ML20033E235

Amends 29 & 10 to Licenses NPF-68 & NPF-81,respectively, Revising Rod Insertion Limits to Allow Withdrawn Range of 222 Steps to 231 Steps

ML20033E235
Person / Time
Site:	Vogtle
Issue date:	02/20/1990
From:	Matthews D Office of Nuclear Reactor Regulation
To:	Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA
Shared Package
ML20033E236	List: ML20033E235 ML20033E238
References
CON-IIT05-169-90, CON-IIT5-169-90, NPF-68-A-029, NPF-81-A-010 NUDOCS 9003090453
Download: ML20033E235 (16)

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UNITO STATES NUCLEAR REGULATORY COMMISSION h

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i GEORGIAPOWERCOVPANl OGLETHORPE POWER CORPORATION l

1 HUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA

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V0GTLE ELECTRIC GENERATING PLANT, UNIT 1 AMENDMENT TO FACILITY OPERAT!NG LICENSE Amendment No. 29 License No. NPF-68 i

1.

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

A.

The application for amendment to the Vogtle Electric Generating Plant, i

-Unit 1 (the f acility), Facility Operating License No. NPF-68 filed by the Georgia Power Company, acting for itself. Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (thelicensees),datedAugust 25, 1989, complies with the standards and requirenents 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; 6

B.

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

Thereisreasonableassurance(1)thattheactivitiesauthorizedby 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 set forth in 10 CFR Chapter I; i

i D.

The issuance of this license 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-68 is hereby j

amended to read as follows:

Technical Specifications and Environmental Protection Plan f

The Technical Specifications contained in Appendix A, as revised through Amendment No. 29 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. GPC shall operate the l

facility in accordance with the Technical Specifications and the Environmental Prvtection Plan, j

3.

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

i FOR THE NUCLEAR REGULATORY COMMIS$10N i

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David B. Ha thews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance: February 20, 1990 i

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UNITED STATES

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NUCLE AR REGULATORY COMMISSION j

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c-t WASHINGTON, D. C, 70665 3

o*..,+E GEORGIA POWER COMPANY i

OGLETHORPE POWER CORPORATION l

MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA j

CITY 0F DALTON, GEORGIA i

YOGTLE ELECTRIC GENERATING PLANT, UNIT 2 AMENDMENT _TO FACILITY OPERATING LICENSE j

Amendment No.10 License No. NPF-81 1.

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

h The app (lication for amendment to the Vogtle Electric Generating Plant.

A.

Unit 2 the facility), Facility Operating License No. NPF-81 filed by the Georgia Power Company, acting for itself Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, t

Georgia (the licensees), dated August 25, 1989, 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

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forth in 10 CFR Chapter I; S.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commir,sion; i

C.

Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted l

in compliance with the Commission's regulations set forth in 10 CFR Chapter 1; D.

The issuance of this license 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 hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-81 is hereby l

amended to read as follows:

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised j

through Amendment No.10

, and the Environmental Protection Plan i.

contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. GPC shall operate the i

facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

1. 0 id B.

atthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of Issuance: February 20, 1990 Y

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i ATTACSIENT TO LICENSE ANEND4ENT NO. 29 FACILITY OPERATING LICENSE NO. NPF-68 AND LICENSE AMENDMENT NO 10 FACILITY OPERATING LICENSE NO. NPF-81 1

DOCKETS NOS. 50-424 AND 50-425 j

Replace the following pages of the Appendix "A" Technical Specifications with l

the enclosed pages. The revised pages are identified by Amenenent number and contain vertical lines indicating the areas of change.

The corresponding i

overleaf pages are also provided to maintain document completeness.

i Ananded Page Overleaf Page 3/4 1-19 3/4 1-k0 3/4 1-21 3/ 4 1-22 3/4 10-1 3/4 10-2 B 3/4 2-2 B 3/4 2-1 5-4 I

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REACTIVITY CONTROL SYSTEMS R0D OROP TIME i

i LIMITING CONDITION FOR OPERATION i

i 3.1.3.4' The individual shutdown and control rod drop time from the physical l

fully withdrawn position shall be less than or equal to 2.2 seconds from beginning of decay of stationary gripper coil voltage to. dashpot entry with:

T,y (TI-0412, TI-0422, TI-0432, TI-0442) greater than or equal to a.

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b.

All reactor coolant pumps operating.

APPLICABILITY: MODES I and 2.

ACTION:

With the drop time of any rod determined to exceed the above limit, restore the i

rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

J SURVEILLANCE REQUIREMENTS f

4.1.3.4 The rod drop time shall be demonstrated through measurement prior to reactor criticality:

a.

For all rods following each removal of the reactor vessel head, b.

for specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could I

affect the drop time of those specific rods, and c.

At least once per 18 months.

V0GTLE UNITS - 1 & 2 3/4 1-19 Amendment No29 (Unit 1)

Amendment NoJO (Unit 2)

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REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT

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LIMITING CONDITION FOR OPERATION f

i 3.1.3.5 All shutdown rods shall be withdrawn to a position greater than or equal to 222 steps.

APPLICABILITY:

MODES 1* and 2* #.

ACTION:

'With a maximum of one shutdown rod inserted to a position less than 222 steps, l

except for surveillance testing pursuant to Specification 4.1.3.1.2, within I hour either:

a.

Withdraw the rod to a position greater than or equal to 222 steps, or l

j b.

Declare the rod to be inoperable and apply Specification i

3.1.3.1.

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i SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be withdrawn to a position I

greater than er equal to 222 steps:

a.

Within 15 minutes prior to withdrawal of any rods in Control Bank A, B. C, or D during an approach to reactor criticality, and

b. -

At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

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• See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

1. With K,ff greater than or equal to 1.

V0GTLE UNITS - 1 & 2 3/4 1-20 Amendment No. 29 (Unit 1)

Amendment No.10 (Unit 2)

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, ; s, REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1. 3. 6 The control banks shall be limited in physical insertion as shown in Figure 3.1-3.

APPLICABILITY: MODES la and 2* #.

ACTION:

With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

f a.

Restore the control banks to within the limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank posi-tion using the above figure, or c.

Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1; 3. 6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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• See Special Test Exceptions Specifications 3.10.2 and 3.10.3.

1. With K,ff greater than or equal to 1.

V0GTLE UNITS - 1 & 2 3/4 1-21

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RELATIVE POWER (percent) i I

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ROD BANK li1SERTION LIMITS VERSUS THERMAL POWER i

V0GTLE UtilTS - 1 & 2 3/4 1-22 Amendment No. 29 (Unit 1)

Amendment No.10 (Unit 2) l'

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.4 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s).

APPLICABILITY:

MODE 2.

ACTION:

a.

With any control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b.

With all control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS r

4.10.1.1 The position of each control rod not fully inserted shall be l

. determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each control rod not fully inserted shall be demonstrated capable of full. insertion when tripped from at least the 50% withdrawn position within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1.

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V0GTLE UNITS - 1 & 2 3/4 10-1 Amendment No. 29 (Unit 1)

Amendment No.10 (Unit 2) l

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l' SPECIAL TEST EXCEPTIONS 3/4.10.2 GROUP HEIGHT, INSERTION. AND POWER DISTRIBUTION LIMITS f

LIMITING CONDITION FOR OPERATION l

3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:

The THERMAL POWER is maintained less than or equal to 85% of RATED l

a.

THERMAL POWER, and b.

The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.

APPLICABILITY:

MODE 1.

ACTION:

With any of the limits of Specification 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either:

Reduce THERMAL POWER sufficient to satisfy the ACTION requirements a.

of Specifications 3.2.2 and 3.2.3, or b.

Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be less than or equal to 85% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.2.2 The requirements of the below listed specifications shall be performed at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during PHYSICS TESTS:

a.

Specifications 4.2.2.2 and 4.2.2.3, and b.

Specification 4.2.3.2.

H VUGTLE UNITS - 1 & 2 3/4 10-2 r

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i 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(1) maintaining the minimum DNBR in the core greater than or equal to 1.30 during normal operation and in short-term transients, and (2) limiting the fission gas re' ease, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200'F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat j

N flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; FfH Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power; and Fxy(Z)

Radial Peaking Factor, is defined as the ratio of peak power densit to average power density'in the horizontal plane at core elevation 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper 9

bound envelope of 2.30 times the normalized axial peaking factor 1s not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux dif ference is determined at equilibrium xenon conditions.

The rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level.

The periodic updating of the target flux difference value-is necessary to reflect core burnup considerations.

V0GTLE UNITS - 1 & 2 B 3/4 2-1

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l AXIAL FLUX DIFFERENCE (Continued)

Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux t

difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels.

This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a l

subsequent return to RATED THERMAL POWER (with the AFD within the target band) i provided the time duration of the. deviation is limited.

Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided for l

operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target ba.id are less significant.

The penalty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l

actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm.

The computer deter-mines the 1-minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at-THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computet outputs an alarm message when the penalty deviation accumulates beyond the l

limits of I hour and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR - Ffg The limits on heat flux hot channel factor and nuclear enthalpy rise hot channel factor ensure that:

(1) the desi and minimum DNBR are not exceeded and (2)gn limits en peak local power density in the event of a LOCA the peak fuel clad teinperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than i 12 steps, indicated, from the group demand position; b.

Control rod banks are sequenced with a constant tip-to-tip distance between banks as defined by Figure 3.1-3.

V0GTLE UNITS - 1 & 2 B 3/4 2-2 Amendment No. 29 (Unit 1)

Amendment No.10 (Unit 2)

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DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES I

5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly

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containing 264 fuel rods clad with Zircaloy-4.

Each fuel rod shall have a nominal active fuel length of 144 inches.

The initial core loading shall i

have a maximum enrichment not to exceed 3.2 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment not to exceed 4.55 weight percent U-235.

CONTROL ROD ASSEMBLIES i

5.3.2 The core shall contain 53 full-length control rod assemblies.

The control rod assemblies shall contain a nominal 142 inches of absorber material.

The nominal absorber composition shall be 95.5% natural halfnium and 4.5%

natural zirconium and/or 80% silver, 15% indium, and 5% cadmium.

All control rods shall be clad with stainless steel.

F 5.4 REACTOR COOLANT SYSTEM i

DESIGN PRESSURE AND TEMPERATURE l

5.4.1 The Reactor Coolant System is designed and shall be maintained:

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 f

c.

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

VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 12,240 + 100 cubic feet at a nominal T,yg of 588.5'F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1' The meteorological tower shall be located as shown on Figure 5.1-1 and 5.1-2.

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L V0GTLE UNITS - 1 & 2 5-4 Amendment No.29 (Unit 1)

Amendment No.10 (Unit 2)

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