ML20117D310
| ML20117D310 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/09/1996 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20101R197 | List: |
| References | |
| NUDOCS 9604160107 | |
| Download: ML20117D310 (30) | |
Text
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Technical SpecNication Pages D
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2-4a Amendment flo.
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LIMITIls SAFETY SYSTEM SETTINGS t
SASES
\\_.
emTOR EwitCTION $YSTEM INSTRWWITATION SETp0!NTS (Continued) 4.
Reacter Vessel Water Level-Lok The reacter vessel water level trip setpoint was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.
5.
Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events. The MSIV's are closed automatically /from measured parameters such as high steae i
flow, thigh steam line radiation,Ilow reacter water level, high steam tunnel toeperature and low steam 11ne pressure. The MSIV's closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal /tqydraulic Safety Limits.
t 6.
lainSteamLineRadiation-Nish The main steam'line radiation detectors are provided to detect a gross i
failure of the fuel cladding. When the high radiation is detected a trip is initiated to reduce.the continued failure of fuel cladding.
At the same time
~,
the main steam line 1 solation valves are closed to limit the release of fission
(* products.
The trip setting is high enough above background radiation levels f
to prevent spuriegs;. trips yet low enough to promptly detect gress failures in the fuel cladding. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall tw11 ability of the Reactor Protection System.
7.
Primary Containment Pressure-High High pressure in the drywell could indicate a break in the primary pres-sure boundary systems..The reactor.is tripped in order to minimize the possi-bility of fuel damage and reduce the amount of energy being added to the coolant. The trip. setting was selected as low as possible without causing spur.ious. trips.
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LA SALLE - UNIT 1 B 2-11
~<x A
f gj NJ TABLE 3.3.1-1 5
REACTOR PROTECTION SYSTEM INSTENENTATION y,
r-l;;
- APPLICABLE NININ#1 OPERABLE
- OPERATIONAL,
CNAIRIELS PER FUNCTIONAL UNIT CONDITIONS TRIP SYSTEM (a)
ACTION g
U 1.
Intermediate Range Moniters:
a.
Neutren Flux - Hi#.
r; 2 3
1 3, 4 2
2 5(b) 3 3
,i,i
'2 b.
Inoperative 3
1
!: 3, 4 2
2
.'t.
5 3
3
<r 2.
Average Power Range McIniter:I*)
![ 2 '
, eutron Flux - Hi p, Setdeun
'd 2
1 N
a.
'3 k
s 2
2 w
- 5(b) 2 3
A b.
FlowBlased$1mudtedThe'6mel F
Power-lhcale li 1-2 4
c.
Fixed Neutron Fluit-High 1
2 4
~
d.
Inoperative 1, 2 2
1 3
2 t
5 2
3 m
3.
Reacter vessel Steen Dome Pressure - Nigh 1,2(g) 2 1
seector vessel unter Level - Lew.
4.
Level 3 1, 2 2
1 5.
Main Stese Line Isolatten Valve -
I y
I 'I 4
4
,f Closure 6.
Main Steam Line Radiation -
Id)
/
High 1, 2 2
5 beleted
(
(
4 i
j c.
~d REACTOR PA0TECTION SYSTEM IIISTRUMENTATICII
- us j
AElm
.Se in at least HDT SIRITDelet within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTICII 1 i
Verify all insertable control rods to be inserted in'the core, ACTI0li 2 and lack the reactor modo switch in the Shutesun position within j
one hour.
Suspend all operettens involving CORE ALTERATIONS 8 insert all ACTICII 3 j
insertable control rods within one hour.
gg l
ACTION 4 Se in at least STARTUP withis 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
1 i
ACTI0ff 5
- I Se in STARTUP with the asia stems line isolation valves closed I within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least HDT SitfTD0lAl within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Initiate a reduction la TIEM4L POWER within 15 minutes an(
ACTION 5 j
reduce tuttine first stage pressuru ta < 140 psig, equivalent l
~
to TIEM4L POWER 1ess than 315 of RATED TIENI4L POWER, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
1 Verify all insertable' control rods to be insertad within I hour.
l l
ACTION 7 t
/
ACTI001 8 Lock the reactor mode switch in the Shutdown positten within f
1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
g i
Suspend all operations involving CORE ALTERATI0fts,* and insert i
ACTION 9 i
all insertable control rods and lock the reactor mode switch in
~
j the SWTDOWN position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
g
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i "Except movement of I M, SIDI or special movable detectors, or replacement of
]
LPAM strings provided SM instrumentation is OPERABLE per Specification 3.9.2.
i r'
j LA SALLE - UNIT 1 3/4 3-4 Amerwhent No.
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l TABLE 3.3.1-2 l
1 E
REACTOR PROTECTION SYSTEM RESPONSE T M S h
i E
RESPONSE T M i
(Seconds)
FUNCTIOML UNIT 4
g 1.
Intermediate Range Monitors-M l
Neutron Flux - High*
M l
a.
b.
Insperative j
2.
Average Power Range Moniter*
M Neutron Flux - High, Setdown
< 0.09 i
a.
b.
Flow Biased Simulated Thermal Power-Upscale 7 0.09 i
Fixed Neutron Flux - High b
}
c.
d.
Inoperative
< 0.55 3.
Reacter Vessel Steam Dome Pressure - Nigh 7 1.05 w) 4.
Reacter Vessel ideter Level
. Low, Level 3 7 0.86 i
5.
Mein Steam 1.ine Isolation Valve Closure L
4 6.
t hia "-
i== -
- =H a= - "'*
l w
M 7.
Primary Containment Pressure - Nigh M
t'b
'd 8.
Scram Discharge Volume Ideter Level - Nigh
< 0.06 9.
Turbine Stop Valve - Closure
- 10. Tudine Control Valve Fast closure.
< 0.0S Trip 011 Pressure - Low b
- 11. Reacter Mode. Switch Shutdown Position M
12: Manuel Scram l
- 13. Control Rod Drive i
M Charging Ideter Feeder Pressure - Low a.
M b.
Delay Timer
" Neutron detectors are esempt from response' time testing. Response time shall be measured from the detector output er from the input of the first electronic component in the channel.
- Not including simulated thermal power time constant.
m fMeasured from start of turbine control valve fast closure.
.F 4
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T m r d.3.1.1-1 pracToR PROTECTION SYSTE3R IMSTRIEEElrFATION SURVEILLARACE REDDIREMENTE CHANNEL OPERATIONAL CHAMEL FUNCTIONAL CHANNEL COMITIONS FOR WIM FUIACTIGIRAL UMIT CHECK TEST CALIBRATION *I MEWEILLhaE'E REEMf1 BED I
1.
Intermediate Range Monitors a.
Neutron Flux - High S/UN.S S/UICI, W R
2+
S W
R 3*.
4, 5 b.
Inoperative NA W
NA 2*.
3*,
- 4. 5 2.
Average Power Range Monitor:"I a.
Neutron Flux - High, Setdown S/UN,8 S/U 'I, W
SA 2*-
I S
W SA 3*,
5 b.
Flow Blased simulated ThermalD '3 S / U '* 3, Q WId3IC3 SA, Rihl g
I Power-Upscale S.
e.
Fixed Neutron Flux -
High S
S/U *3, Q W 'd '. SA 1
I d.
Inoperative NA Q
NA 1,2,3,5 3.
Reactor vessel Steam Dome Pressure - High NA Q
Q 1, 2 4.
Low, Level 3 S
Q R
1, 2 5.
Main Steam Line Isolation valve - Closure NA Q
R 1
6.$ Nain Steam Line Radiation -
)
High S
Q R
1, 2 7.
Primary Containment Pressure -
High NA Q
Q 1, 2 W&~$
3/4 3-7 Amendment No. 104 LA SAI.LE - (TNIT 1
(
I 1
\\g TABLE 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICABLE OPERATED BY CHANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL TRIP SYSTEM (b)
C0fSITION ACTION l
A.
AUTOMATIC INITIATION 1.
PRIMARY CONTAlletENT ISOLATION a.
Low, Level 3 7
2 1, 2, 3 20 J
Low Low, Level 2 2, 3 2
1, 2, 3 20 d
Low Low Low, Level 1 1, 10 2
1, 2, 3 20 b.
Drywell Pressure - High 2, 7, 10 2
1, 2, 3 20 M
c.
Main Steas Line 1)
Radialton - High 1
2 1, 2, 3 21 3
.s-2
- 1. 2. 3 22 2)
Pressure - Low I
2 1
23 3)
Flow - High 1
2/11ne
- 1. ?. 3 21 d.
Main Steam Line Tunnel SELETE Temperature - High 1
2 1 ',8833 2OHj)
,3,3 e.
Main Steam Line Tunnel ATemperature - High 1
2 1 ',8',3 3 2dHj)
,,3 f.
Condenser Vacuum - Low 1
2 1,
2*, 3*
21 2.
SICONDARY CONTAllelENT ISOLATION a.
Reactor Building Vent Exhaust Plenum Radiation - High 4'*"*8 2
1, 2, 3 and **^
24 b.
Drywell Pressure - High 4 ' * "
2 I, 2, 3 24 c.
Reactor Vesse! Water Level - Low Low, Level 2 4 '*" **
2 1, 2, 3, and
- 24 d.
Fuel Pool Vent Exhaust Radiation - High 4'*"*3 2
1, 2, 3, and **
24 LA SALLE - UNIT 1 3/4 3-11 Amendment No.Jee-
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TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATIM RESP 0NSE Ty m>
d TRIP FLNICTION RESPtMISE TIME fSecondsW A.
AUTpMATIC INITIATION 1.
PRIMARY CONTAllBENT IS0LATIM a.
Low, Level 3 N/A 2)
Low Low, Level 2 N/A 3)
Low Low Low, Level 1 g 1.0*
b.
Drywell Pressure - High N/A c.
Maim St! e Line 1
Radiation - HiglF' s 1.0*
2 Pressure - Low
$ Z.0*
ELErEb Flow - High s 0.5*
d.
Main Steam Line Tunnel Temperature - High N/A e.
Condenser Vacu m - Low N/A f.
Main Steam Line Tunnel ATemperature - High N/A 2.
SECONDARY CONTAlletENT ISOLATION N/A i
Reactor Building Vent Exhaust Plenum a.
Radiation - High b.
Drywell Pressure - High
~
c.
Reactor Vessel Water Level - Low, Level 2 d.
Fuel. Pool Vent Exhaust Radiation - High 3.
REACTOR WATER CLEAN'UP SYSTEN ISOLATI M N/A a.
AFlow - High b.
Heat Exchanger Area Temperature - High c.
Heat Exchanger Area Ventilation AT-High d.
SLCS Initiation e.
Reactor Vessel Water Level - Low Low, Level 2 REACTORCn8IISOLATIONCMINGSYSTEMISOLATIbf N/A
~
4.
a.
RCIC Steam Line Flow - High b.
RCIC Steam Supply Pressure - Low c.
RCIC Turbine Exhaust Diaphragm Pressure - High d.
RCIC Equipment Room Temperature - High e.
RCIC Steam Line Tunnel Temperature - Hifih-e f.
RCIC Steam Line Tunnel ATemperature - HLgh g.
Drywell Pressure - Hifih h.
RCIC Equipment Room A"emperature - High 5.
RHR SYSTEM STEAN CONDENSING MODE ISOLATION N/A a.
RHR Equipment Area ATemperature - High b.
RHR Area Cooler Temperature - High c.
RHR Heat Exchanger Steam Supply ~ Flow High v
eaem.e
. wry i i n i in aurunurwT un av
TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEM INSTRtalENTATION RESPONSE TIME TRIP FUNCTION RESPONSE Tiff (Seconds)*
v 6.
RHR SYSTEM SHUTDOWN COOLING lannE ISOLATION N/A a.
Reactor Vessel Water Level - Low, Level 3 b.
Reactor Vessel (RHR Cut-In Permissive RHR Pump Suction Flow -) HighPressure - High c.
d..
RHR Area Cooler Temperature High l
e.
RHR Equipment Area AT High i
l B.
MANUAL INITIATION N/A 1.
Inboard Valves 2.
Outboard Valves 3.
Inboard Valves 4.
Outboard Valves 5.
Inboard Valves 0.
Outboard Valves 7.
Outboard Valve TABLE NOTATIONS Isolation system instrumentation response time for MSIVs only.
No diesel generator delays assumed.
p Radiation detectors are exempt from response time testing. Response time shall be measured.from detector mutput or the input of the first electronic component in the channel.
6 Isolation system instrumentation response time specified for the Trip Function actuating the MSIVs shall be.added'to'MSIV holation time to l
obtain ISOLATION SYSTEN RESPONSE TIME for each valve.
hELETEb
.';t:;
l N/A Not Applicable.
v LA SALLE - UNIT 1 3/4 3-19 Amendment No. g
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CHANNEL OPERATICERL I
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR OWIICH-TRIP FUETIM CHECK TEST PM EEENEl m $3 nit 1 RED
(
A.
AUTosakTfc IMITIATICIE 1.
PRItshRY CGAITAIREEEarF IBGLATIGIf
~
I a.
1)
Low, Level 3 S
Q R
1,2,3 l
2)
Low Low, Level 2 NA Q
R 1,2,3
[
3)
Low Low law. Level 1 S
Q R
.1, 2, 3'
[
b.
. Drywell Pressure - High NA O
O 1, 2, 3 c.
/'
1)
(Radiation - Kleh S
Q R
1, 2, 3
2)
Pressure - Low NA Q
Q 1
60
.i 3)
Flow - High NA Q
R 1, 2, 3.
I i
d.
Main Steam Line Tunnel Temperature - High NA Q
R 1, 2 3 g
e.
Condenser Vacuum - Low NA 0
0 1, 2, 3, t.
Main steam Line Tunnel
& Temperature - High NA O
R 1,2,3 r
I 2.
SEcMDhRY CCMTAIISIEffT ISOLATION
}
a.
Reactor Building vent Exhaust Plenum Radiation - High S
Q R
1, 2, 3, and *
- i b.
Drywell Pressure - High NA Q
Q 1,2,3 j
c.
Reactor Vessel Water Level - Low Low, Level 2 NA Q
R 1, 2, 3, and 9 j
d.
Fuel Fool Vent Exhaust Radiation 6 High S
O R
1, 2, 3, and "
3.
REACTOR M&TER CLEARIUP SYSTERE ISOLhTIOBf l
a.
A Flow - High S
O R
1, 2, 3 i
b.
Neat Exchanger Area l
Temperature - High NA Q
Q 1,2,3 l
l c.
Nest Exchanger Aree Ventilation A1' - High NA Q
Q 1,2,3 d.
SLCS Initiatios MA R
NA 1,2,3 l
e.
Reactor Vessel Water Level - Lev Low, Level 2 NA Q
R 1,2,3
[
l l
i LA SALLE - UNIT 1 3/4 3-20 Amendment No.
104
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i LIMITIldi SAFETY SYSTEM SETTINGS V
jg[S REACTOR PROTECTION SYSTEM INSTRipENTATION SETPollffS (Continued) i 4.
Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to assure that there is adequate protection for the fuel and pressure limits.
5.
Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit
)
the amount of fission product release for certain postulated events. The M
i MSIV's are closed automatically /from measured parameters such as high steam flow. thigh steam line radiationalow reactor water level, high steam tunnel temperature and low steam l' ne pressure. Tlie MSIV's closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby i
protects reactor vessel pressure and fuel theriaal/lydraulic Safety Limits.
1 6.
' Main Steam Line Radiation-High' r
The main steam line radiation detectors are provided to detect a gross failure of the fuel cladding. When the high radiation is detected a trip is initiated to reduce the continued failure of fuel cladding. At the same time the main steam line isolation valves are closed to limit the release of fission products. The trip setting is high enough above background radiation levels to prevent spurious trips yet low enough to promptly detect gross failures in the fuel cladding. No credit was taken for operation of this. trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection Systan.
7.
Primary Containment Pressure-High High pressure in the drywell could indicate a break in the primary pres-sure boundary systems. The reactor is tripped in order to minimize the possi-bility of fuel damage and reduce the amount of enery being added to the coolant. The trip setting was selected as low as possible without causing spurious trips.
O Y
Dele +d m
LA SALLE - UNIT 2 8 2-11
(a, _
b
}
\\
TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRNENTATION
[
APPLICABLE MINIORM OPERABLE OPERATIONAL CHANNELS PER FUNCTIONAL INIIT ColWITIONS TRIP SYSTEM (a)
ACTION E
U 1.
a.
Neutron Flux - High 2 :.
3 1
- A t
t 3$I SI.
~
3 3
3 1
b.
Insperative p
2 4
- 3. 4 2
2 5
3 3
5 I"I 2.
Average Powef Range Moniter:
s.
Neutron Flux - High. setdown 2
2 1
M 3
2 2
5(b) 2 3
e w
4.
~b... Flow Slased' Simulated Thermal.
Power-Upscale 1
2 4
I c.
Fixed Itsutron Flux-Migh 1
2 4
j d.
Insperative
- 1. 2 2
1 I
3 2
2 5
2 3
3.
Reacter Vessel Steam Dome Pressure - High
- 1. 2(d) 2 1
4.
Reacter Vessel Water Level - Law,
- Level 3
- 1. 2 2
1 be]tY.
1 5.
Main Steam Line Isolation Valve -
I *I 4
4 I
Closure
/
6.IMainSteamLineRadiation-1,2(*)
2 5
\\
Hiah
4 4
E l!I i
a
- . R7 >
TABLE 3.3.1-f (Continued)
I ilEACTM PROTECTIN SYSTBI Ill5TRLpWITAT10N i
ACTI0ll STATB Wif$.
l ACTICII 1 Be in at least IST SMTD0lel within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i Verify all insertable control rods to be inserted in the core ACTICII 2 l
and lack the reacter made switch in the Shutdeun position within I hour.
l ACTION 3 Suspend all operatient involving CDilE ALTERATI0Il5*_and insert all l
insertable control rods within one hour.
{g Be in at leekt STARTUP within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
ACTI0li 4 Be in STAltTUP with the sein steen line isolation valves closed ACTI0li 5 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or in at least BET SWTD0601 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Initiate a reduction in THENihL POWER within 15 minutes and ACTIoll 6 reduce turbine first stage pressure to < 140 psis, equivalent to TIE 244L POWER less than SIB of RATED ~ HEW 14L POWER, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
X ACTION 7 Verify. all insertable control rods to be inserted within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Lock the reactor mode switch.in the Shutdown positi,on within ACTION 8 I hour.
+
4 Suspend all operations involving CORE ALTERATI0ll5,* and insert ACTION 9 all insertable control rods and lock the reactor mode switch in the*S WTD0let position within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
- Except movement of 151 SWI, or special movable detectors, or replacement of LPNI strings provided SWI instrumentation is OPERABLE per Specification 3.9.2.
(^
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l TA8tE 3.3.1-2 REACTOR PROTECTf511' W 5fEF RESPONSE TIMES E
TABLE 3.3.1-2
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REACTOR PROTECTION SYSTEN RESPONSE TIES i
l E;
RESPONSE TIE FUNCTIONAL UNIT (Seconds) 1.
a.
Neutron Flux 'High*
m b.
Inoperative m
2.
Average Power Range Monitor
- a.
Neutren. Flux - High, Setdown m
b.
Flow Stased Simulated Thersel Power-Upscale
< 0.09 w
c.
Fixed Neutron Flen - High I 0.09 1
d.
Inoperative NA w
J, 3.
Reacter Vessel Steam Dome Pressure - High
< 0.55 4.
Reacter Vessel Water Level - Low, Level 3 i 1.05 5.
Main Steam Line Isolation Valve - Closure 10.06 6.
inmin v-- una --nauen - mm m
_i
~
7.
PrimaryContainmentPressure-Nip m
(
8.
Scram Olscharge Volume Water Lewes - High m
(
De(ded 9.
Turbine Step Valve - Closure
< 0.06 l
~
- 10. Tertine Control Valve Fast Closure.
i
' trip 011 Pressure - Low
< 0.00
- 11. Reacter Mode Switch Shutdeun Position H4
- 12. Manual Scram m
i
- 13. Control Rod Drive Charging Water Needer Pressure - Low m
g a.
b.
Delay Timer M
iN
= neutron detectors are exempt from reopense time testing.
Response time shall be measured N
from the detector output:er from the input of the first electronic component in the channel.
/
- Net including simulated theries) power time constant.
- Nessured from start of turbine centrol valve fast closure.
m.
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,N R
2*
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R 3*,
4, 5 b.
Inoperative IIA N
NA 2 *, 3 *, 4, 5 P
2.
Average Power Nanage Monitor III a.
Neutron Flux - High, i
setdown '
s/uth) s s/u(CI, W sA 2*
l s
W sA 3*,
5 i
b.
Floor Blased simaalated thermal power-opeoale s, DIGI s/UICI, 9 WIdI I*I, sA R(h) g e.
Fixed neutron Flux -
Mieh a
steICI, O WIdI, sa 1
d.
Inoperative NA 0
38 4 1, 2, 3, 5 f
3.
Reacter vessel steen Dome Pressure - Mish NA 0
0 1, 2
[
t 4.
meester vessel tenator f.evel -
Low, Level 3 3
0 R
1, 2 5.
Semin stees Line Isolation Valve - Closure NA Q
R 1
6.
Main steen Line Radiation -
l l
Nigh s
G R
1, 2 i
bELE1"Eb 7.
Primary Contailuneet Pressure -
l
.. Nigh 38 4 9
Q 1, 2 l
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TABLE 3.3.2-3 1
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ISDLATIM SYSTDI INSTMMENTATION RESPONSE TI4 TRIP FLEETINI RESPGtSE TI4 (Seconds)#
v A.
AtmBIATIC INITIATION 1.
PRIMARY CGITADDENT IS014 TION a.
Reactor Vessel Nater Level 1)
Low, Level 3 N/A 2)
Low Low, '.evel 2 N/A 3)
Low Low i.ow, Level I s 1.0*
b.
Drywell Pressure - High N/A c.
Main Steam Line I) l Radiation - High'~~'
s. 0*
2)
Pressure - Low 5 U.0*
3)
Flow - High 5 0.5*
d.
Main Steam Line Tunnel Temperature - High N/A e.
Condenser Vacuum - Low N/A f.
Main Steam Line Tunnel ATemperature - High N/A AELErdb 2.
MCONDARY CONTAllelENT ISOLATION N/A a.-
Reactor Building Vent Exhaust Plenum Radiation - Hirt b.
Drywell Pressure - High c.
Reactor Vessel Water Level - Low, Level 2 d.
Fuel Pool Vent Exhaust Radiation - High 3.
REACTOR WATER CLEANUP SYSTDI ISOLATION N/A a.
AFlow - High s.
b.
Haat Exchanger Area Temperature - High c.
Heat Exchanger Area.. Ventilation AT-High d.
SLCS Initiation
~
e.
Reactor Vessel Water Level - Low Low, Level 2 4.
REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION N/A a.
RCIC Steam Line Flow - High b.
RCIC Steam Supply Pressure - Low c.
RCIC Turbine Exhaust Diaphragm Pressure - High d.
RCIC Equipment Room Temperature - High e.
RCIC Steam Line Tunnel Temperature - High -
f.
RCIC Steam Line Tunnel ATemperatura - High -
g.
Drywell Pressure - High h.
RCIC Equipment Room ATemperature - High 5.
RHR SYSTEM STEAM C0fGENSING MODE ISOLATION N/A a.
RHR Equipment Area ATemperature - High y--
b.
RHR Area Cooler Temperature - High c.
RHR Heat Exchanger Steam Supply Flow High U
LA SALLE - UNIT 2 3/4 3-18 AMENDHENT NO. g
g TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEN INSTRUMENTATION RESPONSE TINE TRIP FIRICTION MSPONSE TIME fSeconds)#
6.
RHR SYSTEN SHUTDOWN C00 LING fEIDE ISOLATION N/A l
a.
actor Vesse" Nater f.evel - Low, Level 3 oN{h'1
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MANUAL INITIATION N/A lnboardValves wtboardVgives I
$b!ardVfes 4'
l l$boardV es Outboard Va've TABLE NOTATIONS Isolation system instrumentation response time for MSIVs only. No diesel generator d'elays assumed.
p e ectronic compo.37rTd.imor'o7o,.I'U"in.H 7, l'oPtle rffsP" "
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Yh b
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N/A Not Applicable.
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'OPgRATIONRL CNANNEL FUNCTIONAL CNANNEL CONDITIONS FOR-WICM TRIP FUNCTIGIl CHErE TEST CALIBRATION SDRWRILL&Mct REDulaED A.
AUTOMATIC IMITIATION 1.
PRIM &RY CONTAISARENT ISOL&TIGAR a.
Reactor vessel. Water Level 1)
Low. Level 3 S
O R
- 1. 2, 3 2)
Low Low. Level 2
, MA Q
R
- 1. 2, 3 L
31 Low Low Low. Level 1 8
O R
1,2,3 b.
Drywell Pressure - Nigh NA
-Q Q
- 1. 2, 3
[
c.
Main Steen Line W
11 lRadiatten - Nigh S
9 R
1,2,3L 21 Fressure - 3.ew an 9
e 1
i 3)
Flow - Nigh NA O
R 1, 2 3 d.
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R
- 1. 2. 3
- ~
l e.
condenser vacuan - Low M4 9
Q
- 1. 2*.
3*
i C.
Main Steen Line Tunnel
& Tenperature - Nigh NA O
R
- 1. 2, 3 I
2.
maconsnaa? coarvaramanne tanLawrnas i
a.
Reactor Building Vent Exhaust l
Planen Radiation - Nigh 8
O R
- 1. 2. 3 and **
j b.
Drywell Pressure - Nigh NA 9
0
- 1. 2, 3 l
c.
Reactor vessel water Level - Low Low, Level 2 NA 9
R
- 1. 2, 3, and 9 d.
Fuel Fool Vent Exhaust Radiation - Nigh S
Q R
1, 2. 3.and **
l 3.
RancTom ts&TER cLE&aRIP SYSTERI tacLATIcel a.
A Flow - Nigh S
9 R
1,2,3 b.
Heat Exchanger Area Tenperature - Nigh NA 0
0
- 1. 2, 3 c.
Heat Exchanger Area j
ventilation AT - Nigh NA Q
Q
- 1. 2. 3 i
d.
sLc5 Initiation NA R
NA
- 1. 2. 3 e.
Reactor vessel water Level - Low Low. Level 2 NA Q
R 1, 2, 3
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l f
LA SALLE - UNIT 2 3/4 3-20 Amendment No. M
L Page C-1
{om)
ATTACHMENT C SIGNIFICANr HAZARDS CONSIDERATION
)
' Ibis pmposed Technical Specincation change removes the scram function and oortain Gmups 1 and 8 isolation valve closum fhan+1ans nasociated with the Main Steam Line Radiation Monitoring Od812M) '
spetam at la Salle County Station (LSCS) on high radiatian. N shanges am as follows:
Beesove the mactor scram function Remove the automatic closum of the Main Steam Isolation Valves (MSIVs)
Remove the automatic closum of the Reactor Racinulation Water Sample Line Isolation Valves and Main Steam Line Drain Isolation Valves.
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 dadining a significant hazards consideration established in 10 CFR 50.92, q
operation of LaSalle County Station Units 1 and 2 in acconiance with the proposed amendment will not:
g, 1)
Involve a significant incmase in the probability or consequences of an accident previously evaluated because:
Redefining the full power radiation background, thus changing the MSLRM alarm setpoint, does not change the probability of occurrence of any accident which has been postulated and analyzed in the UFSAR, but will reduce the probability of the inadvertent MSIV closure transient which is an analyzed transient in the UFSAR. It does not change the probability of malfunction of any equipment important to safety associated with IDCA, Fuel Handling Accident or CRDA. It also does not change the resultant offsite radiological dose from the bounding design basis CRDA. This is based upon all radioactivity, I resulting from the design basis CRDA, going to the condenser instantaneously (or independent of the actual MSLRM setpoint) in the offsite dose calculation.
'Ibe elimination of reactor scram and isolation of MSIVs, isolation of main steam line drain valves and reactor water sample line valves, associated with the MSLRM system
[)
actuation do not introduce, mitigate, or reduce the probability V
of any design basis accident, or any accident, evaluated in the UFSAR. The topical report NED43M00A has shown that there is essentially no reasonable radiological consequence
4 i
Page C-2 i e@
ATTACHMIDFF C i
SIGNIMCANT HAZARDS CONSIDMIATION i
beneGt in a design basis CRDA of staining the MSLRM l
associated macter scram and MSIV isolation function. In addition, the pmbebility d inadvertent scram and isolation is l
mduced. N proposed change will not adversely hapact the i
operation of the RPS or PCIS with mopect to perdenning its l
ether in*andad safety fane*==
%e papoemd change will not i
asest the operation of other plant systeens or +n
-d important to safety. he consequences of eliminating the automatic closum of the main steam line drain isolation valves and reactor recirculation water sample line isolation valves along with the MSIVs has been evaluated to be negligible l
additions to the CRDA doses. A LSCS unique analysis has demonstrated that the radiological doses as a result of design l
basis CRDA are acceptable.
l he MSLRM system high radiation trip was intended to function in response to a CRDA which has been previously ilq evaluated. No cndit for MSIV closure was taken in the CRDA analysis since it postulates that all the radiandive material l
l assumed to be released from the fuel is transported to the main condenser prior to MSIV closure. Furthermore, the probability of a fuel failure is independent of the operation of the MSLRM system.
By eliminating the MSLRM induced MSIV closure, the Ofigas system can be utilized to reduce potential offsite doses after a CRDA. The MVP is tripped no later than 15 minutes of a Hi-Hi radiation alarm but analytically results in acceptable offsite doses.
%us the pmposed amendment will not increase the probability of any accident previously evaluated, and the elimination of the M8LRM isolation signal for MSIVs and other small containment valves will not significantly increase the consequences of a CRDA as previously evaluated, j
2)
Create the possibility of a new or different kind of accident from any accident previously evaluated because:
/N Redefining the full power radiation background, thus changing C
the actual MSLRM alarm setpoint, does not alter the configuration of the plant. It does not revise any logic or function of the MSLRM trip channels or add, replace, or delete
.w--y v
I gj Page C-3 r-ATTACHMIDff C SIGNIFICANT HAZARDS CONSIDERATION i
any equipment important to safety. %erefom it does not introduos any new failum modes or arente any possibility of a new meeldant which may challenge safety to the public and has not been previously analysed. It ales does not involve any equipment which either has not been evaluated pavviously, or may have any esfety conesqueness to the public.
%e proposed Technical Specification changes involve eliminating the MSLHM system high radiation trip function for I initiating an automatic mactor scram, and automatic isolations.
%e proposed changes will not affect the operation of other plant systems or equipment important to safety. The MSLRM system will continue to initiate alarms as before. Piant procedums will be in place to take appropriate mitigative measures in response to a high alarm.
%e isolation and reactor scram functions associated with the r3 MSLRM system actuation was originally intended to mitigate, not pavent, a potential accident acanario such as a CRDA or gross fuel failure event. Adding or removing an electronic signal, such as the one from the MSLRM system, does not change system or hardware design within the reactor vessel pressure boundary, and therefore will not emate the possibility of a new or different kind of accident from those evaluated in the UFSAR like a LOCA or CRDA during power operation. It also does not create the possibility of a new or different kind of accident outside the reactor vessel pressure boundary from those evaluated in the UFSAR, such as a IDCA or Puel Handling Accident. Removing the isolation signal also reduces the probability of inadvertent scram and isolation.
%erefore the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously analyzed.
3)
Involve a significant reduction in the margin of safety because:
%e current MSLRM trip Hi-Hi alarm setpoint (about 4 IVhour with full power background at 1.3 IVhour) is at 3 times the full
/ N power radiation background. As indicated in the plant unique d
analytical result for LSCS, the radiological reading at the MSLRMs for design basis CRDA is equivalent to over 1200 times the normal full power radiation background (1600 IVhour
1
'c Page C-4 E
ATFACHMIDir C SIGNIFICANT HAEARDS OORWIDERATION divided by 1.8 IVhour), or 150 tienee the full power radiation backymund during peak HWC envire==aat (sines the radiation backsmund is 8 times the normaal liaaksmund). hus the sefsty asargin was very large, and would still be quite large with the HWC backsmund factored into the Mmm setmation setpoint (8 x 8 x 1.8 = about 50). He Hi alarm estpoint of 1.5 tisses full power backsmund likewise will have a bigbar safety margin. Dus there is basically no adverse consequence to the margin of safety in the basis for the LSCS techniemi specifications.
He pmposed Tachnical Specification changes to eliminate the M8LRM system high radiation trip function for initiating an automatic reactor scram, and automatic closum of the MSIVs, main steam line drain isolation valves, and mactor recirculation water sample line isolation valves do not cause radiological dose consequences to aveaad the limit established r
by SRP 15.4.9.
Per NEDO-31400A, the elimination of MSLRM tripfscram signal will result in the reduction of potential inadvertent scrams, unnecessary safety-related actuations, undue vessel isolation, and duty challenges during normal plant operation.
These can be interpreted to be a potential reduction in core damage frequency, which translates to an improvement in the margin of safety.
%us the margin of safety as defined in the basis of the technical specifications is essentially unaffected, and is therefore acceptable.
Guidance has been provided in " Final Pmcedures and Standards on No Significant Hazards Considerations," Final Rule, 51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hasards considerations.
His d-ment provides examples of amendments which are and are not considered likely to involve significant hazards considerations.
%ese proposed amendments most closely fit the example of a change which may either result in some increase to the probability or
/)
consequences of a previously analyzed accident or may reduce in some G'
way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in Standard Review Plan.
k A
. J y SIGNIFICANT HAEARDS CONSIDERATION 11de proposed==aad==at does not involve a signi5 cant relaxation of the criteria used to establish eefety limite, a significant relaxation of the basse for the li=l*ta-esfety system settinge or a signi5 cant relemation of the bases for the limiting conditions for operations.
1 Thandere, based on the guidance pewided in the Federal Register and the criteria establinhad in 10 CFR 50.92(e), the proposed change does not eenetitute a i-aihant hazards anamidaration.
r 4
l l
i I
ATTACHMENT D Q-1 BlVDIONMENTAL ASSESSMENT STATIBGNP APPLICABILITY 1
i REVIEW e
a i
Cammanwealth Edison has evaluated the proposed amandmant against the criteria for identification of licensing and agulatory estians mquiring envimamental sama=== ant in accordanam with 10 CFR Pad 51.21. It has been determined that the paposed l
changes meet the criteria for categorical exclusion as pmvided for under 10 CFR Part 51.22(cX9). 'Ibe requested changes will have no impact on the environment. This conclusion has been determinad because the changes requested do not pose significant hazards considerations or do not involve a significant increase in the amounts, and no significant changes in the types of any effluent that may be mieased off-site. Additionally, this request does not involve a significant increase in i.ndividual or cumulative occupational radiation exposure.
s ge s
i ll. tl L :ly'.
l 1
1 J
ATTACHMENT E 1
- 'T i
.,,./
NEDO-81400A, Safety Evaluation ibr Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor, dated October 1992.
/T
I; 7 '
This proposed Technical Specification change removes the scram function and the Group 1 and 3 isolation valve closure functions associated with the MSLRM system for high radiation. The changes are as follows:
Remove the reactor scram function Remove the automatic closure of the Main Steam isolation Valves (MSIVs)
Remove the automatic closure of the Reactor Recirculation Water Sample Line Isolation Valves and Main Steam Line Drain isolation Valves Elimination of these functions willimprove plant availability by reducing spurious scrams and isolations caused by MSLRM system. Since LaSalle is proposing to eliminate automatic reactor scram and closure of the MSIVs on high radiation or inoperable trips, the references to MSLRM trip instrumentation will be removed from the Technical Specifications.. The existing alami signals, which are not part of the current Technical Specifications, will remain functional.
This proposed amendment request is subdivided as follows:
1.
Attachment A is a Description of the Safety Analysis of the Proposed Changes.
2.
Attachment B are Proposed Changes to the Technical Specifications for Operating Licenses NPF-11 and NPF-18.
i 3.
Attachment C is a Significant Hazards Consideration.
4.
Attachment D is an Environmental Assessment Statement Applicability Review.
5.
Attachment E is NEDO-31400A, Safety Evaluation for Eliminating the Boiling Water Reactor Main Steam Line Isolation Valve Closure Function and Scram Function of the Main Steam Line Radiation Monitor, dated October 1992.
6.
Attachment F is a General Electric Letter, J. Chase to S. P. Brown, Regarding MSL Radiation Trip Removal.
7.
Attachment G is the LaSalle MSLRM Scram and Isolation Trip Elimination - Offsite Dose Analysis for NEDO-31400 Scenario 2.
8.
Attachment H is the Offsite Dose Impact of Eliminating Automatic Isolation of LaSalle Sample Lines and Main Steam Drain Lines.
"