ML20118B132

Amends 166 & 146 to Licenses NPF-4 & NPF-7,respectively, Revising Wording in TS for RCS Vol in Design Features Section

ML20118B132
Person / Time
Site:	North Anna
Issue date:	09/21/1992
From:	Berkow H Office of Nuclear Reactor Regulation
To:
Shared Package
ML20118B133	List: ML20118B132
References
NUDOCS 9209300191
Download: ML20118B132 (14)

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UNITED STATES i

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NUCLEAR REGULATORY COMMISSION

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VIRGINIA _[LECTRIC AND POWER COMPANY DMLD2 BION ELECTRIC COOPERATIVE DQ1KET NO. 50-338 NORTH ANNA POWER STATION. UNIT NO. 1 AMENDMENT _TO FACILITY OPERATING tICENSE Amendment No. 166 License No. NPf-4 1.

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

A.

The application for amendment by Virginia Electric and Power Company et al., (the licensee) dated July 16, 1992, complies with the i

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 Ch pter I; The facility will operate in conformity with the application, the B.

provision of the Act, and the rules and regulations of the Commission;

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C.

There is reasonable assurance (i) that the activities authorized by l

this amendme'* can be conducted without endangering the health and safety of the public, and (ii) that such activities will be j

conducted in compliance with the Commission's regulations; D.

The isruance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i

E.

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

been satisfied.

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PDR ADOCK 05000338 P

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Accordingly, the license is amended by changes to the Technical Speci-fications a. indicated in the attachment to this license amendment, anJ j

paragraph 2.0.(2) of facility Operating License No. NPf A is hereby amended to read as follows:

(2) Itchnical Setsifications

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The Technical Specifications contained in Appendices A and B. as j

revised through Amendment No.166, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

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FORTHENUCLEARREGULAf=cyCOMMISSION 4}4

.YW lerbert N. Berkow, Director l

Proj.ct Directorate 11-2 l

Division of Reactor Projects - I/II j

Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical j

Specifications l

Date of Issuance: September 21, 1992 i

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'Q FALlillY OPERAlIt10_11[J1d(,J1Q. tiPF-4 DOCKET tio. 50-338 Replace the following pages of the Appendix "A" Technical Specificattuns with the enclosed pages as indicated.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

The corrvsponding overleaf pages are also provided to maintain document completeness.

flerpye Pagei Insert Paan B 3/; l-1 B 3/4 1-1 i

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____._____._____.__.______.._._y T4.1 REACTMTY CONTROL SYSTEMS B A S E__S 3!4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life hs a function of fuel depletion, RCS boron concentration, and RCS Tavg. The most restrictive condition occurs at EOL, with Tavy at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown, in the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.77% Ak/k is inltlally required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With Tavg

< 200*F, the reactivity transients resulting from a pcstulated steam line break cooldown are minimal. A 1.77% AA/k shutdown rnargin provides adequate protection for the boron dilution accident.

3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that - reactivity changes will be gradual during boron concentration reductions in the Reactor Coolant System. A flow rate of at least 3000 CPM l

will circulate the Reactor Coolant System volume in approximately 30 minutes. The reactivity chante rhte associated with boron reductions will therefore be within the capability for operator rocognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIDIT (MTC)

The limitations on MTC are proGded to ensure that the value of this coefficient remains withiri the limiting conditions assumed for this parameter in the FSAR accident and transient analyses.

The MTC values of this specification are applicable to a specific set of plant operations; accordingly, verification of MTC values at NORTH ANNA UNIT 1 B 3/4 1 1 Amendment No.166,

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311 RE ACTIVITY CONTROL SYSTEMS MSES 3/4.1 1 4 MODERATOR TEMPERATURE COEFFICIENJ_rMTC) (Continued) conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

The inost negative MTC value was obtained by incrementally correcting the MTC used in the FSAR analyses to nominal operating conditions. These corrections involved adding the incroitiental change in the MTC associated with a core condition of Bank D inserted to an all rods withdrawn condition and an incremental change in MTC to accour.1 for measurement uncertainty at RATED THERMAL POWER conditions. These corrections result in the End of Cycle (EOC) MTC 1

limit. The 300 ppm surveillance limit MTC value represents a conservative value (with j

corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron j

concentration and is obtained by making these octrections to the EOC MTC limit.

Once the equilibrium boron concentration falls be!ow about 60 ppm, dilution operations take an extended amount of time and rehable MTC measurements become more difficult to obtain 4

due to the potential for fluctuating core conditions over the test interval. For this reason, MTC measurements may be suspended provided Ine measured MTC value at an equilibrium full power i

boron concentration s 60 ppm is less negative than the 60 ppm surveillance limit. The j

difference between this value and the EOC MTC tmit conservatively bounds the maximum i

credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POV.ER ccnditions) and the licensed end of cycle, inciuding the effect of boron concentration, burnup, and end of cycle coastdown.

The surveillance requirements for measurement of the MTC at the beginning and near the j

end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slow!y due principally to the reduction in RCS boron concentration l

associated with fuel burnup.

2 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICAllTY This specification ensures that the reactor will not be made critical with the Reactor l

Coolant Syttem average temperature less than 541'F. This limitation is required to ensure 1) i the moderator temperature coefficient is within its analyzed temperature range,2) the protective instrumentation is withia its normal operating range, and 3) the P 12 interlock is above its setpoint, and 4) compliance with Appendix G to 10 CFR Part 50 (see Bases 3/4.4,9),

f 3/4.1.2 BORATION SYSTEMS I

The boron injection system ensures that negative reactivity control is avaltable during each mode of facility operation. The components required to perform this function include 1) borated water sources,2) charging pumps,3) separate flow paths,4) boric add transfer.

j pumps,5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

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i NORTH ANPA UNIT 1 B 3/4 12 Amendment No. M,111,177.146,

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05S@.FEATlf3ES

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650'F, except for the pressurizer which is 680'F.

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j 5.4.2 The total water and sicam volume of the reactor coolant system is approximately 10.000 cubic feet at nominal operating conditions.

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5.5 METEOROLCGCALTOWER LOCATION i

j 5.5.1 The meteorological tower shall be located as shown on Figure 5.11.

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5.6 FUEL STORAGE i

l CRITICALITY 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

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a. A Keff equivalent to less than or equal to 0.95 when flooded with unborated i

water, which includes a conservat!ve allowance of 3.4% delta k/k for l

uncertainties, I

b. A nominal 10 9/16 inch center to center distance between fuel assemblies placed in the storage racks.

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5.6.1.2 The new fuel pit storage racks are designed and shall be maintained with a nominal 21 inch center to center distance between new fuel assernblies such that, on a j

best estimate basis, Kett will not exceed.98, with fuel of the higt.est anticipatt.d enrichment in place, when aqueot,s loain moderation is assumed.

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5.6.1.3 If new fuel for the first core loading is stored dry in the spent fuel storage racks, the centel to center distance between the new fuel assemblies will be j

administratJvely limited to 28 inches and the kett shall not exceed 0.98 when aqueous foam inoderation is assumed.

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NORTH ANNA UNIT 1 55 Amendment No. U,27.E7, 4

166, 4

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DESIGN FEATURES DRAINAGE l

5.6.2 The spent fuel pit is designed and shall be traintained to prevent inadvertent draining of the pool below elevation 288.83 feet.

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Level U5b5 datum.

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CApACI.TY i

5.6.3 The fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1737 fuel assemblies.

l 5.7 COMPONENT CYCLIC OR TRAfd,;ENT LIMIT a

j 5.7.1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

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N0RTH ANNA - UNIT 1 5-6 Amendment No. 14.61

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i UNITED STATES n

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!.'",i NUCLEAR REGULATORY COMMISSION 1

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WASHINoTON O C. FJM4 i

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j VIRGINIA ELECTRIC AND POWER COMPAfq QLD DOMINION ELECTRIC COOPERATIVE QQClET NO 50-339 i

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NORTH ANNA POWER STATION. UNIT NO. 2 AMENDMENT TO FAClllTY OPERATING Ll(R{SL I

Amendment No. 146 l

Licenso No. NPF-7 l

1.

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

A.

The application for amendment by Virginia Electric and Power Company et al., (the lict.ee) dated July 16, 1992, complies with the standards and re,,irements 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 ccaformity with the application, the i

i 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 l

conducted in compliance with the Commissicn's regulations; l

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 2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of facility Operating License No. NPf-7 is hereby amended to read as follows:

(2)

Technical Specificatir,n1 The Technical Specifications contained in Appendices A and B, as revised through Amendment No.146

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

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION i

ierbert N. Berkow, Director l

Project Directorate 11-2 Division of Reactor Projects - i/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical 4

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Specifications Date of issuance: September 21, 1992 i

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ATTACHMEllT 10 LL(MSE AMENDMENT N0.

TO FACillTY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-332 i

Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated.

The revised pages are identif',ed by i

aniendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

Remove Page.1 Insert Praes B 3/4 1-1 8 3/4 1-1 5-4 5-4 t

l 3'4.1 REACTMTY CONTROL SYSTEMS l

BASES 3/4.1.1 BORATION CONTROL 3!411 1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) thq re3ctivity transients associated with postulated accident conditions are controllable within acceptable !imits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg. The most restrict!ve condition occurs at EOL, with Tavg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown, in the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.77% ak/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With Tavg less than 200'F, the reactivity transients resulting from a postulated steam line break cooldown are minimal.

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3/4.1.1.3 BORON DILUTION A m!nimum flow rate of at least 3000 GPM as provided by either one RCP or one RHR pump as required by Specification 3.4.1.1, provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration reductions its the Reactor Coolant System. A flow rate of at 16ast 3000 GPM will circulate the Reactor Coolant System volume in approximately 20 minutes. The l

reactivity change rate associated with boron reductions will therefore be within tha capability for operator recognition and control. The requirement that certsin vnhes remain closed at cll times axcept during planned boron dilution or makeup, activitics provides assurance that an inadvertent boron dilution will not occur.

1 NORTH ANNA UNIT 2 B 3/4 11 Amendment No.146,

l 3'4.1 REACTIVITY CONT ROL 3YSl t-Mb BASES 3/4.1.1.4 MODER ATOR TEMPERATURE COEFF!CIENT fMTC) (Continued)

The iimitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed for this parameter in the FSAR accident and transient analyses.

The MTC valui.; of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those erplicitly stated will require extrapolation to those conditions in order to permit an accurato comparison.

The most negative MTC value was obtained by incrementally correcting the MTC used in the FSAR analyses to nominal operating conditions. These corrections involved adding the incremental change in the MTC associated with a core condition of Bank D inserted to an all rods withdrawn condition and an incremontal change in M1C to account for measurement uncertainty at RATED THERMAL POWER conditions. These conectons result in the End of Cycle (EOC) MTC limit. The 300 ppm surveillance lirnit MTC value represents a coaservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equ;iibdem boron concentration and is obtained by making these corrections to the EOC MTC limit.

Once tha equilibrium boron concentration falls below about 60 ppm, dilution operations take an extended amount of time and reliable MTC measurements become more difficult to obtain due to the potential for fluctuating core conditions over the test interval. For this reason, MTC measurements may be suspended provided the measured MTC value at an equilibrium full power boron concentration s 60 ppm is less negative than the 60 ppm surveillance limit. The ditterence between this value and the EOC MTC limit conservatively bounds the maximum credible change in MTC between the 60 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER conditions) and the licensed end-of cyrle, including the effect of boron concentration, burnup, and end of cycle coastdown.

The surveillance requirements for measurement of the MTC at the begnning and near the end of each fuoi cycle are adequate to confirm that the MTC remains within its limits since this coefficient changet t-ily due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not bv made critical with the Reactor Coolant System average temp"ature less than 541*F. T:.is limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range,2) the protective instrumentation is within its normal operating range, and 3) the P 12 interlock is above its setpoint,4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is aoove its minimum RTNDT emperature.

t NORTH ANNA UNIT 2 B 3/4 12 Amendment No. 72,100, 130, 1

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MA ANNA

• WIT 2 5-3 Amendment No, 7;,53 j

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DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEVBUES 5.3.1 The reactor core shall contain 157 feel assemolies with each fus: assembly containing 264 fuel rods c; ' with Zircaloy 4. Each fuel rod shall have a nominal active fuel length of 144 inches and contain a maximum total weight of 1780 grams uranium.

The initial core loading shall have a maximum enrichment of 3.2 weight percent U 235.

Reload fuel shall be similar in physical design to the initial core lo_ ding and shall have a maximum enrichment of 4.3 weight percent U 235.

i CCNTROL ROD ASSEMBUES 5.3.2 The reactor core shall contain 48 full length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material.

The nominal valuU, of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium. AC controf rods shall be clad with stainless steel tubing.

l 5 4 REACTOR COOLANT SYSTEl DES!GN PRESSURE AND TEMPERATURE l

The reactor coolant system is designed and shall be maintained:

5.4.1 l

a. In accordance with the code requirements specified in Section 5.2 of the i

l FSAR, with allowance for normat degradation pursuant to the applicable l

Surveillance Requirements,

b. For a pressure of 2485 psig, and

c. For a temperature of 650'F. except for the pressurtzer which is 680'F.

E'll.fE l

l 5.4.2 The total water and steam volume of the reactor coolant system is approximately 10,000 cubic feet at nominal operating conoitions.

NOR1H ANNA UNIT 2

!; 4 Amendment No. 8,7f>,777, 146,