ML103550180

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Cycle 23 Core Operating Limits Report, Revision 2
ML103550180
Person / Time
Site: Surry Dominion icon.png
Issue date: 12/20/2010
From: Funderburk C
Virginia Electric & Power Co (VEPCO), Dominion Resources Services
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
10-691
Download: ML103550180 (11)
v • d • e


Text

D ominion Resources Services, Inc.

Innsbrook Technical Center 5000 Dom inion Boulevard, 25E , Glen Allen, VA 23060 December 20, 2010 United States Nuclear Regulatory Commission Serial No.10-691 Attention: Document Control Desk NL&OS/GDM: RO Washington, D.C. 20555 Docket No. 50-281 License No. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNIT 2 CYCLE 23 CORE OPERATING LIMITS REPORT, REVISION 2 Pursuant to Surry Technical Specification (TS) 6.2.C, enclosed is a copy of Dominion's Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 23 Pattern BOA, Revision 2. This revision to the COLR incorporates updates to TS references, reactor core safety limits, overtemperature 11T and overpower 11T setpoints, power distribution limits, and departure from nucleate boiling (DNB) parameters consistent with implementation of recently approved TS amendments 269 and 270.

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.

Serial No.10-691 Docket No. 50-281 COLR-S2C23 Rev. 2 Page 2 of 2 cc: U.S. Nuclear Regulatory Commission Region II Marquis One Towe r 245 Peachtree Center Avenue , NE Suite 1200 Atlanta, Georgia 30303-1257 Ms. K. R. Cotton NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Surry Power Station

Serial No.10-691 Docket No. 50-281 Enclosure COLR-S2C23, Revision 2 CORE OPERATING LIMITS REPORT Surry 2 Cycle 23 Pattern BOA

Serial NO.1 0-691 Docket No. 50-281 Enclosure

1.0 INTRODUCTION

This Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 23 has been prepared in accordance with the requirements of Technical Specification 6.2.C.

The Technical Specifications affected by this report are:

TS 2.1 - Safety Limit , Reactor Core TS 2.3.A.2.d - Overtemperature L1T TS 2.3.A.2.e - Overpower L1T TS 3.1.E - Moderator Temperature Coefficient TS 3.12.A.1, TS 3.12.A.2, TS 3.12.A.3 and TS 3.12.C.3.b.1(b) - Control Bank Insertion Limits TS 3.12.A.1.a, TS 3.12.A.2.a, and TS 3.12.G - Shutdown Margin TS 3.12.B.1 and TS 3.12.B.2 - Power Distribution Limits TS 3.12.F - DNB Parameters TS Table 4.1-2A - Minimum Frequency for Equipment Tests: Item 22 - RCS Flow

2.0 REFERENCES

1. VEP-FRD-42, Rev. 2. I-A, "Reload Nuclear Design Methodology," August 2003 Methodology for:

TS 3.1.E - Moderator Temperature Coefficient TS 3.12.A.l , TS 3.12.A.2, TS 3.12.A.3 and TS 3.12.C.3.b.1(b) - Control Bank Insertion Limit TS 3.12.A.1.a, TS 3.12.A.2.aand TS 3.12.G-ShutdownMargin TS 3.12.B.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor and Nuclear Enthalpy Rise Hot Channel Factor TS 3.12.F - DNB Parameters TS Table 4.1-2A - Minimum Frequency for Equipment Tests: Item 22 - RCS Flow

2. 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.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor

3. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code ," (Westinghouse Proprietary), August 1985 Methodology for :

TS 3.12.B.1 and TS 3.12 .B.2 - Heat Flux Hot Channel Factor COLR-S2C23, Rev. 2 Page l of S

Serial No.10-691 Docket No. 50-281 Enclosure

4. WCAP-10079-P-A , "NOTRUMP, A Nodal Transient Small Break and General Network Code," (Westinghouse Proprietary), August 1985 Methodology for:

TS 3.l2.B.l and TS 3.12.B.2 - Heat Flux Hot Channel Factor

5. WCAP-12610-P -A, "VANTAGE+ Fuel Assembl y Report ," (Westinghouse Proprietary),

June 1990 Methodolog y for:

TS 3.l2.B.1 and TS 3.l2.B.2 - Heat Flux Hot Channel Factor

6. VEP-NE-2-A, Rev. 0, "Statistical DNBR Evaluation Methodology," June 1987 Methodology for:

TS 3.l 2.B.1 and TS 3.12.B.2 - Nuclear Enthalp y Rise Hot Channel Factor

7. VEP-NE-3-A , Rev. 0, "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.l2.B.2 - Nuclear Enthalp y Rise Hot Channel Factor

8. WCAP-8 745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function," September 1986.

Methodology for :

TS 2.3.A.2 .d - Overtemperature b..T TS 2.3.A.2.e - Overpower b..T COLR-S2C23 , Rev. 2 Page 2 of8

Serial No.10-691 Docket No. 50-281 Enclosure 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 Safety Limit, Reactor Core (TS 2.1) 3.1.1 The Reactor Core Safety Limits are presented in Figure A-I.

3.2 Overtemperature AT (TS 2.3.A.2.d)

!1T ::;; !1To [ K1 - Kz ( 1 + t 1 S),

(T - T) + K3 (P - P,) - [(M) ]

1 + tzs Where:

!1T is measured RCS !1T, of .

~To is the indicated ~T at RATED POWER, of .

s is the Laplace transform operator, sec" :

T is the measured RCS average temperature (T avg) , OF.

T' is the nominal Tavg at RATED POWER, :s 573.0°F.

P is the measured pressurizer pressure, psig.

P' is the nominal RCS operating pressure 2: 2235 psig.

K 1 :S 1.1425 Kz 2: 0.01059 lOP K3 2: 0.000765 Ipsig tl 2: 29.7 seconds tz:S 4.4 seconds f(LlI) 2: 0.0268 { (qt - qb)} , when (qt - qb) < -24.0% RATED POWER 0, when -24.0% RATED POWER:S (qt - qb) :S 8.0% RATED POWER 0.0188 {(qt - qb) - 8.0}, when (qt - qb) > +8.0% RATED POWER Where qt and qb are percent RATED POWER in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RATED POWER.

COLR-S2C23 , Rev . 2 Page 3 of8

Serial No.10-691 Docket No. 50-281 Enclosure 3.3 Overpower AT (TS 2.3.A.2.e)

~T 5,~To [K4 - Ks (1 :~3J T - K 6(T - T') - f(~I)]

Where: ~ T is measured RCS ~ T, of .

~To is the indicated ~T at RATED POWER, of .

s is the Laplace transform operator, sec-I.

T is the measured RCS average temperature (Tavg) , of .

T' is the nominal T avg at RATED POWER, :S 573.0°F.

x, ~ 1.0965 x, 2: 0.0198 /oF for increasing T avg K 6 2: 0.001074 /oF for T > T' 2: 0 /oF for decreasing Tavg 2: 0 for T ~ T' t3 2: 9.0 seconds f(AI) = as defined above for OTAT 3.4 Moderator Temperature Coefficient (TS 3.1.E) 3.4.1 The Moderator Temperature Coefficient (MTC) limits are:

+6.0 pcm/°F at less than 50 percent ofRATED POWER, and

+6.0 pcmfF at 50 percent of RATED POWER and linearly decreasing to a pcm/°F at RATED POWER 3.5 Control Bank Insertion Limits (TS 3.12.A.l , TS 3.12.A.2 and TS 3.12.C.3.b.l(b))

3.5.1 The control rod banks shall be limited in physical insertion as shown in Figure A-2.

3.5.2 The rod insertion limit for the A and B control banks is the fully withdrawn position as shown on Figure A-2 .

3.5.3 The rod insertion limit for the A and B shutdown banks is the fully withdrawn position as shown on Figure A-2.

3.6 Shutdown Margin (TS 3.12.A.1.a, TS 3.12.A.2.a and TS 3.12.G) 3.6.1 Whenever the reactor is subcritical the shutdown margin (SDM) shall be 2: 1.77 %Ak/k.

COLR-S2C23, Rev. 2 Page 4 of 8

/ Serial NO.1 0-691 Docket No. 50-281 Enclosure 3.7 Power Distribution Limits (TS 3.12.B.l and TS 3.12.B.2) 3.7.1 Heat Flux Hot Channel Factor - FQ(z)

CFQ FQ(z) s pK(z) for P > 0.5 CFQ FQ(z) s ~K(z) for P ~ 0.5 THERMAL POWER where: P = RATED POWER 3.7.1.1 CFQ=2.32 3.7.1.2 K(z) is provided in Figure A-3 3.7.2 Nuclear Enthalpy Rise Hot Channel Factor - F.6.H(N)

F!1H(N) s CFDH * {1 + PFDH(1- P)}

THERMAL POWER where: P = RATED POWER 3.7.2.1 CFDH = 1.56 for Surry Improved Fuel (SIF) 3.7.2.2 PFDH=0.3 3.8 DNB Parameters (TS 3.12.F and TS Table 4.1-2A) 3.8.1 Departure from Nucleate Boiling (DNB) Parameters shall be maintained within their limits during POWER OPERATION:

  • Pressurizer Pressure 2: 2205 psig

Serial No.10-691 Docket No. 50-281 Enclosure Figure A-I REACTOR CORE THERMAL AND HYDRAULIC SAFETY LIMITS THREE LOOP OPERATION, 100% FLOW

I -

670 ************_ * * - T - -**_**

660

.L 2385 Pi ig  !

I 650

- !I 640 r--.  ; I 2235 '~ I r---

r--- --- I~

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11l 630 o""

11l 1985 ~ sig

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11l E 610 620

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~

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11l E 600 11l ct i, _._------

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580 .>>.__. _._.*.*.*. ............._*.*.*......... ** * * ** * * * * * *** * * ** n * * ** * *

  • _ m n

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o 10 20 30 40 50 60 70 80 90 100 110 120 Percent of Rated Power COLR-S2C23 , Rev . 2 Page 6 of8 I

Ser ial NO.1 0-691 Docket No. 50-281 Enclosure Figure A-2 SURRY UNIT 2 CYCLE 23 ROD GROUP INSERTION LIMITS Fully wid pos ition = 230 steps 230

~I I (0.4938, 230) 220 V

210

/

200

/

190 V / C-BANK (1.0, 183)

~ ~

180 170

/ ,/

/" /

160 150 V V

'C j 140 III (0, 151)

V

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Co s 130 /

~

1£ 120 o

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+:0 Ow 110 o

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a..

§-100 ~

V D-BANK o

e

'C 90 io"'"

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o 0::: 80 /

70

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60 V

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50

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40 /

30

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20 V

f---- (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 COLR-S2C23 , Rev. 2 Page 7 of8

Serial NO.1 0-691 Docket No. 50-281 Enclosure Figure A-3 K(Z)

  • Normalized FQ as a Function of Core Height 1.2 1.1 1.0 0.9 (6, 1.0)

(12, 0.925) 0.8 N

(50.7 u..

c W

N

~0 .6 0:::

oZ I 0.5 N

~

0.4 0.3 0.2 0.1 0.0 a 2 3 4 56 7 8 9 10 11 12 13 CORE HEIGHT (FT)

COLR-S2C23 , Rev. 2 Page 8 of8

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