ML20128K165

Proposed Tech Specs Modifying Automatic Depressurization Sys Logic

ML20128K165
Person / Time
Site:	LaSalle
Issue date:	07/15/1985
From:	COMMONWEALTH EDISON CO.
To:
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ML20128K105	List: ML20128K101 ML20128K165
References
NUDOCS 8507240053
Download: ML20128K165 (36)

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,. -

4 ATTACHENT B PROPOSED CHANGE TO APPENDIX A TECHNICAL SPECIFICATIONS TO OPERATING LICENSES NPF-ll and NPF-18 Revised Pages:

NPF-11 WF-18 3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 3-29 3/4 3-29 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 3/4 5-1 3/4 5-1 B 3/4 5-1 B 3/4 5-1 B 3/4 5-2 B 3/4 5-2 J

0329K 8507240053 850715 PDR ADOCK 05000373 P

PDR.

\\

• C TABLE 3.3.3-1 EMERGENCY. CORE COOLING SYSTEM ACTUATION INSTRilMENTATION MINIMUM OPERABLE APPLICABLE c ~

CHANNELS PER TRIP OPERATIONAL FUNCTION {a)

COH0lT10NS ACTION 1.

TRIP FUNCTION i

A.

DIVISION I TRIP SYSTEM 1.

RilR-A (LPCI MODE) & LPCS SYSTEM a.

Reactor Vessel Water Level - Low Low Low, level 1 2(b) 1, 2, 3, 4*, 5*

30 b.

Drywell Pressure - liigh 2(h) 1, 2, 3 30 c.

LPCS Pump Discharge Flow-Low (Bypass) 1 1, 2, 3, 4 *, 5*

31 d.

LPCS and LPCI A Injection Valve Injection Line 1/ valve 1, 2, 3 32

-i Pressure-Low (Permissive) 4*, 5*

33 jo

'D}

N.

L, e.

LPCS and LPCI A Injection Valve Reactor Pressure-Low 2

1,2,3 38 z

, l,'

(Permissive)

~4*, 5*

33 y

Y f.

LPCI' Pump A Start Time Delay Relay 1

1, 2, 3, 4*, 5*

32 rrt g

m g.

LPCI Pump A Discharge Flow-Low (Bypass) 1 1, 2, 3, 4*, 5*

31 q

h.

Manual Initiation 1/ division 1, 2, 3, 4*, 5*

34 Atl10HATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#

2.

a.

Reactor Vessel Water Level - Low Low Low, level 1 2(b) 1, 2, 3 30 coincident with 2(b) 1, 2, 3 30 Drge{1 Pressure - liigh b.

c.

-AOS T imer 1

1, 2, 3 32 l

r; d.

Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 132

}

LPCS Pump Discharge Pressure-liigh (Permissive) 2 1,2,3 32 A

e.

g f.

LPCI Pump A Discharge Pressure-liigh (Permissive) 2 1,2,3 32 g.

Manual Initiation 1/ division 1, 2, 3 34 h.

Dq ell Pressure. Bgass Timec i

I, 2, 3 32.

E.

tie =al Lui+

1//wisia I, 2. 3 3Y

a.

h TABLE 3.3.3-1 (Continued) u,

,'?

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION G

MINIMUM OPERABLE APPLICABLE CHANNELS PER TRIP OPERATIONAL TRIP FUNCTION FUNCTION ")

CONDITIONS ACTION I

B.

HIVISION 2 TRIP SYSTEM 1.

RilR B & C (LPCI MODE) a.

Reactor Vessel Water Level - Low, Low Low, Level 1 2(b) 1 2, 3, 4*, 5*

30 l

N) b.

Drywell Pressure - High 2

1, 2, 3 30 i

c.

LPCI B and C Injection Valve Injection Line 1/ valve 1, 2, 3 32 no l

Pressure-Low (Permissive) 4*, 5*

33 d.

LPCI Pump B Start Time Delay Relay 1

1,2,3,4*,5*

32 l

l m

c.

LPCI Pump Discharge Flow - Low (Bypass) 1/ pump 1, 2, 3, 4*, 5*

31 I

~

f.

Manual Initiation 1/ division 1, 2, 3, 4 *, 5*

34 l

i w

L g.

LPCI B and C Valve Reactor Pressure-Low (Permissive) 2 1, 2, 3 38 g,

4*, 5*

33

{

I 2.

AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#

5 a.

Reactor Vessel Water Level - Low Low Low, Level 1 2(b) 1, 2, 3 30 coincident with b.

D ell Pressure - High 2(b) 1, 2, 3 30 n+ws c.

Timer 1

1,2,3 32 l

d.

Reactor Vessel Water Level - Low, level 3 (Permissive) 1 1,2,3

,, 32 Q

e.

LPCI Pump 8 and C Discharge Pressure - High (Permissive) 2/ pump 1, 2, 3 32 l

f.

Manual Initiation 1/ division 1, 2, 3 34 l

g, T>creit Nss ore. Byrass Tw."

i i, 2, 3 32.

A.

Manual Tdb*+

J

_1/ :vistou 1, 2., 3 3'l i,

i

TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 -

With the number of OPERASLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a.

With one channel inoperable, place the inoperable channel in the tripped condition within one hour

• or-declare the associated system inoperable.

b.

With more than one channel inoperable, declare the associated system inoperable.'

ACTION 31 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip Function, place the inoperable channel in the tripped condition within one hour; restort the inoperable channel to OPERABLE status within 7 days or declare

~

the associated system inoperable.

ACTION 32 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ADS trip system or ECCS inoperable.

ACTION 33 -

With the number of OPERABLE channels less than the Minimum OPERABl.E Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.

ACTION 34 -

With the number of CPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Functicn requirement, restere the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or declare the associated A05 valv+ or ECCS incperable.

1 E 3'r54*"

f ACTION 35 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per. Trip Function requirement a.

For one trip system, place that trip system in the tripped condition within one hour

• cr declare the HPCS system inoperable.

b.

For both trip systems, declare the HPCS system inoperable.

l ACTION 36 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour" or declare the HPCS system inoperable.

ACTION 37 -

With the number of OPERABLE channels less than the Total Number l

of Chr.ne's, declare the associsted cer;e : Ob.sel gsentar

/

inopaa:alganctake:,eACTI^NrequiraceySceciMcati:r3.3.1.1 i

or 4.

...c, as appr:priate.

"The provisions of Specification 3.0.4 are not applicable.

I LA SALLE - UNIT 1 3/4 3-27

- ~

A10 Ct4AN&es 408 REPGA&NCR QNo.Y TABLE 3.3.3-1 (Continued)

EMERGENCYCORECOOLING'SYSTEMACTUATIONINSTRUMENTATION ACTION ACTION 38 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per trip function requirements:

a.

With one channel inoperable, remove the inoperable channel within one hour; restore the inoperable channel to OPERABLE status within 7 days or declare the associated ECCS systems inoperable.

b.

With both channels inoperable, restore at least one channel to OPERABLE status within one hour or diciare the associated ECCS systems inoperable.

s I

\\

LA SALLE,- UNIT 1 3/4 3-27(a)

Amenc:nent 10

.w,..,

_.,w..

. -...,,.. ~,.,

-.,~,_c..,.,---_n..

E TABLE 3.3.3-2

'O LMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS r

A.10WABLE ci TRIP FilNCTION TRIP SETPOINT VAll)E y

A.

DIVISION 1 TRIP SYSTEM

.1.

RilR-A (LPCI MODE) AND LPCS SYSTEM a.

Reactor Vessel Water Level - Low Low Low, Level 1 1-129 inches

• 1-136 inches

• b.

Drywell Pressure - High 5 1.69 psig i 1.89 psig c.

LPCS Pump Discharge Flow-Low 1 750' gpm 1 640 gpm d.

LPCS and LPCI A Injection Valve Injection 500 psig 500 1 20 psig Line-Low Pressure Interlock go e.

LPCS and LPCI A Injection Valve Reactor 500 psig 500 1 20 psig Pressure-Low Pressure Interlock f.

LPCI Pump A Start Time Delay Relay 1 5 seconds 1 6 seconds g.

IPCI Pump A Discharge Flow-Low 1 1000 gpm 1 550 gpm

,,g h.

Manual Initiation NA NA 2.

AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A" a.

Reactor Vessel Water Level - Low Low Low, level 1

>- 129 inches *

>- 136 inches

• b.

Drywell Pressure - High 31.69psig 51.89psig c.

Timer

< 105 seconds

< 117 seconds l

Isl4; Shoo eactor Vessel Water Level-Low, Level 3

[12.5 inches *

[11 inches

• e.

LPCS Pump Discharge Pressure-liigh 1 146 psig, increasing 1 136 psig, increasing f.

LPCI Pump A Discharge Pressure-liigh 1 119 psig, increasing 1 106 psig, increasing g.

Manual Initiation NA NA

h. Drywett Prescore Byf*55 T*'e 6 9.0 4;, des F tu.te.un

$y E. Mr,4, val Tak;b;&

gj,g,

g,g, il";

3 (A The. Sun

-the. 4im e da,l A yf ASsCCiAla. w".4b 4ke. Ab s lastliafra 4imer and & drywell psson bypass f.=ec skall be less Aac or opal -lo 6 8*! seconds.

.,. J TABLE 3.3.3-2 (Continued) en EMERGENCY EORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS f*1 ALLOWARLE f[

IRIP FUNCTION TRIP SETPOINT, VALUE i

M B.

DIVl510N 2 TRIP SYSTEM i

1.

RilR B AND C (LPCI MODE) a.

Reactor Vessel Water Level - Low Low Low, Level 1

>- 129 inches *

>- 136 inches

• I b.

Drywell Pressure - High 51.69psig 51.89psig c.

LPCI B and C Injection Valve Injection

'I Line-Low Pressure Interlock 500 psig 500 psig 120 psig so i

d.

LPCI Pump 8 Start Time Delay Relay 5 5 seconds 5 6 seconds j

e.

LPCI Pump. Discharge flow-tow

> 1000 gpm

) 550 gpm

~.

i f.

Manual Initiation NA NA i

g.

LPCI B and C Injection Valve Reactor 500 psig 500 1 20 psig lo Pressure low Pressure Interlock u.

.~

f 2.

AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTIM "B" L

~

f a.

Reactor Vessel Water Level - Low Low Low, level 1

>- 129 inches *

>- 136 inches

• i h.

Drywell Pressure - High 1 1.69 psig 1 1.89 psig i

e c.

Timer

< 105 seconds

< 117 seconds l

(.

Reactor Vessel Water Level-Low, level 3

[12.5 inches

• I 11 inches *

)

p,.4.gg,

LPCI Pump B and C Discharge Pressure-High

> 119 psig, increasing

[106psig, increasing j

e.

f.

Manual Initiation HA NA f

g Dryweit ?resme. Byfast hi 3 9.o p;,wf44,

,p,,4,,gg h t%wat r G.4 p A.

g,4, i.'a j

J a

f a

associnksl. w;4k ~4ke. ADS in:linkoa };see opd 4he derwell pressore.,

(A) Tk son f +ke 4;me <laisys j

bygAss 4/me.r slAll be. less eau er egi=I fo 6 8 7 5ecoads.

i i

E TABLE 4.3.3.1-1 "I.

' ~.

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRilMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL i

's CHANNEL FUNCTIONAL CilANNEL CONDITIONS FOR WillCH y

TRIP It#4CTION CilECK TEST CALIBRATION SilRVEll_ LANCE REQUIRED A.

DIVISION I TRIP SYSTEM l

1.

RilR-A (LPCI MODE) AND LPCS SYSTEM i

a.

Reactor Vessel Water Level -

j Low Low Low, Level 1 5

M i

R 1,2,3,4*,5*

i b.

Drywell Pressure - High NA M

Q 1,2,3 c.

LPCS Pump Discharge Flow-Low NA M

Q 1, 2, 3, 4", 5*

l i

d.

LPCS and LPCI A Injection Valve

. Injection Line Pressure Low Interlock NA M

R 1, 2, 3, 4*, 5*

g, I

e.

LCPS and LPCI A Injection Valve

' Reactor Pressure Low Interlock NA M

R I, 2, 3, 4*, 5*

o, 2

f.

LPCI Pump.A Start Time Delay l

Relay NA M

Q 1, 2, 3, 4*, 5*

l J,

g.

LPCI Pump A Flow-Low NA M

Q 1, 2, 3, 4*, 5*

m N

h.

Manual Initiation NA R

NA 1, 2, 3, 4*, 5*

l AtlT0HATIC DEPRESSilRIZATION SYSTEM TRIP SYSTEM "A"#

i 2.

a.

Reactor Vessel Water Level -

Low Low Low, Level 1 5

M R

1, 2, 3 b.

Drywell Pressure-liigh NA M

Q 1, 2, 3

- y y,.f;4} fog c.

A95' Timer HA M

Q I, 2, 3 l

i

. Reactor Vessel Water Level -

y Low, Level 3 S

M R

1,2,3 g

e.

LPCS Pump Dis ~ charge

']

Pressure-High NA M

Q 1,2,3 7

g f.

LPCI Pump A Discharge Pressure-liigh NA M

Q 1, 2, 3 I

g.

Manual Initiation NA R

NA 1, 2, 3 Dryv ett Pmsm Sgast T;mec.

NA et a

I, 2, 3 h.

e l

Me**'A I I"V M

_ g g

g 3

l

t TABLE 4.3.3.1-1 (Continued) g EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 2

k CilANNEL OPERATIONAL CllANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WillCil TRIP FilNCTION CilECK TEST CALIBRA' TION SURVEILLANCE REQUIRED U

B.

DIVISION 2 TRIP SYSTEM

~

1.

RilR B AND C (LPCI MODE) a.

Reactor Vessel Water _ Level -

R 1, 2, 3, 4*, 5" Low Low Low, Level 1 S

M b.

Drywell Pressure - High NA M

Q 1, 2, 3 c.

LPCI B and C Injection Valve Injection Line Pressure so Low Interlock NA M

R 1, 2, 3, 4*, 5*

d.

LPCI Pump B Start Time Delay

.g Relay NA M

Q 1, 2, 3, 4 *, 5*

l e.

LPCI Pump' Discharge Flow-Low NA M

Q 1, 2, 3, 4 *, 5"

, s, f.

Manual Initiation NA R

NA 1,

2., 3, 4*, 5*

<a g.

LPCI B and C Injection Valve i

Reactor Pressure Low Interlock NA M

R 1, 2, 3, 4*, 5*

l' j

2.

AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#

a.

Reactor Vessel Water Level -

Low Low Low, Level 1 5

M R

1, 2, 3 l

ddiAEM Drywell Pressure-liigh NA H

Q 1, 2, 3 i.

c.

ADS Timer NA H

Q 1, 2, 3 l

2 d.

Reactor Vessel Water level -

E Low, Level 3 5

H R

1, 2, 3 S

e.

LPCI Pump B and C Discharge S

Pressure-liigh NA H

Q 1, 2, 3 f.

Manual Initiation NA R

NA 1, 2, 3

g. Dryweli Pmeare. Bqass T;e yA g

dl I' z' 3 A, Moon ( roL1:4 up E

. AIA 1, 2, 3 P

e

3/4.S EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION 3.5.1 ECCS* divisions 1, 2 and 3 shall be OPERABLE with:-

a.

ECCS division 1 consisting of:

1.

The OPERABLE low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

2.

The OPERABLE low pressure coolant injection (LPCI) subsystem "A".

of the RHR system with a flow path capable of-taking suction' from the suppression chamber and transferring the-water to the reactor vessel.

su 3.

At least 6 OPERABLE ADS valves.

l b.

ECCS division 2 consisting of:

1.

The OPERABLE low pressure coolant injection (LPCI) subsystems "B" and "C" of the RHR system, each with a flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

w 2.

At least 6 OPERABLE ADS valves.

[

c.

ECCS division 3 consisting of the OPERABLE high pressure core spray (HPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

APPLICABILITY: OPERATIONAL CONDITION 1, 2*# and 3*.

a The ADS is not required to be OPERABLE when reactor steam dome pressure i

is less than or equal to 122 psig.

i 9

i See Special Test Exception 3.10.6.

l 4

l f

\\"see sgu;6w 313 & up swh opmLI: 3 i

LA SALLE - UNIT 1 3/4 5-1

3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN ECCS Division 1 consists of the low pressure core spray system, low pressure coolant injection subsystem "A" of the RHR system, and the automatic depressurization system (ADS) as actuated by ADS trip system "A".

ECCS Division 2 consists of low pressure coolant injection subsystems "B"and "C" of the RHR system and the automatic depressurization system as actuated by ADS trip system "B".

The low pressure core spray (LPCS) system is provided to assure that the core is adequately' cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following l

j depressurization by the ADS.

9,4 g The LPCS is.a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequate assurance that the LPCS system will be OPERA 8LE when required. Although all active components are testable and full flow can'be demonstrated by recirculation through a test

-loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

The low pressure coolant injection (LPCI) mode of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Three subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for.small breaks following depressurization by the A05.

[

cwom w The surveillance requirements provide adequate assurance that the LPCI system will be OPERA 8LE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

ECCS Division 3 consists of the high pressure core spray system. The i

high pressure core spray (HPCS) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel. The HPCS system permits the reactor to be shut down while maintaining sufficient reactor l

vessel water level inventory until the vessel is depressurized. The HPCS system operates over a range of 1160 psid, differential pressure between l

reactor vessel and HPCS suction source, to 0 psid.

The capacity of the HPCS system is selected to provide the required core cooling. The HPCS pump is designed to deliver greater than or equal to 516/1550/6200 gpa at differential pressures of 1160/1130/200 psid.

Initially, water from the condensate storage tank is used instead of injecting water from lit i

LA SALLE - UNIT 1 B 3/4 5-1 Amendment No.18 i

EMERGENCY CORE C00LXNG 5YSTEMS BASES

~

ECCS-OPERATING and SHUTDOWN (Continued) the suppress' ion pool into the reactor, but no credit is taken in the hazards analyses for the condensate storage tank water.

With the HPCS system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems.

In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the hazards analysis, will automatically provi'de makeup at reactor operating pressures on a reactor low water level condition.

The HPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.

The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required.

Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown.

The pump discharge piping is maintained full to prevent water hammer. damage and to provide cooling at the earliest moment.

' i4 g roi n J.

Upon failure of the HPCS system to function properly n er ; :,::11 breat les:-cf-ccc'Ont accid:nt, the automatic depressurization hstem (ADS) automati-cally causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel claading temperature to less than 2200 F.

ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 122 psig even tnough low pressure core cooling systems provide adequate core cooling up to m

350 psig.

ADS automatically controls seven selected safety-relief valves.

Six valves are required to be OPERABLE although the hazards analysis only takes credit for five valves.

It is therefore appropriate to permit one of the required valves to be out-of-service for up to 14 days without materially reducing system reliability.

3/4.5.3 SUPPRESSION CHAMBER The suppression chamber is also required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA.

This limit on suppression enamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core.

The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.2.1.

Repair work might require making the suppression chamber inoperable.

This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made. inoperable, including draining, in OPERATIONAL CONDITION 4 or 5.

In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required water volume is reduced because the reactor coolant is maintained at or below 200 F.

Since pressure suppression.is not required below 212 F, the minimum water vulume is based on NPSH, recirculation volume, vortex prevention plus a T -4" safety margin for conservatism.

LkSALLE-UNIT 1 B 3/4 5-2

9 TA8LE 3.3.3-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION f

MINIMUM OPERABLE APPLICA8LE I

m i

CHANNELS PER TRIP OPERATIONAL E

FUNCTION *I CONDITIONS ACTION E

7 TRIP FUNCTION l

m A.

DIVISION I TRIP SYSTEM

.t, 1.

RHR-A (LPCI MODE) & LPCS SYSTEM I

a.

Reactor Vessel Water Level - Low Low Low, level 1 2

1, 2, 3, 4*, 5" 30 I

ID) r 2(b)-

1,2,3 30 i

l; b.

Drywell Pressure - High c.

LPCS Pump Discharge Flow-Low (Bypass) 1 1, 2, 3, 4*, 58 31 i

I d.

LPCS and LPCI A Injection Valve Injection Line 1/ Valve 1, 2, 3 32 2

Pressure-Low (Permissive) 4 *, 5 *-

33 I

w 2

1,2,3 38 l

. 's" e.

LPCS and LPCI A Injectitin Valve Reactor 48, 58 33 Pressurd-Low (Permissive) f.

LPCI Pump A Start Time Delay Relay 1

1, 2,' 3, 4*, 5*

32 i

l i

g.

LPCI Pump A Discharge Flow-Low (Bypass) 1 1, 2, 3, 4*, 5*

31 I

h.

Manual Initiation 1/ division 1, 2, 3, 4", 5*

,34 i

2.

AUTOMATIC-DEPRESSURIZATION SYSTEM TRIP SYSTEM "A" 2(b) 1, 2, 3 30 Reactor Vessel Water Level - Low Low Low, Level 1 a.

j coincident with 2(b)'

1, 2, 3 30 b.

Dr 11 Pressure - H10h c.

-AB& Timer 1

1,2,3 32 l

l

~hn a b.o n s

i d.

Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 12 LPCS Pump Discharge Pressure-liigh (Permissive) 2 1,2,3 32 e.

f.

LPCI Pump A Discharge Pressure-High (Permissive) 2 1,2,3 32 9

Manual Initiation 1/ division 1, 2, 3 34 h.

Dryo, cit Prenoce. Bypos:. Thec

[

1, 2, 3 72 i.

r% val 14:htt 3 / J:o:s;ou i, 2, 3 34 5

I E

TABLE 3.3.3-1,(Continued)

W EMERGENCY CORE C00L1HG SYSTEM ACTUATION INSTRUMENTATION y

en HINIMUM OPERABLE APPLICABLE CllANNELS PER TRIP OPERATIONAL i

c5 FUNCTION "I CONDITIONS ACTION I

b 1 RIP FUNCTION ro B.

DIVISION 2 TRIP SYSTEH i

1.

RilR 8 & C (LPCI H0DE) 2(b) 1, 2, 3, 4*, Sa 30 a.

Reactor Vessel Water Level - Low, Low Low, level 1 2(b) 1, 2, 3 30 b.

Drywell Pressure - liigh 1,2,3 32

?

LPCI B and C Injection Valve Injection Line Pressure-Low 1/ valve 4*, 5*

33 I

c.

(Permissive) t d.

LPCI Pump B Start T 6e Delay Relay 1

1, 2, 3, 4*, 5*

32 g

e.

LPCI Pump Discharge Flow - Low (Bypass) 1/ pump 1,2,3,4",5*

31

[

52 f.

Manual Initiation 1/ division 1, 2, 3, 4*, 5*

34 9

LPCI B and C Injection Valve Reactor 2

1, 2, 3, 3B Pressure-Low (Permissive) 4 *, 5*

33 AU10MATIC DEPRESSURIZATION SYSTEH TRIP SYS1EH "B"#

2.

}

a.

Reactor Vessel Water Level - Low Low Low, Level 1 2

1,2,3 30 coincident with SI b.

D ell Pressure - liigh 2

1,2,3 30 k

'iner 1

1,2,3 32 I

c.

d.

Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 32 LPCI Pump B and C Discharge Pressure - liigh e.

(Permissive) 2/ pump 1, 2, 3 32 f.

Manual Initiation 1/ division 1, 2, 3 34 g.

Depe.it Pressore. Bygass Tuer 1

f, 2, 3 32.

h.

M Anual I,d,;t,:t

.1/J:ursion I, 2, 3 34

TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a.

With one channel inoperable, place the inoperable channel in the tripped condition within one hour

• or declare the associated system inoperable.

b.

With more than one channel inoperable, declare the aso ciated system inoperable.

ACTION 31 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip Function, place the inoperable, channel in the tripped condition within one hour; restore the inoperable channel to OPERABLE status within 7 days or declare the associated system inoperable.

ACTION 32 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ADS trip system or ECCS inoperable.

ACTION 33 -

With the number of OPERABLE channels less than the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.

ACTION 34 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or declare the associated ADS valve or ECCS inoperable.

I

-Me sydem ACTION 35 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement a.

For one trip system, place that trip system in the tripped condition within one hour

• or declare the HPCS system inoperable.

b.

For both trip systems, declare the HPCS system inoperable.

ACTION 36 -

With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour

• or declare the HPCS system inoperable.

ACTION 37 -

With the number of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2, as appropriate.

"Tne provisions of Specification 3.0.4 are not applicable.

LA SALLE - UNIT 2 3/4 3-27

A)0 C H A nt (riiS CM TH\\ S PA GG Ft)2 CEPGCEn1CB O!JLY TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 38 With the number of OPERABLE channels less than required by the Minirvm OPERABLE Channels per trip function requirements:

a.

With one channel inoperable, remove the inoperable channel within one hour; restore the inoperable channel to OPERABLE status within 7 days or declare the associated ECCS systems inoperable.

b.

With both channels inoperable, restore at least one channel to OPERABLE status within one ho0r or declare the associated ECCS system inoperable.

9 9

LA SALLE - UNIT 2 3/4 3-27(a)

TABLE 3.3.3-2 9

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPolNTS m

r"-

ALLOWA8LE

, TRIP SETPolNT VALUE TRIP flNICTION k

A.

DIVISION 1 TRIP SYSTEM a

I to 1.

RilR-A (LPCI M00E) AND LPCS SYSTEM a.

Reactor Vessel Water Level - Low Low Low, Level 1 1-129 inches

• 1-136 inches

• b.

Drywell Pressure - liigh i 1.69.psig 5 1.89 psig i

c.

LPCS Pump Discharge Flow-Low 1 750 gpa 1 640 gge d.

LPCS and LPCI A Injection Valve Injection Line-500 psig 500 + 20 psig Low Pressure Interlock e.

LPCS and LPCI A Injection Valve Reactor 500 psig 500 + 20 psig

- Pressure-Low Pressure Interlock 1 5 seconds 1 6 seconds i

f.

LPCI Pump A Start Time Delay Relay g

g.

LPCI Pump A Discharge Flow-Low 1 1000 gpa 1 550 gpa m

1 h.

Manual Initiation N.A.

N.A.

m 2.

AHIONATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A" s

a.

Reactor Vessel Water Level - Low Low Low, Level 1 1-129 inches

• 1-136 inches

• f

h. _ Drywell Pressure - liigh

$ 1.69 psig i 1.89 psig

< 105 seconds

< 117 seconds l

1

_M

c. '-A05 Iiner

[12.5 inches *'

[11 inches

• I d % AIi j d.

Reactor Vessel Water Level-Low, Level 3 LPCS Pump Discharge Pressure-High 1 146 psig, increasing 1 136 psig, increasing e.

f.

LPCI Pump A Discharge Pressure-liigh 1 119 psig, increasing 1 106 psig, increasing j

g.

Manual Initiation N.A.

N.A.

]

h. Dqwelt Preuore Bvgass rtuer S q,o y;. ohs g 4,g g,(,3 L

'- Ha" val ruk; bit u.A.

n.4.

l

~

i Y

fWWM M

MM N A is oce MM W,'k b NI M>'bh 4

UM Me e AyS l

hvfna 4 mec shis 6a less k.a or ayaal k se, seconds.

i

1ABLE 3.3.3-2 (Continued')

I E

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETP0lNIS y,

?

ALLOWABLE j

i 1 RIP IUNCTION TRIP SETPOINT VALUE l

c-5 8.

DIVISION 2 TRIP SYSTEM l.

RllR B AND C (LPCI MODE)

Reactor Vessel Water Level - Low Low Low, level 1

>- 129 inches *

>- 136 inches

• a.

b.

Drywell Pressure - liigh 5 1.69 psig 51.89psig LPCI 8 and C Injection Valve Injection Line 500psip 500 1 20 psig c.

Low Pressure Interlock d.

LPCI Pump B Start Time Delay Relay 5 5 seconds 5 6 seconds e.

LPCI Pump DischarDe Flow-Low

> 1000 gpa

> 550 gpa f.

Manual Initiation N.A.

N.A.

g.

LPCI 8 and C Injection Valve Reactor 500 psig 500 1 20 psig w

)

Pressure-Low Pressure Interlock y

E 2.

Alli0MATIC DEPRESSURIZATION SYSTEM IRIP SYSTEM "B" a.

Reactor Vessel Water Level - Low Low Low, Level 1

>- 129 inches *

>- 136 inctes*

h.

Drywell Pressure - liigh 51.69psig 51.89psig yy;4;g4 o,f c.

Timer

< 105 seconds

< 117 seconds I

d.

Reactor Vessel Water Level-Low, Level 3

> 12.5 inches

• i 11 inches

• LPCI Pump B and C Discharge Pressure-liigh

[119psig, increasing

[106psig, increasing e.

f.

Manual Initiation N.A.

N.A.

g Dcyweit Pressore Bqass Tg" 9.0 Advofes bekud, Cr4 h

Manual nk. bit M.A.

el. a.

my rk saa al A. 4tue delags assoc. aid a.4L.4Se Abs is:kk a kee osd k desweII peswre.

byps: 46ecsi,it h4 (ess h oc egual 4.

6 8 7 secouJs.

-I G'

TABLE 4.3.3.1-1 f

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS l

v I

E CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 5,

TRIP IUNCTION CilECK TEST CALIBRATION SURVEILLAMCE REQUIRED l

e c

A.

DIVISION I TRIP SYSTEM 1.

RHR-A (LPCI MODE) AND LPCS SYSTEM a.

Reactor Vessel Water Level -

Low Low Low, Level 1 5

M R

1, 2, 3, 4*, 5" b.

Drywell Pressure - liigh NA M

Q 1, 2, 3 c.

LPCS Pump Discharge Flow-Low NA M

Q 1, 2, 3, 4*, 5*

j d.

LPCS and LPCI A Injection Valve injection Line Pressure Low Interlock NA M

R 1, 2, 3, 4", 5*

e.

LPCS and LCPI A Injection Valve Reactor Pressure Low Interlock NA M

R 1, 2, 3, 4*, 5*

w 4

f.

LPCI Pump A Start Time Delay 2,

Relay NA M

Q 1, 2, 3, 4 *, 5*

l w

g.

LPCI Pump A Flow-Low HA M

Q 1, 2, 3, 4*, 5*

^2 h.

Manual Initiation NA R

NA 1, 2, 3, 4*, 5*

i AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#

2.

a.

Reactor Vessel Water Level -

i Low Low Low, Level 1 5

M R

1, 2, 3 b.

Drywell Pressure-liigh NA M

Q 1, 2, 3 j

y,,; t;p,4;ea

c. "-ABS Timer NA M

Q 1, 2, 3 I

l d.

Reactor Vessel Water Level -

Low, Level 3 S

M R

1, 2, 3 e.

LPCS Pump Discharge i

Pressure-liigh NA M

Q 1, 2, 3 f.

LPCI Pump A Discharge t

Pressure-Illoh NA M

Q 1, 2, 3 g.

' Manual Initiation NA R

NA 1, 2, 3

h. Drywen Pressoce Ovf a55 Ibec MA f4 Ct Il 3 s a l

Mwv4I Ich:1,; t NA R

ng I,1, 3 j

i

I C

TABLE 4.3.3.1-1 (Continued)

W EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS r-CHANNEL OPERATIONAL m

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WilICH f

.5 TRIP FUNCTION CllECK TEST CALIBRATION SURVEILLANCE REQUIRED e

I B.

DIVISION 2 TRIP SYSTEM 1.

RilR B AND C (LPCI MDDE) i a.

Reactor Vessel Water Level -

l Low Low Low, Level 1 S

M t

-R 1, 2, 3, 4*, 5*

l' b.

Drywell Pressure - High NA M

Q 1, 2, 3 f

c.

LPCI 8 and C Injection Valve Injection Line Pressure Low Interlock NA M

R

,1, 2, 3, 4", 5" i

d.

LPCI Pump 8 Start Time Delay Relay NA M

Q 1, 2, 3, 4*, 5*

l m

2, e.

LPCI Pump Discharge Flow-Low NA M

Q 1, 2, 3, 4*, 5" i

f.

Manual Initiation NA R

NA 1, 2, 3, 4*, 5*

g.

LPCI B and C Injection Valve Reactor Pressure Low Interlock NA M

R 1,2,3,4*,5*

AUTGHATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#

2.

a.

Reactor Vessel Water Level -

Low Low Low, Level 1 S

M R

1, 2, 3 b.

Drywell Pressure-liigh NA M

Q 1, 2, 3 T.,;{;ghera c N A05 Timer HA M

Q 1, 2, 3 I

~~

d.

Reactor Vessel Water Level -

Low, Level 3 S

M R

1, 2, 3 e.

LPCI Pump 8 and C Discharge Pressure-liigh NA M

Q 1,2,3 f.

Manual Initiation NA R

NA 1, 2, 3

q. Ocywett Pressucc. 6 pass he gg g

U Q

)'2.,3 I

h. Mauvat Tak: bit MA R

. NA 1, 2., 3

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR OPERATION l

3.5.1 ECCS divisions 1, 2 and 3 shall be OPERABLE with:

a.

ECCS division 1 consisting of:

1.

The OPERABLE low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

i 2.

The OPERABLE low pressure coolant injection (LF".I) subsystem "A" of the RHR system with a flow path capable U' taking suction from the suppression chamber and transferring the water to the reactor vessel.

3.

At least 6 OPERABLE ADS valves.

l b.

ECCS division 2 consisting of:

1.

.The OPERABLE low pressure coolant injection (LPCI) subsystems "B" and "C" of the RHR system, each with a flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.

se 2.

At least 6 OPERABLE ADS valves.

l c.'

ECCS division 3 consisting of the OPERABLE high pressure core spray (HPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.

APPLICABILITY: OPERATIONAL CONDITION 1, 2*# and 3*.

" The AD5 is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 122 psig.

I See Special Test Exception 3.10.6.

I

'"See Spe:ficah> 3.3.3 4 +c r SY54e" eFmb;l:4g LA SALLE - UNIT 2 3/4 5-1 l

3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN ECCS Division 1 consists of the low pressure core spray system, low pressure coolant injection subsystem "A" of the RHR system, and tne automatic depressurization system (ADS) as actuated by A05 trip system "A".

ECCS Division 2 consists of low pressure coolant injection subsystems "B"and "C" of the RHR system and the automatic depressurization system as actuated by ADS trip system "B".

The low pressure core spray (LPCS) system is provided to assure that the core is adequately cooled,following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and~ including the

. double-ended reactor recirculation 1ine break, and for, smaller breaks following' l

dep'ressurization'by the ADS.

Qgs,g The LPCS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.

The surveillance requirements provide adequate assurance that the LPCS system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated 'by recirculatirsn through a test loop during reactor operation, a complete functional test requires reactor-shutdown.

The pump discharge piping is maintained full to prevent water hamer damage to piping 'and to start'ccoling at the earliest moment.

The low pressure coolant injection (LDCI) modo of the RHR system is provided to assure that the core is adequately cooled following a loss-of-coolant accident. Three subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for#small breaks following depressurization by the ADS.

1

... es o n The surveil an e requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.- The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.

ECCS Division 3 consists of the high pressure core spray system. The high pressure core spray (HPCS) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result'in rapid depressurization of the reactor vessel. The HPCS system perinits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized. The HPCS system cperates over a range of 1160 psid, differential pressure between reactor vessel and HPCS suction source, to O psid.

The capacity of the HPCS system is selected to provide the required core cooling. The HPCS pump is designed to deliver greater than or equal to 516/1550/6200 gpm at differential pressures of 1160/1130/200 psid.

Initially, water from the condensate storage tank is used instead of injecting water from LA SALLE - UNIT 2 8 3/4 5-1

EMERGENCY CORE COOLING SYSTEMS BASES ECCS-OPERATING and SHUTDOWN (Continued) the suppression pool into the reactor, but no credit is taken in the hazards analyses for the condensate storage tank water.

• With the HPCS system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems.

In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the hazards analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.

The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

Upon failure of the HPCS system to function properly 3:is 5. $.9eus&,e;;11 br::k

7 ;

-1;;;-Of :;;l;r.t :::id:-t, the automatic depressurization system (ADS) automati-cally causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can. enter the core in time to limit fuel cladding temperature to less than 2200*F.

ADS is conservatively required to be' OPERABLE whenever reactor vessel pressure exceeds 122 psig even though low pressure core cooling systems provide adequate core cooling up to 350 psig.

ADS automatically controls seven selected safety-relief valves. Six valves are required to be OPERABLE although the hazards analysis only takes credit for five valves.

It is therefore appropriate to permit one of the required valves to be out-of-service for up to 14 days without materially reducing system reliability.

3/4.5.3 SUPPRESSION CHAMBER The suppression chamber is also required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA.

This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit recirculation cooling flow to the core. The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.2.1.

Repair work might require making the suppression chamber inoperable.

This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5.

In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required water volume is reduced because the reactor coolant is maintained at or below 200*F.

Since pressure suppression is not required below 212*F, the minimum water volume is based on NPSH, recirculation volume, vortex prevention plus a 2'-4" safety margin for conservatism.

LA SALLE - UNIT 2 8 3/4 5-2

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated the proposed Technical Specification Amendment and determined that it does not represent a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Units 1 and 2 in accordance with the proposed amendment will not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated because the revised ADS logic doesn't affect automatic depressurization for events where high drywell pressure occurs. This modification automates the function of reactor vessel blowdown for events where high drywell pressure does not occur. Under these conditions, manual operation of the ADS system is called for by the emergency operating procedures and was assumed in Chapter 15 of the UFSAR.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated because automatic depressurization is analyzed and required for events where high pressure coolant sources are unavailable and reactor vessel level is low. This change only automates what were previously manual operator actions.

3) Involve a significant reduction in the margin of safety because the upgraded logic provides additional margin of safety for events where high drywell pressure does not occur while still providing the same level of protection for events where high drywell pressure does occur.

Based on the preceding discussion, it is concluded that the proposed system change clearly falls within all acceptable criteria with respect to the system or component, the consequences of previously evaluated accidents will not be increased and the margin of safety will not be decreased.

Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(e), the proposed change does not constitute a significant hazards consideration.

0329K L'

i.

ATTACHENT D Schematic diagrams 4201 AB, AH, AM are included.

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6 st44 (ac0 mqm GEN ER AL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY 175 CURTNER AVE., SAN JOSE. CALIFORNIA 9s12s MC 391, (408) 925-3798 A TT A CH Me+>T F

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September 29, 1983 EkUCof) h Tab ~~ ~

LS-2894 6 am~j Mr.T.EIWatts faa

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2c Sy/ e Commonwealth Edison' Company One First National Plaza Chicago, IL 60690

SUBJECT:

LASALLE' COUNT STATION UNITS 1 & 2 ADS LOGIC MODIFICATION - TECHNICAL SPECIFICATION MODIFICATIONS AND TIMER SETTINGS s

This letter provides the basis for selecting the timer settings for the ADS Logic Drywell Pressure Bypass Timer (DPBT) for the LaSalle units and provides technical specification revisions to reflect the modification.

Maximum and minimum setpoints for the DPBT are determined based on minimizing the operator actions related to the ADS logi.c, minimizing the chance for unintentional ADS and limiting the impact on design basis ECCS analysis.

The nominal trip setpoints and allowable values are derived from maximum or minimum value after taking setpoint drift, accuracy and calibration prevision into account.

DPBT Analytic limit The DPBT Analytic Limit is based on design bases FSAR calculation of the main steam line break outside of containment.

For this event, wit.h HPCS assumed to be unavailable, manual ADS is currently modeled at a maximum of 10 minutes after the event begins.

To account for tne revised ADS logic, a revised calculation has been made with ADS occurring about 10 minutes after RPV level reaches Level 1.

The design goal is to limit the PCT to about 1500 F.

This limit is somewhat arbitrary, but has been set so that this event does not become the limiting design bases accident (DBA).

The results of the calculation (Figures 1 through 4) show that the final Peak Cladding Temperature (PCT) with the DPBT set at 10 minutes plus the existing 2 minute ADS initiation delay is 1493 F.

The analytical limit for the DPBT is set at 10 minutes.

GENERAL h ELECTRIC Page 2 September 29, 1983 Trip Setpoint and Allowaole Value The trip setpoint for the DPBT is determined after consideration of the analytical limit, instrument setpoint accuracy', calibration and drift.

The allowable value determined by the methodology described in GE document 22A5261 yields a trip setpoint of 9.0 minutes and an allowable value of 9.5 minutes.

(5% setpoint accuracy,.1 minute calibration error and.5 minute drift is assumed based on manufacturer information.)

With these values the LER avoidance probability is greater than 90%.

Minimum Setpoint The minimum setpoint is such that the operator is not unduly burdened during an ATWS event by having to repeatedly reset the A05 logic initiation delay.

The time delay should be long enough so that the operator has time to stabilize RPV water level (per EPGs Contingency 7) near the top of active fuel before manually inhioiting ADS with the new inhibit-switch which is part of the-logic modification.

With this approach, the operator is keyed by the startup of the ADS initiation delay to either inhibit A05 or prepare for depressurization, depending on whether RPV water level is stabilized or decreasing.

To determine the minimum time delay value, several assum made with regard to operator actions and responsiveness:ptions have been 1.

Following an ATWS event from full power with MSIV closure, water level will drop to near the RPV level 1 trip setpoint (L1) and then start to increase due to automatic initiation of HPCS at L2 and injection of baron through the Standby Liquid Control System (SLCS).

No operator action is expected until suppression pool water temperature is observed to be greater than 110 F.

If the RPV water level has dropped below L1, which starts the OPBT, RPV level is expected to increase above L1 before the operator manually controls RPV level.

This allows the OPBT to reset.

LaSalle Emergency Procedures are assumed to specify this action even though restoration above L1 is not specifically' called for in the EPGs.

Engineering judgment has been used to determine that for a limiting ATWS event, the RPV level will recover about L1 (if it goes below L1) before the OPBT runs out without operator action.

This judgment, which is based on generic BWR/5 evaluations of ATWS events (NEDE 24222), assumes that:

The proposed ATWS rule is implemented (draft 10CFR50.62, a.

August 3, 1983).

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GENER AL h ELECTRIC Page 3 September 29, 1983 b.

SLCS injection location is repiped from the current lower plenum standpipe to the HPCS(msel s,h d rk Nks,y h v@ l.

The operator follows the BWR EPGs and initiates SLCS when the c.

suppression pool temperature reaches 110 F.

With these assumptions, it has been judged that a 5 minute DPBT timer setpoint would allow avoidance of ADS during the initial stages of an ATWS event.

2.

Af ter taking manual RPV level control, the EPGs instruct the operator to lower indicated RPV water level to near the top of active fuel.

When water level decreases below L1, the DPBT will start.

Since the RPV wide range level indication is at the bottom of its indicating range, the operator will need to stabilize the RPV level based 'on the fuel zone indicator (which will indicate well below the top of active fuel due to lack of pressure compensation relative to the fuel zone calibration condition).

L1 to stabilization of the water level is somewhat uncertain.The time requi It is judged that 5 minutes is adequate for stabilization of the RPV level under these conditions.

The above two considerations place a minimum setpoint at about 5 minutes for LaSalle.

Since this is less than the trip setpoint determined above, ATWS events without becoming unduly burdened.use of the trip setpoint a Technical Specification Modifications The LaSalle technical specification and bases were reviewed to determine p any modifications required due to implementation of this modification to Np<cy#O the ADS logic.

Tables 3.3.3-1, 3.3.3-2 and 4.3.3.1-1 require modification.

Attachment A includes a markup of these two tables showing the required Y

changes.

d made.A slight clarification to the bases 3/4.5.1 and 3/4.5.2 should also be T

O N

t.9 A markup showing these changes is also included on Attachment A.

l

GENERAL h ELECTRIC Page 4 September 29, 1983 Finally, it should be noted that although no chance to the technical specification is needed, the functional test procedure to conform with Section 4.5.1.d should be modified to include the DPBT function.

Very truly your,

H. R. Peffe, Project Manager LaSalle County Station Units 1 & 2 HRP: cal /K09294 Attachments cc:

J. E. Ellis D. C. Haan R. Janecek P. B. Kavanagh F. Leone B. R. Shelton R. E. Spencer DBL /CTK e

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