ML11110A111

Issuance of Amendments Regarding Reactor Vessel Heatup and Cooldown Curves for 48 Effective Full-Power Years

ML11110A111
Person / Time
Site:	Surry
Issue date:	05/31/2011
From:	Cotton K Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO)
Cotton K
References
TAC ME3920, TAC ME3921
Download: ML11110A111 (19)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 31,2011 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 SUB..IECT:

SURRY POWER STATION, UNIT NOS. 1 AND 2 -ISSUANCE OF AMENDMENTS REGARDING REACTOR VESSEL HEATUP AND COOLDOWN CURVES FOR 48 EFFECTIVE FULL-POWER YEARS (TAC NOS. ME3920AND ME3921)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 274 to Renewed Facility Operating License No. DPR-32 and Amendment No. 274 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2 (Surry 1 and 2), respectively. The amendments change the Technical Specifications in response to your application dated May 6, 2010.

These amendments revise the pressure and temperature limit curves to provide new limits that are valid to 48 effective full-power years for Surry 1 and 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,

~ (2t;;J 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. 274 to DPR-32

2. Amendment No. 274 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. 274 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) dated May 6,2010, 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 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.

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2.

Accordingly, the license is 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. 274

, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION C(~

Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. DPR-32 and the Technical Specifications Date of Issuance: ~1ay 31, 2011

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. 274 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) dated May 6, 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 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.

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2.

Accordingly, the license is 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. 274

, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION G lC---,_~-

Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

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

May 31, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 274 RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND TO LICENSE AMENDMENT NO. 274 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 License License License No. DPR-32, page 3 License No. DPR-32, page 3 License No. DPR-37, page 3 License No. DPR-37, page 3 TSs TSs TS 3.1-9 TS 3.1-9 TS 3.1-11 TS 3.1-11 Fig 3.1-1 Fig 3.1-1 Fig 3.1-2 Fig 3.1-2

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3. This renewed license shall be deemed to contain and is subject 10 the conditions specified in the following Commission regulations: 10 CFR Part 20, Section 30:34 of 10 CFR Part 30, SectiOn 40.41 01'0 CFR Part 40, Sectbns 50.54 and 50.59 01 10 CFR Pan SO, and Section 70.32 of 10 CFR Part 70; and is subject to ali applicable provisions 01 tne AC'I Bnd the (ules, regulations, and orders of the Commission row or hereafter in effect; and is subject to Ihe additional conditions speclned below:

A. M§ximum Powe r Leve!

The licensee is au1horized to operate the facility at steady slate reactor core power levels not In excess of 25S7 megawatts (thermal).

S, Technical SPfts;tfications The Technical S~cificB1ions contained In Appenctlx Ai as revised through Amendment No. 274 are hereby Incorporated in the renewed license. Th~

licensee shall operate the facility In accordance with 1he Technical Specifications.

C. Reports the licensee shall maKe certain reports In accoroance with the requirements of the Technical Specifications.

D. Records The licensee shall keep facllity operat"mg records in accordance with the requirements of 1he Techrllcsl Specifications.

E. *Deleted by Amendment 65 F. Deleted by Amendment 7~

G. Deleted by AmenOment 227 H. Deleted by Amer¥:lment 227 Fire Proter:tlon The licensee shall Implement and main1ain in effect the provisions of the approved fire protection. program as descrtbed In the Updated Flnal Safety AnalysIs Repon and as approved In the SER osted September 19. 1979, (and Suppiements daled May 29,1980, October 9,1980, December 18.1980.

February n, 198'. DecemOer 4, 198', April 27, 1982, November 18, '982.

january n. 1984, February 25, '988, and SURRY UNIT ~

Renewed License No. DPR*32 Amendment No, 274

• 3 E. Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materiajs as may be produced by the operation of the facility, 3, This renewed license shall oe deemed to contain and is subject to the conditions specified in the following Commission regula1ions: 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 authoJized to ooerate the facllity at steady state reactor core Dower levels not in excess of 2557 megawatts (thermal).

B. Technical Spacifications The Technical Sf"Pl"':'ifications contained in Appendix A, as revised through Amendment No. 274

. are hereby incorporated in thIs renewed license. The licensee shall operate the facility in accordance with the Technical Specifications.

C. Reports The licensee shall make certain reports in accordance with the requirements of the Technical Speci*f!cations.

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. 274

TS 3.1-9 Heatup and cool down limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RT NOT, at the end of 48 Effecti ve Full Power Years (EFPY) for Units I and 2. The ~ost limiting value of RTNOT (222.5°F) occurs at the 1I4-T, 0° azimuthal location in the Unit 2 intermediate-to-lower shell circumferential weld. The limiting RTNOT at the 1/4-T location in the core region is greater than the RTNOT of the limiting un irradiated material. This ensures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RT NOT; the results are presented in UFSAR Section 4.1. Reactor operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RT NOT. Therefore, an adjusted reference temperature, based upon the copper and nic~el content of the material and the fluence was calculated in accordance with the recommendations of Regulatory Guide 1.99, Revision 2 "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cool down limit curves of Figures 3.1-1 and 3.1-2 include predicted adjustments for this shift in RTNOT at the end of 48 EFPY for Units 1 and 2 (as well as adjustments for location of the pressure sensing instrument).

Surveillance capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR 50, Appendix H. The surveillance specimen withdrawal schedule is shown in the UFSAR. The heatup and cool down curves must be recalculated when the ~RTNOT determined from the surveillance capsule exceeds the calCulated

~RTNDT for the equivalent capsule radiation exposure, or when the service period exceeds 48 EFPY for Units 1 and 2 prior to a scheduled refueling outage.

(

Amendment Nos. 274, 274

TS 3.1-11 KIt is the stress' intensity factor caused by the thermal gradients KIR is provided by the code as a function of temperature relative to the RT NDT of the material.

c =2.0 for level A and B service limits, and c ;;;;;; 1.5 for inservice hydrostatic and leak test operations.

At any time during the heatup or cooldown transient, KIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RT NDT. and the reference fracture toughness curve. The thermal stresses resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor. KIt. for the reference flaw is computed. From Equation (2) the pressure stress intensity factors are obtained and, from these. the allowable pressures are calculated.

The heatup limit curve, Figure 3.1-1, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60°F per hour. The cooldown limit curves of Figure 3.1-2 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The cool down limit curves are valid for cooldown rates up toJOO°F/hr.

The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of 48 EFPY for Units 1 and 2. The adjusted reference temperature was calculated using materials properties data from the B&W Owners Group Master Integrated Reactor Vessel Surveillance Program (MIRVSP) documented in the most recent revision to BAW -1543 and reactor vessel neutron fluence data obtained from plant-specific analyses.

Amendment Nos. 274, 274

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Figure 3.1-1 : Surry Units 1 and 2 Reactor Coolant Sys:em Heatup Limitations (Heatup Rates up to 60°Flhr) Applicable for 48 EFPY 400 Amendment Nos. 2]4" 274

Figure 3.1*2 Surry Units 1 and 2 Reactor Coolant System Cool down Limitations iMaterial Property BasIs Limiting Material: Surry Unit 1 Intermediate to Lower Shen eire Weld Limiting ART Values for Surry 1 al. 48 EFPY: 1/4-T, 228.4F 3/4-T, 169.5 F 2500.00

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Figure 3.1*2: Surry Units 1 and 2 Reactor O.olant System Cool down Limitations (Cooldown Rates up to lOOOP/1Ir) Applicable for 48 EFPY Amendment Nos. 274, 274

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 274 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 274 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

1.0 INTRODUCTION

By letter dated May 6,2010,1 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 proposed changes would revise the pressure-temperature (P-T) limit curves to provide new limits that are valid to 48 effective full power years (EFPY) for Surry 1 and 2.

2.0 REGULATORY EVALUATION

The Nuclear Regulatory Commission (NRC, the Commission) has established requirements in Title 10 of the Code of Federal Regulations (10 CFR), Part 50, to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The staff evaluates the P-T limit curves based on the following NRC regulations and guidance: 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements"; Generic Letter (GL) 88-11,2 "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations"; GL 92-01, Revision 1,3 "Reactor Vessel Structural Integrity"; GL 92-01, Revision 1, Supplement 1,4 Regulatory Guide (RG) 1.99, Revision 2,5 "Radiation Embrittlement of Reactor Vessel Materials";

and Standard Review Plan (SRP), Revision 2, Section 5.3.2.6 Appendix G to 10 CFR Part 50 requires that P-T limit curves be at least as conservative as those obtained by applying the 1 Agencywide Documents Access and Management System (ADAMS) Accession No. ML101310604, May 6,2010 2 ADAMS Accession No. ML031150357, July 12,1988 3 ADAMS Accession No. ML031070438, March 6,1992 4 ADAMS Accession No. ML031070449, May 19,1995 5 ADAMS Accession No. ML003740284, May 1988 6 ADAMS Accession No. ML080940723, March 2007

- 2 methodology of Appendix G to Section XI of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME B&PV Code). Appendix G to 10 CFR Part 50 also provides minimum temperature requirements that must be considered in the development of the P-T limit curves. GL 88-11 advised licensees that the NRC staff would use RG 1.99, Revision 2 to review P-T limit curves. RG 1.99, Revision 2 contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy (USE) resulting from neutron radiation. GL 92-01, Revision 1, requested that licensees submit their reactor pressure vessel (RPV) materials property data for review. GL 92-01, Revision 1, Supplement 1 requested that licensees provide and assess data from other licensees that could affect their RPV integrity evaluations.

SRP Section 5.3.2 provides an acceptable method of determining the P-T limit curves for ferritic materials in the beltline of the RPV based on the linear elastic fracture mechanics methodology of Appendix G to Section XI of the ASME B&PV Code. The basic parameter of this methodology is the stress intensity factor K" which is a function of the stress state and flaw configuration. ASME B&PV Code,Section XI, Appendix G requires a safety factor of 2.0 on stress intensities resulting from reactor pressure during normal and transient operating conditions, and a safety factor of 1.5 on these stress intensities for hydrostatic testing curves. The flaw postulated in the ASME B&PV Code,Section XI, Appendix G has a depth that is equal to 114 of the RPV beltline thickness and a length equal to 1.5 times the RPV beltline thickness. The critical locations in the RPV beltline region for calculating heatup and cooldown P-T limit curves are the 1/4 thickness (1/4T) and 314 thickness (3/4T) locations, which correspond to the maximum depth of the postulated inside surface and outside surface defects, respectively. The methodology found in Appendix G to Section XI of the ASME Code requires that licensees determine the adjusted reference temperature (ART or adjusted RTNDT) by evaluating material property changes due to neutron radiation. The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RTNDT), the mean value of the adjustment in reference temperature caused by irradiation

(~RTNDT) and a margin term. The ~RTNDT is a product of a chemistry factor (CF) and a fluence factor. The CF is dependent upon the amount of copper and nickel in the material and may be determined from tables in RG 1.99, Revision 2, or from surveillance data. The fluence factor is dependent upon the neutron fluence at the maximum postulated flaw depth. The margin term is dependent upon whether the initial RTNDT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Revision 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RTNDT, the copper and nickel contents, the neutron fluence and the calculational procedures. RG 1.99, Revision 2 describes the methodology to be used in calculating the margin term.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Evaluation The proposed P-T limit curves in the licensee's letter dated May 6,2010, extend the period of applicability of the current P-T limit curves, which were approved for 28.8 EFPY and 29.4 EFPY for Surry 1 and 2, respectively.

The licensee updated material property calculations using revised initial material properties and margins of Topical Report (TR) BAW-2308, Revision 2-A (publicly available through the letter from PWR (Pressurized-Water Reactor) Owners Group, to NRC, dated April 30, 2008,

Subject:

PWR Owners Group, Transmittal of NRC Approved Topical Report BAW-2308-NP, Revision 2,

- 3 "Initial RTNDTof Linde BO Weld Materials" (TAC No. MD4241), PA-MSC-0229).7 NRC staff had previously approved the use of TR SAW-230B, Revision 1-A (publicly available through the letter to Mr. Jerald S. Holm, Framatome ANP, from Herbert N. Serkow, NRC, transmitting the Final Safety Evaluation for Topical Report SAW-230B, Revision 1, "Initial RTNDT of Linde BO Weld Materials (TAC No. MS6636), dated August 4, 2005)6 for Surry 1 and 2 analyses. The revised initial RT NOT values and initial margin terms for Linde BO weld materials in SAW-230B, Revision 2-A differ slightly from the values contained in SAW-230B, Revision 1-A. The licensee also provided updated f1uence values, applicable to 4B EFPY, in performing the calculations.

3.2 Staff's Evaluation 3.2.1 ART Value and P-T Limit Curves Sy letter dated June 13, 2006,9 VEPCO submitted an exemption request from the requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61. Appendix G provides fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the system may be subjected over its service lifetime.

Section 50.61 provides fracture toughness requirements for protection against pressurized thermal shock (PTS). Section 50.61 (a)(5) and 10 CFR Part 50, Appendix G(II)(D)(i) require that preservice or unirradiated condition RT NOT be evaluated according to the procedures in the ASME Code,Section III, Paragraph NS-2331, from Charpy V-notch impact tests and drop-weight tests.

Framatome ANP TR SAW-230B, Revision 1-A provides an alternate method for determining adjusted RT NOT (reference nil-ductility temperature) and margins of the Linde BO weld materials present in the beltline region of the Surry 1 and 2 RPVs. Sy exemption dated June 27,2007,10 NRC staff approved the exemption request for the alternate material properties basis per 10 CFR 50.60(b). The exemption stated:

The exemptions are granted to the licensee to utilize the most recent staff-approved version of SAW-230B (currently SAW-230B, Revision 1). Future revisions of SAW-230B could affect fracture toughness data and analyses for Surry 1 and 2. Therefore, the licensee must review any future staff-approved revisions of SAW-230B and update the units' fracture toughness assessments, based on the information in any staff-approved revision of SAW-230B.

Sy letter dated March 24, 200B,11 the NRC staff approved the Final Safety Evaluation for Pressurized Water Reactor Owners Group TR SAW-230B, Revision 2. The values of initial RT NDT and initial margin terms for the Linde BO weld materials in SAW-230B, Revision 2-A were changed slightly from the values approved in TR SAW-230B, Revision 1-A.

7 ADAMS Accession No. ML081270388. April 30. 2008 B ADAMS Accession No. ML052070408, August 4, 2005 9 ADAMS Accession No. ML06165OO80. June 13,2006 10 ADAMS Accession No. ML071160287, June 27, 2007 11 ADAMS Accession No. ML080770349, March 24, 2008

- 4 For Surry 1 and 2, the applicable updated values for the Topical Reports are as follows:

Linde 80 Weld Wire Heat BAW-2308, Revision 1 Initial RTTOCF)

BAW-2308, Revision 1 Margin CF)

BAW-2308, Revision 2 Initial RTTOCF)

BAW-2308, Revision 2 Margin CF) i Generic

-47.6 17.2

-48.6 18.0 299L44

-81.8 11.6

-74.3 12.8 72445

-72.5 12.3

-72.5 12.0 The following Linde 80 weld materials are contained in Surry 1: 8T1554, 299L44, and 72445; and Surry 2 are: 8T1762 and 72445. Linde 80 weld wire heat numbers not listed specifically in the Table above must use the Generic values provided.

The staff evaluated the licensee's P-T limit curves for acceptability by performing independent calculations using the methodologies of Appendix G of Section XI of the ASME B&PV Code and 10 CFR Part 50, Appendix G. Based on the 48 EFPY neutron fluence projections and the updated initial values for the Linde 80 weld materials provided in TR BAW-2308, Revision 2-A, the staff's calculated ART values were in good agreement with the licensee's calculated ART values for the Surry 1 and 2 beltline materials. The limiting material for the 114 T RT NOT was determined to be the Surry 2Intermediate-to-Lower Shell Circumferential Weld R3008/0227, with a calculated value of 222.5 of. The limiting material for the 3/4T RTNOT was determined to be the Surry 2 Intermediate-to-Lower Shell Circumferential Weld R3008/0227, with a calculated value of 188.6 oF.

The current P-T limit curves for Surry 1 and 2 are based on limiting 1/4T and 3/4T RTNOT values of 228.4 of and 189.5 of, respectively. The values are higher than the values calculated using the 48 EFPY neutron fluence projections and the revised Linde 80 weld material values contained in TR BAW-2308, Revision 2-A. Therefore, the NRC staff concludes that the existing P-T limit curves in the current Surry 1 and 2 TSs remain valid for cumulative core burnups up to 48 EFPY.

Based upon the aforementioned limiting ART values, NRC staff verified that the licensee's proposed P-T limits are in accordance with Appendix G to Section XI of the ASME B&PV Code and satisfy the requirements in Paragraph IV.A.2 of Appendix G to 10 CFR Part 50.

3.2.2 Pressurized Thermal Shock The NRC staff evaluated the RPV beltline materials to ensure adequate resistance to failure during PTS events. The staff performed PTS reference temperature (RT PTS) calculations.

Projected values of RTPTS for PWR reactor vessel beltline materials are determined in accordance with 10 CFR 50.61. The limiting material with respect to PTS for Surry 1 is Lower Shell Longitudinal Weld SA-1526/299L44. The value of RT PTS for this material was determined to be 210.3 'F, which is below the PTS screening criterion of 270 'F for plates, forgings, and axial welds.

The limiting material with respect to PTS for Surry 2 is Intermediate-to-Lower Shell Circumferential Weld R3008/0227. The value of RTPTS for this material was determined to be 236.4 of, which is below the PTS screening criterion of 300 'F for circumferential welds.

The projected values of RTPTS for Surry 1 and 2 RPV beltline materials at 48 EFPY are below the 10 CFR 50.61 PTS screening criteria through 48 EFPY. Therefore, the NRC staff concludes that the submittal is in accordance with the requirements of 10 CFR 50.61.

- 5 3.2.3 Upper-Shelf Energy Charpy USE calculations for Surry 1 and 2 remain consistent with the USE calculations previously submitted by letter dated June 13, 2006. 12 The NRC staff reviewed the USE calculations and determined that the calculations were acceptable and in accordance with the USE requirements of Appendix G to 10 CFR Part 50. The exemption dated June 27, 2007, approved the USE calculations. The 48 EFPY neutron fluence projections and updated Linde 80 weld materials values contained in TR BAW-2308, Revision 2-A do not alter the neutron fluence or copper composition values previously submitted and approved. Therefore, the USE percentage drops, calculated using the RG 1.99, Revision 2, Position 1.2 methodology, and equivalent margins analyses (EMAs) for the Surry 1 and 2 RPV beltline materials remain unchanged from the calculations and EMAs previously approved. The NRC staff concludes that the USE analyses are consistent with the previously approved USE analyses, demonstrating that the Surry 1 and 2 RPV beltline materials meet the requirements of Appendix G to 10 CFR Part 50 by satisfying the 50 ft-Ib USE, or demonstrating through EMAs that acceptable results relative to Appendix K to Section XI of the ASME Code at 48 EFPY.

3.3 Conclusions The staff concludes that the proposed P-T limit curves for Surry 1 and 2 satisfy the requirements in Appendix G to 10 CFR Part 50 and Appendix G to Section XI of the ASME Code. Hence, the proposed P-T limit curves may be incorporated into the Surry 1 and 2 TSs and are valid through 48 EFPY. The staff also concludes that the PTS and USE evaluations for the Surry 1 and 2 RPVs continue to meet the requirements of 10 CFR 50.61 and 10 CFR Part 50, Appendix G, respectively.

4.0 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.

5.0 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 54396) published on September 7,2010.

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.

12 ADAMS Accession No. ML071010469. June 13, 2006

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6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: there is reasonable assurance that (1) 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 Contributor: Carolyn Fairbanks Date: May 31. 2011

May 31,2011 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 REACTOR VESSEL HEATUPAND COOLDOWN CURVES FOR 48 EFFECTIVE FULL-POWER YEARS (TAC NOS. ME3920 AND ME3921)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 274 to Renewed Facility Operating License No. DPR-32 and Amendment No. 274 to Renewed Facility Operating License No. DPR-37 for the Surry Power Station, Unit Nos. 1 and 2 (Surry 1 and 2), respectively. The amendments change the Technical Specifications in response to your application dated May 6, 2010.

These amendments revise the pressure and temperature limit curves to provide new limits that are valid to 48 effective full-power years for Surry 1 and 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, IRA!

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. 274 to DPR-32

2. Amendment No. 274 to DPR-37

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

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