ML20003H803

Amends 63 & 60 to Licenses DPR-39 & DPR-48,respectively, Incorporating Requirements for Implementation of TMI Lessons Learned Items Including Emergency Power Supply, Valve Position Indication & Containment Isolation

ML20003H803
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	04/20/1981
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20003H792	List: ML20003H791 ML20003H803 ML20003H807
References
NUDOCS 8105070467
Download: ML20003H803 (25)

Text

___

j*g UNITED STATES NUCLEAR REGULATORY COMMISSION e

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-295 ZION STATION UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 63 License No. DPR-99 i

1.

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

A.

The application for amendment by Comonwealth Edison Company (the licensee) dated February 17, 1981, complies with the standards and I

requirements o.' the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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; t

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 Comission's regulations; i

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 Comission's regulations and all applicable requirements have been satisfied.

l l

81050706E

2-2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-39 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63..are hereby incorporated in the license. The licensee shall operate the facility in i

accordance with the Technical Specifications.

In addi' ion, the following Licensing Conditions are hereby added as follows:

3.

t i

"2.C.(9) The Licensee shall implement a program to reduce leakage from systems outside containment that would or could contain hiahly radioactive flui.ds during a serious transient or accident to as low as practical levels. This program shall include the following:

1.

Provisions establishing preventive maintenance and periodic visual inspection requirements, and

, 2.

Integrated leak test requirements for each system at a frequency not to exceed refuel:ing cycle intervals.

I 2.C.(10) The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concen-tration in vital areas under accident conditions. This program l

shall include the following:

1.

Training of personnel.,

2.

Procedures for monitoring, and 3.

Provisions for mairtenance of sampling and analysis equipment."

4 This license amendment is effective as of the date of its. issuance.

FbRTHENUCLFARREGULATORYCOMMISSION l0

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fT n

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%r Operating React Branch No. 1 Division of Lic ng

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Attachment:

Changes to the Technical Specifications Date of Issuance:

APR 2 0 M

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'o UNITED STATES

{ g}g,//,,ogj,I NUCLEAR REGULATORY COMMISSION wAsmwoTow, o. c.20ses y w]

COMMONWEALTH EDISON COMPANY DOCKET NO. 50-304 ZION STATION UNIT 2 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 60 License No. OPR-48 1.

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

A.

The application for amendment by Comonwealth Edison Company (the licensee) dated February 17, 1981, complies with the standards and requt'rements of the Atc=ic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confomity with the application, the provisions of the Act, and the rules and regulations of the Cc raission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and seiety of the public, and (ii) that such activities will be conducted in compliance with the Ccmission'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 puSlic; and l

D.

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

of the Ccmission's regulations and all applicable requirements have I

been satisfied.

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4 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-48 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained -in Appendices A and B, as revised through Amendment No. 60, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

In addition, the following Licensing Conditions are hereby added as follows:

"2.C.(9) The Licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

1.

Provisions establishing preventive maintenance and periodic visual inspection requirements, and

. 2.

Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

2.C.(10) The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concen-tration in vital areas under accident conditions. This program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment."

1 4

This license amendment is effective as of the date of its, issuance.

FOA.THE NUCLEA REGULATORY COMMISSION h

)I k dyga, i

e Operating Reactors ranch No. 1 Division of Licensivig

Attachment:

Changes to the Technical Specifications APR 2 01981 Date of Issuance:

ATTACHMENT'TO LICENSE AMENDMENTS AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. DPR-39 AMENDMENT NO. 60 TO FACILITY OPERATING LICENSE NO. DPR-48 DOCKET NOS. 50-295 AND 50-304 Revise Appendix A as follows:

l Remove Pages Insert Pages vii vii viii viii i

ix ix x

x 75 75 75a 76 76 77 77 78 78 t

1 31 131 131b l

133 133 133a 135 135 184 184 192a 192b 195 195 l

300 300 331 331 1

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LIST OF FIGlRES Figure g

3.3.2-1 Reactor Coolant System lieatop Limitations 84 3.3.2-2 Reactor Coolant System Cooldown Limita&. ions 85 3.3.2-3 Ef fect of Fluence and Copper Content on Shift of 86 RTmy for Reactor Vessel Steels Exposed to 550 degrees F Temperature 3.3.2-4 Fluence at 1/4T and 3/4T as a function of Full Power 87 Service Years 3.4-1 liigh Steam Line Flow Setpoint 131a 4.16-1 Location of Flyed Environmental Radiological Monitoring Station 278 6.1.1 Commonwealth Edison Corporate and Station Organization 329 6.1.2 Minimum Shift Crew Composition 331

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i Amendments 63 and 60 yit

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LIST OF TABLES (cont)

J$d!.

Pagg 4.4-7 Engineered Safety Equipme_mt Actuation Test 136 j

4.5-1 Containment Fan Cooler Lomponents 148 4.6-1 Containment Spray System Components 153 4./-l Steam Generator Safety Valves, Set Pressures, Orifice Sizes 160 and Steam Flows 1

4.7-2 Auxiliary Feedwater Pumps 16L 4.8-1 Centrifugal Criarging Pump System 185 i

4.8-2 Safety Injection Pump System 186

}

4.8-3 Residual Heat Removal Pump System 187 i

4.8-4 Accumulator Tanks 188 4.8-5 Component Cooling Pump Syrtem 189 4.8-6 Service Water Pump System 190 4.8-7 Hydrogen Control System 192 4.8.9-1 Accident Monitorin0 Instrumentation Surveillance Requirements 192b j

4.9-1 Isolation Seal Water System 203 4.9-2 Penetration Pressurization System 204 4.9-3 Containment Isolation Valves 205 4.9-4 Main Steam Isolation Valves 208 Amendments 63 and 50 1x

LIST OF TABLES (cont)

Table

Pa_gg, 4.11-1 Radioactive Liquid Waste Sampling and Analysis 226 4.12-1 Pathways of Release 236

~

4.12-2 Radioactive Gaseous Waste Sampling and. Analysis 237 4.12-3 Effluent Caseous Waste Monitor 239 4.14-1 Process and Internal Monitoring 252 4.15-1 4160-Volt Engineered Safeguard Bus Main, Reserve and Standby feeds 270 4.16-1 Zion Standard Radiological Monitoring Program 276 4.17-L K PA/ Charcoal filter Systems Surveillance Requirements 284 4.17-2 Particulate filter System Surveillance Requirements 286 4.19-1 Failed fuel Monitoring Instruments 295 6.6.2 Special Reports 328a 6.3.1 Doundary Doors for Flood Conditions 332 i

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Amendments 63 and 60 l

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LIMITING CONDITION FOR OPERATION SWVEILLANCE REQUIREMENT 3.3.1 C.

Pressurizer Safety valves 4.3.1 C.

Pressurizer Safety Valves 1.

At least one code safety valve shall 1.

Safety valves R' 1010A, RCbdiOG and RC8010C shall bu 'ested for set be operable whenever the vessel is closed, except during hydrostatic pressure at each afueling tests.

outage.

Testing shall * ' done by a calibrated auxiliary 4 '.

'ng 2.

All code safety valves shall be device or by bench teiting using operable whenever the reactor compressed gas.

The valve coalant temperature is above 200'F.

setpoints shall be 2485 + 1% psig.

0.

Pressurizer 2.

Not Applicable.

The pressurizer shall be OPERABLE with at D.

Pressurizer least 150kw of pressurizer heaters and a water level not to exceed 92%.

1.

The pressurizer volune shall be determined to be witnin its limit APPLICABILITY: Modes 1, 2, 3 at least once per snift.

ACTIDH:

a.

With the standby AC on-site power 2.

The standby AC on-site power supply (diesel generators) to the supply (diesel generators) for the I

l pressurizer heaters inoperable; pressurizer heaters shall be either restore the inoperable demonstrated OPERABLE at least standby AC on-site power supply once per 18 months by transferring (diesel generators) within seven power from the nornel to the days, or be in at least IOT SiUTDOWN standby AC on-site power supply within the followl,og four hours.

(diesel generators) and energizing j

the heaters.

u.

With the pressurizer otherwise inoperable, be in at least HOT 4

SHUTDOWN with the rdactor trip breakers open within six hours and in the ICT SIUIDOWN condition with Tavg less than 350'F within the following six hours.

Amendments 63 and 60 75

i LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT s 3.3.1 E.

Loop Stop Valves 4.3.1 E.

Loop Stop Valves a

j 1.

No more than one loop shall be isolated 1.

Not hpplicable.

unless the Reactor Coolant System is connected to the Residual teat Removal System and the Residual that Removal System is operable.

l 2.

Whenever a reactor coolant loop is 2.

The boren concentration in tne l

1solated, the boron concentration in the isolated loop shall be determineu isolated loop shall be maintained at a at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

value greater than or equal to the baron concentration in the active loops.

3.

Isolated Reactor Coolant Loop Startup 3.

Isolated Reactor Coolant Loop Startup Whenever startup of an isolated reactor coolant loop is initiated the following conditions shall be met:

a.

All the channels, including a.

The interlocks associated with redundant channels, of the Loop Stop the loop stop valves shall be Valve Interlock System of the verified every refueling isolated loop are operable, except

outage, i

as specified in 3.3.1.E.3. f.

In the event this condition is not satisfied, the loop must remain isolated.

b.

Tne reactor shall be in a shutdown b.

Not Applicable.

conditions prior to opening either stop valve and througnout the timing interval required by 3.3.1.E.3.a prior to opening the cold leg stop valve.

[

Amendments 63 and 60 75a

s

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LlHITING C0!lDITION FO.1 OIli. tnt 10H SuflVEILL.'.CC7. !!L*.;UI.P.l:E:1T 3.3.1.E.3 c.

The hot Icg stop valve and the 4.3.1.E.3 c.

Prior to opening the hot leg.

cold Icg stop valve shall not,

valve and again prior to be opened unicss the baron concentration of the isolated opening the cold leg valve, the tioron concentration of

~

loop is greater than or equal the isolated loop must be to the boron concentration in verified as greater than or the unisolated loops.

equal to the boron concen-tration in the operating loops.

l d.'

If the neutron count rate f

d.

The count rate, as given by increases by more thars a ti.e nuclear instrumentation factor of two over the initial shall oe log 6ed every five.

count rate durini; the time 8';18.utes durind. the timing interval af ter the hot leg Interval required prior to valve is opened and before opening the cold. leg stop the cold leg is opene.d, the valve.

i hot leg stop valve should be l

re-closed t nd, no a t.t.cinpt.

shall be made to open the stop valves unt.11 the reason for the count rate increase i.

has been determined.

O

'e.

All reur reactor coolant e.

Not Applicable D.::nps shall be running, during the time a hot leg stop valve is being opened:, tihen a hot r

b.

leg stop valve is open and the i

cold let; s tos: valve in the

)

saue loop is clos'ed, and dur-

}

ing t'.ie tis.:e a colu leg st.op valve is being opened; except when the reactor coolant j

. tempera ture arul pressure are citual to or less than 350*F und 450 psig, res pec t.1vely.

~

l Amendments 63 and 60 7b' I

a l

LIMITING C0t01 TION FOR OPERATION SURVEILLANCE REQUIREMENT f.

The Loop Stop Valve Interlock System f.

The volume of water needed to of tne isolated loop is not required refill tne isolated loop shall be ta be operable when openlon reactor measured to ensure tnat the 25%

coolant loop islation valves additional baron concentration is provided that the reactor is in cold maintained taking into acount the shutdown condition with sufficient residual water in the drained boron concentration for 70'F loop.

The required boron operation, and the isolated loop has concentration shall be verifico been drained and refilled, borated during refilling by sampling the to a boron concentration of 25%

refilling water, prior to opening greater than that required for cold tne isolation valves t'y sampling

shutdown, the loop, and by sampling the loop every four hours prior to initial F.

Relief Valves pump operation.

The refilling water sample and tne isolateo loop Two power operated relief valves (PORV) sample shall be taken and analyzed and their associated block valves shall independently by different be OPERABLE.

personnel.

j APPLICABILITY: Modes 1, 2, 3 F.

Relief Valves

~ ACTION:

a.

With one or more PORV inoperable, 1.

Each PORY shall be demonstrateo within one hour either restore the OPERABLE at least once per 18 i

PORV to OPERABLE status or close tne months by performance of a CHANNEL j

associated block valve (s) and remove CALIDRATION.

l power from the block valve (s);

otherwise, be in at least HOT 2.

If open, block valve (s) shall De SHUTDOWN within the next six hours demonstrated OPERABLE at least and in COLD SHUTDOWN within the once per quarter by operating the follor!cy 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

open valve (s) tnrou0h one complete cycle of full travel.

b.

With one or more block valve (s) inoperable, within one hour either 3.

The standby AC on-site power restore the block valve (s) to supply (diesel generators) for tne OPERABLE status or close the block POHV and block valves shall be valve (s) and remove power from the demonstrated OPERABLE at least block valve (s); otherwise, be in at once per 18 months by transferring least IOT SHUIDOWN within the next motive and control power from the l

six hours and in COLD SHUTDOWN normal to the standby AC on-site within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

power supply and operating the s

valves througn a complete cycle of 77 full travel.

Anendments 63 and 60

Bases:

3.3.1 & 4.3.1 1

]

Operational Components A

Specifications 3.3.1. A.1 requires that a sufficient The requirement that a maximum number of nut -r of reactor ccclant pumps be operating to pressurizer heaters are OPERABLE assures that. the provide coastdown core cooling flow in the event of plant will be able to establish natural a loss of reactor coolant flow accident.

This circulation.

i provided flow will maintain the DtH1 above 1.30.

(1) Heat transfer analyses also show that reactor The reactor coolant loop isolation valve heat equivalent to approximately 10% of rated power specifications will insure that adequate cooling 4

can De removed with natural circulation; however, e is available for the reactor wnen the resioual i

reactor is not designed for critical operation with heat removal system is not in operation.

The j

natural circulation or less than three loop requirements relating to boron concentration and 4

eperation and will not be operated with these startup of an isolated loop will ensure that no i

conditions.

reactivity insertion takes place wnen tne loop l

stop valves are opened. (2) 1 One steam generator capable of performing its heat transfer function will provide sufficient heat Startup of a loop will inject cool water into tne removal capability to remove core decay heat after core wnich will result in a power spike.

4 j

a normal reactor shutdowr...The requirement for Limiting the power to 25% at tne time of startup three operable steam generators, combined witn the of a loop will minimize the effect the cool water j

requirements of Specification 3.7, ensure adequate will have on the core. (3) heat removal capabilities for reactor coolant system temperatures of greater than 350*F.

The specification for refilling a drained and isolated loop when tne cold snutdown conoition i

Operation with a primary system safety valve ensures that proper boron concentration is l

setting of 2485 psig is less than the safety limit maintained when reactor coolant loop isolation of 2735 psig.

(See Bases 1.2 and 2.2).

valve interlocks are bypassed.

Pressurizer Relief Valves (PORV) j The limit on the maximum water volume in the The power operated relief valves (PORV) and steam pressurizer assures that the parameter is bubble both function to relieve RCS pressure maintained within the normal steady-state envelope during all design transients up to and including i

of operation assumed in the FSTR(1). The 12-hour the design step load decrease with steam dump.

]

periodic surveillance is sufficient to ensure that This prevents actuation of the fixed the parameter is restored to within its limit high-pressure reactor trip. Operation of the i

following expected transient operation.

The PORV minimizes the undesiracle opening of the maximum water volume also ensures that a steam spring-loaded pressurizer code safety valves.

i bubble is formed and thus the RCS is not a Each PORV has a remotely operated block valve to j

hydraulically solid system.

provide a positive shutoff capaollity should a reller valve become inoperable or if excessive

References:

(1) FSAR Section 14.1.6, (2) FSAR leakage should occur.

Section 4.2.2.2, (3) FSAR Section 14.1.7 Amendments 63 and 60 78

1.

2.

3.

4.

5.

6.

Actuation Channel tb. of Minimum Minimum Operatcr Action 07scription (Per Unit)

No. of Channels Operable Degree of if column 3 or 4 Channels to trip Channels m Rcdondancy m cannot be met +

Setpoints.+

V.

Auxiliary Feedwater 1.

Maruai 1/ pump 1/ pump 1/ pump 0

Maintain Hot Snutdown m N.A.

2.

Automatic 2

1 2

1 Maintri.9 Ibt Shutdownm N.A.

3.

Steam Generator (S/G)

Water Level low-low i Start Motor 2 per S/G Driven Pumps 3 per S/G any 1/4 S/G 2 per S/G 1 per S/G Maintain tot Snutdown " 10%

Narrow Ran(

11 Start Turbine 2 per S/G Driven Pumps 3 per S/G any 2/4 S/G 2 per S/G 1 per S/G Maintain lot Snutdown m 104 Narrow Rang 4.

Undervoltage-RCP busses 75%

Start Turbine Driven Pump 4-1/ bus 2

3 1

Maintain fut Shutdown *** RCP Bus 4

Voltage 5.

S.I. Start. Motor and Turbine Driven Puinps 2

1 2

1 Maintain Fet Shutoown m N.A.

6.

Station Blackout 3-1/ bus 2

2 1

Maintain ibt Shutdownm Time Start Motor end Turbine Driven Pump Dependent on Voltage PERMISSIVES Setpaint +

i P-ll Pressurizer pressure (2/3) below 1915 psig allows manual block of safety injection actuation during a plant cooldown.

}l P-12 Tavg (2/4) below 540'F allows manual block of High Steam Flow safety' injection actuation if borated to greater than cold shutdown conditions.

SEE F001 TOTES ON FOLLOWING PAGE.

4 ENGINEERED SAFEGUARDS ACTUATION SYSTEM - LIMITING CONDITIONS FOR OPERATION Ato SETFO TABLE 3.4-1 (CONTINUED) 131 Amendments 63 and 60

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Chennni D92cription Dtvice D231gnntion III.

CONTAIt#4ENT ISOLATION A)

Phase A 1.

Manual Actu Lica N.A.

2.

Safety Injection Section I of this table..

B)

Phase B 1.

Manual Actuation N.A.

2.

Automatic Actuation N.A.

3.

High-eligh Containment Pressure PT-CS19, PT-CS20, PT-CS21, PT-CS22 IV.

STEAM LINE ISOLATION 1.

Manual Actuation N.A.

2.

Automatic Actuation N.A.

3.

High-High Containment Pressure PT-CS19, PT-CS20, PT-CS21, PT-CS22 4.

liigh Steam Line Flow in

a. Flow:

FT-512, FT-513, FT-522, FT-523 Coincidence with Low-Low Tavg FT-532, FT-533, FT-542, FT-543 or Low Steam Line Pressure

b. TemPoratures TE-411A, TE-411B, TE-421A, TE-421B TE-431A, TE-431B, TE-441A, TE-441B

c. Pressure:

Pr-516, PT-526, PT-536, PT-546 ENGINEERED SAFEGUARDS SYSTEM INSTRIMENT NUMBERS TABLE 3.4-2 Continued Amendments 63 and 60 133

I Channel Description Device Description l

V.

AUXILIARY FEEDWATER 1.

Manual NA 2.

Automatic NA l

3.

Steam Generator LC-5178, LC-5270, LC-5370, LC-5470 i

Water Level Low-low LC-5180, LC-5280, LC-5388, LC-5480 LC-5190, LC-5290, LC-5390, LC-5498 4.

Uadervoltage - RCP Busses Start Turbine Driven Pump 27-1, 27-2, 27-3, 27-4 5.

SI Start Motor and See Section I of this Table.

Turbine Driven Pumps 2

i 6.

Station Blackout Start Unit I 427x2-142, 427x2-143, 427x2-144 Motor and Turbine Unit II 427x2-242, 427x2-243, 427x2-244 Driven Pumps

]

PERMISSIVES P-ll Pressurizer pressure - PT-455, PT-456, PT-457 P-12 Temperature - TE-All A, TE-4118, TE-421A, TE-4218, TE-431A, TE-4318, TE-441A TE-4418 i

i 1

j I

i i

I ENGINEERED SAFECUARDS SYSTEM INSTRLDENT NUMBERS TABLE 3.4-2 (Continued)

Amenstents 63 and 60 133a 4

ACTUATION CHANtEL CHANNEL CHAtNEL CHANNEL DESCRIPTION ClIECK CALIDRATION FUNCTION TEST IV.

STEAMLINE ISOLATION 1.

Manual Actuation N.A.

N.A.

R 2.

Automatic Actuation N.A.

N.A.

M High-H1 h Containment 3.

0 Pressure See Item II Above 4.

High Steam Line Flow in Coincidence with Low-Low Tavg See Item I Above i

or Low Steam Pressure V.

AUXILIARY FEE 0 WATER 1.

Manual N.A.

N.A.

R 2.

Automatic H.A.

N.' A.

M 3.

Steam Generator S

R M

i Water Level Low-Low 4.

Undervoltage - RCP Busses N.A.

R R

\\

5.

Safety Injection i

See Item I on Page 134 j

6.

Station Blackout N.A.

R R

1 FERMISSIVES 1.

P-ll N.A.

N.A.

M 2.

P-12 N.A.

N.A.

M i

NOTE: Specified intervals may be adjusted 225% to accommodate test seneoules I

S - Once per shif t M - Once per month N.A. - Not applicable i

R - Once per refueling shutdown - calibration of these instruments may be done as such as six months prior to the star t of refueling outage and still satisfy this requirement.

i ENGINEERED SAFEGUARDS SYSTEM TESTING AND CALIBRATION REQUIREMENTS TABLE 4.4 (Continued)

,....,,-,e n,., o,i c n rw

l LIMITING C0toITION FOR OPERATION SlHVEILLANCE REQUIREMENT s

3.8.9 The accident monitoring instrumentation 3.8.9 Each accident monitoring ir'strumentation channels sh m n jn Table 3.8.9-1, hall be OPERABLE.

channel shall be demonstrated OPERA 0LE by performance of the ItaiRtiENT CHANNEL CHECK AND CHANNEL CALIBRATION at the frequencies APPLICABILITY:

Modet. 1, 2 and 3 shown in Table 4.8.9-1.

ACTIGh a.

With the numbe' of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.8.9-1, (Col. 2) either restore the inoperable channel (s) to OPERABLE status within seven days, or the

)'

unit shall be placed in at least the IOT SlHTOOWN mode, with T less than 350'F (Mode 4), withir5 the next I

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.*

l b.

With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum tomber of Channels shown in Table 3.8.9-1, (Col. 3) either restore the inoperable channel (s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or the unit shall be placed in at least the HOT SIUT00WN mode, with Tavn less than 350'F, (Mode 4) within the next 12 nours.

c.

The provisions of Specifications 3.0.4 are not applicable.

• This action does not apply to the PORV Position Indication or the PORV Block Valve Position Indication if the Block Valve on the i

associated line is known to be closed either by verification within seven days or by system status knowledge prior to indication failure.

184 Anendments 63 and 60

TABLE 3.8.9-1 ACCIDENT MONITORING INSTRLSENTATION 1.

INSTRtKNT (PARAMETER) 2.

TOTAL NO.

3.

_OF CHANNELS REQUIRED HO.

1.

Containment Pressure (Narrow Range)

_OF CHANNELS MINIMLN NO.

_O_F CHANNELS 2.

4 Reactor Coolant Outlet Temperature - T 2

1 ibt (Wide Ra)ge) 3.

Reactor Coolant Inlet Temperature - T 4-1/ Loop 2

1 Cold (Wide Range) 4.

Reactor Coolant Pressure (Wide Range) 4-1/ Loop 2

1 i

5.

Steam Line Pressure 2

2 1

6.

Pressurizer Water Level 12-3/SG 2/SG 1/SG 3

7.

Stewn Generator Water Level (Narrow Range) 2 1

8.

12-3/SG Steam Generator Water Level (Wide Range) 2/SG 1/SG 4

9 Refuelleg Water Storage Tank Level 4-1/SG 2

1 10.

2 Auxiliary Feedwater Flow 2

i 1

11.

4-1/SG Reactor Coolant System Subcooling Margin

• 2

\\

1 12.

2*

PORV Position Indication **

2*

l' 13.

2/ valve PORV Plock Valve Position Indication 2/ valve 1/ valve 14.

1/ valve Safety Valve Position Indicqtion 1/ valve 0

3-1/ valve

{

2 procedure performed by the opetator.The Reactor Coolant Subcooling Margin is dete 1

y two methods; l) computer analysis or, 2) a

==

The PORY Position Indication consists of 1) stem acoustical monitoring system (backup indication) -mounted limit switch (primary indica i

1 Amendments 63 and 60 192a i

4

N O

LI ET NA NR R

R R

R R

R R

R R

R R R R

R AB HI CLAC S

TN E

M ER IUQ E

R E

L C

EK_

N NC A

NE M

M M

M M

M H

M M

M M

M M

M L

At I

C L

D I

EV R

t 1

S

)

)

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g

8.,

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(

T

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)

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W L

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P u

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(

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C n

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G G i t

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R R

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P S

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1 2

3 4

5 6

7 8

9 1

1 1

1 1

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i i

A hydrogen reconbiner system is installed to remove the hydrogen and oxygen gases that accumulate in the If during this seven day period a third service water containment atmosphere following a loss-of-coolant pump becomes inoperable, the unit will be brought to HOT SHUTDOWN within four hours unless one of the accident. (9)

The system operability requirement inoperable pumps are returned to operable status.

becomes effective following successful preoperational The requirement foi independent stancby AC and DC

testing,

]

power supplies for an operating service water pump ironi the unit precludes the possibility of loeing two antil completion of circulatory water pump maintenance service water pump dedica.t,ed to the same unit because program, scheduled for completion about December 31, of the failure of a common power supply.

1981, one service water pump at a time may be taken out of service to perform maintenance on the associated The OPERABILITY of the accident monitoring i

circulating water pump.

In order to work on a instrumentation ensures that sufficient information circulating water pump, an intake plenum must be is available on selected plant parameters to monitor l drained. The intake plenum is a common reservoir for a and assess these variables during and following an

} service water pump and a circulating water pump.

This accident.

This capability is consistent with the leaves five service water pumps in service, which will

' orovide sufficient coverage for any postulated loss of recommendations of Regulatory Guide 1.97, caolant accident coincident with a loss of off-site

" Instrumentation for Light-Water-Cooled Nuclear Power

power and any other single failure of an active Plants to Assesf Plant Conditions During and component. With this temporary specifi' cation, should a Following an Accident," December 1975 and NUREG-0578, second service water pomp become inoperable, it must be "TMI-2 Lessons Learned Task Force Status Report and Short-Terro Recommendations".

returned to operable status within seven days or.the unit wl)I be brought to. HOT SHUTDOWN.

(1)

FSAR Chapter 9 (2)

FSAR Section 6.2 (3)

FSAR Section 6.2.3 (4)

FSAR Section 14.3 (5)

FSAR Section 9.3 (6)

FSAR Section 9.6 & FSAR answer to question 9.1 (7)

FSAR Section 14.3.6 (8)

FSAR Answer to Question 9.9 (9)

FSAR Section 6.8 a.

j 195 Amendments 63 and 60 i

t l

6.0 ADMINISTRATIVE CONTROLS l6.1 Organization, Review, Investigation, and Audit 3

j A.

The Station Superintendent shall have D.

Qualificattuns of the Station management 4

overall full-Line responsibility for and operating staff shall meet minimum 4'

safe operations of the facility.

acceptable levels as described in ANSI During periods when the Station N18.1 Selection and Training of Nuclear Superintendent is unavailable, he shall Power Plant Personnel, dated March 8, designate this responsibility to an 1971 with the exception of the Rad-Chen l

established alternate who satisfies' Supervisor or Lead Health Physicist, who the ANSI N18.1 experience requirenents l

shall meet or exceed the qualifications of for plant manager.

I Radiation Protection Manager of Regulatory 1

i Guide 1.8 September 1975 The Shift Contrcl B.

The organization chart of the corporate Room Engineer

• shall hue a bachelors degree or equivalent in a scientific or engineering l

management which relates to the operation; of this station and the normal functional discipline with specific training in plant i

organization chart for operation of the design, and response and analysis of the station is shown in Figures 6.1.1.

plant for transients and accidents. The individual filling the position of Admin-1strative and Support Services Assistant j

Superintendent shall meet the minimum 1

i acceptable level for " Technical Manager" as i

C.

The shift manning for the station shall be described in 4.2.4 of ANSI N18.1,1971.

l as shown in Figure 6.1.2.

The Operating Assistant Superintendent, Operating Engineer, E.

Retraining and replacenent training of Shift Engineers, and Shift Foreman shall Station personnel shall be accordance have a senior operating license. The Fuel gg

,,3 g

llandling foreman has a limited Senior Operatjng of Nuclear Power Plant Personnel," dated License. The Division Vice President, Nuclear g

Stations on the corporate lev,el has responsibility Fire Brigade shall be maintained under the for the Fire Pmtection Pmgram. An Operating Engineer direction of the Station Fim Marshall and at the station will be responsible for implenentation shall neet or exceed the requirements of of the Fire Protection Program.

'A Fire Brigrade Section 27 of the NFPA Code - 1975 except of at least 5 cmebers shall beimaintained onsite at that Fire Brigade training will be conduc-i all times. Tlic Fire Brigade shall not include ted quarterly.

I the minimum shift crew necessary for safe shutdown of the plant (4 members) or any personnel required F.

I sid k M edd h w 1

for other essential functions during a fire emergency.

not exceeding two years.

l

~

G.

The Review and Investigative Function and i

i the Audit Fune.ti6n of activities affecting

• The Shift Control Room Engineer training quality during facility operations shall requirement will be errective after be constituted and have the responsibilities

. in n n 1, 1931 ann anil ant hm i t ieu: n o t i i nn.1 i...i no -

thit in Modes 1, 2, 3, 4 thit I thit 1 Position No or and thit thit 2

_thit 2 Shift Engineer or Shift Foreman 1

1 2

l Shift Control Room Engineer None 1

1 Required Nuclear Station Operator 1

2 3

Equipment Operator or Equipment Attendant 2

3 4

R diation Protection Person 1

1 1

TOTAL 6

8 11 MINIMLM' 6

7 10

• The rainimum number refers only for the case of shift shortage caused by a sudden sickness or home emergency.

tut.3s :

1.

Senior reactor Operator (SRO) shall be present :

when there is fuel in the reactor.

11-site at all times 2.

A licensed person shall be in the control room at all time whenever fuel is in either reactor.

s 3.

Two licensed people shall be in the control room during react startups, shutdowns, operation, and other periods such as planned or control rod manipulations.

l MINIMLN SHIFT CREW COMPOSITION Figure 6 4 2 Amendments 63 and 60 331

,