ML16042A403

Core Operating Limits Report, Cycle 27 Pattern Hgg, Revision 2

ML16042A403
Person / Time
Site:	Surry
Issue date:	02/04/2016
From:	Huber T Dominion Resources Services, Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
15-564A
Download: ML16042A403 (10)

Text

Dominion Resources Services, Inc. mno 0*9PD Innsbrook Technical Center M fl lI l 5000 Dominion Boulevard, 2SE, Glen Allen, VA 23060 February 4, 2016 United States Nuclear Regulatory Commission Serial No. 15-564A Attention: Document Control Desk NL&OS/GDM: R0 Washington, D.C. 20555 Docket No. 50-281 License No. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNIT 2 CORE OPERATING LIMITS REPORT SURRY 2 CYCLE 27 PATTERN HGG REVISION 2 Pursuant to Surry Power Station (Surry) Technical Specification (TS) 6.2.C, attached is a copy of Dominion's Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 27, Pattern HGG, Revision 2. The revision was prepared to provide Surry Unit 2 with additional operational flexibility. Specifically, Dominion implemented an increased statistical/deterministic FAH limit of 1.635/1.70. To implement the increased FAH limit, Dominion also implemented the use of the ABB-NV and WLOP CHF correlations in place of the W-3 CHF correlation at Surry. The DNB statepoint and reload analyses for Surry Unit 2 were reanalyzed using the increased FAH and the ABB-NV and WLOP CHF correlations, and the reanalysis confirmed the applicable limits were met. The other licensing basis accident analyses continue to support the increased FAH limit and therefore were not required to be reevaluated. The use of the ABB-NV and WLOP CHF correlations were previously approved by the NRC in a letter dated August 12, 2014 (ADAMS Accession No. ML14169A359).

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

Sincerely, T. R. Huber, Director Nuclear Licensing and Operations Support Dominion Resources Services, Inc. for Virginia Electric and Power Company

Attachment:

Core Operating Limits Report, Surry Unit 2 Cycle 27, Pattern HGG, Revision 2, January 2016 Commitment Summary: There are no new commitments contained in this letter.

Serial No. 15-564A Docket No. 50-281 COLR S2C27 R2 Page 2 of 2 cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1 257 Ms. K. Cotton Gross NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector Surry Power Station

Serial No. 1 5-564A Docket No. 50-281 Attachment CORE OPERATING LIMITS REPORT Surry Unit 2 Cycle 27 Pattern HGG Revision 2 January 2016 Page 1 of 8

Serial No. 15-564A Docket No. 50-281 Attachment

1.0 INTRODUCTION

This Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 27 has been prepared in accordance with the requirements of Surry 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 AT TS 2.3 .A.2.e - Overpower AT 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.l(b) - Control Bank Insertion Limi'ts TS 3.12.A.1.a, TS 3.12.A.2.a, TS 3.12.A.3.c and TS 3.12.G- Shutdown Margin TS 3.12.B.1 and TS 3.12.B.2 - Power Distribution Limits (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

2.0 REFERENCES

1. VEP-FRD-42, Rev. 2.1l-A, "Reload Nuclear Design Methodology," August 2003.

Methodology for:

TS 2.1 - Safety Limit, Reactor Core 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.l(b) - Control Bank Insertion Limit TS 3.12.A.l1.a, TS 3. 12.A.2.a, TS 3.1 2.A.3.c and TS 3.1 2.G - Shutdown Margin TS 3.12.B.1 and TS 3.12.11.2 - 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.B1.1 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.B1.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor Page 2 of 8

Serial No. 15-564A Docket No. 50-281 Attachment

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

Methodology for:

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

5. WCAP-126 10-P-A, "VANTAGE+ Fuel Assembly Report," (Westinghouse Proprietary),

April 1995.

Methodology for:

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

6. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO,"

(Westinghouse Proprietary), July 2006.

Methodology for:

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

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

Methodology for:

TS 3.12.B.1 and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor

8. DOM-NAF-2-P-A, Rev. 0.3, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix B, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code," September 2014.

Methodology for:

TS 3.12.B.1 and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor

9. 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 AT TS 2.3 .A.2.e - Overpower AT Page 3 of 8

Serial No. 15-564A Docket No. 50-281 Attachment 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 and repeated in Section 2.0.

3.1 Safe ty Limit, Reactor Core (TS 2.1)

The Reactor Core Safety Limits are presented in Figure A-i.

3.20Overtemperature AT (TS 2.3 .A.2.d)

- *2](-T') +K3 (P-P')-fA)

Where:

AT is measured RCS AT, 0F.

AT0 is the indicated AT at RATED POWER, 0F.

s is the Laplace transform operator, see1 T is the measured RCS average temperature (Tavg), 0 F.

T' is the nominal Tavg at RATED POWER, < 573.00 F.

P is the measured pressurizer pressure, psig.

P' is the nominal RCS operating pressure > 2235 psig.

K 1

• 1.1425 K2 >_0.01059 /°F K3 _>0.000765 /psig tl > 29.7 seconds t2 _<4.4 seconds f(AI) > 0.0268 { (qt - qb)}, when (qt - qb,) < -24.0% RATED POWER 0, when -24.0% RATED POWER *_(qt - qb,) *--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.

Page 4 of 8

Serial No. 15-564A Docket No. 50-281 Attachment 3.3 Overpower AT (TS 2.3.A.2.e) where:

AT is measured RCS AT, °F.

AT0 is the indicated AT at RATED POWER, °F.

s is the Laplace transform operator, sec- 1*

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

T' is the nominal Tavg at RATED POWER, < 573.00 F.

K4 _<1.0965 K5 >_0.0198 /°F for increasing Tavg K(6> 0.001074/TF for T > T'

> 0/TF for decreasing Tayg >0 for T _<T' t3 9.0 seconds f(AI) =as defined above for OTAT 3.4 Moderator Temperature Coefficient (TS 3.1.E)

The Moderator Temperature Coefficient (MTC) limits are:

+6.0 pcm/'F at less than 50 percent of RATED POWER, and

+6.0 pcm/°F at 50 percent of RATED POWER and linearly decreasing to 0 pcm/'F at RATED POWER 3.5 Control Bank Insertion Limits (TS 3.12.A.1, TS 3.12.A.2, TS 3.12.A.3, and TS Yl12.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 Margzin (TS 3.12.A.l.a, TS 3.12.A.2.a, TS 3.12.A.3.c and TS 3.12.G)

Shutdown margin (SDM) shall be > 1.77 %Ak/k.

Page 5 of 8

Serial No. 15-564A Docket No. 50-281 Attachment 3.7 Power Distribution Limits (TS 3.12.B.1 and TS 3.12.B.2) 3.7.1 Heat Flux Hot Channel Factor - FQ(z)

CFQ ...

FQ(z) <_ K--(z) for P >0.5 FQ(z) <_* K(z) for P *_0.5 THERMAL POWER where: P =

RATED POWER CFQ=2.5 K(z) --1.0 for all core heights, z 3.7.2 Nuclear Enthalpy Rise Hot Channel Factor - FAH(N)

FLAH(N) *_ CFDH * (1 + PFDH(1 - P)}

THERMAL POWER where: P =

RATED POWER CFDH=1.635 PFDH= 0.3 3.8 DNB Parameters (TS 3.12.F and TS Table 4.1-2A)

Departure from Nucleate Boiling (DNB) Parameters shall be maintained within their limits during POWER OPERATION:

• Reactor Coolant System Tavg _<577.0 0 F

• Pressurizer Pressure >Ž2205 psig

• Reactor Coolant System Total Flow Rate > 273,000 gpm (Tech Spec Limit) and > 274,000 gpm (COLR Limit)

Page 6 of 8

Serial No. 15-564A Docket No. 50-281 Attachment Figure A-i REACTOR CORE SAFETY LIMITS THREE LOOP OPERATION, 100% FLOW 670 660 650 LL U'

aj 640 a,

L.

a, 630 0

L.

4-, 620 610 2

a, I-a, 600 Cu I-a, 590 a,

In 580 U) a, 570 560 550 0 10 20 30 40 50 60 70 80 90 100 110 120 Percent of Rated Power Page 7 of 8

Serial No. 15-564A Docket No. 50-281 Attachment Figure A-2 Surry 2 Cycle 27 Rod Group Insertion Limits Max w/d position = 228 step*s 230 220 *O.4813, 2 8) 210 200 C-Bank 190 180 (t4o, 170 160 "o 150

", 140 (0.0, 151)

U130 C 120 S110 S100

• 0 90 70 60 50 70 30 20 10 0

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 8 of 8

Dominion Resources Services, Inc. mno 0*9PD Innsbrook Technical Center M fl lI l 5000 Dominion Boulevard, 2SE, Glen Allen, VA 23060 February 4, 2016 United States Nuclear Regulatory Commission Serial No. 15-564A Attention: Document Control Desk NL&OS/GDM: R0 Washington, D.C. 20555 Docket No. 50-281 License No. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)

SURRY POWER STATION UNIT 2 CORE OPERATING LIMITS REPORT SURRY 2 CYCLE 27 PATTERN HGG REVISION 2 Pursuant to Surry Power Station (Surry) Technical Specification (TS) 6.2.C, attached is a copy of Dominion's Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 27, Pattern HGG, Revision 2. The revision was prepared to provide Surry Unit 2 with additional operational flexibility. Specifically, Dominion implemented an increased statistical/deterministic FAH limit of 1.635/1.70. To implement the increased FAH limit, Dominion also implemented the use of the ABB-NV and WLOP CHF correlations in place of the W-3 CHF correlation at Surry. The DNB statepoint and reload analyses for Surry Unit 2 were reanalyzed using the increased FAH and the ABB-NV and WLOP CHF correlations, and the reanalysis confirmed the applicable limits were met. The other licensing basis accident analyses continue to support the increased FAH limit and therefore were not required to be reevaluated. The use of the ABB-NV and WLOP CHF correlations were previously approved by the NRC in a letter dated August 12, 2014 (ADAMS Accession No. ML14169A359).

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

Sincerely, T. R. Huber, Director Nuclear Licensing and Operations Support Dominion Resources Services, Inc. for Virginia Electric and Power Company

Attachment:

Core Operating Limits Report, Surry Unit 2 Cycle 27, Pattern HGG, Revision 2, January 2016 Commitment Summary: There are no new commitments contained in this letter.

Serial No. 15-564A Docket No. 50-281 COLR S2C27 R2 Page 2 of 2 cc: U.S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1 257 Ms. K. Cotton Gross NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 Dr. V. Sreenivas NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 08 G-9A 11555 Rockville Pike Rockville, Maryland 20852-2738 NRC Senior Resident Inspector Surry Power Station

Serial No. 1 5-564A Docket No. 50-281 Attachment CORE OPERATING LIMITS REPORT Surry Unit 2 Cycle 27 Pattern HGG Revision 2 January 2016 Page 1 of 8

Serial No. 15-564A Docket No. 50-281 Attachment

1.0 INTRODUCTION

This Core Operating Limits Report (COLR) for Surry Unit 2 Cycle 27 has been prepared in accordance with the requirements of Surry 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 AT TS 2.3 .A.2.e - Overpower AT 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.l(b) - Control Bank Insertion Limi'ts TS 3.12.A.1.a, TS 3.12.A.2.a, TS 3.12.A.3.c and TS 3.12.G- Shutdown Margin TS 3.12.B.1 and TS 3.12.B.2 - Power Distribution Limits (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

2.0 REFERENCES

1. VEP-FRD-42, Rev. 2.1l-A, "Reload Nuclear Design Methodology," August 2003.

Methodology for:

TS 2.1 - Safety Limit, Reactor Core 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.l(b) - Control Bank Insertion Limit TS 3.12.A.l1.a, TS 3. 12.A.2.a, TS 3.1 2.A.3.c and TS 3.1 2.G - Shutdown Margin TS 3.12.B.1 and TS 3.12.11.2 - 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.B1.1 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.B1.1 and TS 3.12.B.2 - Heat Flux Hot Channel Factor Page 2 of 8

Serial No. 15-564A Docket No. 50-281 Attachment

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

Methodology for:

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

5. WCAP-126 10-P-A, "VANTAGE+ Fuel Assembly Report," (Westinghouse Proprietary),

April 1995.

Methodology for:

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

6. WCAP-12610-P-A and CENPD-404-P-A, Addendum 1-A, "Optimized ZIRLO,"

(Westinghouse Proprietary), July 2006.

Methodology for:

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

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

Methodology for:

TS 3.12.B.1 and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor

8. DOM-NAF-2-P-A, Rev. 0.3, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix B, "Qualification of the Westinghouse WRB-1 CHF Correlation in the Dominion VIPRE-D Computer Code," and Appendix D, "Qualification of the ABB-NV and WLOP CHF Correlations in the Dominion VIPRE-D Computer Code," September 2014.

Methodology for:

TS 3.12.B.1 and TS 3.12.B.2 - Nuclear Enthalpy Rise Hot Channel Factor

9. 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 AT TS 2.3 .A.2.e - Overpower AT Page 3 of 8

Serial No. 15-564A Docket No. 50-281 Attachment 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 and repeated in Section 2.0.

3.1 Safe ty Limit, Reactor Core (TS 2.1)

The Reactor Core Safety Limits are presented in Figure A-i.

3.20Overtemperature AT (TS 2.3 .A.2.d)

- *2](-T') +K3 (P-P')-fA)

Where:

AT is measured RCS AT, 0F.

AT0 is the indicated AT at RATED POWER, 0F.

s is the Laplace transform operator, see1 T is the measured RCS average temperature (Tavg), 0 F.

T' is the nominal Tavg at RATED POWER, < 573.00 F.

P is the measured pressurizer pressure, psig.

P' is the nominal RCS operating pressure > 2235 psig.

K 1

• 1.1425 K2 >_0.01059 /°F K3 _>0.000765 /psig tl > 29.7 seconds t2 _<4.4 seconds f(AI) > 0.0268 { (qt - qb)}, when (qt - qb,) < -24.0% RATED POWER 0, when -24.0% RATED POWER *_(qt - qb,) *--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.

Page 4 of 8

Serial No. 15-564A Docket No. 50-281 Attachment 3.3 Overpower AT (TS 2.3.A.2.e) where:

AT is measured RCS AT, °F.

AT0 is the indicated AT at RATED POWER, °F.

s is the Laplace transform operator, sec- 1*

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

T' is the nominal Tavg at RATED POWER, < 573.00 F.

K4 _<1.0965 K5 >_0.0198 /°F for increasing Tavg K(6> 0.001074/TF for T > T'

> 0/TF for decreasing Tayg >0 for T _<T' t3 9.0 seconds f(AI) =as defined above for OTAT 3.4 Moderator Temperature Coefficient (TS 3.1.E)

The Moderator Temperature Coefficient (MTC) limits are:

+6.0 pcm/'F at less than 50 percent of RATED POWER, and

+6.0 pcm/°F at 50 percent of RATED POWER and linearly decreasing to 0 pcm/'F at RATED POWER 3.5 Control Bank Insertion Limits (TS 3.12.A.1, TS 3.12.A.2, TS 3.12.A.3, and TS Yl12.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 Margzin (TS 3.12.A.l.a, TS 3.12.A.2.a, TS 3.12.A.3.c and TS 3.12.G)

Shutdown margin (SDM) shall be > 1.77 %Ak/k.

Page 5 of 8

Serial No. 15-564A Docket No. 50-281 Attachment 3.7 Power Distribution Limits (TS 3.12.B.1 and TS 3.12.B.2) 3.7.1 Heat Flux Hot Channel Factor - FQ(z)

CFQ ...

FQ(z) <_ K--(z) for P >0.5 FQ(z) <_* K(z) for P *_0.5 THERMAL POWER where: P =

RATED POWER CFQ=2.5 K(z) --1.0 for all core heights, z 3.7.2 Nuclear Enthalpy Rise Hot Channel Factor - FAH(N)

FLAH(N) *_ CFDH * (1 + PFDH(1 - P)}

THERMAL POWER where: P =

RATED POWER CFDH=1.635 PFDH= 0.3 3.8 DNB Parameters (TS 3.12.F and TS Table 4.1-2A)

Departure from Nucleate Boiling (DNB) Parameters shall be maintained within their limits during POWER OPERATION:

• Reactor Coolant System Tavg _<577.0 0 F

• Pressurizer Pressure >Ž2205 psig

• Reactor Coolant System Total Flow Rate > 273,000 gpm (Tech Spec Limit) and > 274,000 gpm (COLR Limit)

Page 6 of 8

Serial No. 15-564A Docket No. 50-281 Attachment Figure A-i REACTOR CORE SAFETY LIMITS THREE LOOP OPERATION, 100% FLOW 670 660 650 LL U'

aj 640 a,

L.

a, 630 0

L.

4-, 620 610 2

a, I-a, 600 Cu I-a, 590 a,

In 580 U) a, 570 560 550 0 10 20 30 40 50 60 70 80 90 100 110 120 Percent of Rated Power Page 7 of 8

Serial No. 15-564A Docket No. 50-281 Attachment Figure A-2 Surry 2 Cycle 27 Rod Group Insertion Limits Max w/d position = 228 step*s 230 220 *O.4813, 2 8) 210 200 C-Bank 190 180 (t4o, 170 160 "o 150

", 140 (0.0, 151)

U130 C 120 S110 S100

• 0 90 70 60 50 70 30 20 10 0

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 8 of 8