ML111010523

Kewaunee Issuance of Amendment 208 Regarding Technical Specification Change Regarding Reactor Vessel Heatup and Cooldown Curves and Low Temperature Overpressure Protection for 52.1 Effective Full Power Years

ML111010523
Person / Time
Site:	Kewaunee
Issue date:	04/29/2011
From:	Feintuch K D Plant Licensing Branch III
To:	Heacock D A Dominion Energy Kewaunee
Feintuch K D, NRR/DORL/LPL3-1, 415-3079
Shared Package
ML111010552	List: ML111010523
References
TAC ME3777
Download: ML111010523 (13)

Text

UNITED NUCLEAR REGULATORY WASHtNGTON, D.C. 20555*0001 April 29, 2011 Mr. David A. Heacock President and Chief Nuclear Officer Dominion Energy Kewaunee, Inc. Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 SUBJECTKEWAUNEE POWER STATION -ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST TO CHANGE TECHNICAL SPECIFICATIONS REGARDING REACTOR VESSEL HEATUP AND COOLDOWN CURVES AND LOW TEMPERATURE OVERPRESSURE PROTECTION FOR 52.1 EFFECTIVE FULL POWER YEARS (TAC NO. ME3777)

Dear Mr. Heacock:

By application dated April 13, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 101110123), as supplemented by a letter dated January 18, 2011 (ADAMS Accession No. ML 110250293), Dominion Energy Kewaunee, Inc. (the licensee), requested an amendment to Renewed Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 208 to Renewed Facility Operating License No. DPR-43. The amendment provides new pressure-temperature (P-T) limit curves and low temperature overpressure protection system requirements.

The licensee revised the P-T limit curves to provide new limits that are valid to 52.1 effective full power years for KPS. 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, Karl Feintuch, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-305 Enclosures

Amendment No. 208 to License No. DPR-43 2. Safety Evaluation cc w/encls: Distribution via ListServ UN I TED NUCLEAR REGULATORY COMM IWAS H I NGT O N, D.C. 20555*0001 DOMINION ENERGY KEWAUNEE, DOCKET NO. KEWAUNEE POWER AMENDMENT TO RENEWED FACILITY OPERATING Amendment No. 208 Renewed License No. DPR-43 1The U.S. Nuclear Regulatory Commission (the Commission) has found that: The application for amendment by Dominion Energy Kewaunee, Inc. dated April 13, 2010, as supplemented by a letter dated January 18, 2011 , 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; BThe facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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

DThe issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and EThe issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requ i rements have been satisfied.

2Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-43, is hereby amended to read as follows:

-2 (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 208 ,are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. This license amendment is effective as of its date of issuance and shall be implemented within 120 days of the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Robert J. Pascarelli, Chief Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance:

April 29, 20ll ATTACHMENT TO LICENSE AMENDMENT NO. TO RENEWED FACILITY OPERATING LICENSE NO. DOCKET NO. Replace the following page of the Renewed Facility Operating License No. DPR-43 with the attached revised page. The changed area is identified by a marginal line. REMOVE INSERT Page 3 Page 3 Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change. REMOVE INSERT Page 3.4.3-3 Page 3.4.3-3 Page 3.4.3-4 Page 3.4.3-4 Page 3.4.5-1 Page 3.4.5-1 Page 3.4.6-1 Page 3.4.6-1 Page 3.4.10-1 Page 3.4.10-1 Page 3.4.12-1 Page 3.4.12-1 Page 3.4.12-2 Page 3.4.12-2 Page 3.4.12-3 Page 3.4.12-3 Page 3.4.12-4 Page 3.4.12-4 Page 3.5.2-1 Page 3.5.2-1

-3 Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive , possess and use in amounts as required any byproduct, source, or special nuclear material without restriction to chemical or physical form for sample analysis or instrument calibration or associated with radioactive apparatus or components; 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. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1 : (1) Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50 , and Section 70.32 of Part 70; and (2) is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and (3) is subject to the additional conditions specified or incorporated below: Maximum Power Level The licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 1772 megawatts (thermal). Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 208, are hereby incorporated in the renewed license. The licensee shall operate the facility in accordance with the Technical Specifications. Fire Protection The licensee shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the licensee's Fire Plan, and as referenced in the Updated Safety Analysis Report (USAR), and as approved in the Safety Evaluation Reports, dated November 25, 1977, and December 12,1978 (and supplement dated February 13,1981), subject to the following provision: The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission , only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Amendment No. 208 Renewed Operating License DPR-43

_ tI.O 'iii a.-QI... ::J QI... a. QI ... :cI: o ,I I ' -: i _=L+/-t-Material Property Basis --t --f-I I-Weld Metal ,-' I ........I ' I n -Cu = 0.287 wt% Ni = 0.756 wt% .-' -JI"r '\ -1 _ __I At 52.1 Effective Full Power Years -\. Adj. RT NOT at 1/4T = 278°F In-Service Leak Test ' 1 ,...... Adj. RT NDT at 3/4T = 230°F .." Minimum Temperature Upper Shell Forging -I .... 1 Cu = 0.12 wt% Ni = 0.71 wt% -_ Initial RT NOT = 60°F ---I-t ./ 1 CF = 84.65°F l-J _ Margin = 34 of =f--Heatup Rates _ -up to 5°F/Hr -C( t.::..:!----r-C(--.-,_ .-L (" to * + ---t--'r-) --f -, , . I I _ :=--"-1 F t -I , I At 52.1 Effective Full Power Years -V" / -" --...... V I,;" -Adj. RT NOT at 1/4T = 155°F Adj. RT NOT at 3/4T = 138°F -Vessel Flange -Initial RT NOT = 60°F --"-T' I -._._" '---1 If I I : T , I , l I I , I , ... h I I I A -./ -' I , \. -...... 17 '" T -I /. --.: lilt-t\.\ \ I 1 J I " Acceptable 1---, If' \ ! Criticality Limit Operation

-I \ . I I -H-r I I I Unacceptable I I-Heatup Rates I -I I I' I I Operation up to 100°F/Hr -r I I I I I I I ---i t I I I I I----' I I Margins for Instrumentation Error and I Pressure Drop Across RV Core I , l-I -\_. +13 0 F Instrumentation I I I -30 psi Instrumentation 1 I -I -70 psi .-1 I F I I-I-I :;I I I I 50 Kewaunee Power Station 200 Indicated Temperature (OF) Figure 3.4.3-1 (page 1 of 1) RCS PIT Limits 3.4.3 300 350 KPS Heatup, Criticality, and In-Service Leak Test Limitation Curves Applicable for Periods up to 52.1 Effective Full Power Years (EFPY) 3.4.3-3 Amendment No.

RCS PiT Limits 3.4.3 2500 2250 2000 1750 1500 -Cl "in Q. 1250-Q)... Q)... 1000 Il. "0 Q) co 750 -"0 c: 500 250 o I r-r-r-r-r-I I I I I I I Material Property Bas i s Weld Metal Cu =0.287 wt% Ni =0.756 wt% At 52.1 Effective Full Power Years Adj. RT NDT at 1/4T =278°F Adj. RT NDTat 3/4T =230°F Upper Shell Forging Cu =0.12 wt% Ni =0.71 wt% Initial RT NDT = 60°F CF =84.65°F Margin =34 of At 52.1 Effective Full Power Years Adj. RTNDTat 1/4T = 155°F i.oo""'" Adj. RT NDT at 3/4T = 138°F -i.oo""'"'

Vessel Flange /. " ". " Initial RT NDT =60°F foooo"""" "'-./' ...tIfI!'

-' ..... Unacceptable Operation .",. ./' ./ I I L.. II:.L r-.O°Flhr -.// " /"",.-.",!/720°F/hr 'r""/ "",.-./ / -40°F/hr i""'" 60°F/hr./ ---...... r-. 100°F/hr Acceptable Operation Margins for Instrumentation Error and Pressure Drop Across RV Core +13°F Instrumentation

-30 psi Instrumentation

-70 psi ll P I I I I J. I I I --50 100 150 200 250 Indicated Temperature (OF) Figure 3.4.3-2 (page 1 of 1) KPS Cooldown and L TOP Event Limitation Curves Applicable for Periods up to 52.1 EFPY 300 350 Kewaunee Power Station 3.4.3-4 Amendment No. 2.08 RCS Loops -MODE 3 3.4.5 REACTOR COOLAI\lT SYSTEM (RCS) 3.4.5 RCS Loops -MODE 3 LCO 3.4.Two RCS loops shall be OPERABLE, and one RCS loop shall be in operation. ---------------------------------------------NOT E S--------------------------------------All reactor coolant pumps (RCPs) may be removed from operation for:::; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period provided: No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1, "SHUTDOWN MARGIN"; and Core outlet temperature is maintained at least 10 "F below saturation temperature. No RCP shall be started with any RCS cold leg temperature:::;

356°F unless the secondary side water temperature of each steam generator (SG) is < 100°F above each of the RCS cold leg temperatures.

MODE 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RCS loop inoperable.

A.1 Restore RCS loop to OPERABLE status. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Required Action and associated Completion Time of Condition A not met. B.1 Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Kewaunee Power Station Amendment No. 208 RCS Loops -MODE 4 3.4.6 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Loops -MODE 4 LCO Two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be OPERABLE , and one loop shall be in operation. --------------------------------------------NOT E S-----------------------------------------1All reactor coolant pumps (RCPs) and RHR pumps may be removed from operation for:::; 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period provided:

aNo operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1, "SHUTDOWN MARGIN"; and Core outlet temperature is maintained at least 1Q°F below saturation temperature.

2No RCP shall be started with any RCS cold leg temperature:::;

356 °F unless the secondary side water temperature of each steam generator (SG) is < 1 00 above each of the RCS cold leg temperatures.

APPLICABILITYMODE 4. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required loop inoperable.

A.1 Initiate action to restore a second loop to OPERABLE status. Immediately AND A.2 NOT E Only required if RHR loop is OPERABLE. Be in MODE 5. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Kewaunee Power Station Amendment No. 208 Pressurizer Safety Valves 3.4.10 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 3.4.Two pressurizer safety valves shall be OPERABLE with lift settings 2410.45 psig and :5 2559.55 psig. ---------------------------------------------NOT E ------------------------------------------The pressurizer safety valves may be inoperable during a hydro test of the RCS provided the pressurizer power operated relief valves and the safety valves on the discharge pump are set at > the test pressure +35 psi. APPLICABILITYMODES 1, 2, and MODE 3 with both RCS cold leg temperatures>

356 °F. ---------------------------------------------NOT E ----------------------------------------The lift settings are not required to be within the LCO limits during MODE 3 (with both RCS cold leg temperatures>

356°F) for the purpose of setting the pressurizer safety valves under ambient (hot) conditions.

This exception is allowed for 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following entry into MODE 3 provided a preliminary cold setting was made prior to heatup. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pressurizer safety valve inoperable.

A.1 Restore valve to OPERABLE status. 15 minutes B. Required Action and associated Completion Time not met. OR Two pressurizer safety valves inoperable.

B.1 AND B.2 Be in MODE 3. Be in MODE 3 with any RCS cold leg temperatures

5 356°F. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours Kewaunee Power Station 3.4.Amendment No. 208 LTOP System 3.4 .12 REACTOR COOLANT SYSTEM (RCS) 3.4 .12 Low Temperature Overpressure Protection (L TOP) System LCO 3.4.12 APPLICABILITY

An L TOP System shall be OPERABLE with a maximum of one safety injection (SI) pump capable of injecting into the RCS and the accumulators isolated and one of the following pressure relief capabilities:

aThe Residual Heat Removal (RHR) System L TOP overpressure relief valve with a lift setting :-:::; 500 psig and two RHR suction flow paths OPERABLE; or An RCS vent of:::>: 6.4 square inches. ------------

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NOT E S ---Accumulator may be unisolated when accumulator pressure is less than the maximum RCS pressure for the existing RCS cold leg temperature allowed by the PIT limit curves provided in Figure 3.4.3-1 and Figure 3.4.3-2. A reactor coolant pump shall not be started with one or more RCS cold leg temperatures

-:::; 356°F unless the secondary water temperature of each steam generator is < 100°F above each of the RCS cold leg temperatures.

MODE 3 when any RCS cold leg temperature is :-:::; 356°FMODES 4 , MODE 6 when the reactor vessel head is on Kewaunee Power Station Amendment 208 LTOP System 3.4.12 ACTIONS -----------------------------

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LCO 3.0.4.b is not applicable when entering MODES 3 or 4. CONDITION REQUIRED ACTION COMPLETION TIME A. Two SI pumps capable of injecting into the RCS . A.1 Initiate action to verify a maximum of one SI pump is capable of injecting into the RCS. Immediately B. An accumulator not isolated when the accumulator pressure is greater than or equal to the maximum RCS pressure for existing cold leg temperature allowed in Figure 3.4.3-1 and Figure 3.4.3-2. B.1 Isolate affected accumulator.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. Required Action and associated Completion Time of Condition B not met. C.1 OR C.2 Increase RCS cold leg temperature to > 356°F. Depressurize affected accumulator to less than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the maximum RCS pressure for existing cold leg temperature allowed in Figure 3.4.3-1 and Figure 3.4.3-2 . 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Kewaunee Power Station 3.4.12-2 Amendment 208 LTOP System 3.4.12 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One RHR suction flow path inoperable. 0 .1 AND 0.2 Verify suction valves in the other RHR suction flow path are locked open with the motive power removed . Restore RHR suction flow path to OPERABLE status. Immediately 5 days E. Two RHR suction flow paths inoperable. OR Required Action and associated Completion Time of Condition A , C , or 0 not met. OR RHR System LTOP overpressure relief valve inoperable. E.1 Depressurize RCS and establish RCS vent of z 6.4 square inches. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Kewaunee Power Station 3.4.12-3 Amendment 208 LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of one SI pump is capable of injecting into the RCS. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.12.2 Verify each accumulator is isolated.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.12.3 Verify RHR suction valves are open for each RHR suction flow path. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.12.4 Verify required RCS vent ?: 6.4 square inches is open. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for unlocked open vent valve(s) AND 31 days for other vent path(s) Kewaunee Power Station 3.4.12-4 Amendment 208 ECCS -Operating 3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.2 ECCS -Operating LCO 3.5.2 Two ECCS trains shall be OPERABLE.

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NOT E S---------------------------------------A safety injection (SI) train may be considered OPERABLE for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> when being used to fill an SI accumulator , provided the other SI train is OPERABLE. In MODE 3, an SI pump may be made incapable of injecting to support transition into or from the Applicability of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System", for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or until the temperature of all RCS cold legs exceeds 381 °F, whichever comes first. APPLICABILITY

MODES 1,2 , and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains inoperable.

A.1 Restore train(s) to OPERABLE status. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> B. Required Action and associated Completion Time not met. B.1 AND B.2 Be in MODE 3. Be in MODE 4. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 12 hours C. Less than 100% of the ECCS flow equivalent to a single OPERABLE ECCS train available.

C.1 Enter LCO 3.0.3. Immediately Kewaunee Power Station Amendment No. 208 UNITED NUCLEAR REGULATORY WA S HINGTON, D.C. 2055 5-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 208 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-43 DOMINION ENERGY KEWAUNEE, INC. KEWAUNEE POWER STATION DOCKET NO. 50-305

1.0 INTRODUCTION

By application dated April 13 , 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 101110123), as supplemented by a letter dated January 18 , 2011 (ADAMS Accession No. ML 110250293), Dominion Energy Kewaunee, Inc. (the licensee), requested an amendment to Renewed Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). The supplement dated January 18 , 2011, 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) staff's original proposed no significant hazards consideration determination as published in the Federal Register on June 29,2010 (75 FR 37473). 1.1 Explanation for the need for the January 18 , 2011 Supplement At the time of the April 13, 2010 application , Facility Operating License No. DPR-43 was in effect with Custom Technical Specifications (CTS) as its Appendix A. The application document referenced CTS sections and supplied changed information to be reviewed that was based on the CTS. During the period of time preceding the January 18 , 2011 supplement , the CTS were converted to Improved Technical Specifications (ITS), which was the subject of license amendment No. 207 that was issued on February 2 , 2011 (ADAMS Accession No. ML 102020662). Further , on February 24, 2011, the NRC issued Renewed Operating License No. DPR-43 (ADAMS Accession No. ML 110100600), which was the successor to Operating License No. DPR-43. Thereafter, and including this issuance, the operating license to be amended is referred to as "Renewed Facility Operating License DPR-43." The purpose of the January 18, 2011, supplement was to convert the originally proposed CTS changes to a format and content consistent with the ITS format as approved on February 2, 2011, by NRC for KPS in response to the licensee's amendment request 249, submitted on August 24, 2009 (as amended). Whereas the application document of April 13 , 2010, discussed changes to CTS sections 3.1 and its associated Technical Specifications (TS) Bases, the supplement document of January 18 , 2011, discussed the equivalent changes in ITS section 3.4 and section 3.5 and their

-2 associated TS Bases. The supplement also provided additional restrictions on reactor coolant system (RCS) mass addition by (1) reactor coolant pump starting limits , (2) limits on the initial number of safety injection pumps capable of injecting reactor coolant, and (3) initially blocking coolant addition from accumulators, all until RCS cold leg temperature is greater than 356 OF, consistent with Improved Standard Technical Specifications.

This Safety Evaluation utilized both the April 13, 2010 and January 18, 2011 submittals in its analysis.

1.2 Purpose of the Proposed Amendment The proposed amendment would provide new pressure-temperature (P-T, or the equivalent PIT, if reading the licensee's application, supplement or other technical documents) limit curves and low temperature overpressure protection (L TOP) system requirements. The licensee revised the P-T limit curves to provide new limits that are valid to 52.1 effective full power years (EFPY) for KPS .

2.0 REGULATORY EVALUATION

The NRC 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 NRC 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, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations";

GL 92-01, Revision 1, "Reactor Vessel Structural Integrity";

GL 92-01, Revision 1, Supplement 1; Regulatory Guide (RG) 1.99, Revision 2 (Rev. 2), "Radiation Embrittlement of Reactor Vessel Materials";

and Standard Review Plan (SRP) Section 5.3.2. Appendix G to 10 CFR Part 50 requires that P-T limit curves be at least as conservative as those obtained by applying the methodology of Appendix G to Section XI of the American Society of Mechanical Engineers (ASME) 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, Rev. 2 to review P-T limit curves. RG 1.99, Rev. 2 contains methodologies for determining the increase in transition temperature and the decrease in upper-shelf energy resulting from neutron radiation.

GL 92-01, Rev. 1 requested that licensees submit their reactor pressure vessel (RPV) materials property data for their plants to the NRC staff for review. GL 92-01 , Rev. 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 Code. The basic parameter of this methodology is the stress intensity factor (K 1), which is a function of the stress state and flaw configuration. ASME 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 Code,Section XI, Appendix G has a depth that is equal to 1/4 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

-3/4 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 reference temperature nil ductility (RT NOT) by evaluating material property changes due to neutron radiation.

The ART is defined as the sum of the initial (unirradiated) reference temperature (initial RT NOT) , the mean value of the adjustment in reference temperature caused by irradiation (LlRT NDT) and a margin term. The LlRT NOT 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, Rev. 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 RT NOT is a plant-specific or a generic value and whether the CF was determined using the tables in RG 1.99, Rev. 2, or surveillance data. The margin term is used to account for uncertainties in the values of the initial RT NDT, the copper and nickel contents, the neutron fluence and the calculational procedures.

RG 1.99, Rev. 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 letters dated April 13, 2010, and January 18, 2011, are based on the current P-T limit curves which were approved for 31.1 EFPY. The licensee generated new heatup and cooldown curves using the most limiting ART values, the NRC-approved methodologies cited in the letter dated May 1, 2001 (ML011210180), "Kewaunee Nuclear Power plant -Exemption from the Requirements of 10 CFR Part 50, Appendix G, Appendix H, and Section 50.61 (Master Curve)," and the "axial flaw" and "circumferential flaw" methodologies of the 1998 Edition through the 2000 Addenda (which allows the use of the K 1c methodology) of ASME Code,Section XI, Appendix G. The licensee determined that the highest ART values at 52.1 EFPY are for the upper shell forging 123W250VA1

("axial flaw" orientation) and the RPV circumferential weld manufactured from weld wire heat 1 P3571 ("circumferential flaw" orientation). The licensee's ART determinations for these materials at the 1/4T and 3/4T locations were as follows: Material Location Neutron Fluence at Inside Surface (n/cm2) (E>1 MeV) ART CF) Upper Shell Forging 123W250VA1 1/4T 5.37 x 10 19 155 RPV Circumferential Weld Heat 1 P3571 1/4T 5.37 x 10 19 278 Upper Shell Forging 123W250VA1 3/4T 5.37x1019 138 RPV Circumferential Weld Heat 1 P3571 3/4T 5.37x 10 19 230

-4 The TS changes submitted by the letters dated April 13, 2010, and January 18, 2011 , include: Modified P-T limit curves , extended to 52.1 EFPY. L TOP system Tenable (that is , the enabling temperature for LTOP system activation) values to reflect the extended period of 52.1 EFPY. 3.2 NRC Staff's Evaluation 3.2.1 ART Value and P-T Limit Curves The licensee stated that the proposed P-T limit curves were based on the methodologies of Appendix G of Section XI of the ASME Code, 1998 Edition with the 2000 Addenda, which utilizes alternative reference fracture toughness (Krc) curve instead of the Kia fracture toughness curve for RPV materials in determining the P-T limit curves. The 1998 Edition with the 2000 Addendum of the ASME Code,Section XI has also incorporated the provisions of ASME Code Case N-588 into Appendix G which modified the methodology endorsed by the ASME Code regarding the postulation of circumferential weld flaws for the purpose of P-T limit generation.

NRC Regulatory Issue Summary 2004-04, "Use of Code Cases N-588, N-640, and N-641 in Developing Pressure-Temperature Operating Limits," dated April 5, 2004 (ML040920323), states that the ASME Code,Section XI, Appendix G, 1998 Edition with the 2000 Addenda may be used without the need for an exemption.

To assess the validity of the licensee's proposed curves , the NRC staff performed an independent assessment of the licensee's submittal.

Previously, by letter dated May 1,2001 (ADAMS Accession No. ML011210180), NRC staff approved the following exemptions for KPS, which addressed the licensee's evaluation of the RPV circumferential weld manufactured from weld wire heat 1 P3571:

• An exemption to establish the use of a new methodology to meet requirements of 10 CFR Part 50, Appendix

• An exemption to modify the RPV surveillance program of 10 CFR Part 50, Appendix H, to incorporate acquired toughness data,

• An exemption to establish the use of a new methodology to meet requirements of 10 CFR The RPV surveillance capsule results for the KPS RPV circumferential weld meet the credibility criteria of 10 CFR 50.61, paragraph (c)(2)(i) and were therefore used to determine a material-specific chemistry factor. Enclosure 1 (WCAP-15074, Revision 1) to the submittal contained an evaluation of the 1 P3571 weld metal from the RPV surveillance programs at KPS and Maine Yankee (MY). The licensee presented different averaging methods to assess best estimate chemistry values. The NRC staff agreed that the copper coil-weighted average of 0.287 weight percent and the simple averaging of 0.756 weight percent for nickel were acceptable best estimate values for weld metal heat 1 P3571 .

-5 NRC staff examined the heat-adjusted RT NDT values for the KPS RPV circumferential weld which were derived using the Master Curve method previously approved by the NRC staff. Enclosure 3 to the submittal (WCAP-16641) contained the data and analysis of KPS Capsule T. This capsule was removed with a calculated fluence of 5.62 x 10 19 n/cm 2 (E > 1.0 MeV), which is close to 52.1 EFPY calculated fluence of 5.37 x 10 19 n/cm 2 (E > 1.0 MeV) for the RPV. In Enclosure 2 (WCAP-16609) to the submittal, Master Curve fracture toughness data from irradiated KPS Surveillance Capsule T weld metal heat 1 P3571 were evaluated to derive a direct measurement of RT To to use in place of adjusted RT NOT for the RPV limiting circumferential weld. The RT To at end of life (EOL) fluence was determined by making direct measurement of irradiated fracture toughness using precracked Charpy specimens.

Fracture toughness data generated on the same weld wire heat number from KPS Surveillance Capsule S and from MY Surveillance Capsule A-35 were also included in the evaluation.

Based on these calculations, the NRC staff verified that the licensee's limiting material in the circumferential flaw case was the KPS RPV circumferential weld manufactured from wire heat 1 P3571. The staff's calculated ART values were in good agreement with the licensee's calculated ART values of 278 of and 230 of for the 1/4T location and the 3/4T location, respectively, for the KPS circumferential weld manufactured from weld wire heat 1 P3571 . Since there are no RPV surveillance data available for the upper shell forging or vessel flange, Regulatory Positions 1.1 and 1.2 within the limitations in Regulatory Position 1.3, as described in RG 1.99, Revision 2, Section C "Regulatory Position," apply. As a consequence, chemistry factors were determined from RG 1.99, Revision 2, Table 1. For the upper shell forging 123W250VA 1, ART values were determined using the methodology in RG 1.99, Revision 2. Based on these calculations, the NRC staff verified that the licensee's limiting material in the axial flaw case is upper shell forging 123W250VA1.

The NRC staff's calculated ART values were in good agreement with the licensee's calculated ART values of 155 of and 138 of for the 1/4T location and 3/4T location, respectively, for upper shell forging 123W250VA

1. The NRC 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 °Code and 10 CFR Part 50, Appendix G. The proposed P-T limit curves in the letters dated April 13, 2010, and January 18, 2011, modified the P-T limit curves using approved methodologies, extending the curves to 52.1 EFPY. The P-T limit curves apply to both heatup and cooldown and for both 1/4T and 3/4T locations.

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 Code and satisfy the requirements in Paragraph IV.A.2 of Appendix G to 10 CFR Part 50. Additionally, RT PTS is defined as the reference temperature, RT NOT, evaluated for the EOL Fluence for the limiting KPS RPV circumferential weld manufactured from weld wire heat 1 P3571. The NRC staff noted that RT PTS for KPS is 297.5 of at EOL with a 95.6 percent capacity factor. The RT PTS was calculated using the methodology in the NRC safety evaluation dated May 1, 2001 (ML011210180). This temperature is below the 300 OF screening criteria contained in 10 CFR 50.61.

-6 3.2.2 Low Temperature Overpressure Protection (L TOP) The L TOP system, provided by the power operated relief valves and also by the shut down cooling relief valves, ensures RCS over pressurization below certain temperatures would be prevented, thus maintaining reactor coolant pressure boundary integrity. The L TOP analysis yields Limiting Conditions for Operation that constitute L TOP system alignments for the period of applicability. The NRC staff reviewed the LTOP analysis using the guidance contained in Branch Technical Position 5-2, "Overpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures," of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." The current L TOP enabling temperature is 200 of. The proposed L TOP enabling temperature is 343 of based on the criteria specified in Appendix G of the ASME code,Section XI, 1998 Edition through 2000 Addenda. Accordingly, this value was determined using a value of RT NDT for the 1/4T position of the circumferential weld projected to 52.1 EFPY of 278 of, plus 50 of, plus 15 of to compensate for the maximum potential difference between the coolant temperature and the metal temperature. Factoring in instrument uncertainty (13 OF), the L TOP will be enabled when the indicated RCS temperature reaches 356 of. The L TOP setpoint is unaffected by this proposed license amendment, and remains at 500 pounds per square inch gauge. Therefore, the NRC staff accepts the licensee's analyses and concludes that the requested L TOP system limitations are acceptable. The NRC staff has reviewed the L TOP System changes, as expressed on the revised TS pages 3.4.12-1, 3.4.12-2, 3.4.12-3 (on which only CONDITION and ACTION labeling was affected), 3.4.12-4 and 3.5.2-1, associated with limiting mass injection capability to a maximum of one safety injection (SI) pump capable of injecting into the RCS. The NRC staff acknowledges that the increased L TOP enabling temperature of 356 of required adding requirements derived from Standard Technical Specifications, NUREG-1431 that were previously not applicable when the L TOP enabling temperature was 200 of. Those newly effective requirements are restrictions on the SI pumps, as expressed on revised TS pages 3.4.12-1, 3.4.12-2, 3.4.12-4, and 3.5.2-1, and restrictions on the accumulators, which are expressed on TS pages 3.4.12-1, 3.4.12-2, and 3.4.12-4.

These revisions addressed the requirements, the conditions if requirements are not met, and the surveillance to verify that the requirements are met. All of these changes are more restrictive than the previous requirements.

The NRC staff concludes that the requested changes to the revised TS pages are acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Wisconsin State official was notified on April 12, 2011, of the proposed issuance of the amendment.

The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION This amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes a surveillance requirement.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no Significant change in the types, of any effluent that may be released offsite and that there is no significant increase in individual or cumulative

-7 occupational radiation exposure.

The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding (75 FR 37473 dated June 29,2010). Accordingly, this amendment meets 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 this amendment.

6.0 CONCLUSION

The NRC staff concludes that the proposed P-T limit curves and L TOP system requirements for KPS 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 and L TOP system requirements may be incorporated into the KPS TS and are valid through 52.1 EFPY. 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 this amendment will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributor:

Carolyn Fairbanks, NRR Date: April 29, 2011 Mr. David A. Heacock April 29, 2011 President and Chief Nuclear Officer Dominion Energy Kewaunee, Inc. Innsbrook Technical Center 5000 Dominion Boulevard Glen Allen, VA 23060-6711 KEWAUNEE POWER STATION -ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST TO CHANGE TECHNICAL SPECIFICATIONS REGARDING REACTOR VESSEL HEATUP AND COOLDOWN CURVES AND LOW TEMPERATURE OVERPRESSURE PROTECTION FOR 52.1 EFFECTIVE FULL POWER YEARS (TAC NO. ME3777)

Dear Mr. Heacock:

By application dated April 13, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 101110123), as supplemented by a letter dated January 18, 2011 (ADAMS Accession No. ML 110250293), Dominion Energy Kewaunee, Inc. (the licensee), requested an amendment to Renewed Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). The U.S. Nuclear Regulatory Commission (NRC, the Commission) has issued the enclosed Amendment No. 208 to Renewed Facility Operating License No. DPR-43. The amendment provides new pressure-temperature (P-T) limit curves and low temperature overpressure protection system requirements.

The licensee revised the P-T limit curves to provide new limits that are valid to 52.1 effective full power years for KPS. 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! Karl Feintuch, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-305 Enclosures

Amendment No. 208 to License No. DPR-43 2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRIBUTION

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