ML20043B504

Proposed Tech Specs Change Re Incorporation of post-accident Monitoring Instrumentation Into Specs

ML20043B504
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	05/23/1990
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20043B500	List: ML20043B499 ML20043B504
References
NUDOCS 9005300115
Download: ML20043B504 (53)

Text

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j ATTACHMENT 5 l

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j

1. Copy of the current Technical Specifications, marked-up j

f to denote the chan9es being proposed.

Pages: 111 y

x 184 V:

192a l

192b i

L 192c-195 L

236a 236b 237 251

(.

2524 309 i

310

[

324 i

2. Copy of the typed change being proposed.

l-Pages: 111 l

y x

184

.l 192a 192b 192c t

192d

-i 192e.

'195-2364

{

236b 237 251 252a 1'

309:

310 t

324 i

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'$$f3$$b$ $O0 f I

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1 Table of Contents (Continued)

SURVELLLANCE LIMITING CONSITION F0d OPERATION REQUIRUENT PAGE 3.7 Steam Generator Emergency Heat Removal 4.7 156 3.7.1 Steam Generator Safety Valves 4.7.1 156 3.7.2 Auxiliary Feedwater Puny System 4.7.2 158 3.7.3 Auxiliary Feedwater Supply System 4.7.3 159a i

Bases

%? "

3.8 Emergency Core Cooling and Core Cooltag Support 4.8 164 3.8.1 Centrifugal Charging Punp System 4.8.1 164

~3.8.2 Safety Injection Pump System 4.8.2 164 j

3.8.3 Residual Heat Removal Puny System 4.8.3 170

)

3.8.4 System Testing of Centrifugal Charging, Safety 4.5.4 173 t

i Injection, and Residual Neat Removal Punp Systems 3.8.5 Accumulator System 4.8.5 174 3.8.6 Component Ceoling System 4.8.6 175 3.8.7 Service Water System

/

4.8.7 178 Mydrogen Control Systems Audo b" 4.8.8 100 3.8.9 C;;% ; fer C'Je1 11, Teal ".GCA-Iw*"*--**^$

1 3.8.9 4.5.9 184 Bases 3.9 Containment Isolation Systems H3 '

}

/

4.9 197 3.9.1 Isolation Valve Seal Water System 4.9.1 197 L

l 3.9.2 Penetration Pressur1 ration Systems 4.9.2 194 3.9.3 Containment Isolation Valves 4.9.3 1994 3.9.4 Main Steam Isolation Valves and Sypasses 4.9.4 200 l

3.9.5 Containment Integrity 4.9.5 201

' Bases ao J' 3.10 Containment Structural Integrity 4.10 212 3.10.1 Containment Leakage Rate Testing 4.10.1 212

.i 3.10.2 Containment Air Locks 4.10.2 214a 3.10.3 Containment Tendons 4.10.3 215 3.10.4 EndAnchoragegandConcrete 4.10.4 217

[

3.10.5 Containment Pressure 4.10.5 219 i

3.10.6-Containment Temperature 4.10.6 219 8ases 2 ro -

l 3.11 Radioactive Ligslds 4.11 222 Bases 3.12 Radioective-9sses 4 esdeas Nd#ACI zz9 -

4.12 230 Bases 2 #2. "

73<. 90-o2 11610 til Ta ^ _1.

a. ;Z " G i

i

Table East 4.8-1 Centrifugal Charging Pump System 185 g

4.8-2 Safety Injection Pump System 186 4.8-3 Residual Heat Removal Pump System 187 4.8-4 Accumulator Tanks 188 5

4.8-5 Component Coollag Pump System 189 4.8-6 Service Mater Pump System 190 4.8-7 Hydrogen Control System 192 4.8.9-1 Accident Monitoring Instrumentation Surveillance Requirements 192K <;,

4.

3-z.

nwdear w mc.- imrnouc><r-4 uns -

14sd -

4.

1 Isolation Seal Mater System 203 4.9-2 Penetration Pressurization System 204

[

i 4.11-1 Radioactive Llquid Effluent Sampling and Analysis 227 Surveillance j

4.11-2 Radioactive Liguld Effluent Monitoring Instrumentation 228b Serve 111ance 4.12-1 Radioactive Gaseous Effluent Samp11ns and Analysis Program 238 l

4.12-2 Radioactive gaseous Effluent Monite

'.'nstrument 240

{

Surveillance 4.14-1 Plant Radiation Monitoring Inst):nentation Surveillance 253 4.15-1 4160-Volt Engineered Safeguard Bus Main, Reserve, and 270 i

Standby Feeds 4.16-1 Maximum Values for the Lower Limits of Detection (LLD) 290 i

'[

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LIST OF TABLES (Continued) l 7sc % -ot I

17510/17560 x

W -Nos H OS-& 95~ t j

. -., ~.

.o.

i 5.0 Gesien Features 296 5.1 Site 296 5.2 Reacter Coolant System 296 s

5.3 Reactor Core 296 I

5.4 Containment System 296 5.5 Feel Storage 290 t

5.6 Seismic Design 299 6.0 M IRISTRATIVE CONTOOLS 300 i

r l

6.1 Organization, Review. Invest 1gatten and Audit....................

300 30' i

6.2 Plant Operating Procedures 300 1 )

f 6.3 Action [tobeTakenintheeventofaReportable

}

Event in Plant Operetten 310 I 6.4 Actlen to be Taken in the Event A Safety Llett is Exceeded 31/

i t

349-l 6.5 Plant Operating Records............................... M i

6.6 Reporting Requirements 312 l l

I 6.7 Offsite Dese Calculation Mensal (OSCM).......................

325 l i

4 i

l 6.0 Fleeding Protectlen

......... ~...................... ".

326 g I

t Tatie ef Contents (Centlemed)

V 06430 tsc 90-02.

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LIMITIfC 00POITION FOR OPERATION SLRVEILLANCE RE9115EPENT 3.8.9 The accident monitoring instrumentation (f.8.9 Each accident monitoring instrumentation channels sh m n.in Table 3.8.9-1 shall be OPERABLE.

channel shall be demonstrated OPEfDELE by performance of the 49efRtM5W CHDSEL CNECK APPLICABILITY:

Modes 1, 2,eed. 3,4e 7.

Ann CHANNEL CALIB8tATION at the frewfes shown in Table 4.8.9-1.

ACTION:

a.

With the number of OTRABLE accident monitoring Instrumentation channels less than the Required Weber of Channels shown in Table 3.8.9-1, (Col. 2)444her restore the inoperab,le channel (s) to OPERABLE status within.seven7 days, or.4ho ba** M f

it m M p';;e4 1. at lee.t t'

)

b E. h [ [""

I) with$nNnext i

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

b.

With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum.mmbeenefoPERA19E j

Channels shown in Table 3.8.9-1, (Col. 3),sitt:: 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 />,1;;;,t t:.; ;"J'or in; _ it k,A N da.

j m . M ;!:::2 in at

}

l

-SMt1YDC -- e, with-T l

.yg e, ( L. O withisy b e55 tr.s.

next 12 nours.

c.

The provisions of Specifications 3.0.4 are not applicable.

• Ini acti does not appi'y o the

_f, Po tion Icati or tit VB

' vaMe i

j siti. Indica on if Ane al Val /on t I

[ assoc ted 11 is kref*;n to clo eit r

i by rifica on wipdn sevfff days r b-ys i-s tus ledge ptfor torindi lon f 11

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184 a==nAmants M and 60 i

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n w-TABLE 3.8.9-1

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ACCIDENT PONITORING INSTRLDENTATION 1.

2.

3.

INSTRAENT (PAR 4ETER)

TOTAL NO.

REQUIRED PC.

MINIMLM NO.

OF CHAPOEL_S 0F CHUGELS OF CHUGELS Containment Pressure (=vLclo 1.

Range) 42-2 1

t 2.

Reactor Coolant Outlet Teaperature - T et (Wide Range) i 4-1/ Loop 2

1 3.'

Reactor Coolant Inlet Temperature - TCold (Wide Range) 4-1/ Loop 2

1 4.

Reactor Coolant Pressure (Wide Range) i 2

2 S.

Steam Line Pressure 1

12-3/SG 2/SG 6.

Pressurizer Water Level 1/SG 3

2 1

7.

Steam Generator Water Level (Narrow Range) 12-3/SG 2/SG 1/SG i

8.

Steam Generator Water Level (Wide Range) 4-1/SG g

p 9.

Refueling Water Storage Tar

• Level 2

2 i

10.

Auxillary Feedwater Flow r

4-1/SG gI 4-pA/4 ll.

Reactor Coolant System Subcooling Margin 2k 2f 1$

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12. PORY Position Indication M df/ valve fg/valvek

13. PORV Block Valve Position Indication T 1/ valve 1/ valve 1/ valve T W l[f*I 14.

Safety Valve Position Indicgtlan*hby ' 'WCMt 3-1/ valve Add I.s - 2. /

Ne %

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V,* b 1 y,/.k Tim ";eeMr Ocal mi precedure performed by the operator.3uve.ooling Margin is uei.et-incu uy L., -di uds;-1) co-yuter-onelysis-or,-2)_a-8%The PORY Position -Indication consists of 1) seewstical w itoring-system-(&g f~'!c2 tim) _ stem-mounted limit switch (primary-indication) and 2) en i

n n 4,J.t wh nz NM *gi k-SL as cl~/+s & w p

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TABLE 4.1

-1 SO*~ [

,g ACCIDENT MONITORING INSTRipENTATION SLRVEILLANCE RE9JIREMENTS CHRPoEL CHWeEL INSTRlp(NT (PAR 4ETER)

DECK CALIERATION u)</4.

1.

Contalruent Pressure Ct;;; Raq)

M R

2.

Reactor Cholant Outlet Temperature - THot (Wide Range)

M R

3.

Reactor Coolant Inlet Temperature - TCold (Wide Range)

M R

4.

Reactor Coolant Pressure (Wide Range)

M R

5.

Steam Line Pressure M

R 6.

Pressurizer Water Level M

R i

7.

Steam Generator Water Level (Narrow range)

M R

l 8.

Steam Generator Water Level (Wide range)

M R

9.

Refueling Water Storage Tank Level M

R I

M R

l 10 Auxiliary Feedwater Flow

11. Reactor Coolant System Subcooling Margin M

R 12 PORY Position Indication I M

R

13. PORY Block Valve Position Indication M M

R Sarety valve Position Indication

• h'-*C//[ofN M*) M R

lo.

al/ /r-a. I l

na. man,anty

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t Bases 3.8 and 4.8 (Continued)

The operability of the service water system ensures The channel calibration of the hydrogen monitors that sufficient cooling capacity is available for requires disassembly and electronic testing, continued operation of safety related equipment adjustment and reassently of the instrument during normal and accident conditions.

In the therefore the refueling cycle frequency is deemed split discharge header operating mode (Unit 1 & 2 adequate.

l with separate discharge headers), two operating pumps and one standby are required for normal The instrumentation, equipment, and procedures for operation.

In the cross-tied discharge header the tests which are required on the ventilation operating mode (Unit 1 & 2 with common discharge filter system will generally conform to the header), three pumps have sufficient capacity for recommendations of ANSI N510-1975.

normal operation but five pumps to be operable, with only'one pump from either unit satisfying this The OPERABILITY of the accident monitoring-requirement for the other unit, are required to instrumentation ensures that sufficient information provide suff1cient redundancy. In an accident or is available on selected plant parameters to shul'down mode, only one pump per unit is required.

monitor and assess these variables during and following an accident. This capability is a

For the service water system, the operability ith the recommendations of Regulatyry

_co q w" Instrumentation for Light-Water-Cooled l

requirements include consideration of the standby vQ_

Guide 1.gh, i

1 AC & DC power supplies so that a single failure of

~ Nuclear Power Plants to Assess Plant Conditions Dece,.J.c

• I the "0" diesel will not cause a common mode failure During and Following an Accident," Secember-4975-N *M' 7

j in the service water system; "0" diesel powers the and NUREG-0578, *TMI-2 Lessons learned Task Forge I

emergency busses for service water pumps lA and 2A.

Status Report and Short-Term Recommendations", <-

% k7) nu N2c 6eme' L&<

0%-37 ~ wur_,ogy, I

A hydrogen Recombiner system 1s installed to remove j eg,% f sp,1.m >

'Q 4 f,fg the hydrogen and oxygen gases that accumulate in

-l the containment atmosphere following a loss-of-coolant accident. (7) The containment (1) FSAR Chapter g hydrogen monitoring system is used to determine the (2) FSAR Section 6.2 effectiveness of this system.

(3) FSAR Section 6.2.3 (4) FSAR Section 14.3 (5) FSAR Section 9.3 j

(6) FSAR Section 9.6 & FSAR answer to question g.1 l

(7) FSAR Section 6.8 b

l

• sendment,-e7-mum 7 -

07670/07680 195

irJ Minimum Ch;nnels'

- Appilcchl2 Instrument Operable

-Action Modes 4.

Auxiliary Building Veutilation and Miscellaneous Ventilation Stack

.A.

Gas Activify~ Monitor 1.

OR-0014 or 1'

6 All 2.

1R-PR25 and 2R-PR25 1

6

.All 3.

OR-PR188 Gas 1

6 All

'4.

1R-PR49E (Channel 5) 1 6

All 5.

2R-PR49E (Channel 5) 1 6

All 8.

Iodine Monitor 1.

1R-PR490 (Channel 3) 1 8

All 2.

2R-PR49C (Channel 3) 1 8

All C.

Particulate Monitor i

1.

OR-PR18A Particulate 1

6 All 1.

1R-PR49A (Channel 1) 1 8

All 2.

2R-PR49A (Channel 1) 1 8

All D.

Flev Rate Monitor 1.

1LP-084 1

9 All 2.

2LP-084 1

9 All

~ 5.

Service Building Ventilation A.

Gas Activity Monitor 1

8 All 1.

OR-PR22 1

8 All 8.

Particulate / Iodine Monitor 1.

OR-PR36 1

8 All 6.

Steam Generator Atmospheric

' Relief and Safety Valves 1.

1R-PR58 1

13 4 1,2,3,7 2.

2R-PR58 1

1 )

1.2.3,7 3.

1R-PR59 1

-1 )

1,2,3,7 4.

2R-PR59 1

1{

1,2,3,7 5.

1R-PR60 1-11 1,2,3,7 6.

2R-PR60 1

1 l

1,2,3,7 7.

1R-PR51 1

1,2,3,7 l

8.

2R-PR61 1

1,2,3,7 Radioactive Gaseous Effluent Monitor Instrumentation (Continued) i Table 3.12-1 (Continued) 7g "n 7.in.g %.t-Mos-$6-and - 8'r l

j 0664t/06651

-236a

^

c Minimum Channels Applicable

. Instrument' Operable; Action Modes 7.

Accident _ Monitoring A.

Containment 1.

1R-PR40G (Channe1 7) 1 10 1,2,3,4,7

2. -2R-PR40G (Channel 7)

.1 10 1,2,3,4,7 3.

1R-PR401 (Channel 9) 1 10 1,2,3,4,7 4.

2R-PR401 (Channel 9) 1 10 1,2,3,4,7 B.

Miscellaneous Vent Stack 1.

1R-PR49G (Channel'7) 1 Y

1,2,3,4,7 2.

2R-PR49G (Channel 7) l' 1,2,3,4,7 3.

1R-PR491 (Channel-9) 1 1,2,3,4,7 4.

2R-PR491 (Channel 9) 1 1,2,3,4,7 C.

Containmeni. Fuel Handling Area Monitor 1.

1R-AR04A 1

11 6

. hen purging during W

l 2.

1R-AR048 1

11 6

fuel handling t

3.

2R-AR04A 1

11 6

operations

(

4.g/R-AR048 1

11 6

+

~

Radioactive Gaseous Effluent Monitor Instrumentation (Continued)

Table 3.12-1 :(Continued) l T5c 9c> c 't

'---t Nos. 96 and 86 0664t/0665t 236b

~

Table..otation 3 kY Aeboh f conW w.rf. A,f:O rO g &.L 83-37

' ACTION 5 - With' the number of channels OPERABLE less than the minimum number required, the contents of the tank may be released to the environment provided that prior to initiating the release:.

1.

At least two independent samples of the tank's content are analyzed, and 2.

At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge flow path valving; otherwise, suspend release of radioactive ef fluents via this pathway.

dACTION6-With the number of channels OPERABLE less than the minimum number required, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per shift and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l ACTION 7 - With the number of channels OPERABLE less than the minimum number required, and no redundant monitor OPERABLE in this flow path, inneediately suspend PURGING of radioactive ef fluents via this pathway.

[

ACTION 8 - With the number of channels OPERABLE less than the minimum number required, ef fluent releases via this pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling l

equipment as required in Table 4.12.1.

ACTION 9 - With the number of OPERABLE channels less than the minimum number required, effluent releases via this pathway may continue provided the flow rate is estimated at least once per shift while release is in progress.

ACTION 10 - With the number of channels OPERABLE less than.the minimum number required, restore.the inoperable monitor to OPERABLE status within 30 days or establish an alternate means of monitoring the parameter.

ACTION 11 - With the number of OPERABLE channels less than the minimum number required, suspend vent and purge operations and close each vent and purge valve providing direct access from the containment atmosphere to the outside atmosphere or suspend the movement of nuclear fuel and reactor components in the vicinity of the reactor, refueling csvity, and transfer canal (containment side).

ACTION 12 - With the number of OPERABLE channels less than the minimum number required, ef fluent releases via this pathway may coitinue provided the effluent flow is being accounted for in the total plant effluent.

Radioactive Gaseous Effluent Monitor Instrumentation (Continued) t TABLE NOTATION Table 3.12-1 75(_ 90 -c> >

-Amendment-Mes-.-96-and '

237 06641/0665t 7

.~:.

~;'.,

~ ::._

~,

Min k m

. Channels Applicable Instrument Operable Action #

Modes 1.

Area Monitors R

A.

Fuel Storage Pool Area 1.

OR-0005 1

24 All

~

2.

OR-ARO3 1

21' During Fuel Handling-Operations or Crane Operation in or near.

SFP.

3.

OR-AR13 1

21

-During Fuel Handling Operations B.

Containment Purge Isolation 1.

1R-AR04A -

1 22 6

When purging during 2.

1R-AR048 1

22 6

fuel handling 3.

2R-AR04A 1

22 6

operations 4

2R-AR048 1

22 6

C.

Containment Area.(High Range) 1.

1R-AR02 1

30 '3l 1,2,3,4,7 2.

2R-AR02<

1 3@,

1,2,3,4.7 L

3.

1R-AR03-1 p

1,2,3,4,7 I

l

-4.

2R-AR03 1

^

30 -

1,2,3,4,7 l

l~

D.

Control Room 1.

OR-0001 1

24 All l

l E.

Technical Support Center a

1.

Portable Area Monitor 1

24 All RADIATION MONITORING INSTRUMENTATION TABLE 3.14-1 17~fC %R J -....L..i

m. 36...J 0707t/0708t 251

x

~a

~

~

Action 27:

With-the number of channels'0PERABEE less than the minimum number regstred, ef fluent via this pathway-may continue provided the gross radioactivity level (beta / gamma or isotopic)-is j.

determined at least once per day. :If 'the inoperable channel is not returned to OPERA 8tE

. status within 30 days conduct a Station Review to detemine a plan of action to restore the -

channel to operability..

Action 28:

With the number of channels OPERABLE less than the minimum number required, comply with'the~

surveillance requirements 4.3.3.A.2 and 4.3.3.8.

Action 30:

With the number of channels OPERABLE less than the minimum number required, initiate an alternate _ method (if feasible) of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and :

1)

Either restore the inoperable Channel (s) to OPERABLE status within 7 days of the event,? or 2)

Conduct a Station Review within 14 days to determine a plan of action to restore the channel to OPERABLE status.

Ah & 3/

t%us,,d an2 AcS h { <5#A /&w BF17 t

TABLE NOTATION (Continued)

' TABLE 3.14-1 (Continued) 75C 1C) C A

^_..b.d ev>. No ano sb -

_.g07t/0708t J52a-

i 6.2.

Plant operatina Procedgrgi 1.

Written procedures including applicable checkoff lists covering items listed below shall be prepared, implemented, and maintained:

A.

Normal startup, operation, and shutdown of the reactor and other systems and components involving nuclear safety of the facility.

B.

Refueling operations.

C.

Actions to be taken to correct specific and foreseen potential malfunctions of systems or components including responses to alarms, suspected primary system leaks, and abnormal reactivity changes.

s O.

Emergency conditions involving potential or actual release of radioactivity

" Generating Stations Emergency Plan" and station emergency and abnormal procedures.

E.

Instrumentation operation which could have an effect on the safety of the facility.

+

F.

Preventive and corrective maintenance operations which could have an effect on the safety of the facility.

8 G.

Surveillance and testing requirements.

j H.

Tests and experiments.

I.

Procedures to ensure safe shutdown of the plant.

J.

Station Security Plan and implementing procedures.

K.

Fire Protection Program implementation.

6.6.VA) b G.uri 4A sy-y y

l..

Post Accident Sampling Program & A M.

Working hours of the Shift Engineer, Shift Control Room Engineer, Shift Foreman, and Nuclear Station Operator such that the heavy use of overtime is not routinely required.

I 2.

RaiM tion control procedures shall be prepared, implemented and maintained. These procedures shall specify permissible radiation exposure limits and shall be consistent with the requirements of 10CFR 20.

The radiation protection program shall meet the requirements of 10CFR 20.

i 73'c jD -oC 309.

-Amendmenr1Rr--ti5-4 104 14870 15460(29) l

.___..a.

c..

W L Tc>)

gl) om

, re.m 6.2 an tinued)

Procedures for items identified in Specification 6.2.1 and any changes to such prncedures shall be reviewed 3.

~

and approved by the Operating Engineer and the Technical Staff Supervisor in the areas of operation and fuel l

handling, and by the Maintenance Assistant Superintendent and Technical Staff Supervisor in the areas of plant maintenance, instrument maintenance, and plant inspection. Procedures for items identified in Specification 6.2.2 and any changes to such procedures shall be reviewed and approved by the Technical Staff Supervisor and the Health Physics Supervisor / Chemistry Supervisor or designees. At least one person In addition.

approving each of the above procedures shall hold a valid Senior Reactor Operator's license.

l these procedures and changes thereto must have the authorization of the Station Manager or designee before.

being implemented.

Mork and instruction type procedures which implement approved maintenance or modification procedures shall be The " Maintenance / Modification Procedure" utilized approved and authorized by the Production Superintendent.

for safety related work shall be so approved only if procedures referenced in the " Maintenance / Modification Procedures which do not fall within the requirements of Procedure" have been approved as required by 6.2.1.

6.2.1 or.6.2.2 may be approved by the Department Heads.

Temporary changes to procedures identified in Specifications 6.2.1 and 6.2.2 above may be made provided:

4.

A.

The intent of the original procedure is not altered.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior B.

Reactor Operator's License on the unit affected.

C.

Tiie change is documented, reviewed by the Onsite' Review and Investigative Function and approved by the Station Manager or designee withire 14 days of implementation.

Drills of the emergency procedures described in Specification 6.2.1.0 shall be conducted at the frequency 5.

specified in the Generatirig Station Emergency Plan. These drills will be planned so that during the course of the year, communication links are tested and outside agencies are contacted.

6.3 Action to be Taken in the Event of a Resortable Event th Plant Operation:

Any Reportable Event shall be promptly reported to the Vice President PMR Operations or his designated alternate.

The incident shall be promptly reviewed pursuant to Specification 6.1.7.8.2.(j) and a separate report for each reportable event shall be prepared in accordance with the requirements of 10CFR 50.73.

7fC ' 9C) - O L _

310

f. r t t N. : : S a.

0' 14870 15460(30)

=

W j

6~.6.3.8 Special Reports (Continued)

TOPIC SUBRITTAL DATE 1.

PressurizerPORVafSafetyValve Document the event in the Annual Report k

l challenees g f j.

(Future) kt Steam geneetor tube inspection and/or Per surve111ance requirements 4.3.1.s.5 plugging. 4/

1.

Emergency p re Cooling System (ECCS)

-Within 30 days as LER - include nozzle usage i

actuation ard injection when RCS temp >

factor per T.S. 3.3.2.F.3

-3 W P G) m.

Fire detector inoperability Within 30 days per T.S. 3.21.1.C n.

Fire pump system inoperability

~ Within 30 days per T.S. 3.21.2.9 c.

Fire suppression system inoperability Within 30 days per T.S. 3.21.2.C or 3.21.2.5 p.

Sprinkler system ino'perability Within 30 days per T.S. 3.21.3.C q.

Low pressure CO2 system inoperability Within 30 days per T.S. 3.21.4.C I

~

(-

%r Acc.<b f ZW "h" rm---

b/A-t.~ H f r %z Tr. i-W s, z./m 7. 47,o swp A gresueviu7a%p A k w/

kl-..s man. g x ne_

n g

SPECIAL REPORTS-Table 6.6-1 (Continued) 11320/11330 324.

Amendment Nos. 103 & 93

\\_

-. - -. -.- - =

= ~ - - -~

^ ~ ~ ' ~ ~ ' ~ '~~'~

^

SURVEILLANCE ~

LIMITING CONDITION FOR OPERATION REOUIREMENT PAGE 3.7

. Steam Generator Emergency Heat Removal 4.7 156

'3.7.1 Steam Generator Safety Valves-4.7.1 156 3.7.2 ' Auxiliary:Feedwater Pump. System 4.7.2 158 3.7.3 Aux 111ary Feedwater Supply System 4.7.3 159a l

' Bases-162 3.8 Emergency Core Cooling and Core Cooling Support 4.8 164 3.8.1 Centrifugal Charging Pump System 4,3.1 164 3.8.2 Safety Injection Pump System 4.3.2 1 68 3.8.3 Residual Heat Removal Pump System 4.8.3 170 3.8.4 System Testing of Centrifugal Charging, Safety 4.8.4 173 Injection, and Residual Heat Removal Pump Systems 3.8.5 Accumulator System 4.8.5 174 3.8.6 Component _ Cooling System 4.8.6 175 3.8.7 Service Hater System 4.8.7 178 3.8.8 Hydrogen Control Systems 4.8.8 180 3.8.9 Accident Monitoring Instrumentation 4.8.9 184 Bases 193 3.9 Containment Isolation Systems 4.9 197 3.9.1 Isolation Valve Seal. Hater System 4.9.1 197 3.9.2 Penetration Pressurization Systems

-4.9.2 198 3.9.3 Containment Isolation Valves 4.9.3 199a 3.9.4 Main Steam Isolation Valves and Bypasses 4.9.4 200 3.9.5 Containment Integrity 4.9.5 201 l

Bases 209 3.10 Containment Structural Integrity 4.10 212 3.10.1 Containment Leakage Rate Testing 4.10.1 212 3.10.2 Containment Air Locks 4.10.2 214a 3.10.3 Containment. Tendons 4.10.3 215 3.10.4 End Anchorage and Concrete 4.10.4 217 3.10.5 Containment Pressure-4.10.5 219 3.10.6 Containment Temperature 4.10.6 219 l

Bases 220 3.11 Radioactive Liquids 4.11

'222 Bases 229 3.12 Gaseous Effluents 4.12 230 Bases 242 TABLE OF CONTENTS (Continued)

~18760 til TSC 90-02

4 5.0 Desian Features 296-5.1 Site 296 5.2 Reactor Coolant System..........-.............-.....-....

296 5.3 Reactor Core 296 296 5.4 Containment System 298 5.5 Fuel Storage 299 5.6 Seismic Design 300 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization, Review, Investigation and Audit....................

300 309 l

6.2 Plant Operating Procedures l

6.3 Action to be Taken in the event of a Reportable Event in Plant Operation 310.

311 6.4 Action to be Taken in the Event A Safety Limit is Exceded I

(

l 6.5 Plant Operating Records.............-..................

311

.312 6.6 Reporting Requirements 325 6.7 Offsite Dose Calculation Manual (00CM) 326 6.8 Flooding Protection Table of Contents (Continued) v 18760 TSC 90-02 w

,_.. a u.

u..

a

~'

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 4

I l

ACCIDENT MONITORING INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTAT10H-l 3.8.9 The accident monitoring instrumentation 4.8.9 Each accident _ monitoring _ instrumentation channels shown in Table 3.8.9-1 shall be channel shall be demonstrated OPERABLE by CPERABLE.

-performance-of the CHANNEL CHECK and Instrument.

CHANNEL CALIBRATION operatio.is at the l

APPLICABILITY:

MODES 1, 2, 3 and 7 frequencies shown in Table'4.8.9-1.

ACTION:

a.

Nith the number of OPERABLE' accident monitoring instrument channels less than the Rgquired Number of Channels shown in Table 3.8.9-1, (Col. 2), restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least MODE 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

With the number of'0PERABLE accident monitoring instrumentation chanriels less than the Minimum Ooerable Channels requirements of Table.3.8.9-1,~(Col. 3),

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 be in at least MODE 4 within'the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l l

c.

The provisions'of Specifications 3.0.4 are not applicable.

j i

i f-t i

f

'1 18760 184?

-TSC 90-02

-l

.l

,m.

u, n c,

n,,-.-c

- ~,.

n

~

m.

. ~.

i a

Table Eagit 4.8-1 Centrifugal Charging Pump System

-185 4.8-2 Safety Injection Pump System 186 4.8-3 Residual Heat Removal Pump System 187 4.8-4 Accumulator Tanks 188-4.8-5 Component Cooling Pump System 189 4.8-6 Service Water Pump System 190 4.8-7 Hydrogen Control System 192 l

4.8.9-1 Accident Monitoring-Instrumentation Surveillance Requirements 192c l

4.8.9-2 Accident Monitoring Instrument Numbers 192d

,.9-1 Isolation Seal Water System 203 4

4.9-2 Penetration Pressurization System 204 4.11-1 Radioactive Liquid Effluent Sampling and Analysis 227 Surveillance 4.11-2 Radioactive Liquid Effluent Monitoring Instrumentation 2285 Surveillance 4.12-1 Radioactive Gaseous Effluent Sampling and Analysis Program 238 4.12-2 Radioactive Gaseous Effluent Monitoring Instrumentation 240 Surveillance.

4.14-1 Plant Radiation Monitoring Instrumentation Surveillance 253 Requirements.

4.15-1 4160-Volt Engineered Safeguard Bus Main, Reserve, and 270 Standby Feeds 4.16-1 Maximum Values for the' Lower Limits of Detection (LLD) 280 List of Tables (Continued) 18760 x

'TSC 90-02

1.

2.

3.

TOTAL MO.

REQUIRED NO.

MINIMUM OPERABLE INSTRUMENT (PARAMETER)

OF CHANNELS OF CHANNELS

-CHANNELS l

1. Containment Pressure (Hide Range) 2 2

1

2. Reactor Coolant Outlet T9mperature - THOT (Hide Range) 4-(1/ loop) 2 1

3. Reactor Coolant Inlet Temperature - TCOLD (Hide Range) 4-(1/ loop) 2 1

4. Reactor Coolant Pressure (Hide Range) 2 2

1

5. Steam Line Pressure 12-(3/SG) 2/SG 1/SG

6. Pressurizer Water Level 3

2 1-

7. Steam Generator Water Level (Narrow Range) 12-(3/SG) 2/SG 1/SG l

8. Steam Generator Water Level (Wide Range) 4-(1/SG) 1/SG N/A

9. Refueling Hater Storage Tank Level.

2 2

1 l

l

10. Auxillary Feedwater Flow Rate 4-(1/SG) 1/SG N/A l

11. Reactor Coolant System Subcooling Margin 2

2 1

Accident Monitorina Instrumentation Table 3.8.9-1 18760

.192a

.TSC 90-02

~

1.

2.

13.

TOTAL NO.

REQUIRED NO.

MINIMUM OPERABLE INSTRUMENT 1 PARAMETER) 0F CHANNELS OF~ CHANNELS CHANNELS

12. PZR'PORV Position Indicator

• 1/ valve

'l/ valve

• 1/ valve *

13. PZR PORV Block Valve Position Indicator *

'1/ valve 1/ valve

• 1/ valve *

14. PZR Safety Valve Position Indicator **(Primary: Temperature Detectors) 3-(1/ valve) 1/ valve 1/ valve 3

15. PZR Safety Valve Position Indicator (Backup:

4 Acoustic Monitors) 3-(1/ valve) 1/ valve

.M/A

16. Core Exit Thermocouples 65 4/ core quadrant 2/ core quadrant'

17. Containment Water Level (Narrow Range) 2 l#

l#

18. Containment Hater Level (Hide Range) 2 2

1

19. Reactor Vessel Water Level (Hide Range)

[ t least one RCP running]

2 2

1

20. Reactor Vessel Water Level-(Narrow Range)

[All RCPs not running]

2 2

1

21. Condensate Storage Tank Level 2

2 1

• Not required if the PZR PORV Block Valve is closed and de-energized in accordance with specification 3.3.1.F.

• • Direct indication of PZR Safety Valve Position - NUREG 0578, Item 2.1.3.a.

1. Operation may continue up to 30 days with less than the REQUIRED NO. OF CHANNELS or less than the MINIMUM t

OPERABLE CHANNELS.

Accident Monitorina Instrumentation (Continued)

Table 3.8.9-1 (Continued) t 18760/11880 192b TSC 90-02

=

CHANNEL' CHANNEL'.

INSTRUMENT (PARAMETER)

CHECK CALIBRATION ll'

1. Containment Pressure (Hide Range)

M R-

2. Reactor Coolant' Outlet Temperature - THOT (Hide Range)

-M R.

~

3. Reactor Coolant Inlet Temperature - TCOLD.(Hide Range)

M R;

4. Reactor Coolant Pressure (Hide Range)

.M R

5. Steam Line Pressure

.M R'

6. Pressurizer Water Level M

R

7. Steam Generator Water Level (Narrow Range)

M R

8. Steam Generator Hater Level (Hide Range)

M R

9. Refueling Water Storage Tank Level M

R

10. Auxiliary Feedwater Flow Rate M

R 11 Reactor Coolant System Subcooling Margin M

R

12. PZR PORV Position Indicator *

'M R

13. PZR PORV Block Valve Position Indicator

• M R

14. PZR Safety Valve Postilon. Indicator **(Primary: Temperature Detectors)

M R

IS. PZR Safety Valve Posi"tien Indicator **(Backup: Acoustic Monitors)

M R

16. Core Exit Thermocouples M

R

17. Containment Water Level (Narrow Range)

M R

18. Containment Water Level (Hide Range)

M R

19. Reactor Vessel Water Level (Hide Range)

M R

20. Reactor Vessel Water Level (Narrow Range)

M R

21. Condensate Storage Tank Level M

R Not required if the PZR PORV Block Valve is closed and de-energized in accordance with Specification 3.3.1.F.

l Direct indication of PZR Safety Valve Position - NUREG 0578, Item 2.1.3.a.

Accident Monitorina Instrumentation Surveillance Reautrements.

Table 4.8.9-1 18760 192c TSC 90-02 a.

O

~

ACCIDENT PARAMETER SENSOR INSTRUMENT NUMBER

1. Containment Pressure (Hide Range)

PT-CS50, PT-CSSI

2. Reactor Coolant Outlet. Temperature - THOT (Hide Range)

TE-413A, TE-423A, TE-433A, TE-443A

3. Reactor Coolant Inlet Temperature - TOOLD (Hide Range)

TE-4138 TF-4238. TE-4338. TE-443B

4. Reactor Coolant Pressure (Hide Range)

PT-403, PT-405

5. Steam Line Pressure PT-514, PT-515 PT-516, PT-524, PT-525, PT-526, PT-534, PT-535, PT-536, PT-544, PT-545,-PT-546

6. Pressurizer Water Level LT-459, LT-460, LT-461

7. Steam Generator Water Level (Narrow Range)

LT-517,'LT-518, LT-519, LT-527, LT-528,

.LT-529, LT-537, Lt-538, LT-539, LT-547, LT-548, LT-549

8. Steam Generator Water Level (Hide Range)

LT-501, LT-502, LT-503, LT-504

9. Refueling Hater Storage Tank Level LT-920 LT-921

10. Auxiliary Feedwater Flow Rate FT-FH02, FT-FH03, FT-FH04, FT-FW25

11. Reactor Coolant System Subcooling Margin RCC01A, RCC01B Pressure Inputs: FT-403, PT-405, PT-456, PT-457, PT-458.

Temperature Inputs: T-0001 thru T-0065'

12. PZR PORV Position Indicator Valve Limit Switches

13. PZR PORV Block Valve Position Indicator Valve Limit Switches

14. PZR Safety Valv'e Position Indicator (Primary: Temperature Detectors)

TE-464, TE-465 TE-466 Accident Monitorina Instrument N::mbers Table 4.8.9-2 18760 192d TSC 90-02 u

_________,u

ym@MRx:

~.

=-

w-

~

~[

-ACCIDENT PARAMETER

SENSOR INSTRUMENT NUMBER 15.:PZR Safety Valve Position Indicator-(Backup: ' Acoustic Monitors)-

Acoustic Monitors LLO25, LLO26, LLO27 ^

16. Core Exit Thermocouples T-0001 thru T-0065

17. Containment Water Level (Narrow Range)

LT-CS48, LT-CS49

18. Co'ntainment Water Level (Wide Range)

LT-CS46, LT-CS47

19. Reactor Vessel Water Level (Wide Range)

LT-1321, LT-1322

20. Reactor Vessel Water Level (Narrow Range)

LT-1311, LT-1312

21. Condensate Storage Tank Level LT-CD94, LT-CD139 Accident Monitorina Instrument Numbers

-Table 4.8.9-2.(continued) 18760

-192eT TSC 90-02

?

~

^

Bases 3.8 and 4.8.(Continued)

The OPERABILITY of the Service Hater System ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions.

In the split discharge header.

OPERATING mode (Unit 1 & 2 with separate discharge headers), two OPERATING pumps and one standby are required for normal operation.

In the cross-tied discharge header OPERATING mode (Unit 1 & 2 with common discharge header). -

three pumps have sufficient capacity for normal operation, but five pumps being OPERABLE with only one pump from either unit satisfying this requirement for the other unit, are required to provide sufficient redundancy.

In an accident or shutdown mode,.only one pump per unit.is required.

For the Service Hater System, the OPERABILITY requirements include consideration of the standby AC & DC power supplies so that a single failure of the "0" diesel will not cause a common mode failure in the Service Water System; "0" diesel powers the emergency busses for Service Water pumps 1A and 2A.

A Hydrogen Recombiner System is installed to remove the hydrogen and oxygen gases that accumulate in the containment atmosphere following a loss-of-coolant accident.

(7) The containment Hydrogen Monitoring System is used to determine the effectiveness of this system.

The CHANNEL CALIBRATION of theihydrogen monitors. requires disassembly'and electronic testing, adjustment and-

, reassembly of the instrument, therefore the REFUELING CYCLE frequency is deemed adequate.

The instrumentation, equipment, and procedures for the tests which are required on the ventilation filter system will generally conform to the recommendations of ANSI M510-1975.

The OPERABILITY of the accident monitoring instrumentation. ensures that sufficient information is avaliable on selected plant parameters to monitor and assess these variables during and following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97 Rev. 2, " Instrumentation for Light-Hater-Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1980, NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations", July 1979.and NRC Generic Letter 83-37 "NUREG-0737 Technical Specifications," November 1, 1983.

(1)

FSAR Chapter 9 v

(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 6.8 18760 195 TSC 90-02

~'

. 5. -

L

. ~ - ~

Minimum Channels ~

-Applicable Instrument Ooerable-

. A<:ftlpJ '

Modes

4.

AuxillkryBuildingVentilation and Miscellaneous Ventilation' Stack-A.

Gas Activity Monito*-

1.

OR-0014 or 1

6 All 2.

1R-PR25 and 2R-PR25 1

6 All 3.

OR-PR18B Gas 1

6 All 4.

'1R-PR49E (Channel 5)-

1 6

All 5.

2R-PR49E (Char:nel 5) 1 6

All B.

Iodine Monitor 1.

IR-PR49C (Channel 3) 1 8

All 2.

2R-PR49C (Channel 3) 1-8 All C.

Particulate Monttor 1.

OR-PR18A Particulate 1

6 All 1.

iR-PR49A (Channel 1) 1 8

All 2.

2R-PR49A (Channel 1) 1 8 ~

All D.

Flow Rate Monitor 1.

ILP-084 1

9 All i

2.

2LP-084 1

9 All 5.

Service Buildina Ventilation A.

Gas Activity Monitor 1.

OR-PR22 1

8 All B.

Particulate / Iodine Monitor 1.

OR-PR36 1

8 All 6.

Steam Generator Atmospheric Relief and Safety Valves 1.

1R-PR58 1

4 1,2,3,7 2.

2R-PR58 1

4 1,2,3,7 3.

1R-PR59 1

4 1,2,3,1 4.

2R-PR59 1

4 1,2,3,7 5.

1R-PR60 1

4 1,2,3,7 6.

2R-PR60 1

4 1,2,3,7 7.

IR-PR61 1

4 1,2,3,7 8.

2R-PR61 1

4 1,2,3,7 Radioactive Gaseous Effluent Monitor Instrumentation (Continued)

Table 3.12-1.(Continued)

TSC 90-02 236a 18760

m,e y.

Minimum Channels.

' Applicable Instrument Ooerable-Action Modes

'7.

Accident Monitorina A.

Containment 1.

1R-PR40G (Channel 7) 1 10 1,2,3,4,7 2.

2R-PR40G (Channel 7) 1 10 1,2,3,4,7 3.

1R-PR40I.(Channel-9) 1 10 1,2,3,4,7-4.

2R-PR40I (Channel 9) 1 10 1,2,3,4,7 B.

Miscellaneous Vent Stack 1.

IR-PR49G (Channel 7)

'l 4

1,2,3,4,7 2.

2R-PR49G (Channel 7) 1 4

1,2,3,4,7 3.

IR-PR49I'(Channel 9) 1 4

1,2,3,4,7 4.

2R-PR49I (Channel 9) 1 4

1,2,3,4,7 C.

Containment Fuel Handling Area Monitor 1.

1R-AR04A 1

11 6

When purging during 2.

1R-AR048 1

11 6

fuel handling 3.

2R-AR04A'

'l 11 6

operations l

4.

2R-AR04B 1

11 6

i l

l l

i Radioactive Gaseous Effluent Monitor Instrumentation (Continued) l Table 3.12-1(Continued)

(

'TSC 90-02 18760 236b

..u....

_.=

. a-ACTION 4 - Hith the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirements, Initiate an alternate method of monitoring the. appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1.

either restore the inoperable Channel (s) to OPERABLE status within 7 days of the event, or 2

prepare and submit a Special Report to the Commission pursuant to Specification 6.6.3.B within 14 days following the event outlining the-action taken, the cause of the inoperability and the plans and schedule for-restoring the system to OPERABLE status.

ACTION 5 - Hith the number of channels OPERABLE less than the minimum number required, the contents of the tank may be released to the environment provided that prior to initiating the release:

1.

At least two independent samples of the tank's content are analyzed, and 2.

At least two technically quallfled members of the facility staff independently verify the release rate calculations and discharge flow path valving; otherwise suspend release of radioactive effluents via this pathway.

ACTION 6 - Hith the number of channels OPERABLE less than the minimum number required, effluent releases via this pathway may continue for up to 30 days provided grab samples are taken at least once per shift and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

ACTION 7 - Hith the number of channels OPERABLE less than the minimum number required, and no redundant monitor OPERABLE in this flow path, immediately suspend PURGING of radioactive effluents via this pathway.

ACTION 8 - With the number of channels OPERABLE less.than the minimum number required, effluent releases via this pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 4.12.1.

ACTION 9 - Hith the number of OPERABLE channels less than the minimum number required, effluent releases via this pathway may continue provided the flow rate is estimated at least once per shift while release is in progress.

ACTION 10 - With the number of OPERABLE channels less than the minimum number required, restore the inoperable monitor to OPERABLE status within 30 days or establish an alternate means of monitoring the parameter.

l ACTION 11 - Hith the number of OPERABLE channels less than the minimum number required, suspend vent and purge operations and close each vent and purge valve providing direct access from the containment atmosphere to the outside atmosphere or suspend the movement of nuclear fuel and reactor components in the vicinity of the reactor, refueling cavity, and transfer Canal (containment side).

ACTION 12 - With tan number of OPERABLE channels less than the minimum number required, effluent releases via this pathway may continue provided the effluent flow is being accounted for in the total plant effluent.

Radioactive Gaseous Effluent Monitor Instrumentation (Continued)

Table Notation Table 3.12-1 (Continued) 18760 237.

TSC 90-02

i D Minimum Channels Applicable Instrument Ooerable Action #

Modes

~

1.

Area' Monitors

-A.

Fuel -Storage Pool Area 1.

OR-0005 1

24

.All 2.

OR-AR03 1-21 During Fuel Handling i

Operations or Crane Operation in or near SFP.

3.

OR-AR13 1

21 During Fuel Handling Operations B.

Containment Purge Isolation 1.

'1R-AR04A 1

22 ~

6 When purging during 2.

1R-AR048

-1 22 6

fuel handling.

3.

2R-AR04A 1

22 6

operations 4.

2R-AR04B 1

22 6

C.

Containment Area.(High Range) 1.

1R-AR02 1

31 1,2,3,4,7 2.

2R-AR02 1

31 1,2,3,4,7 3.

IR-AR03 1

31 1,2,3,4,7 l

4.

2R-AR03 1

31 1,2,3,4,7 D.

Control Room

'1.

OR-0001 1

24 All E.

Technical Support Center 1.

Portable Area Monitor-

-1 24 All RADIATION MONITORING-INSTRUMENTATION TABLE 3.14-1 TSC.90-02 18760 251

.m.

Action 27:

Hith.the number of channels OPERABLE 'less than~ the minimum number required, effluent via this pathway may continue:provided the-gross radioactivity level (beta / gamma or isotopic) is determined at least once per day.

If the Inoperable channel'Is not returned to OPERABLE 1 status within 30 days conduct a Station Review to determine a plan of action to restore the channelLto operability.

Action 28:

lHith the number of channels 0PERABLE~1ess.than the minimum number required.. comply with the surveillance requirements 4.3.3.A.2 and 4.323.B.

' Action 30:

Hith the number of channels OPERABLE-less than the minimum number required.. Initiate an alternate-method-(if feasible) of monitoring the appropriate parameter (s).wlthin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and :

1)

Either restore the inoperable Channel (s) to OPERABLE status with1r. 7 days of the event, or 2)

Conduct a' Station Review within 14 days to. determine a plan of action to restore the channel.

to OPERABLE status.

ACTION 31:

With the number of OPERABLE Channels less than required by-the Minimum Channels OPERABLE requirements, initiate an alternate method of monitoring the appropriate parameter (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and:

1.

either restore the. Inoperable Channel (s) to OPERABLE status within 7. days of the event, or 2.

prepare and submit a'Special Report'to the Commission pursuant to Specification 6.6.3.B within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

TABLE NOTATION (Continued)

TABLE 3.14-1 (Continued)

TSC 90-02 18760 252a:

E,,

,,c t,-w4 y

-~h

~,--r ar

• w--

4re v--

+%',,,-*

mw-e E'

6.2.

Plant Ooeratina Procedures 1.

Written procedures including applicable checkoff lists covering items listed below shall be prepared, implemented, and maintained:

A.

Normal startup, operation, and shutdown of the reactor and other systems and components involving nuclear safety of the facility.

B.

Refueling operations.

C.

Actions to be taken to correct specific and foreseen potential malfunctions of systems or components including responses to alarms, suspected primary system leaks, and abnormal reactivity changes.

D.

Emergency conditions involving potential or actual release of radioactivity

" Generating Stations Emergency Plan" and station emergency and abnormal procedures.

E.

Instrumentation operation which could have an effect on the safety of the facility.

F.

Preventive and corrective maintenance operations which could have an effect on the safety of the facility.

G.

Surveillance and testing requirements.

H.

Tests and experiments.

I.

Procedures to ensure safe shutdown of the plant.

J.

Station Security Plan and implementing procedures.

K.

Fire Protection' Program implementation.

t L.

Post Accident Sampling Program which will ensure the capability to: obtain and analyze reactor coolant and containment atmosphere samples, and collect and analyze or measure radioactive lodine and-particulates in plant gaseous effluents under accident conditions. The program shall include the following:

(1) Training of personnel, (11) Procedures for sampling and analysis, (iii) Provisions for maintenance of sampling and analysis equipment.

M.

Working hours of the Shift Engineer, Shift Control Room Engineer, Shift Foreman, and Nuclear Station-Operator such that the heavy use of overtime is not routinely required.

18760 309

.TSC 90-02

~.

6.2 (Continued) 2.

Radiation control procedures shall be prepared, implemented and maintained. These procedures shall specify permissible radiation exposure limits and shall be consistent with the requirements of 10CFR 20.

The radiation protection procram shall meet the requirements of 10CFR 20.

3.

Proc.edures for items identified in Specification 6.2.1 and any changes to such procedures shall be reviewed and approved by the Operating Engineer and the Technical Staff Supervisor in the areas of operation and fuel handling, and by the Maintenance Assistant Superintendent and Technical Staff Supervisor in the areas of plant maintenance, instrument maintenance, and plant inspection. Procedures for items identified in Specification 6.2.2 and any changes to such procedures shall be reviewed and approved by the Technical Staff Supervisor and the Health Physics Supervisor / Chemistry Supervisor or designees. At least one person approving each of the above procedures shall hold a valid Senior Reactor Operator's license. In addition, these procedures and changes thereto must have the authorization of the Station Manager or designee before i

being implemented.

Work and instruction type procedures which implement approved maintenance or modification procedures shall be approved and authorized by the Production Superintendent. The " Maintenance / Modification Procedure" utilized for safety related work shall be so approved only if-procedures referenced in the " Maintenance / Modification Procedure" have been approved as required by 6.2.1.

Procedures which do not fall within the requirements of 6.2.1 or 6.2.2 may be approved by the Department Heads.

4.

Temporary changes to procedures identified in Specifications 6.2.1 and 6.2.2 above may be made provided:

A.

The intent of the original procedure is not altered.

l B.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

l C.

The change is documented, reviewed by the Onsite Review and Investigative Function and approved by the Station Manager or designee within 14 days of implementation.

5.

Drills of the emergency procedures described in Specification 6.2.1.0 shall be conducted at the frequency specified in the Generating Station Emergency Plan. These dril, will be planned so that during the course of the year, communication links are tested and outside agencies are contacted.

6.3 Action to be Taken in the Event of a Reportable Event'in Plant Ooeration:

Any Reportable Event shall be promptly reported to the Vice President PWR Operations or his designated alternate.

The incident shall be promptly reviewed pursuant to Specification 6.1.7.B.2.(j) and a separate report for each reportable event.shall be prepared in accordance with the requirements of 10CFR 50.73.

18760 310 TSC 90-02

6.6.3.B Soecial Reports (Continued)

TOPIC SUBMITTALJATE 1.

Pressurizer PORV or Safety Valve challences Document the event in the Aisnual Report j.

(Future) k.

Steam generator tube inspection and/or Per surveillance requirements 4.3.1.B.5 plugging.

1.

Emergency Core Cooling System (ECCS)

Hithin 30 days as LER - include nozzle usage actuation and injection when RCS temp 2 factor per T.S. 3.3.2.F.3 350*F m.

Fire detector inoperability Within 30 days per T.S. 3.21.1.C n.

Fire pump system inoperability Nithin 30 days per T.S. 3.21.2.B o.

Fire suppression system inoperability Within 30 days per T.S. 3.21.2.C or 3.21.2.D l

p.

Sprinkler system inoperability Hithin 30 days per T.S. 3.21.3.C l

q.

Low pressure CO2 system inoperability Within 30 days per T.S. 3.21.4.C r.

Post Accident Radiation monitor Hithin 14 days per T.S. Tables 3.12-1 and 3.14-1 to the inoperable greater than 7 days.

Regional Administrator of the NRC Regional Office.

SPECIAL REPORTS Table 6.6-1 (Continued) 18760

-324 TSC 90-02

i ATTACHMENT 6 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS t

For ease of review, the proposed changes have been broken down into 5 j

' distinct categories, based on the changes made.

The No Significant Hazards Considerations has been broken down into separate assessments, to reflect these categories, and to appropriately address all the changes made.

The alphabetic designators following the categorical

- title, designates the corresponding proposed change that is addressed in the No Significant Hazards Considerations.

Commonwealth Edison has evaluated this proposed amendment and l

determined that it involves no significant hazards considerations.

t':ording to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

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e ADMINISTRATIVE CHANGES (A)

)

The following changes have been categorized as administrative in-nature.

These changes do not involve a change in intent, requirement, or regulation.

As such, this No Significant Hazards Considerations 1

will address these items:

i Page number changes and additions, Title changes, relocations, and additions,

. Location and wording of notes associated with.the PORVs and PORV Block Valves, Clarified proper mode to eliminate-ambiguity Changes in terminology, Generation of new table to be used for reference purposes,

- Addition of notes clarifying current exemptions.to performing periodic surveillances,

?

Changes providing reference to base documents and revision

numbers, a

Deletion of surveillance frequency notation addressed 1

appropriately elsewhere in the Technical Specifications, Reformating of pages,

- Equipment identification changes that are typographical in nature, and Inclusion of reporting requirements into the administrative-section-of the Technical Specifications, required to be performed under an action statement.

i The proposed changes do not result in a significant increase in the probability or consequences of accidents previously evaluated.

The proposed changes to the~ Technical Specification listed above have been-determined to be administrative in nature. -The proposed amendment to y

the Technical Specifications does not change or alter any current operator actions, or requirements for the mitigation of any previously l:

evaluated accident. These changes have no impact on assumed margins 1

l, or actions during an evaluated accident.

Plant response to previously evaluated accidents will not be altered as a result of these changes.

I__

As such, the probability _and consequences associated with evaluated accidents have remained unchanged.

The proposed changes do not create _the possibility of a new or different kind of accident from any accident previously evaluated.

1 1

These proposed changes will not impose or result in plant operation that differ from any existing requirements.

No new equipment is being L

introduced as a result of these changes.

As such, the possibility of-L a new or different kind of accident from any accident previously evaluated will not occur as a result of these changes.

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The proposed changes do not involve a significant reduction in a margin of safety.

These changes do not alter the manner in which equipment required for safe operation of the plant is operated.

There are no setpoint, or operational limitations being altered or changed e

as a result of these revisions. As such, these changes have no l

significant effect on the margiu of safety.

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o ADDITION OR CHANGES OF INSTRUMENTATION (B)

The following changes involve the addition of instrumentation as a result of the guidance given in, Generic Letter 83-37 regarding "NUREG 0737 Technical Specifications", and NRC Memorandum D.M. Crutchfield to T.M. Novak regarding relief and safety valve position indication.

In addition, instrumentation has been added and/or changed to assure that all Type A variables are addressed, and reference the appropriate qualified instrumentation. As such, this No Significant Hazards Considerations will address the following items:

Containment Pressure (Hide Range) changing from narrow range to wide range.

In addition, the total number of instrument channels has been changed from 4 to 2 to address the appropriate number of qualified instruments.

Changing of the Pressurizer Safety Valve Position Indicator to include both a Technical Specification for the primary and the backup indicators.

Revision e' the RCS Subcooling Monitor, to delete previously approved manual calculations in lieu of utilization of installed instrumentation, and The addition of the following instrumentation into the Technical Specifications:

Core Exit Thermocouples, Containment Water Level Hide Range,

'Tntainment Water Level Narrow Range, Reactor Coolant Inventory Tracking System, and Condensate Storage Tank Level.

The proposed change does not result in a significant increase in the probability or consequences of accioents previously evaluated.

The purpose of the Accident Monitoring Instrumentation is to provide i

l sufficient information to perform required manual actions, and to monitor and assess plant status and behavior during and following an i

accident.

The instrumentatien referenced in the Technical Specifications to perform these functions must be capable of providing this information under assumed accident conditions.

Therefore, there must both be an adequate number of parameters monitored, ano these parameters must be monitored by appropriately qualified instrumentation.

The probability for an evaluated accident is independent of the number and type of instruments designated for monitoring purposes.

The probability for an accident is linked to the precursor events leading to an accident.

None of the accident monitoring instrument changes addressed here are linked in any fashion to these precursor events.

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As such, the probability for previously evaluated events have remained unchanged.

The consequences of evaluated events are lessened by t

assuring that the minimum number of parameters required to perform required manual actions, ano monitor and assess plant status following an accident are available.

The proposed changes do not create the possibility of a new or I

different kind of accident from any accident previously evaluated.

The proposed changes will not approve or result in plant operation i

that could create a new or different kind of accident.

The addition of these instruments to the technical specifications will provide assurance in regards to their availability, through the establishment of Limiting Conditions for Operation, Remedial Actions, and appror fate Surveillance Requirements.

The accident monitoring instrucantation is directly tied to accident mitigation, through the accomplishment of manual actions and assessment.

Through proper decision making the possibility of different type of accident or accident conditions propagating in a differing manner will be prevented.

The proposed change does not involve a significant reduction in a margin of safety. The inclusion of these instruments into the Technical Specification will provide assurance that an adequate number i

of the appropriate instruments are available to monitor required plant parameters in the event of an accident.

By referencing instrumentation that is appropriately qualified for the expected condition, assurance is gained that these parameters will be available

+

if required.

By reducing the total number of channels for the Containment Hide Range Pressure Instruments from 4 to 2 the appropriately qualified instruments are reference.

However, the margin of safety has remained unchanged by virtue d the required number of channels remaining the same.

For the instrumentation being added, these changes will enhance the safety of the plant through the establishment of minimum acceptable levels of performance and availability for these instruments in the Technical Specifications.

Through the establishment of these minimum levels, information required for appropriate decision making during and following an accident can be assured.

Surveillance requirements are being specified for the purpose of determining operability.

The remedial actions associated with the added Specifications will require the plant to be placed into a shutdown condition, after providing an appropriate time frame to restore the inoperable instrument (s) to operable status.

As such, these proposed changes to the Technical Specifications will resu't in an increase in the current margin of safety.

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ACTION AND PROGRAMATIC CHANGES (C) l i

The foliowing assessment involves the revisiok of the action statements associattd with the accident radiation monitoring instruments.

This assessment also addresses ie addition of i

programatic r.ontrol to assure these mGnitort-c capabilities. As such, I

this No Significant Hazards Considerations will address the following items:

s Addition of action statement 4 regarding the actions associated with the Steam Generator Atmospheric Relief and Safety Valve Radiation Monitors, and the Vent Stack Noble Gas k;diation

Monitors, Addition of action statement 31 regarding the actions associated with the Containment High Range Radiation Monitors, and The addition of the requirement to have Post Accident sampling programs' capable of obtaining and analyzing reactor coolant and containnent atmosphere, and collect and analyze or measure radioactive iodines and particulates in plant gaseous effluents samples under accident conditions.

The proposed changes do not result in a significant ' increase in the probability or consequences of accidents previously evaluated.

The radiation monitor action statement changes and programs being added to the administrative section of the Technical Specifications of i

themselves have no impact on the probability of any events that are e

' inputs to evaluated accidents. As such, the probability for evaluated events has remained the sanie.

The changes made are providing i

assurance to the availability of these monitors or alternative methods of providing comparable information during and following an accident.

Through assering the availability of these parameters and programs, the consequences of evaluated accidents remain the same.

The prc4 posed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes address the revisions to programs and action statements.

There is no new equipment being proposed for installation that could present the possibility for a new or different type of accident, as a result of these changes.

Plant operations will not be altered in any fashion that could create the possibility for a new or different type of accident.

Through assuring the availability of these parameters during an accident, proper decision making can be assured. As such, these changes will not create the possibility for a new or different type of accident.

In addition, the potential for accident conditions propagating in a differing menner will be prevented.

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-The proposed change does not involve a significant reduction in a margin of safety.

The proposed actions will result in the establishment of alternate methods of monitoring the effluent release paths for the Steam Generator Atmospheric Relief and Safety Valves, I

and Vent Stack within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> instead of 30 days as addressed in the current Technical Specifications.

The reduction in these time frames will provide for an increase in the readiness to monitor these parameters in 'the unilkely event of an accident.

The administrative changes being added to the Technical Specifications involve the addition of the requirements to have Post Accident sampling programs capable of obtaining and analyzing reactor coolant and containment atmosphere, and collect and analyze or measure radioactive iodines and

=

particulates in plant gaseous effluents samples under accident conditions. These programs include; the training of personnel, 1

b procedures for sampling and analysis, and provisions for maintenance of sampling and analysis equipment.

These changes provide assurance that these capabilities and programs will be established and maintained.

The results of these changes are viewed as an improvement in the plants overall ability.to assess accident conditions, and the potential for or quantification of releases. As such, these changes do not result in a significant reduction in the margin of safety.

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1 CHANGES TO REQUIRED NUMBER OF INSTRUMENTS (D)

The following assessment involves changes to the required number of instruments, and parameters that must be monitored. As such, this No Significant Hazards Considerations will address the following items:

Changes to the required and minimum operable number of instruments for the Steam Generator Hide Range Level indicators, Changes to the required and minimum operable number of instruments for the Auxiliary Feedwater Flow Rate Monitoring j

instruments, j

-- The dr : tion the PZR PORV acoustic monitors form the Technical Specifications, and Revision of the PZR PORV Glock Valve Position Indicator, j

increasing the minimum operable channels from 0 to 1.

i The proposed changes do not result in a significant increase in the probability or consequences of accidents previously evaluated.

The

)

above noted instruments provide monitoring capabilities of parameters 1

important for the oroper assessment of plant status.

These proposed 1

changes are not 1mked to increasing the probability of any accident.

These changes do not alter any of the assumed initial conditions for evaluated accidents at the Zion Station. As such, the probability for an evaluated accident has remained unchanged.

The consequence of evaluated accidents will remain the same.

In the case of the Steam Generator Hide Range Level, and the Auxiliary Feedwater Flow monitoring instruments, the required number of instruments have become more restrictive.

For the PZR PORV Block Valve Position Indicator, the minimum operable number has become more restrictive. The proposed change will relax-the minimum operable number of channels. required for the Steam Generator Hide Range Level and Auxiliary Feedwater Flow l

Monitoring instruments based on the more restricive limits imposed on the required number of channels.

In addition, the Steam Generator Narrow Range Level instruments provide redundant indication of heat sink status for both the Steam Generator Hide Range Level instruments, and the Auxiliary Feedwater Flow Rate instruments. As such, there is redundancy available for monitoring these parameters.

The Steam Generator Narrow Range Level instruments have a required number of 2 and a minimum number of 1, and are powered from redundant safety related power supplies.

In this fashion, it can be shown that there is sufficient redundancy available to justify a 7 day action statement alone for the Steam Generator Hide Range Level and Auxiliary feedwater Flow Rate instruments.

In reference to the PORV position indicator, the safety related limit switch position indicators will still be retained in the Technical Specifications.

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These indicators on their own are sufficient to provide valve status i

indication in the Control Room.

In this fashion, the consequences of

[

previously evaluated accidents will remain unchanged.

Based on the diversity available for monitoring the parameter associated with the Steam Generator Hide Range Level and the Auxiliary Feedwater Flow Rate indicator, in addition to the safety related limit switches associated with the PZR PORV Position indication, sufficient monitoring capability will be maintained. As such, the consequences of evaluated events will remain the same through the assurance that the minimum number of parameters required to perform required manual actions, and monitor and assess plant status following an accident are available.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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The proposed changes will not impose or result in plant operation that will challenge the integrity of any fission product barriers.

Post accident monitoring instrumentation is directly tied to accident f

mitigation in regards to assessment and decision making process.

The number.and type of instruments specified in the Technical Specifications, of there own will not result in the possibility of a

-3 different type of accident. All of these items as proposed, will y

still have adequate indication available through either safety related r

or diverse indication. Through the assurance that the required parameters necessary for appropriate decision making during and l

following an accident are available, it can be determined that the l'

possibility of a different type of accident or an accident conditions propagating in a differing manner will be prevented.

The proposed change does not involve a significant reduction in a margin of safety.

These changes will enhance the safety of the-plant through the establishment of more restrictive minimum acceptable levels of performance and availability for the Steam Generator Hide Range Level and Auxiliary Feedwater Flow Rate instruments in the Technical Specifications.

The current Technical Specifications do not invoke any required remedial actions until at least 3 instruments fall.

In the eveni of 3 failures, a 7 day action statement would be entered.

In the event of all 4 instruments falling, a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> action statement time clock is entered.

The proposed changes would invoke a i

7 day time clock based on a single failure, and the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> time clock would be eliminated.

The Steam Generator Narrow Range Level instruments provide redundant indication of heat sink status for both the Steam Generator Hide Range Level instruments, and the Auxiliary Feedwater Flow Rate instruments. As such, there is redundancy available for monitoring these parameters. The Steam Generator Narrow Range Level instruments have a required ntmber of 2, and a minimum number of 1.

In this fashion, it can be shown that there is sufficient redundancy available to justify a 7 day action statement alone for the Steam Generator Hide Range Level and Auxiliary feedwater Flow Rate instruments.

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Based on the diversity available for monitoring this parameter, it can be concluded that sufficient monitoring capability will be maintained. As such, hased on the redundancy in indication, and the j

lower threshold for action statement entry, it has been concluded that the overall margin of safety has been enhanced.

In regards to the PZR PORV Valve Position Indicators, valve position indication will be available from the safety related limit switches associated with these valves.

Based on the current Technical Specifications, operations j

could continue for a period of 7 days based on the inoperability of either a stem mounted limit switch, or an acoustical monitor.

If both of these indjeelors were inoperable, operations could continue for a period of tika not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

If, the required number of channels or t.finimum operable channels are not restored within these time frames, the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The proposed changes will allow operations to continued for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on the inoperability of the stem mounted limit switch alone.

If, the required number of channels (same as the minimum operable number) are not restored within this time frame the unit will be placed in Mode 4 within the next 12 j

hours.

In the event that the acoustical monitor is inoperable, the actions required will remained the same. As such, it can be concluded that the restoration time frame for an inoperable limit switch is the same in the proposed change, is the same as the current Technical Specifications assuming loss of the acoustical monitor.

Regarding the PZR PORV Block Valve Position Indicators, valve position indication will be available from the safety related limit switches associated with these valves.

Based on the current Technical Specifications, operations could continue for a period of 7 days based on the inoperability of a stem mounted limit switch.

If, the required number of channels channels is not restored within this time frame the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The proposed changes will allow operations to continued for a period of time not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> based on the inoperability of the stem mounted limit switch.

If, the required number of channels (same as the minimum operable number) are not restored within this time frame, the unit will be placed in Mode 4 within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

In this fashion, the margin of safety will remain the same, I

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MODE OF APPLICABILITY CHANGE (E)

The following No Significant Hazards Considerations addresses the addition of Mode 7 as a required mode for the accident monitoring instrumentation.

i The Mode of Applicability for Specification 3.8.9, Accident Monitoring Instrumentation has been revised to include Mode 7.

Generic Letter 83-37 was written to address the Modes of Applicability as defined in the Standardized Technical Specifications.

Mode 7 in the Zion Station Technical Specifications is defined ast less than or egaal to 5% power with reactivity and temperature stated per the specific test.

This mode is synonymous with Mode 2, incorporating specified tests from the Special Test Exemptions in the Standard Technical Specift'ations.

This change will reouire the Accident Monitoring Instruments to be operable in Modes 1, 2, 3, and 7.

The proposed change does not result in a significant increase in the probability or consequences of accidents previously evaluated. Mode 7 at the Zion Station is Low Power Physics Testing.

Low Power Physic Testing i

involves operation at power levels not to exceed 5% of rated thermal power.

In order to establish these conditions, the plant must transition Mode 3.

In transitioning Mode 3 the plant is in a condition where the accident' monitoring instruments are required to be operable.

The probability for a previously evaluated accident has remained unchanged.

The addition of this mode has no effect on an events or conditions that are precursors to any evaluated accidents.

By assuring the operability of

'1 these instruments, in all appropriate modes, the consecuences of previously evaluated accidents will remain the same.

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed change does not approve or result in the installation of new equipment.

The proposed changes will not impose or result in plant operation that will deviate from current practices.

In order to establish Mode 7, the plant must first enter Mode 3.

The Accident Monitoring Instruments are currently required operable in Mode 3.

As such, this change does not i

result in operations different from current. As such, the proposed change will not create the possibility for a new or different kind of accident from those previously evaluated.

The proposed change does not involve a significant reduction in a margin of safety.

This proposed change will result in the accident monitoring instruments specified in table 3.8.9, being required to be operable in Mode 7.

The inclusion of this mode provides assurance that these instruments would be maintained operable during low power physics testing. As such, the margin of safety will not be significantly reduced.

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L ATTACHMENT 7 ENVIRONMENTAL ASSESSMENT STATEMENT Zion Station has evaluated the proposed amendment against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR51.21.

The proposed changes do not individually or cumulatively have a significant impact on the human environment. As such, this change has been evaluated against the criteria for a categorical exclusion.

It has been determined that the proposed change meets the criteria for a categorical exclusion as provided for under 10 CFR51.21(c)(9).

This detera;.iation is based on; involving the issuance of an amendment to a license for a reactor issued pursuant to 10 CFR50, which changes both the requirements for and the surveillance requirements associated with components located within the restricted area, do not involvt any significant hazards considerations, there is no change in the amount or type of releases made offsite, and there is no increase in individual or cumulative occupational radiation exposure.

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ATTACHMENT 8 ZION STATION TYPE A VARIABLES This attachment provides a listing of the Type A variables identified in Zion Station's Regulatory Guide 1.97 submittal.

The instrument numbers and Technical Specification locations are provided for cross referencing purposes.

SPECIFICATION I

VARIABLE INSTRUMENT TABLE 3.8.9-1 ITEM

1) RCS Hot Leg Temperature TE-413A 2

(Hide Range)

TE-423A 2

f TE-433A 2

i TE-443A 2

2) RCS Cold Leg Temperature TE-413B 3

l (Hide Range)

TE-423B 3

TE-433E 3

TE-443B 3

3) RCS Pressure PT-403 4

(Hide Range)

PT-405 4

4) Steam Generator Level LT-501 8

(Hide Range)-

LT-502 8

LT-503 8

LT-504 8

5) Steam Generator Level LT-517, 518, 519 7

(Narrow Range)

LT-527, 528, 529 7

i LT-537, 538, 539 7

LT-547, 548, 549 7

6) Pressurizer Level LT-459 6

LT-460 6

LT-461 6

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7) Steam Line Pressure PT-514, 515, 516 5

PT-524, 525, 526 5

l PT-534, 535, 536 5

L PT-544, 545, 546 5

8) Containment Pressure PT-CS50 1

i PT-CSS 1 1

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9) Refueling Water Storage LT-920 9

Tank Level (Note 1)

'T.921 9

..t.

10) Containment Water Level LT-CS46 18 (Wide Range)

LT-CS47 18

11) Auxiliary feedwater Flow FT-FWO2 10 FT-FH03 10 FT-FH04 10 FT-FH25 10

12) Containment Area Radiation RE-AR02 Note 3 RE-AR03 Note 3

13) Core Exit Temperature T-0001 thru 16 T-0065

14) RCS Subccoling PT-403, 405, 456, 11 457, 458, and T-0001 thru T-0065

15) Condensate Storage Tank LT-CD94 21 Level (Note 2)

LT-CD139 21 Notc 1: As addressed in Regulatory Guide 1.97 submittal dated August 1, 1986, these transmitters do not met Environmental qualification.

These instruments will be upgraded during the Unit I and 2 Cycle 13 Refueling Outages.

Note 2: As addressed in Regulatory Guide 1.97 submittal dated August 1, 1986, neither of these level indicators met the criteria for Type A variables.

Instrumentation upgrades will be performed during the Unit 1 i

and 2 Cycle 13 Refueling Outages.

Note 3: Containment High Range Radiation Monitors addressed in Specification 3.14.1, Table 3.14-1 item 1.C.

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.3 ATTACHMENT 9 POST ACCIDENT MONITORING INSTRUMENTATION i

This attachment provides a listing of the Type B through E variables referenced in this change.

The instrument numbers and Technical Specification locations are provided for cross referencing purposes.

These instruments are cont.istent with those specified in Zion Station's Regulatory Guide 1.97 submittal.

TECHNICAL SPECIFICATION TABLE ITEM INSTRUMENT 3.8.9-1 ITEM

1) PORV Position Indicator Direct Lihiit Switch 12 i

2) PORV Block Valve Position Direct Limit Switch 13

3) Safety Valve Position TE-464 14 (Primary Detector)

TE-465 14 (Note 6)

TE-466 14 1

4) Safety Valve Position Acoustic Mo')itor LLO25 15 (Backup Detector)

Acoustic Monitor LLO26 15 Acoustic Monitor LLO27 15

5) Containment Water Level LT-CS48 17 (Narrow Range)

LT-CS49

6) Reactor Vessel Water Level LT-1311 20 (Narrow Range)

LT-1312

7) Reactor Vessel Water Level LT-1321 19 (Hide Range)

LT-1322 19

8) Vent Stack Monitor RIA-PR49 channels 7, and 9 Note 4

9) Steam Generator Atmospheric RIM-PR58 Note 5 Relief and Safety Valve RIM-PR59 Note 5 Monitors RIM-PR60 Note 5 RIM-PR61 Note 5 Note 4: Vent Stack Monitors addressed in Specification 3.12.3, Table 3.12-1 items 4.A.5, and 7.B.

Note 5: Steam Generator Atmospheric Relief and Safety Valve Monitors addressed in Specification 3.12.3, Table 3.12-1 item 6.

Note 6: These transmitters do not met the criteria for a Category 2 variable.

The date for modification of these instruments has not been determined yet.

These dates will be submitted by June 30, 1990 in Zion Station's Regulatory Guide 1.97 compliance response.

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o ATTACHMENT 10 UNCHANGED PARAMETERS This attachment provides a listing of parameters that have remained unchanged.

These items are currently in conformance with the guidance given in Generic Letter 83-37. A brief statement summarizing compliance to the design and qualification criteria established in Regulatory guide 1.97 is provided for each of these parameters.

These items have not been changed, as such no justifications have been provided.

Reactor Coolant Outlet Temperature - THOT (Hide Range) Table 3.8.9-1 item 2 has remained the same.

These instruments are Type A Category 1 variables.

The Reactor Coolant Outlet Temperature - THOT (Hide Range) instruments referred to by this table meet Category 1 design and qualification criteria.

Reactor Coolant Inlet Temperature - Tc0Ln (Hide Range) Table 3.8.9-1 item 3 has remained the same.

These Instruments are Type A Category i variables. The Reactor Coolant Outlet Temperature - TCOLD (Hide Range) instruments referred to by this table meet Category 1 design and qualification criteria.

Reactor Coolant Pressure (Hide Range) Table 3.8.9-1 item 4 has remained the same. These instruments are Type A Category i variables. The Reactor Coolant Pressure (Hide Range) instruments referred to by this table meet Category 1 design and qualification criteria.

Steam Line Pressure Table 3.8.9-1 item 5 has remained the same.

These instruments are Type A Category i variables.

The Steam Line Pressure instruments referred to by this table meet Category 1 design and qualification criteria.

Pressurizer Water level Table 3.8.9-1 item 6 has remained the same.

These instruments are Type A Category 1 variables.

The Pressurtier Water Level instruments referred to by this table meet Category 1 design and qualification criteria.

Steam Generator Water Level (Narrow Range) Table 3.8.9-1 item 7 has remained the same.

This specification deviates from the required number of channels specified in Generic Letter 83-37.

In Generic Letter 83-37 the required number of chhnnels is 1, while the current Technical Specifications require 2.

These instruments are Type A Category 1 variables.

Regulatory Guide 1.97 requires Category 1 instrumentation to be redundant.

In order to assure redundancy is maintained, 2 Steam Generator Narrow Range Water Level Instruments are required. The Steam Generator Narrow Range Water Level Instruments referred to by this table meet Category 1 design and qualification criteria.

4

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,,, o Refueling Hater Storage Tank level Table 3.8.9-1 item 9 has remained the same.

These instruments are Type A Category 1 variables.

The Refueling Water Storage Tank Level instruments referred to by this table meet Category 1 criteria, with the exception of environmental qualifications.

This exception was discussed in Zion Station's Regulatory Guide 1.97 submittal.

Modifications to upgrade these instruments are being processed at this time.

The commitment made to upgrade these instruments to Regulatory Guide 1.97 revision 2 qualifications was summarized in Attachment 3 of this submittal.

The modifications to upgrade these instruments will be completed for Unit

1. cod 2 during their cycle'13 Refueling Outages.

After completion of these modifications, the Refueling Hater Storage Tank Level Indicators will meet Category I design and qualification criteria, i

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