ML20155E026

Amend 11 to License NPF-68,modifying Tech Specs to Allow Slightly Positive Moderator Temp Coefficient & Revised Shutdown Margin Requirements

ML20155E026
Person / Time
Site:	Vogtle
Issue date:	10/04/1988
From:	Matthews D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20155E029	List: ML20155E026 ML20155E627
References
NPF-68-A-011 NUDOCS 8810120086
Download: ML20155E026 (23)

Text

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L NUCLEAR REGULATORY COMMISSION e

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w ASHING TON. D. C. 20666 i

s+....j GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION I

4 NUNICIPAL ELECTIC AUTHORITY OF GEORGIA 1

CITY OF DALTON, GEORGIA l

V0GTLE ELECTRIC GENERATING PLANT, UNIT 1 I

j AMENDMENT TO FACILITY OPERATING LICENSE l

b I

Amendment No. 11 License No. NPF-68 i

1.

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

l A.

The opplication for amendment to the Yogtie Electric Generating Plant, Unit 1 (the facility) Facility Operating License No. NPF-68 filed by j

the Georgia Power Company acting for itself, Oglethorpe Power Corpo-t l

ration, Municipal Electric Authority of Georgia, and City of Dalton.

l Georgia. (the licensees) dated May 19, 1988, as supplernented 6

i August 12 and October 3, 1988, complies with the standards and requirements of the Atom k Energy Act of 1954, as amended (the Act),

and the Comission's rules and regulations set forth in 10 CFR l

l Chapter I; j

B.

The facility will operate in conformity with the application, as I

amended, the provisions of t!)e Act, and the rules and regulations l

of the Comission; C.

There is reasonable assurance (1) that the activities authorized by l

l this amendinent can be conducted without endangering the health and j

safety of the public, and (11) that such activities will be conducted j

in compliance with the Corsaission's regulations se'. forth in 10 CFR 1

Cnapter I;

[

D.

The issuance of this license amendment will not be inimical to the l

comon defense ard security or to the health and safety ot' the public; and 1

l E.

The issuance of this amendment is in accordance with 10 CFR Part 51 l

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of the Cocenission's regulations and all applicable requirements have l

been satisfied.

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Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-68 is hereby amended to read as follows:

(2) Technical Specifications and Environmntal Protection Plan The Technical Specifications contained in Appendix A, as revised through Amerdment No.11 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into ti.is license.

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

3.

This license amendment is effective as of its date of issuance end shall be implemented within 60 cays of issuance, i

FOR THE NUCLEAR REGULATORY COMMISSION oricInal stenea By:

David B. Platthews, Director Project Directorate !!-3 Division of Reactor Projects Attachent:

Technical Specification Changes Date of Issuance:

October 4. 1988 fSQ] 45 s

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ATTACHMENT TO LICENSE AffENDNENT NO.

11 FACILITY OPERATING LICENSE NO. NPF-68 DOCKET NO. 50-424 Replace the following pages of the Appendix "A" Technical Specifications with the enclusco pages.

The revised pages are identified by An,endment number and contain vertical lines indicating the areas of change.

The corresponding i

overleaf pages are also provided to maintain cocument completeness.

I Amended Page Overleaf Page 3/4 1-3.

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0 400 800 1200 1600 2000 2400 E,

2 RCS BORON CONCENTH ATION (ppm)

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REACTIVITY CONTROL SYSTEMS i

MODERATOR TEMPERATURE COEFFICIENT l

)

I l

i l

LIMITING CONDITION FOR OPERATION l

t i

3.1.1.3 The moderator temperature coefficient (M1C) shall be:

l' f

{

a.

Less positive than + 0.7 x 10 4 ok/k/'F for the all rods withdrawn, beginningofcorelife(BOL)inearrampto0ak/k/' Fat 100% RATED condition for power levels up to 70%

RATED THERMAL POWER with a l THERMAL POWER; and j

b.

Less negative than - 4.0 x 10 4 ak/k/'F for the all rods withdrawn,

[

end of cycle life (EOL), RATED THERMAL POWER condition.

l APPLICABILITY:

Specification 3.1.1.3a. - MODES 1 and 2* only**.

Specification 3.1.1.3b. - MOCES 1, 2, and 3 only**.

]

ACTION:

a.

With the MTC more positive than the limit of Specification 3.1.1.3a.

3 above, operation in MODES 1 and 2 may proceed provided:

j i

1.

Control rod withdrawal limits are established and maintained I

sufficient to ristore the MTC to within n: above limits j

within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANOBY with.n the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

These withdrawal limits shall be in additiut, to the insertion j

l limits of Specification 3.1.3.6; I

2.

The control rods are maintained within the withdrawal limits

'l established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and I

3.

A Special Report is prepared and submitted to the Commission, I-pursuant to Specification 6.8.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal j

l limits, and the predicted avera e core burnup necessary for L

restoring the positive MTC to w thin its limit for the all rods j

withdrawn condition, b.

With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUIDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1 I

I

• With K,gf greater than or equal to 1.

)

• • See Special Test Exceptions Specification 3.10.3.

}

V0GTLE - UNIT 1 3/4 1-4 Amendment No. 11 t

i

4 REACTIVfTY CONTROL SYSTEMS BORATED WATER SOURCE - SHUT 00WN i

LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

A Boric Acid Storage Tank with:

1)

A minimum contained borated water volume of 9504 gallons (19%

of inst.ument span) (LI-102A, LI-104A),

1 l

2)

A boron concentration batween 7000 ppm and 7700 ppm, and 3)

A minimum solution tempereture of 65'F (TI-0103).

i l

b.

The refueling water storage tank (RWST) with:

f 1)

A minimum contained borated water volume of 99404 gallons (9%

l of instrument span) (LI-0990A&B, LI-0991A&B, LI-0992A, l

LI-0993A),

2)

A boron concentration between 2400 ppm and 2600 ppm, and l

3)

A minimum solution temperature of 54*F (TI-10982).

1 APPLICABILITY:

MODES 5 and 6.

ACTION:

l With no borated water source OPERABLE, suspend all operations involving CORE j

ALTERATIONS or positive reactivity coanges.

i SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

I l

1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water volume, and l

?)

When the boric :cid storage tank is the source of borated water I

and the ambient temperature of the boric acid storage tank roem (TISL-20902, TISL-20903) is (72*F, verify the boric acid storage tank solution temperature is } 65'F.

i b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature (TI-10982) when it is the source of borated water and the outside air temperature is less than 30*F.

i V0GTLE - UNIT 1 3/4 1-11 A.tendment No. 11 i

i

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following bor r water source (s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

A Boric Acid Storage Tank with:

1)

A minimum contained borated water volume of 36674 gallons (81%

of instrument span) (LI-102A, LI-104A),

2)

A boron concentration between 7000 ppm and 7700 ppm, and 3)

A minimum solution temperature of 65 F (TI-0103).

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 631478 gallons (86%

of instrument span) (LI-0990A&B, LI-0991A&B, LI-0992A, LI-0993A),

2)

A boron concentration between 2400* ppm and 2600 ppm, l

3)

A minimum solution temperature of 54 F, and 4)

A maximum solution temperature of L16*F (TI-10982).

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

With the Boric Acid Storage Tank inoperable and being used as one of the above required borated water sources, restore the tank to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTOOWN MARGIN as required by Figure 3.1-2 at 200 F; restore the Boric Acid Storage Tank to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

• Until concentration is initially raised to 2400 ppn fron the naxinun lintt authorized nrior to Anendnent No.11, the ninieue boron concentratico limit

i. a 2000 onn.

P V0GTLE - UNIT 1 3/4 1-12 Amendment No. 11

TABLE 4.3-1 (Continued)

TABLE NOTATIONS a When the Reactor Trip System breakers are closed and the Control Rod Drive System is capable of rod withdrawal.

b Above P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint.

c Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

d Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

e A.bove P-7 (Low Power Reacto: Trip Block) Setpoint.

(1) If not perfo.med in previous 31 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.

The provisions of Specification 4.0.4 are not applicable to entry into MODE 2 or 1.

(3) Single point comparison of incore to excore AXIAL FLUX OIFFERENCE 1

l above 15% of RATED THERMAL POWER.

Recalibrate if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

For the purpose of this surveillance requirement, monthly shall mean at least once per 31 EFPD.

(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector plateau curves shall be obtained, and evaluated.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(6)

Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

This is the determination of the response of the excore power range detectors to the incore measured axial power distribution to generate setpoints for the CHANNEL CALIBRATION.

The provisions of Specification 4.0.4 are not appli-cable for entry into MODE 2 or 1.

For the purpose of this surveillance requirement, quartarly shall mean at least once per 92 EFPD.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

(8) Not used a

a a

(9) Quarterly surveillance in MODES 3, 4, and S shall al

. a verifi-cation that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive window.

Quarterly surveillance shall include verification of the Snurce Range High i

Flux at Shutdown Alarm Setpoint of less than or equal to 2.30 times l

background, r

V0GTLE - UNIT 1 3/4 3-13 Amendment No.11

4 TABLE 4.3-1 (Continued)

TABLE NOTATIONS (Continued)

(10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the OPERABILITY of the Undervoltage and Shunt trip of the Reactor Trip Breaker.

(12) CHANNEL CALIBRATION shall include the RTO bypass loops flow rate.

(13) Not used (14) The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERABILITY of the undervoltage and shunt trip ci.cuits for the Manual Reactor Trip Function.

The test shall also verify the OPERABILITY of the Bypass Breaker trip circuit (s).

(15) Local manual shunt trip prior to placing breaker in service.

(16) Automatic undervoltage trip.

(17) Each channel shall be tested at least every 92 days on a STAGGERED TEST BASIS.

(18) The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 are more restrictive and, therefore, applicable.

l V0GTLE - UNIT 1 3/4 3-14

7

• .q' t

3/4.5 EMERGENCY CORE CC3 LING SYSTEMS 3/4.5.1 ACCUMULATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

a.

The isolation valve open, b.

A r,ontained borated water volume of between 6616 (36% of instrument span) and 6854 (64% of instrument span) gallons (LI-0950, LI-0951, LI-0952, LI-0953, LI-0954, LI-0935, LI-0956, LI-0957),

c.

A boron concentration of between 1900 and 2600 ppm, and d.

A nitrogen cover pressure of between 617 and 678 psig.

(PI-0960A&B, PI-0961A&B, PI-0962A&B, PI-0963A&B, PI-0964A&B, PI-0965A&B, PI-0966A&B, PI-0967A&B)

APPLICABILITY:

MODES 1, 2, and 3*.

ACTION:

a.

With one accumulator inoperable, except as a result of a close1 isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or 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 /> and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQtIREMENTS 4.5.1.1 Each accumulator shall be demonstrated OPERABLE:

i a.

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

1)

Verifying the contained borated water volume and nitrogen l

cover pressure in the tanks, and 2)

Verifying that each accumulator isolation valve is open.

(HV-8808A,B,C,0)

• Pressurizer pressure above 1000 psig.

l V0GTLE - UNIT 1 3/4 5-1 Amendment No. 11

~

o EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> af ter each solution volume increase of greater than or equal to 1% of tank volume (67 gallons) by verifying the boron concentaation of the accumulator solution; and At least once per 31 days when the RCS pressure is above 1000 psig c.

by verifying that the circuit breaker supplying power to the isolation valve operator is open.

4.5.1.2 Each accumulator water level and pressure channel shall be demon-strated OPERABLE at least once per 18 months by the performance of a CHANNEL t

CALIBRATION.

8 b

V0GTLE - UNIT 1 3/4 5-2

EMERGENCY CORE COOLING SYSTEMS LESS THAN 350*F 3/4.5.3 ECCS SUBSYSTEMS - T,yg SAFETY INJECTION PUMPS LIMITING CONDITION FOR OPERATION 3.5.3.2 All Safety Injection pumps shall be inoperable.

APPLICABILITY MODES 4, 5, and 6 with the reactor vessel head on.

ACTION:

With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVE!LLANCE REQUIREMENTS 4.5.3.2 All Safety Injection pumps shall be demonstrated inoperable" by verifying that the motor circuit breakers are secured '.i the open position within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entering MODE 4 from MODE 3 prire to the temperature of one or more of the RCS cold legs decreasing below 3'.5'F, and at least once per 31 days thereafter.

• An inoperable pump may be energized for testing or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the volve operator, or by a manual isolation valve secured in the closed position, f

V0GTLE - UNIT 1 3/4 5-9

o BORON INJECTION SYSTEM 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contair,ed borated water volume of 631,478 gallons (86% of instrument span) (LI-0990A&B, LI-0991A&B, LI-0992A, LI-0993A).

b.

A boron concentration of between 2400* ppm and 2600 ppm of boron, l

c.

A minimum so:ation temperature of 54 F, and a.

A maximum solution temperature of 116 F (TI-10982).

RWST Sludge Mixing Pump Isolation valves capable of closing on RWST e.

low-level.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

1 l

\\

l a.

With the RWST inoperable except for the Sludge Mixing Pump Isolation l

Valves, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at I

least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUT 00WN within the following 30 ba rs.

j b.

With a Sludge Mixing Pump Isolation Valve (s) inoperable, restore the valve (s) to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or isolate the slud y mixing system by either closing the manual isolation valves or deenergizing the OPERABLE solenoid pilot valve within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and maintain closed..

I SURVEILLANCE REQUIREMENTS 1

4.5.4 The RWST shall be demonstrated OPERABLE:

l a.

At least once per 7 days by:

{

1)

Verifying the contained borated water volume in the tank, and 2)

Verifying the boron concentration of the water, b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when the outside air temperature is less than 50*F.

c.

At least once per 18 months by verifying that the sludge mixing pump isolation valves automatically close upon an RWST low-level test signal.

• L'ntil concentration is initia11" rained to 2'4M nm f rom the naxi.un linit authorized nrior to Arend ent "o.

li, the nininun noron concentratior. li-it is l'"'o,~.

V0GTLE - UNIT 1 3/4 5-10 Amendment No. 11 l

REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued)

MARGIN from expected operating conditions as defined by Specification 3/4.1.1.1 (MODES 1 and 2) and Specification 3/4.1.1.2 (MODES 3 and 4) after xenon decay and cooldown to 200 F.

The maximum expected boration capability requirement i

occurs at EOL from full power equilibrium xenon conditions and requires 31740 gallons usable volume of 7000 ppm borated water from the boric acid 1

storage tanks or 178182 gallons usable volume of 2400 ppm borated water from l

the refueling water storage tank (RWST).

With the RCS temperature below 200*F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The boration capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN as defined by Specification 3/4.1.1.2 (MODE 5) af ter xenon decay and cooldown from 200 F to 140*F.

This condition requires either 4570 gallons usable volume of 7000 ppm borated water from the boric acid storage tanks or 41202 gallons usable volume of 2400 ppm borated water from the l

RWST.

The contained water volume limits provided in Specifications 3/4.1.2.5 and 3/4.1.2.6 include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.0 and 10.5 for the solution recirculated l

within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the offect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures i

that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES ine specifications of this section are necessary to ensure that the follow-ing requirements are met at all times during.1ormal operation.

By observing that the RCCAs are positioned above their respective insertion limits during normal operation, 1.

At any time in life for M3DE 1 and 2 operation, the minimum SHUTDOWN MARGIN will be maintained.

For operational MODES 3, 4, 5, and 6, the reactivity condition consistent with other specifications will be main-tained with all RCCAs fully inserted by observing that the boron concentra-tion is always greater than an appropriate minimum value.

2, During normal operation the enthalpy rise hot channel factor, F3g, w W j

be maintained within acceptable limits.

V0GTLE - UNIT 1 B 3/4 ?-3 Amendment No.

11

REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued) 3.

The consequences of an ejected RCCA accident will be restricted below the limiting consequences referred to in the ejected rod analysis.

4.

The core can De made subcritical by the required SHUT 00WN MARGIN with one RCCA stuck.

In che event of an RCCA ejection, the core can be made suberitical with two RCCAs stuck, where one of the RCCAs is assumed to be the worst ejected rod control assembly.

5.

The trip reactivity assumed in the accident analysis will be available.

6.

Dropping an RCCA into the core or statically misaligning an RCCA during normal operation will not violate the thermal design basis with respect to ONBR.

7.

The uncontrolled withdrawal of an RCCA will result in consequences no more severe than presented in the accident analysis.

8.

The uncontrolled withdrawal of a control assembly bank will not result in' a peak power density that exceeds the center line melting criterion.

OPERABILITY of the control rod position indicator channels (LCO 3.1.3.2) is required to determine control rod positions and thereby ensure compliance with the control rod alignment.

OPERABILITY of the Demand Position Indication System (LCO 3.1.3.2) is required to determine bank demand positions and thereby ensure compliance with the insertion limits.

The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that some of the original criteria are met.

Misalignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER, either Jf these restric-tions provide assurance of fuel rod integrity during continued operation provided no further abnormal condition develops.

For Specification 3.1.3.1 ACTIONS b and c it is incumbent upon the plant to verify the trippability of the inoperable control rod (s).

This may be by veri-fication of a control system failure, usually electrical in nature, or that the failure is associated with the control rod stepping mechanism.

In the event the plant is unable to verify the rod (s) trippability, it must be assumed to be un-trippable and thus fall under the requirements of ACTION a.

Assuming a controlled shutdown from 100% RATED THERMAL POWER, this allows approximately four hours for thit verification.

The maximum rod drop time permitted by (LCO 3.1.3.4) is consistent with the assumed rod drop time used in the accident analyses.

Measurement with T,yg V0GTLE - UNIT 1 8 3/4 1-4

INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

(7) steam line isolation, (8) turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment fan coolers start and automatic valves position, (11) Nuclear Service Cooling and Component Cooling water pumps start and automatic valves position, and (12) Control Room Ventilation Emergency Actuation Systems start.

The Engineered Safety Features Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater valves on T,yg below Setpoint, prevents the opening of the main feedwater valves which were closed by a Safety Injection or High Steam Generator Water Level signal, allows Safety Iajection block so that components can be reset or tripped.

Reactor not tripped prevents manual block of Safety Injection.

P-11 With pressurizer pressure below the P-11 setpoint, allows manual block of safety injection actuation on low pressurizer pressure signal.

Allows manual block of safety injection actuation and steam line isolation on low compensated steam line pressure signal and i

allows steam line isolation on high steam line negative pressure rate. With pressurizer pressure above the P-11 setpoint, defeats manual block of safety injection actuation on low pressurizer pres-l sure and safety injection and steam line isolation on low steam line pressure and defeats steam line isolation on high steam line negative pressure rate.

P-14 On increasing steam generator water level, P-14 automatically trips all feedwater isolation valves, initiates a turbine trip, and inhibits feedwater control valve modulation.

The Source Range Higt. Flux at Shutdown Alarm Setpoint is an analysis assumption for mitigation of a Boron Oilution Event in MODES 3, 4, and 5.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that:

(1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance.

The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions.

Once the required logic combination is i

completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems.

j j

V0GTLE - UNIT 1 B 3/4 3-3 Amendment No. 11 1

i

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INSTRUMENTATION BASES 3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

For the purpose of measuring F (Zl or F a full incore flux map is used.

9 H

Quarter-core flux maps, as defined in UCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may he used for monitaring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capa-bility is required to permit comparison of the measured response to that used in the design basis for the facility to determine if plant shutdown is required.

pursuant to Appendix A of 10 CFR Part 100.

The instrumentation is consistent with the recommendations of Regulatory Guide 1.12 "Instrumentation for Earth-quakes," April 1974.

3/4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

3/4.3.3.5 REMOTE SHUT 00WN SYSTEM The OPERABILITY of the Remote Shutdown System ensures that sufficient capability is available to permit safe shutdown of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.

The OPERABILITY of the Remote Shutdown System ensures that a fire will not preclude achieving safe shutdown.

The Remote Shutdown System instrumentation, i

V0GTLE - UNIT 1 8 3/4 3-4

I 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators.

This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the. assumptions used for accumulator injection in the cafety analysis are met.

The minimum boron concentration must ensure that the reactor core will remain subcritical during the accumulator injection period of a small break LOCA.

l The accumulator power operated isolation valves are cor.sidered to be "operating bypasses" in the context of IEEE Std. 279-1971, which requires that 4

bypasses of a protective function be removed automatically whenever permissive e

conditions are not met.

In addition, as these accumulater isolation valves fail to meet single failure criteria, removal of power to the valves is required.

l The limits for operation with an accumulator inoperable for any reason

[

4 except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.

If a closed isolation valve cannot be immediately opened, the full capability of one l

accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure considr ation, Either subsystem operating in conjunction with the accumulators is capable of c

supplying sufficient core cooling to limit the peak cladding temperatures i

within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward.

In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period.

With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is i

acceptable without single failure consideration nn the basis of the stable l

j reactivity condition of the reactor and the i h lted core cooling requirements.

5

-l V0GTLE - UNIT 1 B 3/4 5-1 Amendment No. 11

I EMERGENCY CORE COOLING SYSTEMS BASES 4

1 ECCS SUBSYSTEMS (Continued)

The limitation for all safet/ injection pumps to be inoperable below 350'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.

Surveillance Requirements for throttle valve position stops and flow balance testing provide q

assurance that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of preper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points j

in accordance with the assumptions used in the ECCS-LOCA analyses, (3) provide

]

an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses and (4) to ensure that centrifugal charging pump injection flow wmch is directed through the seal injection path is less than or equal to the amount assumed in the safety analysis.

3/4.5.4 REFUELING WATER STORAGE TANK The OPERABILITY of the Refueling Water Storage Tank (RWST) as part of the ECCS ensures that sufficient negative reactivity is injected'into the core to counteract any positive increase in reactivity caused by RCS cooldown.

RCS cooldcwn can be caused by inadvertent depressurization, a loss-of-coolant accident, or a steam line rupture.

The limits on RWST minimum volume and boron concentration ensure that

1) suf ficient water is available within containment to permit recirculation cooling flow to the core, 2) the reactor will remain subcritical in the cold condition following a small LOCA or steamline break, assuming complete mixing of the RWST, RCS, and ECCS water volumes with all control rods inserted except t

the most reactive control assembly (ARI-1), and 3) the reactor will remain suberitical in the cold condition fcilowing a large break LOCA (break flow

> 3.0 FT2) assuming complete mixing of the RWST, RCS, ECCS water and other i

sources of water that may eventually reside in the sump, post-LOCA with all control rods assumed to be out.

The contained water volume limit includes an allowance for water tot usable i

l because of tank discharge line location or other physical characteristics.

l The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.0 and 10.5 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and tuinimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

1 l

i V0GTLE - UNIT 1 B 3/4 5-2 Amendment No. 11 i

l l

CONTAINMENT SYSTEMS BASES CONTAINMENT VENTILATION SYSTEM (Continued)

The use of the containment purge lines is restricted to the 14-inch purge supply and exhaust isolation valves since, unlike the 24-inch valves, the 14-inch valves are capable of closing during a LOCA or steam line break accident.

There-fore, the SITE BOUNDARY dose guideline of 10 CFR Part 100 would not be exceeded in the event of an accident during containment PURGING operation.

Only safety-related reasons; e.g., containment pressure control or the reduction of air-borne radioactivity to facilitate personnel access for serveillance and main-tenance activities, should be used to justify *,he opening of these isolation valves.

Leakage integrity tests with a maximum allowable leakage rate for containmer..

purge supply and exhaust supply valves will provide early indication of resilient material seal degradation and will allow opportunity for repair be' ore gross leak-age failures could develop.

The 0.60 L leakage limit of Specification 3.6.1.2b.

~

shall not be exceeded when tha leakage fates determined by the lo kage integrity tests of these valves are added to the previously determined total for all valves and penetrations subject to Type B and C tests.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the Containment Spray System ensures that containment depressurization and cooling capability will be available in the event of a LOCA or steam line break.

The pressure reduction and resultant lower containment leakage rate are consistent with the assumptions used in the safety analyses.

The Containment Spray System and the Containment Cooling System both provide post-accident cooling of the containment atmosphere.

However, the Containment Spray System also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to OPERABLE status have been maintained cansistent sith that assigned other inoperable ESF equipment.

3/4.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the Spray Additive System ensures that sufficient daOH is added to the containment spray in the event of a LOCA.

The limits on Na0H i

volume and concentration ensure a pH value of between 8.0 and 10.5 for the l

l solution recirculated within containment after a LOCt,.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The solution volume limits (3700-4000 gallons) represent the required solution to be delivered (i.e., the delivqred solution volume is that volume above the tank discharge).

These assumptions are consistent with the iodine removal efficiency assumed in the j

safety analyses.

V0GTLE - UNIT 1 B 3/4 6-3 Amendment No.

11

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3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:

(1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water l

volume having direct access to the reactor vessel.

The locking closed of the l

required valves during refueling operations precludes the possibility of I

uncontrolled boron dilution of the filled portions of the Reactor Coolant System.

This action prevents flow to th. RCS of unborated water by closing flowpaths from sources of unborated water.

These limitations are consistent with the initial conditions assumed for the Boron Dilution Accident in the safety analysis.

The boron concentration value of 2000 ppm or greater ensures aK of 0.95 or less and includes 1 conservative allowance for calculational uncN(aintyof100ppmboron.

3/4.9.2 INf.TRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 OECAY TIME

{

The minimum requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.

This decay time is consistent with the assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT BUILDING DENETRATIONS The requirements on containment building penetration closure and OPERABILITY l

ensure that a release of radioactive material within containment will be restricted from leakage to the environment.

The OPERABILITY and closure l

restrictions are sufficient to restrict radioactive material release frem a l

fuel element rupture based upon the lack of containment pressurization I

potential while in the REFUELING MODE.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station persennel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

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V0GTLE - UNIT 1 B 3/4 9-1 Amendment No. 11

REFUELING OPERATIONS BASES l

3/4.9.6 REFUELING MACHINE The OPERABILITY requirements of the refueling machine and auxiliary hoist ensure that:

(1) The refueling machine will be used for the movement of fuel assemblies l

and/or rod control cluster assemblies (RCCA) or thimble plug assemblies, and the auxiliary hoist will be used for the movement of contral rod drive shaf ts, (2) the refueling machine will have sufficient load capacity to lift a fuel ast,embly and/or a rod control cluster assembly or thimble plug assembly, l

and the auxiliary hoist will have sufficient load capacity to lift a control l

rod drive shaft and attached RCCA, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL, SPENT FUEL STORAGE AREAS The restrict Nn on movement of loads in excess of the nominal weight of a I

fuel and control rod assembly and associated handling tool over other fuel i

assemblies in the storage pool ensures that in the event this load is dropped:

(1) the activity release will be lim lted to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will nct result in a critical array.

This assumption is consistent with the activity release assumed in the safety analyses.

3/4.9.8 RESI0,UAL HEAT REMOVAL AND COOLANT CIRCULATION i

The requirement that at least one residual heat removal (RHR) train be in operation ensures that:

(1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.

The requirement to have two RHR trains OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that e sjngle iailure of the operating RHR train will not result in a complete loss of residual heat removal capability.

With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large haat sink is avail-abla for core cooling.

Thus,11 the event of a failure of the operating RHR train, adequate time is provided to initiate emergency procedures to cool the core.

V0GTLE - UNIT 1 B 3/4 9-2