ML20249A729

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Amends 127 & 112 to Licenses NPF-11 & NPF-18,respectively, Relocating Certain Requirements Re Fire Protection from TS to Updated Final Safety Analysis Rept
ML20249A729
Person / Time
Site: LaSalle  
Issue date: 06/10/1998
From: Skay D
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20249A730 List:
References
NUDOCS 9806180153
Download: ML20249A729 (33)
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Text

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UNITED STATES p

NUCLEAR REGULATORY COMMISSION O

WASHINGTON, D.C. 20066 4001 COMMONWEALTH EDISON COMPANY

, DOCKET NO. 50-373 LASALLE COUNTY STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.127 License No. NPF-11 1.

The Nuclear Regulatory Commission (the Commission) has found that:

l l

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee), dated July 15,1996, as supplemented on June 19,1997, and l

February 2,1998, complies with the standards and requirements of the Atomic l

Energy Act of 1954, as amended (the Act), and the Commission's regulations set i

forth in 10 CFR Chapter I-1 B.

The facility will operate in conformity with the application, the provisions of the Act, and the reguls.tions 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 paragraphs 2.C.(2) and 2.C.(25) of the Facility Operating License No. NPF-11* are hereby amended to read as follows:

i I

' License page 9 is provided, for convenience, for the composite license to reflect this change.

9806180153 980610 PDR ADOCK 05000373 P

PDR

2-(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.127, 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.

(25)

Fire Protection Pronram The licer.see shallimplement and maintain all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for LaSalle County Station, and as approved in NUREG-0519, " Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2," dated March 1981; Supplement 2 dated February 1982; Supplement 3 dated April 1982; Supplement 5 dated August 1983; Supplement 7 dated December 1983; Supplement 8 dated March 1984; and SERs for the following:

LaSalle Unit 1 License Amendment 1, dated June 18,1982; LaSalle Unit 1 License Amendment 18, dated August 8,1984; LaSalle Unit 1 License Amendment 23, dated May 22,1985; LaSalle Unit 1 License Amendment 44, dated June 20,1986; LaSalle Unit 1 License Amendment 127, dated June 10,1998 ; and j

NRC Evaluation of the Consequences of Postulated Failures of 1 Hour Fire Rated i

Darmatt KM-1 Fire Barrier under Seismic Loading at LaSalle County Station, dated March 29,1996.

The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION

@ b-f Donna M. Skay, Project Maryger Project Directorate lll-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation Attachments:

1.

License page 9 2.

Changes to the Technical Specifications Date ofissuance: June 10,1998 i

1 9-

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(25) Fire Protection Procram The licensee shallimplement and maintain all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for LaSalle l

County Station, and as approved in NUREG-0519, " Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2," dated March 1981; Supplement 2 dated February 1982; Supplement 3 dated April 1982; Supplement 5 dated August 1983; Supplement 7 dated December 1983; Supplement 8 dated March l

1984; and SERs for the following:

l LaSalle Unit 1 License Amendment 1, dated June 18,1982; LaSalle Unit 1 License Amendment 18 dated August 8,1984; l

LaSalle Unit 1 License Amendment 23, dated May 22,1985;

(

LaSalle Unit 1 Ucense Amendment 44, dated June 20,1986; L

LaSalle_ Unit 1 License Amendment 127, dated June 10, 1998

and NRC Evaluation of the Consequences of Postulated Failures of 1 Hour Fire Rated Darmatt KM-1 Fire Barrier under Seismic Loading at LaSalle County Station, dated March 29,1996.

l The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability

- to achieve and maintain safe shutdown in the event of a fire.

I I

i Amendment No. 127 e-

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l ATTACHMENT TO LICENSE AMENDMENT NO.127 FACILIW OPERATING LICENSE NO. NPF-11 DOCKET NO. 50-373 4

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.

l REMOVE INSERT V

V Vill Vill Xill Xill XV XV XXil XXll XXill XXIll 3/4 3-73 3/4 3-73 3/4 3-75 3/4 3-76 3/4 3-77 3/4 3-78 3/4 3-79 3/4 3-80 3/4 3-81 3/4 7-11 3/4 7-11 3/4 7-12 3/4 7-13 3/4 7-14 3/4 7-15 3/4 7-16 3/4 7-17 3/4 7-18 3/4 7-19 3/4 7-20 3/4 7-21 3/4 7-22 3/4 7-23 B 3/4 SS B 3/4 S S B 3/4 7-2 B 3/4 7-2 B 3/4 7-3 8 3/4 7-3 S2 62 I

INDZX LfMfTINo CONDITIONE FOR OPERATION AND SURVEILLANCE REOUTREMENTS SECTION EAEK 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................

3/4 3-1 3/4.3.1.

ISOLATION ACTUATION INSTRUMENTATION......................

3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..........................................

3/4 3-23 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation......

3/4 3-35 End-of-Cycle Recirculation Pump Trip System Instrumentation........................................

3/4 3-39 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION........................................

3/4 3-45 3/4.3.6 CONTROL ROD WITEDRAWAL BLOCK INSTRUMENTATION.............

3/4 3-50 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.....................

3/4 3-56 Meteorological Monitoring Instrumentation................

3/4 3-63 Remote Shutdown Monitoring Instrumentation...............

3/4 3-66 Accident Monitoring Instrumentation......................

3/4 3-69 Source Range Monitors....................................

3/4 3-72 Deleted..................................................

3/4 3-73 Deleted..................................................

3/4 3-74 Deleted..................................................

3/4 3-81 Explosive Gas Monitoring Instrumentation.................

3/4 3-82 Loose-Part Detection System..............................

3/4 3-85 3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION I N S TRUENTAT I ON........................................

3/4 3-86 LA SALLE - UNIT 1 V

Amendment No. 127

s INDEX LTMYTfNC CONDITIONE FOR OPERATION AND SURVEfttANCE REOUTREMENTE SECTION E&fsE 3/4.7 Pf_ ANT EYSTEME 3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS Residual Heat Removal Servica 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 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM..........................

3/4 7-4 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM..................

3/4 7-7 3/4.7.4 SEALED SOURCE CONTAMINATION............................

3/4 7-9 l

3/4.7.5 DELETED................................................

3/4 7-11 3/4.7.6 DELETED................................................

3/4 7-11 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 3/4.7.9 SNUBBERS...............................................

3/4 7-27 3/4.7.10 MAIN TURBINE BYPASS SYSTEM.............................

3/4 7-33 4

LA SALLE - UNIT 1 VIII Amendment No.127 l

INDEX RAEFE SECTION R&GE INSTRUMENTATION (Continued)

MONITORING INSTRUMENTATION (Continued)

Meteorological Monitoring Instrumentation.........

B 3/4 3-4a Remote Shutdown Monitoring Instrumentation........

B 3/4 3-4a Accident Monitoring Instrumentation...............

B 3/4 3-5 source Range Monitors.............................

B 3/4 3-5 Deleted...........................................

B 3/4 3-5 Explosive Gas Monitoring Instrumentation..........

B 3/4 3-6 Loose-Part Detection System.......................

B 3/4 3-6 3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION.................................

B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM..............................

B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES..............................

B 3/4 4-2 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.........................

B 3/4 4-2 Operational Leakage...............................

B 3/4 4-2 3/4.4.4 CHEMISTRY.........................................

B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY.................................

B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS.......................

B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES..................

B 3/4 4-5 3/4.4.8 STRUCTURAL INTEGRITY..............................

B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REMOVAL.............................

B 3/4 4-5 LA SALLE - UNIT 1 XIII Amendment No. 127

IMDEX Ramen BECTION

}&gg 3/4.7 PLANT BYSTEMS 3/4.7.1 CORE STANDBY COOLING SYSTEM - E WATER SYSTEMS..................Q.UIPMENT COOLING B 3/4 7-1 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM..................

B 3/4 7-1 3/4.7.3 REACTOR 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 DELETED...........................................

B 3/4 7-2 3/4.7.6 DELETED...........................................

B 3/4 7-3 3/4.7.7 AREA TEMPERATURE MONITORING.......................

B 3,*4 7-3 3/4.7.8 8'cRUCTURAL INTEGRITY OF CLASS I STRUCTURES........

B 3/4 7-3 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 BYSTEMS 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 8-3 3/4.9 REFtfELTMC OPERA'rIONE 3/4.9.1 REACTOR MODE SWITCH...............................

B 3/4 9-1 3/4.9.2 INSTRUMENTATION...................................

B 3/4 9-1 3/4.9.3 CONTROL ROD POSITION..............................

B 3/4 9-1 l

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 HOIST...................................

B 3/4 9-1 3/4.9.7 CRANE TRAVEL......................................

B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE POOL.........

B 3/4 9-2 3/4.9.10 CONTROL ROD REMOVAL...............................

B 3/4 9-2 3/4.9.11 RESIDUAL HEAT RENOVAL COOLANT CIRCULATION.........

B 3/4 9-2 l

LA SALLE - UNIT 1 XV Amendment No.

127

e M

LTET OF TABtFM (Continued)

M EhGK 4.3.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................

3/4 3-65 3.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION..............

3/4 3-67 4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................

3/4 3-68 I

3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION.....................

3/4 3-70 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION l

SURVEILLANCE REQUIREMENTS...............................

3/4 3-71 l

I 3.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION................

3/4 3-83 4.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................

3/4 3-84 3.3.8-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION...............................

3/4 3-87 3.3.8-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS.....................

3/4 3-88 4.3.8.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............

3/4 3-89 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES..................................................

3/4 4-9 l

3.4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS.................

3/4 4-12 4.4.5-l' PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM........................................

3/4 4-15 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE.....................................

3/4 4-19 l

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LA SALLE - UNIT 1 XXII Amendment No.127 u____________-.._

^

M LTET OF TARTRE fCentinued1 TABLE R&g g 3.6.5.2-1 SECONDARY CONTAINMENT VENr1LATION SYSTEM Au - ATIC IsotATION DAxPERS.............................

3/4 6-3, I

3.2.2-1 m A m, m TORE oNI m INo.............................

3/4 2-2.

4.8.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS......................

3/4 8-18 3.8.3.3-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION..............................................

3/4 8-22 33/4.4.6-1 REACTOR VESSEL TOUGHNESS................................

B 3/4 4-6 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS....................

5-6 l

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LA SALLE - UNIT 1 XXIII Amendment No.

127 l

a

1 3/4.3.7.7,3/4.3.7.8,3/4.3.7.9 AND 3/4.3.7.10 INTENTIONALLY LEFT BLANK PAGES 3/4 3-74 THRU 3/4 3-81 ARE DELETED l

I LASALLE - UNIT 1 3/4373 Amendment No.127 NEXT PAGE IS 3/4 3-82 l

i.

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1 3/4.7.5 AND 3/4.7.6 INTENTIONALLY LEFT BLANK PAGES 3/4 7-12 THROUGH 3/4 7-23 ARE DELETED I

. LA SALLE - UNIT 1 3/4 7-11 Amendment No.127 NEXT PAGE IS 3/4 7-24

INSTRUMENTATION BASEE MONITORING INSTRUMENTATION (Continued)

- 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables Endg an accident. This capability is consistent with the recommendations of Regulatory

. Guide 1.97 " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Leamed Task Force Status Report and Short-Term Recommendations."

3/4.3.7.6 SOURCE RANGE MONITORS

.. The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown. At these power l

levels, reactivity additions should not be made without this flux level information available to the l

operator. When the intermediate range monitors are on scale adequate information is available without the SRMs and they can be retracted.

3/4.3.7.7 DELETED 3/4.3.7,8 DELETED

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3/4 3.7 g DELETED j

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l LA SALLE - UNIT 1 B 3/4 3 5 Amendment No. 127 i

PLANT SYSTEMS BASES 3/4.7.4 SEALED SOURCE CONTAMINATION The limitations, on removable contamination for sources requiring leak testing, including alpha emmitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure

. that leakage from byproduct, source, and special nuclear' material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use, with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring or boron measuring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4,7.5 DELETED LA SALLE - UNIT 1 B 3/4 7-2 Amendment No. 127

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PLANT SYSTEMS BASES 3/4.7.6 DELETED 3/4.7.7 AREA TEMPERATURE MONITORING The area temperature ! imitations ensure that safety'-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY. The temperature limits include allowance for an instrument error of i 7'F.

3/4.7.8 STRUCTURAL INTEGRITY OF CLASS 1 STRUCTURES in order to assure that settlement does not exceed predicted and allowable settlement values, a program has been established to conduct a survey at the site. The allowable total differentini settlement values are based on original settlement predictions. In establishing these tabulated values, an assumption is made that pipe and conduit connection have been designed to safely withstand the stresses which would develop due to total and differential settlement.

3/4 7.9 SNUBRERS All snubbers are required OPERABLE to ensure that the structuralintegrity of the Reactor Coolant System and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety-related system.

Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip,10-kip, and 100-kip capacity manufactured by Company "A" are of the same type. The surne design mechanical snubbers manufactured by Company "B" for the purpose of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.

A list of individual snubbers with detailed information of snubbers location and size and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. The accessibility of each l

l LA SALLE - UNIT 1 B 3/4 7-3 Amendment No.127

ADMINISTRATIVE CONTROLS '

1.

At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor in addition, while the reactor is in OPERATIONAL CONDITION 1,2 or 3, at 1

least one licensed Senior Reactor Operator who has been designated by the Shift Supervisor to assume the control room direction responsibility shall be in the Control Room.

2.

A radiation protection technician

  • shall be on site when fuel is in the reactor.

3.

All CORE ALTERATIONS shall be observed and directly supervised by either a i

licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel i

Handling who has no other concurrent responsibilities during this operation.

4.

DELETED i

5.

The Independent safety Engineering Group (ISEG) shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports an1 other sources of plant design and operating experience information, including pants of similar design, which may indicate areas for improving unit safety.

The ISEG aall be composed of at least three, dedicated, full-time engineers of multi-disciplines located on site and shall be augmen;od on a part-time basis by personnel l

from other parts of the Commonwealth Edison Company organization to provide i

expertise not represented in the group. The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification # that these activities are performed correctly and that human errors are reduced as much as practical. The i

ISEG shall make detailed recommendations for revised procedures, equipment i

modifications, maintenance activities, operations activities or other means of j

improving unit safety to the Site Quality Verification Director and the Station Manager.

1 6.

The Station Control Room Engineer (SCRE) may serve as the Shift Technical Advisor (STA) during abnormal operating and accident conditions. During these conditions, the SCRE or other on duty STA shall provide advisory technical support to the Shift Supervisorin the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

i

  • The radiation protection technician position may be less than the minimum requirement for a l

period of time not to exceed two hours in order to accommodate unexpected absence l

provided immediate action is taken to fill the required position.

  1. Not responsible for sign-off feature.

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LA SALLE - UNIT 1 6-2 Amendment No. 127 l

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1 UNITED STATES g

j NUCLEAR REGULATORY COMMISSION r

wAswiwoTow, p.c. sonas.cooi COMMONWEALTH EDISON COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.112 License No. NPF-18 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment Gled by the Commonwealth Edison Company (the licensee), dated July 15,1996, as supplemented on June 19,1997, and February 2,1998, 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; i

B.

The fccility will operate in cor.formity 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 Commisalon's regulations and al; applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes the Technical Specifications as indicated in the enclosure to this license amendment and paragraphs 2.C.(2) and 2.C.(15) of the Facility Operating License No. NPF-18' are hereby amended to read as follows:

l

  • License pages 6 and 7 are provided, for convenience, for the composite license to reflect this change.

I I

b-

a (2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, s.n revised through Amendment No.112, and the Environmental Protection F'an contained in Appendix B, are hereby incorporated in the license. The teensee shall operate the facility in accordance with the Technical Specifications ad the Environmental Protection Plan.

(15)

Fire Pietection Prooram The licensee shall implement and maintain all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for LaSalle County Station, and as approved in NUREG-0519, " Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2," dated March 1981; Supplement 2 dated February 1982; Supplement 3 dated April 1982; i

supplement 5 dated August 1983; Supplement 7 dated December 1983; Supplement 8 dated March 1984; and SERs for the following:

LaSalle Unit 2 License Amendment 11, dated May 22,1985; LaSalle Unit 2 License Amendment 14, dated October 2,1985; LaSalle Unit 2 License Amendment 112, dated June 10,1998; and NRC Evaluation of the Consequences of Postulated Failures of 1 Hour Fire Rated Darmatt KM-1 Fire Barrier under Seismic Loading at LaSalle County Station, dated March 29,1996.

The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the eve,nt of a fire.

3.

Sections E and F (pages 2 and 3) of Attachment 1 to this license are deleted.

4.

This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION pd Donna M. Skay, Project Mans'ger Project Directorate ill-2 Division of Reactor Projects Ill/IV Office of Nuclear Reactor Regulation Attachments:

1. License pages 6 and 7
2. Changes ta the Technical Specifications Date of issuance: June 10,1998

)

. 1 l-(1) battery current (ammeter-charge / discharge), (2) battery charger output voltage (voltmeter), (3) battery charger output current (ammeter), (4) battery high discharge rate alarm, and (5) battery charger trouble alarm. In the interim, the licensee shall implement approved procedures to monitor battery current, battery charger output voltage, and battery charger output current at the local panels at least once per eight-hour shift.

(14)

Control of Heaw Loads (Section 9.1. SSER #1. SER #5)

Prior to startup after the first refueling, the licensee shall have made commitments acceptable to the NRC regarding the guidelines of Sechon 5.1.2 through 5.1.6 of NUREG-0612.

(15)

Fire Protection Proaram I

The licensee shallimplement and maintain all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for LaSalle County Station, and as approved in NUREG-0519, " Safety Evaluation Report related to the operation of LaSalle County Station, Units 1 and 2," dated March 1981; Supplement 2 dated February 1982; Supplement 3 dated April 1982; i

Supplement 5 dated August 1983; Supplement 7 dated December 1983; I

Supplement 8 dated March 1984; and SERs for the following:

)

LaSalle Vnit 2 License Amendment 11, dated May 22,1985; LaSalle Unit 2 License Amendment 14, dated October 2,1985:

LaSalle Unit 2 License Amendment 112, dated June 10, 1998

and NRC Evaluation of the Consequences of Postulated Failures of 1 Hour Fire l

Rated Darmatt KM-1 Fire Barrier under Seismic Loading at LaSalle County Station, dated March 29,1996.

The licensee may make changes to the approved Fire Pro' 7 tion Program without prior approval of the Commiscion only if those changes would not i

adversely affect the ability to achieve and maintain safe shutdown in the event l-of a fire.

l I

Amendment No. 112

. (16)

Industrial Security (Section 13.6. SER. SSER #3. SSER #5)

CECO shall fully implement and maintain in effect all provisions of the Commission approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards information protected under 10 CFR 73.21, are entitled 'LaSalle County Nuclear Station Security Plan," with revisions submitted through June 1,1988, "LaSalle County Nuclear Power Station Security Personnel Training and Qualification Plan," with revisions submitted through June 13,1986, and "LaSalle County Nuclear Power Station Safeguards Contingency Plan," with revisions submitted through February 16,1984. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

Amendment No.112

ATTACHMENT TO LICENSE AMENDMENT NO.112 FACILITY OPERATING 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 vertical line indicating the area of change.

REMOVE INSERT V

V Vill Vill Xill XIll XV XV XXil XXil XXill XXIll 3/4 3-73 3/4 3-73 3/4 3-75 3/4 3-76 3/4 3-77 3/4 3-78 3/4 3-79 3/4 3-80 3/4 3-81 3/4.7-11 3/4 7-11 3/4 7-12 3/4 7-13 3/4 7-14 3/4 7-15 3/4 7-16 3/4 7-17 3/4 7-18 3/4 7-19 3/4 7-20 3/4 7-21 3/4 7-22 3/4 7-23 3/4 7-24 B 3/4 3-5 B 3/4 3-5 8 3/4 7-2 8 3/4 7-2 B 3/4 7-3 8 3/4 7-3 S2

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O INDEX t IMITL3 CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION RAgg 3/4 3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION......................

3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION......................................

3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTR UM ENTATION............................................................................... 3/4323 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.

ATWS Recirculation Pump Trip System Instrumentation......................... 3/4 3 35 End-of-Cycle Recirculation Pump Trip System Instrumentation............... 3/4 3-39 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTR U M E NTATION............................................................................. 3/4 3 4 5 3/4.3.6 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION...............

3/4 3-50 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.......................................................

3/4 3-57 Meteorological Monitoring instrumentation.............................................. 3/4 3-63 Remote Shutdown Monitoring Instrumentation.........................................

3/4 3-66 Accident Monitoring Instrumentation..................................................... 3/4 3 69 Source Range Monitors........................................................................... 3/4 3-72 Delet ed................................................................................................... 3/4 3 Deleted................................................................................................ 3/4 3-74 i

Deleted.................................................................................................... 3/4 3-81 Explosive Gas Monitoring instrumentation............................................. 3/4 3-82 Loose-Part Detection System................................................................ 3/4 3-85 3/4.3.8 FEEDWATER/ MAIN TURBlNE TRIP SYSTEM ACTUATION INSTR UM E NTATION............................................................................... 3/4 3-86 LA SALLE - UNIT 2 V

Amendment No.112

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 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 Ultim ate H e at Sink............................................................................

3/4 7-3 3/4.7.2 CONTROL ROOM AND AUXlLIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM..........................................

3/4 7-4 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM..............

3/4 7-7 3/4.7.4 SEALED SOURCE CONTAMINATION...

3/4 7-9 3/4.7.5 DELETED...............

3/4 7-11 3/4.7.6 DE LET E D..........................,...

3/4 7-11 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 S N U B B E R S.........

3/4 7-28 3/4.7.10 MAIN TURBINE BYPASS SYSTEM.............................................

3/4 7-34 LA SALLE - UNIT 2 Vill Amendment No.112

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l lNDEX BASES SECTION PAGE INSTRUMENTATION (Continued)

MONITORING INSTRUMENTATION (Continued)

Meteorological Monitoring Instrumentation................................

B 3/4 3-4a

]

Remote Shutdown Monitoring instrumentation..........................

B 3/4 3-4a Accident Monitoring Instrumentation.........................................

B 3/4 3-5 l

Source Rang e Monitors...........................................................

B 3/4 3-5 Deleted.............................................................................

B 3/4 3-5

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i Explosive Gas Monitoring Instrumentation...............................

B 3/4 3-6 Loo se-Part Detection System...............................................

B 3/4 3-6

-3/4.3.8 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION l

l N ST RUMENTATION........................................................... B 3/4 3-6 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECI RCULATION SYSTEM..................................................

B 3/4 4-1 I

I 3/4.4.2 SAFETY / RELIEF VALVES...................................................

B 3/4 4-1a

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3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...................................................

B 3/4 4-2 j

l Ope rational Le a ka ge...............................................................

B 3/4 4-2 3/4.4.4 CHEMISTRY..........................................................................

B 3/4 4-2

]

3/4.4.5 S P ECI FIC ACTIVITY...............................................................B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS......................................

B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES...............................

B 3/4 4-5 3/4.4.8 STR UCTU RAL I NTEG RITY.....................................................

B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REMOVAL...............................................

B 3/4 4-5 i

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LA SALLE - UNIT 2 Xill Amendment No.112

INDEX BASES 3.ECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 CORE STANDBY COOLING SYSTEM - EQUIPMENT COOLING WATER SYSTEMS.............................................

B 3/4 7-1 3/4.7.2 CONTROL ROOM AND AUXILIARY ELECTRIC EQUIPMENT ROOM EMERGENCY FILTRATION SYSTEM.......................

B 3/4 7-1 3/4.7.3 REACTOR 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 D5LETED.................................................................

B 3/4 7-2 3/4.7.6 DELETED......

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 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 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 R EACTOR MODE SWITCH.............................................

B 3/4 9-1 3/4.9.2 INSTRUMENTATION................

B 3/4 9-1 3/4.9.3 CONTROL ROD POSITION......................................

B 3/4 9-1 3/4.9.4 D E CAY TI M E...................................................................

B 3/4 9-1 3/4.9.5 COM M U N I CATI O N S............................................................

B 3/4 9-1 3/4.9.6 C RA N E AN D HOI ST.........................................................

B 3/4 9-1 3/4.9.7 C RAN E TRAVE L..................................................................

B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE POOL.........

B 3/4 9-2 3/4.9.10 CONTROL ROD R EMOVAL..................................................

B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL COOLANT CIRCULATION.....

B 3/4 9-2 LA SALLE - UNIT 2 XV Amendment No.112

INDEX LIST OF TABLES (Continued)

TABLE PAGE 3.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION..............

3/4 3-67 4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SU RVEI LLANCE REQ UIREMENTS...................................................

3/4 3-68 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION..............................

3/4 3-70 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQ UIREMENTS...........................................

3/4 3-71 3.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION.............

3/4 3-83 4.3.7.11-1 EXPLOSlVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..

3/4 3-84 3.3.8-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION...............................................

3/4 3-87 3.3.8-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SETPOINTS..............

3/4 3-88 i

4.3.8.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............

3/4 3-89 3.4.3.2-1' REACTOR COOT. ANT SYSTEM PRESSURE ISOLATION VALVES....

3/4 4-10 3.4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS........................

3/4 4-13 l

l 4.4.51 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALY SI S P ROG RAM.............................................................

3/4 4-16 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITH DRAWAL S C H E DU LE..........................................................

3/4 4-20 LA SALLE - UNIT 2 XXil Amendment No.112

e INDEX LIST OF TABLES (Continued)

TABLE PAGE 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS...........................................

3/4 6-42 3.7.7-1 AREA TEMPERATURE MONITORING..............................................

3/4 7-26 4.7.9-1 SNUBBER VISUAL INSPECTION INTERVAL......................................

3/4 7-33a 4.8.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS...................................

3/4 8-18 3.8.3.3-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD P R OT E CTI O N........................................................................

3/4 8-27 B3/4.4.6-1 REACTOR VESS EL TOUG HNESS............................................

B 3/4 4-6 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS........................

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LA SALLE - UNIT 2 XXill Amendment No.112 1

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LA SALLE - UNIT 2 3/4 3-73 Amendment No.112 NEXT PAGE IS 3/4 3-82

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l LA SALLE UNIT 2 3/4 7-11 AMENDMENT NO.112 NEXT PAGE IS 3/4 7-25 i

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INSTRUMENTATIC J.

BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, " instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "TMl-2 Lessons Leamed Task Force Status Report and Short-Term Recommendations."

3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown. At these power levels, reactivity additions should not be made without this flux levelinformation available to the I

operator. When the intermediate range monitors are on scale adequate information is available without the SRMs and they can be retracted.

3/4.3.7.7 DELETED 3/4.3.7.8 DELETED 3/4.3.7.9 DELETED 3/4.3.7.10 DELETED LA SALLE-UNIT 2 B 3/4 3-5 Amendment No.112

e PLANT SYSTEMS BASES 3/4.7.4 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emmitters, is based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values. Sealed sources are classified into three groups according to their use,

{

with surveillance requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which era continuously enclosed within a shielded mechanism, i.e., sealed sources within radiation monitoring or boron measuring devices, are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

3/4.7.5 DELETED 1

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LA SALLE - UNIT 2 B 3/4 7-2 Amendment No.112

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PLANT SYSTEMS BASES 3/4,7.6 DELETED 3/4.7.7 AREA TEMPERATURE MONITORING The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause loss of its OPERABILITY. The temperature limits include allowance for an instrument error ofi 7'F.

3/4.7.8 STRUCTURAL INTEGRITY OF CLASS 1 STRUCTURES in order to assure that settlement does not exceed predicted and allowable settlement values, a program has been established to conduct a survey at the site. The allowable total di#erential settlement values are based on original settlement predictions. In establishing these tabulated values, an assumption is made that pipe and conduit connection have been designed to safely withstand the stresses which would develop due to total and di#erential settlement.

3/4.7.9 SNUBBERS All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant Syste n and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse e#ect on any safety-related system.

Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip,10-kip, and 100-kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purpose of this Technical Specification would be of a di#erent type, as would hydraulic snubbers from either manufacturer.

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LA SALLE - UNIT 2 B 3/4 7-3 Amendment No.112

ADMINISTRATIVE CONTROLS

\\

1.

At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor. In addition, while the reactor is in OPERATIONAL CONDITION 1,2 or 3, at least one licensed Senior Reactor Operator who has been designated by the Shift Supervisor to assume the control room direction responsibility shall be in the Control Room.

2.

A radiation protection technician

  • snall be on site when fuel is in the reactor.

3.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

4.

DELETED 5.

The Independent Safety Engineering Group (ISEG) shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event Reports and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improving unit safety.

The ISEG shall be composed of at least three, dedicated, full-time engineers of multi-disciplines located on site and shall be augmented on a part-time basis by personnel j

from other parts of the Commonwealth Edison Company organization to provide expertise not represented in the group. The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification # that these activities are performed correctly and that human errors are reduced as much as practical. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities or other means of improving unit safety to the Site Quality Verification Director and the Station Manager.

6.

The Station Control Room Engineer (SCRE) may serve as the Shift Technical Advisor (STA) during abnormal operating and accident conditions. During these conditions, the SCRE or other on duty STA shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

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  • - The radiation protection technician position may be less than the minimum requirement for a period of time not to exceed two hours in order to accommodate unexpected absence provided immediate action is taken to fill the required position, j
  1. Not responsible for sign-off feature.

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I LA SALLE - UNIT 2 6-2 Amendment No.112

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