ML103550180

Cycle 23 Core Operating Limits Report, Revision 2

ML103550180
Person / Time
Site:	Surry
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)

Text

Dominion Resources Services, Inc.

Innsbrook Technical Center 5000 Dom inion Boulevard, 25E, Glen Allen, VA 23060 December 20, 2010 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Serial No.10-691 NL&OS/GDM: RO 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.

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 COLR-S2C23 Rev. 2 Page 2 of 2

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

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 ofTechnical 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 Assembly Report," (Westinghouse Proprietary),

June 1990 Methodology 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 Enthalpy 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 Enthalpy Rise Hot Channel Factor 8.

WCAP-8745-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)

[

( 1 + t 1S),

]

!1T ::;; !1To K1 - Kz (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 (Tavg), 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.

K1 :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 qbare 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 - K6(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 Tavg at RATED POWER, :S 573.0°F.

x, ~ 1.0965 x, 2: 0.0198 /oF for increasing Tavg 2: 0 /oF for decreasing Tavg K6 2: 0.001074 /oF for T > T' 2: 0 for T ~ T' t32: 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 apcm/°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

3.7.1.1

/

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 CFQ=2.32 Serial NO.1 0-691 Docket No. 50-281 Enclosure 3.7.1.2 K(z) is provided in Figure A-3 3.7.2.1 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 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:

Reactor Coolant System Tavg:S 577.0 OF Pressurizer Pressure 2: 2205 psig Reactor Coolant System Total Flow Rate 2: 273,000 gpm and 2: 276,000 gpm COLR-S2C23, Rev. 2 Page 5 of8

670 660 650 640 u..

11le 630 11loe 620 BE 11lE 610

~

11lE 600 11l

>ct

~

590 11l>

580 570 560 550 Serial No.10-691 Docket No. 50-281 Enclosure Figure A-I REACTOR CORE THERMAL AND HYDRAULIC SAFETY LIMITSTHREE LOOP OPERATION, 100% FLOW

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110 120 COLR-S2C23, Rev. 2 Percent of Rated Power Page 6 of8 I

Figure A-2 SURRY UNIT 2 CYCLE 23 ROD GROUP INSERTION LIMITS Fully wid position = 230 steps Serial NO.1 0-691 Docket No. 50-281 Enclosure 230 220 210 200 190 180 170 160 150

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140 III Cos 130

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+:0 Ow 110 oa..

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V f---- (0, 23) o 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 COLR-S2C23, Rev. 2 Page 7 of8

Figure A-3 K(Z)

• Normalized FQ as a Function of Core Height Serial NO.1 0-691 Docket No. 50-281 Enclosure 1.2 1.1 1.0 0.9 0.8 N

(50.7 u..

c W

N

~0.6 0:::o Z

I 0.5 N

~

0.4 0.3 0.2 0.1 0.0 (6, 1.0)------------

(12, 0.925) a 2

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56 7

8 CORE HEIGHT (FT) 9 10 11 12 13 COLR-S2C23, Rev. 2 Page 8 of8