ML073400528

Cycle 21 Core Operating Limits Report

ML073400528
Person / Time
Site:	Surry
Issue date:	12/05/2007
From:	Funderburk C Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
07-0770
Download: ML073400528 (8)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 December 5, 2007 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555-0001 Serial No.

07-0770 NLOS/vlh Docket No.

50-281 License No.

DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNIT 2 CYCLE 21 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 21 Pattern KB, Revision 1.

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

Very truly yours,

~/~

~. ~"F~~er~urk, 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. Siva P. Lingam U. S. Nuclear Regulatory Commission One White Fli nt North 11555 Rockville Pike Rockville, MD 20852-2738 Mr. Richard A. Jervey U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 Mr. C. R. Welch NRC Resident Inspector Surry Power Station Serial No. 07-0770 Cycle 21 Core Operating Limits Report Page 2 of 2

COLR-SU-SPEC-OOO-COLR-S2C21, Revision 1 CORE OPERATING LIMITS REPORT Surry 2 Cycle 21 Pattern KB Page 10f6

1.0 INTRODUCTION

This Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 21 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-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

(Westinghouse Proprietary), January 2005 (Methodology for TS 3.12.B.l and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2b. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 (W Proprietary)

(Methodology for TS 3.12.B.l and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2c. 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) 2d. WCAP-12610, "VANTAGE+ Fuel Assembly Report," June 1990 (Westinghouse Proprietary)

(Methodology for TS 3.12.B.l 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.l and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor) 3b. VEP-NE-3-A, "Qualification of the WRB-l CHF Correlation in the Virginia Power COBRA Code,"

July 1990 (Methodology for TS 3.12.B.l and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor)

Page 20f6

3.0 OPERATING LIMITS 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 pcm/F at less than 50 percent of RATED POWER, or

+6.0 pcm/F at 50 percent of RATED POWER and linearly decreasing to 0 pcm/F 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-I.

3.2.2 The rod insertion limit for the A and B control banks is the fully withdrawn position as shown on Figure A-I.

Page 3 of6

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

CFQ FQ(z):S;-- K(z) for P > 0.5 P

CFQ FQ(z):S; --K(z) for P:S; 0.5

0.5 where

P ::: Thermal 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-FMI(N) (TS 3.12.B.l)

FMi(N):S; CFDH x{l+ PFDH(l-P)}

where: P ::: Thermal Power Rated Power 3.4.1 CFDH::: 1.56 for Surry Improved Fuel (SIF) 3.4.2 PFDH::: 0.3 Page 4 of6

Figure A-I S2C21 ROD GROUP INSERTION LIMITS Fully wid position = 226 steps 230 220 210 200 190 180 170 160 150

~

140 IIIc..

130 III

II:

.2 120

t::

III

~

110 c..5 100 C) 1j 90 ocr:

80 70 60 50 40 30 20 10

/

(0.4688, 226)

V

/

/

/

C-BM K V

(1.0, 183)

./

/

/

/

VV V

V

[7 (0,151)

/

/ V

/

/

D-BANK

/

/

/

,/

V V

/

V

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V (0,23) a 0.0 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 Page 5 of6

Figure A-2 K(Z) - Normalized FQ as a Function of Core Height 12 13 10 11 9

5 678 CORE HEIGHT (FT) 4 3

2 1

6,1.0-----r----~

(12 0.925 I

I 0.0 o

1.2 0.2 0.3 0.4 0.8 0.1 0.9 1.0 1.1 N

0' LL 0.7 cw N

i

<C 0.6

2 II:o Z

~ 0.5 Page 60f6