ML091130631

Cycle 23 Core Operating Limits Report

ML091130631
Person / Time
Site:	Surry
Issue date:	04/23/2009
From:	Funderburk C Dominion, Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
09-269 COLR-S1C23, Rev 0
Download: ML091130631 (8)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 April 23, 2009 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555-0001 Serial No.09-269 NLOSlvlh Docket No.

50-280 License No. DPR-32 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNIT 1 CYCLE 23 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 23 Pattern NXS, Revision O.

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

Sincerely,

/~--

(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. J. F. Stang, Jr.

NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North MaiI Stop 0-8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 MS. D. N. Wright NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North MaiI Stop 0-8 H4A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Surry Power Station Serial No.09-269 Cycle 23 Core Operating Limits Report Page 2 of 2

COLR-SIC23, Rev. a COLR-SIC23, Revision 0 CORE OPERATING LIMITS REPORT Surry 1 Cycle 23 Pattern NXS Page 1 of 6

1.0 INTRODUCTION

This Core Operating Limits Report (COLR) for Surry Unit 1 Cycle 23 has been prepared in accordance with the requirements ofTechnical Specification 6.2.C.

The Technical Specifications affected by this report are:

TS 3.1.E - 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 - 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-I0054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985 (W Proprietary)

(Methodology for TS 3.12.B.I and TS 3.12.B.2 - Heat Flux Hot Channel Factor) 2c. WCAP-I0079-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-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)

COLR-SIC23, Rev. 0 Page 2 of6

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) 3.1.1 The Moderator Temperature Coefficient (MTC) limits are:

+6.0 pcrnfF at less than 50 percent of RATED POWER, and

+6.0 pcm/T' at 50 percent of RATED POWER and linearly decreasing to a pcmfF at RATED POWER 3.2 Control Bank Insertion Limits (TS 3.I2.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.

COLR-SIC23, Rev. a Page 3 of6

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) 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-FAII(N) (TS 3.12.B.1)

FM!(N) s CFDH x{l+PFDH(l-P)}

h P

Thermal Power were:

=-----

Rated Power 3.4.1 CFDH = 1.56 for Surry Improved Fuel (SIP) 3.4.2 PFDH = 0.3 COLR-SIC23, Rev. 0 Page 4 of6

Figure A-I SURRY UNIT 1 CYCLE 23 ROD GROUP INSERTION LIMITS Fully wid position = 228 steps 230 220 210 200 190 180 170 160 150 1J

~ 140 lfJ Q.

Gliii 130

~

g 120

'iii

~ 110 Q.

J 100 ee 1J 90 o

II:

80 70 60 50 40 30 20 10

/

(0.4813, 228)

./V

/

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C-BANK

~V (1.0, 183)

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/

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/

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/r

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V (0, 151)

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D-BANK

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r

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/(0,23) o 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 COLR-SIC23, Rev. 0 Fraction of Rated Thermal Power Page 5 of6

1.2 1.1 1.0 0.9 0.8 N

~ 0.7 cw N

i c( 0.6

iEa:oz

..:... 0.5 N

~

0.4 0.3 0.2 0.1 Figure A-2 K(Z) - Normalized FQ as a Function of Core Height 6,1.0-r---t---r---r-----.....

(12 0.925 0.0 o

2 3

4 5

6 7

8 CORE HEIGHT (FT) 9 10 11 12 13 COLR-SIC23, Rev. 0 Page 60f6