ML20034B031
| ML20034B031 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/16/1990 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20034B030 | List: |
| References | |
| NUDOCS 9004250232 | |
| Download: ML20034B031 (36) | |
Text
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AITAcilPUDIT_B ER0t0 SED _CilANGES TO THE TECHNICAL _SEEclElCATIONS FOR OPERATING LICENSE NPF-11 AND NPF-1B-HPF-11 HEE _lB REVISED PAGES:
VIII VIII 3/4 6-41 XV<
3/4 6-42 3/4 6-44 3/4 7-4 3/4 6-45 3/4 7-5 3/4 7-4 3/4 7-6 3/4 7-5 B 3/4 7-1 3/4 7-6 B 3/4 7-1 i
0513T:6 j
9004250232 900416 PDR ADOCK 05000373 PDC p
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INDEX
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS t
SECTION PAGE 3/4.7 PLANT SYSTEMS L
3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS Residual Heat Removal Service Water System...................
3/4 7 1 Diesel Generator Cooling Water System........................
3/4 7-2
]
Ultimate Heat Sink...........................................
3/4 7-3 3/4.7.2 CONTROLROOMkNDAUXILIARYELECTRICEnUIPMENTponu EMERGENCY FILTRATION 5YditM..................................
3/4 7-4 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM........................
3/4-7-7 3/4.7.4 SEALED SOURCE C0NYAMINATION..................................
3/4 7-9~
3/4.7.S FIRE SUPPRESSION SYST'MS E
I Fire Suppression Water System................................-
3/4 7-11 Deluge and/or Sprinkler. Systems..............................
3/4 7-14 CO Sy s tem s..................................................
3 /4 7 - 17 2
l Fire Hose Stations...........................................
3/4 7 18 3/4.7.6 FIRE RATED ASSEMBLIES........................................
3/4 7-22 3/4.7.7 AREA TEMPERATURE MONITORING..................................
3/4 7-24 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS I STRUCTURES...................
3/4 7-26 i
l 3/4.7.9 SNUB B E RS....................................................... 3 /4 7-2 7 3/4.7.10 MAIN TURBINE BYPASS SYSTEM....................................
3/4 7-33
[
i LA SALLE - UNIT 1 VIII Amendment No. 18 i-
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7,
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.
At least 'once per 18 months or (1) after any structural maintenar.ce.
i on the HEPA filter or charcoal adsorber housings, or (2) followin painting, fire, or chemical release in any ventilation zone j
communicating with the subsystem by:
b eDANCE 1.
Verifying that the subsystem satisfies the in pla sting of Q33 acceptance criteri and uses the test procedure of egulatory NQM 0 05 /o 8
Re on 2 ch 19, a d t, syst t is'4h0 cfm i 10%.
fW37R4T10M 2.
Verifying within 31 days after removal that a laboratory. analysis of a' representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2.
March 1978. seetm the laboratorv tentina :riteria of1 Regulatory ]A Position c.6.a o" Regulatory Guide
.52, tev'sion 2, March 1978. ; -
3.
Verifying a subsystem flow rate of 4000 cfm + 10% during systemo
' operation when tested in accordance with ANSI N510-1975.
Mgg y [c.
After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a i
l A
representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Ravinion 2. March 1978.
meets the laboratory testina cri'eria oflRegu' atory Position C.6.a )(
of Regulatory Guide 1.52, Revision 2, March 1978 7 d.
- At least once per 18 months by:-
~
1.
Verifying that the pressure drop across the combined HEPA filters-and charcoal adsorber banks is less than or equal to 8 inches Water Gauge while operating the filter train at a flow rate of 4000 cfm i 10%.
~
2.
Verifying that the' filter train starts and isolation dampers open on each of the following test signals:
a.
Reactor Building exhaust plenum radiation - high, b.
Drywell pressure - high, c.
Reactor vessel water level - low low, level 2, and d.
Fuel pool vent exhaust radiation - high.
3.
Verifying that the heaters dissipate 23 1 2.0 kw when tested in j
accordance with ANSI N510-1975.
This reading shall include the-appropriate correction for variations from 480 volts at the bus.
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LA SALLE - UNIT 1 3/4 6-41 Amendment No. 21
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CONTAINMENT SYSTEMS
[
SURVEILLANCE REQUIREMENTS (Continued)
After each complete or partial replacement of a HEPA filter bank by.
verifying that the HEPA filter banks remove greater than or equal:to 99% of the DOP when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm 1 10%.
l f.
. After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers-remove greater than or-i equal to 99% of a halogenated hydrocarbon refrigerant test gas.when l
they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of'4000_cfm i 10%.
1" I
L Rapt.Aca wiya.wszer : 3 t
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J LA SALLE - UNIT 1 3/4 6-42
'c
- PLANT SYSTEMS A
3/4.7.2-CONTROL 'R00MhND AUXILIARY ELECTRIC EQUIPMENT RO FILTRATION SYSTEM T
q.l LIMITING CONDITION FOR OPERATION-3.7.2 Two. independent control room (and auxiliary electric equipment roo emergency filtration system trains shall be OPERABLE."-
APPLICABILITY:
All OPERATIONAL. CONDITIONS and *.
ACTION:
l With one emergency filtration system train inoperable, restore the a.
inoperable train to OPERABLE status within 7 days or:
1.
In OPERATIONAL CONDITIONS 1, 2, 3, be in at least HOT SHUiOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
In OPERATIONAL CONDITION 4, 5 or
- operation of the OPERABLE emergenc, initiat'e and maintain pressurization mode of operation. y filtration system in the b.
With both emergency filtration s OPERATIONAL CONDITION 4, 5-or *,ystem trains inoperable, in of irradiated fuel in' the secondary containment and operations withs a potential for draining the reactor vessel.
G The provisions of Specification 3.0.3 are not applicable in c.
Operational condition *,
SURVEILLANCE REQUIREMENTS 4.7.2 Each control roomCand auxiliarv electric ecuSment roomJemergency; C
filtration system train shall be demonstrated OPEFA0.E:
)
a.
At least once per 31 days.on a STAGGERED TEST BASIS by initiating, l
from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the train operates for at least l
10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters OPERABLE.
i
?When irradiated fuel is being handled in the secondary containment.
i The normal or emergency power source may be inoperable in OPERATIONAL.
CONDITION 4, 5 or
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4 I
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-LA SALLE - UNIT 1 3/4 7-4 1.
.n f,.
I PLANT SYSTEMS l
SURVEILLANCE REQUIREMENTS (Continued) [
At least once per 18 monthor (1) after any structural maintenance b.
on the HEPA filter or charcoal adsorber housings, or (2) following i
painting, fire or chemical release in any ventilation zone communicating with the train by:
1.
Verifying that the train satisfies the in place g MES THAQ acceptance criteri and uses the test procedures aulatory 4
i Positions C.5.a.
.c and C.5.d of Regulatory G 2.52, j g*f pansTRKf80#
Revision 2, Mar 1978,'and the train flow rate is 4000 cfm i 10%.
gAnysr 2.
Verifying within 31 days after removal that a laborator IN-analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2. March 1978, meets the laboratorv tadiac criterinfef Regulatory Position C.5.a of Regulatory.
Guide 1.52, Rev ~sson 2, March 1978.
3.
Verifying a train flow rate of 4000 cfm + 10% during subsystem operation when tested in accordance with~ ANSI N510-1975.
c.
After every 72 urs of charcoal adsorber operation by verifying fNSERT "ithi" 3 d'Y' 'ft'r "'" V*l th*t *
'6 "#
'"Y'i' representative carbon sample'obtained in accordance with Regulatory
.L
(
Positon C.6.b of Regulatory Guide 1.52. Revision 2. March 1978.
j meets the laboratory testina criteria [of Regulatory Position C.6.a]
Q of Regulatory Guide 1.52, Revision 4,- March 19787 d.
At least once per 18 months by:
l-1.
Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 8 inches Water Gauge while operating the train at a flow rate of 4000 cfm i 10%.
"This surveillance shall include the recirculating charcoal filter, " odor eater,"
in the normal control room supply filter train using ANSI N510-1975 as a guide
'l to verify >_ 70% efficiency in removing freon test gas.
- Except that recirculating charcoal filter samples,shall be removed and analyzed at least once per 18 months.
I LA SALLE - UNIT 1 3/4 7-5 Amendment No. 61 I
l
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1 PLANT SYSTEMS f;
SURVEILLANCE REQUIREMENTS (Continued):
\\[ l 2.
Verifying that on each of the below pressurization mode actuation d
test signals, the emergency train automatically switches to the
-J pressurization mode of operation and the control room is maintained at a positive pressure of 1/8 inch W.G. relative to
'the adjacent areas during emergency train operation at a flow rate less.than or equal to 4000 cfm:
a) -Outside air smoke detection, and 4
b)
Air intake radiation monitors.
3.
Verifying that the heaters dissipate 20 t 2.0 Kw-when tested-in accordance with ANSI N510-1975.
This reading shall include the appropriate' correction' for variations from 480 volts at the bus I.
After each complete ~or partial replacement of a HEPA filter bank by e
verifying that.the HEPA filter banks remove greater than'or equal to-99% of the DOP when they are tested:in place in accordance with ANSI.
.5 N510-1975 while operating the system at a flow rate of 4000 cfm i 10%.
f.
Aftegeach complete or partial replacement' of a charcoal adsorber:
l bank by verifying that the charcoal.adsorbers remove 99% of a.
}
halogenated hydrocarbon refrigerant test gas when.they are tested; in place in accordance with ANSI N510-1975 while operating the
(
system at a flow rate of 4000 cfm i 10%.
i t,
w RaptAcs v ma L
N WSERT D
I "This surveillance shall include the recirculating charcoal filter, " odor' eater,"
in the normal control room supply filter train using ANSI N510-1975 as'a guide to verify > 70% efficiency in removing freon test gas.
2
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LA SALLE - UNIT 1 3/4 7-6 Amendment No. 61 L
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3/4.7 PLANT SYSTEMS
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i BASES
[
3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS-The OPERABILITY of the core standby cooling system equipment cooli.ng water systems and the ultimate heat sink ensure that sufficient cooling capacity
'{
j is available for continued operation of safety-related equipment during normal.
1 L
and accident conditions. The redundant cooling capacity of these systems, assuming a single failure, is consistent with the assumptions used in.the l
accident conditions within acceptable limits.
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3/4.7.2 CONTROLROOM[ANDAUXILIARYELECTRICEQUIPMENTROOMEMERGENCY FFLTRATION SYSTEM
=9 i
The OPERABILITY of the control roomfand aavi' inrv alactric a?"'inment room 'I
- emergency filtration system ensures that the roost mill remain habitable for i
operations personnel during and following all-des'gn basis accident conditions..
The OPERABILITY of this system in conjunction with room design provisions based on limiting the radiation' exposure to personnel occupying.the; 5 rem or less whole body, or its equivalent.. This limitation is consistent t
with the requirements of General Design Criteria 19 of Appendix "A",
Cumulative operation of the system with the heaters OPERABLE for-i 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture j g on the adsorbers and HEPA filters.
s 3/4.7.3 REACTOR CORE ISOLATION C0OLING SYSTEM J
The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the mactor vessel without requiring actuation of any of the Emergency care Cooling System equipment. The.RCIC
}
system is conservatively required to be OPERABLE whenever reactor pressure j-exceeds 150 psig even though the LPCI mode of the the residual heat removal (RHR) system provides adequate core cooling up to 350 psig.
i The RCIC system specifications are applicable during 0PERATIONAL:
CONDITIONS 1, 2 and 3 when reactor vessel pressure exceeds 150 psig because RCIC is the primary non-ECCS source of core cooling when the reactor -is i
pressurized.
With the RCIC system inoperable, adequate core cooling is : assured by the OPERABILITY of the HPCS systens and justifies the specified 14 day out-of-service period.
The surveillance requirements provide adequate assurance that RCICS will be OPERABLE when required.
Although all active components are testable and-full, flow can be demonstrated by recirculation during reactor operation', a complete functional test requires reactor shutdown.
Initial startup test' program data may be used to determine equivalent turbine / pump capabilities between test flow path and the vessel injection flow path.
The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment.
LA SALLE - UNIT 1 B 3/4 7-1 Amendment 5
M SERI_A ASIM D 3803-79, for a methyl iodide penetration of less than-0.5% when tested at a temperature of 30'C and a relative humidity of 70%.
l HSERT_.B l
e.
After each complete or partial replacement of a HEPA filter bank by E
verifying that the HEPA filter banks have a DOP penetration of less than 0.05% when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 CFR & 10%.
i f.
After each complete or_ partial replacement of a charcoal adsorption bank l
by verifying that the charcoal adsorption beds have a halogenated hydrocarbon refrigerant test gas penetration of less than 0.05% when they are tested in-place in_accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 CFR i 10%.
E SERT _C ASIM D 3303-79, for a methyl iodide penetration of less than 10.0% when tested at a temperature of 30'C and a relative humidity of 70%.
l H SERI_D e.
.After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks have a DOP penetration of less than 1.0% when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 CFR & 10%.
f.
After each complete or partial replacement of a charcoal adsorption bank by verifying that the charcoal adsorption beds have a halogenated hydrocarbon refrigerant test gas penetration of less than 1.0% when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 CFR i 10%.
0513T:9
l INDEX LIMITIN5CONDITIONSFOROPERATIONANDSURVEILLANCEREQUIREMENTS l
SECTION PAGE 3/4.7 PLANT SYSTEMS i
3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS Residual Heat Removal Service Water System...................
3/4 7-1 Diesel Generator Cooling Water System........................
3/4 7-2 Ul ti mate Heat S i nk...................................jd>.....
3/4 7-3 o
3/4.7.2 CONTROLROON(EANDAUXILIARYELECTRICEQUIPMENTR00Mli
EMERGENCY FILTRATION sY51tM..................................
3/4-7-4 3/4.7.3 REACTOR CORE ISO LATION COOLING SYSTEM.......................
3/4 7 3/4.7.4 SEALED SOURCE CONTAMINATION..................................
3/4 7-9 l
l 3/4.7.5 FIRE SUPPRESSION SYSTEMS Fi re Suppression Water System................................. - 3/4 7-11 Del uge and/or Spri nkl er Systems..............................
3/4 7 CO Systems..................................................
3/4'7-17 i
2 1
Fire Nose Stations...........................................
3/4 7-18 3/4.7.6 FIR E RATED AS S EMB LI ES....................................... - 3/4.7-23 3/4.7.7 AREA TEMPERATURE MONITORING..................................
3/4~7-25 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS 1 STRUCTURES...................
3/4 7-27 3/4.7.9 SNUBBERS.....................................................
3/4 7-28 i
3/4.7.10 MAIN TURBINE BYPASS SYSTEM...................................
3/4 7-34 LA SALLE UNIT 2 VIII
i i
INDEX
(
BASES l
SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATERSYSTEMS......................................ggy B 3/4 7-1 3/4.7.2 CONTROL ROOM (AND AUXILIARY ELECTRIC EQUIPMENT ROOM l[ "
EMERGENCY FILTRATION SY5 TEM...........................
'B 3/4 7-1:
3/4.7.3 REAGTOR CORE ISOLATION COOLING SYSTEM...................
B 3/4.7-1 3/4.7.4 SEALED SOURCE CONTAMINATION.............................
B_3/4 7-2 3/4.7.5 FIRE SUPPRESSION SYSTEMS................................
B'3/4 7-2 3/4.7.6 FIRE RATED ASSEMBLIES...................................
B 3/4 7-3 3/4.7.7 AREA TEMPERATURE MONITORING.............................
B 3/4'7-3 3/4.7.8 STRUCTURAL INTEGRITY OF CLASS 1' STRUCTURES..............
-B 3/4 7.3/4.7.9 SNUBBERS................................................
B'3/4 7-3
'3/4.7.10 MAIN TURBINE BYPASS SYSTEM..............................
B 3/4 7-5 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 and 3/4.8.2 A.C. SOURCES AND ONSITE POWER DISTRIBUTION SYSTEMS....................................
B 3/4 8-1 3/4.8.3 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES..................
B 3/4 8 '
L 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH.....................................
B 3/4 9 ' -
3/4.9.2 INSTRUMENTATION..........................................
B 3/4 9-1 3/4.9.3 CGNTROL R0DLP0SITION....................................
B 3/4 9-1 3/4.9.4 DECAY TIME..............................................
B 3/4 9-1 3/4.9.5 COMMUNICATIONS..........................................-
B 3/4 9-1 3/4.9.6 CRANE AND H01ST.........................................
B 3/4 9-1 3/4.9.7 CRANE TRAVEL............................................
B 3/4'9-2 i
3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL. - SPENT FUEL STORAGE P00L...............
B 3/4 9-2.
3/4.9.10 CONTROL ROD REM 0 VAL.....................................
B 3/4 9-2 6
3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT-CIRCULATION...........B 3/4 9 i l
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LA SALLE - UNIT 2 XV I
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' CONTAINMENT SYSTEMS.
'3
.i SURVEILLANCE REQUIREMENTS (Continued)-
I b.
At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following-painting, fire, or chemical release in any ventilation zone e
communicating with the subsystem by:
talD4WC2 ;
.sw 1.
Verifying that the subsystem satisfies the in p1 tin op (gg hg acceptance criteri nd uses the test procedures ory -
l Positions C.5.a,
..c and C.S.d of Regulatory G e 1.52, a
0.05 %
Revision 2, Ma 1978, and the system flow rate is 4000 cfm i 10%.
i Pmm d
2.
Verifying within 31 daLys after removal that a laboratory analysis of a representative carbon sample obtained in accordance with n
Regulatory Position C.6.b of Regulatory Guide 1.52, *-"if on ? - &
March 1978.seetuthelaboratorytestinacri"eriaoffRegu"atoryl
( Position C.6.a o" Regulatory Guide 1.52 RevTsion z, March 1978.
n 3.
Verifying a subsystem flow rate of 4000 cfm + 10% during system.
operation when tested in accordance with ANST N510-1975.
i c.
After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal.adsorber operation by verifying.
INSERT within.31 days after removal. that a laboratory analysis of a g-representative carbon sample obtained in accordance with Regulatory 9
1 i.
t Position C.6.b of Regulatory Guide 1.52, L-"hian 2. March 1978.
A
_ meets the laboratory testing criteria off tegulatory Position C.6 a]
3
{ oT Negulatory Gulae 1.52, Kevision z, narch 1978, f d.
At'least once per 18 months by:
1.
Verifying that the pre'ssure drop'across the combined HEPA filters and charcoal adsorber banks is less than or equal to 8-inches I
water gauge while operating the filter train at a flow rate of 4000 cfm i 10%.
2.
Verifying that the filter train starts and isolation dampers -
open on each of the following test signals:
a.
Reactor Building exhaust plenum radiation - high, b.
Drywell pressure - high,-
c.
Reactor vessel water level - low low, level 2, and I
d.
Fuel pool vent exhaust radiation - high.'.
3.
Verifying that the heaters dissipate 23
- 2.0 kW when tested in
~
I accordance with ANSI N510-1975. This reading shall include the-
~
appropriate correction for variations.from 480 volts at the bus.
i.
i LA SALLE - UNIT 2 3/4 6-44 Amendment No. 9
"o.
6 3
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
(
After each complete or partial replacement of a HEPA filter bank by e.
[-
verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm i 10%,
f.
After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove. greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while-operating the system at a flow rate of 4000 cfm i 10L 4
I V
i 4,uc, #ra msenT B
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LA SALLE - UNIT 2 3/4 6-45
f 3
PLANT SYSTEMS 3/4.7.2 CONTROL ROOMi UXILIARY ELECTRIC EQUIPMENT ROOM (MERGENCY FILTRATION SYSTEM q
LIMITING CONDITION FOR OPERATION 3.7.2 TwoindependentcontrolroomQndauxiliaryelectricequipmentroom
(
emergency filtration system trains'shall be OPERABLE."
APPLICABILITY:
All OPERATIONAL CONDITIONS and *.
l.
ACTION:
With one emergency filtration system train inoperable, restore' the a.
i inoperable train to OPERABLE status within 7 days or:
1.
In OPERATIONAL CONDITIONS 1, 2, 3, be in at least HOT SHUTDOWN-within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following-24 hours.
2.
In OPERATIONAL CONDITION 4, 5 or *, initiate and maintain operation of the OPERABLE emergency filtration system in the pressurization mode-of operation.
b.
With both emergency filtration system trains. inoperable, in OPERATIONAL CONDITION 4,.5 or *, suspend CORE ALTERATIONS, handling
+
of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
The provisions of Specification 3.0.3 are not applicable. in-c.
Operational Condition *.
SURVEILLANCE REQUIREMENTS 4.7.2 Each control room (and auxiliarv electric eouhment roo gency-filtration system train snail be demonstrated OPERAB.t:
At least once per 31 days on a STAGGERED TEST BASIS by initiating, a.
from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the train operates for at least.
10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> with the heaters OPERABLE.
t
- When irradiated fuel is being handled in the secondary containment.
- The normal or emergency power source may be inoperable in OPERATIONAL CONDITION 4, 5 or *.
LA SALLE - UNIT 2 3/4 7-4
4
.s i
PLANT SYSTEMS
{
$URVEILLANCE REQUIREMENTS (Continued)
O.
(
b.
'At least once per 18 month r (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following i
painting, fire or chemical release in any ventilation zone communicating with the train by:
Verifying'that the train satisfies t n place testing g hSI M N acceptance criteri and uses the procedures of Regulatory
\\,0To g y ppl l Positions C.S.a.
..c and C.5.d julatory Guide 1.52,.
I Revision 2, 1978, and the te is 400 t 105.
Ag gpger(1,-
'O W
1 2.
Verify ng within 31 days'after removal that a laborato analys s of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory n
Guide 1.52, Revistor 2, March 1978, meets the laboratnev A
tantino criteria o Regulatory rosition c.6.a of Regula' tory _)
Q u'de 1.52 Rev s n 2, March 1978. j 3.
Verifying a train flow rate of 4000 cfm + 10% 'during subsystem operation who tested in accordance with ANSI'N510-1975.
I 4
c.
After every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of charcoal adsorber operation by verifying i }MRT within 31 days after removal.that a laboratory analysis of a
~
representative carbon sample obtained in accordance with Regulatory O
Positon C.6.b of Regulatory Guide 1.52, Revir on 2. March 1978.
A meets the laboratory testing criteria offRegu' atory Position C.6.a r g
vCpf Keguiswry uume 4.x, neymon z, riarch 1978.J d.
At least once per 18 months by:
1.
--Verifying that the pressure drop across the combined HEPA -
filters and charcoal adsorber banks is less than 8 inches Water Gauge while operating _the train at a flow rate of 4000 cfm-1 10%.
l l
- This surveillance shall include the recirculating charcoal filter, " odor eater,"
in the normal control room supply filter train using ANSI N510-1975 as a guide to verify >, 70% efficiency in removing freon test gas.
I
- Except that recirculating charcoal filter samples shall be removed and i
{
analyzed at least once per 18 months.
I j
l LA SALLE - UNIT 2 3/4 7-5 Amendment No.- 42 3
c.
i o.
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.
Verifying that on each of the below pressurization mode actua-I tion test signals, the emergency train automatically switches to the pressurtration mode of operation and the control room is l
i.
maintained at a positiva pressure of 1/8 inch W.G. relative to the adjacent areas during emergency train operation at a flow i
t rate less than or equal to 4000 cfe:
a)
Outside air smoke detection, and b)
Air intake radiation monitors.
3.
Verifying that the heaters dissipate 2012.0 Kw when tested in accordance with AN$1 N510-1975. This reading shall include the appropriate correction for variations from 480 volts at the bus.
l After each complete or partial replacement of a HEPA filter bank by
[
verifying that the HEPA filter banks remove greater than or equal to t
99% of the 00P when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm 1 105.
I f.
Aftegeach complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove 99% of a
(
halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the
(
system at a flow rate of 4000 cfm i 10L REptAcg wtrH W5ERT D
l I
t e
This surveillance shall include the recirculating charcoal filter " odor en r,"
in the normal control room supply filter train using ANSI N510-1975 as a guide to verify ?, 70E efficiency in removing freon test gas.
(
i LA SALLE - UNIT 2.
3/4 7-6 Amendment No. 42 i
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3/4.7 PLANT SYSTEMS BASES j
3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS The OPERABILITY of the core standby cooling system - equipment cooling water systems and the ultimate heat sink ensure that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.
The redundant cooling capacity of these systems, i
assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.
p
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3/4.7.2 CONTROLR00MkDAUXILIARYELECTRICEQUIPMENTR00DMMERGENCY FILTRATION SYSTEM L
The OPERABILITY of the control roomCd nuviliarv alectr_ic equipment room emergency filtration system ensures that tie roogw111 remain naoueuse ror -
operations personnel during and following all design basis accident conditions.
The OPERABILITY of this system in conjunction with room design provisions based on limiting the radiation exposure to personnel occupying the roo l
s o
l 5 rem or less whole body, or its equivalent.. This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",10 CFR Part 50.
I Cumulative operation of the system with the heaters OPERABLE for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> over a 31 day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA. filters.
3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM The reactor core isolation cooling (RCIC) system is provided to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the Emergency Core Cooling System equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig even though the LPCI mode of the the residual heat removal (RHR) system provides adequate core cooling up to 350 psig.
The RCIC system specifications are applicable during OPERATIONAL CONDITIONS 1, 2 and 3 when reactor vessel pressure exceeds 150 psig because l
RCIC is the primary non-ECCS source of core cooling when the reactor is pressurized.
i With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCS system and justifies the specified 14 day out-of-service period.
The surveillance requirements provide adequate assurance that RCICS will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation during reactor operation, a i
complete functional test requires. reactor shutdown.
Initial startup test program data may be used to determine equivalent turbine / pump capabilities between test flow path and the vessel injection flow path.
The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment.
LA SALLE - UNIT 2 B 3/4 7-1
--.- ~ - -.
9 0
INftR T_A ASTM D 3803-79, for a methyl iodide penetration of less than 0.5% when tested at a temperature of 30*C and a relative humidity of 70%.
INSRT_B e.
Af ter each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks have a DOP penetration of less than 0.05% when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 CFR A 10%.
f.
Af tcr each complete or partial replacen.ent of a charcoal adsorption bank by verifying that the charcoal adsorption beds have a halogenated hydrocarbon refrigerant test gas penetration of less than 0.05% when they are tested in-place in accordance with ANSI N510-1975 while operating the syste:a at a flow rate of 4000 CFR A 10%.
INSRT_C ASIN D 3803-79, for a methyl iodide penetration of less than 10.0% when tested at a temperature of 30'c and a relative humidity of 70%.
.INSRTJ After each complete or partial replacement of a HEPA filter bank by e.
verifying that the HEPA filter banks have a DOP penetration of less than 1.0% when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 CFR t 10%.
f.
After each complete or partial replacement of a charcoal adsorption bank by verifying that the charcoal adsorption beds have a halogenated hydrocarbon refrigerant test gas penetration of less than 1.0% when they are tested in-place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 CFR 1 10%.
0513Tt9
ATIACRUNLC t
$1CNIFICANT HA' LARDS CONS.lDEEAll0B i
l t
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 hasards I
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 j
accident previously evaluated because The proposed amendments to the testing criteria for the VG and VC charcoal adsorption beds revises the acceptance criteria for methyl lodide penetra-j tion and specifies the laboratory testing method. The proposed Technical Specification penetration acceptance criterion is less conservative than the existing values. This reduction in conservatism will be addressed by l
station procedures which will increase the testing frequency as the l'
penetration values increase (See Table 1 attached).
The laboratory testing method thet the station currently uses is ASTM D 3803-79.
This test is not referenced in the applicable regulatory documents to which the t
station is committed (Regulatory Guide 1.52, Revision 2 and ANSI l
N510-1975). However, this test is referenced in ANSI N510-1980 and is the current industry standard.
+
L References to the " control room and auxiliary electric equipment room emergency filtration system" are being revised to the " control room I
emergency filtration system because the emergency filter system is part of the VC system. This revision is not removing any equipment only clarifying the nomenclature of an existing piece of equipment.
Analysis has shown that the control room recirculating charcoal filter is not required to maintain the control room environment under accident conditions. Therefore, removal of the references to this filter from the Technical Specifications will not increase the probability or consequences of a previously evaluated event.
2.
Create the possibility of a new or different kind of accident from any accident previously evaluated because l
The VG and VC systems are intended to mitigate the consequences of an accident and cannot, by themselves, initiate an accident. No new accidents are postulated to occur as a result of this proposal.
P t
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4 4
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- 3) Involve a significant reduction in the margin of safety because J
The proposed amendment explicitly states the acceptance criteria used for VG and VC system HEPA and charcoal bed in-place testing.
These values were derived from the referenced document and do not represent a change to the Technical Specifications. Addition of the acceptance criteria values will help to ensure that the requirements of the Technical Specifications and all governing documents are met, therefore, the margin of saf ety is unaffected.
The relaxation of the laboratory analysis acceptance criteria will i
decrease the margin of safety slightly.
However, since.the assumed analytical values for the charcoal bed efficiencies will still result in a dose well below the limits established in 10 CFR 50, Appendix A, GDC 19, t
the decrease in the margin of safety will be offset in part by the establishment of proceduralized controls.
These controls will establish acceptance criterion (action levels) conservative to the proposed Technical Specification values. While the relaxed Technical Specification acceptance criteria will allow more operational flexibility, the action levels establish requirements for increased testing frequencies and actions which will help ensure that the Technical Specification and the analytical limits are not exceeded.
The removal of references to the AEER emergency flitration system from the Technical Specifications will not reduce the margin of safety because this is only an editorial change.
The control room recirculating charcoal filter is not addressed in the Technical Specification bases and has been shnwn by analysis to be unnecessary to ensure control room habitability under accident conditions, therefore, removal of references to this equipment from the Technical Specifications does not reduce the margin of safety.
l
(
Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations". Final Rule, 51 FR 774, for the application i
of standards to license change requests for determination of the existence of significant hazards considerations.
This proposed amendment most closely l
resembles Example 1.C.2.e.vi of the examples which do not involve a significant hazards consideration, "a change which either may r(sult in some increases to the probability or consequences of a previously analyzed accident or may reduce in some 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 the Standard Review Plan, e.g., a change resulting from the application of a small refinement of a previously used calculational model er design method. This proposed amendment does not involve a signifcant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations.
Therefore, based on the guidance provided in the Federal Register and the criteria established in 100FR50.92(e), the proposed change does not l
constitute a significant hazards consideration.
0513T 8
ATIACREDtI_D REEERENCES a.
10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criteria 19 - Control Room.
b.
USAEC-xDT Standard RDT M 16-1T, October 1973, " Gas-Phase Absorbents for Trapping Radioactive Iodine and Iodine Compounds."
c.
ANSI /ASME N510-1975, " Testing of Nuclear Air-Cleaning Systems".
d.
ANSI ASME N50f 1976, " Nuclear Power Air Cleaning Units and Components".
e.
NRC Regulatory Guide 1.52 Revision 2, March 1978, " Design. Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants".
f.
ASTM D 3803-79, " Standard Test Methods for Radio-iodine Testing of Nuclear-Grade Gas-Phase Absorbents".
g.
ANSI /ASME N510-1980 (Revision of N510-1975), " Testing of Nuclear Air-Cleaning Systems".
h.
Generic Letter 83-13, dated March 2, 1983, " Clarification of Surveillance l
Requirements for llEPA Filters and Charcoal Absorber Units in Standard Technical Specifications on ESF Cleanup Systems".
1.
NRC Information Notice 87-32, dated July 10, 1987, " Deficiencies in the Testing of Nuclear-Grade Activated Charcoal".
J.
Dr. J. Benton, S&L HVAC Project Engineer, letter to M.L. Reed dated June 21, 1989, " Commonwealth Edison Company LaSalle County Station, Unita 1 and 2 - System Code VC, VE, VG and VQ, WIN 1060".
k.
G.J. Diederich letter to Director of Nuclear Reactor Regulation, dated June 26, 1989, transmitting Licensee Event Report #89-019-00, Docket No.
50-373, " Charcoal Laboratory Sample Results Out-of-Tolerance".
1.
L.R. Gregor, NRC Chief Reactor Programs Branch, letter to Mr. Cordell Reed, Commonwealth Edison dated June 30, 1989, transmitting Inspection Report 50-373/89014 (DRSS); 50-374/89014 (DRSS).
0513T 10
4 m.
T.J. Kovach letter to A.B. Davis, NRC Regional Administrator, dated July 31, 1989 "LaSalle Counth Station Units 1 and 2, Response to Inspection Report Nos. 50-373/89014 and 50-374/89014.
n.
M.L. Reed letter to G.J. Diederich dated, October 16, 1989 "LaSalle County Station Units 1 and 2."
1 I
o-,
Report SL-7232 i
REFERENCE. n.
09-28-89 Page 1 C_.
It4TRODUCT1011 A review of the La Salle County Station (LSCS) Technical Specifications, Test Procedures, and Updated Final Safety Analysis Report ~(UFSAR) has identified the need to revise these documents for internal consistency.
Radiological analyses, which are based on some of the identified parameters being changed in the Technical Specification and<UFSAR, would also require revision.
This report summarizes the principal assumptions used to recalculate the design-basis post accident radiological doses in the plant's main control room and at the offsite points of. interest, namely, the Exclusion Area Boundary.
(EAB) and Low Population Zone (LPZ). The resulting doses are then compared to the licensing acceptance criteria of 10CFRSO (Appendix A, Criterion 19) and 10 CFR100.
s.
RADIOLOGICAL DOSE tt0DElli1G The calculation of postaccident radiological-doses requires a number of separate steps, each of which is governed by plant or site parameters. Each will be discussed separately.
Briefly, the dose assessment process consists of 1.
Identifying the design-basis accident and core release of radionuclides (that is, identifying the " Source Term").
2.
Quantifying the removal of radionuclides in the primary containment by mechanisms such as plate-out, spray removal, and suppression pool scrubbing.
3.
Quantifying the release from the primary containment.
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Quantifying the movement (and removal) of radioactivity in the secondary containment -(reactor building) and the _ release to the-environment.
5.
Estimating atmospheric dispersion to the sites of interest.
- 6.
For control rooms, model the movement (and removal) of radioactivity-
.by the control room ventilation system.
7.
For'offsite locations, estimate local concentrations.
8.
Given the radionuclide> concentrations, calculate the resulting doses 1
over the time periods of interest
-9.
Compare the resulting doses with appropriate' regulatory acceptance-p4 criteria.
1 s PRINCIPAL ASSUllPTIONS:
~
Source Term j
1 The design-basis accident. selected for this assessment is the' loss-of--
~
coolant accident.(LOCA) quantified in Regulatory. Guide 1.3 (Reference'1)'.- The radionuclide releases given in this reference aro used:in the subsequent dose assessments.
-Removal of Radionuclides Within Containment Standard Review Plan 6.5.5 (Reference 2) permits removal.ofc90% of the 4
elemental and particulate forms of iodine-by scrubbing in.BWR llark 11 design-suppression pools.
Thiscredit(a" decontamination-factor,"DFof10)1st taken in the present assessment.
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Release of Radioactivity from Primary Containuent i
Two principle' pathways for release of radioactivity from the primary
~
containment have been identified and will be addressed. Others will be discussed and disposed of.
Leakage from Primary _ Containment.
Radioactivity in the primary containment is~.
presumed to leak through multiple pathways to the reactor building (secondary; containment) at the primary containment Technical Specification; leak: rate of.
0.6357 per day (Reference 3). This leak rate is assumed to persist throughout the course of the accident.
Leakage through flSIV's. Radioactivity is also presumed.to leak from primary containment through the main steam isolation valves (11SIV).
The Technical:
Specification leak rate for the itS!V's is a total of 100 scfh for 4 lines (or l
25 scfh per line) (Reference 4).
Following the accident scenario of the LSCS L Safety Evaluation Report (SER) prepared by the_ Nuclear Regulatory Commission (fiRC),theilSIVicakagescenariosproceedsas.follows:
One' line leaks at 25 scfh for 20 minutes prior to initiation of the llSIV leakage control system (11SlV-LCS).
Thereaf ter all four lines leak at 25 scfh each.. but this leakage I
is picked up by the itS!V-LCS and returned to the reactor building (see.
Reference 5). The t1SIV-LCS is assumed to be made operational 20 minutes after the onset of the LOCA (Reference 6).
In the present assessment, the leakage through the 11SIV's is not released'
+
unencumbered to the environment.
Creditforelementaliodineplate-outlinthe steam line is taken.
A maximum of 100 is taken for this effect.
The plate 4
out model was taken from f1VREG/CR-0009 (Reference 7) and was previously used.
in the Dresden and Quad Cities Control Room Habitability assessments.
(References 8 and 9).
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Report SL-7232 q
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Before the itSIV-LCS is started, llSIV leakage moves through long lengths (more than.100 feet) of large diameter pipe (22"' diameter)'before approaching a release point such as turbine gland-seals or the main condenser. 'At the low leak rates.(25 sefh) considered, ample opportunity exists for iodine removal. Further, iodine reaching the ~ condenser has a large, cool wet surface area on'which to plate-out.
Leakage from the turbine gland seals'may then plate out in the turbine building or be exhausted by the ventilation system.-
In short, ample opportunity for iodine removal in the t1SIV leakage pathway
~
exists. Credit was taken only for' steam line' plate-out.
Leak _ age through,,0ther pipinL ystems Byp,a,sl ng Secondary _ Containment. The i
S i
possibility of leakage through specific piping systems, other than the HSIV's-has been addressed.
These pathways and the-reason for them not to be considered viable leakage pathways are given in'the FSAR (Reference 10).
Leaka_ge_throug M D Hydraulic Lines.
General Electric and the NRC have generically identified another possible leakage pathway that may not have been' previously considered.
That pathway is via the Control-Rod Drive (CRD) hydraulic lines under certain circumstances (Reference 11).
This pathway is-being reviewed for LSCS to determine if it is significant to control room habitability.
RELEASE TO TsiE ENVIR0NitENT Radioactivity in the reactor building is exhausted through the Standby Gas Treatment System (SGTS) to a 113 meter tall stack.
Although the normal reactor building vent'ilation system maintains this building at -0.25 inch WG, this system is presumed to be non-operational:at the time of SGTS initiction. The SGTS then draws the building pressure down'to -0.25 inch WG in 5 minutes as required by Technical Specification (Reference 12).
In the 5 minutes prior to having a fully operational SGTS, radioactivity released to the reactor building is irmiediately assumed to be released, unfiltered, through the building walls at ground level.
-i R202/ PRE RSAR-GLCRH.
o c-Report SL-7232 09-28-89 Page 5-fZy The SGTS filter unit contains an 8-inch thick _ nuclear grade carbon filter
-(Reference 13) designed in accordance with Regulatory Guide 1.52 (Reference 14). When operational, it is assumed that the SGTS removes iodin'e with 99%
efficiency as shown in Table 1.
This assigned removal ef ficiency is consistent with the, Specifications of Table 2 in Regulatory Guide;1.52.
Releases from f1SIV leakage in the 20 minutes prior to operation of: the 11SIV-LCS are assuned to be unfiltered and at ' ground level. After the itSIV-LCS is operational, this leakage is exhausted by the:SGTS.
-Conservatively, no-credit is taken for the mixing or holdup of' radionuclides in the reactor building.-
ATil0SPilERIC DISPERSION Atmospheric dispersion factors for offsite' dose assessments were taken-s from Regulatory Guide 1.3 (Reference 1)'.and' were conservatively higher thang those in the UfSAR (Reference 15).-
Atmospheric dispersion factors for the*controliroom habitability assessments were recalculated from a 5 year meteorology; data file 1for the LSCS--
.y site. Atmospheric dispersion factors (X/Q) for ground level-releases were n
calculated using the methodology of the llurphy-Campe paper (
Reference:
16)
Releases from the 115 m tall stack ordinarily would _not enter the control. room-i intakes. Ilowever, for conservatism,- a fumigation condition;is assumed to
-occur bringing the stack release into the building wake, which wouldithen-be available for intake to the control. room.
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' Report SL-7232-09-28-89 Page 6-(i CONTROL R00ll DOSE: 110DELING.
In order to calculate thet radiological doses 'in the control room,,a model of its hVAC system is necessary..The LSCS.HVAC~ system is shown..in Figure 1,.
Under accident conditions the emergency makeup _fari supplies air to the control l
' room at the rate of'1500 cim.
The' control room emergency filter uni.t contains-.
a two-inch thick nuclear grade carbon filter (Reference'17) designed in accordance with Regulatory Guide 1.52 (Reference 14). This1 filter train is!
assumed to have an iodine removal efficiency of:90% (See Table _1).
This value is conservative as it is less than that allowed by Table 2:of Regulatdry Guide 1.52.
Unfiltered air inleakage at various points in'the system are also noted, on Figure 1. _These data are taken from pre-op tests-(for ductwork) and manufacturers' dats (for dampers).
No credit. is' taken for _1odine removal byi the. recirculation (" odor eater.")- filter train which contains~ a two-inch thick nuclear grade carbon bed.
-The llurphy-Campe paper (Reference 16) parameters for control room occupancy factors were'used in the assessment.
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'q o 8 Report SL-7232 09-28-89 Page 7
..C 0FFSITE DOSE 110DELING Offsite doses were calculated using the rnethodology and data of Regulatory Guide 1.3 (Reference 1).
RESULTS
~
The results of this assessment are given in Table 2.
All results are within applicable regulatory criteria.
A detailed calculation supporting these results is available.
(Reference 18)
Prepared by: h dwes..I b/bM,D' 9-28-89 I
G. P. Lahti Assistant Division Head Nuclear Safeguards and Licensing Division
.o n d..
9-28-89 Reviewed by:
t D. J. Benton HVAC Project Engineer Heating, Ventilating and Air Condition Division Approved by: 28-89
[O*A*
YO R. A. Parson Project flanager e f 62 37214 l
l ?.
Project Management & Engineering Division er Heuimnu L PA0 FESS 10NA l
LNG tjl5R 4
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t
' Report SL-7232' 09-28-89 Page'8-b REFERENCES-
- I
. I.'
US Nuc1 car' Regulatory Commission, " Assumptions Used for Evaluating ther 1
Potential Radiological Consequences of a' Loss of Coolant Accident for:
~
?-
IBoiling Water Reactors," Regulatory Guide 1.3, Revision. 2,4 June 1974'.-
4 2.
US Huclear Regulatory Commission, " Pressure Suppression Poo1 A$' A' Fission '
L Product Cieanup System," Standard Review Plant 6.5.5, Revision 0,
- December 1988.
j 3.
La Salle County Station, Technical Specification 3/4;6.1, " Primary Containment " Section 3.6.1.2.a.
I i
1 A..
La Salle County Station, Technical-Specification' 3/4.6.1, " Primary Containment," Section 3.6.1.2.b.
t 5.
US Nuclear Regulatory Commissioni" Safety Evaluation Report related to the Operation of La Salle County Station. Units 11and 2,".NUREG-0519..
llarch 1981, Subsection 15.3.2.2.
q
- 6. -
La Salle County Station, Updated Final Safety Analysis Report (UFSAR),
Subsection 6.7.1.1.f.1.
7.-
A. K. Postma, R. R. Sherry and.P. S. Tam, " Technological Bases: for.ilodels
'i of Spray Washout of Airborne Contaminants in Containment' Vessels,"
~
NUREG/CR-_0009, October 1978.
- o l
8.
-Dresden Control Room liabitability Assessment. Attachment,to s
Letter, E. D. Swartz of Commonwealth Edison To D. G. Eisenhut,-.USNRC, "Dresden Station. Units 2 and 3 Quad Cities Station, Units 1 and 2. and-1 Zion Station Units 1 and 2 Supplemental Response to NUREG-0737, item' 1
111.D.3.4.-
s l
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l-pi w o-s Report SL-7232 09-28-89 Page 9 l[D.
9.
Quad Cities Control Room Habitabilit'y Assessment. Attachment to letter, E. D. Swartz of Conaonwealth Edison To D. G. Eisenhut, USNRC, "Dresden Station, Units 2 and 3, Quad Cities Station, Units l' and 2, and Zion Station Units 1 and 2, Supplemental Response to.NUREG-0737, 1 tem-111.D.3.4.
10.
LaSalleCounty,FinjilSafetyAnalysisReport, Questions 021.11and.
021.25 and their responses.
11.
Letter, J. E. Ilorrison of General Electric to T. J. Kovach of-Connonwealth Edison, "GE PRC 89-15, CRD System Leakage During LOCA,.
July 13, 1989.
12.
LaSalle County Station,; Technical ~ Specification 3/4.6.5, " Secondary Containment "-Section 4.6.5.1.c.
t 13.
LaSalle County Station, UFSAR. -Table 6.5-1..
14.
US Huclear Regulatory Connission, " Design. Testing and itaintenance Criteria for Postaccident Engineered-Safety-Feature Atmospheric Cleanup.
System Air Filtration and Adsorptice Units of Light-Water-Cooled. Nuclear Power Plants," Regulatory Guide 1-
>2, Revision.2,.flarch 1978.
15.
LaSalle County Station, UFSAR, Table 15.6 9._
16.
K. G. 'llurphy Snd K.11. Campe, " Nuclear. Power Plant Control Room' Ventilation System Design for fleeting Genera 1' Criterion 19," 13th AEC Airt Cleaning Conference, 1974.
- 17.. La Salle-County' Station, UFSAR, Table 9.4-1.
18.
Sargent & LunJy Calculation 1-CT-2, " Control Room Doses from.Inside -
E Atmosphere After LOCA," Revisions 1 and 3, 1989.
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'o Report SL-7232 09-28-89 Page 10
(_I Table 1 Charcoal Performance Criteria (Proposed)
Standby Gas Treatnient Control Room System Filter Emergency Filter
,_,__ { 8-i nch b ed )
(2-inch bed)
Efficiency Penetration Efficiency Penetration Procedural Initial Action Level
< 99.825%
> 0.175%
< 98.0%.
> 2.0%
Technical Specification
< 99.5%
> 0.5%
< 90%
> 10.0%
LC0 Action Level Radiological Dose 99.0%
1%
90%
10%
Assessment Basis Applicable Technical 3/4.6.5 3/4.7.2 Specification (Subsections'4.6.5.3 (Subsections 4.7.2 b, c and f) b, c and f) q
~
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Report SL-7232 09-28 Page 11
- - -C.
V; Table 2 RESULTS h
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s 1.
ATl10 SPHERIC DISPER$10N FACTOR.
t i
'X/Q.
- s/m3-E Release
' Cont rol-
'EAB-LPZ
-l Room' (509 m)
(6400m) t Stack Release l
Fumigation - first 30 minutes 2.65x10'4 1.85x10'4-2.3x10-5:
}
n Elevated Release
}
30 min -
8 hr 0
-1.7x10 6.0x10-6 8 hr 24 hr 0
'1.9x10-6:
24-hr 96 hr 0-6.1x10-7
'I I
90 hr - 720 hr 0
'1.9x10-7 Ground-Level Releases
]/
Reactor Building Exfiltration.
2.65x10 16.8x10'4 3a9x10 5 flSIV-Leakage 2.65x10'4 6.8x10'4 3.9x10-5' l
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_-y{
j Table 2, continued RESULTS
't J
t, 2.
RADIOLOGICAL DOSES _IN CONTROL R00l1 Dos'es, rem 3.
JSource Thgid Whol B
l ody Skin-Primary: Cont'ainment
- 5 minute exfilthation-3.938:
0.021 0.27
- SGTS stack = release"
. 196
- 0.086 1.03, f1SIV Leakage s
- 20 minute initial release 2.161 l0.014
.0.17-e'
- SGTS stack release
. 0'4 3_
10.024
= 0.29]
I
- t
. Total 6.34 ~
0.145:
-1.76L
?
[G 10CFR50, Appendix A, GDC19 Limits 30 5
,30 -
j
~
-1{
N d
~
'{
3
-' I 1
m s
z a ;-
- R2D2/ PRE.~RSAR-GLCRH-
!I
~
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r
+
1*
g..
e e, Report SL-7232 09-28-89 Page 13 P'
Table 2, continued RESULTS 3.
OFFSITE DOSES Exclusion _ Area Boundary Low Population Zone Source Thyroid tlhole Body,,,,
Thy ro,i,d_
llhole Body Primary Containment
- 5 minute exfiltration 34.8 2.7 1.99 0.15
- SGlS stack release 0.7 2.8 0.59 0.80 ftSIV Leakage
- 20 minute initial 19.1 1.5
~1.10 0.08 release
- SGTS stack release 0.2 1.1 0.31 0.48 Total 54.8 8.1
_... 4. 0 1.5 10 CFR 100 Limits 300 25 300 25 R202/ PRE RSAR-GLCRH
Control Doom HVAC (k ntical to Train C)
- 1 wans Inleakage
,2 FIGURE 1 3
ReMgerant UmtA To seniliary l4 ognipment room ll ll 45.2 CFM ll HVAC system n
iI 01YA 02YA THer fin r c:
c s Ill2.500CFMll
- m (:
>7 ig EE vP
/
11 r
d 11 e
~;
m
==
i e
==
tmeriency gi HTG MED Charcoal malieup Fan A c:
l1.503CFMI coil EFF filter Qg r filter ll a T5 A Il Ii M0 d MO 13YA fl 7.0CFM ]
52YA Ii T
08YA 53YA T
r 11YA 12YA Supply '
ll (
(
.f
_ Q' Mo MO knA r
s T~
TJ MO MO Control g7 7v O" f_1 v
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,[
Hearstg Homedifier L
I, Nlinim8m Firer Charcoal n
coil ll 21.5 CFMl Met-m MD I
air supply filter to seniliary C#3mf cil From system B 4 jf 135 CFM ogeipment
- c o *.4 room HVAC
$ 74 system A Typical for5 zenes c wo
- t-MO 18YA y"
l 292 CFM l
- $ tn 15YA To seniliary d
building "
7
]
7 to Co U
Fan A MO To retumm T m
FanB um7m ow,
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