ML20216C679
| ML20216C679 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/13/1998 |
| From: | Skay D NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216C684 | List: |
| References | |
| NUDOCS 9805190386 | |
| Download: ML20216C679 (18) | |
Text
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t UNITED STATES j
,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30666-0001
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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-373 LASALLE COUNTY STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 126 License No. NPF-11 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by tne Commonwealth Edison Company (the licensee), dated September 26,1997, as supplemented on April 7,1998, and May 1,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter 1, i
B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; 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 compliance with the Commission's regulations set forth in 10 CFR Chapter ll 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 Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:
9005190386 980513 PDR ADOCK 05000373 P
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. (2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.126
. 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 Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and shall be implemented prior to restart from L1F35.
FOR THE NUCLEAR REGULATORY COMMISSION hk Donna M. Skay, Project Manager Project Directorate lil-2 Division of Reactor Projects -Ill/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: May 13, 1998 i
i
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ATTACHMENT TO LICENSE AMENDMENT NO.
126 FACILITY OPERATING LICENSE NO. NPF-11 DOCKET NO. 50-373 Replace the following pages of the Appendix "A" Tec 'nical Specifications with the enclosed pages. The revised pages are identified by amendmer.' nu mber and contain a vertical line indicating the area of change.
FEMOVE INSERT 3/4 7-4 3/4 7-4 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6 B 3/4 7-1 B 3/4 7-1 6-20a 6-20a 6-20b 6-20b
l PLANT SYSTEMS 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION 3.7.2 Two independent control room and auxiliary electric equipment room emergency filtration system trains shall be OPERABLE.8 APPLICABILITY: All OPERATIONAL CONDITIONS and *.
ACTION:
a.
With one emergency filtration system train inoperable, restore the 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 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 *, 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 fuelin the secondary containment and operations with a potentin.1 for draining the reactor vessel.
c.
The provisions of Specification 3.0.3 are not applicable in Operational Condition *.
SURVEILLANCE REQUIREMENTS 4.7.2 Each control room and auxiliary electric equipment room emergency filtration system train shall be demonstrated OPERABLE:
a.
At least once per 31 days on a STAGGERED TEST BASIS:
1.
Operate each Control Room and Auxiliary Electric Equipment Room Emergency Filter System for greater than or equal to 10 continuous hours with the heaters operating, and 2.
Manually initiating flow through the control room and auxiliary electric equipment room recirculation filters for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
- 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 *.
l LA SALLE - UNIT 1 3/4 7-4 Amendment No. 126
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i PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.
Perform required control room and auxiliary electric equipment room filter testing in accordance with, and at the frequency specified by, the Ventilation Filter Testing Program.
c.
Deleted.
d.
At least once per 18 rnonths by-1.
Deleted.
LA SALLE - UNIT 1 3/4 7-5 Amendment No. 126 i
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.
Verifying that on each of the below pressurization mode actuation test signals, the emergency train automatically switches to the pressurization mooc of operation. Manually initiate flow through the control room and auxiliary electric l
equipment room recirculation filters line and then verify that the control room j
and auxiliary electric equipment rooms are maintained at a positive pressure of greater than or equal to 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:
l l
a)
Outside air smoke detection, and l
b)
Air intake radiation monitors.
3.
Deleted.
1 l
e.
Deleted.
1 f.
Deleted.
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l LA SALLE - UNIT 1 3/4 7-6 Amendment No. 126
t.
3/4.7 PLANT SYSTEMS BASES 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, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.
3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM The OPERABILITY of the control room and auxiliary electric equipment room emergency filtration system, which includes the control room and auxiliary electric equipment room recirculation filters, ensures that the rooms will remain habitable for operations personnel during and following all design basis accident conditions. The OPERABILITY of this system in conjunction with room design provisions is based on limiting the radiation exposure to personnel occupying the rooms to 5 rem or less whole body, or its equivalent. This limitation is consistent I
with the requirements of General Design Criteria 19 of Appendix "A",10 CFR Part 50.
Continuous operation of the system with the heaters operating 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 l
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 RCIC is the primary non-ECCS source of core cooling when the reactor is pressurized.
With the RCIC system inoperable, adequate core cooling is assured by the OPERABILITY of the HPCS system and justifies the specified I4 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. The low pressure setpoint allowable value for the discharge line " keep-filled" alarm is based on the head of water between the centerline of the pump discharge and the system high point vent.
LA SALLE - UNIT 1 B 3/4 7-1 Amendment No.126
ADMINISTRATIVE CONTROLS PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) 7.
Primary Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Testing Program," dated September 1995.
The peak calculated primary containment intemal pressure for the design basis loss of coolant accident, P., is 39.6 psig.
j The maximum allowable primary containment leakage rate, L., at P, is 0.635% of primary containment air weight per day.
Leakage rate acceptance criteria are:
a.
Primary containment overallleakage rate acceptance criterion is s1.0 L,. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are s 0.60 L, for the combined Type B and Type C tests, and s 0.75 L, for Type A tests, b.
Air lock testing acceptance criteria are:
1)
Overall air lock leakage rate is so.05 L, when tested at 2 P.
2)
For each door, the seal leakage rate is s 5 scf per hour when the gap between the door seals is pressurized to 210 psig.
The provisions of specification 4.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.
The provisions of specification 4.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.
8.
Ventilation Filter Testing Program (VFTP)
A program shall be astablished to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2, dated March 1978, and in accordance with ASME N510-1989.
The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the VFTP test frequencies.
s.
Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass
<0.05% when tested in accordance with ASME N510-1989, at the system flowrate specified below:
ESF Ventilation Flowrate (cfm)
System SBGT System 2 3600 and s 4400 CREF System 2 3600 and c 4400 LA SALLE - UNIT 1 6-20a Amendment No. 126
ADMINISTRATIVE CONTROLS PLANT OPERATING PROCEDURES AND PROGRAMS (Continuid) b.
Demonstrate for each of the ESF system filter units that an inplace test of the charcoal adsortier shows a penetration and system bypass less than the value
)
specified below, when tested in accordance with ASME N510-1989, at the system flowrate specified below:
ESF Ventilation Penetration and Flowrate (cfm)
System System Bypass SBGT System 0.05 %
2 3600 and s 4400 CREF System 0.05 %
2 3600 and s 4400 CRRF System 2.0 %
218000 and s 28900 AEERRF System 2.0 %
214000 and s 22800 c.
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30'C, a relative humidity of 70 % and a face velocity as specified below.
i ESF Ventilation Penetration Face System Velocity (fpm)
SBGT System 0.5 %
40 CREF System 2.5 %
40 i
CRRF System 15.0 %
80 AEERRF System 15.0 %
80 d.
Demonstrate for each of the ESF systems that the pressure drop across the combined l
moisture separator, heater, prefilter, HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below:
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ESF Ventilation Delta P Flowrate (cfm)
System (inches wg)
SBGT System 8
2 3600 and s 4400 CREF System 8
2 3600 and s 4400 CRRF System 3.0 218000 and s 28900 AEERRF System 3.0 214000 and s 22800 e.
Demonstrate that the heaters for each of the ESF systems dissipate the electrical l
power specified below when tested in accordance with ASME N510-1989. These readings shallinclude appropriate corrections for variations from 480 Volts at the bus.
ESF Ventilation Wattage (kw)
System SBGT System 2 21 and s 25 l
CREF System 218 and s 22 l
6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE EVENT IN PLANT OPERATION l
The following actions shall be taken for REPORTABLE EVENTS:
a.
The Commission shall be notified and a Licensee Event Report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b.
Each REPORTABLE EVENT shall be reviewed by the Onsite Review and investigative Function.
LA SALLE - UNIT 1 6-20b Amendment No.126
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UNITED STATES g
,g NUCLEAR REGULATORY COMMISSION
'f WASHINGTON, D.C. 20066-0001
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i I
COMMONWEALTH EDISON COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION. UNIT 2 l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.111 License No. NPF-18 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment filed by the Commonwealth Edison Company (the licensee), dated September 26,1997, as supplemented on April 7,1998, and May 1,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 Commission; 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 compliance 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 Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to read as follows:
l
l
. i (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
111, 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 Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and shall be implemented prior to restart of LaSalle, Unit 2, from the current outage.
FOR i HE NUCLEAR REGULATORY COMMISSION l
fi W
j Donna M. Skay, Project Manager Project Directorate 111-2 Division of Reactor Projects - til/IV l
Office of Nuclear Reactor Regulation l
Attachment:
Changes to the Technical Specifications Date of issuance: May 13, 1998 l
l l
L
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ATTACHMENT TO LICENSE AMENDMENT NO. 111 FACILITY ODERATING LICENSE NO. NPF-18 DOCKET NO. 50 374 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 verticalline indicating the area of change.
REMOVE INSERT 3/4 7-4 3/4 7-4 3/4 7-5 3/4 7-5 3/4 7-6 3/4 7-6 B 3/4 7-1 B 3/4 7-1 6-20b 6-20b 6-2Cc O
l PLANT SYSTEMS 3L(J.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION i
3.7.2 Two independent control room and auxiliary electric equipment room emergency filtration system trains shall be OPERABLE.'
APPLICABILITY: All OPERATIONAL CONDITIONS and *.
ACTION j
a.
With one emergency filtration system train inoperable, reatore the inoperable train to OPERABLE status within 7 days or.
l 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 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 *, 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 l
CONDITION 4,5 or *, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor l
vessel.
c.
The provisions of Specification 3.0.3 are not applicable in Operational Condition *.
SURVEILLANCE REQUIREMENTS 4.7.2 Each control room and auxiliary electric equipment room emergency filtration system train shall be demonstrated OPERABLE:
s.
At least once per 31 days on a STAGGERED TEST BASIS:
1.
Operate each Control Room and Auxiliary Electric Equipment Room Emergency Filter System for greater than or equal to 10 continuous hours l
with the heaters operating, and l
2.
Manually initiating flow through the control room and auxiliary electric equipment room recirculation filters for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
- When irradiated fuel is being handled in the secondary containment.
l
- The normal or emergency power source may be inoperable in OPERATIONAL CONDITION 4,5 or *.
LA SALLE - UNIT 2 3/4 7-4 AMENDMENT NO.111 l
PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.
Perform required control room and auxiliary electric equipment room filter testing in accordance with, and at the frequency specified by, the Ventilation Filter Testing Program.
c.
- Deleted, d.
At least once per 18 months by:
1.
Deleted.
LA SALLE - UNIT 2 3/4 7-5 AMENDMENT NO.111
i PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2.
Verifying that on each of the below pressurization mode actuation test signals, the emergency train automatically switches to the pressurization mode of operation. Manually initiate flow through the control room and auxiliary electric equipment room recirculation filters and then verify that the control room and auxiliary electric equipment rooms are maintained at a positive pressure of greater than or equal to 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 b)
Air intake radiation monitors.
3.
Deleted.
e.
Deleted.
f.
Deleted.
l LA SALLE - UNIT 2 3/4 7-6 AMENDMENT NO.111
3/4.7 PLANT SYSTEMS BASES 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, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits.
' 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM The OPERABILITY of the control room and auxiliary electric equipment room emergency filtration system, which includes the control room and auxiliary electric equipment room recirculation filters, ensures that the rooms will remain habitable for operations personnel during and following all design basis accident conditions. The OPERABILITY of this system in conjunction with room design provisions is based on limiting the radiation exposure to personnel occupying the rooms to 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.
Continuous operation of the system with the heaters operating 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 I
equipment. The RCIC system is conservatively required to be OPERABLE whenever reactor pressure exceeds 150 psig even though the LPCI mode of the residual heat removal (RHR) l system provides adequate core cooling up to 350 psig.
l The RCIC system specifications are applicable during OPERATIONAL 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 pressurized.
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 sunteillance 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 vtssel injection flow path. The pump discharge piping is maintained full to prevent water hammer damage and to start cooling at the earliest possible moment. The low pressure setpoint allowable value for the discharge line " keep-filled" alarm is based on the head of water between the centerline of the pump discharge and the system high point vent.
l l
LA SALLE - UNIT 2 B 3/4 7-1 Amendment No. 111
ADMINISTRATIVE CONTROLS PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) a.
Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.05 %
when tested in accordance with ASME N510-1989, at the system flowrate specified below:
ESF Ventilation Flowrate (cfm)
System SBGT System a 3600 and s 4400 CREF System 2 3600 and s 4400 b.
Demonstrate for each of the ESF system filter units that an inplace test of the charcoal adsorber shows a penetration and system bypass less than the value specified below, when tested in accordance with ASME N510-1989, at the system flowrate specified below:
ESF Ventilation Penetration and Flowrate (cfm)
System System Bypass SBGT System 0.05 %
2 3600 and s 4400 CREF System 0.05 %
2 3600 and s 4400 CRRF System 2.0 %
218000 and s 28900 AEERRF System 2.0 %
214000 and s 22800 c.
Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyliodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30 C, a relative humidity of 70 % and a face velocity as specified below.
ESF Ventilation Penetration Face Velocity (fpm)
System SBGT System 0.5 %
40 CREF System 2.5 %
40 CRRF System 15.0 %
80 AEERRF System 15.0 %
80 d.
Demonstrate for each of the ESF systems that the pressure drop across the combined moisture separator, heater, prefilter, HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below:
ESF Ventilation Delta P Flowrate (cfm)
System (inches wg)
SBGT System 8
2 3600 and s 4400 CREF System 8
2 3600 and s 4400 CRRF System 3.0 218000 and s 28900 AEERRF System 3.0 214000 and s 22800 LA SALLE - UNIT 2 6-20b Amendment No. 111
l ADMINISTRATIVE CONTROLS PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) e.
Demonstrate that the heaters for each of the ESF systems dissipate the electrical power specified below when tested in accordance with ASME N5101989. These readings shall include appropriate corrections for variations from 480 Volts at the bus.
ESF Ventilation Wattage (kw) l System SBGT System 2 21 and s 25 CREF System 218 and s 22 l
l 6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE EVENT IN PLANT OPERATION The following actions shall be taken for REPORTABLE EVENTS:
The Commission shall be notified and a Licensee Event Report submitted pursuant to the a.
requirements of Section 50.73 to 10 CFR Part 50, and b.
Each REPORTABLE EVENT shall be reviewed by the Onsite Review and Investigative Function.
l LA SALLE - UNIT 2 6-20c Amendment No.111 i