ML073410347

Cycle 22 Core Operating Limits Reports

ML073410347
Person / Time
Site:	Surry
Issue date:	12/05/2007
From:	Funderburk C Dominion, Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
07-0771
Download: ML073410347 (8)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 December 5, 2007 United States Nuclear Regulatory Commission Serial No.: 07-0771 Attention: Document Control Desk NLOS/VLH Washington, D. C. 20555-0001 Docket No.: 50-280 License No.: DPR-32 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNIT 1 CYCLE 22 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 1 Cycle 22 Pattern EON, Revision O.

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

Very truly yours, (F/y-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.

Serial No. 07-0771 Cycle 22 Core Operating Limits Report Page 2 of 2 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 Flint 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

COLR-SU-SPEC-OOO-COLR-SIC22, Revision 0 CORE OPERATING LIMITS REPORT Surry 1 Cycle 22 Pattern EON Page 10f6

1.0 INTRODUCTION

This Core Operating Limits Report (COLR) for Surry Unit 1 Cycle 22 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.l 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.l 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.l and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2d. WCAP-1261O, "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)

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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.

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

CFQ FQ(z) ~ --K(z) for P > 0.5 P

CFQ FQ(z) ~--K(z) for P ~ 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)

FMf(N) ~ CFDHx{l+PFDH(l-P)}

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

Figure A-I S1 C22 ROD GROUP INSERTION LIMITS Fully wid position = 226 steps 230

/ (0.468 3,226) 220 /

/

210 200

/

190

/ C-BA NK 180

// (1.0 , 183) /

170 / /

~

// /v 3: 160 en / /

'l.

150

/

Q) en (0,151

'*I: 140 s::::

/

o 130

/

Q.

en o 120

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J o 110 / O-BI NK e"

~

/

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100 "C

o a:: 90

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/v 80 70 V

60

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50 /

40

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30

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20 (0,23) 10 o

0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Fraction of Rated Thermal Power Page 5 of6

Figure A-2 K(Z) - Normalized FQ as a Function of Core Height 1.2 1.1 6,1.0 1.0

~

0.9 (12 0.925 0.8 N

LL C

w N

0.7

i

<C 0.6

2 a:

o z

~ 0.5

~

0.4 0.3 0.2 --

0.1 --

0.0 a 2 3 4 5 678 9 10 11 12 13 CORE HEIGHT (FT)

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