ML051590292

Cycle 20 Core Operating Limits Report

ML051590292
Person / Time
Site:	Surry
Issue date:	05/31/2005
From:	Funderburk C Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
05-249
Download: ML051590292 (7)

Text

2I VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 May 31, 2005 U. S. Nuclear Regulatory Commission Serial No.05-249 Attention: Document Control Desk NLOS/vlh Washington, D. C. 20555-0001 Docket No.

50-281 License No. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNIT 2 CYCLE 20 CORE OPERATING LIMITS REPORT Pursuant to Surry Technical Specification 6.2.C, enclosed is a copy of Dominion's Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 20 Pattern BP, Revision 0.

If you have any questions or require additional information, please contact Mr. Gary Miller at (804) 273-2771.

Very truly yours, C. L. Funderburk, Director Nuclear Licensing and Operations Support Dominion Resources Services, Inc. for Virginia Electric and Power Company Enclosure Commitment Summary: There are no new commitments as a result of this letter.

cc:

U. S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, S. W.

Suite 23T85 Atlanta, GA 30303-8931 Mr. S. R. Monarque U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 Mr. N. P. Garrett NRC Senior Resident Inspector Surry Power Station IV

CORE OPERATING LIMITS REPORT Surry 2 Cycle 20 Pattern BP Revision 0 March 2005

1.0 INTRODUCTION

This Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 20 has been prepared in accordance with the requirements of Technical Specification 6.2.C.

The Technical Specifications affected by this report are:

TS 3.1.E and TS 5.3.A.6.b - Moderator Temperature Coefficient TS 3.12.A.2 and TS 3.12.A.3 - Control Bank Insertion Limits TS 3.12.B.1 and TS 3.12.B.2 - Power Distribution Limits

2.0 REFERENCES

1. VEP-FRD-42, Rev. 2.1-A, "Reload Nuclear Design Methodology," August 2003 (Methodology for TS 3.1.E and TS 5.3.A.6.b - Moderator Temperature Coefficient; TS 3.12.A.2 and 3.12.A.3 - Control Bank Insertion Limit; TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor) 2a. WCAP-9220-P-A, Rev. 1, "Westinghouse ECCS Evaluation Model -

1981 Version,"

February 1982 (W Proprietary)

(Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2b. WCAP-9561-P-A, ADD. 3, Rev. 1, "BART A-1: A Computer Code for the Best Estimate Analysis of Reflood Transients-Special Report: Thimble Modeling in W ECCS Evaluation Model,"

July 1986 (W Proprietary)

(Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2c. WCAP-10266-P-A, Rev. 2, 'The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," March 1987 (W Proprietary)

(Methodology for TS 3.12.B. 1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2d. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 (W Proprietary)

(Methodology forTS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2e. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code,"

August 1985 (W Proprietary)

(Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor)

I 2f. WCAP-126 10, "VANTAGE+ Fuel Assembly Report," June 1990 (Westinghouse Proprietary)

(Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 3a. VEP-NE-2-A, "Statistical DNBR Evaluation Methodology," June 1987 (Methodology for TS 3.12.B.1 and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor) 3b. VEP-NE-3-A, "Qualification of the WRB-1 CHF Correlation in the Virginia Power COBRA Code," July 1990 (Methodology for TS 3.12.B.1 and TS 3.12.B.2 - NuclearEnthalpy Rise Hot Channel Factor) 3.0 OPERATING LIITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. These limits have been developed using the NRC-approved methodologies specified in Technical Specification 6.2.C.

3.1 Moderator Temperature Coefficient (TS 3.1.E and TS 5.3.A.6.b) 3.1.1 The Moderator Temperature Coefficient (MTC) limits are:

+6.0 pcmfPF at less than 50 percent of RATED POWER, or

+6.0 pcm/F at 50% of Rated Power and linearly decreasing to 0 pctiVF at Rated Power 3.2 Control Bank Insertion Limits (TS 3.12.A.2) 3.2.1 The control rod banks shall be limited in physical insertion as shown in Figure A-1.

3.3 Heat Flux Hot Channel Factor-FO(z) (TS 3.12.B.1)

FQ(z) < CFQ K(z) for P > 0.5 P

FQ(z) 0 CFQ K(z) for P < 0.5 where

= Th=enrnal Power Rated Power 3.3.1 CFQ = 2.32 3.3.2 K(z) is provided in Figure A-2.

3.4 Nuclear Enthalpy Rise Hot Channel Factor-FAH(N) (TS 3.12.B.1)

FAH (N) < CFDH x {1 + PFDH (1 - P))

wvhzere

-P

=T7hennal Power Rated Power 3.4.1 CFDH = 1.56 for Surry Improved Fuel (SWE) 3.4.2 PFDH = 0.3

Figure A-I S2C20 ROD GROUP INSERTION LIMITS 230 220 210 200 190 180 170 160

• g 150 Ad 140 C,

W ul 130 o 120

.2-0 11 0

. 100 0

g TV 90 0CCso A:

80 70 60 50 40 30 20 10 0

C Fully w/d position = 229 steps

/

0.4875, 29)

-C-BAN

/1",

(1.0 183)

(0, 151_

0- ANK

(

0 2

3

)0 7~~~

z 7

<=-_DXA1"I

=

I=Z =

_7__

z X /_<<<

AT__

23_

no0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 Fraction of Rated Thermal Power 0.8 0.9 1.0

Figure A-2 K(Z) - Normalized FQ as a Function of Core Height 1.2 1.1 1.0 0.9 0.8 LC 0.7 aw N

E0.6 0

A~ 0.5 0.4 0.3 0.2 0.1 0.0 6, 1.0

__~~1 __

__9_2__

0 1

2 3

4 5

6 7

8 9

10 11 12 13 CORE HEIGHT (FT)