ML102530115

Issuance of Amendments Regarding Request for Technical Specification Revisions Related to the Core Operating Limits Report(Tac Nos. ME2591 and ME2592)

ML102530115
Person / Time
Site:	Surry
Issue date:	10/19/2010
From:	Cotton K Plant Licensing Branch II
To:	Heacock D Virginia Electric & Power Co (VEPCO)
cotton K, NRR/DORL, 415-1438
References
TAC ME2591, TAC ME2592
Download: ML102530115 (35)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 19, 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 SUB"IECT: SURRY POWER STATION, UNIT NOS. 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REQUEST FOR TECHNICAL SPECIFICATION REVISIONS RELATED TO THE CORE OPERATING LIMITS REPORT (TAC NOS. ME2591 AND ME2592)

Dear Mr. Heacock:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 270 to Renewed Facility Operating License No. DPR-32 and Amendment No. 269 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 October 16, 2009, as supplemented by letter dated May 7, 2010.

These amendments revise the following licensing basis changes as follows:

(1) Implementation of WCAP-8745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function,"

(2) Implementation of NRC-approved Dominion Fleet Report DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code,"

(3) Implementation of a Statistical Design Limit for the analytic code and critical heat flux correlations that are being used in DOM-NAF-2-A, and (4) Implementation of Dominion TR VEP-NE-2-A, "Statistical DNBR Evaluation Methodology. "

The requested change also affects the facility TSs. Items 1 and 2 in the above list are methodologies that are used in the determination of core operating limits; hence, an amendment to the TSs was also requested to insert the items into the reference list contained in TS 6.2.C, "CORE OPERATING LIMITS REPORT."

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,

~~

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. 270 to DPR-32

2. Amendment No. 269 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. 270 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 October 16,2009, as supplemented by letter dated May 7,2010, 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. 270 , 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 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION 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: October 19, 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. 269 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) October 16,2009, as supplemented by letter dated May 7,2010, 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. 269 , 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 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION C ICv.-....--.

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: October 19, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 270 RENEWED FACILITY OPERATING LICENSE NO. DPR-32 DOCKET NO. 50-280 AND TO LICENSE AMENDMENT NO. 269 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 TS 2.1-1 TS2.1-1 TS 2.1-2 TS2.1-2 TS 2.1-3 TS2.1-3 TS 2.1-4 TS2.1-4 TS 2.1-5 TS Fig. 2.1-1 TS Fig. 2.1-2 TS Fig. 2.1-3 TS Fig. 2.1-4 TS 2.3-2 TS 2.3-2 TS 2.3-3 TS 2.3-3 TS 2.3-5 TS 2.3-5 TS Fig. 2.3-1 TS 3.12-12 TS 3.12-12 TS3.12-12a TS 3.12-12a TS 3.12-20 TS.3.12-20 TS 4.1-5 TS 4.1-5 TS 4.1-5a TS 4.1-9d TS4.1-9d TS 6.2-1 TS 6.2-1 TS 6.2-2 TS 6.2-2

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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 SO. and Section 70.32 of 10 CFR Part 70; and is subject to all applicable provisJo ns of .tl1e ACl 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 to The licensee is authorized operate the facility at steady state reactor core power levels not In excess of 2546 megawatts (thermal).

B. Technical Soeclfications The Technical S~cifications contained In Appendix A, as revised through Amendment No. 270 are hereby Incorporated in the renewed license. The licensee shall operate the facility In accordance with the Tttchnical Specifications.

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

D. Records The Ilcensee shall keep facility operating records in accordance 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

[. Flr8 ProtectIon The licensee shall Implement and maintain in effect the provisions of the approved fire protection. program as descrlbed In the Updated 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 17,1964. February 25, 1988. and SURRY UNIT 1 Renewed License No. DPR-32 Amendment No. 270

-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~hor1zeq to operate the facility at steady state reactor core power levels not in excess of 2546 megawatts (thermal).

B. Technical Specifications The Technical Specifications contained in Appendix A, as rev\sed through Amendment No. 269 , are hereby incorporated in this renewed license. The licensee shall operate the facUity in accordance with the Technlcal*Specifications.*

C. Reports The licensee sttall make certain reports in accordance with the reqUirements of the Technical Speci'ftcations.

D. Records The licensee shall keep facility operating records tn 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. De\eted by Amendment 227 SURRY - UNIT 2 Renewed License No. OPR-37 Amendment NO.269

TS 2.1-1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT, REACTOR CORE Applicability Applies to the limiting combinations of THERMAL POWER, Reactor Coolant System pressure, coolant temperature and coolant flow when a reactor is critical.

Objective To maintain the integrity of the fuel cladding.

Specification A. The combination of reactor THERMAL POWER level, pressurizer pressure, and Reactor Coolant System (RCS) highest loop average temperature shall not:

I. Exceed the limits specified in the CORE OPERATING LIMITS REPORT when full flow from three reactor coolant pumps exists, and the following Safety Limits shall not be exceeded:

a. The design limit for departure from nucleate boiling ratio (DNBR) shall be maintained ~ 1.27 for transients analyzed using the Statistical DNBR Evaluation Methodology and the WRB-l DNB correlation. For transients analyzed using the deterministic methodology, the DNBR shall be maintained greater than or equal to the applicable DNB correlation limit ~ 1.17 for WRB-l, ~ 1.30 for W-3).

b. The peak fuel centerline temperature shall be maintained < 5080°F, decreasing by 58°F per 10,000 MWDIMTU of burnup.

2. The reactor THERMAL POWER level shall not exceed 118% of rated power.

Amendment Nos. 270 and 269

TS2.1-2 B. In the event the Safety Limit is violated, the facility shall be placed in at least HOT SHUTDOWN within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The safety limit is exceeded if the combination of RCS highest loop average temperature and THERMAL POWER level is at any time above the appropriate pressure line as specified in the CORE OPERATING LIMITS REPORT; or the co~e THERMAL POWER exceeds 118% of the rated power.

To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under all operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is very large and the clad surface temperature is only a few degrees Fahrenheit above the reactor coolant saturation temperature. The upper boundary of the nucleate ,boiling regime is termed Departure From Nucleate ""' Boiling (DNB) and at this point there is a sharp reduction of the heat transfer coefficient, which would result in high clad temperatures and the possibility of clad failure. DNB is not, however, an observable parameter during reactor operation. Therefore, DNB has been correlated to thermal power, reactor coolant temperature and reactor coolant pressure which are observable parameters. This correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the DNB heat flux at a particular core location to the local heat flux, is indicative of the margin to DNB. The DNB basis is as follows: there must be at least a 95% probability with 95% confidence that the minimum DNBR of the limiting rod during Condition I and IT events is greater than or equal to the DNBR limit ofthe DNB correlation being used. The correlation DNBR limit is based on the entire applicable experimental data set to meet this statistical criterion. (l) I The figure provided in the CORE OPERATING LIMITS REPORT shows the loci of points of THERMAL POWER, RCS pressure, and average temperature for which the minimum DNBR is not less than the safety analyses limit, that fuel centerline temperature remains below melting, that the j

average enthalpy in the hot leg is less than or equal to the enthalpy of saturated liquid, or that the exit quality is within the limits defined by the DNBR correlation. The area where clad integrity is assured is below these lines. The temperature limits are considerably more conservative than would Amendment Nos. 270 and 269

TS 2.1-3 be required if they were based upon the design DNBR limit alone but are such that the plant conditions required to violate the limits are precluded by the self-actuated safety valves on the steam generators. The effects of rod bowing are also considered in the DNBR analyses.

The reactor core Safety Limits are established to preclude violation of the following fuel design criteria:

a. There must be at least a 95% probability at a 95% confjdence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB and

b. There must be at least a 95% probability at a 95% confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

The reactor core Safety Limits are used to define the various Reactor Protection System (RPS) functions such that the above criteria are satisfied during steady state opera~ion, normal operational transients, and anticipated operational occurrences (AOOs). To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied. to the Overtemperature and Overpower AT reactor trip functions. That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the cdre exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that the variations~in the THERMAL POWER, RCS pressure, RCS average temperature, RCS flow rate, and M that the reactor core Safety Limits will be satisfied during steady state operation, normal operational transients, and AOOs.

The Reactor Control and Protection System is designed to prevent any anticipated combination of transient conditions for Reactor Coolant System temperature, pressure and thermal power level that would result in a DNBR less than the design DNBR limit(3) based on steady state nominal operating power levels less than or equal to 100%, steady state nominal operating Reactor Coolant System average temperatures less than or equal to 573.0°F and a steady state nominal operating pressure of 2235 psig. For deterministic DNBR analysis, allowances are made in initial conditions assumed for transient analyses for steady state errors of +2% in power, +4°F in Reactor Coolant System average temperature and +/-30 psi in pressure. The combined steady state Amendment Nos. 270 and 269

TS 2.1-4 errors result in the DNB ratio at the start of a transient being 10 percent less than the value at nominal full power operating conditions.

For statistical DNBR analyses, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% probability that the minimum DNBR for the limiting rod is greater than or equal to the statistical DNBR limit. The uncertainties 'in t4e plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, e~tablishes a statistical DNBR limit which must be met in plant safety analyses using values of input parameters without uncertainties. The statistical DNBR limit also ensures that at least

/

99.9% of the core avoids the onset of DNB when the limiting rod is at the DNBR limit.

)

The fuel overpower design limit is 118% of rated. power. The overpower limit criterion is that core power be prevented from reaching a value at which fuel pellet melting would occur. The value of 118% power allows substantial margin to this limiting criterion. Additional peaking factors to account for local peaking due to fuel rod axial gaps and reduction in fuel pellet stack length have been included in the calculation of this limit.

References

1) FSAR Section 3.4

2) FSAR Section 3.3

3) FSAR Section 14.2 Amendment Nos. 270 and 269

TS 2.3-2 (b) High pressurizer pressure - ~ 2380 psig.

(c) Low pressurizer pressure - ~ 1875 psig.

(d) Overtemperature liT

~T ~ ~To[K:I - K 2C: :~D (T - T) + K3 (P -PI) - f(~I)J Where: ,

. ~T is measured RCS ~T, oF.

~To is the indicated ~T at RATED POWER, oF.

s is the Laplace transform operator, sec-I.

T is the measured RCS average temperature (Tavg)' oF.

T' is the nominal Tavg at RATED POWER, ~ [*] oF.

P is the measured pressurizer pressure, psig.

P' is the nominal RCS operating pressure, ~ [*]psig.

K 1 ~ [*] Kz ~ [*WP K3 ~ [*]/psig t1 ~ [*] sec tz ~ [*] sec f(AI) = [*] {[*] - (qt - qb)} when qt - qb < [*]% RATED POWER o when [*]% RATED POWER ~ qt - qb ~ [*]% RATED POWER

[*]{ (qt - qb)-[*]) when qt - qb > [*]% RATED POWER Where qt and qb are percent RATED POWER in.the upper and lower halves of the core, respectively, and 'It + qb is the total THERMAL POWER in percent RATED POWER.

The values denoted with [*] are specified in the CORE OPERATING LIMITS REPORT.

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.0% of the ~T span. (Note that 2.0% ofthe ~T span is equal to 3.0% ~T PoweL) .

(e) Overpower ~T Amendment Nos. 270 and 269

TS 2.3-3 Where:

~T is measured RCS ~ T, oF.

~T 0 is the indicated ~ T at RATED POWER, oF.

s is the Laplace transform operator, sec-I.

T is the measured RCS average temperature (Tavg )' oF.

T' is the nominal Tavg at RATED POWER, 2 [*] OF.

K s ::=: [*]jOF for decreasing Tavg K 6 ::=: [*]jOF when T > T'

[*] / of for increasing Tavg [*]jOF when T2 T' t3 ::=: [*] sec f(M) = [*]

The values denoted with [*] are specified in the CORE OPERATING LIMITS REPORT.

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.0% of the ~T span. (Note that 2.0% of the ~T span is equal to 3.0% of ~T Power.)

(f) Low reactor coolant loop flow - 2=: 91 % of normal indicated loop flow as measured at elbow taps in each loop (g) Low reactor coolant pump motor frequency - 2=: 57.5 Hz (h) Reactor coolant pump under voltage - 2=: 70% of normal voltage

3. Other reactor trip settings (a) High pressurizer water level- :s 89.12% of span (b) Low-low steam generator water level - 2=: 16% of narrow range instrument span (c) Low steam generator water level - 2=: 19% of narrow range instrument span in coincidence with steam/feedwater mismatch flow - :S 1.0 x 106 lbs/hr (d) Turbine trip (e) Safety injection - Trip settings for Safety Injection are detailed III TS Section 3.7.

Amendment Nos. 270 and 269

TS 2.3-5 The overtemperature /),.T reactor trip provides core prote~tion against DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided only that the transient is slow with respect to piping transit delays from the core to the temperature *detectors (about 3 seconds), and pressure is within the range between high and low pressure reactor trips.

With normal axial power distributio~; the reactor trip limit; with allowance for errors, (2) is' always below the core safety limit as specified. in the CORE OPERATING LIMITS REPORT. If axial peaks are greater than design,. as indicated by the difference between top and bottom, power range nuclear detectors, the reactor limit is automatically reduced.(4)(5)

The overpower and oveI1emperature ~rotection system setpoints have been revised to include effects of fuel densification on core safety limits and to apply to 100% of design flow. The revised setpoints in the Technical Specifications will ensure that the combination of power, temperature, and pressure will not exceed the revised core safety limits as specified in the CORE OPERATING

. '. . I LIMITS REPORT. The reactor is prevented from reaching the overpower limit condition by action of the Tluclear overpower and overpower /),.T trips. The overpower limit criteria is that core power be prevented from reaching a value at which fuel pellet centerline melting would occur.

The overpower protection system set points include the effects of fuel densification.

The overpower /),.T reactor trip prevents power density anywhere in the core from exceeding 118% )

of design power density as discussed Section 7 and specified in Section 14.2.2 of the FSAR and includes corrections for axial power dJstribution, change in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop' temperature detectors. The specified setpoints meet this requirement and include allowance for instrument errors. (2)

Refer to Technical Report EE-0116 for justification of the dynamic limits (time constants) for the Overte~pera~e /),.T and Overpower /),.T Rteactor Trip functions.

Amendment Nos. 270 and 269

)

TS 3.12-12

3. If more than one rod position indicator per group is inoperable, place the control rods un~er manual control immediately, monitor and record RCS Tavg once I?er hour, verify the position of the control rod assemblies indirectly using the movable" incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and restore inoperable position indicators to OPERABLE status such that a maximum of one position indicator c

per group is inoperable within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

\

4. If one or more rods with inoperable position indicators have been moved in excess of 24 steps in one direction since the last determination of the rod's position, verify the position of the control rod assemblies indirectly using the movable incore detectors within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce power to less than 50% of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. .

5. If one group step demand counter per bank: for more than one or more banks is /

inoperable, verify that all rod position indicators for the affected bank(s) are OPERABLE once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and verify that the most withdrawf!. rod and the least withdrawn ro.d of the affected bank:(s) are less than or equal to 12 steps apart once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Alternatively, reduce power to less than 50% of RATED POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

6. If the requirements of Specification 3.12.E.2, 3.12.E.3, 3.12.E.4, or 3.12.E.5 are not satisfied, then the unit shall be placed in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

F. DNB Parameters

1. The following DNB related parameters shall be maintained within their limits during POWER OPERAnON:

• Reactor Coolant System Tavg $: the limit specified in the CORE OPERATING LIMITS REPORT

• Pressurizer Pressure ~ the limit specified in the CORE OPERATING LIMITS REPORT

• Reactor Coolant System Total Flow Rate ~ 273,000 gpm and ~ the limit

(

specified in the CORE OPERATING LIMITS REPORT Amendment Nos. 270 and 269

TS 3.12-12a

a. The Reactor Coolant System Tavg

• Pressurizer Pressure, and Reactor Coolant System Total Flow Rate shall be verified to be within their limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

)

.b. TheReactor Coolant System Total Flow Rate shall be detennined to be within its limit by precision heat balance with the frequency specified in TS Table 4.1-2A.

2.. When any of the parameters in Specification 3.12.F.l has been determined to exceed its limit, either restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3. The limit for Pressurizer Pres*sure in Specification 3.I2.F.I is not applicable during either a THERMAL POWER ramp increase in excess of 5% of RATED POWER per minute or a THERMAL POWER step increase in excess of 10% of RATED POWER.

G. Shutdown Margin v 1. Whenever the reactor is subcritical, the shutdown margin shall be . within the limits specified in the CORE OPE~TING LIMITS REPORT. If the shutdown margin is not within limits, within 15 minutes, initiate boration to restore shutdown margin to within limits.

Amendment Nos. 270 and 269

TS 3.12-20 A 2% QUADRANT POWER TILT allows th~t a 5% tilt might/actually be present in the core because of insensitivity of the e.xcore detectors for disturbances near the core center such as misaligned inn~r control rod assembly and an error allowance. No increase in FQ' occurs with tilts up to 5% because misaligned control rod assemblies producing such tilts do not extend to the unrodded plane, where the maximum FQ occurs... , .,

The QPTR limit must be maintained during power operation with THERMAL POWER> 50% of RATED POWER to prevent core power distributions from exceeding the design limits.

Applicability during power operation ~ 50% RATED POWER or when shut down is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the FN Mf and FQ(Z) LeOs still appl;, but allow progressively higher peaking factors at 50%

RATED. POWER or lower.

The limits of the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accidenf I

analyses. The limits are consistent with the UFSAR assumptions and have been analytically demonstrated to be adequate to maintain a minimum DNBR which is 'greater than the design limit throughout each analyzed transient. Measurement uncertainties are accounted for in the DNB design margin. Therefore, measurement values are compared directly to the surveillance limits without applying instrument uncertainty.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of temperature 3;nd pressure through instrument readout is sufficient to ensure that these param~ters are restored to within their limits following load changes and other expected transient operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of RCS total flow rate, by installed flow instrumentation, is sufficient to regularly assess potential degradation and to verify' operation within safety analysis assumptions. Measurement of RCS total flow rate by performance of a precision calorimetric heat balance specified in TS Table 4.1-2A allows for the installed RCS flow instrumentation to be calibrated and verifies that the aC,tual RCS flow rate is greater than or equal to the minimum required RCSflow rate.

Amendment Nos. 270 and 269 I

1 TS4.I-S The refueling water storage tank is sampled weekly for Cl- and/or F contaminations. Weekly sampling is adequate to detect any inleakage of contaminated water.

Control Room Bottled Air System The control room bottled air system is required to establish a positive differential pressure in the control room for one hour following a design basis accident. The ability of the system to meet this requirement is verified by: I) checking air bottle pressurization, 2) demonstrating the capability to pressurize the control room pressure boundary, 3) functionally testing the pressure control valve(s), and

4) functionally testing the manual and automatic actuation capability. The test requirements and frequency are specified in Table 4.l-2A.

Pressurizer PORV, PORV Block Valve, and PORV Backup Air Supply The safety-related, seismic PORV backup air supply is relied upon for two functions - mitigation of a design basis steam generator tube rupture accident and low temperature overpressure protection (LTOP) of the reactor vessel during startup and shutdown. The surveillance criteria are based upon the more limiting requirements for the backup air supply (i.e. more PORV cycles potentially required to perfonn the mitigation function), which are associated with the LTOP function.

The PORV backup air supply system is provided with a calibrated alann for low air pressure. The alarm is located in the control room. Failures such as regulator drift and air leaks which result in l~w pressure can be easily recognized by alarm' or annunciator action. A periodic quarterly verification of air pressure against the surveillance limit supplements this type of built-in surveillance. Based on experience in operation, the minimum checking frequencies set forth are deemed adequate.

RCS Flow The frequency of 18 months for RCS flow surveillance reflects the importance of verifying flow after a refueling outage when the core has been altered, which may Amendment Nos. 270 and 269

TS 4.1-5a have caused an alteration of the flow resistance. This surveillance requirement in Table 4.1-2A is modified by a note that allows entry into POWER OPERATION, without having performed the surveillance, and placement of the unit in the best condition for performing the surveillance. The note states that the surveillance requirement is not required to be performed until 7 days after reaching a THERMAL POWER of ~ 90% of RATED POWER. The 7 day period after reaching 90% of RATED POWER is reasonable to establish stable operating conditions, install the test equipment, perform the test, and analyze the results. The surveillance shall be performed within 7 days after reaching 90% of RATED POWER.

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l Amendment Nos. 270 and 269

TABLE 4.1-2A(CONTlNUED)

MINIMUM FREQUENCY FOR EQUIPMENT mSTS UFSAR SECTION DESCRIPTION TEST FREQUENCY REFERENCE

19. Primary Coolant Functional 1. Periodic leakage testing(a)(b) on each valve System listed in Specification 3.l.C.S.a shall be accomplished prior to entering POWER OPERATION after every time the plant is placed in COLD SHUTDOWN for refueling, after each time the plant is placed in COLD SHUTDOWN for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the preceding 9 months, and prior to returning the valve to service after maintenance, repair or replacement work is performed.

20. ContainmelJt Purge Functional Semi-Annual (Unit at power or shutdown) if MOV Leakage purge valves are operated during interval(c)

21. Deleted

22. RCS Flow Flow ~ 273,000 gpm and 2: the limit as Once per 18 months (d) 14 specified in the CORE OPERATING LIMITS REPORT

23. Deleted (a) To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and sllPported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

(b) Minimum differential test pressure shall not be below 150 psid.

S (c) Refer to Section 4.4 for acceptance criteria.

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l (d) Not required to be performed until 7 days after 2: 90% RATED POWER.

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TS 6.2-1 6.2 GENERAL NOTIFICATION AND REPORTING REQUIREMENTS Specification A. The following action shall be taken for Reportable Events:

A report shall be submitted pursuant to the requirements of Section 50.73 to 10 CFR.

B. Immediate notifications shall be made in accordance with Section 50.72 to 10

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CFR.

C. .CORE OPERATING LIMITS REPORT Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.

Parameter limits for the following Technical Specifications are defined in the CORE OPERATING LIMITS REPORT:

• 1. TS 3.1.E - Moderator Temperature Coefficient

2. TS J.12.A.2 and TS 3.12.A.3 -Control Bank Insertion Limits

3. TS 3;12.B.l and TS 3.12.B.2 - Power Distribution Limits

4. TS 3.12.F - DNB Parameters

5. TS 2.1 - Safety Limit, Reactor Core

6. TS 2.3.A.~.d - Overtemperature ~T

7. TS 2.3.A.2.e - Overpower ~T

8. TS Table 4.l-2A - Minimum Frequency' for Equipment Tests: Item 22 - RCS Flow

.J Amendment Nos. 270 and 269

TS 6.2-2 The analytical methods used to detennine the core operating limits identified above shall be those previously reviewed and approved bythe NRC, and identified

. r below.

The CORE OPERATING LIMITS REPORT will contain the complete identification for each of the TS referenced topical reports used to prepare the CORE OPERATING LIMITS REPORT (i.e., report number, title, revision, date, ~nd any supplements). The core operating limits shall be determined so that applicable limits *(e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis .are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided for information for each

.reload cycle to the NRC Do'cument Control Desk with copies to the Regional Administrator and Residentlnspector.

REFERENCES

1. VEP-FRD-42-A, "Reload Nuclear Design Methodology"

2. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM),"

(Westinghouse Proprietary).

,3. WCAP-lOO54-P-A, "Westinghouse Small Break ECCS",Evaluation Model Using the NOTRUMP Code," (W Proprietary)

4. WCAP-lO079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code;" (W Proprietary)

5. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Report," (Westinghouse Proprietary) .

6. VEP-NE'::2-A, "Statistical DNBR Evaluation Methodology"

7. VEP-NE-3-A, "Qualification of the WRB-I CHF Correlation in the Virginia Power COBRA Code"

8. DOM-NAF-2-A, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," including Appendix B, "Qualification of the Westinghouse WRB-I CHF Correlation -in the Dominion VIPRE-D Computer Code."

I' 9.. WCAP-8745-P-A, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function."

Amendment Nos. 270 and 269

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1-') ****... ~o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-32 AND AMENDMENT NO. 269 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 October 16,2009, as supplemented by letter dated May 7,2010, Virginia Electric Power Company (VEP), the licensee for Surry Power Station, Units 1 and 2, requested to amend facility operating license numbers DPR-32 and DPR-37 as necessary to implement thermal hydraulic methodologies as required for analysis of Surry Improved Fuel (SIF) with intermediate flow mixer (IFM) grids. Specifically, the licensee requested the following licensing basis changes:

(1) Implementation ofWCAP-8745-P-A\ "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function,"

(2) Implementation of NRC-approved Dominion Fleet Report DOM-NAF-2-A2 ,

"Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code,"

(3) Implementation of a Statistical Design Limit for the analytic code and critical heat flux correlations that are being used in DOM-NAF-2-A, and (4) Implementation of Dominion Topical Report VEP-NE-2-A, "Statistical DNBR Evaluation Methodology."

The requested change also affects the facility Technical Specification (TS). Items 1 and 2 in the above list are methodologies that are used in the determination of core operating limits; hence, an amendment to the TS was also requested to insert the items into the references list contained in TS 6.2.C, "CORE OPERATING LIMITS REPORT." Although VEP-NE-2-A is already included in TS 6.2.C, the licensee determined that NRC approval was required to apply this methodology to the SIF in concert with the requirements of section 59, part 50 to Title 10 of the Code of Federal RegUlations (10 CFR 50.59). The supplement dated May 7,2010, provided clarifying information that did not change the scope of the original application and the initial proposed no significant hazards consideration determination.

1 Ellenberger, S. L., et aI., Westinghouse Electric Company, "Design Bases for Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Function," WCAP-8745-P-A, September, 1983. ML073521507.

2 Brackmann, R. S., Dominion, "Reactor Core Thermal-Hydraulics Using the VIPRE-D Computer Code," DOM-NAF-2, Revision 0.1-A, with Appendices A, B, and C, July, 2009. ML092190894.

-2 The subject licensing action request also seeks to broaden the scope of the Core Operating Limits Report by relocating parameters referenced by safety limits, limiting safety system settings, departure from nucleate boiling parameters, and equipment test intervals to the core operating limits report.

2.0 REGULATORY EVALUATION

The Nuclear Regulatory Commission (NRC) staff's review of this license amendment request (LAR) is based on the conditions and limitations found in each of the licensing topical reports (TRs) requested for implementation. Proper implementation of the licensing topical reports provides the NRC staff with reasonable assurance that the licensee has determined, and will continue to determine on a cycle-specific basis, acceptable operating limits with respect to the amount of margin on the departure from nucleate boiling ratio (DNBR) to allow for normal operation and anticipated operational occurrences (AOOs). In this sense, the margin is evaluated against the fuel cladding integrity safety limit (FCISL) contained in TS 2.1.A.1.a.

A demonstration that adequate margin to the FCISL is maintained under conditions of normal operation and AOOs assures the NRC staff that the implementation of the requested methodologies at Surry 1/2 will maintain adherence to Updated Final Safety Analysis (UFSAR)

Criterion 1.4.6, "Reactor Core Design," which states:

The reactor core with its related controls and protection systems is designed to function throughout its design lifetime without exceeding acceptable fuel damage limits that have been stipulated and justified. The core and related auxiliary system designs provide this integrity under all expected conditions of normal operation with appropriate margins for uncertainties and for specified transient situations that can be anticipated.

3.0 TECHNICAL EVALUATION

Specific changes requested by the licensee have been excerpted from the LAR:

1. TS 2.1, Safety Limit, Reactor Core

a. Revise TS 2.1.A.1 to refer to the CORE OPERATING LIMITS REPORT instead of TS Figure 2.1-1, and to add the statement "and the following Safety Limits shall not be exceeded."

b. Revise TS 2.1.A.1 to add TS 2.1.A.1.a and TS 2.1.A.1.b, which support relocating the Reactor Core Safety Limits to the CORE OPERATING LIMITS REPORT.

c. Revise TS 2.1.B to refer to the CORE OPERATING LIMITS REPORT instead of TS Figures 2.1-1, 2.1-2 or 2.1-3.

d. Delete TS Figure 2.1-1 since this change relocates TS Figure 2.1-1 to the CORE OPERATING LIMITS REPORT.

e. Delete obsolete TS 2.1.A.2 and TS 2.1.A.3, and TS Figures 2.1-2, 2.1-3, and 2.1-4.

f. Revise TS 2.1 Basis to reflect TS changes and delete obsolete information.

-3

2. TS 2.3, Limiting Safety System Settings, Protective Instrumentation

a. Revise TS 2.3.A.2.d and 2.3.a.2.e to make the format consistent with NUREG-1431, Rev. 3 and relocate the cycle-specific parameters to the CORE OPERATING LIMITS REPORT. .

b. Delete TS Figure 2.3-1.

c. Revise TS 2.3 Basis to reflect TS changes.

3. TS 3.12.F, Departure from Nucleate Boling (DNB) Parameters

a. Revise TS 3.12.F.1 to remove numerical limits for Reactor Coolant System (RCS)

Tavg and pressurizer pressure and replace with reference to CORE OPERATING LIMITS REPORT.

b. Revise TS 3.12.F.1 to include "and ~ the limit specified in the CORE OPERATING LIMITS REPORT," in addition to the current limit listed for RCS Total Flow Rate.

c. Revise TS 3.12.F.1.a to add RCS Total Flow Rate to list of parameters that are to be verified to be within their limits once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

d. Revise TS 3.12.F.1.b to clarify that RCS Total Flow Rate is determined to be within its limit by precision heat balance with the frequency specified in TS Table 4.1-2A.

e. Revise TS 3.12.F.2 to change the time requirement to reduce THERMAL POWER to make it consistent with NUREG-1431, Rev. 3.

f. Revise TS 3.12 Basis to reflect TS requirements.

4. Table 4.1-2A, Minimum Frequency for Equipment Tests

a. Revise Item 22 to include "and ~ the limit as specified in the CORE OPERATING LIMITS REPORT."

b. Revise Item 22 to include a note that allows entry into POWER OPERATION, without having performed the Surveillance Requirement (SR), and placement of the unit in the best condition for performing the SR.

c. Revise TS 4.1 Basis to reflect TS requirements.

5. TS 6.2.C, Core Operating Limits Report

a. Revise TS 6.2.C to include the following TS Sections:

i. TS 3.12.F - DNB Parameters, TS 2.1.A.1 - Safety Limit, Reactor Core, ii. TS 2.3.A.2.d - Overtemperature llT, TS 2.3.A.2.e - Overpower llT, and iii. TS Table 4.1-2A Minimum Frequency for Equipment Tests: Item 22 RCS Flow.

b. Revise TS 6.2.C to include references to DOM-NAF-2-A, including Appendix B, and WCAP-8745-P-A. The COLR references have been renumbered.

-4 3.1 TS 2.1, Safety Limit, Reactor Core The licensee references TSTF Traveler 339-A3 and WCAP-14483-A 4 as the basis for requested changes to TS 2.1. The TSTF Traveler proposes the relocation of TS parameter to the COLR consistent with WCAP-14483-A.

The requested changes will relocate the Reactor Core Safety Limits figure to the COLR, add new Safety Limits for departure from nucleate boiling ratio and peak fuel centerline temperature, and delete obsolete limits associated with n-1 Loop operation, since two-loop operation is prohibited for Surry, and fuel densification.

The limit for the departure from nuclear boiling ratio (DNBR) is different from that discussed in WCAP-14483-A, in that it specifies limits on the DNBR that are specific to the VEP methodology.

This includes a design limit for statistically analyzed transients, and a correlation limit for deterministically analyzed transients.

This deviation is acceptable because the generic limit is based solely on the correlation limit, and the design limit is typically controlled by the licensee. Because the design limit is specified in the TS, it is, in a sense, more restrictive than specification of the correlation limit.

N-1 loop restrictions have been removed. The licensee stated that Surry TS 3.1.A.1.a proscribes operation with other than three reactor coolant loops in service, and hence the specification applies only to three loop operation. Therefore, the NRC staff finds that the proposed specification differs from the standard language in the Improved Technical Specifications (ITS),

but is appropriate for the allowed operational configuration at Surry.

The NRC staff finds that the remaining changes to TS 2.1 are consistent with the ITS and with WCAP-14483-A. The licensee will be using NRC-approved methodology to determine the cycle-specific parameter operating limits that will be specified in the COLR, which provides adequate assurance that cycle-specific limits on the DNBR will provide appropriate margin to account for uncertainties and specified transient situations that can be anticipated.

The approved methodology for determining cycle-specific DNBR limits referenced by the licensee will be VEP-NE-2-A and DOM-NAF-2-A. VEP-NE-2-A currently appears in the Surry Core Operating Limits Report - References section of the TS, and the NRC staff finds the licensee's addition of DOM-NAF-2-A acceptable herewith. Although the safety evaluation approving WCAP-14483-A states that the DN B lim its relocation is acceptable based on use of the Westinghouse-supplied, NRC-approved reload method, it also specifies that use of other approved reload licensing methods is acceptable. Because the licensee references an NRC-approved methodology to determine cycle-specific parameter operating limits, the NRC staff finds the requested modification acceptable.

3 Technical Specification Task Force Traveler, 339 Revision 2 , "Relocate TS Parameters to COLR" FSTF-339-A, Rev. 2), May 26, 2000, ML003723269, approved by NRC July 6, 2000, ML003730788 Westinghouse Electric Company, "Generic Methodology for Expanded Core Operating Limits Report,"

WCAP-14483-A, January, 1999. ML020430092.

-5 3.2 TS 2.3, Limiting Safety System Settings, Protective Instrumentation The proposed change to TS 2.3 will relocate the Overtemperature Oelta-T (OTLiT) and Overpower Oelta-T (OPLiT) setpoint parameters, including RCS average temperature (Tavg ),

nominal RCS operating pressure, K values, dynamic compensation time constants, and the breakpoint and slope values for the f(LiI) penalty function from TS 2.3.A.2.d and TS 2.3.A.2.e to the COLR. According to WCAP-14483-A, f(LiI) is a modification applied to the OP and OTLiT trip functions to provide conservative values for adverse axial power distributions. Figure 2.3-1 will be deleted.

This revision is consistent with Notes 1 and 2 of Table 3.3.1-1 of the Standard Technical Specifications contained in TSTF Traveler 339-A. The licensee has modified the OTLiT and OPLiT equations to match those of Surry; since the OTLiT equation in use at Surry is plant-specific, the NRC staff finds this difference acceptable. Also, YEP has elected to correct an error where the current Surry TS refer to f(LiI) as a percentage of rated core power, and in actuality the f(LiI) parameter is dimensionless. This is viewed by the NRC staff as an administrative correction to the TS and is hence acceptable.

OTLiT and OPLiT setpoints may be determined using cycle-specific parameter inputs consistent with the methodology described in WCAP-8745-P-A, and the licensee is adding a reference to WCAP-8745-P-A to the Surry TS. This is consistent with the approach delineated in both WCAP-14483-A and TSTF Traveler 339-A. Because of this consistency, the NRC staff finds the requested changes to TS 2.3 acceptable.

3.3 TS 3.12.F, ONB Parameters The proposed changes to TS 3.12.F, "ONB Parameters," will relocate the limits for RCS T avg ,

pressurizer pressure, and RCS total flow rate to the COLR. The proposed changes will also add the requirement to verify that RCS flow is within its limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and modify the requirement to perform a precision RCS flow measurement with the frequency defined in TS Table 4.1-2A.

The NRC staff verified that, while Surry does not contain the same format as the Standard Technical Specifications (STS) and TSTF, the requested change will align the Surry TS with the functional specifications contained in the STS and TSTF Traveler 339-A. Limits on pressurizer pressure, RCS Tavg , and RCS total flow rate will be subject to the same requirements, and relocated to the COLR. For the Surry TS, this will result in the addition of a new requirement to verify that the RCS flow is within its limits at least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, which is a more restrictive change. This will also result in an increase in the Completion Time to reduce core thermal power if any RCS parameter exceeds its limit and cannot be restored within two hours from four hours to six hours. This change is less restrictive, although it is consistent with ITS, and allows for an acceptable time to reduce core power in an orderly manner without challenging plant systems.

The relocation of these parameter limits to the COLR is consistent with both NRC-approved WCAP-14483-P-A and TSTF Traveler 339-A. Because of this consistency and because the parameter limits will be set or confirmed on a cycle-specific basis in accordance with NRC-approved methodology, the NRC staff finds the requested parameter limit relocation acceptable.

- 6 3.4 TS 4.1-2A: Minimum Frequency for Equipment Tests The proposed change to TS 4.1, "Operational Safety Review," modifies Item 22 in Surry TS Table 4.1-2A by adding "and ~ the limit as specified in the CORE OPERATI NG LIMITS REPORT" to the flow requirement, and adding a note that allows entry into POWER OPERATION, without having performed the surveillance requirement and placement of the unit in best condition for performing the SR. The licensee stated that the first addition is for consistency with the changes proposed for TS 3.12.F. The note states that the SR is, "Not required to be performed until 7 days after ~ 90%

RATED POWER." The addition of the note means that the precision flow measurement is required to be made within 7 days after reaching 90% RATED POWER.

The requested TS revision is largely consistent with the generic STS and TSTF models. For instance, the power level is consistent with TSTF Traveler 339-A, Revision 2, and WCAP-14483-A. The time limit is inconsistent with TSTF Traveler 339-A, however, in that the traveler specifies a time limit of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and VEP has requested to increase the time limit to seven days.

The licensee justified the completion time extension by stating that seven days will allow the establishment of stable operating conditions, installation of test equipment, performance of the test, and analysis of the results. However, the generic TS Bases contained in NUREG-1431, Revision 3, does not contain this consideration. The bases state only that the exception allowing entry into MODE 1, power operation, without having performed the SR to place the unit in the best condition for performing the surveillance requirement, since the surveillance requires the plant to be at a certain minimum RTP to obtain the stated RCS flow accuracies.

The licensee also stated that the requested change was appropriate since it was less than the 30 day completion time that is incorporated in the North Anna TS. The NRC reviewed the safety evaluation approving the incorporation of a 30 day completion time at North Anna 5 . The NRC staff concluded that the basis for a 30 day completion time was not explained in sufficient detail to warrant its use as a basis for a seven day completion time at Surry, and requested additional information (RAI) dated May 7,2010 6 .

In the response to the RAI dated May 25, 20107 , the licensee stated that current TS 3.12.F.1.b does not specify a time for completing this surveillance. The licensee stated that the 7-day LCO will allow for escalation to a steady-state, 1DO-percent full power level prior to completing the surveillance. Conducting the surveillance at a higher power level is conducive to a higher-accuracy RCS flow rate measurement, and the evolution period from 90- to 1DO-percent power requires 7 days to allow the core to reach eqUilibrium xenon and samarium concentrations, put equipment in place to perform the measurement, and analyze and record data. An accurate flow rate measurement is desirable because it better establishes the RCS flow rate throughout the operating cycle, which increases the quality of confirmation available to the licensee that it is operating its facility within safety analysis limits. Because the selected LCO supports a more 5 Monarque, S., USNRC, letter to D. A. Christian, VEP, "North Anna Power Station, Units 1 and 2 Issuance of Amendments Regarding Conversion to Improved Technical Specifications (TAC Nos. MB0799 and MB0800)," Dockets 50-338/50-339, April 5, 2002. ML021200265.

6 NRC, Request for Additional Information, e-mail dated May 7,2010 ML102720368 7VEPCO, Response to NRC RAI dated May 25,2010 ML101470905

-7 accurate flow rate measurement, the staff finds the selected completion time acceptable.

3.5 TS 6.2.C: Core Operating Limits Report Proposed changes to TS 6.2.C, "Core Operating Limits Report," will add Fleet Report DOM-NAF-2-A including Appendix B, and Westinghouse Licensing Topical Report WCAP-8745-P-A to the list of NRC-approved methodologies used to determine core operating limits.

WCAP-8745-P-A describes the generic method for determining OTilT and OPilT setpoints and provides a technical basis for determining the input parameters used to establish those setpoints.

It served as the generic bases for these reactor trips, and it is being included in the TS COLR References as an associated requirement of relocating the setpoint parameters to the COLR.

This TS revision is viewed by the NRC as an administrative change required to relocate the setpoint parameters to the COLR and is hence acceptable.

Fleet Report DOM-NAF-2-A and Appendix B are being newly implemented at Dominion, and NRC approval is also requested for a change in the way that current COLR reference VEP-NE-2-A is implemented. The implementation of these new methods is the subject of further NRC staff evaluation, as described in the following subsections. A table is provided to summarize the use of the methods as they apply to each fuel type that exists or will exist in the Surry core:

Fuel Product Surry Improved Fuel (Current) Surry Upgrade Fuel (Proposed)

Statistical DNB Analytic VEP-NE-2-A VEP-NE-2-A Method Deterministic DNB Analytic DOM-NAF-2-A DOM-NAF-2-A Method DNB Correlations WRB-1/W-3 WRB-1/W-3 Thermalhydraulic Code COBRA VIPRE-D Statistical Desiqn Limit DNBR 1.27 1.27 Deterministic Correlation 1.17 -WRB-1 1.17 - WRB-1 Limit DNBR 1.30 - W-3 1.30 - W-3 In order to establish that the addition of DOM-NAF-2-A to the Surry TS COLR References would not result in the inclusion of outdated, unused methodology in the existing TS COLR References, the NRC staff requested VEP to conduct a review of all references currently included in Surry TS COLR References and confirm that all remaining references remain in use after implementation of the new methods. The licensee responded by letter dated May 25, 2010, stating that this review had been completed, and that all included references remained current and in-use. The NRC staff finds, therefore, that the inclusion of DOM-NAF-2-A would not result in the inclusion of unnecessary references in the Surry TS COLR References section.

The licensee provided a technical justification for analyzing the Surry Upgrade Fuel product using NRC-approved methods VEP-NE-2-A and DOM-NAF-2-A. This information is required because both methodologies require confirmation of their applicability to new fuel designs prior to use. The NRC staff reviewed the information as discussed in the following subsections.

-8 3.5.1 Implementation of DOM-NAF-2-A for the Westinghouse 15x15 Surry Upgrade Fuel Product Section 2.1 of DOM-NAF-2-A listed information to be provided to the USNRC for review and approval of any plant-specific application of the VIPRE-D code:

1. Technical Specifications change request to add DOM-NAF-2-A and relevant appendices to the plant's COLR list.

2. Statistical Design Limits for the relevant code/correlations.

3. Any TS changes related to OT~T, OP~T, F~I or other protection function, as well as revised Reactor Core Safety Limits.

4. List of Updated Final Safety Analysis Report (UFSAR) transients for which the code/correlations will be applied.

VEP provided this information, relevant to the Surry Upgrade Fuel product, in Attachment 4 to its license amendment request. Items 1, 2, and 3 of the above list are discussed in preceding sections of this Safety Evaluation Report; Item 4 was provided in Table 3.9-1 of Attachment 4 to the license amendment request.

The NRC staff reviewed the table of UFSAR transients analyzed with VIPRE-D and compared it to the table of UFSAR transients contained in the NRC staff's safety evaluation approving DOM-NAF-2-A. The tables were largely consistent, except as noted in the paragraph below regarding the loss of all alternating current power to the station auxiliaries transient. The NRC staff finds that Item 4 has been provided and the information is acceptable for Surry, because the Surry-specific list of analyzed transients mirrors the NRC staff's generic SER approving DOM-NAF-2-A.

The licensee stated that the DOM-NAF-2-A methodology will be used to analyze the loss of all alternating current power to the station auxiliaries transient. This transient is not included in the table of UFSAR transients in the NRC staff's safety evaluation approving DOM-NAF-2-A, however, it is acceptable for analysis using the DOM-NAF-2-A methodology because the transient is included in the loss of normal feedwater flow transient, one case of which includes the loss of reactor coolant pumps. Similar conditions - no normal feedwater flow accompanied by a reactor coolant pump trip - would characterize the loss of all alternating power to the station auxiliaries transient and hence its analysis IJsing DOM-NAF-2-A is acceptable.

Based on the considerations described above, the NRC staff concludes that VEP has provided sufficient and acceptable information to implement DOM-NAF-2-A at Surry Power Station.

3.5.2 Implementation of VEP-NE-2-A for the Westinghouse 15x15 Surry Upgrade Fuel Product The safety evaluation approving the use of VEP-NE-2-A contained four limitations:

1. The selection and justification of the Nominal Statepoints used to perform the plant specific implementation must be included in the submittal.

-9

2. Justification of the distribution, mean, and standard deviation for all the statistically treated parameters must be included in the submittal.

3. Justification of the value of model uncertainty must be included in the plant-specific submittal.

4. For the relevant CHF correlations, justification of the 95/95 DNBR limit and the normality of the M/P distribution, its mean and standard deviation must be included in the submittal, unless there is an approved Topical Report documenting these.

To satisfy the condition on nominal statepoints, the choice of the nominal statepoints must be shown to maximize the DNBR standard deviation (and therefore, the DNBR limit) over the proposed range of applicability. The licensee's approach was to perform Monte Carlo analysis of nine statepoints covering the full range of normal operation and anticipated transient conditions.

The Monte Carlo analysis consisted of 2000 calculations performed around each of the nine statepoints. The DNBR standard deviation was augmented by code/correlation uncertainty, the sample correction factor, and the code uncertainty to obtain the total DNBR standard deviation.

The nine statepoints spanned the four reactor core safety limit pressures, power levels up to 118%, and included a loss of flow statepoint. The NRC staff finds that the licensee has, by consideration of these nine statepoints, satisfied Condition 1 above.

The statistically treated parameters are core power, pressurizer pressure, inlet temperature, vessel mass flow, core bypass flow and the nuclear and engineering enthalpy rise factors. The licensee established uncertainties for each of these parameters using a rigorous analysis of plant hardware and measuremenUcalibration procedures. The licensee's discussion identified significant attributes of the uncertainties, characterized the distributions, and compared current values to those used previously. The NRC staff finds that the description of the analyses provided by the licensee is sufficiently comprehensive that Condition 2, above, is satisfied.

The licensee stated that the VIPRE-D 19-channel production model for Surry 15x15 Upgrade fuel was used in the development of the VIPRE-DIWRB-1 code/correlation pair statistical design limit for Surry. In other words, the statistical design limit on the departure from nucleate boiling ratio already incorporates VI PRE uncertainties, and the uncertainty need not be included in the calculation of the DNBR standard deviation. The NRC staff finds that Condition 3, above, is satisfied because the model uncertainty has been appropriately justified.

The NRC staff finds that Condition 4 is satisfied for two reasons: (1) Fleet report DOM-NAF-2-A and its appendices have been reviewed and approved by the NRC staff and Appendix B documents the calculation of a deterministic design limit for the VIPRE-DIWRB-1 code/correlation pair; and (2) Remaining information provided by the licensee, particularly discussions regarding the detailed uncertainty evaluations and Monte Carlo analyses, justified the chosen statistical design limit of 1.27.

The licensee stated that VEP-NE-2-A has already been implemented at Surry; hence the NRC staff's evaluation of the acceptability of VEP-NE-2-A associated with this licensing action request is limited to the applicability of the methodology to the 15x15 Upgrade fuel product.

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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 and change surveillance requirements. 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 (74 FR 62838). The amendment also relates to changes in recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the amendments meet the eligibility criteria for categorical exclusions set forth in 10 CFR 51.22(c)(9) and (c)(10). 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.

6.0 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 Contributor: B. Parks Date: Q:tober 19, 2010

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, 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. 270 to DPR-32

2. Amendment No. 269 to DPR-37

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

Public RidsNrrLAMO'Brien (hard copy) RidsNrrDirsltsb (TKobetz)

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DATE 10/13/10 10/18/10 06/25/10 9/30/10 10/19/10 10/19/10 OFFICIAL RECORD COpy