ML101750002

Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate

ML101750002
Person / Time
Site:	Surry
Issue date:	09/24/2010
From:	Cotton K Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO)
Cotton K, NRR/DORL, 301-415-1438
References
TAC ME3293, TAC ME3294
Download: ML101750002 (93)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 24, 2010 Mr. David A. Heacock President and Chief Nuclear Officer Virginia Electric and Power Company Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711

SUBJECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE (TAC NOS. ME3293 AND ME3294)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 269 to Renewed Facility Operating License No. DPR-32 and Amendment No. 268 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2, respectively. The amendments change the Technical Specifications (TSs) in response to your application dated January 27, 2010, as supplemented by letters dated February 4 and April 29, 2010.

These amendments would increase each unit's rated power (RP) level from 2546 megawatts thermal (MWt) to 2587 MWt and make TSs changes as necessary to support operation at the uprated power level. The proposed change is an increase in RP of approximately 1.6 percent.

The increase is based upon increased feedwater flow measurement accuracy achieved by using high-accuracy Cameron CheckPlus' Leading Edge Flow Meter ultrasonic flow measurement instrumentation.

D. Heacock

- 2 A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

{o./lhLa~

Karen Cotton, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-280 and 50-281

Enclosures:

1. Amendment No. 269 to DPR-32

2. Amendment No. 268 to DPR-37

3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-280 SURRY POWER STATION, UNIT NO.1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 269 Renewed License No. DPR-32

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Virginia Electric and Power Company (the licensee) January 27, 2010, as supplemented by letters dated February 4 and April 29, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR)

Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is amended by adding License Condition T.3 to Renewed Facility Operating License No. OPR-32, as indicated in the attachment to the amendment, and reads as follows:

COMMITMENT SCHEDULED COMPLETION DATE

1. VEPCO will perform the final acceptance of the Surry 1 uncertainty analysis to ensure the results are bounded by the statements contained in the LAR (Attachment 5, Section 1.1.0.4.1).

Prior to operating above 2546 MWt (98.4% RP).

2. Technical Requirements Manual Prior to operating above 2546 MWt (TRM) will be revised to include UFM administrative controls (Attachment 1 Section 3.0).

(98.4% RP).

3. Revise procedures, programs, and documents for the new UFM (including transducer replacement)

(Attachment 5, Sections 1.1, 1.1.0.1.1, 1.1.H, VI1.1, V11.2.A, and VII.4).

Prior to operating above 2546 MWt (98.4% RP).

4. Appropriate personnel will receive Prior to operating above 2546 MWt training on the UFM and affected procedures (Attachment 5, Sections 1.1.0.1.1, V11.2.A, VI1.2.0, and VI1.3).

(98.4% RP).

5. The FAC CHECWORKS SFA Prior to operating above 2546 MWt models will be updated to reflect the MUR PU conditions (Attachment 5,Section IV.1.E.iii).

(98.4% RP)

6. Simulator changes and validation will be completed (Attachment 5, Section VI1.2.C).

Prior to operating above 2546 MWt (98.4% RP).

7. Revise existing plant operating procedures related to temporary operation above full steady-state licensed power levels (Attachment 5,Section VII.4).

Prior to operating above 2546 MWt (98.4% RP).

- 3

2.

(Continued, DPR-32)

COMMITMENT SCHEDULED COMPLETION DATE

8. Process UFSAR changes in accordance with 10 CFR 50.59 (Attachment 1, Section 3.0).

In accordance with 10 CFR 50.71(e).

9. UFM commissioning and calibration will be completed (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

10. Confirm flow normalization factors (Attachment 5, Section 1.1.G).

Prior to operating above 2546 MWt (98.4% RP).

11. Rescaling and calibration of main turbine first stage pressure input to AMSAC (Attachment 5, Sections 11.2.28, V11.2.B, V1I1.2, and VII1.3).

Prior to operating above 2546 MWt (98.4% RP).

12. Determine EO service life for excore detectors (Attachment 5, Sections 1I1.2.A and V.1.C).

Prior to operating above 2546 MWt (98.4% RP).

13. The excore neutron detectors are scheduled to be replaced (Attachment 5,Section V.I.C).

Unit 1: fall 2010 Refueling Outage.

Unit 2: spring 2011 Refueling Outage.

14. Revise EOP setpoints (Attachment 5, Section VI1.2.A).

Prior to operating above 2546 MWt (98.4% RP).

15. The UFM feedwater flow and temperature data will be compared to the feedwater flow venturis output and the feedwater RTD output (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

- 4

2.

(Continued, DPR-32)

COMMITMENT SCHEDULED COMPLETION DATE

16. For the applicable UFSAR Chapter 14 events, Surry 1 will re-analyze the transient consistent 2546 MWt (98.4% RP) with VEPCO's NRC-approved reload design methodology in VEP-FRD-42, Rev.

2.1-A.

If NRC review is deemed necessary pursuant to the requirements of 10 CFR 50.59, the accident analyses will be submitted to the NRC for review prior to operation at the uprate power level. These commitments apply to the following Surry 1 UFSAR Chapter 14 DNBR analyses that were analyzed at 2546 MWt consistent with the Statistical DNBR Evaluation Methodology in VEP-NE-2-A:

• Section 14.2.7 - Excessive Heat Removal due to Feedwater System Malfunctions (Full Power Feedwater Temperature Reduction case only);

• Section 14.2.8 - Excessive Load Increase Incident;

• Section 14.2.9 - Loss of Reactor Coolant Flow; and

• Section 14.2.10 - Loss of External Electrical Load Prior to operating above 2546 MWt (98.4% RP).

3.

The license is also amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-32 is hereby amended to read as follows:

(B)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 269, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

- 5

4.

The license is also amended by a change to paragraph 3.A. of Renewed Facility Operating License No. DPR-32 as indicated in the attachment to this license amendment, and is hereby amended to read as follows:

(A)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

5.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days. Accordingly, the scheduled completion dates listed in License Condition 3.T, shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the condition and shall determine the environmental qualification service life of the excore detectors and incorporate changes in the qualified lifetime of this equipment into environmental qualification program documentation, prior to operating above the current maximum operating level of 2587 MWt, as described in Virginia Electric and Power Company's letters dated January 27, February 4, and April 29, 2010.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to License No. DPR-32 and the Technical Specifications Date of Issuance: September 24, 2010

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 VIRGINIA ELECTRIC AND POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO.2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 268 Renewed License No. DPR-37

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Virginia Electric and Power Company (the licensee) January 27, 2010, as supplemented by letters dated February 4 and April 29, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR)

Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is amended by adding License Condition 3.T to Renewed Facility Operating License No. OPR-37, as indicated in the attachment to the amendment, and reads as follows:

SCHEDULED COMPLETION COMMITMENT DATE

1. VEPCO will perform the final Prior to operating above 2546 MWt acceptance of the Surry 2 (98.4% RP).

uncertainty analysis to ensure the results are bounded by the statements contained in the LAR (Attachment 5, Section 1.1.0.4.1).

2. Technical Requirements Manual Prior to operating above 2546 MWt (TRM) will be revised to include (98.4% RP).

UFM administrative controls (Attachment 1, Section 3.0).

Prior to operating above 2546 MWt and documents for the new UFM

3. Revise procedures, programs, (98.4% RP).

(including transducer replacement) (Attachment 5, Sections 1.1, 1.1.0.1.1, 1.1.H, V11.1, VI1.2.A, and VII.4).

Prior to operating above 2546 MWt receive training on the UFM and

4. Appropriate personnel will (98.4% RP).

affected procedures (Attachment 5, Sections 1.1.0.1.1, VI1.2.A, VI1.2.0, and VI1.3).

Prior to operating above 2546 MWt models will be updated to reflect

5. The FAC CHECWORKS SFA (98.4% RP) the MUR PU conditions (Attachment 5,Section IV.1.E.iii).

Prior to operating above 2546 MWt will be completed (Attachment 5,

6. Simulator changes and validation (98.4% RP).

Section VI1.2.C).

Prior to operating above 2546 MWt procedures related to temporary

7. Revise existing plant operating (98.4% RP).

operation above full steady-state licensed power levels (Attachment 5,Section VII.4).

- 3

2.

(Continued, DPR-37)

COMMITMENT SCHEDULED COMPLETION DATE

8. Process UFSAR changes in accordance with 10 CFR 50.59 (Attachment 1, Section 3.0).

In accordance with 10 CFR 50.71(e).

9. UFM commissioning and calibration will be completed (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

10. Confirm flow normalization factors (Attachment 5, Section 1.1.G).

Prior to operating above 2546 MWt (98.4% RP).

11. Rescaling and calibration of main turbine first stage pressure input to AMSAC (Attachment 5, Sections 11.2.28, V11.2.B, VII1.2, and VII1.3).

Prior to operating above 2546 MWt (98.4% RP).

12. Determine EO service life for excore detectors (Attachment 5, Sections 1I1.2.A and V.1.C).

Prior to operating above 2546 MWt (98.4% RP).

13. The excore neutron detectors are scheduled to be replaced (Attachment 5,Section V.I.C).

Unit 1: fall 2010 Refueling Outage.

Unit 2: spring 2011 Refueling Outage.

14. Revise EOP setpoints (Attachment 5, Section VI1.2.A).

Prior to operating above 2546 MWt (98.4% RP).

15. The UFM feedwater flow and temperature data will be compared to the feedwater flow venturis output and the feedwater RTD output (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

- 4

2.

(Continued, DPR-37)

COMMITMENT SCHEDULED COMPLETION DATE

16. For the applicable UFSAR Chapter 14 events, Surry 2 will re-analyze the transient consistent 2546 MWt (98.4% RP) with VEPCO's NRC-approved reload design methodology in VEP-FRD-42, Rev.

2.1-A.

If NRC review is deemed necessary pursuant to the requirements of 10 CFR 50.59, the accident analyses will be submitted to the NRC for review prior to operation at the uprate power level. These commitments apply to the following Surry 2 UFSAR Chapter 14 DNBR analyses that were analyzed at 2546 MWt consistent with the Statistical DNBR Evaluation Methodology in VEP-NE-2-A:

• Section 14.2.7 - Excessive Heat Removal due to Feedwater System Malfunctions (Full Power Feedwater Temperature Reduction case only);

• Section 14.2.8 - Excessive Load Increase Incident;

• Section 14.2.9 - Loss of Reactor Coolant Flow; and

• Section 14.2.10 - Loss of External Electrical Load Prior to operating above 2546 MWt (98.4% RP).

3.

The license is also amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Renewed Facility Operating License No. DPR-37 is hereby amended to read as follows:

(B)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 268, are hert:by incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

- 5

4.

The license is also amended by a change to paragraph 3.A of Renewed Facility Operating License No. DPR-37 as indicated in the attachment to this license amendment, and is hereby amended to read as follows:

(A)

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

5.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days. Accordingly, the scheduled completion dates listed in License Condition 3.T, shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the condition and shall determine the environmental qualification service life of the excore detectors and incorporate changes in the qualified lifetime of this equipment into environmental qualification program documentation, prior to operating above the current maximum operating level of 2587 MWt, as described in Virginia Electric and Power Company's letters dated January 27, February 4, and Apri/29, 2010.

FOR THE NUCLEAR REGULATORY COMMISSION

\\'\\

~A-1 ~-U~

~)".~\\.~~.~~

(

J~\\eph G. Giitter, Director Di ision of Operating Reactor Licensing

\\"".0 ice of Nuclear Reactor Regulation

Attachment:

Changes License No. DPR-37 and the Technical Specifications Date of Issuance:

September 24, 2010

ATIACHMENT TO LICENSE AMENDMENT NO. 269 RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND TO LICENSE AMENDMENT NO. 268 RENEWED FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NO. 50-281 Replace the following pages of the Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages Licenses Licenses License No. DPR-32 Page 3 Page 3 Page 6 Page 6 Page 7 Page 8 Page 9 License No. DPR-37 Page 3 Page 3 Page 6 Page 6 Page 7 Page 8 Page 9 TSs TSs TS 1.0-1 TS 1.0-1 TS Figure 2.1-1 TS Figure 2.1-1 TS 2.3-2 TS 2.3-2

-3

3. This renewed license shall be deemed to contain an~ Is sUbject to the condlt(ons specified in the following Commission regulations: 10 CFR Part 20, Section 30:34 of 10 CFR Part 30, Section 40.41 of 10 CFR Part 40, Sections 50.54 and 50.59 of 10 CFR Part 50. and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provtslons of the Act and tne rules, regulations, and olders of the Commission now or hereafter in effect and is subject to the additional conditions specJfled below:

A. Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not In excess of 2587 megawatts (thermal).

a. Technical SDectfications The Technical SPAcifications contained 11"1 Appendix A. as revised through Amendment No. 269 are hereby Incorporated in the renewvd license. The IIcel"lsee shall operate the facility In accordance with the Technical Specifications.

C. Reports T.he licensee shall make certain reports In accordance with the requirements of the Technical Specifications.

O. Records The licensee shall keep facility operating records in acamlance with the requirements of the Technical Specifications.

E.. Deleted by Amendment 65 F. Deleted by Amendment 71 G. Deleted by Amendment 227 H. Deleted by Amendment 227

l.

Fire Prot"ctlon The licensee shall Implement and maintain in effect the provisions of the approved fire protection. program as described In the Update*d Rnal Safety Analysis Report" and as approved In the SeR dated September 19, 1979, (and Supplements dated May 29,1980, October 9, 1980, December 18, 1980, February 13,1981, December 4,1981, April 27. 1982, November 18,1982, January 17, 1984, February 25. 1988. and SURRY UNIT' Renewed L.icense No. D~R*32 Amendment No. 269

- 6 R. As discussed in the footnote to Technical Specifications 3.23.C.2.a.1 and 3.23.C.2.b.1, the use of temporary 45-day and 14-day allowed outage time extensions to permit replacement of the Main Control Room and Emergency Switchgear Room Air Conditioning System chilled water piping shall be in accordance with the basis, risk evaluation, equipment unavailability restrictions, and compensator actions provided in the licensee's submittal dated February 26,2007 (Serial No. 07-0109) and in the associated supplemental transmittals, as approved by the NRC Safety Evaluation.

S. Upon implementation of Amendment No. 260 adopting TSTF-448, Revision 3, the determination of Main Control Room/Emergency SWitchgear Room (MRC/ESGR) envelope unfiltered air inleakage as required by TS SR 4.18 in accordance with TS 6.4.R.3.a, the assessment of MCRIESGR envelope habitability as required by Specification 6.4.R.3.b, and the measurement of MCRIESGR envelope pressure as required by Specification 6.4.R.4, shall be considered met. Following implementation:

(1) The first performance of SR 4.18, in accordance with Specification 6.4.R.3.a, shall be within the specified frequency of 6 years plus the 18-month allowance of SR 4.0.2, as measured from January 18, 2004, the date of the most recent successful tracer gas test, as stated in the April 22, 2004 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(2) The first performance of the periodic assessment of MCRIESGR envelope habitability, Specification 6.4.R.3.b, shall be within 3 years, plus the 9-month allowance of SR 4.0.2, as measured from January 18, 2004, the date of the most recent successful tracer gas test, as stated in April 22, 2004 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(3) The first performance of the periodic measurement of MCRIESGR envelope pressure, Specification 6.4.R.4, shall be within 18 months, plus the 138 days allowed by SR 4.0.2, as measured from January 19. 2007, the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

Renewed License No. DPR-32 Amendment No. 269

- 7 T.

COMMITMENT SCHEDULED COMPLETION DATE

1. VEPCO will perform the final Prior to operating above 2546 MWt acceptance of Surry 1 uncertainty (98.4% RP).

analysis to ensure the results are bounded by the statements contained in the LAR (Attachment 5, Section 1.1.0.4.1 ).

2. Technical Requirements Manual Prior to operating above 2546 MWt (TRM) will be revised to include UFM (98.4% RP).

administrative controls (Attachment 1 Section 3.0).

3. Revise procedures, programs, and Prior to operating above 2546 MWt documents for the new UFM (including (98.4% RP).

transducer replacement) (Attachment 5, Sections 1.1,1.1.0.1, 1.1.H, VI1.1, VI1.2.A, and VII.4).

4. Appropriate personnel will receive Prior to operating above 2546 MWt training on the UFM and affected (98.4% RP).

procedures (Attachment 5, Sections 1.1.0.1.1, VI1.2.A, V11.2.0, and V11.3)

5. The FAC CHECWORKS SFA models Prior to operating above 2546 MWt will be updated to reflect the MUR PU (98.4% RP).

conditions (Attachment 5,Section IV.1.E.iii).

6. Simulator changes and validation will Prior to operating above 2546 MWt be completed (Attachment 4, Section (98.4% RP).

VI1.2.C).

7. Revise existing plant operating Prior to operating above 2546 MWt procedures related to temporary (98.4% RP).

operation above full steady-state licenses power levels (Attachment 5,Section VII.4).

8. Process UFSAR changes in In accordance with 10 CFR 50.71(e) accordance with 10 CFR 50.59 (Attachment 1, Section 3.0).

I Renewed License No. OPR-32 Amendment No. 269

- 8 T.

(Continued)

9. UFM commissioning and calibration will be completed (Attachment 5 Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

10. Confirm flow normalization factors (Attachment 5, Section 1.1.G).

Prior to operating above 2546 MWt (98.4% RP).

11. Rescaling and calibration of main turbine first stage pressure input to AMSAC (Attachment 5, Sections 11.2.28, VI1.2.B, V1I1.2, and VII1.3).

Prior to operating above 2546 MWt (98.4% RP).

12. Determine EO service life for excore detectors (Attachment 5, Sections 1I1.2.A and V.1.C).

Prior to operating above 2546 MWt (98.4% RP).

13. The excore neutron detectors are scheduled to be replaced (Attachment 5,Section V.I.C).

Un~t1: fall.201 0 Refueling Outage.

Unit 2: spring 2011 Refueling Outage.

14. Revise EOP setpoints (Attachment 5 Section VI1.2.A).

Prior to operating above 2546 MWt (98.4% RP).

15. The UFM feedwater flow and temperature data will be compared to the feedwater flow venturis output and the feedwater RTD output (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

Renewed License No. DPR-32 Amendment No. 269

- 9 T.

(Continued)

16. For the applicable UFSAR Chapter 14 events, Surry 1 will re-analyze the transient consistent 2546 MWt (98.4%

RP) with VEPCO's NRC-approved reload design methodology in VEP-FRD-42, Rev. 2.1-A.

Prior to operating above 2546 MWt (98.4% RP).

If NRC review is deemed necessary pursuant to the requirements of 10 CFR 50.59, the accident analyses will be submitted to the NRC for review prior to operation at the uprate power level.

These commitments apply to the following Surry 1 UFSAR Chapter 14 DNBR analyses that were analyzed at 2546 MWt consistent with the Statistical DNBR Evaluation Methodology in VEP-NE-2-A:

• Section 14.2.7 - Excessive Heat Removal due to Feedwater System Malfunctions (Full Power Feedwater Temperature Reduction case only);

• Section 14.2.8 - Excessive Load Increase Incident;

• Section 14.2.9 - Loss of Reactor Coolant Flow; and

• Section 14.2.10 - Loss of External Electrical Load

4.

This renewed license is effective as of the date of issuance and shall expire at midnight on May 25, 2032.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

Samuel J. Collins, Director Office of Nuclear Reactor Regulation

Attachment:

Appendix A, Technical Specifications Date of Issuance: March 20, 2003 Renewed License No. DPR-32 Amendment No. 269

-3 E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

3. This renewed license shall be deemed to contain and is sUbject to the conditions specified in the folloWing Commission regulations: 10 CFR Part 20, Section 30.34 of 10 CFR Part 30, Section 40,41 of 10 CFR Part 40, Sections 50.54 and. 50;59 of 10 CFR Part 50, and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisions of the Act and the rules, 'regulations; and orders of the Commission now or hereafter In effect; and is subject to the additional conditions specified below:

A. Maximum Power Level The licensee is aU'honzeq to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained *in Appendix A, as revised through Amendment No. 268

. are hereby incorporated In thIs renewed license. The I licensee shall operate the facUity in accordance with the Technlcal*Specifications.*

C. Reports The licensee stlall make certain reports in accordance with the reqUirements of the Technical Specmcations.

D. Records The licensee shall keep facility operating records In accordance with the requirements of the Technical Specifications.

E. Deleted by Amendment 54 F. Deleted by Amendment 59 and Amendment.65 G. Deleted by Amendment 227 H. Deleted by Amendment 227 SURRY* UNIT 2 Renewed LIcense No.. OPR-37 Amendment No. 268 I

- 6 (3) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders R. As discussed in the footnote to Technical Specifications 3.23.C.2.a.1 and 3.23.C.2.b.1, the use of temporary 45-day and 14-day allowed outage time extensions to permit replacement of the Main Control Room and Emergency Switchgear Room Air Conditioning System chilled water piping shall be in accordance with the basis, risk evaluation, equipment unavailability restrictions, and compensator actions provided in the licensee's submittal dated February 26,2007 (Serial No. 07-0109) and in the associated supplemental transmittals, as approved by the NRC Safety Evaluation.

S. Upon implementation of Amendment No. 260 adopting TSTF-448, Revision 3, the determination of Main Control Room/Emergency Switchgear Room (MRC/ESGR) envelope unfiltered air inleakage as required by TS SR 4.18 in accordance with TS 6.4.R.3.a, the assessment of MCRIESGR envelope habitability as required by Specification 6.4.R.3.b, and the measurement of MCRIESGR envelope pressure as required by Specification 6.4.R.4, shall be considered met. Following implementation:

(1) The first performance of SR 4.18, in accordance with Specification 6.4.R.3.a, shall be within the specified frequency of 6 years plus the 18-month allowance of SR 4.0.2, as measured from January 18, 2004, the date of the most recent successful tracer gas test, as stated in the April 22, 2004 letter response to Generic Letter 2003-01, or within the next 18 months if the time period since the most recent successful tracer gas test is greater than 6 years.

(2) The first performance of the periodic assessment of MCRIESGR envelope habitability, Specification 6.4.R.3.b, shall be within 3 years, plus the 9-month allowance of SR 4.0.2, as measured from January 18, 2004, the date of the most recent successful tracer gas test, as stated in April 22, 2004 letter response to Generic Letter 2003-01, or within the next 9 months if the time period since the most recent successful tracer gas test is greater than 3 years.

(3) The first performance of the periodic measurement of MCRIESGR envelope pressure, Specification 6.4. R.4, shall be within 18 months, plus the 138 days allowed by SR 4.0.2, as measured from January 19, 2007, the date of the most recent successful pressure measurement test, or within 138 days if not performed previously.

Renewed License No. DPR-37 Amendment No. 268

- 7 T.

COMMITMENT SCHEDULED COMPLETION DATE

1. VEPCO will perform the final acceptance of Surry 2 uncertainty analysis to ensure the results are bounded by the statements contained in the LAR (Attachment 5, Section 1.1.0.4.1 ).

Prior to operating above 2546 MWt (98.4% RP).

2. Technical Requirements Manual Prior to operating above 2546 MWt (TRM) will be revised to include UFM administrative controls (Attachment 1 Section 3.0).

(98.4% RP).

3. Revise procedures, programs, and Prior to operating above 2546 MWt documents for the new UFM (including transducer replacement) (Attachment 5, Sections 1.1, 1.1.0.1, 1.1.H, VI1.1, V11.2.A, and VII.4).

(98.4% RP).

4. Appropriate personnel will receive Prior to operating above 2546 MWt training on the UFM and affected procedures (Attachment 5, Sections 1.1.0.1.1, VII.2.A, VI1.2.0, and V11.3)

(98.4% RP).

5. The FAC CHECWORKS SFA models Prior to operating above 2546 MWt will be updated to reflect the MUR PU conditions (Attachment 5,Section IV.1.E.iii).

(98.4% RP).

6. Simulator changes and validation will be completed (Attachment 4, Section VI1.2.C).

Prior to operating above 2546 MWt (98.4% RP).

7. Revise existing plant operating Prior to operating above 2546 MWt procedures related to temporary operation above full steady-state licenses power levels (Attachment 5,Section VII.4).

(98.4% RP).

8. Process UFSAR changes in accordance with 10 CFR 50.59 (Attachment 1, Section 3.0).

In accordance with 10 CFR 50.71(e)

Renewed License No. OPR-37 Amendment No. 268

- 8 T.

(Continued)

9. UFM commissioning and calibration will be completed (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

10. Confirm flow normalization factors (Attachment 5, Section 1.1.G).

Prior to operating above 2546 MWt (98.4% RP).

11. Rescaling and calibration of main turbine first stage pressure input to AMSAC (Attachment 5, Sections 11.2.28, VII.2.B, VII1.2, and VII1.3).

Prior to operating above 2546 MWt (98.4% RP).

12. Determine EO service life for excore detectors (Attachment 5, Sections 1I1.2.A and V.1.C).

Prior to operating above 2546 MWt (98.4% RP).

13. The excore neutron detectors are scheduled to be replaced (Attachment 5,Section V.I.C).

Unit 1: fall 2010 Refueling Outage.

Unit 2: spring 201 'I Refueling Outaqe.

14. Revise EOP setpoints (Attachment 5, Section VI1.2.A).

Prior to operating above 2546 MWt (98.4% RP).

15. The UFM feedwater flow and temperature data will be compared to the feedwater flow venturis output and the feedwater RTD output (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

Renewed License No. DPR-37 Amendment No. 268

- 9 T.

(Continued)

16. For the applicable UFSAR Chapter 14 events, Surry 2 will re-analyze the transient consistent 2546 MWt (98.4% RP) with VEPCO's NRC-approved reload design methodology in VEP-FRD-42, Rev. 2.1-A.

If NRC review is deemed necessary pursuant to the requirements of 10 CFR 50.59, the accident analyses will be submitted to the NRC for review prior to operation at the uprate power level. These commitments apply to the following Surry 2 UFSAR Chapter 14 DNBR analyses that were analyzed at 2546 MWt consistent with the Statistical DNBR Evaluation Methodology in VEP-NE-2-A:

• Section 14.2.7 - Excessive Heat Removal due to Feedwater System Malfunctions (Full Power Feedwater Temperature Reduction case only);

• Section 14.2.8 - Excessive Load Increase Incident;

• Section 14.2.9 - Loss of Reactor Coolant Flow; and

• Section 14.2.10 - Loss of External Electrical Load Prior to operating above 2546 MWt (98.4% RP).

4.

This renewed license is effective as of the date of issuance and shall expire at midnight on January 29, 2033.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

Samuel J. Collins, Director Office of Nuclear Reactor Regulation

Attachment:

Appendix A, Technical Specifications Date of Issuance: March 20, 2003 Renewed License No. DPR-37 Amendment No. 268

TS 1.0-1 1.0 DEFINITIONS The following frequently used terms. are defined for the uniform interpretation of the "

specifications.

A.

RATED POWER A steady state reactor core heat output of 2587 MWt.

B.

THERMAL POWER The total core heat transferred from the fuel to the coolant.

C.

REACTOR OPERATION

1.

REFUELING SHUTDOWN When the reactor is subcritical by at least 5% ~klk and Tavg is ~ 140° F and fuel is scheduled to be moved to or from the reactor core.

2.

COLD SHUTDOWN When the reactor is subcritical by at least 1% 6k1k and Tavg is ~200°F.

3.

INTERMEDIATE SHUTDOWN When the reactor is subcritical by at least 1.77% 6k1k and 200°F < Tavg < 547°F.

4.

HOT SHUTDOWN L

. When the reactor is subcritical by at least 1.77% 6k1k and Tavg is ~ 547°F.

Amendment Nos. 269 and 268

670 660 650 640

......... 630 Q

~

~ 620

I i!..

~ 610

~ 600 J

"ii 590 580 570 560 550 TS FIGURE 2.1-1 REACTOR CORE THERMAL AND HYDRAULIC SAFETY LIMITS THREE LOOP OPERATION. 100% FLOW T--*

.--r----

r---r--

r--

2385 psig

"""""'---r---.

~

2235

~r--

~ -r--

r--..... ~

~

--~ ~

.~._-~-

r--.-.

1985 psig r---...

~

-;ro~

~'"

r--

-r---..,'""

---......... -i"--..--0 "

1\\0.

"~ "

J I

o 10 20 30 40 50 60 70 80 90 100 110 120 Percent of Rated Thermal Power Amendment Nos. 269 and 26n

(b) High pressurizer pressure - ~2380 psig.

(c) Low pressurizer pressure - ~ 1875 psig.

(d) Overtemperature i1T where 11TO::: Indicated 11T at rated thermal power, of T = Average coolant temperature, of T' = S73.0°F P = Pressurizer pressure, psig (,

P' = 2235 psig Kj ::: 1.135 K2 ::: 0.01072 K 3 = 0.000770 111 = qt - qb' where qt and qb are the percent power in the top and bottom halves of the core respectively, and ql + qb is total core power in percent of rated power f(I1I)::: function of 111, percent of rated core power as shown in Figure 2.3-1 t) ~ 29.7 seconds t2 ~4.4 seconds The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.0% of the 11T span. (Note that 2.0% of the 11T span is equal to 3.0% 11T Power.)

(e) Overpower I1T Amendment Nos. 269 and 268

1.0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 269 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 268 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION, UNIT NOS. 1 AND 2 DOCKET NOS. 50-280 AND 50-281 INTRODUCTION By letter dated January 27,2010 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML100320264), as supplemented by letters dated February 4 (ADAMS Accession No. ML100480781) and April 29, 2010 (ADAMS Accession No. ML101200269),

Virginia Electric and Power Company (VEPCO, the licensee) submitted a request for changes to the Surry Power Station, Unit Nos.1 and 2 (Surry 1 and 2), Technical Specifications (TSs). The supplements provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the Nuclear Regulatory Commission (NRC, the Commission) staff's original proposed no significant hazards consideration determination.

The proposed change is an increase in Rated Power (RP) of approximately 1.6 percent. The proposed changes would revise the license and TSs to reflect an increase in the rated power (RP) from 2546 to 2587 MWt. The maximum electrical generator output will be revised from 842 megawatts electric (MWe) to 857 MWe (an increase of 15 MWe) for each unit. The increase is based upon increased feedwater (FW) flow measurement accuracy achieved by using high-accuracy Cameron CheckPlus' Leading Edge Flow Meter (LEFM) ultrasonic flow measurement (UFM) instrumentation. This type of application is commonly referred to as a measurement uncertainty recapture (MUR) power uprate (PU). The licensee developed the application using the guidance of NRC Regulatory Issue Summary (RIS) 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications" (ADAMS Accession No. ML013530183).

- 2 Specifically, the licensee proposes the following changes:

Description of Technical Specifications Changes Change No.

Change Description 1

Facility Operating License, Paragraph 3.A rhe licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2587 megawatts (thermal).

2 TS Section 1.0, Definitions RATED POWER RATED POWER shall be a steady state reactor core heat output of 12587 MWt.

TS Figure 2.1 Reactor Core Thermal and Hydraulic Safety Limits Three Loop Operation, 100% Flow Revised reactor core safety limit lines reflect MUR operating conditions at 12587 MWt.

TS Section 2.3.A.2.{d) - Overtemperature flT rhe OT6T pressure adjustment term, K3' is determined to equal 0.000770.

3 4

Neutron flux instrumentation is calibrated to the core thermal power which is determined by an automatic or manual calculation of the energy balance around the plant nuclear steam supply system (NSSS). This calculation is called "secondary calorimetric" for a pressurized-water reactor (PWR). The accuracy of this calculation depends primarily upon the accuracy of FW flow and FW net enthalpy measurements. The FW flow is the most significant contributor to the core thermal power uncertainty. An accurate measurement of this parameter will result in an accurate determination of core thermal power.

This MLIR PU is based on a reduced measurement uncertainty of core thermal power resulting from the installation of a Cameron LEFM CheckPlus System to measure FW flow and temperature at Surry 1 and 2. The licensee submitted Cameron (formerly Caldon) Topical Report ER-80P, Revision 0, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the System," issued March 1997 (ADAMS Legacy Library Accession No. 9903190053, Non-Public), and its supplement, Topical Report ER-157P, Revision 5, "Supplement to Topical Report ER-80P Basis for a MUR' PU with the LEFM '1/ ' or LEFM CheckPlus' System," issued October 2001 (ADAMS Accession No. ML013440134, Non-Public).

These topical reports, document the ability of the Cameron LEFM Check and CheckPlus Systems to increase the accuracy (and reduce the uncertainty) of flow measurement. The NRC-approved Topical Report ER-80P (ADAMS Legacy Library Accession No. 9903190053, Non-Public), and its supplement, ER-157P in the Safety Evaluation (SE) were issued March 1999 and December 2001 (ADAMS Accession No. ML013540256), respectively. Topical Report ER-80P describes the LEFM technology, includes calculations of power measurement uncertainty using a Cameron LEFM Check System in a typical two-loop pressurized-water reactor (PWR) or two-feedwater-line boiling-water reactor (BWR), and provides guidelines and equations for determining the plant-specific power calorimetric uncertainties. Its supplement, Topical Report ER-157N (ADAMS Accession No. ML013440078), describes the Cameron LEFM CheckPlus System and lists nonproprietary results of a typical PWR or BWR thermal power measurement uncertainty

2.0

- 3 calculation using either the Cameron LEFM Check or LEFM CheckPlus System. Together, these two reports provide a generic basis for an MUR PU.

Cameron Engineering Reports ER-650, Revision 2, "Bounding Uncertainty Analysis for Thermal Power Determination at Surry Unit 1 Using the LEFM CheckPlus System" (ADAMS Accession No. ML100321412, Non-Public), and ER-651, Revision 1, "Bounding Uncertainty Analysis for Thermal Power Determination at Surry Unit 2 Using the LEFM CheckPlus System" (ADAMS Accession No. ML100321413, Non-Public), both issued September 2009, describe the plant-specific bases for the proposed uprate at Surry 1 and 2 and were provided with the January 27,2010, license amendment request (LAR).

The NRC has recently issued similar MUR PU license amendments for North Anna Power Station, Unit Nos. 1 and 2 on October 22,2009 (ADAMS Accession No. ML092250616), the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 on July 22, 2009 (ADAMS Accession No. ML091820366),

Crystal River, Unit 3 on December 26,2007 (ADAMS Accession No. ML073600419), Vogtle Electric Generating Plant, Units 1 and 2 on February 27,2008 (ADAMS Accession No. ML080350347), Cooper Nuclear Station on June 30,2008 (ADAMS Accession No. ML081540280), and for Davis Besse Nuclear Power Station, Unit 1 on June 30,2008 (ADAMS Accession No. ML081410652).

BACKGROUND Nuclear power plants are licensed to operate at a specified maximum core thermal power, often called rated power (RP). Title 10 of the Code ofFederal Regulations (10 CFR), Part 50, AppendiX K, "ECCS [emergency core cooling system] Evaluation Models," formerly required licensees to assume that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level when performing loss-of-coolant accident (LOCA) and ECCS analyses. This requirement was included to ensure that instrumentation uncertainties were adequately accounted for in the analyses. In practice, many of the design-bases analyses assume a 2 percent power uncertainty, consistent with 10 CFR Part 50, AppendiX K.

A revision to 10 CFR Part 50, Appendix K, effective on July 31,2000, allows licensees to use a power level less than 1.02 times the RP, but not less than the licensed power level, based on the use of state-of-the-art FW flow measurement devices that provide a more accurate calculation of power. Licensees can use a lower uncertainty in the LOCA and ECCS analyses provided the licensee has demonstrated that the proposed value adequately accounts for instrumentation uncertainties. Because there continues to be substantial conservatism in other Appendix K requirements, sufficient margin to ECCS performance in the event of a LOCA is preserved.

However, the final rule by itself did not allow increases in licensed power levels. Because the licensed power level for a plant is a TS limit, proposals to raise the licensed power level must be reviewed and approved under the license amendment process. Surry 1 and 2 were initially licensed to operate at a maximum of 2441 MWt. In Amendment Nos. 203 and 203 (ADAMS Accession No. ML012710328), for licenses DPR-32 and DPR-37, dated August 3, 1995, the NRC approved Surry 1 and 2 operation at the current power level of 2546 MWt. The proposed MUR PU is based on a redistribution of analytical margin which originally required of ECCS evaluation models performed per the requirements of 10 CFR Part 50, Appendix K, originally mandated 102 percent of licensed power level for light-water reactor emergency core cooling system (ECCS) evaluation models. The NRC approved a change to Title 10 of the Code of Federal Regulations

- 4 (10 CFR), Part 50, Appendix K requirements on June 1, 2000, effective July 31, 2000. This change provided licensees the option of maintaining the 2-percent power margin between licensed power level and the ECCS evaluation assumed power level, or applying a reduced ECCS evaluation margin based on an accounting of uncertainties due to instrumentation error.

Implementing the FW UFM (Cameron LEFM CheckPlus System) is found to be an effective way to obtain additional plant power without significantly affecting plant design margins. The FW flow measurement uncertainty is the most significant contributor to core power measurement uncertainty. The UFM provides a more accurate measurement of FW flow and thus reduces the uncertainty in the FW flow measurement. This reduced uncertainty, in combination with other uncertainties, results in an overall power level measurement uncertainty of 0.35 percent at RP.

The UFM will provide online main FW flow and temperature measurement to determine reactor thermal power. This system uses acoustic energy pulses to determine the main FW mass flow rate and temperature. The UFM consists of a measuring section containing 16 ultrasonic multi-path transit time transducers, one dual resistance temperature detector (RTD), and two pressure transmitters installed in each of the three FW lines, and an electronic signal processing cabinet.

The UFM will be used in lieu of the current feed flow venturi indication and RTD temperature indication to perform the plant calorimetric measurement calculation. The currently installed steam or feed flow instruments will continue to provide inputs to other indication, protection and control systems, and will be used if the UFM is not functional.

The NRC staff finds that the Cameron LEFM-assisted core thermal power measurement uncertainty is limited to 0.35 percent of actual reactor thermal power and, therefore, can support the proposed 1.6 percent MUR PU. This results in the proposed increase of 1.6 percent in the Surry 1 and 2 license power level using current NRC-approved methodologies.

3.0 TECHNICAL EVALUATION

3.1 Human Factors 3.1.1 Regulatory Evaluation NRC's human factors reviews address programs, procedures, training, and plant design features related to operator performance during normal and accident conditions. The NRC staff's human factors evaluation was conducted to confirm that operator performance would not be adversely affected as a result of system and procedure changes made to implement the proposed MUR PU.

The scope of the review included changes to operator actions, human-system interfaces, and procedures and training needed for the proposed MUR PU.

3.1.2 Technical Evaluation The NRC staff has developed a standard set of questions for human factors reviews

( RIS 2002-03, Requested Information (RI), Attachment 5,Section VII, Items 1 through 4) of MUR PUs. The following sections evaluate the licensee's response to these questions in the license amendment request (LAR).

- 5 3.1.2.1 Operator Actions The licensee stated in its submittal that the existing required operator actions are not affected by the MUR PU. The licensee also stated that there is no reduction in the time required for the necessary operator actions. The licensee also determined that there were no new manual actions or automation of existing operator actions required as a result of the MUR PU. Operator actions included in the safety analyses were reviewed for potential MUR PU impact. The following design-basis events were reviewed:

Appendix R Fire Updated Final Safety fA,nalysis Report (UFSAR)

Section 9.1 0 Boron Dilution UFSAR Section 14.2.5 SGTR [Steam Generator Tube Rupture]

UFSAR Section 14.3.1 Rupture of Main Steam Pipe UFSAR Section 14.3.2 Fuel-Handling Accident UFSAR Section 14.4.1 VCT Rupture [Valve Control Tank]

UFSAR Section 14.4.2.1 Large Break Loss-of-Coolant Accident (LOCA) UFSAR Section 14.5.1 SBLOCA [Small Break LOCA]

UFSAR Section 14.5.2 Loss of Normal Feedwater UFSAR Section 14B.6 The licensee also stated that there is no reduction in the time required for the operator actions.

The licensee's analysts also determined that no new manual actions or automation of existing operator actions will be required as a result of the MUR PU.

The NRC staff has reviewed the licensee's statements in its submittal relating to any impacts of the MUR PU to existing or new operator actions credited in the safety analyses. The NRC staff concludes that the proposed MUR PU will not adversely impact operator actions and its response times because there were no changes identified. The NRC staff finds that the statements provided by the licensee are in conformance with Section VI1.1 of Attachment 5 of the LAR, RIS 2002-03, RI.

3.1.2.2 Emergency and Abnormal Operating Procedures VEPCO stated in its submittal that the current emergency operating procedures (EOPs) and abnormal operating procedures (AOPs) were reviewed for potential changes due to the proposed MUR PU. VEPCO concluded that no changes to operator actions, timing, or sequence are required to support the MUR PU. The licensee identified EOP setpoints that require revision.

These setpoints will be revised to reflect the new total core power. VEPCO stated that operator training on the procedure changes will be provided as a part of the MUR PU implementation.

The NRC staff concludes that the proposed MUR PU does not present any adverse impacts to the EOPs and AOPs. This conclusion is based upon two licensee statements: (1) VEPCO will revise the EOPs and AOPs to reflect the new power level and revised setpoints, and (2) the minor

- 6 changes being made to the EOPs and AOPs will be reflected in the operator training program prior to MUR PU implementation. Additionally, VEPCO will revise the existing plant operating procedure related to temporary operation above full steady-state licensed power levels.

Precautions will be revised to account for the uprate power level.

The NRC staff finds that the statements provided by VEPCO are in conformance with Sections VI1.2.A, VI1.3, and VilA of Attachment 5 of the LAR, RIS 2002-03, RI.

3.1.2.3 Changes to Control Room Controls, Displays, and Alarms In its submittal, VEPCO described changes to control room controls, displays (including the safety parameter display system (SPDS)), and alarms related to the proposed MUR PU.

Notable proposed changes to controls, displays, and alarms include:

Instruments associated with turbine first stage pressure will require scaling change Instrument loops are affected by the MUR PU (indicator replacement, calibration span, and/or scaling)

Plant computer points will be added and/or changed for the revised calorimetric algorithm and the FW ultrasonic flow meter The new UFM electronic cabinet, located in the cable spreading room, is used to display and control aspects of FW flow data. The display provides system status or monitored process parameters. The display is typically used for maintenance purposes and not for control of plant operation The system alerts operations personnel of UFM trouble through the main control room overhead annunciator "Feedwater Ultrasonic Flow Meter Trouble." The main control room overhead annunciator "Feedwater Ultrasonic Flow Meter Failure" alerts the operators when the system loses a plane of operation, suffers a loss of alternating current (AC) power or other total failure. Any UFM condition that increases FW flow uncertainty is considered a "Feedwater Ultrasonic Flow Meter Failure" alarm condition No significant SPDS changes are anticipated as a result of the MUR PU. Critical safety function status trees will be reviewed and revised All changes to the control room will be reflected in the operator training program prior to MUR PU implementation.

The NRC staff has reviewed the licensee's evaluation of the proposed changes to the control room. The NRC staff concludes that the proposed changes do not present any adverse effects to the operators' functions in the control room. VEPCO committed to making all modifications to the control room and providing training on these changes prior to MUR PU implementation.

The NRC staff finds that the statements provided by VEPCO are in conformance with Sections V11.2.B and VI1.3 of Attachment 5 of the LAR, RIS 2002-03, RI.

3.1.204 Control Room Plant Reference Simulator and Operator Training VEPCO stated that potential simulator changes will be identified as part of the plant modification process. The submittal also stated that these modifications will be evaluated, implemented and tested per the approved plant procedures. The submittal stated that simulator fidelity will be

- 7 revalidated using the approved plant procedures. The licensee stated that these modifications will be completed in time to support operator training prior to the MUR PU implementation.

VEPCO stated in its submittal that operator training will be developed and the operations staff will be trained on the plant modifications, TSs, technical requirements manual (TRM), and procedural changes prior to MUR PU implementation. The licensee stated that these changes, along with the plant simulator modifications, will be made prior to MUR implementation.

The NRC staff has reviewed VEPCO's evaluation of proposed changes to the operator training and plant simulator related to the MUR PU. The licensee concluded that the changes are appropriate and do not present any adverse effects on the plant simulator or the operator training program. VEPCO committed to making all modifications to the plant simulator and incorporating these changes into the operator training program prior to MUR PU implementation. The NRC staff finds that the statements provided by VEPCO are in conformance with Sections V11.2.C, V11.2.D, and VI/.3 of Attachment 5 of the LAR, RIS 2002-03, RI.

3.1.2.5 Conclusion The NRC staff has completed its human factors review of VEPCO's proposed changes and conclude that the licensee has adequately considered the impact of the proposed MUR PU on operator actions, EOPs and AOPs, control room components, the plant simulator and operator training programs.

3.2 Dose Consequences Analysis 3.2.1 Regulatory Evaluation RIS 2002-03, recommends that to improve the efficiency of the NRC staff's review, licensees requesting an MUR PU should identify existing design-basis accident (DBA) analyses of record which bound plant operation at the proposed uprated power level. For any existing DBA analyses of record that do not bound the proposed uprated power level, the licensee should provide a detailed discussion of the reanalysis.

This safety evaluation (SE) input addresses the impact of the proposed changes on analyzed DBA radiological consequences. Previously, in Amendment Nos. 230 and 230 for licenses DPR-32 and DPR-37 of March 8, 2002 (ADAMS Accession No. ML020710159), Surry 1 and 2 was granted implementation of a full-scope alternative source term in accordance with 10 CFR 50.67, "Accident source term," and following the guidance of Regulatory Guide (RG) 1.183 (ADAMS Accession No. ML003716792), "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." Therefore, this evaluation has been conducted to verify that the results of the licensee's DBA radiological dose consequence analyses continue to meet the dose acceptance criteria given in 10 CFR 50.67 for offsite doses and 10 CFR Part 50, Appendix A, general design criterion (GDC) 19, "Control room" (or equivalent for plants licensed before the GDC were in existence) with respect to control room habitability. The applicable acceptance criteria are 5 roentgen equivalent man (rem) Total Effective Dose Equivalent (TEDE) in the control room (CR), 25 rem TEDE at the exclusion area boundary (EAB),

and 25 rem TEDE at the outer boundary of the low population zone (LPZ). Except where the licensee proposed a suitable alternative, the NRC staff utilized the regulatory guidance provided in applicable sections of RG 1.183 and NUREG-0800 (ADAMS Accession No. ML051580396),

"Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power

- 8 Plants: LWR [Light-Water Reactor] Edition," Chapter 15, for design-basis accidents, and SRP Chapter 6.4, for control room habitability, in performing this review.

3.2.2 Technical Evaluation The NRC staff reviewed the regulatory and technical analyses performed by the licensee in support of its proposed MUR PU license amendment, as they relate to the radiological consequences of DBA analyses. Information regarding these analyses was provided by the licensee in Attachment 5 of the LAR. The findings of this SE are based solely on the descriptions and results of the licensee's analyses and other supporting information docketed by the licensee.

The NRC staff reviewed the impact of the proposed 1.6 percent MUR PU on DBA radiological consequence analyses, as documented in Chapter 14 of the Surry 1 and 2 Updated Final Safety Analysis Report (UFSAR). The specific DBA analyses that were reviewed were as follows:

Loss-of-Coolant Accident (LOCA)

Fuel-Handling Accident (FHA)

Main Steam Line Break (MSLB) Accident Locked-Rotor Accident (LRA)

Steam Generator Tube Rupture (SGTR) Accident In the LAR submittal, the licensee stated that the current dose analyses of record for the FHA and LOCA were performed at 2605 MWt, or 102.3 percent of the currently licensed thermal power (CLTP) of 2546 MWt. The licensee also stated that the current dose analysis of record for the LRA was performed with steam releases consistent with 2596.9 MWt, or 102 percent of the CLTP and a source term consistent with 2605 MWt, or 102.3 percent of the CLTP of 2546 MWt. The licensee also stated that the current dose analyses of record for the SGTR and MSLB were performed at 2596.9 MWt, or 102 percent of the CLTP. The assumptions used in the licensee's dose consequence analyses associated with the LOCA, FHA, and LRA are conservative such that they result in the determination of bounding consequences when compared to those doses which would be assessed at the MUR uprated power. Based upon licensing basis documentation and DBA descriptions in the Surry 1 and 2 UFSAR, the NRC staff has confirmed that this is indeed the case for the three aforementioned DBAs. Therefore, for these three DBAs, the current analyses bound the analyses that would be performed at the proposed MUR uprated power level of 2587 MWt, as the current power assumptions are either at least 100.38 percent of the proposed uprated power or unrelated to the resulting dose consequence. The 0.38-percent margin is within the assumed uncertainty associated with advanced flow measurement techniques, including use of the Cameron CheckPlus' LEFM System credited by the licensee.

For the Surry 1 and 2 design-basis MSLB and SGTR accidents, the licensee determined that the reactor coolant system (RCS) source term needed to be revised to reflect changes to the fuel management scheme (18-month fuel cycle length), core power, and the resulting relative concentration of specific radioactive isotopes that contribute to the TS-controlled dose equivalent (DE) 1-131 activity concentration. As a Surry 1 and 2 design basis, no fuel failure is postUlated for these two accidents. Typically, for accidents involving only RCS activity, a licensee will base the source term upon those isotopes present in the coolant that are controlled by the plant TSs (e.g.,

iodine and noble gases), in accordance with the guidance of RG 1.183. However, in the VEPCO analysis of both the MSLB and SGTR, the licensee accounted for far more fission product

- 8 Plants: LWR [Light-Water Reactor] Edition," Chapter 15, for design-basis accidents, and SRP Chapter 6.4, for control room habitability, in performing this review.

3.2.2 Technical Evaluation The NRC staff reviewed the regulatory and technical analyses performed by the licensee in support of its proposed MUR PU license amendment, as they relate to the radiological consequences of DBA analyses. Information regarding these analyses was provided by the licensee in Attachment 5 of the LAR. The findings of this SE are based solely on the descriptions and results of the licensee's analyses and other supporting information docketed by the licensee.

The NRC staff reviewed the impact of the proposed 1.6 percent MUR PU on DBA radiological consequence analyses, as documented in Chapter 14 of the Surry 1 and 2 Updated Final Safety Analysis Report (UFSAR). The specific DBA analyses that were reviewed were as follows:

Loss-of-Coolant Accident (LOCA)

Fuel-Handling Accident (FHA)

Main Steam Line Break (MSLB) Accident Locked-Rotor Accident (LRA)

Steam Generator Tube Rupture (SGTR) Accident In the LAR submittal, the licensee stated that the current dose analyses of record for the FHA and LOCA were performed at 2605 MWt, or 102.3 percent of the currently licensed thermal power (CLTP) of 2546 MWt. The licensee also stated that the current dose analysis of record for the LRA was performed with steam releases consistent with 2596.9 MWt, or 102 percent of the CLTP and a source term consistent with 2605 MWt, or 102.3 percent of the CLTP of 2546 MWt. The licensee also stated that the current dose analyses of record for the SGTR and MSLB were performed at 2596.9 MWt, or 102 percent of the CLTP. The assumptions used in the licensee's dose consequence analyses associated with the LOCA, FHA, and LRA are conservative such that they result in the determination of bounding consequences when compared to those doses which would be assessed at the MUR uprated power. Based upon licensing basis documentation and DBA descriptions in the Surry 1 and 2 UFSAR, the NRC staff has confirmed that this is indeed the case for the three aforementioned DBAs. Therefore, for these three DBAs, the current analyses bound the analyses that would be performed at the proposed MUR uprated power level of 2587 MWt, as the current power assumptions are either at least 100.38 percent of the proposed uprated power or unrelated to the resulting dose consequence. The 0.38-percent margin is within the assumed uncertainty associated with advanced flow measurement techniques, including use of the Cameron CheckPlus ' LEFM System credited by the licensee.

For the Surry 1 and 2 design-basis MSLB and SGTR accidents, the licensee determined that the reactor coolant system (RCS) source term needed to be revised to reflect changes to the fuel management scheme (18-month fuel cycle length), core power, and the resulting relative concentration of specific radioactive isotopes that contribute to the TS-controlled dose equivalent (DE) 1-131 activity concentration. As a Surry 1 and 2 design basis, no fuel failure is postulated for these two accidents. Typically, for accidents involving only RCS activity, a licensee will base the source term upon those isotopes present in the coolant that are controlled by the plant TSs (e.g.,

iodine and noble gases), in accordance with the guidance of RG 1.183. However, in the VEPCO analysis of both the MSLB and SGTR, the licensee accounted for far more fission product

- 9 radionuclides. Of the greatest impact, was the licensee's inclusion of many alkali metal isotopes.

Nevertheless, the assumption by the licensee is conservative and therefore acceptable to the NRC staff. In addition to the source term adjustments, the licensee revised their analyses of the MSLB and SGTR accidents to reflect EAB x/a values (radiological atmospheric dispersion factor) previously approved by the NRC staff in an SE dated August 10, 2006 (ADAMS Accession No. ML062220194) and updated post-accident steam flows from the steam generator (SG) relief valves to incorporate newly calculated relief valve flow data as a function of steam pressure.

The licensee's reanalysis of the MSLB and SGTR accidents determined new design basis dose consequences for both the pre-accident iodine spike and concurrent iodine spike scenarios of both accidents. The following table presents those doses:

Control Room EAB LPZ Total Dose Acceptance Criteria Total Dose Acceptance Criteria Total Dose Acceptance Criteria (rem TEDE)

(rem TEDE)

(rem TEDE)

(rem TEDE)

(rem TEDE)

(rem TEDE)

MSLB PIS 1.4E+00 5.0 4.0E-01 25 1.0E-01 25 CIS 1.6E+00 5.0 5.0E-01 2.5 1.0E-01 2.5 SGTR PIS 4.3E+00 5.0 1.2E+00 25 2.0E-01 25 CIS 1.3E+00 5.0 1.7E+00 2.5 2.0E-01 2.5 Upon review of the accident analysis summaries, the NRC staff concluded that the licensee has appropriately applied acceptable EAB x/a values to postulated releases of airborne reactor coolant system (RCS) coolant activity resulting from the design basis MSLB and SGTR accidents.

3.2.3 Conclusion As described above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to reassess the radiological consequences of the postulated DBA with the proposed uprated power level. The NRC staff finds that the licensee will continue to meet the applicable dose acceptance criteria, as identified in Section 3.2 of this evaluation, following implementation of the proposed 1.6-percent MUR PU. The NRC staff further finds reasonable assurance that Surry 1 and 2, as modified by this approved license amendment, will continue to provide sufficient safety margins, with adequate defense-in-depth, to address unanticipated events and to compensate for uncertainties in accident progression, analysis assumptions, and input parameters. Therefore, the proposed license amendment is acceptable with respect to the radiological dose consequences of the DBAs.

3.3 Fire Protection 3.3.1 Regulatory Evaluation The purpose of the fire protection program is to provide assurance, through a defense-in-depth design, that a fire will not prevent the performance of necessary plant safe-shutdown functions

3.3.2

- 10 nor will it significantly increase the risk of radioactive releases to the environment. The NRC staff's review focused on the effects of the increased decay heat due to the MUR PU on the plant's safe-shutdown analysis to ensure that structures, systems, and components (SSCs) required for the safe shutdown of the plant are protected from the effects of the fire and will continue to be able to achieve and maintain safe shutdown following a fire. The NRC's acceptance criteria for the fire protection program are based on (1) 10 CFR 50.48, "Fire Protection," insofar as it requires the development of a fire protection program to ensure, among other things, the capability to safely shutdown the plant; (2) GDC 3 of Appendix A to 10 CFR Part 50, "Fire Protection," insofar as it requires that [a] SSCs important to safety be designed and located to minimize the probability and effect of fires, [b] noncombustible and heat resistant materials be used, and fire detection and suppression systems be provided and designed to minimize the adverse effects of fires on SSCs important to safety; and (3) GDC 5 of Appendix A to 10 CFR Part 50, "Sharing of structures, systems, and components," insofar as it requires that SSCs important to safety not be shared among nuclear power units unless it can be shown that sharing will not significantly impair their ability to perform their safety functions.

Technical Evaluation The LAR was consistent with the guidelines in RIS 2002-03. In the LAR, the licensee re-evaluated the applicable SSCs and safety analyses at the proposed MUR core power level of 2587 MWt against the previously analyzed core power level of 2546 MWt.

The NRC staff reviewed Attachment 5 of the LAR, Section 11.2.31, "Safe Shutdown Fire Analysis (Appendix R Report) UFSAR 9.1," Section VI1.6.A, "Fire Protection Program," and Section VI1.6.A.i, "Fire Protection Systems". The NRC staff also reviewed the licensee's commitment to 10 CFR 50.48 (Le., approved fire protection program). The review covered the impact of the proposed MUR PU on the results of the safe-shutdown fire analysis as noted in Sections II and III, of the LAR, RIS 2002-03. The review focused on the effects of the MUR PU on the post-fire safe-shutdown capability and increase in decay heat generation following plant trips.

In a letter dated April 14, 2010 (ADAMS Accession No. ML101020452), the NRC staff issued a request for additional information (RAI). The NRC staff noted that Attachment 5 to the LAR, Section 11.2.31, "Safe Shutdown Fire Analysis (Appendix R Report) UFSAR 9.1," states that

"...Operator actions in response to an Appendix R fire are not adversely impacted. The MUR PU does not affect the worst case fire location or the post-fire local operations and capability to complete repairs..." The NRC staff requested the licensee to verify that (1) the MUR PU will not require any new operator actions; (2) any effects from additional heat in the plant environment from the increased power will not prevent required post fire operator manual actions, as identified in the Surry 1 and 2 fire protection program, from being performed at and within their designated time; and (3) procedures and resources necessary for systems required to achieve and maintain safe shutdown will not change and are adequate for the MUR PU. In a letter to the NRC dated April 29, 2010 (ADAMS Accession No. ML101200269), the licensee provided additional information outlined in this paragraph in response to the above RAI. In its response, the licensee stated that the existing operator actions are not affected and no reduction in operator action time was identified. Further, the licensee indicated that no new operator actions were identified and no eXisting manual actions were automated. The licensee stated that the there are minor changes in temperature and pressure conditions, but these will not diminish the ability to perform post-fire operator actions. Further, the licensee has analyzed procedures and resources to achieve and maintain post-fire shutdown conditions and concluded that they are still adequate for plant safe

- 11 shutdown for the MUR PU conditions.

The licensee's response satisfactorily addresses the NRC staff's concerns, and this RAI issue is considered resolved based on the following: The licensee indicated that proposed MUR PU does not impact the required operator manual action times since the MUR PU does not create any adverse environmental condition which would impact performance of operator manual actions.

The response times are not impacted for the post MUR PU conditions nor is operator access affected to those areas where the actions must be performed. Further, the licensee has evaluated procedures and resources for systems required to achieve and maintain safe shutdown as a result of the MUR PU and determined that the procedures and resources do not adversely impact the post-fire safe-shutdown capability of the plant. Therefore, the NRC staff finds the response to the RAI acceptable because the licensee evaluated and determined no impact on the existing procedures and resources for post-fire safe-shutdown capability by the proposed MUR PU.

In a letter to the NRC dated April 14, 2010 (ADAMS Accession No. ML101020452), the NRC staff noted that Attachment 5 to the LAR, Section VI1.6.A.i, "Fire Protection Systems," states that

"...The fire protection subsystems remain unchanged as a result of the MUR power uprate..."

However, this section does not discuss the changes in the physical plant configuration related to the fire protection program or changes to the combustible loading at MUR PU conditions. The NRC staff requested the licensee clarify whether this request involves any changes in plant configurations related to the fire protection program or changes to the combustible loading. If any, the staff requested the licensee to identify proposed changes and discuss the impact of these changes on the plant's compliance with the fire protection program licensing basis, 10 CFR 50.48, or applicable portions of 10 CFR Part 50, Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979."

In a letter to the NRC dated April 29, 2010 (ADAMS Accession No. ML101200269), the licensee provided additional information outlined in this paragraph in response to the above RAI. In its response, the licensee stated that no changes to the installed fire protection systems are required as a result of the MUR PU mechanical and electrical installation design modifications. The mechanical portion of the installations does not affect combustible loading. The physical requirements and response of the station fire protection systems remain unchanged. The modifications implemented by these design change packages will not adversely impact the Station's design basis for compliance with Appendix R to 10 CFR Part 50. The mechanical installation will have a slight effect on one scenario assumed for the Surry 1 and 2, B.5.b response, but the overall scenario and its goals are unchanged by these modifications. The licensee indicated that increases in combustible loading (22-pound cables) in the main control room, computer room, and cable spreading room will not adversely impact the plant's ability to achieve and maintain safe shutdown at the increased MUR power level.

The licensee's response satisfactorily addresses the NRC staffs concerns, and this RAI issue is considered resolved based on the following. The licensee's review indicates no change to the installed fire protection system for the MUR PU condition. The licensee's review also indicates that the increase in combustible loading (cables) in the main control room (CR), computer room, and cable spreading room as result of the MUR PU will not affect plant compliance to Appendix R to 10 CFR Part 50. Therefore, the NRC staff finds the response to the RAI acceptable because these changes do not impact fire protection features and post-fire safe-shutdown capability.

- 12 Based on the licensee's fire-related safe-shutdown assessment and responses to the RAls, the NRC staff concludes that the licensee has adequately accounted for the effects of the 1.6-percent increase in decay heat on the ability of the required systems to achieve and maintain safe-shutdown conditions. The NRC staff finds this aspect of the capability of the associated SSCs to perform their design basis functions at an increased core power level of 2587 MWt acceptable with respect to fire protection.

3.3.3 Conclusion Based on its review, the NRC staff has concluded that the proposed MUR PU will not have a significant impact on the fire protection program or post-fire safe-shutdown capability and, therefore, finds the proposed amendment acceptable.

3.4 Chemical Engineering 3.4.1 Flow-Accelerated Corrosion 3.4.1.1 Regulatory Evaluation Flow-accelerated corrosion (FAC) is a corrosion mechanism occurring in carbon steel components exposed to flowing single-or two-phase water. Components made from stainless steel are immune to FAC, and FAC is significantly reduced in components containing small amounts of chromium or molybdenum. The rates of material loss due to FAC depend on flow velocity, fluid temperature, steam quality, oxygen content, and pH. During plant operation, flexibility to control these parameters to minimize FAC is limited. Loss of material by FAC will therefore, occur. The NRC staff has reviewed the effects of the proposed MUR PU on FAC and the adequacy of the licensee's FAC program. The intent of the FAC program is to predict the rate of loss so that repair or replacement of damaged components can be made before they reach critical thickness. The licensee's FAC program is based on NRC Generic Letter (GL) 89-08, "Erosion/Corrosion -Induced Pipe Wall Thinning," May 1989 (ADAMS Accession No. ML031200731), Electric Power Research Institute Report NSAC-202L, "Recommendations for an Effective Flow-Accelerated Corrosion Program," and the guidelines in the Institute of Nuclear Power Operations (INPO) EPG-06, "INPO Engineering Guide - Flow Accelerated Corrosion." It consists of predicting loss of material using the CHECWORKS steam or feedwater application (SFA) FAC monitoring computer code, visual inspection, and volumetric examination of the affected components. The NRC's acceptance criteria are based on the structural evaluation of the minimum acceptable wall thickness for the components undergoing degradation by FAC.

3.4.1.2 Technical Evaluation The licensee utilizes a FAC program that provides a standardized method of identifying, inspecting, and tracking components susceptible to FAC wear in both single-and two-phase flow conditions. Aspects of the program include: FAC susceptibility analysis and modeling; FAC inspection and evaluation; operational experience reviews; cross-over and -under main steam piping; moisture separators and reheaters inspections and evaluations. The licensee indicated that the criteria used to determine system susceptibility to FAC include: material, moisture content, temperature, dissolved oxygen, frequency of system usage, plant-specific operating experience, and industry operating experience.

- 13 The licensee stated the CHECWORKS SFA FAC monitoring computer code is utilized in predicting and tracking FAC susceptible components. Additionally, the computer code is used to create unit-specific databases, which are used for analyses and data interpretation. The licensee stated that analytical models are used in wear rate analysis, which ranks components in order of predicted FAC wear and predicted time to reach minimum code wall thickness. The licensee further stated that the CHECWORKS SFA models will be updated to meet the changes associated with the MUR PU.

The licensee provided two tables that summarized the impact that the proposed MUR PU had on remaining service life as a result of the increase in wear rates. The licensee reported that the 4th-point extraction pipe in Surry Unit 1 and the 5th-point extraction pipe in Surry Unit 2 in the extraction steam systems were predicted to have the greatest increase in wear rates. The predicted increase in wear rates are 37 and 17 percent in Surry 1 and 2, respectively. The tables contained representative piping components in the five systems expected to experience the greatest increase in FAC wear as a result of the MUR PU. The licensee further stated that any piping components with a low or a negative time for remaining service life will be evaluated for future inspection. The licensee stated that the CHECWORKS SFA database for Surry 1 and 2 will be updated to account for MUR PU conditions. Based 011 reviews conducted, the licensee stated that there is no significant impact due to the MUR PU. The NRC staff finds this acceptable.

The NRC staff has reviewed the licensee's evaluation and confirms that the applicable regulatory guidance was followed. The NRC staff has also reviewed the calculation results in the application.

The licensee has demonstrated that the FAC program is adequate for managing the potential effects on the piping components susceptible to FAC. The NRC staff's acceptance of the licensee's use of CHECKWORKS can be found in NUREG-1766, "Safety Evaluation Report Related to the License Renewal of North Anna Power Station, Units 1 and 2, and Surry Power Station, Units 1 and 2," December 2002 (ADAMS Accession Nos. ML030160804, ML030160825, ML030160848). The NRC staff finds that the FAC program is adequate in predicting the rate of material loss.

3.4.1.3 Conclusion The NRC staff has reviewed the licensee's evaluation of the effect of the proposed MUR PU on the FAC analysis for the plant and concludes that the licensee has adequately addressed the impact of changes in the plant operating conditions on the FAC analysis.

3.4.2.

Coatings 3.4.2.1 Regulatory Evaluation Protective coating systems (paints) protect the surfaces of facilities and equipment from corrosion and radionuclide contamination. Protective coatings systems also provide wear protection during plant operation and maintenance activities. The NRC staff's review covered protective coating systems used inside containment, including the coating's suitability for, and stability under, design-basis loss-of-coolant accident (DBLOCA) conditions, considering radiation and chemical effects. The NRC's acceptance criteria for protective coating systems are based on (1) 10 CFR, Part 50, Appendix B, "Quality Assurance Criteria For Nuclear Power Plants and Fuel Reprocessing Plants," and (2) Regulatory Guide (RG) 1.54, Revision 1, "Service Levell, II, and III Protective Coatings Applied to Nuclear Power Plants," JUly 2000 (ADAMS Accession No.

- 14 ML003714475). Specific review criteria are contained in Standard Review Plan (SRP) Section 6.1.2, "Protective Coating Systems (Paints) - Organic Materials Review Responsibilities,"

(ADAMS Accession No. ML052070455).

3.4.2.2 Technical Evaluation The licensee stated that the coating systems used inside containment are used to protect equipment and structures from corrosion and radionuclide contamination. It was further stated that the coatings also provide wear protection during plant operation and maintenance activities.

The Surry 1 and 2 containment Service Level I coatings are qualified to withstand a LOCA environment and meet American National Standards Institute (ANSI) Standards N5.12, "Protective Coatings (Paints) for the Nuclear Industry," N1 01.2, "Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities," and N101.4, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities."

The licensee stated that the UFSAR LOCA containment response analyses remain bounding for MUR PU conditions. The licensee stated that there were no changes to the containment analyses that would require a change to the containment design pressure and temperature. Therefore, since the containment design pressure and temperature limits are not changing, the licensee concluded that the Service Level I containment coatings remain qualified under MUR PU conditions. The NRC staff finds this evaluation acceptable.

The NRC staff has reviewed the licensee's evaluation and has confirmed that the applicable regulatory guidance was followed. The NRC staff finds that the coatings will not be adversely impacted by the MUR PU and that temperature and pressure limits under MUR PU conditions continue to be bounded by the conditions to which the coatings were qualified.

3.4.2.3 Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed MUR PU on protective coating systems. The NRC staff concludes that the licensee has appropriately addressed the impact of changes in conditions following a DBLOCA and their effects on the protective coatings. The NRC staff further concludes that the licensee has demonstrated that the protective coatings will continue to be acceptable following implementation of the proposed MUR PU. Specifically, the protective coatings will continue to meet requirements of 10 CFR Part 50, Appendix B. Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to protective coating systems.

Steam Generator Program 3.4.3.1 Regulatory Evaluation Steam generator (SG) tubes constitute a significant part of the reactor coolant pressure boundary (RCPB). The SG inservice inspections provide a means for assessing the structural and leak-tight integrity of SG tubes through periodic inspection and testing of critical areas and features of the tubes. The NRC staff reviewed the effects of changes in differential pressure, temperature, and flow rates resulting from the proposed MUR PU on the design and operation of SGs. Specifically, the NRC staff evaluated whether changes to these parameters continue to be bounded by those considered in the plant design and licensing basis (i.e., the TS plugging limits).

3.4.3

- 15 3.4.3.2 Technical Evaluation Surry 1 and 2 each have three Westinghouse model 51 F SGs (i.e., A, B, and C) that were installed in 1981 and 1980, respectively for Units 1 and 2. Each SG nominally contains 3,342 thermally treated Alloy 600 tubes. Each tube has a nominal outside diameter of 0.875 inches and a nominal wall thickness of 0.050 inches. The tubes were hydraulically expanded at both ends for the full length of the tubesheet and are supported by a number of stainless steel tube support plates. The U-bends of the tubes installed in rows 1 through 8 were thermally stress-relieved after bending.

The licensee performed a thermal-hydraulic evaluation that focused on changes to the secondary-side operating characteristics at MUR PU conditions. It was reported that the proposed MUR PU will potentially affect SG secondary-side performance characteristics which include steam pressure and flow, circulation ratio, bundle mixture flow, heat flux, secondary-side pressure drop, moisture carryover, hydrodynamic stability, and secondary-side mass. The licensee determined that the operating characteristics remain acceptable for the MUR PU. Based on its review of the licensee's thermal hydraulic evaluation, the NRC staff finds the thermal hydraulic operating characteristics acceptable.

The licensee performed a structural evaluation that focused on the critical SG components as determined by the design analysis stress ratios and fatigue usages. The SG primary-side components were evaluated to determine the impact of the uprated conditions' differential pressures. The licensee stated that the scale factors used for the primary-side components were the ratios of the baseline condition primary-to-secondary side differential pressures to the uprated conditions' differential pressures. The scale factors were applied conservatively to both the thermal and pressure stresses. The decrease in pressure on the secondary-side components at the uprated conditions was used in determining the applicable scale factors. The licensee calculated the stress increase due to the pressure reduction. The stress increase was then used in calculating the resulting fatigue usage changes for operation at uprated conditions. The scale stresses were used in determining the stress ranges involving transients associated with full power.

The licensee performed an analysis to determine if the American Society of Mechanical Engineering (ASME), Boiler and Pressure Vessel Code (ASME Code) limits on design primary-to-secondary ~P [differential pressure] are exceeded for any applicable transient at MUR PU conditions. The licensee reported that the maximum primary to secondary-side differential pressures are below the applicable design pressure limits of 1600 pounds per square inch differential (psid) and 1760 psid for normal and postulated transient conditions, respectively. As a result, the licensee determined that the ASME Code design pressure requirements are satisfied.

The NRC staff finds this evaluation acceptable.

The licensee performed an evaluation on the primary-plus-secondary stress range for primary-side and secondary-side components. The primary-plus-secondary stress was determined to meet the three times mean-stress limit of the ASME Code limits. Additionally, the cumulative usage factors for affected components remain below 1.0, which satisfies the ASME Code. The NRC staff finds this acceptable.

The licensee stated that outside diameter stress corrosion cracking and pitting are potential tube

- 16 degradation mechanisms that may affect SG tubes as a result of the MUR PU. Accordingly, the current operating and maintenance practices are applicable and adequate in minimizing the degradation due to these mechanisms. Based on its review of the impact of SG chemistry control on potential tube degradation mechanisms, the NRC staff finds the licensee's assessment acceptable.

The licensee performed a SG tube wear analysis in consideration of SG thermal-hydraulic changes resulting from the MUR PU and determined that the fluid-elastic stability will increase by 13 percent. This includes an increase in vibration amplitude due to turbulence and an increase in tube wear of as much as 29 percent. The maximum stability ratio was determined to be 0.57 and the turbulence induced amplitude < 0.006 inches, which are less than their allowable limits. In addition, the licensee calculated the maximum post-uprate wear over 40 years to be < 0.002 inches, which is a 50-percent increase from the pre-uprate value. The NRC staff reviewed the licensee's SG tube wear analysis and agrees that tube wear resulting from MUR will not significantly affect tube integrity.

The licensee also performed reviews on tube stress and fatigue. The tube stress from flow-induced vibration after the MUR PU was reported to be approximately 0.2 kilo pounds per square inch (ksi), which is below the ASME Code stress limits and the fatigue endurance limit.

Therefore, the NRC staff finds this acceptable.

The licensee performed an analysis on the mechanical repair hardware (Le., plugs, sleeves, and stabilizers) that are installed in the SGs to address tube degradation and reported that the fatigue usage values adjusted for MUR PU conditions remain less than the 1.0 ksi fatigue limit. The licensee stated that the SG mechanical repair hardware continues to meet ASME Code limits for plant operation at MUR PU conditions. The fatigue usage values and the SG mechanical repair hardware continues to meet ASME Code limits, therefore, NRC staff finds the licensee's evaluation acceptable.

The licensee stated foreign object search and retrieval operations during the Surry 1 and 2 refueling outages determined that four irretrievable objects are present in the Unit 1 SGs and five in the Unit 2 SGs with no indication of wear present. The licensee performed an NRC RG 1.121 "Bases for Plugging Degraded PWR Steam Generator Tubes" (ADAMS Accession No. ML003739366), evaluation on the SG tubing structural limits. The allowable stress limits were established using the ASME Code,Section III, 1986, Code Case N-20-3 minimum strength properties. A wear time analysis was performed assuming a 20-percent initial tube wear on the limiting tube location. Although SG secondary-side conditions will change due to the MUR PU, the licensee stated that the changes do not affect the evaluation conclusions. The licensee's analysis determined that the amount of time required for the limiting foreign object orientation to wear a tube down to a minimum allowable tube wall thickness under conservative secondary-side conditions was greater than 3 years or two operational cycles. The TSs require that SG tube integrity be maintained between SG operational cycles. The licensee determined that operation at MUR PU conditions is acceptable with respect to the current foreign object inventory in the SGs. The licensee stated that the analysis showed that the existing tube repair limit is unaffected by the MUR PU and remains valid at uprated conditions. The NRC staff finds this acceptable.

The NRC staff has reviewed the licensee's evaluation and calculation results found in the amendment request and have confirmed that the applicable regulatory guidance was followed.

The NRC staff agrees that the proposed MUR PU will introduce only insignificant changes as it

3.4.4

- 17 relates to tube stresses resulting from tube vibration, cumulative fatigue usage factors, and potential tube wear. The MUR PU will not affect the satisfactory performance in maintaining SG tube integrity.

3.4.3.3 Conclusion The NRC staff reviewed the licensee's evaluation of the effect of the proposed MUR PU on SG tube integrity and concludes that the licensee has adequately assessed the continued acceptability of the plant's TSs. Specifically, the licensee has an ongoing periodic inspection program which will continue to be utilized to detect degradation that may occur. Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to the SG program.

Chemical and Volume Control System 3.4.4.1 Regulatory Evaluation The chemical and volume control system (CVCS) provides a means for: (1) maintaining water inventory and quality in the RCS, (2) supplying seal-water flow to the reactor coolant pumps and pressurizer auxiliary spray, (3) controlling the boron neutron absorber concentration in the reactor coolant, (4) controlling the primary-water chemistry and reducing coolant radioactivity level, and (5) supplying recycled coolant for demineralized water makeup for normal operation and high-pressure injection flow to the ECCS in the event of postulated accidents. The NRC staff has reviewed the safety-related functional performance characteristics of CVCS components. The NRC's acceptance criteria are based on (1) General Design Criterion (GDC)-14, "Reactor Coolant Pressure Boundary (RCPB)," as it requires that the RCPB be designed to have an extremely low probability of abnormal leakage, or rapidly propagating fracture, and of gross rupture, and (2)

GDC-29, "Protection Against Anticipated Operational Occurrences," as it requires that the reactivity control systems be designed to assure an extremely high probability of accomplishing their functions in the event of condenser in-leakage or primary-to-secondary leakage. Specific review criteria are contained in SRP Section 9.3.4, "Chemical and Volume Control System (PWR)."

3.4.4.2 Technical Evaluation The licensee stated that UFSAR Section 14.2.5 for Surry 1 and 2 contain information on boron dilution during refueling, cold shutdown, intermediate shutdown, hot shutdown, reactor critical, and power operation. The analysis on boron dilution events is conducted to verify that limits of the departure from nucleate boiling ratio (DNBR), RCS and main steam system over-pressurization are met. The licensee stated that the administrative shutdown margin requirements are verified for each reload core and will continue to be verified for the MUR PU cores.

In addition, the licensee stated that consequences that may come about at reactor critical conditions are bound by the Rod Withdrawal from Subcritical (RWSC) event in UFSAR Section 14.2.1. The consequences are bounding because the maximum achievable reactivity insertion rate experienced during a boron dilution event is less than the reactivity insertion rate assumed in the RWSC analysis.

The licensee stated that the "at power" boron dilution transient is identical to that of a control rod withdrawal at power (RWAP), which is analyzed in UFSAR Section 14.2.2. Since the boron

3.4.5

- 18 dilution reactivity insertion rate is within the range analyzed for the RWAP event, the DNBR, RCS pressure, and main steam safety (MSS) pressure responses would be bounded by the RWAP event. The licensee has concluded that fuel cladding integrity is maintained during postulated boron dilution events in all operating modes, such that RCS and MSS pressures remain below 110 percent of design pressure during postulated boron dilution events. Therefore, the licensee has not performed any explicit analysis of the boron dilution event and has determined that none is required. In addition, the licensee stated that the 15-minute operator response time for postulated boron dilution events in all operating modes will not be affected by the MUR PU. The NRC staff has reviewed the licensee's assessment of the potential impact of a boron dilution event under MUR conditions and finds it acceptable.

The NRC staff has reviewed the licensee's evaluation and has confirmed that the applicable regulatory guidance was followed. The licensee has demonstrated that the CVCS will continue to maintain RCS inventory and water chemistry. The NRC staff finds that the CVCS will continue to meet system design requirements and that no new design transients will be created at the MUR PU conditions.

3.4.4.3 Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed MUR PU on the CVCS and concludes that the licensee has adequately addressed changes in the temperature of the reactor coolant and its effects on the CVCS. The staff further concludes that the licensee has demonstrated that the CVCS will continue to be acceptable and will continue to meet the requirements of GDC-14 and GDC-29 following implementation of the proposed MUR PU.

Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to the CVCS.

Steam Generator Blowdown System 3.4.5.1 Regulatory Evaluation Control of secondary-side water chemistry is important for preventing degradation of SG tubes.

The SG blowdown system (SGBS) prOVides a means for removing SG secondary-side impurities and, thus, assists in maintaining acceptable secondary-side water chemistry in the SGs. The design basis of the SGBS includes consideration of expected and design flows for all modes of operation. The NRC staff's review covered the ability of the SGBS to remove particulate and dissolved impurities from the SG secondary side during normal operation, including condenser in-leakage and primary-to-secondary leakage. The NRC's acceptance criteria for the SGBS are based on GDC-14, "Reactor Coolant Pressure Boundary," as it requires that the RCPB be designed so as to have an extremely low probability of abnormal leakage, of rapidly propagating fracture, and of gross rupture. Specific review criteria are contained in SRP 10.4.8, "Steam Generator Blowdown System (PWR)" (ADAMS Accession No. ML052350093).

3.4.5.2 Technical Evaluation The licensee stated that the SGBS flow rates during plant operation are based on chemistry control and tubesheet sweep necessary to control solids buildup. The SGBS was analyzed for a blowdown flowrate increase of approximately 4 gallons per minute (gpm).

The licensee stated that the SGBS flowrate will continue to be operated with no change under

- 19 MUR PU conditions. It was further stated that SGBS operating temperatures and pressures will decrease, but remain bounded by the existing design parameters under uprate conditions. The licensee stated that the uprate will not significantly increase the potential for FAC on the blowdown system piping and components. The SGBS system will continue to be monitored by an acceptable FAC monitoring program. The NRC staff finds this acceptable.

The NRC staff has reviewed the licensee's evaluation and has confirmed that the applicable regulatory guidance was followed. The licensee has demonstrated that the SGBS is adequate for maintaining secondary-side water chemistry within industry guidelines for maintenance of controlled corrosion rates in secondary system components. The NRC staff finds this evaluation acceptable and the SGBS will continue to meet system design requirements at MUR PU conditions.

3.4.5.3 Conclusion The NRC staff has reviewed the licensee's evaluation of the effects of the proposed MUR PU on the SGBS and concludes that the licensee has adequately addressed changes in system flow and impurity levels and their effects on the SGBS. The NRC staff further concludes that the licensee has demonstrated that the SGBS will continue to be acceptable and will continue to meet the requirements of GDC 14 following implementation of the proposed MUR PU.

3.4.6 Overall Chemical Engineering Conclusion In the areas of SGs and chemical engineering, the NRC staff concludes that the licensee has adequately addressed (1) the changes in the plant operating conditions for the FAC program, (2) the effects on protective coatings, (3) the changes in the SG operating parameters, the effects on the SGs and the determination that the SG tube integrity will continue to be maintained, (4) the changes of the reactor coolant and their effect on the CVCS, and (5) the changes in the system flow and impurity levels, and their effects on the SGBS.

3.5 Mechanical and Civil Engineering 3.5.1 Regulatory Evaluation Nuclear power plants are licensed to operate at a specified core thermal power, referred to as the CLTP. Appendix K to 10 CFR Part 50 requires licensees to assume that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level when performing ECCS analyses for LOCAs. This requirement is included to ensure that instrumentation uncertainties are adequately accounted for in these analyses. Appendix K to 10 CFR Part 50 allows licensees to assume a bounding power level less than 1.02 times the licensed power level (but not less than the licensed power level), provided the licensee has demonstrated that the proposed value adequately accounts for instrumentation uncertainties.

The licensee has proposed to use a power measurement uncertainty of 0.38 percent based on the installation of the Cameron system which provides a more accurate measurement of FW flow than the FW flow measurement accuracy assumed during the development of the original 10 CFR Part 50, Appendix K, requirements and that of the current method of FW flow measurement used to calculate reactor thermal power output.

The NRC staff's review in the areas of civil and mechanical engineering covers the structural and

- 20 pressure boundary integrity of the NSSS and balance-of-plant (BOP) systems, and SSCs.

Specifically, this review focuses on the impact of the proposed MUR PU on the structural integrity of (1) reactor pressure vessel; (2) reactor vessel internals; (3) control rod drive mechanisms (CRDMs); (4) NSSS piping, componenets, and supports; (5) SG components and the replacement steam generator (RSG) components and their supports; (6) the pressure retaining portions of the reactor coolant pumps (RCPs) and their supports; (7) the pressurizer and its supports, and (8) BOP piping systems. Technical areas covered by this review include stresses, fatigue and corresponding cumulative usage factors (CUFs), flow-induced vibration (FIV),

high-energy line break (HELB) locations and corresponding jet impingement and thrust forces, safety-related valves and pumps the Inservice Testing (1ST) Program.

Surry 1 and 2 received construction permits prior to May 21, 1971, which is the date the GDC in Appendix A of 10 CFR Part 50 became effective. Although the plant is exempt from the current GDC, the licensee states it is in compliance with the intent of the current GDC and also meets the design criteria that were in effect when Surry 1 and 2 was licensed. VEPCO discusses how it meets the design criteria for Surry 1 and 2 in Section 1.4 of the UFSAR. As such, the NRC staff's evaluation in the areas of civil and mechanical engineering considered 10 CFR 50.55a and GDC 1, 2, 4, 10, 14, and 15. The NRC staff's review focused on verifying that the licensee has provided reasonable assurance of the structural and functional integrity of the aforementioned piping systems, components, component internals and their supports under normal and vibratory loadings, including those due to fluid flow, postulated accidents, and natural phenomena such as earthquakes.

The acceptance criteria are based on continued conformance with the requirements of the following regulations: (1) 10 CFR 50.55a, and GDC 1 as they relate to SSCs being designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed; (2) GDC 2 as it relates to structures and components important to safety being designed to withstand the effects of earthquakes combined with the effects of normal or accident conditions; (3) GDC 4 as it relates to structures and components important to safety being designed to accommodate the effects of, and to be compatible with, the environmental conditions of normal and accident conditions and these structures and components being appropriately protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids; (4) GDC 10 as it relates to the reactor core being designed with appropriate margin to assure that specified acceptable fuel design limits (SAFDLs) are not exceeded during any condition of normal operation, inclUding the effects of anticipated operational occurrences; (5) GDC 14 as it relates to the reactor coolant pressure boundary being designed, fabricated, erected, and tested to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture; and (6) GDC 15 as it relates to the RCS being designed with sufficient margin to ensure that the design conditions are not exceeded.

3.5.2 Technical Evaluation The NRC staff's technical review in the areas of civil and mechanical engineering focused on the effects of the MUR PU on the structural and pressure boundary integrity of piping systems and components, their supports, the reactor vessel and internal components, the pressure retaining portions of the CRDMs and the BOP and I'JSSS interface piping systems. The proposed 1.6 percent MUR PU will increase the RP level from 2546 MWt to 2587 MWt at Surry 1 and 2. The MUR PU will be achieved by an increase in demand to the turbine-generator. In turn, an increase

- 21 in steam flow will occur due to the increased demand on the secondary side of the plant. In addition, there will be an increased temperature difference across the core with the RCS pressure remaining the same.

Table 4.0-2 of Attachment 1 in the LAR shows the pertinent temperatures, pressures, and flow rates for the current and uprated conditions. VEPCO evaluated four bounding cases to evaluate the effects of the proposed MUR PU on various SSCs. For the purposes of the structural and mechanical evaluations performed in support of the proposed MUR PU fragment. VEPCO utilized the most appropriate case to perform bounding analyses for the aforementioned SSCs. In all cases, there is no change in the RCS operating pressure (2250 [pounds per square inch atmosphere] psia). At full power for Case 4 (highest average vessel temperature, 7 percent SG tubes plugged), the hot-leg temperature increases from 605.6 to 609.1 degrees Fahrenheit (OF) while the cold-leg temperature increases from 540.4 of to 542.9 of. The SG pressures decrease from 784.0 psia to 781.0 psia and the steam flow increases from 11.26 million pounds per hour (Mlbm/hr) to 11.63 Mlbm/hr. The FW temperature increases from 443.0 of to 452.0 of. The design parameters for the primary system at Surry 1 and 2 are found in Chapter 4 of the Surry 1 and 2 UFSAR. The RCS components are designed to 650 of (except the pressurizer and pressurizer surge line, which are designed to 680 OF) and 2,485 pounds per square inch gage (psig). Chapter 10 of the Surry 1 and 2 UFSAR provides the design-basis information for the secondary side systems, including the main steam (MS) system and the FW and condensate system.

3.5.2.1 Reactor Pressure Vessel (RPV)

The licensee evaluated the effects of the proposed MUR PU on the structural integrity of the RPV, including its nozzles, in Section IV.1.A.i of Attachment 5 in the LAR. Table IV-3 of Attachment 5 provides the design codes of record for the SSCs evaluated in support of the MUR PU at Surry 1 and 2. The design code of record for the RPV is the ASME Code,Section III, 1968 Edition through the winter 1968 Addenda. The code of record for the replacement reactor vessel closure heads (RVCHs) at Surry 1 and 2 is the ASME Code,Section III, 1995 Edition with 1996 Addenda.

The licensee compared the expected temperatures and pressures for the proposed MUR PU condition against the analyses of record. The licensee indicated that neither the minimum vessel inlet temperature nor maximum vessel outlet temperature were bounded by the current analysis of record, which was performed in support of the stretch power uprate (SPU) implemented at Surry 1 and 2 in 1995. For the purposes of evaluating the effects of the increased maximum vessel outlet temperature under operating conditions, the licensee evaluated the reactor vessel main closure flange assemblies and outlet nozzles to ensure that these components would continue to meet the design requirements of their corresponding code of record. The licensee also evaluated the remaining components to determine whether the lower minimum vessel inlet temperature would have any effect on these components' abilities to meet their design requirements, per the code of record. Based on its evaluations, the licensee confirmed that the RPV and its associated components described above will continue to meet the stress and fatigue design requirements of the design codes of record for these components following the implementation of the MUR PU. The licensee also confirmed in its submittal that there is no change in RCS design or operating pressure, and the effects of operating temperature changes for the cold-and hot-legs are within the design limits. To date. no additional transients have been proposed as a result of the MUR PU at Surry 1 and 2.

- 22 The NRC staff has reviewed the licensee's evaluations related to the structural integrity of the RPV and its associated components, including nozzles. For the reasons set forth above, which demonstrate that the RPV will continue to meet its design basis acceptance criteria under the conditions of the proposed MUR power level, the NRC staff concludes that the licensee has adequately addressed the effects of the proposed MUR PU on these components. The NRC staff further concludes that the licensee has demonstrated that the RPV and its associated components will continue to meet the applicable regulatory requirements, described above, following implementation of the proposed MUR PU. Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to the structural integrity of the RPV.

3.5.2.2 Reactor Vessel Internals (RVls)

The licensee evaluated the effects of the proposed MUR PU on the structural integrity of the RVls in Section IV.1.A.ii of Attachment 5 in the LAR. Section 15A.2.3 of the Surry 1 and 2 UFSAR details the ASME Code,Section III criteria used in the structural design and analysis of the Surry 1 and 2 RVls. Mechanical and structural evaluations were performed by the licensee to determine any effects on the RVls due to the conditions which would be present following the implementation of the proposed MUR PU. Based on the analysis, the MUR PU will not affect the design basis seismic and LOCA loads for Surry 1 and 2, negating the need to re-assess the structural integrity of the RVls with regards to the LOCA-induced hydraulic and dynamic loads and seismic loads. Due to the increased hot-leg temperature and corresponding reduction in RCS density, the licensee evaluated the possible effects of FIV on the reactor vessel internals to ensure that fatigue endurance limits for the RVI components under review would not be approached. It was confirmed by the licensee that the RVls, including the internals most vulnerable to FIV (the core barrel assembly and upper internals), contain adequate structural margin such that FIV will have a negligible impact on the structural integrity of the RVls due to the MUR PU.

Structural evaluations performed by the licensee demonstrated that all but two components making up the RVls would be bounded by the current stress and fatigue analyses of record.

However, the upper and lower core plates, are directly affected by the increased heat generation rates at the conditions which exist at the MUR PU level. The licensee performed additional confirmatory structural analyses on the upper and lower plates to demonstrate that they would maintain their structural integrity at the proposed MUR PU conditions. By comparing the maximum primary plus secondary stress intensities of these two components to their respective code allowable values, the licensee demonstrated that the lower and upper core plates are structurally adequate for operation at the MUR PU conditions and will continue to meet their design basis criteria. Additionally, the licensee stated that the CUFs for these two components remain acceptable under MUR PU conditions.

The NRC staff has reviewed the licensee's evaluations related to the structural integrity of the RVls and concludes that the licensee has adequately addressed the effects of the proposed MUR on these components based on the licensee's demonstration that the RVls will continue to meet their design basis acceptance criteria under the conditions of the proposed MUR power level.

The NRC staff further concludes that the licensee has demonstrated that the reactor internals and core supports will continue to meet the applicable regulatory requirements following implementation of the proposed MUR PU. Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to the design of the RVls.

- 23 3.5.2.3 Control Rod Drive Mechanisms The licensee evaluated the effects of the proposed MUR PU on the structural integrity of the CRDMs in Section IV.1.A.iii of Attachment 5 of the LAR. Table IV-3 of Attachment 5 in the LAR provides the design codes of record for the SSCs evaluated in support of the MUR PU at Surry 1 and 2. The design codes of record for the pressure retaining components of the CRDMs is the ASME Code, Section 111,1965 Edition through the Summer 1966 Addenda for Unit 1, and the ASME Code, Section 111,1995 Edition through the 1996 Addenda for Unit 2. Based on the location of the CRDMs, these components are affected primarily by the RCS pressure and the RCS hot leg temperature, of which only the hot-leg temperature changes as a result of the MUR PU. The licensee indicated that the current analyses of record for the pressure retaining portions of the Unit 1 and Unit 2 CRDMs were reviewed to evaluate the effects of design parameter changes on these components. The current analysis of record for Unit 1 utilizes a bounding hot leg temperature of 609.1 of while the current analysis of record for Unit 2 utilizes a bounding hot leg temperature of 610°F. Both of these values are either equal to or bounded by the maximum expected temperatures following the implementation of the proposed MUR PU at Surry 1 and 2.

Therefore, the licensee confirmed that the stresses in the pressure retaining portions of the CRDMs remain acceptable for both units at the conditions following the MUR PU implementation.

The licensee also confirmed that the fatigue analyses for both units remain acceptable for the MUR PU conditions. In addition, no additional transients have been proposed.

The NRC staff has reviewed the licensee's evaluations related to the structural integrity of the pressure retaining portions of the CRDMs. For the reasons set forth above, which demonstrate that the CRDMs will continue to meet their design-basis acceptance criteria under the conditions of the proposed MUR power level, the NRC staff concludes that the licensee has adequately addressed the effects of the proposed MUR on these components. Based on the above, the NRC staff further concludes that the licensee has demonstrated that the CRDMs will continue to meet the applicable regulatory requirements following implementation of the proposed MUR.

Therefore, the NRC staff finds the proposed MUR acceptable with respect to the structural integrity of the pressure retaining portions of the CRDMs.

3.5.2.4 Reactor Coolant System Piping, Components and Supports In support of the proposed MUR PU at Surry 1 and 2, the licensee evaluated the various piping systems, components and supports which make up the RCS. This evaluation included the RCS main loop piping and corresponding branch piping, the original and replacement, primary and secondary side SG components, the pressure retaining portions of the reactor coolant pumps, and the pressurizer and corresponding pressurizer surge line. The NRC staff's review of these SSCs is detailed in the following subsections.

3.5.2.4.1 Reactor Coolant System Piping and Supports The licensee evaluated the effects of the proposed MUR PU on the structural integrity of the RCS piping and supports in Section IV.1.A.iv of Attachment 5 of the LAR. Table IV-3 of Attachment 5 provides the design codes of record for the SSCs evaluated in support of the MUR PU at Surry 1 and 2. The RCS piping was designed in accordance with the USA Standard (USAS) B31.1 Code for Power Piping, 1955 Edition, with the exception of the pressurizer surge line, which is discussed in SE Section 3.5.2.4.4. The licensee reviewed the revised design conditions for impact on the existing design basis analyses for the reactor coolant piping and supports. The

- 24 NRC staff has noted that the design code of record does not require fatigue evaluations to be performed for the Surry 1 and 2 RCS piping systems.

For the purposes of evaluating the effects of the MUR PU on the RCS piping and supports, there is no change in RCS design or operating pressure, and the effects of the variance in operating temperature for the hot-legs (increases at uprated conditions) and cold-legs (decreases at uprated conditions) are within design limits. The licensee also indicated that the operating transient effects due to the MUR implementation continue to be bounded by the analyses of record and no additional transients have been proposed. In response to an NRC staff request for additional information regarding the evaluations performed for the RCS loop piping and components, the licensee confirmed that the current analysis of record for the pressure retaining portions of the RCS, including the RCS loop piping, primary equipment nozzles, primary equipment supports, auxiliary piping lines attached to the RCS loops, and the auxiliary branch nozzles attached to the RCS loops remain bounding under the proposed MUR PU conditions.

The MUR PU conditions are bounded by the design conditions and the RCS piping remains within the allowable stress limits provided by the applicable codes of record. Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to the structure integrity reactor coolant system piping and supports.

3.5.2.4.2 Steam Generator Components The licensee evaluated the effects of the proposed MUR PU on the structural integrity of critical primary and secondary side components of the SGs in Section IV. 1.A.vL2, Attachment 5 of the LAR. Table IV-3 of Attachment 5 of the LAR provides the design codes of record for the SSCs evaluated in support of the MUR PU at Surry 1 and 2. The RSGs at Surry were installed in 1981 and 1980 at Units 1 and 2, respectively. The new SGs replaced the original tube bundles, lower shells, primary channel head regions and primary moisture separators and feedrings while the original upper shells and secondary moisture separators (steam drums) remained. The design code of record for the shell side of the RSGs at Surry 1 and 2 is the ASME Code,Section III, 1974 Edition through the Winter 1976 Addenda. The original SG components, which were not replaced, were reanalyzed in accordance with the same design code of record as that indicated for the RSGs.

The revised design conditions were reviewed by the licensee to determine the impact on the existing design basis analyses for the critical original and RSG primary and secondary side components affected by the proposed MUR PU. By utilizing scale factors, the licensee determined the increases in the primary and secondary stresses on the critical SG components, which resulted from the MUR PU conditions. The licensee also evaluated the change in the primary-to-secondary differential pressure due to the proposed MUR PU. The NRC staff finds that all of the primary and secondary SG components will continue to meet the design basis code stress and differential pressure requirements (Le. are bounded by the design code of record) following the implementation of the proposed MUR PU. Additionally, the NRC staff finds that these components will continue to maintain acceptable CUFs following the MUR PU implementation.

3.5.2.4.3 Reactor Coolant Pump Components The licensee evaluated the effects of the proposed MUR PU on the structural integrity of the pressure retaining portions of the RCPs, including the nozzles and supports, in Sections IV.1.A.iv

- 25 and IV.1.A.vii, Attachment 5 of the LAR. Table IV-3 of Attachment 5 of the LAR provides the design codes of record for the SSCs evaluated in support of the MUR PU at Surry 1 and 2. The pressure retaining parts of the RCPs at Surry 1 and 2 were designed in accordance with the Article 4 of the ASME Code,Section III. The licensee reviewed the revised design conditions to determine the impact on the existing design basis analyses for the RCPs, their nozzles and corresponding supports. Given that the RCP structural evaluations are highly dependent on the vessel inlet temperature, the licensee noted that the maximum vessel inlet temperature analyzed in support of the proposed MUR PU (542.9 OF) is essentially equal to the value used for these components in the current analyses of record (543 OF). The licensee noted that all of the pressure retaining components of the RCPs are bounded by the current analyses of record with the exception of the weir plate. The licensee indicated that additional verification was performed to demonstrate the acceptability of the weir plate at the conditions following the implementation of the MUR PU. Additionally, it was noted that the original analyses for the weir plate utilized a fatigue waiver (per NB-3222.4 of Section III of the ASME Code). The licensee indicated that this waiver remains valid for the purposes of the weir plate evaluation in support of the proposed MUR PU. In response to an NRC RAJ regarding the evaluations performed for the RCS loop piping and components, the licensee confirmed that the current analyses of record for the pressure retaining portions of the RCS and corresponding primary component supports, including the RCPs, remain bounding under the proposed MUR PU conditions. Therefore, the NRC staff finds the structural integrity of the pressure retaining portions of the RCP components acceptable.

3.5.2.4.4 Pressurizer The licensee evaluated the effects of the proposed MUR PU on the structural integrity of the pressurizer and the pressurizer surge line in Sections IV.1.A.viii and IV.1.A.iv, respectively, of of the LAR. Table IV-3 of Attachment 5 provides the design codes of record for the SSCs evaluated in support of the MUR PU at Surry 1 and 2. The design code of record for the pressurizer at Surry 1 and 2 is the ASME Code,Section III, 1965 Edition through the Winter 1965 Addenda. The current design code of record for the pressurizer surge line is provided as the 1986 Edition, with 1987 Addenda, of Section III of the ASME Code. The surge line was re-evaluated to address NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification" (ADAMS Accession No. ML031220290). With regards to the structural integrity of the pressurizer, the governing loading conditions occur when the RCS pressure is elevated and the hot and cold leg temperatures are low. The licensee reviewed the revised design conditions to determine the impact on the existing structural design basis analyses for the pressurizer. Given that the RCS pressure remains unchanged at the uprated conditions and the hot-and cold-leg temperature variations at these conditions remain bounded, the licensee concluded that the current design basis analyses remain bounding for the Surry 1 and 2 pressurizers.

The licensee confirmed that the NSSS design transients and insurge/outsurge transient effects are not affected by the MUR PU and no additional transients have been proposed. The existing loads, stresses, and fatigue CUF values for the pressurizer remain valid for the proposed MUR PU. In response to an RAI regarding the evaluations performed for the RCS loop piping and components, the licensee confirmed in their April 29, 2010 (ADAMS Accession No. ML101200269), RAI response that the current analyses of record for the primary equipment supports and piping systems attached to the RCS loops remain bounded under the proposed MUR PU conditions. Therefore, for the pressurizer surge line and the pressurizer component supports, the MUR PU conditions are bounded by the current analyses of record. The allowable stress limits for these components, provided by the applicable codes of record identified above,

- 26 will continue to be satisfied following the implementation of the proposed MUR PU. Therefore, the NRC staff finds this acceptable.

3.5.2.4.5 Conclusion The NRC staff has reviewed the licensee's evaluations related to the structural integrity of the RCS piping, components, and supports. The NRC staff concludes that the licensee has adequately addressed the effects of the proposed MUR PU on the current analyses of record for these SSCs based on the licensee's demonstration that the RCS piping, components, and supports will continue to meet their design basis acceptance criteria under the conditions of the proposed MUR power level. The NRC staff further concludes that the licensee has demonstrated that SSCs important to safety will continue to meet the applicable regulatory requirements following implementation of the proposed MUR PU. Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to the st~ctural integrity of the RCS piping, components, and supports.

3.5.2.5 BOP Piping Systems The licensee evaluated the effects of the proposed MUR PU on the structural integrity of the BOP piping, including NSSS interface systems, safety-related cooling water systems and containment systems, in Section IV.1.A.v in the LAR. A complete listing of the systems evaluated in support of the proposed MUR PU at Surry 1 and 2 is provided in Section IV.I.A.v of Attachment 5 of the LAR.

Table IV-3 of Attachment 5 of the LAR provides the design codes of record for the SSCs evaluated in support of the MUR PU at Surry 1 and 2. The BOP piping systems at Surry 1 and 2 were designed in accordance with the 1967 Edition of the ANSI B31.1 Code for Power Piping.

The licensee evaluated the BOP piping systems by comparing the conditions for the proposed MUR PU with the conditions in the analyses of record and the current operating conditions. The licensee stated that change factors (Le., ratios of pre-uprate conditions versus post-uprate conditions) were utilized to analyze the effect that the changes in pressure, temperature and flow would have on the aforementioned piping systems after the proposed MUR PU. If a change factor was greater than one (1) for any of the three parameters for a specific system, an evaluation was performed to determine whether the design basis analyses for the system under review would remain bounding.

In response to an NRC RAI regarding its evaluation utilizing these change factors, the licensee provided additional information regarding systems which were found to have a change factor greater than 1 during the analyses performed to support the proposed MUR PU implementation.

The licensee indicated that six systems were found to have change factors greater than 1 under the proposed MUR PU conditions: the MS and steam dump system, the extraction steam system, the condensate system, the FW system, the heater drain system, and the FW system. Based on these change factors, the licensee stated that the pipe stresses in the current analyses of record were proportionally increased; these increased stresses were found to be within allowable stress values provided in the code of record indicated above. Due to increased steam flow rates, an additional analysis was performed for a portion of the MS system piping between the turbine stop valves (TSVs) and the steam generators in order to evaluate the effects of a sudden TSV closure transient. The licensee stated that the pipe stresses in this portion of the MS system due to this transient loading condition were bounded by the code allowable stress values.

- 27 The NRC staff has reviewed the licensee's evaluations related to the Surry 1 and 2 BOP piping systems and concludes that the licensee has adequately addressed the effects of the proposed MUR on the current BOP piping analyses of record based on the licensee's demonstration that the BOP piping systems will continue to meet their design basis acceptance criteria under the conditions of the proposed MUR power level. The NRC staff further concludes that the licensee has demonstrated that SSCs important to safety will continue to meet the applicable regulatory requirements following implementation of the proposed MUR PU. Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to the structural integrity of the aforementioned BOP piping systems.

3.5.2.6 High Energy Line Break (HELB) Locations The licensee stated that the current HELB analysis for Surry 1 and 2 was reviewed in support of the proposed MUR PU.Section IV.1.B.vii of Attachment 5 of the LAR states that the changes in the NSSS system parameters (e.g., temperature, pressure, and mass flow rates) were reviewed to determine whether the changes in these parameters would have a significant impact on the pipe stresses of the piping systems affected by the proposed MUR PU. It was concluded that the pipe stresses were not significantly affected by the parameter changes resulting from the proposed MUR PU. Therefore, the licensee concluded that the current Surry 1 and 2 HELB analysis remains unaffected by the MUR PU and the existing pipe break, jet impingement and pipe whip design basis evaluations remain valid. Additionally, the licensee indicated that the existing leak before break (LBB) analyses used to justify the elimination of large break pipe ruptures of the RCS loop piping at Surry 1 and 2 remain valid for the conditions under the proposed MUR PU.

The NRC staff has reviewed the licensee's evaluations related to determinations of rupture locations and associated dynamic effects and concludes that the licensee has adequately addressed the effects of the proposed MUR PU on the current HELB and LBB analyses based on the insignificant pipe stress increases due to the MUR PU. The NRC staff further concludes that the licensee has demonstrated that SSCs important to safety will continue to meet the regulatory requirements applicable to HELBs following implementation of the proposed MUR PU. Therefore, the NRC staff finds the proposed MUR PU acceptable with respect to the determination of rupture locations and dynamic effects associated with the postulated rupture of piping and the existing design basis remains valid.

3.5.2.7 Safety-Related Valves In the LAR, the licensee reviewed the impact of the proposed MUR PU conditions on the existing design basis analyses for the safety-related valves. In Sections IV.1.B.iv, IV.1.B.v, and IV.1.B.vi of Attachment 5, "NRC Regulatory Issue Summary 2002-03 Requested Information," the licensee reviewed the revised design and operating conditions resulting from the MUR PU against the eXisting safety-related valve design basis analyses, and concluded that no changes in reactor coolant flow and operating pressure were made as part of MUR PU, and that the temperature changes due to MUR PU were bounded by those used in the existing analyses. Therefore, the licensee states in Section IV.1.A.ix that none of the safety-related valves require a change to their design or operating conditions as a result of the MUR PU. In Section VI.1.A.i.b, the licensee reviewed the MUR PU impact on main steam safety valves (MSSVs), and concluded that the existing MSSVs and associated setpoint pressures were adequate for the MUR PU conditions. In Section V11.6.E, the licensee reviewed the MUR PU impact on the requirements of GL 89-10

- 28 (ADAMS Accession No. ML031150300), GL-95-07 (ADAMS Accession No. ML003689383), and GL 96-05 (ADAMS Accession No. ML993080099). The evaluation shows that no required changes are identified. In Section V11.6.F, the licensee also reviewed the MUR PU impact on the Air-Operated Valves (AOVs) Program. The review indicates that the maximum pressures, flow rates and fluid temperature for all AOVs, remain unchanged or are bounded by the current parameters. The review concluded that the MUR PU does not impact the designs and operations of the safety-related valves since the operating ranges of pressure, temperature, and flow, are bounded by those used in the existing analyses. Therefore, the NRC staff finds the performance of existing safety-related valves and the current 1ST program acceptable with respect to the MUR.

3.5.2.8 Safety-Related Pumps The NRC staff reviewed the licensee's safety-related pumps analysis. The NRC's acceptance criteria for reviewing the safety-related pumps analysis are based on 10 CFR 50.55a. The licensee reviewed the impact of the proposed MUR PU conditions on the existing design-basis analyses for the safety-related pumps. The evaluation showed that there are no significant changes to the maximum operating conditions and no changes to the design-basis requirements that would affect pump performance or the acceptable test criteria for the pumps. Therefore, the NRC staff finds the performance of existing safety-related pumps acceptable with respect to the MUR PU.

3.5.2.9 Inservice Testing (1ST) Program In its submittal, the licensee states that 10 CFR 50.55a(f), "Inservice Testing Requirements,"

mandates the development and implementation of an 1ST Program. Surry 1 and 2 has developed and is implementing an 1ST Program for pumps and valves per the applicable requirements.

Surry 1 and 2 TS 6.4.1 describes the surveillance requirements (SRs) that apply to the inservice testing of ASME Code Class 1, 2, and 3, pumps and valves.

In Section 1V.1.E.i of Attachment 5 of the LAR, for the 1ST Program, the licensee states that there are no significant changes to the maximum operating conditions and no changes to the design basis requirements that would affect component performance or test acceptance criteria. The 1ST Program for safety-related valves will not be affected by the MUR PU. The NRC staff finds the 1ST Programs analysis acceptable.

3.5.3 Conclusion The NRC staff has reviewed VEPCO's assessment of the impact of the proposed MUR PU on the NSSS and BOP SSCs with regard to stresses, CUFs, flow-induced vibration, HELB locations, and corresponding jet impingement and thrust forces and safety-related valves and pumps and the 1ST Program. Based on the review described above, the NRC staff finds the MUR PU acceptable with respect to the structural integrity of the aforementioned SSCs affected by the MUR PU. This acceptance is based on the licensee's demonstration that the SSCs affected by the proposed MUR PU will maintain their structural integrity following the implementation of the MUR PU because the condition of the MUR PU are bounded by the design of record and the ASME Code.

Additionally, the licensee has also demonstrated that the intent of the aforementioned regulatory requirements related to the civil and mechanical engineering purview has been met. Therefore, there is reasonable assurance that these SSCs will be able to maintain their structural integrity in order to perform their intended functions following the implementation of the MUR PU at Surry 1

- 29 and 2.

3.6 Reactor Systems 3.6.1 Regulatory Evaluation Early revisions of 10 CFR 50.46, and Appendix K to 10 CFR Part 50, required licensees to base their LOCA analysis on an assumed power level of at least 102 percent of the licensed thermal power level to account for power measurement uncertainty. The NRC later modified this requirement to permit licensees to justify a smaller margin for power measurement uncertainty.

Licensees may apply the reduced margin to operate the plant at a level higher than the previously licensed power. The licensee proposed to use a Cameron LEFM CheckPlus System to decrease the uncertainty in the measurement of FW flow, thereby decreasing the power level measurement uncertainty from 2.0 percent to 0.35 percent.

The licensee developed its LAR consistent with the guidelines in NRC RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications."

3.6.2 Technical Evaluation The NRC staff reviewed thermal-hydraulic aspects of the Cameron LEFM CheckPlus System installation, including its laboratory calibration, the effects of system changes such as transducer replacement, and the impact the system installation will have, if any, on the applicable plant safety analyses, as discussed in the following subsections.

3.6.2.1 Feedwater Flow Measurement Device Installation The Cameron LEFM CheckPlus Systems at Surry 1 and 2 consist of measurement spool pieces installed in each of the three main FW flow lines between the existing FW venturi flow meters and the main FW check valves. The spool pieces are installed downstream of existing FW venturis in 14-inch Schedule 80 FW piping.

The devices are installed in accordance with the requirements in the approved Cameron Topical Reports ER-80P and ER-157P related to the LEFM Check and LEFM CheckPlus Systems. After plant installation, the licensee compared the "as-installed" measurement uncertainties with that of the measurement uncertainties obtained during testing and calibration at Alden Labs and the measurement uncertainties were found to be consistent.

3.6.2.2 Checkplus Inoperability To operate above the current licensed thermal power of 2546 MWt, the licensee proposes to use the Cameron LEFM CheckPlus System in the normal and the maintenance modes. If the UFM is not functional, the input for the calorimetric will not use the Cameron LEFM data. Instead, the input will revert to the original source from the venturis.

In the normal mode of operation, both planes of transducers are in service and system operations are processed by both central processing units (CPU). The LEFM reading will be used as input for the FW flow in the calorimetric.

- 30 The maintenance mode of operation is defined as a time when there is a failure involving a transducer, failure of one plane of operation or if a CPU-related malfunction occurs. If any of the stated conditions occur, the system reverts to the Cameron Check System or maintenance mode.

In the non-functional mode, the licensee will have 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to return the system to normal status any time the plant is operating above 2546 MWt. If the LEFM cannot be returned to functional status in the 48-hour allotted time, plant power must be reduced to 2546 MWt. Stated differently, the plant will be operated as though the Cameron CheckPlus was never installed and the MUR PU was not in effect. These actions are to be covered in the Technical Requirements Manual (TRM) and limiting condition of operation. The NRC staff finds that operation with an inoperable Cameron CheckPlus has been acceptably addressed.

3.6.2.3 Transducer Replacement Uncertainty associated with transducer replacement was addressed to VEPCO in the Cameron Customer Information Bulletin CIB-125, Revision 0, April 23, 2007. Since the transducer installation uncertainty is incorporated in the Surry 1 and 2 UFM System uncertainty and no additional uncertainty terms need to be applied whenever a transducer is replaced, the NRC staff finds that transducer installation variability has been acceptably addressed.

3.6.2.4 Cameron Checkplus Calibration Cameron CheckPlus calibration was accomplished at Alden Laboratories. The LAR covers the test configuration. The NRC staff reviewed drawings and schematics provided and confirmed that, insofar as configuration is concerned, the laboratory configuration largely matched the in-situ configuration.

The piping configuration used during testing included a full scale model of Surry 1 and 2 hydraulic geometry and a straight pipe. The calibration factor used for the UFM is based on the reports prepared by Alden Laboratories after the testing. Final uncertainty analysis will be completed after the completion of the commissioning process, prior to operation at the MUR PU level.

The tests were completed using previously applied procedures and laboratory measurement elements traceable to the National Institute of Standards and Technology. The NRC staff finds that the licensee's laboratory calibration was sufficiently fabricated to provide meaningful data based on the modeling of piping geometry of the UFM at Surry 1 and 2.

3.6.2.5 Cameron Topical Reports Safety Evaluation Criteria The NRC staff reviewed and approved Cameron Topical Reports ER-80P and ER-157P related to the Cameron LEFM CheckPlus Systems. In approving the Cameron Topical Reports, the NRC staff established four criteria to be satisfied by each licensee as follows:

3.6.2.5.1 Criterion 1 Discuss maintenance and calibration procedures that will be implemented with the incorporation of the Cameron LEFM, including processes and contingencies for inoperable Cameron LEFM instrumentation and the effect on thermal power measurements and plant operation.

- 31 The licensee will develop calibration and maintenance programs in accordance with the LEFM vendor recommendations. Maintenance and calibration will be conducted by site instrumentation and controls personnel qualified per the Surry 1 and 2 Instrumentation and Control Training Program. Examples of preventive maintenance activities that qualified personnel will perform include general terminal cleanliness and inspection, power supply inspections and transducer checks.

Licensee's Response:

Surry 1 and 2 allows an outage time of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for power levels in excess of 2546 MWt. The 48-hour allowed outage time is based on most repairs taking one shift of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to complete and allowing for time to diagnose the problem and plan for the repairs. In the event of a UFM failure, a backup calorimetric will be used. The backup calorimetric will be based on FW venturis and resistance temperature detectors. If the power level drops below 2546 MWt, the maximum power level of the plant will operate at until repair of the UFM is 2546 MWt.

The UFM functional requirements will be contained in the Surry 1 and 2 TRM.

Reactor operators will be provided with procedural guidance to use when the UFM is not functional.

3.6.2.5.2 Criterion 2 For plants that currently have Cameron LEFMs installed, provide an evaluation of the operational and maintenance history of the installed instrumentation and confirmation that the installed instrumentation is representative of the Cameron LEFM System and bounds the analyses and assumptions set forth in "Improving Thermal Power Accuracy and Plant Safety While Increasing Power Level Using the Cameron LEFM Check Systems," Cameron Engineering Report.

Licensee's Response:

Surry 1 and 2 installed the UFM spool pieces in Unit 1 during the spring 2009 refueling outage and in Unit 2 during the fall 2009 refueling outage. Monitoring will begin after final commissioning and calibration which was expected to be completed prior to operating above 2546 MWt (98.4% RP). There have been no maintenance activities completed since installation of the UFM spool pieces.

3.6.2.5.3 Criterion 3 Confirm that the methodology used to calculate the uncertainty of the Cameron LEFM in comparison to the current FW instrumentation is based on accepted plant setpoint methodology (With regard to the development of instrument uncertainty). If an alternative approach is used, the application should be justified and to both venturi and ultrasonic flow measurement (UFM) instrumentation installations for comparison.

Licensee's Response:

VEPCO uses a methodology consistent to the approved methodologies in the Cameron topical reports to determine the uncertainty of the Cameron LEFM.

- 32 VEPCO uses a statistical approach to determine uncertainty in the setpoint methodology. Dependent parameters are arithmetically combined to form statistically independent groups and then combined using the square root of the sum-of-the-squares approach to determine the overall uncertainty. The same fundamental approach was used to determine the UFM-based power calorimetric uncertainty.

3.6.2.5.4 Criterion 4 For plants where the ultrasonic meter (including Cameron LEFM CheckPlus System) was not installed and flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors not representative of the plant-specific installation), additional justification should be provided for its use. The justification should show that the meter installation is either independent of the plant-specific flow profiles for the stated accuracy, or that the installation can be shown to be equivalent to known calibrations and plant configurations for the specific installation including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed calibrated elements, confirm that the piping configuration remains bounding for the original Cameron LEFM CheckPlus System installation and calibration assumptions.

Licensee's Response:

The site-specific calibration was conducted at Alden Labs. The piping configuration used during testing included a full scale model of Surry Unit 1 and Unit 2 hydraulic geometry and a straight pipe. The calibration factor used for the UFM is based on the reports prepared by Alden Labs after the testing. Final uncertainty analysis will be completed after the completion of the commissioning process.

Based on its review of the licensee's responses, the NRC staff finds this evaluation acceptable and has determined that the licensee has addressed the four criteria specified in the NRC staff's evaluation of Topical Reports ER-80P and ER-175P and the evaluation is consistent with the guidelines of RIS 2002-03.

3.6.3 Nuclear Steam Supply System Parameters The NSSS design parameters provide the RCS and secondary system conditions (pressures.

temperatures, and flow) that are used as the basis for the design transients and for systems, components, accidents and transient analyses and evaluations. The parameters are established using conservative assumptions to provide bounding conditions to be used in the NSSS analyses.

3.6.4 Accident Analyses Bounded By Current Analysis of Record Although the licensee concluded that existing analyses were bounding of uprated plant operation with reduced uncertainty, the analyses were shown to be bounding in one of three different ways:

For analyses that assume steady-state plant operation with a core power of 2597 MWt, there is a 2-percent margin for power measurement uncertainty at the CLTP, 2546 MWt.

These analyses bound plant operation at the MUR RP of 2587 MWt, with an operating

- 33 margin of 0.38 percent, which is greater than the stated 0.35-percent calorimetric power measurement uncertainty.

For analyses that assume steady-state plant operation with a core power of 2546 MWt, the licensee evaluated accidents or transients, and reanalyzed as necessary.

• Zero-power transients were not reanalyzed.

The licensing basis transients and accidents are summarized in Table 1. RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," states the following:

In areas (e.g., accident/transient analyses, components, systems) for which the existing analyses of record do bound plant operation at the proposed uprated power level, the staff will not conduct a detailed review.

The NRC staff therefore finds the licensee's existing analyses that were performed at or above 102 percent of the current licensed thermal power level acceptable without detailed review. In the sections that follow Table 1, discussion is provided for those existing analyses that were performed at less than 102 percent of CLTP.

- 34 Table 1 - Accident and Transient Analyses Transient!Accident Analytic Power Level Review Comments

(% CLTP)

Uncontrolled Control-Rod Assembly Withdrawal From a Subcritical Condition 0

See Section 3.6.4.1 Uncontrolled Control-Rod Assembly Withdrawal at Power RCS/MSS Pressure 102 See Section 3.6.4.2 DNBR (Statistical) - 100 Control Rod Assembly Drop/Misalignment 100 See Section 3.6.4.3 CVCS Malfunction 102 Acceptable Excessive Heat Removal Due to RCS/MSS Pressure 102 See Sections 3.6.4.2 and 3.6.4.4 Feedwater System Malfunctions DNBR (Statistical) - 100 Excessive Load Increase Incident 100 See Section 3.6.4 Loss of Reactor Coolant Flow RCS/MSS Pressure 102 See Section 3.6.4.2 (Includes Locked Rotor)

DNBR (Statistical) - 100 Loss of External Electrical Load RCS/MSS Pressure 102 DNBR (Statistical) - 100 Loss of Normal Feedwater 102 Acceptable Loss of All Alternating Current Power to the Station Auxiliaries 102 Likelihood of a Turbine-Generator Unit Overspeed Not Analyzed See Section 3.6.4.5 Steam Generator Tube Rupture 102 Acceptable Rupture of Main Steam Pipe 0

See Section 3.6.4.6 Rupture of a Control Rod Drive Mechanism Housing (Control Rod Assembly Ejection) 102 Acceptable Large Break Loss-of-Coolant Accident 102 Small Break Loss-of-Coolant Accident 102 Natural Circulation Cooldown 102 Anticipated Transients Without Scram/

ATWS Mitigation Systems Actuation Circuitry 109.4 Station Blackout 102

- 35 3.6.4.1 Uncontrolled Control-Rod Assembly Withdrawal from a Subcritical Condition The licensee stated that, because the control rod withdrawal from subcritical is analyzed at zero-power conditions, its analysis is unaffected by the proposed MUR PU. Therefore, the licensee did not reanalyze this transient. The NRC staff finds that the zero-power transient analysis with respect to the proposed MUR PU acceptable because it is unaffected by core power level.

3.6.4.2 Statistical Transients Analyzed at Nominal Power Level Compliance with fuel cladding integrity safety limits, one of the transient analysis acceptance criteria is assured by demonstrating that the DNBR on the analyzed hot fuel rod, remains above a set limit. In analytic space, this limit will incorporate uncertainties associated with the experimental correlation used to determine the DNBR, uncertainties associated with plant parameters, and an amount of retained margin.

Since plant parameter uncertainties are incorporated into the DNBR limits, the transient analyses are performed assuming a nominal power level. At Surry 1 and 2, this is 2546 MWt, an amount that is non-bounding of the proposed uprated power level. To account for this, the licensee determined the effects that the increased power level would have on the transient reductions in DNBR. The licensee determined that the 1.6-percent increase in fuel rod heat flux associated with the MUR PU will cause a 3.3-percent decrease in DNBR. The licensee therefore proposed to retain a 3.3-percent DNBR penalty to account for the MUR PU effects associated with those transients analyzed at 2546 MWt.

The NRC staff requested that the licensee provide an evaluation of other parameters, aside from fuel rod surface heat flux, affecting the DNBR and confirm that the 3.3-percent decrease is bounding of those parameters, as well. The licensee responded, stating that all other parameters affecting the DNBR would either remain the same at uprated conditions, or produce a DNBR benefit that was conservatively ignored.

The licensee stated that the 3.3-percent DNBR penalty is acceptable because there is sufficient retained margin in the DNBR limits to account for it. The NRC staff requested that the licensee provide a detailed DNBR margin evaluation to demonstrate that there is sufficient retained margin to account for the requested MUR PU. The licensee responded by listing all current DNBR penalties and comparing them against the 13-percent retained DNBR margin. In the most-limiting case, including the 3.3-percent DNBR penalty left 0.7 percent of retained margin.

The NRC staff finds the licensee's evaluation of those transients analyzed using statistical methodology at CLTP acceptable with regard to the proposed MUR PU for two reasons. First, the licensee acceptably evaluated the impact that the MUR PU would have on those analyses, and determined that a 3.3-percent DNBR penalty would appropriately account for that impact.

Second, the licensee demonstrated that there is sufficient retained DNBR margin to incorporate the 3.3-percent penalty. Based on the consideration above, the NRC staff finds the licensee's evaluation acceptable because the licensee demonstrated that the existing analyses bound operation at the proposed uprated power level.

- 36 3.6.4.3 Rod Cluster Control Assembly Misalignment (System Malfunction or Operator Error)

The licensee evaluated the rod cluster control assembly misalignment at the dropped rod limit lines. The dropped rod limit lines determine the allowable radial peaking factor at the limiting point of the transient. The NRC staff requested additional information to determine whether the dropped rod limit lines are confirmed on a cycle-specific basis, and the licensee responded that this event is evaluated during each fuel cycle and has been evaluated for the MUR PU and remains bounding. Therefore, the NRC staff finds the rod cluster control assembly misalignment analysis acceptable.

3.6.4.4 Excessive Heat Removal Due to Feedwater System Malfunction There are several types of FW system malfunctions that are analyzed in the Surry 1 and 2 licensing basis. The licensee reviewed existing analyses for an excessive FW flow event, and for a FW temperature reduction event. Both of these transients would increase the core cooling, causing an insertion of positive reactivity and a corresponding power increase.

The analysis for the excessive FW flow event was performed to confirm conformance to DNBR limits prior to the implementation of statistical methods, such that the analysis was performed assuming a core power level of 102 percent of the CLTP. This analysis is bounding of implementation of the proposed MUR PU and is hence acceptable.

The FW temperature reduction was performed using the statistical analysis method. This means that the RCS and MSS pressure analysis was performed using a bounding power level, and the DNBR analysis was performed assuming the nominal CLTP level. In addition, licensee has assessed a 3.3-percent DNBR penalty to account for uprated operation. The NRC staff finds this analysis acceptably bounding of the requested MUR PU, as discussed in Section 3.6.4 of this SE.

3.6.4.5 Likelihood of a Turbine-Generator Unit Overspeed The Surry 1 and 2 FSAR contains this transient; however, the licensee stated that the turbine-generator speed is constant and not dependent upon reactor power level. The licensee characterized this event as a turbine missile analysis that accounts for the effects of turbine-generator speeds in excess of 120 percent of the rated speed. The licensee also stated that the evaluations and conclusions related to turbine overspeed protection are not affected by the proposed uprated conditions. The NRC staff finds the disposition provided by the licensee acceptable because it accounts for 120 percent overspeed events while the turbine-generator speed will remain constant at the proposed uprated conditions.

3.6.4.6 Rupture of a Main Steam Pipe The licensee stated that the steam line break is limiting at zero power conditions, and that a re-analysis was therefore not necessary. The NRC staff requested additional information about the non-limiting nature of the full-power MSLB analysis to determine why re-analysis is not necessary.

- 37 The licensee provided additional information, explaining that the full-power MSLB analysis relies on initial conditions that include more stored energy in the RCS, higher average coolant temperature, and appreciable stored energy in the fuel when compared to the no-load conditions of the zero-power analysis. The licensee also stated that these sources of additional stored energy are removed via the cooldown caused by the MSLB before the no-load conditions are reached. From that point, the licensee stated that the transient proceeds in a manner similar to the zero-power transient. However, the full-power analysis starts with significantly less SG liquid mass, which causes the magnitude and duration of the RCS cooldown to be greater for the break at zero-power conditions.

The licensee also stated that the basis for this conclusion is contained in WCAP-9226-PA, "Reactor Core Response to Excessive Secondary Steam Releases" (ADAMS Accession No. ML093630006, Non-Public). Chapter 3 of this report explains how, following a plant trip for postulated breaks occurring from power, the plant transient is bounded by the more limiting breaks from hot shutdown. The end conclusion of the report is that, because the rods and core following the plant trip are in the same configuration as the cases initiated from zero power, the plant parameters approach those of the breaks from zero power.

The licensee stated that the generic evaluations in WCAP-9226-PA are applicable and bounding for Surry 1 and 2 operation at the proposed uprated power level because the generic study was performed using a 3-loop Westinghouse NSSS like Surry 1 and 2, and analyzed at a maximum power level of 2840 MWt, which is bounding of the proposed power level and measurement uncertainty at Surry 1 and 2, which is 2597 MWt. The NRC finds the analysis acceptable because the conditions remain bounded for operation at the proposed MUR PU.

3.6.4.7 Large Break Loss-of-Coolant Accident Surry 1 and 2 UFSAR Section 14.5.1 describes the large break LOCA analysis for the Westinghouse Surry Improved Fuel product. The analysis applies the NRC-approved Westinghouse ASTRUM Best-Estimate LOCA (BELOCA) analysis methodology for calculation of Peak Cladding Temperature and oxidation (local and whole-core). The NRC approved the use of the ASTRUM methodology for the Surry 1 and 2 BELOCA analysis. Since NRC approval, the BELOCA analysis has been augmented by evaluations under 10 CFR 50.46. The analysis of record uses a core power of 2597 MWt, which is 102 percent of 2546 MWt, with no additional core power uncertainty applied. Therefore, the analyzed core power is bounding for the MUR PU.

UFSAR Section 14.5.1.6 concludes that the LOCA long-term core cooling requirement of 10 CFR 50.46(b)(5) is met. Implicit in that conclusion is the acceptability of the ECCS long-term water supply to the core and the procedures to mitigate the build-up of boric acid in the core. The analysis of record for post-LOCA containment sump boron concentration (subcriticality),

containment sump pH, and hot leg switchover time were reviewed for the MUR PU. The most recent NRC approval of these analyses was in a license amendment that increased the boron concentration limits of the refueling water storage tank and accumulators. These analyses have been supplemented by additional evaluations performed under the provisions of 10 CFR 50.59.

The following evaluations confirm that the analyses of record remain bounding for the proposed MUR PU, and that long-term cooling is assured.

- 38

• The containment sump pH calculation does not explicitly include a core power level. The methodology normalizes the contributing inventories to a sump temperature of 70°F. The proposed core power increase does not affect the analysis that determines the post-LOCA sump pH.

• The minimum containment sump boron concentration calculation that ensures post-LOCA subcriticality does not explicitly include a core power level. Each core reload confirms that the post-LOCA sump boron concentration provides adequate subcriticality during the vessel reflood stage, the switchover to cold leg recirculation, and during long-term core cooling. The core reload confirmation of post-LOCA sump boron concentration limits accounts for the core power level.

• The hot-leg switchover time calculation uses a core power level of 2597 MWt, or 102 percent of 2546 MWt, to determine the post-LOCA core steaming rate.

The NRC staff finds that this analysis will continue to bound operation of the plant at the uprated power level of 2587 MWt.

3.6.4.8 Waste Gas Decay Tank Rupture The waste gas decay tank (WGDT) rupture analysis was submitted as part of the UFSAR and approved in the associated NRC SE dated April 11, 1991, (ADAMS Accession No. ML012540130). The WGDT analysis was subsequently updated under the provisions of 10 CFR 50.59. The WGDT rupture was analyzed consistent with the activity limit defined in TS 3.11.B. The result of the analysis was an EAB whole-body dose of less than 0.5 Rem.

Therefore, the NRC staff finds these analyses acceptable.

3.6.4.9 Fuel-Handing Accident The analysis of record fuel-handling accident is based on the alternate source term as defined in NUREG-1465 (ADAMS Accession No. ML003756729), with acceptance criteria as specified in either 10 CFR 50.67 or RG 1.183. This analysis was completed assuming a core inventory of 102 percent of 2546 MWt with a single failed fuel assembly of 204 rods. The analyses of record bounds the MUR PU, and therefore, NRC staff finds this acceptable.

3.6.4.10 Analysis to Determine Environmental Qualification (EQ) Parameters The licensee's analysis assumed that all normal non-radiological plant operating conditions within all environmental zones (i.e., temperature, pressure, humidity) remained unchanged for MUR PU operation. A separate evaluation was performed to assess potential increase in normal operation radiation dose used in the EQ program. In general, MUR PU operation would be expected to increase the core inventory of radioisotopes by the percentage increase in core power and potentially to increase the normal operation radiation source term. However, this potential increase in radiation source term will not affect the currently estimated normal operation doses used for EQ because of several conservative factors incorporated into the current estimates. The most significant of these considerations are: (a) use of a dose for a given radiation zone designation that represents the maximum end of the normal operation range, and (b) the limitation imposed by plant operations as a result of TS limits on RCS coolant activity (i.e.,

allowable limits of operation are approximately one-third of the value associated with the

- 39 assumed 1 percent fuel defects used in the normal operation source term). The conditions used in the EO program for normal operation therefore remain bounding for the MUR PU, with the exception of the dose levels for the reactor vessel excore neutron detectors. The excore detectors radiation dose increases such that the EO in-service life may be decreased. These excore detectors are scheduled to be replaced on Surry 1 in the fall of 2010 and Surry 2 in the spring of 2011. Prior to operating above 2546 MWt (98.4% RP), VEPCO will determine the EO service life of the excore detectors. A calculation is being developed to evaluate the dose impact on these detectors. Based on results of the comparable North Anna Power Station calculation, there is no anticipated impact on radiation dose margin or qualified life. Therefore, the NRC staff finds this acceptable.

3.6.5 Conclusion The NRC staff reviewed the reactor systems and thermal-hydraulic aspects of the proposed LAR in support of implementation of an MUR PU. Based on the considerations discussed above, the NRC staff determined that the results of the licensee's analyses related to these areas continue to meet applicable acceptance criteria following implementation of the MUR PU. Most of the current analyses of record are based on operation at 2597 MWt, which includes 2.0-percent measurement uncertainty. The proposed amendment is based on the use of a Cameron LEFM CheckPlus System that would decrease the uncertainty in the FW flow, thereby decreasing the power level measurement uncertainty from 2.0 percent to 0.35 percent. In these cases, the proposed MUR PU rated thermal power of 2587 MWt is bounded by the current analyses of record. Therefore, the NRC staff finds this acceptable.

3.7 Reactor Pressure Vessel Integrity The NRC staffs review in the area of reactor vessel (RV) integrity focuses on the impact of the proposed MUR PU on neutron fluence calculations, the RV surveillance capsule withdrawal schedules, RV pressure-temperature (P-T) limits, upper shelf energy (USE) evaluations, and pressurized thermal shock (PTS) calculations. This review was conducted, consistent with the guidance contained in RIS 2002-03, to verify that the results of licensee analyses related to these areas continue to meet the requirements of 10 CFR Part 50, Sections 50.60 and 50.61, and 10 CFR Part 50, Appendices G and H following implementation of the proposed MUR PU.

3.7.1 Reactor Vessel (RV) Material Surveillance Program 3.7.1.1 Regulatory Evaluation The RV material surveillance program provides a means for determining and monitoring the fracture toughness of the RV beltline materials to support analyses for ensuring the structural integrity of the ferritic components of the RV. Appendix H of 10 CFR Part 50 provides the staff's requirements for the design and implementation of the RV material surveillance program.

3.7.1.2 Technical Evaluation In its LAR, the licensee stated that the NRC-approved RV surveillance capsule withdrawal schedules for Surry 1 and 2 are contained in the UFSAR and the surveillance capSUle withdrawal schedules are based on American Standard Testing of Materials (ASTM) ASTM E-185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power

- 40 Reactor Vessels." Per ASTM E-185-82, the withdrawal of a capsule is to be scheduled at the nearest vessel refueling outage to the calculated effective full power years (EFPY) established for the particular surveillance capsule withdrawal.

Sy letter dated October 26, 2009 (ADAMS Accession No. Ml092990570), VEPCO submitted proposed revised RV materials surveillance capsule withdrawal schedules for Surry 1 and 2. The proposed schedules were developed to accommodate the 60-year licensing period for Surry 1 and 2. This submittal is currently under separate review.

The surveillance program requirements in Appendix H of 10 CFR Part 50, were established to monitor the radiation-induced changes in the mechanical and impact properties of the RV materials. Appendix H of 10 CFR Part 50 requires licensees to monitor changes in the fracture toughness properties of ferritic materials in the RV beltline region of light-water nuclear power reactors. Appendix H of 10 CFR Part 50 states that the design of the surveillance program and the withdrawal schedule must meet the requirements of the edition of ASTM E-185 that is current on the issue date of the ASME Code, to which the RV was purchased. later editions of ASTM E-185 may be used including those editions through 1982 (i.e., ASTM E-185-82). NUREG-1801 (ADAMS Accession No. Ml012060521), "Generic Aging lessons learned Report," provides additional guidance for the surveillance program for the 60-year extended period of operation.

The RV materials surveillance programs for Surry 1 and 2 were originally developed under ASTM E-185-73. VEPCO is using the requirements of ASTM E-185-82 as its basis for meeting the RV materials surveillance capsule withdrawal requirements of Appendix H to 10 CFR Part 50. Table 1 of ASTM E-185-82 requires that either a minimum of three, four, or five surveillance capsules be removed from each of the vessels, as based on the projected nil ductility reference temperature shift (f1RTNDT) of the limiting material at the clad-vessel interface location of the RV at the end-of-Iicensed plant life (EOl). ASTM E-185-82 recommends that for Surry 1 and 2, at a minimum, the number of RV materials surveillance capsules consists of at least five capsules.

VEPCO has removed and tested four capsules from each unit and has proposed a change in a withdrawal schedule for the fifth capsule.

3.7.1.3 Conclusion The NRC staff concludes that the licensee's current surveillance capsule withdrawal schedules, approved in the lAR for Surry 1 and 2, are acceptable with the understanding that the revised RV materials surveillance capsule withdrawal schedules for Surry 1 and 2, submitted by letter dated October 26, 2009 (ADAMS Accession No. Ml092990570), are currently under review and any revisions to the RV materials surveillance programs must be incorporated in the appropriate sections (i.e., Tables 4.1-12 and 4.1-13) of the Surry 1 and 2 UFSAR in accordance with the requirements of 10 CFR 50.71(e).

3.7.2 P-T Limits and Use 3.7.2.1 Regulatory Evaluation Appendix G of 10 CFR Part 50, provides fracture toughness requirements for ferritic (low alloy steel or carbon steel) materials in the RCPS, including requirements on the USE values used for assessing the safety margins of the RV materials against ductile tearing and for calculating P-T limits for the plant. These P-T limits are established to ensure the structural integrity of the ferritic

- 41 components of the RCPB during any condition of normal operation, including anticipated operational occurrences and hydrostatic tests. The NRC staff's review of the USE assessments covered the impact of the MUR PU on the neutron fluence values for the RV beltline materials and the USE values for the RV materials through the end of the current licensed operating period. The NRC staff's P-T limits review covered the P-T limits methodology and the calculations for the number of EFPY specified for the proposed MUR PU considering neutron embrittlement effects.

3.7.2.2 Technical Evaluation The current P-T Limits and low temperature overpressurization protection system (lTOPS) setpoints for Surry 1 and 2 were developed in 1998 based on Y<!T and %T adjusted reference temperature (ART) values of 228.4 OF and 189.5 OF, respectively. The limiting material in 1998 was identified as the Surry 1 intermediate-to-Iower shell circumferential weld. SUbsequently, by letter dated June 13, 2006 (ADAMS Accession No. Ml061650080), the licensee revised the initial nil-ductility reference temperature (initial RTNDT) values for the Surry 1 and 2 Linde 80 weld materials based on the use of topical report BAW-2308, Revision 1-A, "Initial RTNDT of Linde 80 Weld Materials." This significantly affected the assessment of the Surry 1 intermediate-to-Iower shell circumferential weld such that it is no longer projected to be the limiting material for Surry 1.

In the attachment to the licensee's letter of June 13, 2006, the limiting material for Surry 1 was redefined to be lower Shell longitudinal Weld l2 with Y<!T and %T ART values 171.3 OF and 113.3 OF, respectively.

In the MUR PU application, ART calculations were performed for the Surry 1 and 2 RV materials based on updated 48 EFPY neutron fluence values. The most limiting Y<!T and %T ART values for either Surry unit were determined to be 222.5 OF and 188.6 OF, respectively, and were based on the evaluation of the Surry 2 intermediate-to-Iower shell circumferential weld. The NRC staff independently checked these ART values and confirmed the licensee's results.

Since the Y<!T and %T ART values through 48 EFPY for the new Surry limiting material (Surry 2 intermediate-to-Iower shell circumferential weld, after consideration of the effects of the MUR PU) remained bounded by the Y<!T and %T ART values used to construct the current Surry 1 and 2 P-T limits and establish the current Surry 1 and 2 lTOPS setpoints, the current P-T limits curves and l TOPS setpoints remain valid through at least their currently approved periods of 28.8 EFPY and 29.4 EFPY for Surry 1 and 2, respectively. The licensee will, however, need to modify the period of applicability of their current P-T limits, or submit revised P-T limits, before the current curves expire.

3.7.2.3 Conclusion The licensee also addressed the impact of the MUR PU on the Surry 1 and 2 USE evaluations.

These analyses are documented in the attachment to the licensee's letter of June 13, 2006, "Update to NRC Reactor Vessel Integrity Database and Exemption Request for Alternative Material Properties Basis per 10 CFR 50.60(b)" (ADAMS Accession No. Ml061650080), for Surry 1 and 2. Since the updated projected EOl neutron fluence values, which reflect the effects of the MUR PU, were lower than the values upon which the prior USE evaluations were based, the projected USE values for this MUR PU application are bounded by the earlier analyses. Therefore, the NRC staff finds the proposed MUR PU to be acceptable with respect to the P-T limits and USE.

- 42 3.7.3 Pressurized Thermal Shock (PTS) 3.7.3.1 Regulatory Evaluation The PTS evaluation provides a means for assessing the susceptibility of PWR RV beltline materials to failure during a PTS event to assure that adequate fracture toughness exists during reactor operation. The staff's requirements, methods of evaluation, and safety criteria for PTS assessments are given in 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressure Thermal Shock Events. " The NRC staff's review covered the PTS methodology and the calculations for the reference temperature for pressurized thermal shock (RTPTS) at the expiration of the license, considering neutron embrittlement effects.

3.7.3.2 Technical Evaluation The PTS calculations were performed for Surry 1 and 2 using the procedures specified in 10 CFR 50.61. Updated neutron fluence projections, corresponding to the EOL conditions, were used in the PTS analyses. These analyses are documented in the attachment to the licensee's letter of June 13, 2006 (ADAMS Accession No. ML061650080), for Surry 1 and 2. In this submittal, RTPTS calculations were performed in accordance with RG 1.99, Revision 2, (ADAMS Accession No. ML003740284) for the Surry 1 and 2 RV beltline materials at 60-year fluences. Unirradiated RTNOT values, conditions, and associated uncertainties contained in BAW-2308, Revision 1-A, "Initial RTNDTof Linde 80 Weld Materials," (ADAMS Accession No. ML051570314) were used in calculations for the Linde 80 RV beltline welds. For Surry 1, the limiting material is Lower Shell Longitudinal Weld L2 with an RTPTS value of 201.8 OF.

This is lower than the PTS screening criterion of 270 OF. For Surry 2, the limiting material is the Intermediate-to-Lower Shell Circumferential Weld with an RTPTS value of 236.4 OF. This is lower than the PTS screening criterion of 300 OF for circumferential welds. Therefore the NRC staff concludes that the Surry 1 and 2 RVs will remain within their limits for PTS after the MUR PU and the RV materials would continue to meet the PTS screening criteria requirements of 10 CFR 50.61.

3.7.4 Reactor Vessel (RV) Internals and Core Support Materials 3.7.4.1 Regulatory Evaluation The RV internals and core support structures include SSCs that perform safety functions or whose failure could affect safety functions performed by other SSCs. These safety functions include reactivity monitoring and control, core cooling, and fission product confinement (within both the fuel cladding and the reactor coolant pressure boundary). The NRC's acceptance criteria for RV internals and core support materials are based on GDC-1 and 10 CFR 50.55a for material specifications, controls on welding, and inspection of RV internals and core supports.

Matrix 1 of NRC RS-001, Revision 0, "Review Standard for Extended Power Uprates," (ADAMS Accession No. ML023610659), provides references to the NRC's approval of the recommended guidelines for RV internals in Topical Reports WCAP-14577, Revision 1-A, "License Renewal Evaluation: Aging Management for Reactor Internals" (ADAMS Accession No. ML010290348, Non-Public), and BAW-2248-A, "Demonstration of the Management of Aging Effects for the Reactor Vessel Internals" (ADAMS Accession No. ML003708443).

- 43 3.7.4.2 Technical Evaluation In an RAI, the NRC staff asked the licensee to provide a description of plant-specific degradation management programs or participation in industry programs to investigate degradation effects and determine appropriate management programs. The licensee discussed the impact of the Surry 1 and 2 MUR PU on the structural integrity of the RV internal components in Attachment 5 of the LAR,Section IV.1.A.ii and the licensee's RAI response dated April 29, 2010 (ADAMS Accession No. ML101200269). The licensee concluded that the RV internals and core support structures are not adversely affected either by the MUR PU reactor coolant system conditions and transients or by secondary effects on reactor thermal-hydraulic or structural performance.

Therefore, the NRC staff found the analyses acceptable and determined that based on the evaluation of the licensee's responses to the RIS-2003 and the RAI, that the existing loads remain valid and the stresses and fatigue values remain valid.

The RV internals of PWR-designed light-water reactors may also be susceptible to the following aging effects:

• cracking induced by thermal cycling (fatigue-induced cracking), SCC, or irradiation assisted stress corrosion cracking (IASCC);

• loss of fracture toughness properties induced by radiation exposure for all stainless steel grades, or the synergistic effects of radiation exposure and thermal aging for cast austenitic stainless steel (CASS) grades;

• stress relaxation in bolted, fastened, keyed or pinned RV internal components induced by irradiation exposure and/or exposure to elevated temperatures; and

• void swelling (induced by radiation exposure).

Matrix 1 of NRC RIS-001, Revision 0 (ADAMS Accession No. ML033640024), provides the NRC staff's basis for evaluating the potential for extended PUs to induce these aging effects. In Note 1 to Matrix 1, the NRC staff stated that guidance on the neutron irradiation-related threshold for IASCC for PWR RV internals are given in BAW-2248-A (ADAMS Accession No. ML003708443) and WCAP-14577, Revision 1-A. This Matrix 1 note further stated that for thermal and neutron embrittlement of CASS, SCC, and void swelling, licensees will need to provide plant-specific degradation management programs or participate in industry programs that investigate degradation effects and determine appropriate management programs.

In its RAI response dated April 29, 2010 (ADAMS Accession No. ML101200269), the licensee discussed its plant-specific RV internals program. The program is based on the Electric Power Research Institute (EPRI) "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines" (Materials Reliability Program (MRP)-227). The MRP-227 guidelines consider various aging factors including neutron fluence exposure, temperature history, and representative stress levels for determining relative susceptibility of PWR internals to postulated aging mechanisms that include SCC, IASCC, wear, fatigue, thermal aging embrittlement, irradiation embrittlement, irradiation-enhanced stress relaxation and creep, and void swelling. In the LAR for Surry 1 and 2, the licensee stated that inspections will be performed to implement industry recommendations and that inspection plans were under development. WCAP-14577, Revision 1-A, "License

- 44 Renewal Evaluation: Aging Management for Reactor Internals," (ADAMS Accession No. ML010290348, Non-Public), was used in the development of the License Renewal Application (LRA) for Surry 1 and 2 (ADAMS Accession No. ML011500496). Table 3.1.3-W1, "FSER [Final Safety Evaluation Report] Response to Applicant Action Items", of WCAP-14577, Revision 1-A confirmed the applicability of WCAP-14577, Revision 1-A, to Surry 1 and 2. The licensee noted that since the issuance of WCAP-14577, further materials research and functional requirements review have resulted in additional recommendations for aging management programs contained in MRP-227. These recommendations are specifically configured for the combinations of degradation mechanisms and component functional requirements of the RV internals, and include recommendations for specific plant design features. Therefore, in accordance with the commitment to consider industry guidance, the inspections planned for Surry 1 and 2 prior to the period of extended operation will be consistent with the augmented examinations required by MRP-227. ASME Code,Section XI, examination programs for internals will continue to apply to other components not requiring augmented examinations. Since the Surry MUR PU results in very small changes to aging parameters such as temperature and neutron flux, the current aging management program for the reactor internals is acceptable. However, the NRC staff will further review the Surry 1 and 2 RV internals program when the licensee submits the program for NRC review to fullll1 its license renewal commitment.

3.7.5 Conclusion The NRC staff has reviewed the licensee's proposed LAR to increase the rated core thermal power by approximately 1.6 percent and has evaluated the impact that the MUR PU conditions will have on the structural integrity assessments for the RV and RV internals. The staff has determined that the changes identified in the proposed LAR will not impact the remaining safety margins required for the following structural integrity assessments: (1) RV surveillance program; (2) RV USE assessment; (3) P-T limits; (4) PTS assessment; and (5) RV internals and core support structures. Therefore, the NRC staff finds that the information provided in response to the RIS 2002-03 and RAls provided supports the requested MUR PU and is acceptable.

3.8 Electrical Systems 3.8.1 Regulatory Evaluation The licensee developed the LAR consistent with the guidelines in NRC RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Uprate Applications."

The regulatory requirements which the staff applied in its review of the application include:

• GDC 24 and 39, require that sufficient alternative sources of power are provided and designed with adequate independence, redundancy, capacity, and testability to permit the functioning of the protection system in the event of loss of all offsite power. Also, both onsite and offsite power systems need to provide the required capacity assuming the failure of a single active component in each power system.

• Section 50.63, "Loss of all AC [alternating current] power," requires, in part, that all nuclear plants have the capability to withstand a loss of all AC power (station blackout (S80)) for an established period of time, and to recover there from.

3.8.2

- 45

• Section 50.49, "Environmental Qualification (EQ) of Electric Equipment Important to Safety for Nuclear Power Plants," requires licensees to establish programs to qualify electrical equipment important to safety.

Technical Evaluation 3.8.2.1 AC Distribution System The AC Distribution System is the source of power for the nonsafety-related buses and for the safety-related emergency buses. It consists of the 4.16 kilovolt (kV), 480 volt (V), and 120 V systems (not including the emergency diesel generators). The licensee stated that the MUR PU will only affect the 4.16 kV buses while the 480 V, and 120 V, buses will not see a load increase or are independent of the MUR PU. The 4.16 kV loads that will be affected are the main FW pumps, condensate pumps, low pressure heater drain pumps, high pressure heater drain pumps, bearing cooling pumps, and RCP. The licensee stated that the increase in these loads will not exceed the motor ratings and the 4.16 kV buses will have sufficient capacity.

The NRC staff requested additional information on the increased loadings of the RCP motors and other loads, and their impact on the electrical distribution system. By letter dated April 29, 2010 (ADAMS Accession No. ML101200269), the licensee provided its response to the RAI. The licensee stated that the RCP motors temperature rise for hot/cold loop loading and starting conditions still complies with the motor specifications. In addition, the licensee stated the following:

The increased loading of Station Service buses has minimal impact on the emergency buses. The voltage profile calculation simulates two types of transients; unit trip or unit accident. For these events, the transfer of Station Service buses to the Reserve Station Service transformers is delayed approximately 30 seconds after a turbine trip. For an accident scenario, the initial large motor starting load block is assumed to occur prior to the load transfer and is therefore unaffected. A calculation was performed to determine the impact on the voltage profiles. As expected, the emergency bus voltages are decreased after the delayed transfer due to the higher loading. The resulting voltages are fully acceptable and do not approach the Degraded Voltage relay settings. The Reserve Station Service transformer automatic load tap changers will continue to correct voltage after the load transfer. No large, challenging, motor starting load blocks occur after the load transfer. The Degraded Voltage relay settings are unaffected. The voltage profile calculation performed to evaluate the impact of the Surry MUR also determined that the Reserve Station Service transformer loading remains well below the transformer ratings. The existing load shedding schemes continue to limit loading adequately. The transformers are rated for 30 MVA [mega volt-ampere] and the maximum calculated loading is 25.8 MVA.

The NRC staff reviewed the licensee's response and concludes that the AC power system will experience minor load changes, however the AC system has adequate capacity to operate the plant equipment within the design to support the MUR PU.

- 46 3.8.2.2 Power Block Equipment (Generator, Exciter, Transformers, Iso-phase bus duct, Generator circuit breaker)

VEPCO assessed the impact of a 180 MWe (Le., 90 MWe per unit), of new generation capacity on the transmission system. VEPCO is anticipating additional plant modifications that would result in additional electrical power increases beyond that proposed by this MUR LAR. Grid stability studies were conducted assuming that power increases were in effect, so the results bound the MUR PU. The transmission system assessment was based on the Pennsylvania, New Jersey, Maryland Interconnection's (PJM) best assumptions at the present time for load growth and new generation through the summer of 2012. The evaluation included load flow studies of import/export system conditions and single-contingency, both normal and stressed, system conditions.

The local generation study assessed station operation at maximum capability. The study identified no transmission deficiencies. The import/export study assessed conditions into and out of the VEPCO system. Any new facility interconnected with the VEPCO system should not significantly decrement First Contingency Incremental Transfer Capability between utilities. The study indicated no decrement to system First Contingency Incremental Transfer Capability. In the summary section of the PJM system impact studies, the current maximum facility output (MFO) is 842 MWe for each unit. The MUR PU will increase each unit's generating capacity by approximately 15 MWe. The PJM system impact studies (SIS) describe the final power output values used in the stability analysis.

Gross and net generator output MWe values used in the P~IM impact and stability studies are bounding values that will not be exceeded during actual operation. These MWe values include expected additional MWe output due to MUR power increase, plus more efficient steam turbines, with additional margin. The design inputs consist of 28 MWe for house loads, 15 MWe for MUR, and 75 MWe for turbine replacement. The gross generator output, for each unit, is derived by adding the house loads and the MUR loads to the MFO of 842 MWe for each unit.

For Surry 1, the generator is rated at 1055 MVA. Currently, the generator output is 850.2 MWe.

At uprated conditions, the output will be 864.7 MWe while being capable of exporting 500 mega volt-ampere reactive (MVAR) and importing 430 MVAR. The licensee stated in the responses to the RAI that the current station operating procedures limit MVAR to 400 lagging and 200 MVAR leading, and that the higher MVAR capability would only be applicable after future turbine upgrade and increase in the hydrogen pressure. However, the 864.7 MWe for Unit 1 is a gross MWe value and does not take into account the approximate 28 MWe of internal electrical loads the plant represents to the generator output for each unit. Subtracting the 28 MWe of internal electrical loads from the Unit 1 heat balance value of 864.7 MWe yields a net MFO of 836.7 MWe, which is below the 857 MWe value in the PJM study (Attachment 5 of the LAR,Section V.1.D.ii). The exciter has the capability to support main generator operation within its restricted operational rating and within the capability curve for leading and lagging power factors. Therefore, the increase from the MUR PU remains below the main generator maximum capability and the MFO for Unit 1 is still bounded by the PJM studies.

Based on its review, the staff concludes that the new operating point of the generator is within the generator capability curve and is below the exciter and the main generator maximum capability.

Therefore, the staff finds that the generator is capable of safe operation at uprated conditions.

- 47 For Surry 2 the generator is also rated at 1055 MVA. Currently, the generator output is 850.7 MWe. At uprated conditions, the output will be 865.6 MWe while being capable of exporting 500 MVAR and importing 430 MVAR. Subtracting the 28 MWe of internal electrical loads from the Unit 2 heat balance value of 865.6 MWe yields a net MFO of 837.6 MWe, which is below the 857 MWe value in the P~'M study (Attachment 5 of the LAR,Section V.1.D.ii). The exciter has the capability to support main generator operation within its restricted operational rating and within the capability curve for leading and lagging power factors. Therefore, the increase from the MUR PU remains below the main generator maximum capability and the MFO for Unit 2 is still bounded by the PJM studies. The Surry 2 generating capacity at uprated conditions is similar to the Surry 1 generating capacity. Therefore, the staff finds that the Surry 2 generator is also capable of safe operation at uprated conditions.

The iso-phase bus is rated for 26,000 amperes (A). The maximum of 1.6-percent increase from the MUR PU will increase the current on the iso-phase bus to 25,214 A for Surry 1 and 25,240 A for Surry 2. Based on this information, the NRC staff finds that the iso-phase bus is capable of safe operation at uprated conditions since the increase from the MUR PU remains below the iso-phase bus rating.

Each main generator feeds electric power through a 22 kV isolated phase bus to a main (step-up) transformer, stepping the generator voltage of 22 kV up to the transmission voltage of 230 kV for Surry 1 and 500 kV for Surry 2. These main (step-up) transformers are rated for 1200 MVA. The licensee stated in the LAR that the uprated loadings of the main transformers are 916 MVA, which is below the rating of the main transformers (1200 MVA). Therefore, the NRC staff finds that the main (step-up) transformers are capable of safe operation at uprated conditions.

Station service transformers, connected to the 22 kV isolated phase bus from each main generator, normally supply power to the auxiliaries of each unit by stepping down the 22 kV to 4.16 kV. During normal operating conditions the six station service transformers (three per unit) power the 4.16 kV switchgear, 480 V load centers and motor control centers. Also, during station startup, shutdown, and hot standby conditions, the 4.16 kV normal switchgear buses are transferred to the reserve station service transformers. The licensee stated that the maximum calculated load for any of the station service transformers is 15.83 MVA, which is within the current rating of the station service transformers (22.4 MVA). Therefore, the NRC staff finds that the station service transformers are capable of safe operation at uprated conditions.

Three reserve station service transformers (connected to the switchyard 500/36.5 kV transformer and two 230/36.5 kV transformers) supply 4.16 kV power for startup, shutdown, and hot standby conditions. In the April 29, 2010 (ADAMS Accession No. ML101200269), letter, the licensee stated that the maximum uprated loading of the reserve station service transformers is 25.8 MVA, which is below the 30 MVA design rating of the reserve station transformers. Therefore, the NRC staff finds that the reserve station service transformers are capable of safe operation at uprated conditions.

3.8.2.3 DC System The Class 1E 125 V direct current (DC) System provides the source of power for operation of switchgear, annunciators, vital bus inverters, and emergency lighting. According to Figure 8.3-1 of the UFSAR, the 125 V DC power source in each unit consists of two independent batteries, four battery rectifier/chargers, and two battery distribution switchboards.

- 48 The licensee stated that the MUR PU does not affect any DC powered indication, control, or protection equipment. The staff reviewed the LAR and UFSAR and confirmed that the MUR PU does not impact DC system loads. Therefore, the staff finds that the analyses for the DC system bound MUR PU conditions.

3.8.2.4 Emergency Diesel Generators (EDG)

The EDG system provides a safety-related source of AC power to sequentially energize and restart loads necessary to shut down the reactor safely, and to maintain the reactor in a safe shutdown condition. The standby emergency AC power source consists of three EDGs. One EDG is dedicated to each unit while the third EDG functions as a backup to either of the units.

There are no changes to the emergency buses loads supported by the EDGs due to the MUR PU stated in the LAR, and thus, the existing accident analyses remain bounding. Hence, the EDG system has adequate capacity and capability to power the safety-related loads at MUR PU conditions.

The NRC staff finds that the analyses for the EDG system bound MUR PU conditions, and the onsite power system will continue to meet the requirements of GDC 24 and 39.

3.8.2.5 Switchyard The switchyard equipment and associated components are classified as nonsafety-related. The switchyard serves three 500 kV lines and elght 230 kV lines. Power output from the plant to the switchyard is supplied at the 230 kV level from Surry 1 and the 500 kV level from Surry 2. The switchyard supplies power to the plant electrical system at the 34.5 kV level.

The licensee stated that the current to the switchyard is bounded by the main transformers' capability. The small increase in plant output does not significantly impact the switchyard equipment. The NRC staff reviewed the analyses for the switchyard system and found that they bound the MUR PU conditions.

3.8.2.6 Grid Stability The grid stability impact of the MUR PU is discussed in the LAR, and the licensee concludes that there is no significant effect on grid stability or reliability. Their assessment, of the impact of the 180 MWe of new generation capacity (MUR plus future turbine replacement), was based on the Pennsylvania, New Jersey, and Maryland Interconnection's (PJM) impact study for load growth and new generation.

Following the review of PJM's impact study, the staff requested additional information, in regard to dual unit trip, consequences of a fault in a Reserve Station Service transformer, and the effect of the main generators' replacement on the grid analyses. In its April 29, 2010 (ADAMS Accession No. ML101200269), response, the licensee stated that a simultaneous trip of both units was not performed since it is not required by the Surry 1 and 2 UFSAR or the North American Electric Reliability Corporation's standards. The impact of the main generator replacement was evaluated under the transmissions' systems analysis, when the generators were replaced several years ago.

- 49 The NRC staff reviewed the licensee's responses, and finds that the MUR PU will allow for continued stable and reliable grid operation.

3.8.2.7 Station Blackout In 10 CFR Section 50.63 it requires, in part, that each light-water cooled nuclear power plant be able to withstand and recover from a loss of all AC power, referred to as a station blackout (SBO).

The Surry 1 and 2 SBO coping duration is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This is based on the licensee's evaluation of the offsite power design characteristics, emergency AC power system configuration, and EDG reliability, in accordance with the evaluation procedure outlined in NUMARC 87-00 (ADAMS Legacy Accession No. 9209020287), and RG 1.155 (ADAMS Legacy Accession No. ML080420185). The offsite power design characteristics include the expected frequency of a grid-related loss of offsite power, the estimated frequency of loss of offsite power from severe and extremely severe weather, and the independence of offsite power.

The licensee stated that the evaluation for SBO included the adequacy of the alternate AC power source, emergency condensate storage tank inventory, Class 1E battery capacity, ventilation, compressed air, and containment isolation. Also, the licensee stated that the alternate AC power source (AC diesel generator) has sufficient capacity to operate systems necessary for coping with an SBO event for the required coping period. The licensee stated that the emergency condensate storage tank inventory is adequate for decay heat removal following an SBO event at uprated conditions. The proposed MUR PU has no affect on Surry 1 and 2 station battery capacity as the MUR PU does not increase DC loads. The ventilation for areas containing SBO equipment, the operation of air operated or containment isolation valves, and the ability to restore compressed air is unaffected by the MUR PU. Based on this information, the staff finds that the MUR PU will have no impact on the Surry 1 and 2 SBO coping duration. Therefore, the NRC staff finds that Surry 1 and 2 will continue to meet the requirements of 10 CFR 50.63 under MUR PU conditions.

3.8.2.8 Environmental Qualification (EQ) Program The licensee stated that the MUR PU does not affect the EQ related non-radiological conditions (temperature, pressure, humidity). The licensee described the adjustments made to the source terms used for the radiation aspects of the EO Program evaluations in order to accommodate MUR PU operation.

The radiation dose qualification is based on the sum of the normal operational dose plus the accident dose. The licensee has evaluated the MUR PU-related increase in the post-accident integrated dose and discussed that the increase in radiation levels will impact the equipment qualification of the following equipment:

Excore Neutron Detectors: The licensee committed (Regulatory Commitment #13) that it will follow its EQ Periodic Maintenance schedule and replace the excore neutron detectors during the fall 2010 Surry 1 outage and the spring 2011 Surry 2 outage. Also, the licensee committed (Regulatory Commitment #12) to incorporate changes in the qualified lifetime of this equipment into EQ Program documentation, prior to operating above the current reactor power. This issue has been included within the scope of a condition to the license as discussed below in Section 4.

- 50 Hydrogen Monitoring Equipment: In the NRC's safety evaluation dated March 22, 2005 (ADAMS Accession No. ML050840168), the NRC staff determined that the hydrogen monitoring equipment no longer met the definition of safety-related component as defined in 10 CFR 50.2.

This equipment will be removed by the licensee from the Surry 1 and 2 EO Program.

The NRC staff finds that the licensee's evaluation of the proposed MUR PU on the EO of electrical equipment has adequately addressed the effects of the proposed MUR PU on the environmental conditions for the qualification of electrical equipment. Therefore, the NRC staff finds that with the exception of the excore neutron detectors, the MUR PU will have no adverse impact on the EO Program and the licensee will continue to meet the requirements of 10 CFR 50.49.

3.8.3 Electrical Systems License Conditions As part of the MUR PU amendment, the following items will be included as conditions in the Renewed Facility Operating Licenses for Surry 1 and 2, as follows:

VEPCO shall complete actions prior to operating above 2546 MWt, to determine EO service life for excore neutron detectors.

Prior to operating above 2546 MWt, VEPCO shall replace excore neutron detectors that have been determined to be at end-of-service life.

3.8.4 Conclusion Based on the technical evaluation provided above, the NRC staff finds that Surry 1 and 2 will continue to meet GDC Criteria 24 and 39, 10 CFR 50.63, and 10 CFR 50.49. Therefore, the NRC staff finds the MUR PU acceptable.

3.9 Instrumentation and Controls 3.9.1 RegUlatory Evaluation Nuclear power plants are licensed to operate at a specified core thermal power. Appendix K, "ECCS Evaluation Models," to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," requires loss-of-coolant accident and ECCS analyses to assume "that the reactor has been operating continuously at a power level at least 1.02 times the licensed thermal power level (to allow for instrumentation error)." Alternately, Appendix K allows such analyses to assume a value lower than the specified 102 percent, but not less than the licensed thermal power level, "provided the proposed alternative value has been demonstrated to account for uncertainties due to power level instrumentation error." This allowance gives licensees the option to justify a MUR PU with reduced margin between the licensed power level and the power level assumed in the ECCS analysis by using more accurate instrumentation to calculate the reactor thermal power.

Because the maximum power level of a nuclear plant is a licensed limit, the NRC must review and approve a proposal to raise the licensed power level under the license amendment process. The LAR should include a justification for the reduced power measurement uncertainty to support the proposed MUR PU.

3.9.2

- 51 Approved Topical Report ER-80P and its supplement, Topical Report ER-157P, describe the Cameron LEFM CheckPlus System for the measurement of FW flow and provide a basis for the proposed 1.6-percent MUR PU of the licensed reactor thermal power. The NRC staff also considered guidance of RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," in its review of the licensee's submittals for the proposed MUR PU request.

Cameron LEFM Technology and Measurement In the LAR, the licensee stated that both the Cameron LEFM Check and Cameron LEFM CheckPlus Systems use transit time methodology to measure fluid velocity. The basis of the transit time methodology for measuring fluid velocity and temperature is that ultrasonic pulses transmitted through a fluid stream travel faster in the direction of the fluid flow than opposite the flow. The difference in the upstream and downstream traversing times of the ultrasonic pulse is proportional to the fluid velocity in the pipe, and the temperature is determined using a pre-established correlation between the mean propagation velocity of the ultrasound pulses in the fluid and the fluid pressure.

Both systems use multiple diagonal acoustic paths instead of a single diagonal path, allowing velocities measured along each path to be numerically integrated over the pipe cross-section to determine the average fluid velocity in the pipe. This fluid velocity is multiplied by a velocity profile correction factor, the pipe cross-section area, and the fluid density to determine the FW mass flow rate in the piping. The mean fluid density may be obtained using the measured pressure and the derived mean fluid temperature as an input to a table of thermodynamic properties of water. The velocity profile correction factor is derived from calibration testing of the Cameron LEFM Check or CheckPlus System in a plant-specific piping model at a calibration laboratory.

The Cameron LEFM Check System consists of a spool piece with eight transducers, two on each of the four acoustic paths in a single plane of the spool piece. The velocity measured by anyone of the four acoustic paths is the vector sum of the axial and the transverse components of fluid velocity as projected onto the path. The Cameron LEFM CheckPlus System uses 16 transducers, 8 each in two orthogonal planes of the spool piece. In the Cameron LEFM CheckPlus System, when the fluid velocity measured by an acoustic path in one plane is averaged with the fluid velocity measured by its companion path in the second plane, the transverse components of the two velocities are canceled and the result reflects only the axial velocity of the fluid. This makes the numerical integration of four pairs of averaged axial velocities and computation of volumetric flow inherently more accurate than a result obtained using four acoustic paths in a single plane.

Also, because there are twice as many acoustic paths and there are two independent clocks to measure the transit times, errors associated with uncertainties in path length and transit time measurements are reduced.

The NRC staff's review in the area of instrumentation and control covers the proposed plant-specific implementation of the FW flow measurement technique and the power increase gained as a result of implementing this technique, in accordance with the guidelines (A through H) in Section I of Attachment 5 of the LAR, RIS 2002-03, RI. The NRC staff conducted its review to confirm that the licensee's implementation of the proposed FW flow measurement device is consistent with staff-approved Topical Reports ER 80P and ER 157P and that the licensee adequately addressed the four additional requirements listed in the staff's SE. The NRC staff also reviewed the power measurement uncertainty calculations to ensure that (1) the conservatively

3.9.3

- 52 proposed uncertainty value of 0.35 percent correctly accounts for all uncertainties associated with power level instrumentation errors, and (2) the uncertainty calculations meet the relevant requirements of Appendix K to 10 CFR Part 50, as described in Section 2 of this SE.

The licensee provided the following information about the Cameron LEFM CheckPlus System FW flow measurement technique and its implementation at Surry 1 and 2.

The Cameron LEFM CheckPlus System consists of an electronic cabinet installed in the cable spreading room and measurement spool pieces installed in each of the three main FW flow lines between the existing FW venturi flow meters and the main FW check valves. The spool pieces are located downstream of the existing venturi flow meters to avoid affecting their performance.

Each measurement section consists of 16 ultrasonic, multipath, transit time transducers, one dual RTD, and two pressure transmitters.

The electronic cabinet controls magnitude, sequences transducer operations, makes time measurements, and calculates vOlumetric flow, temperature, and mass flow. The system software measures velocities at precise locations. The FW mass flow rate and temperature are transmitted to the plant process computer for use in calorimetric measurement of reactor thermal output.

In the event of system failure, including failure of a transducer, failure of one plane of operation, loss of AC power, or a CPU-related malfunction, the control room operators are alerted via an annunciator window for FW UFM failure. An internal high temperature in the electrical cabinet will also result in an alert to operators.

The UFM values for FW mass flow and temperature will be directly substituted for the existing venturi-based flow and RTD temperature inputs used in the plant calorimetric measurement calculations. The existing venturi-based FW flow and RTD temperature will continue to be used for other plant functions and may be used for plant calorimetric calculations in the event of a UFM failure.

LAR Compliance to Guidance A through H, Section 1, Attachment 5 of the LAR Items A through C Items A, 8, and C gUide licensees to identify the approved topical reports, provide references to the NRC's approval of the measurement technique, and discuss the plant-specific implementation of the guidelines in the topical report and the NRC staff's approval of the FW flow measurement technique.

The licensee identified approved Topical Reports ER-80P, Revision 0, and ER-157P, Revision 5, as applicable to the Cameron LEFM CheckPlus System. The licensee also referenced NRC SEs for Topical Reports ER-80P, dated March 8, 1999, and ER-157P, dated December 20, 2001.

The licensee also cited Evaluation of The Hydraulic Aspects of The Caldon Leading Edge Flow Measurement Check And Checkplus Ultrasonic Flow Meters, an NRC SE, (ADAMS Accession No. ML061700222) dated July 5, 2006, which reexamined the performance of the Cameron systems and confirmed the validity of the previously referenced Cameron topical reports.

- 53 The Cameron LEFM CheckPlus System was permanently installed in Surry 1 during the spring of 2009. The Cameron LEFM CheckPlus System for Surry 2 was installed during the spring of 2009 and in Surry 2 during the fall of 2009.

On the basis of its review of the licensee's submittals, as reflected in the above discussion, the NRC staff finds that the licensee has sufficiently addressed the plant-specific implementation of the Cameron LEFM CheckPlus System using topical report guidelines. Therefore, the NRC staff concludes that the licensee's description of the FW flow measurement technique and implementation of the MUR PU using this technique follows the guidance in Items A through C of Section I of Attachment 5 of the LAR.

Item D Item D in Section I of RIS 2002-03, guides licensees to address four criteria when implementing the FW flow measurement uncertainty technique. The NRC staff SEs on approved Topical Reports ER-80P and ER-157P both include these four plant-specific criteria to be addressed by a licensee referencing these topical reports for an MUR PU. The licensee's submittal addresses each of the four criteria as follows:

(1)

The licensee should discuss the maintenance and calibration procedures that will be implemented with the incorporation of the LEFM. These procedures should include processes and contingencies for an inoperable LEFM and the effect on thermal power measurement and plant operation.

Licensee's Response:

Implementation of the PU license amendment will include developing the necessary procedures and documents required for operation, maintenance, calibration, testing, and training at the uprated power level with the new LEFM CheckPlus System. Work will be performed at Surry 1 and 2 by site instrumentation and control personnel qualified by the Surry 1 and 2 Instrumentation and Control Training Program.

In addition, the licensee stated that a preventive maintenance program for the Cameron system will be developed using the vendor's maintenance and troubleshooting manual. Preventive maintenance activities for the Cameron LEFM CheckPlus System at Surry 1 and 2 will include the following:

general terminal and cleanliness inspection power supply inspection central processing unit inspection acoustic processor unit checks alarm relay checks watchdog timer checks to ensure that the software is running communication checks transducer checks calibration checks on each FW pressure transmitter

- 54 On the basis of its review of the licensee's submittal, the NRC staff concludes that the licensee has adequately addressed Criterion 1.

(2)

For plants that currently have LEFMs installed, provide an evaluation of the operational and maintenance history of the installed instrumentation and confirmation that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions set forth in Topical Report ER-80P.

Licensee's Response:

The Cameron LEFM CheckPlus System was installed in Surry 1 during the spring of 2009. The Cameron LEFM CheckPlus System for Surry 2 was installed during the fall of 2009. Commissioning and calibration are expected to be completed in the spring of 2010. The licensee indicated that no maintenance activities have taken place on either unit's system during that time.

Given the very short time since the installation of the systems, the NRC staff finds the licensee's response adequate to address Criterion 2.

(3)

The licensee should confirm that the methodology used to calculate the uncertainty of the LEFM in comparison to the current FW instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty). If the licensee uses an alternative approach, its application should be justified and applied to both venturi and ultrasonic flow measurement instrumentation installations for comparison.

Licensee's Response:

The licensee uses a core thermal power uncertainty calculation consistent with ISA-RP67.04.02-2000. The combination of errors within instrument loops is performed using VEPCO Technical Report EE-0116. The methodology used statistically combines inputs to determine the overall uncertainty. Dependent parameters are arithmetically combined to using the square root of the sum-of-the-squares approach. This methodology is consistent with the vendor determination of the Cameron LEFM CheckPlus System uncertainty, as described in the referenced topical reports, and is consistent with the guidelines in RG 1.105, "Setpoints for Safety-Related Instrumentation" (ADAMS Accession No. ML082120659).

Based on the foregoing, the NRC staff concludes that the licensee has adequately addressed Criterion 3.

(4)

For plant installation where the ultrasonic meter (including Cameron LEFM) was not installed with flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors are not representative of the plant-specific installation), licensees should provide additional justification for its use. The justification should show that the meter installation is either independent of the plant-specific flow profile for the stated accuracy, or that the installation can be shown to be eqUivalent to known calibrations and plant configurations for the specific installation, including the propagation of flow profile

- 55 effects at higher Reynolds numbers. Additionally, for previously installed calibrated elements, licensees should confirm that the piping configuration remains bounding for the original Cameron LEFM installation and calibration assumptions.

Licensee's Response:

The Cameron LEFM CheckPlus System was calibrated using a site-specific piping configuration at Alden Research Laboratories. As part of the LAR, the licensee submitted ER-684: (ADAMS Accession No. ML100321414, Non-Public),

Revision 2, "LEFM CheckPlus Meter Factor Calibration and Accuracy Assessment for Surry 1," and ER-690: (ADAMS Accession No. ML100321415, Non-Public),

Revision 2, "LEFM CheckPlus Meter Factor Calibration and Accuracy Assessment for Surry 2," both issued September 2009.

According to ER-684, the meter factor uncertainty for Surry 1 is 0.20 percent.

ER-690 gives the meter factor uncertainty for Surry 2 as 0.19 percent.

The licensee stated that acceptance of the final site-specific uncertainty analyses will occur following completion of the commissioning process. The commissioning process will verify bounding calibration test data and confirm that actual field performance meets the uncertainty bounds established for the instruments. Final commissioning is expected to be completed in the spring of 2010.

On the basis of the foregoing, the NRC staff concludes that the licensee has adequately addressed Criterion 4. In addition, the licensee has committed to confirming that the in-situ test data are bounded by the calibration test data after final commissioning of the Cameron LEFM CheckPlus System.

Item E Item E guides licensees in the submittal of a plant-specific total power measurement uncertainty calculation, explicitly identifying all parameters and their individual contributions to the power uncertainty.

To address Item E of RIS 2002-03, the licensee provided Cameron engineering reports ER-650 (ADAMS Accession No. ML100321412, Non-Public), Revision 2, and ER-651 (ADAMS Accession No. ML100321413, Non-Public), Revision 2. These reports provide calculations that demonstrate that the total thermal power uncertainty for each unit is 0.35 percent.

The NRC staff reviewed the calculations and determined that the licensee identified all of the parameters associated with the thermal power measurement uncertainty, provided individual measurement uncertainties (including those discussed in Item D(4) above), and calculated the overall thermal power uncertainty.

The licensee's calculations arithmetically summed uncertainties for parameters that are not statistically independent and statistically combined with other parameters. The licensee combined random uncertainties using the square root sum-of-the-squares approach and added systematic biases to the result to determine the overall uncertainty. This methodology is consistent with the vendor determination of the Cameron LEFM CheckPlus System uncertainty,

- 56 as described in the referenced topical reports, and is consistent with the guidelines in RG 1.105, "Setpoints for Safety-Related Instrumentation." (ADAMS Accession No. ML082120659)

The NRC staff finds that the licensee has provided calculations of the total power measurement uncertainty at the plant, explicitly identifying all parameters and their individual contribution to the power uncertainty. The NRC staff concludes that the licensee has adequately addressed the guidance in Item E of Section I of Attachment 5 of the LAR, RIS 2002-03, RI.

Item F Item F guides licensees to provide information to address the specified aspects of the calibration and maintenance procedures related to all instruments that affect the power calorimetric.

In the LAR, the licensee addressed each of the five aspects of the calibration and maintenance procedures listed in Item F of RIS 2002-03, RI:

(1)

Maintaining Calibration The licensee stated that calibration will use procedures based on the appropriate LEFM CheckPlus technical manuals. Other calorimetric process instrumentation and computer points are maintained and calibrated using approved procedures. The response to Item D(4) above addresses the preventive maintenance program.

(2)

Controlling Hardware and Software Configuration The Cameron LEFM CheckPlus System is designed and manufactured in accordance with the vendor's quality assurance program, which meets the requirements of Appendix 8, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

The licensee committed to maintaining, after installation, the software and hardware configuration by using existing procedures and processes, which include verification and validation of software configuration changes. The UFM hardware and the calorimetric process instrumentation will be maintained per the Surry 1 and 2 configuration control processes.

(3)

Performing Corrective Actions Surry 1 and 2 personnel will monitor plant instrumentation that affects the power calorimetric input, including UFM inputs. Any problems detected will be handled according to the Surry 1 and 2 corrective action process.

(4)

Reporting Deficiencies to the Manufacturer The licensee states that any conditions found to be adverse to the quality of the LEFM CheckPlus System will be documented and reported to the vendor, as needed, to support corrective action.

(5)

Receiving and Addressing Manufacturer Deficiency Reports The licensee stated that it has existing processes to address the receipt of manufacturer's deficiency reports. Any such deficiencies will be documented and controlled by existing processes.

- 57 On the basis of its review of the above statements, the NRC staff finds that the licensee has addressed the calibration and maintenance aspects of the Cameron LEFM CheckPlus System and all other instruments affecting the power calorimetric. Therefore, the NRC staff concludes that the licensee meets the guidance in Item F of Section I of Attachment 5 of the LAR, RIS 2002-03, RI.

Items G and H Items G and H in Section I, guide licensees to provide a proposed allowed outage time (AOT) for the instrument and to propose actions to reduce power if the AOT is exceeded.

The licensee proposed a 48-hour AOT for operating above 2,546 MWt (i.e., the current licensed thermal power limit) if the UFM becomes non-operational. The licensee specifically noted that any failure of the UFM, including failures of a single path or plane, will be treated as a complete failure of the UFM system and thus start the plant's allowable 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to remain above 2,546 MWt. In addition, the ability of the plant to stay above 2,546 MWt is contingent upon its ability to maintain steady-state conditions. If the plant should reduce power below 2,546 MWt during the AOT, the plant will not be permitted to return above that value until the UFM function is fully restored.

During the 48-hour AOT in the event of a UFM failure, the plant would use the existing FW venturis and RTDs for the calorimetric calculation. Because the FW venturis are regularly normalized to the UFM measurements, their measurements should be equivalent to the UFM over the 48-hour AOT. Use of calibrated FW venturis to remain at the uprated power level during the 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> that follow a UFM being declared nonoperational is consistent with NRC's approval of previous MUR PU submittals. Venturi nozzle fouling and transmitter drift were considered as potential sources of error within the AOT window. The FW venturis for each unit are regularly inspected. Procedure 0-MPM-1010-01 documents the venturi inspection and cleaning activities.

As per the CABRI Water Loop extension letter (ADAMS Accession No. ML100200269, Non-Public), the most recent inspection of each venturi in both units revealed all of the FW venturis to be at least in satisfactory condition. In addition, monthly trending of FW flow for the FW loops in the Surry units has not revealed any adverse trends in FW venturi performance. Based upon the results of the cited inspection activities and performance monitoring, the likelihood of a FW venturi fouling event that appreciably affects measurements within any given 48-hour period is low. The LAR referred to an FW flow transmitter drift study using data from Surry 1 and 2 measurements of transmitter output voltages. The collected data indicate that, even under maximum observed drift rates, the error associated with transmitter drift over a 48-hour period would be negligible. The licensee's submittal noted that the transmitter drift analyses are documented in an internal engineering document (ET-NAF-09-0013).

A loss of the plant computer is treated as a loss of the UFM and of the ability to obtain corrected calorimetric power using alternate instrumentation. Operation at the uprated power may continue until the time, not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, when the next required nuclear instrument heat balance is to be performed.

On the basis of its review of the licensee's submittals, the staff finds that the licensee has provided sufficient justification for the proposed AOT and the proposed actions to reduce power level if the AOT is exceeded. Therefore, the NRC staff concludes that the licensee has followed the guidance in Items G and H of Section I of Attachment 5 of the LAR to RIS 2002-03, RI.

- 58 3.9.4 Conclusion The NRC staff reviewed the licensee's proposed plant-specific implementation of the feedwater flow measurement device and the power uncertainty calculations and determined that the licensee's proposed license amendment is consistent with the staff-approved Topical Report ER-80P and its supplement, Topical Report ER-157P. The NRC staff has also determined that the licensee adequately accounted for all instrumentation uncertainties in the reactor thermal power measurement uncertainty calculations and demonstrated that the calculations meet the relevant requirements of 10 CFR Part 50, Appendix K, as described in Section 3.9.1 of this SE.

The licensee has committed to verify bounding calibration test data and confirm that actual field performance meets the uncertainty bounds established for the instruments (per Item 0(4) above).

Therefore, the NRC staff finds the instrumentation and control aspects of the proposed 1.6-percent thermal power uprate to be acceptable.

3.10 Plant Systems 3.10.1 Regulatory Evaluation The NRC staff's review in the area of plant systems covers the impact of the proposed MUR PU on the NSSS interface systems, containment systems, safety-related cooling water systems, spent fuel pool (SFP) storage and cooling, radioactive waste systems, and engineered safety feature (ESF) heating, ventilation and air conditioning ( HVAC) systems. The staff's review is based on the guidance in SRP Chapters 3, 6, 9, 10, and 11, and Sections II, III, and VI, of the LAR, RIS 2002-03, RI.

The NRC staff reviewed the licensee's LAR for compliance with the following regulations.

Containment design, GOC 16, requires that the containment shall provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment. Control room, GOC 19, requires that the control room must provide the operators with the capability to operate the nuclear power units safely under normal conditions and maintain the reactor in a safe condition under accident conditions including a LOCA. Containment heat removal, GOC 38, requires that the containment heat removal systems are capable of rapidly reducing the containment temperature and pressure following a LOCA and maintaining them at an acceptably low level.

Containment design basis, GOC 50, requires that the containment accommodate the pressure and temperature conditions resulting from a LOCA without exceeding the design leakage rate.

Control of releases of radioactive materials to the environment, GOC 60, requires that the nuclear power unit have means to control the release of radioactive materials in gaseous and liquid effluents during normal operation and anticipated operational occurrences.

3.10.2 Technical Evaluation 3.10.2.1 Containment Systems The Surry 1 and 2 containment subcompartment structures include the pressurizer cubicle, the steam generator cubicle and the reactor vessel cavity. Surry 1 and 2 UFSAR, Section 15.6.3, describes the analysis of containment subcompartments. Changes to the reactor coolant system temperatures due to the MUR PU affect the coolant density which in turn affects the energy releases for the LOCA. The licensee stated in Attachment 5 to the LAR that the MUR PU

- 59 temperatures were reviewed and confirmed to be bounding by the existing analyses of subcompartment structures. Since the containment subcompartment analyses remain bounding for the MUR PU, the NRC staff finds the subcompartment analyses acceptable.

The NRC staff has previously approved the mass and energy release and containment analyses for LOCA analyses. These are contained in References 11-2 and 11-3 of Attachment 5 to the licensee's LAR. The licensee states in Attachment 5 of the LAR that subsequent to the NRC approval in Reference 11-3 the licensee performed a long-term LOCA mass and energy release re-analysis assuming 102 percent of 2546 MWt in compliance with 10 CFR 50.59 and the mass and energy releases remain bounding for the MUR PU.

The LOCA mass and energy release calculations for the blowdown, reflood and post-reflood phases assumed a core power of 102 percent of 2546 MWt which bounds the MUR PU and does not change the power including uncertainty of the current analysis.

The licensee states and provides a justification that the MUR PU has no effect on the containment analysis other than the mass and energy input and therefore the containment analyses remain bounding. The NRC staff 'finds this acceptable.

The MSLB mass and energy releases are described in the LAR, Attachment 5, Reference 11-2.

The NRC approval is cited in Reference 11-3. The analyses assumed a core power of 116.1 percent of 2546 MWt and therefore bound the proposed MUR PU. (The higher power reflects the fact that the analyses were done for the comparable North Anna Power Station, which operates at a higher rated thermal power. The NRC staff agreed in an SE dated October 12, 2006 (ADAMS Accession No. ML063000024), that this analysis was applicable to Surry 1 and 2.)

3.10.2.2 Heating, Ventilation and Air Conditioning The licensee addressed the effect of the MUR PU on the following nonsafety-related systems:

containment air cooling, control rod drive mechanism ventilation, and the following safety-related systems: control room ventilation system, ESF ventilation system and the fuel-handling area ventilation system.

The licensee states that the MUR PU will have no significant effect on the containment atmosphere.

The important component of the control rod drive mechanism with respect to heat load is the lift coil. The licensee states that the increased temperature due to the MUR PU for the control rod drive mechanism design basis case is less than the coil design temperature.

The licensee evaluated the heat loads due to the MUR PU on the control room ventilation system.

The heat loads covered by this system do not increase due to the MUR PU.

The licensee evaluated the heat loads covered by the ESF ventilation system and found that there is no significant increase.

The fuel-handling area heat load is not impacted by the MUR PU.

- 60 Summarizing, the licensee has evaluated the HVAC Surry 1 and 2 systems and found that the heat load either does not increase or that the increase is not significant. The licensee's analysis indicated that the MUR PU condition produced no significant heat load and is bounded by current design. Therefore, the NRC staff finds the MUR PU acceptable with respect to the Surry 1 and 2 HVAC systems.

3.10.2.3 Radioactive Waste Systems The waste processing systems provide the means to sample, collect, process, store/hold, re-use, and/or release gaseous and liquid low-level effluents. These waste systems and their various subsystems and components were evaluated for the MUR PU. Systems are common to both units and are sized to treat the radioactive solid waste produced during simultaneous operation of both units. System functions and the waste processed volume are unaffected by the uprate. The radioactive waste systems are bounded by the existing system design parameters and are acceptable at MUR PU conditions. The NRC staff reviewed the licensee's assessment. The NRC staff finds that a 1.6-percent increase in power is not likely to result in a significant change to the operation of the radioactive waste systems. Therefore, the NRC staff finds that the radioactive waste systems will function adequately for the MUR PU.

3.10.2.4 Safety-Related Cooling Water Systems The safety-related cooling water systems include the SW system, and the component cooling water (CCW) system, ultimate heat sink, and residual heat removal (RHR) system.

The CCW system is a closed-loop piping system shared between Surry 1 and 2, and rejects heat to the SW system. Th~re are four circulation water (CW) pumps per unit. Each CW pump takes suction from the James River and discharges it into the CW intake canal. The SW flow is provided by gravity feed through valves located upstream of each unit's condenser inlet. The SW system is designed to support a LOCA in one unit, while placing the non-accident unit in a cold shutdown condition in the event of a coincident loss of offsite power. During an accident condition, one or two (depending on the scenario) of three SW pumps are required to provide adequate inventory for heat removal for both units.

Each component cooled by the SW system was evaluated to confirm that the existing flow rate is sufficient to satisfy the MUR PU heat removal requirements during normal power operation, accident, and cooldown conditions. The evaluations determined that the existing SW flows will continue to support the heat removal requirements at uprate conditions. The SW system and component design parameters remain bounding for MUR PU operation. Therefore, the SW system is acceptable for operation at MUR PU conditions.

The ultimate heat sink is comprised of the James River, the CW intake canal, and the discharge canal. The licensee states that the SW system inlet temperature for normal, cooldown, and DBA conditions is bounded for the MUR PU. The ultimate heat sink is capable of cooling the SW system to prevent SW temperature from exceeding the inlet temperature limits during operating conditions. It is further stated that no system modifications are required to support the MUR PU.

Surry 1 and 2 UFSAR, Section 9.3, describes the RHR system. The RHR cooldown performance was analyzed under MUR PU conditions. The normal two-train cooldown, one-train cooldown,

- 61 and accident-case cooldown were analyzed. The analysis showed that each of these cases met the cooldown time requirements.

The NRC staff reviewed the licensee's evaluation of safety-related cooling water systems. Based upon the licensee's determination that the existing analyses for these systems were evaluated for 102-percent rated thermal power, and are bounded for the MUR PU, the staff finds there is reasonable assurance that the systems are acceptable for the MUR PU.

3.10.2.5 Spent Fuel Pool Cooling (SFPC) and Purification System The licensee stated there are no changes to the SFPC system limiting temperatures, pressures or flow rates as a result of the MUR PU. Uprate conditions are bounded by the existing system design conditions. The principal function of the SFPC system is to provide storage and cooling of the spent fuel. The primary impact of an MUR PU would be to the decay heat of the fuel recently discharged from the core. System modifications are not required to support the MUR PU. The limiting case heat loads at uprate conditions remain bounded by the existing analysis. There is no change to the loss of cooling analysis. The uprate is not expected to have any significant impact on the SFP refueling purification or cooling functions. Therefore, the SFP cooling and purification system is acceptable at the MUR PU conditions. The NRC staff does not expect that the MUR PU will result in a significant change to the operation of the SFPC system. Therefore, the NRC staff finds that the SFPC system will not be impacted by the MUR PU.

3.10.3 Conclusion The effect of the MUR PU on containment safety analyses is either bounded by the current containment safety analyses or, in the case of the short-term LOCA, the increase in mass flux is insignificant. Therefore, Surry 1 and 2 remain in compliance with GDCs 16,38 and 50. The increase of heat loads due to the MUR PU in the containment, control room and on the ESF ventilation systems is insignificant. Therefore, Surry 1 and 2 remain in compliance with GDCs 19 and 60. In summary, the licensee reviewed the design and operation of the plant systems and determined that the proposed MUR PU does not adversely impact any of the systems. The NRC staff finds the plant systems acceptable for the MUR PU.

3.11 Changes to Renewed Facility Operating License and TSs 3.11.1 Regulatory Evaluation This LAR revises the licensed power level identified in Condition 3.T of the Renewed Facility Operating Licenses and the TS definition of reactor thermal power.

Licensees may revise the TS content provided that plant-specific review supports a finding of continued adequate safety because: (1) the change is editorial, administrative or provides clarification (Le., no requirements are materially altered), (2) the change is more restrictive than the licensee's current requirement, or (3) the change is less restrictive than the licensee's current requirement, but nonetheless still affords adequate assurance of safety when judged against current regulatory standards. The detailed application of this general framework, and additional specialized guidance, are discussed in this section in the context of the specific proposed changes.

- 62 3.11.2 Technical Evaluation The submittal includes TS requirements that would demonstrate compliance with 10 CFR 50.36, "Technical Specifications," for plant operating conditions related to the requested authorization for a power level increase. The plant modifications will improve the accuracy of the plant power calorimetric measurement based on the Cameron LEFM Checkplus ' System (ultrasonic flow meter) instrumentation.

3.11.3 Renewed Facility Operating License and Definitions - Rated Thermal Power The licensee proposed to revise paragraph 3.T of the Renewed Facility Operating License and TS 1.0-1, Definitions-rated power (RP) to reflect the authorized power level increase. The TS RP will limit the maximum reactor core heat transfer rate to the reactor coolant to 2587 MWt. The NRC staff finds that this change meets 10 CFR 50.36 and is acceptable because the TS limit for operation is derived from the analyses and evaluation included in the Safety Analysis Report as accepted by the SE for the requested power level increase discussed herein.

4.0 LICENSEE'S COMMITMENTS The following table identifies regulatory commitments made by the licensee in the application dated January 27, 2010, and its supplements dated February 4 and April 29, 2010, and evaluated in this SE. This table has been added to licenses DPR-32 and DPR-37 as License Condition 3.T.

COMMITMENT SCHEDULED COMPLETION DATE

1. VEPCO will perform the final acceptance of the Surry 1 and 2 uncertainty analysis to ensure the results are bounded by the statements contained in the LAR (Attachment 5, Section 1.1.D.4.1).

Prior to operating above 2546 MWt (98.4% RP).

2. Technical Requirements Manual (TRM) will be revised to include UFM administrative controls (Attachment 1 Section 3.0).

Prior to operating above 2546 MWt (98.4% RP).

3. Revise procedures, programs, and documents for the new UFM (including transducer replacement) (Attachment 5, Sections 1.1, 1.1.D.1.1, 1.1.H, V11.1, VI1.2.A, and VII.4).

Prior to operating above 2546 MWt (98.4% RP).

4. Appropriate personnel will receive training on the UFM and affected procedures (Attachment 5, Sections 1.1.D.1.1, VII.2.A, VI1.2.D, and VII.3).

Prior to operating above 2546 MWt (98.4% RP).

- 63 COMMITMENT SCHEDULED COMPLETION DATE

5. The FAC CHECWORKS SFA models will be updated to reflect the MUR PU conditions (Attachment 5,Section IV.1.E.iii).

Prior to operating above 2546 MWt (98.4% RP)

6. Simulator changes and validation will be completed (Attachment 5, Section VI1.2.C).

Prior to operating above 2546 MWt (98.4% RP).

7. Revise existing plant operating procedures related to temporary operation above full steady-state licensed power levels (Attachment 5,Section VII.4).

Prior to operating above 2546 MWt (98.4% RP).

8. Process UFSAR changes in accordance with 10 CFR 50.59 (Attachment 1, Section 3.0).

In accordance with 10 CFR 50.71(e).

9. UFM commissioning and calibration will be completed (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

10. Confirm flow normalization factors (Attachment 5, Section 1.1.G).

Prior to operating above 2546 MWt (98.4% RP).

11. Rescaling and calibration of main turbine first stage pressure input to AMSAC (Attachment 5, Sections 11.2.28, VII.2.B, V1I1.2, and VII1.3).

Prior to operating above 2546 MWt (98.4% RP).

12. Determine EO service life for excore detectors (Attachment 5, Sections 111.2.A and V.1.C).

Prior to operating above 2546 MWt (98.4% RP).

13. The excore neutron detectors are scheduled to be replaced (Attachment 5,Section V.I.C).

Unit 1: Fall 2010 Refueling Outage.

Unit 2: Spring 2011 Refueling Outage.

14. Revise EOP setpoints (Attachment 5, Section VI1.2.A).

Prior to operating above 2546 MWt (98.4% RP).

15. The UFM feedwater flow and temperature data will be compared to the feedwater flow venturis output and the feedwater RTD output (Attachment 5, Section 1.1.D.2.1).

Prior to operating above 2546 MWt (98.4% RP).

- 64 SCHEDULED COMPLETION DATE COMMITMENT

16. For the applicable UFSAR Chapter 14 events, Prior to operating above Surry 1 and 2 will re-analyze the transient 2546 MWt (98.4% RP).

consistent 2546 MWt (98.4% RP) with VEPCO's NRC-approved reload design methodology in VEP-FRD-42, Rev. 2.1-A.

If NRC review is deemed necessary pursuant to the requirements of 10 CFR 50.59, the accident analyses will be submitted to the NRC for review prior to operation at the uprate power level. These commitments apply to the folloWing Surry 1 and 2 UFSAR Chapter 14 DNBR analyses that were analyzed at 2546 MWt consistent with the Statistical DNBR Evaluation Methodology in VEP-NE-2-A:

• Section 14.2.7 - Excessive Heat Removal due to Feedwater System Malfunctions (Full Power Feedwater Temperature Reduction case only);

• Section 14.2.8 - Excessive Load Increase Incident;

• Section 14.2.9 - Loss of Reactor Coolant Flow; and

• Section 14.2.10 - Loss of External Electrical Load

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Virginia State official was notified of the proposed issuance of the amendments. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendments change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (75 FR 17447). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendments.

7.0

- 65 CONCLUSION The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: T. Mossman S. Basturescu C. Fairbanks A. Obodoako W.Jessup J. Hung N. Iqbal R. Lobel A. Boatright D. Woodyatt B. Parks G. Lapinsky Y. Huang Date: September 24, 2010

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IAUlses*

IAMendiola GCasto MO'Brien DATE OS/26/10 06/21/10 06/11/10 08/03/10 08/06/10 7/28/10, 8/18/10, 9/21/10 w/ changes SRohrer for 9/7/10)

PFFICE OGC LPL2-1BC DORL/D NAME MDreher GKulesa JGltter DATE 08/12/10 9/24/10 09'24/10