless otherwise stated, and shall be tho previously approved by the NRC in: '

1). WCAP-9272-P-A,

• WESTINGHOUSE RELOAD SAFETY EVAtu4 TION lETH000 LOGY,'

July 1985 (M Proprietary). (Methodology for Specifications 3.1.1.3 -

Moderator Teeperature coefficient, 3.1.3.5 - Shutdoura Bank Insertion Limit. 3.1.3.6 - Control tank Insertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Mot Channel Factor, and 3.2.3 - Nuclear  ;

Enthalpy Rise Mot Channel Factor.)

2). WCAP-8385,

• POWER DISTRIBUTION CONTROL A W LOAD.FOLLOWING PROCEDURES -

TOPICAL REPORT,* September 1974 (W Proprietar (Methodology for Specification 3.2.1 - Axial Flux Difference [y). Constant Axial Offset Control).)

3). T. M. Anderson to K. Kniel (C51ef of Core Performance Branch, NRC

. January 31, 1980--

Attachment:

Operation and safety Analysis Aspects -

of an-Improved Load Follow Package. (Methodology for Specification 3.2.1 - Axial Flux Difference [ Constant Axial Offset control).)

' 4). JIUREG-0800, Standard Review Plan, U.S. Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981. Branch Technical Position CPS 4,3-1, Westinghouse Constant Axial offset Centrol'(CA0C), Rev. 2, July 1981. (Methodology for Specification 3.2.1 - Axial Flux Difference [ Constant Axial Offset Control).)

COMANCHE PEAK - LetITS 1 A m 2 6-20 Unit 1 - Amene ent No. 6, 14

- - - - - _ ~ - - _ - - - - , - - - , - --. - r

Attachment 3 to TXX-94242 Page 4 of 4 ADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (Continued) oup 8 5). WCAP-10216-P-A)"RELAXAT10NOFCONSTANTAXIA 0FFSET CONTROL F, SURVEILLANCE TECHNICAL SPECIFICATION," Ar.c .00P (W Proprietary).

Re.v,pon i A, (Methodology for Specification 3.2.2 - Heat Flux Hot (.hannel Factor (W(z) surveillance requirements for F Methodology).)

1) . WCAP-10079-P-A, "NOTRUMP, A N00AL TRANSIENT SMALL BREAX AND GENERAL NETWORK CODE," August 1985, (M Proprietary).

7). WCAP-10054-P-A, " WESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE", August 1985, W Proprietary).

8). WCAP-lll45-P-A, " WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL GENERIC STUDY WITH THE NOTRUMP CODE", October 1986, M Proprietary).

9). RXE-90-006-P, " Power Distribution Control Analysis and Overtemperature N-16 and Overpower N-16 Trip Setpoint Methodology," February 1991.

(Methodology for Specification 3.2.1 - Axial Flux Difference, 3.2.2 -

Heat Flux Hot Channel Factor.)

10). RXE-88-102-P, "TUE-1 Departure from Nucleate Boiling Correlation",

January 1989.

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

December 1990.

12) . RXE-89-002, "VIPRE-01 Core Thermal-Hydraulic Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications", June 1989.

13). RXE-91-001, " Transient Analysis Methods for Comanche Peak Steam Electric Station Licensing Applications", February 1991.

14) . RXE-91-002, " Reactivity Anomaly Events Methodology", May 1991.

(Methodology for Specification 3.1.1.3 - Moderator Temperature 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 Hot Channel Factor, 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.)

15). RXE-90-007, "Large Break Loss of Coolant Accident Analysis Methodology", December 1990, 16). TXX-88306, " Steam Generator Tube Rupture Analysis", March 15, 1988.

17). RXE-91-005, " Methodology for Reactor Core Pasponse to Steamline Break Events," May, 1991. l CONANCHE PEAK - UyITS 1 AND 2 6-21 Unit 1 - Amendment No.1,5,l',13,21,28 Unit 2 - Amendment No. 4d,14

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