ML20198C404
| ML20198C404 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 11/01/1985 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20198C410 | List: |
| References | |
| NUDOCS 8511120082 | |
| Download: ML20198C404 (13) | |
Text
.
/
UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
y WASHINGTON, D. C. 20555
\\...../
COMMONWEALTH EDIS0N COMPANY DOCKET N0. 50-373 LA SALLE COUNTY STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 29 License No. NPF-11 1.
The Nuclear Regulatory Comission (the Comission or the NRC) having found that:
A.
The application for amendment filed by the Commonwealth Edison Company, dated July 15, 1985, complies with the standards and requirements of the Atomic Energy Act_ of 1954, as amended (the Act), and the Comission's regulations set forth in 10.CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.
There is reasonable assurance:
(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in conpliance with the Commission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Fart 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, th4 license is amended by changes to the Technical Specifications as ind*cated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No, NPF-11 is hereby anended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 29 and the Environmental Protection Plan contained ^ in Appendix B, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical SpecificMions and the Environmental Protection Plan, m2SBBM S$d P
?
. 4 f3.
This amendment is effective upon startup following the first refueling outage.
FOR THE NUCLEAR REGULATORY C0(11ISSION
/
Walter R. Butler, Chief
^
Licensing Branch No. 2.
Division of Licensing
Enclosure:
Changes to the Technical
-Specifications
-'Date of Issuance: 40V01 g f-1 a
i 1
5 4
i l
..m_,
i ENCLOSURE TO LICENSE AMENDMENT NO. 29 FACILITY OPERATING LICENSE NO. NPF-11 DOCKET NO. 50-373 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change.
REMOVE INSERT l
3/4 3-24 3/4 3-24 3/4 3-25 3/4 3-25 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28
-3/4 3-29 3/4 3-29 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 3/4 5-1
- 3/4 5-1 B 3/4 5-1 B 3/4 5-1 B 3/4 5-2 B 3/4 5-2 a
-rna w
--r te
9' TABLE 3.3.3-1 W
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION r-g MINIMUM OPERABLE APPLICABLE CHANNELS PER TRIP OPERATIONAL c-TRIP FUNCTION FUNCTION (a)
CONDITIONS ACTION 5
]
A.
DIVISION I TRIP SYSTEM 1.
RHR-A (LPCI MODE) &_LPCS SYSTEM a.
Reactor Vessel Water Level - Low Low Low, Level 1 2(b) 1, 2, 3, 4*, 5*
30 fD) b.
Drywell Pressure - High 2
1, 2, 3 30 c.
LPCS Pump Discharge Flow-Low (Bypass) 1 1, 2, 3, 4*, 5*
31 d.
LPCS and LPCI A Injection Valve Injection Line 1/ valve 1,2,3 32 Pressure-Low (Permissive) 4*, 5*
33 1
e.
LPCS and LPCI A Injectiors Valve Reactor Pressure-Low 2
1,2,3 38 y
(Permissive) 4*, 5*
33 f.
LPCI Pump A Start Time Delay Relay 1
1, 2, 3, 4*, 5*
32 g.
LPCI Pump A Discharge Flow-Low (Bypass) 1 1, 2, 3, 4*, 5*
31 h.
Manual Initiation 1/ division 1, 2, 3, 4*, 5*
34 2.
AUTOMATIC DEPRE'5URIZATION SYS.31 TRIP SYSTEM "A"#
a.
Reactor Vessel Water Level - Low Low Low, Level 1 2(b) 1, 2, 3 30 coincident with D.
Drywell Pressure - High 2(b) 1, 2, 3 30 c.
Initiation Timer 1
1,2,3 32 l
Ig d.
Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 32
[
e.
LPCS Pump Discharge Pressure-High (Permissive) 2 1,2,3 32 5
f.
LPCI Pump A Discharge Pressure-High (Permissive) 2 1,2,3 32
)
f g.
Manual Initiation 1/ division 1,2,3 34 m
h.
Drywell Pressure Bypass Timer 1
1,2,3 32 e
i.
Manual Inhibit 1/ division 1,2,3 34
TABLE 3.3.3-1 (Continued)
E EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION l
h MINIMUM OPERABLE APPLICABLE CHANNELS PER TRIP OPERATIONAL' E
TRIP FUNCTION FUNCTION (a)
CONDITIONS ACTION Q
B.
DIVISION 2 TRIP SYSTEM g
1.
Reactor Vessel Water Level - Low, Low Low, Level 1 2(b) 1, 2, 3, 4*, 5*
30 b.
Drywell Pressure - High 2(b)
I, 2, 3 30 c.
LPCI B and C Injection Valve Injection Line 1/ valve 2, 2, 3 32 Pressure-Low (Permissive) 4*, 5*
33 d.
LPCI Pump B Start Time Delay Relay 1
1, 2, 3, 4*, 5*
32 mD e.
LPCI Pump Discharge Flow - Low (Bypass) 1/ pump 1, 2, 3, 4*, 5*
31
[
f.
Manual Initiation 1/ division 1, 2, 3, 4*, 5*
34 g.
LPCI B and C Valve Reactor Pressure-Low (Permissive) 2 1,2,3 38 l
4*, 5*
33 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
4 a.
Reactor Vessel Water Level - Low Low Low, Level 1 2(b) 1, 2, 3 30 coincident with b.
Drywell Pressure - High 2(b) 1, 2, 3 30 c.
Initiation Timer 1
1,2,3 32 l
d.
Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 32 e.
LPCI Pump B and C Discharge Pressure - High g-(Permissive) 2/ pump 1,2,3 32 k
f.
Manual Initiation 1/ division 1, 2, 3 34 g
g.
Drywell Pressure Bypass Timer 1
1,2,3 32 h
h.
Manual InM.>it 1/ division 1, 2, 3 34
TABLE 3.3.3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
With one channel inoperable, place the inoperable channel a.
in the tripped condition within one hour
- or declare the associated system inoperable.
b.
With more than one channel inoperable, declare the associated system inoperable.
ACTION 31 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip Function, place the inoperable channel in the tripped condition within one hour; restore the inoperable channel to OPERABLE status within 7 days or declare the associated system inoperable.
ACTION 32 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per T','ip Function requirement, declare the associated ADS trip system or ECCS inoperable.
ACTION 33 -
With the number of OPERABLE channels less than the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.
ACTION 34 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or declare the associated ADS trip system or ECCS inoperable.
l ACTION 35 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement For one trip system, place that trip system in the tripped a.
condition within one hour
- or declare the HPCS system inoperable.
b.
"For both trip systems, declare the HPCS system inoperable.
ACTION 36 -
With the number of OPERABLE channels less than required by the Minimum GPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour
- or declare the HPCS system inoperable.
ACTION 37 -
With the number of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2, as appropriate.
- The provisions of Specification 3.0.4 are not applicable.
LA SALLE - UNIT 1 3/4 3-27 Amendment No. 29
TABLE 3.3.3-2 9
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SET?OINTS m
2 E
ALLOWABLE 7
TRIP FUNCTION TRIP SETPOINT VALUE E
A.
DIVISION 1 TRIP SYSTEM H
1.
RHR-A (LPCI MODE) AND LPCS SYSTEM a.
Reactor Vessel Water 1.evel - Low Low Low, Level 1
>- 129 inches *
>- 136 inches
- b.
Drywell Pressure - High 51.69psig
[1.89psig
- c.
LPCS Pump Discharge Flow-Low
> 750 gpm
> 640 gpm d.
LPCS and LPCI A Injectior Valve Injection 500 psig 500 1 20 psig Line-Low Pressure Interlock e.
LPCS and LPCI A Injection Valve Reactor 500 psig 500 1 20 psig Pressure-low Pressure Interlock i
l f.
LPCI Pump A Start Time Delay Relay
< 5 seconds
< 6 seconds i
{
g.
LPCI Pump A Discharge Flow-Low
[1000gpm E550gpm h.
Manual Initiation NA NA i
o$
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A" a.
Reactor Vessel Water Level - Low Low Low, Level 1
>- 129 inches *
>- 136 inches
- b.
Drywell Pressure - High 51.69psig 51.89psig c.
Initiation Timer
< 105 seconds
< 117 seconds l
d.
Reactor Vessel Water Level-Low, Level 3 i 12.5 inches
- i 11 inches
- e.
LPCS Pump Discharge Pressure-High
[146psig, increasing
[136psig, increasing f.
LPCI Pump A Discharge Pressure-High
> 119 psig, increasing
> 106 psig, increasing g.
Manual Initiation NA NA h.
Drywell Pressure Bypass Timer
< 9.0 minutes Footnote (a) 1.
Manual Inhibit NA NA k
E 5
(a) The sum of the time delays associated with the ADS initiation timer and the drywell pressure bypass g:
time shall be less than or equal to 687 seconds.
TABLE 3.3.3-2 (Continued) 9 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS
,:=
F ALLOWABLE 7
TRIP FUNCTION TRIP SETPOINT VALUE E
B.
DIVISION 2 TRIP SYSTEM w
1.
Reactor Vessel Water level - Low Low Low, Level 1
>- 129 inches *
>- 136 inches
- b.
Drywell Pressure - High 51.69psig 51.89psig c.
LPCI B and C Injection Valve Injection Line-Low Pressure Interlock 500 psig 500 psig 120 psig d.
LPCI Pump B Start Time Delay Relay 1 5 seconds
< 6 seconds e.
LPCI Pump Discharge Flow-Low
> 1000 gpm E550gpm f.
Manual Initiation NA NA g.
LPCI B and C Injection Valve Reactor 500 psig 500 1 20 psig
{
Pressure Low Pressure Interlock 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B" a.
Reactor Vessel Water Level - Low Low Low, Level 1
>- 129 inches *
>- 136 inches
- b.
Drywell Pressure - High 51.69psig 51.89psig c.
Initiation Timer
< 105 seconds
< 117 seconds d.
Reactor Vessel Water Level-Low, Level 3 5 12.5 inches
- 5 11 inches
- l e.
LPCI Pump B and C Discharge Pressure-High
[119psig, increasing
[106psig, increasing f.
Manual Initiation NA NA g.
Drywell Pressure Bypass Timer s 9.0 minutes Footnote (a) h.
Manual Inhibit NA NA N
(a) The sum of the time delays associated with the ADS initiation timer and the drywell pressure bypass e
h time shall be less than or equal to 687 seconds.
$e teb
TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS F
m CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED w
A. DIVISION I TRIP SYSTEM
Low Low Low, Level 1 NA M
R 1, 2, 3, 4*, 5*
b.
Drywell Pressure - High NA M
Q 1,2,3 c.
LPCS Pump Discharge Flow-Low NA M
Q 1, 2, 3, 4*, 5*
d.
LPCS and LPCI A Injection Valve Injection Line Pressure Low Interlock NA M
R 1, 2, 3, 4*, 5^
R e.
LPCS and LPCI A Injection Valve Reactor Pressure Low Inter. lock NA M
R 1, 2, 3, 4*, 5*
T f.
LPCI Pump A Start Time Delay Relay NA M
Q 1, 2, 3, 4*, 5*
U g.
LPCI Pump A Flow-Low NA M
Q 1, 2, 3, 4*, 5*
h.
Manual Initiation NA R***
NA 1, 2, 3, 4*, 5*
- 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
a.
Low Low Low, Level 1 NA M
R 1,2,3 b.
Drywell Pressure-High NA M
Q 1,2,3 c.
Initiation Timer NA M
Q 1,2,3 1
d.
Low, Level 3 NA M
R 1,2,3 e.
LPCS Pump Discharge Pressure-High NA M
Q 1,2,3 5
f.
LPCI Pump A Discharge E
Pressure-High NA M
Q 1,2,3 5
g.
Manual Initiation NA R
NA 1,2,3 5
h.
Drywell Pressure Bypass Timer NA M
Q 1,2,3 z
i.
Manual Inhibit NA R
NA 1,2,3
TABLE 4.3.3.1-1 (Continued)
EMERGENCY CORE COOLING S'tSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS F
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL' CHANNEL CONDITIONS FOR WHICH I
'E TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED w
- 8. DIVISION 2 TRIP SYSTEM
Low Low Low, Level 1 NA M
R 1, 2, 3, 4*, 5*
b.
Drywell Pressure - High NA M
Q 1,2,3 c.
LPCI B and C Injection Valve Injection Line Pressure Low Interlock NA M
R 1, 2, 3, 4*, 5*
d.
LPCI Pump B Start Time Delay Relay NA M
Q 1, 2, 3, 4*, 5*
R e.
LPCI Pump Discharge Flow-Low NA M
Q 1, 2, 3, 4*, 5*
f.
Manual Initiation NA R***
NA 1, 2, 3, 4*, 5*
Y g.
LPCI B and C Injection Valve O
Reactor Pressure Low Interlock NA M
R 1, 2, 3, 4*, 5*
- 2. AiliOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
a.
Low Low Low, Level 1 NA M
R 1,2,3 b.
Drywell Pressure-High NA M
Q 1,2,3 c.
Initiation Timer NA M
Q 1,2,3 l
d.
Low, Level 3 NA M
R 1,2,3 e.
LPCS Pump B and C Discharge g
Pressure-High NA M
Q 1,2,3 e
f.
Manual Initiation NA R
NA 1,2,3 E
h.
Drywell Pressure Bypass Timer NA M
Q 1,2,3 l
i.
Manual Inhibit NA R
NA 1,2,3 e
E
_m
F 3/4.5 EMERGENCY' CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING LIMITING CONDITION FOR CPERATION 3.5.1 ECCS divisions 1, 2 and 3 shall be OPERABLE with:
a.
ECCS division 1 consisting of:
1.
The OPERABLE low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.
2.
The OPERABLE low pressure coolant injection (LPCI) subsystem "A" of the RHR system with a flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
3.
At least 6 OPERABLE ** ADS valves.
l b.
ECCS division 2 consisting of:
1.
The OPERABLE low pressure coolant injection (LPCI) subsystems "B" and "C" of the RHR system, each with a flow path capable of taking suction from the suppression chamber and transferring the water to the reactor vessel.
2.
At least 6 OPERABLE ADS valves.
i ECCS division 3 consisting of the OPERABLE high pressure core spray c.
(HPCS) system with a flow path capable of taking suction from the suppression chamber and transferring the water through the spray sparger to the reactor vessel.
APPLICABILITY:
OPERATIONAL CONDITION 1, 2*# and 3*.
4
- The ADS is not required to be OPERABLE when reactor steam dome pressure is less than or equal to 122 psig.
- See Specification 3.3.3 for trip system operability.
l See Special Test Exception 3.10.6.
LA SALLE - UNIT 1 3/4 5-1 Amendment No. 29
I s
3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHUTDOWN ECCS Division 1 consists of the low pressure core spray system, low pressure coolant injection subsystem "A" of the RHR system, and the automatic depressurization system (ADS) as actuated by ADS trip system "A".
ECCS Divisica 2 consists of low pressure coolant injection subsystems "B"and "C" of the RHR system and the automatic depressurization system as actuated by ADS trip system "B".
The low pressure core spray (LPCS) s'ystem is provided to assure that the core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for transients or smaller l
breaks following depressurization by the ADS.
The LPCS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.
The surveillance requirements provide adequate assurance that the LPCS system will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
The low pressure coolant injection (LPCI) mode of the RHR system is pro-vided to assure that the core is adequately cooled following a loss-of-coolant accident. Three subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor recirculation line break, and for transients or small breaks following depressurization by l
the ADS.
The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage to pip,ing and to start cooling at the earliest moment.
ECCS Division 3 consists of the high pressure core spray system.
The high pressure core spray (HPCS) system is provided to assure that the reactor core is adequately cooled to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not
- result in rapid depressurization of the reactor vessel.
The HPCS system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.
The HPCS system operates over a range of 1160 psid, differential pressure between reactor vessel and HPCS suction source, to O psid.
The capacity of the HPCS system is selected to provide the required core cooling.
The HPCS pump is designed to deliver greater than or equal to 516/1550/6200 gpm at differential pressures of 1160/1130/200 psid.
Initially, water from the condensate storage tank is used instead of injecting water from LA SALLE - UNIT 1 B 3/4 5-1 Amendment No. 29
r.
EMERGENCY CORE COOLING SYSTEMS BASES ECCS-OPERATING and SHUT 00WN (Continued) the suppression pool into the reactor, but no credit is taken in the hazards analyses for the condensate storage tank water.
With the HPCS system inoperable, adequate core cooling is assured by the OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems.
In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the hazards analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition.
The HPCS out-of-service period of 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems.
The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required.
Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel injection equires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.
Upon failure of the HPCS system to function properly, if required, the l
automatic depressurization system (ADS) automatically causes selected safety-t relief valves to open, depressurizing the reactor so that flow from the low presure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200*F.
ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 122 psig even though low pressure core cooling systems provide adequate core cooling up to 350 psig.
ADS automatically controls seven selected safety-relief valves.
Six valves are required to be OPERABLE although the hazards analysis only takes credit for five valves.
It is therefore appropriate to permit one of the required valves to be out of-service for up to 14 days without materially reducing system reliability.
3/4.5.3 SUPPRESSION CHAMBER l
The suppression chamber is also required to be OPERABLE as part of the l
ECCS to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA.
This limit on suppression chamber minimum wate' volume ensures that sufficient water is available to permit r
recirculation cooling flow to the core.
The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.2.1.
Repair work might require making the suppression chamber inoperable.
This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5.
In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required water volume is reduced because the reactor coolant is maintained at or below 200*F.
Since pressure suppression is not required below 212 F, the minimum j
watar volume is based on NPSH, recirculation volume, vortex prevention plus a 2'-A" safety margin for conservatism.
l LA SALLE - UNIT 1 B 3/4 5-2 Amendment No. 29 l
l i
L l