Text

Amends 102 & 87 to Licenses NPF-11 & NPF-18,respectively, Restructuring Primary Containment & Primary Containment Leakage TS to Reduce Repetition of Requirements Contained in NRC Regulations Such as App J to 10CFR50
	ML20081D655
Person / Time
Site:	LaSalle
Issue date:	03/16/1995
From:	William Reckley; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20081D657	List: ML20081D655; ML20081D667;
References
	GL-91-08, GL-91-8, GL-93-05, GL-93-5, NUDOCS 9503200351
	Download: ML20081D655 (70)
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4t UNITED STATES

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,j NUCLEAR REGULATORY COMMISSION -

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-373 LASALLE COUNTY STATION. UNIT 1 6MENDMENT TO FACILITY OPERATING LICENSE Amendment No.'102 License No. NPF-11 1.

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

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee), dated October 24, 1994, 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 paragraphs 2.C.(2) and 2.D* of Facility Operating License No. NPF-11 are hereby amended to read as follows:

Page 16b is attached, for convenience, for the composite license to reflect this change.

9503200351 950316 PDR ADOCK 05000373 p

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2.C.(2).

Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 102, 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.

2.D.

The facility requires exemptions from certain requirements of 10 CFR Part 50, 10 CFR Part 70, and 10 CFR Part 73. These include:

(a)

Exemptions from certain requirements of Appendices G, H and J and 10 CFR Part 73 are described in the Safety Evaluation Report and Supplement No. 1, No. 2 and No. 3 to the Safety Evaluation Report.

(b)

An exemption was requested until the completion of the first refueling from the requirements of 10 CFR 70.24.

(c)

An exemption from 10 CFR Part 50, Appendix E from performing a full scale exercise within one year before issuance of an operating license, both exemptions (b) and (c) are described in Supplement No. 2 of the Safety Evaluation Report.

(d)

An exemption was requested from the requirements of 10 CFR 50.44 until either the required 100 percent rated thermal power trip startup test has been completed or the reactor has operated for 120 effective full power days as specified by the Technical Specifications. Exemption (d) is described in the safety evaluation of License Amendment No. 12.

(e)

An exemption from the requirement of paragraph III.D of Appendix J to conduct the third Type A test of each ten-year service period when the plant is shutdown for the 10-year plant inservice inspections.

Exemption (e) is described in the safety evaluation accompany:ng Amendment No.

to this license. These exemptions are authorized by law an:1 will not endanger life or property or the common defense and security and are otherwise in the public interest.

Therefore, these exemptions are hereby granted. The

.i facility will operate, to the extent authorized herein, in conformity with the application, as amended, and the rules and regulations of the Commission (except as hereinafter exempted therefrom), and the provisions of the Act.

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This anendment is effective upon date of issuance and shall be implemented within 30 days.

i FOR THE NUCLEAR REGULATORY COMMISSION D

William D. Reckley, Pr Manager r

Project Directorate III-2 l

Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Attachments:

1.

License page 16b-2.

Changes to the Technical Specifications-Date of Issuance: March 16. 1995 l

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- 16b -

D.

The' facility requires exemptions from certain requirements of 10 CFR F at 50, 10 CFR Part 70, and 10 CFR Part 73.

These include:

(a)

Exemptions from ceP ain requirements of Appendices G H and J and 10 CFR Part 73 ce described in the Safety Evaluation Report and Supplement No. 1, No. 2 and No. 3 to tle Safety Evaluation Report.

(b)-

An exemption was requested until the completion of the first l

refueling from the requirements of 10 CFR 70.24.

(c)

An exempti n from LO CFR ) art 50, Appendix E from performing I

a full sea e exerc se witiin one year before issuance of an icense both exemptions i

in Supplement No.,2 of the Safety Ev(b? and (c) are described operating a.uation Report.

Anexempt'onwasrekherequired100percentratedthermauested from the req (d) 50.44 unt'l either power trip startup test has been completed or the reactor s as specified has operated for 120 effect<ve full power day;d) is described l

by the Technical Spec' ficat' ons.

Exemption in the safety evaluat on of License Amendmen; No. 12.

(e)

An exemption from the requirement of paragraph III.D of Appendix J to conduct the third Type A test of each ten-year service period when the plant is shutdown for the 10-year plant inservice inspections.

Exemption is described in the safety evaluation accompanying Amendm(e)t No.

en to this license.

These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

l these exemptions are hereby aranted.

The Therefore,ill operate, to the extent authorized herein,les facility w in conformity with the application, as amended, and the ru exempted therefrom), and the provis(except as hereinafter and regulations of the Commission ions of the Act.

E.

This license is subject to the following additional condition for the protection of the environment:

Before engaging in additional construction or operational activities which may result in a sianificant adverse environmental impact that was not evaluated or that is significantly greater than that evaluated in the Final Environmental Statement and its Addendum the licensee shall provide a written notification to the Director of the Office of Nuclear Reactor Regulation and receive written approval from that office before proceeding with such activities.

F.

Reporting to thr # sission:

(a)

The licenso shall report any violations of the requirements Items C(l), C(3 through 133 and E contained in Section 2,twentyilaram(24)) hours by ". el of this 'icense within four and conf rmed by telegram ma or facsimile transmission to the NRC Re,gional Adm,inistrator, Reaion III, or desianee, not later than the first working day following the vioTation, with a written followup report within fourteen (14) working days.

(b)

The licensee shall notify the Commission, as soon as possible but not later than one hour, of any accident at this facility which could result in an unplanned release of ouantities oT fission products in excess of allowable limits for normal operation established by the Commission.

G.

The licensee shall have and maintain financial orotection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

Amendment No. J, L?, 102

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a.

ATTACHMENT TO LICENSE AMENDMENT N0. 102 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. The page marked with an asterisk is provided for convenience.

REMOVE INSERT I

I II II VII VII XXIII XXIII 1-3 1-3 1-4 1-4 1-5 1-5 1-6 1-6 1-7 1-7 1-8 1-8 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 3/4 3-19 3/4 3-19 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 3/4 6-3 3/4 6-3 3/4 6-18 3/4 6-18 3/0 6-19 3/4 6-19 3/4 6-22 3/4 6-22 3/4 6-23 3/4 6-23 3/4 6-24 3/4 6-24 3/4 6-36 3/4 6-36 3/4 6-43 3/4 6-43

3/4 6-44

3/4 6-44 B 3/4 6-1 B 3/4 6-1 B 3/4 6-la B 3/4 6-la B 3/4 6-4 B 3/4 6-4 B 3/4 6-4a B 3/4 6-4a

s 2-INDEX t

DEFINITIONS SECTION 1.0 DEFINITIONS EAEE 1.I' ACTI0N...........................................................

1-1 1.2 AVE RAG E PLANAR EX P0SURE..........................................

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.......................

1-1 1.4 CHANNEL CALIBRATION..............................................

1-1 F

1.S CHANNEL CHECK....................................................

1-1 1.6 CHANNEL FUNCTIONAL TEST..........................................

1-1 1.7 C O R E A L T E RAT I ON..................................................

1 - 2 1.8 CORE OPERATING LIMITS REP 0RT.....................................

1-2 l

1.9 C R I T I C AL POWE R RAT I0.............................................

1 -2 1.10 DOS E EQU I VAL ENT I-131............................................

1 -2 1.11 E-AVERAGE DISINTEGRATION ENERGY..................................

1-2 l

1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME...............

1-2 1.13 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME........

1-2 1.14 FRACTION OF LIMITING POWER DENSITY...............................

1-3 1.15 FRACTION OF RATED THERMAL P0WER..................................

1-3 1.16 FREQUENCY N0 TAT I ON...............................................

1 -3 l

t 1.17 GASE0US RADWASTE TREATMENT SYSTEM................................

1-3 1.18 I D E NT I F I E D L E AKAG E...............................................

1 -3 1.19 ISOLATION SYSTEM RESPONSE TIME...................................

1-3 t

1.20 L,...............................................................

1-4 1.21 LIMITING CONTROL R0D PATTERN.....................................

1-4

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1.22 LINEAR HEAT GENERATION RATE......................................

1-4 1.23 LOGIC SYSTEM FUNCTIONAL TEST.....................................

1-4 1.24 MAXIMUM FRACTION OF LIMITING POWER DENSITY.......................

1-4 1.25 MEMBER (S) 0F THE PUBLIC..........................................

1-4 f

1.26 MINIMUM CRITICAL POWER RATI0.....................................

1-4 i

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LA SALLE - UNIT 1 I

Amendment No. 102 t

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INDEX DEFINITIONS-SECTION DEFINITIONS (Continued)

PAGE l.27 0FFSITE DOSE CALCULATION MANUAL..................................

1 1.28 OPE RAB L E - O PE RAB I L I TY...........................................

1-5 :

1.29 OPERATIONAL CONDITION - CONDITION................................

1-5' l

1.30 PHYSICS TESTS....................................................

1-5 1.31 PRESSURE BOUNDARY LEAKAGE........................................

1-5 1.32 PRIMARY CONTAINMENT INTEGRITY.....................................-1-5 1.33 PROCESS CONTROL PR0 GRAM..........................................

1-6 r

1.34 PURG E - PU RG I N G..................................................

1 - 6 1.35 RAT E D TH E RMAL P0W E R..............................................

1 - 6 f

1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME..........................

1-6 1.37 REPORTABLE EVENT.................................................

1-6 1.38 R00 DENSITY......................................................

1-6 3.39 SECONDARY CONTAINMENT INTEGRITY..................................

1-7 1.40 SHUTDOWN MARGIN..................................................

1-7 1.41 SITE B0VNDARY....................................................

1-7 1.42 SOURCE CHECK.....................................................

1-8 1.43 ST AGG E RE D T E ST BAS I S..............................................

1 - 8 1.44 THERMAL P0WER....................................................

1-8 1.45 TURBINE BYPASS RESPONSE TIME.....................................

1-8 1.46 UNIDENTIFIED LEAKAGE.............................................

1-8 3.47 VENTILATION EXHAUST TREATMENT SYSTEM.............................

1-8 1.48 VENTING..........................................................

1-8 LA SALLE - UNIT 1 II Amendment No. 102 t

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INDEX LIMITING'COND1i.ONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE k

3 /4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity............................

3/4 6-1 g

Primary Containment Air Locks............................

3/4 6-5 MSIV Leakage Control System..............................

3/4 6-7' Drywell and Suppression Chamber Internal Pressure........

3/4 6-13 Drywell Average Ai r Temperature..........................

3/4 6-14 Drywell and Suppression Chamber Purge System............

3/4 6-15 3/4.6.2 DEPRESSURIZATION SYSTEMS Suppression Chamber......................................

3/4 6-16 l

Suppre s s i on Pool Spray...................................

3/4 6-20 Suppres si on Pool Cool i ng.................................

3/4 6-21 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES.....................

3/4 6-22 3/4.6.4 VACUUM RELIEF............................................

3/4 6-35.

3/4.6.5 SECONDARY CONTAINMENT Secondary Containment Integrity..........................

3/4 6-37 Secondary Containment Automatic Isolation Dampers........

3/4 6-38 t

Standby Gas Treatment System...............

3/4 6-40 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL well and Su DrNystems.....ppressionChamberHydrogenRecombiner 3/4 6-43 Drywell and Suppression Chamber Oxygen Concentration.....

3/4 6-44

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LA SALLE - UNIT 1 VII Amendment No. 102

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1 INDEX LIST OF TABLES (Continued) i TABL E PAGE 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM 1

_ AUTOMATIC ISOLATION DAMPERS..............................

3/4 6-39 3.7.5.2-1 DELUGE AND SPRINKLER SYSTEMS............................. 3/4 7-16 3.7.5.4-1

-FIRE H0SE STATIONS....................................... 3/4 7-19 3.7.7-1 AREA TEMPERATURE MONITORING..............................

3/4.7-25 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE...........................

3/4 8-7b 4.8.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS........................

3/4 8-18 3.8.3.3-1 MOTOR-0PERATED VALVES THERMAL OVERLOAD i

PROTECTION...............................................

3/4 8-27 i

B3/4.4.6-1 REACTOR VES S E L TOUGHNES S.................................

B 3/4 4-6 t

5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.....................

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i LA SALLE - UNIT 1 XXIII Amendment No. 102 f

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.o DEFINITIONS DiD-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME (Continued) breaker trip coil from when the monitored parameter exceeds its trip setpoint at the channel sensor of the associated:

a.

Turbine stop valves, and b.

-Turbine control valves.

The response time may be measured.by any series of sequential, overlapping or total steps such that the entire response time is measured.

FRACTION OF LIMITING POWER DENSITY 1.14 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing at a given location divided by the specified LHGR limit for that bundle type.

FRACTION OF RATED THERMAL POWER 1.15 The FRACTION bi RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

FRE0VENCY NOTATION 1.16 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASE0US RADWASTE TREATMENT SYSTEM 1.17 A GASE0US RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting, primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.18 IDENTIFIED LEAKAGE shall be:

a.

Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or b.

Leakage into the containment atmosphere from sources that are both s3ecifically located and known either not to interfere with t1e operation of the leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME 1.19 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall include diesel generator starting and sequence loading delays where a series of sequential, pplicable.

The response time may be measured by any overlapping or total steps such that the entire response time is measured.

LA SALLE UNIT 1 1-3 Amendment No. 102

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DEFINITIONS 1

1.20 The maximum allowable primary containment leakage rate, L,lculated peak shall be 0.635 % of primary containment air weight per day at the ca containment pressure (P, = 39.6 psig).

LIMITING CONTROL R0D PATTERN 1.21 A LIMITING CONTROL R0D PATTERN shall be a pattern which results in the l

core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.22 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit I

length of fuel rod.

It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.23 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, l

1.e, all relays and contacts, all trip units, solid state logic elements, etc. of a logic circuit, from sensor through and including the actuated device to verify 0PERABILITY. THE LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.24 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be the l

highest value of the FLPD which exists in the core.

MEMBERS (S) 0F THE PUBLIC 1.25 MEMBER (S)d with the plant.0F THE PUBLIC shall include all persons who a associate This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.26 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l

exists in the core.

OFFSITE DOSE CALCULATION MANUAL 1.27 The 0FFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology l

and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program.

The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.2.F.4 and (2)Radiolo ical Environmental Operating and Semi-descri tions of the information that should be included in the Annual Annual Radioactive Effluent Re ease Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

LA SALLE UNIT 1 1-4 Amendment No. 102

DEFINITIONS OPERABLE - OPERABILITY 1.28 A system, subsystem, train, component or device shall be OPERABLE or have l

OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of j

performing their rclated support function (s).

OPERATIONAL CONDITION - CONDITION

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1.29 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive l

combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

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PHYSICS TESTS 1.30 PHYSICS TESTS shall be those tests performed to measure the fundamental l

nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE 1.31 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault l

in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY 1.32 PRIMARY CONTAINMENT INTEGRITY shall exist when:

l a.

All primary containment penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE primary containment automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except for valves that are open under administrative control as permitted by Specification 3.6.3.

b.

All primary containment equipment hatches are closed and sealed.

c.

Each primary containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

d.

The primary containment leakage rates are maintained within the limits per Surveillance Requirement 4.6.1.1.b.

LA SALLE UNIT 1 1-5 Amendment No. 102

i DEFINITIONS e.

The suppression chamber is OPERABLE pursuant to Specification I

3.6.2.1.

f.

The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

g.

Primary containment structural integrity has been verified in j

accordance with Surveillance Requirement 4.6.1.1.e.

PROCESS CONTROL PROGRAM 1.33 The PROCESS CONTROL PROGRAM (PCP) shall contain the current-I' formulas, sampling, raalyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes 7

based on demonstrated processing of actual or simulated wet solid -

l wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.34 PURGE or PURGING shall be the controlled process of discharging air or l

l gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.35 RATED THERMAL POWER shall be a total reactor core heat transfer rate to '

l-the reactor coolant of 3323 MWT.

i REACTOR PROTECTION SYSTEM RESPONSE TIME l

1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from l

when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

The response time may be measured by any series of sequential, overlapping or-total steps such that the entire response time is measured.

REPORTABLE EVENT 1.37 A REPORTABLE EVENT shall be any of those conditions specified in l

.f Section 50.73 to 10 CFR Part 50.

i R00 DENSITY 1.38 ROD DENSITY shall be the number of control rod notches inserted as a

l fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% R0D DENSITY.

I LA SALLE UNIT 1 1-6 Amendment No.102

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DEFINITIONS SECONDARY CONTAINMENT INTEGRITY-

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1.39. SECONDARY CONTAINMENT INTEGRITY. shall exist when:

l a.

All secondary containment penetrations required to be closed-t during accident conditions are either:

1.

Capable of being closed by an'0PERABLE secondary i

containment automatic isolation system, or s

2.

Closed by at least one manual valve, blind flange,- or deactivated automatic damper secured in its closed

. position, except as provided in Table 3.6.5.2-1 of i

Specification 3.6.5.2.

3 b.

All secondary containment hatches and blowout panels are closed j

and sealed.

c.

The standby gas treatment system is'0PERABLE pursuant to

-l Specification 3.6.5.3.

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d.

At least one door in each access to the secondary containment is closed.

e.

The sealing' mechanism associated with each secondary i

containment penetration, e.g., welds, bellows or 0-rings, is OPERABLE.

f.

The pressure within the secondary containment is less than or i

equal to the value required by Specification 4.6.5.1.a.

SHUTDOWN tiARGIN 1.40 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is-in the shutdown condition; cold, i.e. 68*F; and xenon free.

HTE BOUNDARY 1.41 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

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l LA SALLE UNIT 1 1-7 Amendment No. 102

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d DFFINITIONS SOURCE CHECK 1.42 A SOURCE CHECK shall be the qualitative assessment of channel response l

l when the' channel sensor is exposed to a radioactive source.

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STAGGERED TEST BASIS 1.43 A STAGGERED TEST BASIS shall consist of:

l.

l a.

A test schedule for n systems, subsydems, trains or other designated components obtained by di ' ding the specified test interval into n equal subintervals.

j b.

The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

l THERMAL POWER 1.44 THERMAL POWER shall be the total reactcr core heat transfer rate to the l

reactor coolant.

TURBINE BYPASS SYSTEM RESPONSE TIME 1.45 The TURBINE BYPASS SYSTEM RESPONSE TIME shall be time interval from when l

the turbine bypass control unit generates a turbine bypass valve flow i

signal until the turbine bypass valves travel to their required positions. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response. time' is measured.

UNIDENTIFIED LEAKAGE 1.46 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED l

LEAKAGE.

t VENTILATION EXHAUST TREATMENT SYSTEM i

1.47 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l

installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of i

removing iodines or particulates from the gaseous exhaust stream prior to i

the release to the environment (such a system is not considered to have i

any effect on noble gas effluents).

Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST l

TREATMENT SYSTEM components.

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VENH NG i

1.48 VENTING shall be the controlled process of discharging air or gas from a confinet.nt to maintain temperature, pressure, humidity, concentration or i

other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

LA SAll.E UNIT 1 1-8 Amendment No. 102

p TABLE 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICABLE OPERATED BY CHANNELS PER OPERATIONAL.

TRIP FUNCTION SIGNAL TRIP SYSTEM (b)

CONDITION ACTION' l'

A.

AUTOMATIC INITIATION 1.

PRIMARY CONTAINMENT ISOLATION i

a.

Reactor Vessel Water Level Low, Level 3 7

2 1, 2, 3 -

20 Low Low, level 2 2, 3 2

1,2,3 20 Low Low Low, Level 1 1, 10 2'

. 1, 2, 3 20 b.

Drywell Pressure - High 2, 7, 10 2

1, 2, 3 20 c.

Main Steam Line 1)

Radiation - High 1

2 1, 2, 3 21 3

2 1, 2, 3 22 2)

Pressure - Low 1

2 1

23 i

3)

Flow - High 1

2/line")

1, 2, 3 21 d.

Main Steam Line Tunnel 3

Temperature - High 1

2

.1 'g)[3,'g ' 2""3),

e.

Main Steam Line Tunnel ATemperature - High 1

2'

- 1"((/[g 2""33, f.

Condenser Vacuum - Low 1

2 1,

2*,.3*

21 2.

SECONDARY CONTAINMENT ISOLATION l

a.

Reactor Building Vent Exhaust i

Plenum Radiation - High 4""

2 1,-2, 3 and **

'24-I b.

Drywell Pressure - High 4""

2 1,2,3 24 L

c.

Reactor Vessel Water l

Level - Low Low, level 2 4""

2 1,-2, 3,~and '

24 d.

Fuel' Pool Vent Exhaust Radiation - High 4""

2 1, 2, 3, and **

24 l

l LA SALLE UNIT 1-3/4'3-11

~

Amendment No. 107

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICA8LE OPERATED BY CHANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL TRIP SYSTEM (b)

CONDITION

-ACTION

~l' 3.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

A Flow - High 5

1 1, 2, 3' 22 b.

Heat Exchanger Area Temperature - High 5

I/ heat-1,2,3 22 exchanger c.

Heat Exchanger Area Ventilation AT - High 5

1/ heat 1, 2, 3 22 exchanger d.

SLCS Initiation 5")

NA 1, 2, 3 22 e.

Reactor Vessel Water

.t Level - Low Low, Level 2 5

2 1, 2, 3 22 4.

REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION a.

RCIC Steam Line Flow - High 8

1 1,2,3 22 i

b.

RCIC Steam Supply-i Pressure - Low

. 8, 9(*

2 I, 2, 3 22-a c.

RCIC Turbine Exhaust 1

Diaphragm Pressure - High 8

2-1, 2, 3 22 3

d.

RCIC Equipment Room 4

Temperature - High 8

1 1,2,3 22 e.

RCIC Steam Line Tunnel Temperature - High 8

1 1,2,3 22 f.

RCIC Steam Line Tunnel A Temperature - High 8

1 1, 2,.3 22 g.

Drywell Pressure --High 9(O 2

1, 2, 3 22 h.

RCIC Equipment Room.

A Temperature - High 8

1 1, 2, 3.

22 i

LA SALLE - UNIT 1 3/4 3-12

' Amendment No. 107 i

L~

z

._____.____---......__.___._.____....-_.._..___.~.__._,._.,._..___._

y,-

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION' VALVE GROUPS HINIMUM OPERABLE APPLICABLE OPERATED BY CHANNELS PER OPERATIONAL

+

TRIP FUNCTION SIGNAL TRIP SYSTEM (b)

CONDITION ACTION l

5, RHR SYSTEM STEAM CONDENSING MODE ISOLALLON a.

RHR Equipment Area a Temperature - High 8

1/RHR area 1, 2, 3 22 b.

RHR Area Temperature -

5 High 8-1/RHR area 1, 2, 3 22.-

c.

RHR Heat Exchanger Steam Supply Flow - High 8

1 1,2,3 22 6.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION l

a.

Reactor Vessel Water 4

Level - Low, level 3 6

2 1, 2, 3 25 b.

Reactor Vessel (RHR Cut-in Permissive)

Pressure - High 6

1 1,- 2,~ 3 25 c.

RHR Pump Suction Flow - High 6

1 1,.2,'3 25 d.

RHR Area Temperature -

High 6

1/RHR area 1, 2, 3 25 i

e.

RHR Equipment Area AT - High 6-1/RHR area

-1, 2, 3 25' B.

MANUAL INITIATION-'

l.

Inboard Valves 1,2,5,6,7 1/ group 1, 2, 3 26' 2.

Outboard Valves 4*)g'5,6,7 1/ group

-1,-2, 3-26 l g 3.

Inboard Valves 1/ group-1, 2, 3 and 26 4.

Outboard Valves 4(**)

1/ group 1, 2, 3 and **,f 26 5.

Inboard Valves 3,8,9

-1/ valve 1, 2, 3 26-6.

Outboard Valves 3 _ 8 9 1/ valve 1, 2,. 3 26 33,

7.

Outboard Valve 8

1/ group 1,-2, 3 26 i

I LA SALLE - UNIT 1 3/4 3-13

-Amendment No. 102~

e TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION l

ACTION ACTION 20 Be in at least H0T SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

'l ACTION 21 Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 22 -

Close the affected system isolation valves within I hour and declare the affected system inoperable.

ACTION'23 -

Be in at least'STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24 Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating.within I hour.

ACTION 25 Lock the affected system isolation valves closed within I hour and declare the affected system inoperable.

ACTION 26 --

Provided that the manual initiation function 'is OPERABLE for i

each other group valve, inboard or outboard, as applicable, in each line, restore the manual initiation function to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise:

a.

Be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or b.

Close the affected system isolation valves within the next hour and declare the affected system inoperable.

NOTES f

May be bypassed with reactor steam pressure s 1043 psig and all turbine stop valves closed.

When_ handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

(a) Deleted.

l (b) A channel may be placed in-an inoperabie status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the channel in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

In addition for those trip systems with a design providing only one channel per trip system, the channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for required surveillance testing without placing the channel in the tripped condition provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is operable and all required actuation instrumentation for that redundant valve is OPERABLE, or place the trip system in the tripped condition.

(c) Also actuates the standby gas treatment system.

(d) A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE.

(e) Also actuates secondary containment ventilation isolation dampers per Tabl e 3.6.5.2-1.

(f) Closes only RWCU system inlet outboard valve.

LA SALLE - UNIT 1 3/4 3-14 Amendment No. 102

TABLE 3.3.2-3 (Continu:d)'

l ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Secondsf g

6.

RH8_ SYSTEM SHUTDOWN COOLING MODE ISOLATION N/A a.

. Reactor Vessel Water Level - Low, Level 3 b.

Reactor Vessel (RHR Cut-In Permissive) Pressure - High c.

RHR Pump Suction Flow - Iligh d.

RHR Area Cooler Temperature High e.

RHR Equipment Area aJ High B.

NANUAL INITIATION

-N/A 1.

Inboard Valves H

2.

Outboard Valves 3.

Inboard Valves 4.

Outboard Valves S.

Inboard Valves 6.

Outboard Valves 7.

Outboard Valve TABLE NOTATIONS Isolation system instrumentation response timm for MSIVs only.

No diesel generator delays assumed.

Radiation detectors are exempt from response time testing.

Response time shall be measured from detector output or the input of the first electronic component in the channel.

j Isolation system instrumentation response time specified for the Trip l

Function actuating the MSIVs shall be added to MSIV isolation time to l

obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

i l

P N/A. Not Applicable.

I LA SALLE - UNIT I 3/4 3-19 Amendment No. 102

. ~,

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2,* and 3.

ACTION:

Without PRIMARY, CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within I hour or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE0VIREMENTS 4.6.1.1 PRIMARY CONTAIHMENT INTEGRITY shall be demonstrated:

a.

At least once, per 31 days by verifying that all. primary containment I

penetrations not capable of being closed by OPERABLE containment automatic isolation valves a:id required to be closed during accident I

conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except for valves that are open under administrative control as permitted by Specification 3.6.3.

b.

Perform required v4sual examinations and leakage rate testing except for primary containment air lock testing and main steam lines through the isolation valves, in accordance with and at the frequency" specified by 10 CFR 50, Appendix J. as modified by approved exemptions.

The overall integrated leakage rate acceptance criterion is 51.0 L,.

The Type B and C combined leakage rate acceptance criterion is performed. However, during the first unit startup following testing s 0.60 L in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, the leakage rate acceptance criteria are < 0.60 L for the combined Type B and Type C tests, and < 0.75 L, for the Ty,pe A test.

i

See Special Test Exception 3.10.1.

- Except valves, blind flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or j

more often than once per 92 days.

"The provisions of Specification 4.02 are not applicable to the frequencies specified by 10 CFR 50, Appendix J.

LA SALLE - UNIT 1 3/4 6-1 Amendment No. 102 I

CONTAINMENT SYSTEMS o.

PRIMARY CONTAINMENT LEAKAGE SURVEILLANCE REQUIREMENTS (Continued) c.

By verifying each primary containment air lock OPERABLE per

'i Specification 3.6.1.3.

d.

By verifying the suppression chamber OPERABLE per Specification 3.6.2.1.

1 i

e.

Verify primary containment structural integrity in accordance with the~ Inservice Inspection Program for Post Tensioning Tendons. The

. frequency shall be in accordance with the Inservice Inspection Program for Post Tensioning Tendons.

t i

?

t LA SALLE - UNIT 1 3/4 6-2 Amendment No. 102 f

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT LEAKAGE i

~ LIMITING CONDITION FOR OPERATION 3.6.1.2 DELETED l

Pages 3/4 6-3 and 3/4 6-4 Deleted i

J i

I LA SALLE - UNIT 1 3/4 6-3 (next page is 3/4 6-5)

Amendment No 102

CONTAINMENT SYSTEMS' SURVEILLANCE RE0VIREMENTS (Continued) c.

By verifying at least two suppression chamber water level instru-mentation channels and at least 14 suppression pool water.

1 temperature instrumentation channels, 7 in each of two divisions, OPERABLE by performance of a:.

1.

' CHANNEL. CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , j

2.

OiANNEL FUNCTIONAL TEST at least once per 31 days, and j

3.

CHANNEL CALIDRATION at least once per 18 months.

The suppression chamber water level and suppression pool temperature alarm setpoint shall be:

a)

High water level s +2 inches

b)

Low water level 1 -3 inches

l c)

High temperature 5 105'F d.

By conducting drywell-to-suppression chamber bypass leak tests at least once per 18 months at an initial differential pressure of 1.5 psi and verifying that the A//k calculated from the measured

.t leakage is within the specified limit.

.j If any 1.5 psi leak test results in a calculated A//k >20% of the specified limit, then the test schedule.for subsequent tests shall be reviewed by the Commission.

If two consecutive 1.5 psi leak tests res"'t in a calculated A//k greater than the specified limit, then:

t 1.

A 1.5 psi leak test shall be performed at least once per 9 months until two consecutive 1.5 psi leak tests result in the calculated A//k within the specified limits, and 2.

A 5 psi leak test, performed with the second consecutive successful 1.5. psi leak test, results in a calculated A//k within the specified limit, after which the above schedule of once per 18 months for only 1.5 psi leak tests may be resumed.

If-any required 5 )si leak test results in a calculated A//k greater than the specified limit, tien the test schedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 5 psi leak tests reult in a calculated A//k greater than the specified limit, then a 5 psi leak test shall be performed at least once per 9 months until two consecutive 5 psi leak tests result in a calculated l

A//k within the specified limit, after which the above schedule of once per 18 months for only 1.5 psi leak tests may be resumed.

Level is referenced to a plant elevation of 699 feet 11 inches (See Figure B 3/4.6.2-1).

1 LA SALLE - UNIT 1 3/4 6-18 Amendment No.102

INTENTIONALLY LEFT BLANK LA SALLE - UNIT 1 3/4 6-19 Amendment No. 102

1 1

4 CONTAINMENT SYSTEMS d

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES LIMITING' CONDITION FOR OPERATION 3.6.3. Each primary containment isolation,, valve and reactor instrumentation line excess flow check valve shall be OPERABLE APPLICABILITY: OPERATIONAL. CONDITIONS 1, 2, and 3.

ACTION:

a.

With one or more of the primary containment isolation valves, except the reactor instrumentation line excess flow check valves, inoperable:

1.

Maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either-

.i a)

Restore the inoperable valve (s) to OPERABLE status, or q

b)

Isolate each affected penetration by use of at least one deactivatpd automatic valve secured in the isolated position, or 4

c)

Isolate each affected penetration by use of at least one l

closed manual valve or blind flange.*

2.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

t b.

With one or more of the reactor instrumentation line excess flow check i

valves inoperable:

l l

1.

Operation may continue and the provisions of Specification 3.0.3 I

are not applicable provided tha.t within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

l 1

a)

The inoperable valve is returned to OPERABLE status, or b)

The instrument line is isolated and the associated instrument is declared inoperable.

2.

Otherwise, be in at least HOT' SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i r

l

-i

Isolation valves closed to satify these requirements may be reopened on an intermittent basis under administrative control.

- Locked or sealed closed valves may be opened on an intermittent basis under i

administrative control.

j i

LA SALLE - UNIT 1 3/4 6-22 Amendment No. 102

[

e 4

-e

.-nu

,-..,w.a.

..wn n,

v

,.m..

n-

,e.

CONTAINMENT SYSTEMS i

SURVEILLANCE RE0UIREMENTS l

4.6.3.1 Each primary containment isolation valve shall be demonstrated l

OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of fM 1 travel and verifying the specified isolation time.

4.6.3.2 Each primary containment automatic isolation valve shall be l

demonstrated OPERABLE during COLD SHUTDOWN or REFUELING at least once per 18 months by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation position.

4.6.3.3 The isolation time of each primary containment power operated or automatic isolation valve shall be determined to be within its limit when l

tested pursuant to Specification 4.0.5.

4.6.3.4 Each reactor instrumentation line excess flow check valve shall be l

demonstrated OPERABLE at least once per 18 months by verifying that the valve checks flow.

4.6.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying the continuity of the explosive charge.

b.

At least once per 18 months by removing the explosive squib from at least one explosive valve such that the explosive squib in each explosive valve will be tested at least once per 90 months, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired. No explosive squib shall remain in use beyond the expiration of its shelf-life and operating-life.

4.6.3.6 At least once per 18 months:

a.

Verify leakage rate through all four main steam lines through the isolation valves is 5100 scfh when tested at 2 25.0 psig.

b.

Verify combined leakage rate of s I gpm times the total number of primary c. <tainment isolation valves through hydrostatically tested lines that penetrate the primary containment is not excee these isolation valves are tested at 1.1 P,, 2 43.6 psig.,ded when

Results shall be excluded from the combined leakage for all penetrations and seals subject to Type B and C tests.

LA SALLE - UNIT 1 3/4 6-23 Amendment No.102

y r:

e, j s

i t

a INTENTIONALLY LEFT BLANK Pages 3/4 6-25 through 3/4 6-34 Deleted LA SALLE - UNIT 1 3/4 6-24 (next page is 3/4 6-35)

Amendment No. 102

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.4.2 The manual. isolation valves on both sides of an inoperable and/or i

open suppression chamber-drywell vacuum breaker shall be verified to be closed-at least once per 7 days.

l i

i f

i i-

}

LA SALLE - UNIT 1 3/4 6-36 Amendment No. 102 6

-r.

y r-w

CONTAINMENT SYSTEMS-3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL DRYWELL AND SUPPRESSION CHAMBER HYDR 0 GEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION 3.6.6.1. Two independent drywell and suppression chamber hydrogen recombiner systems shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one drywell and/or suppression chamber hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE0VIREMENTS 4.6.6.1 Each drywell and suppression chamber hydrogen recombiner system shall be demonstrated OPERABLE:

a.

At least once per 92 days by cycling each flow control valve and recirculation valve through at least one complete cycle of full travel.

b.

At least once per 18 months by verifying, during a recombiner system functional test:

1.

That the heaters are OPERA 9LE by determining that the current in each phase differs by less than or equal to 5% from the other phases and is within 5% of the value observed in the original acceptance test, corrected for line voltage differences.

2.

That the reaction chamber gas temperature increases to 1200 i

25*F within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

c.

At least once per 18 months by:

1.

Performing a CHANNEL CAllBRATION of all recombiner operating instrumentation and control circuits.

2.

Verifying the integrity of all heater electrical circuits by performing a resistance to ground test within 30 minutes following the above required functional test. The resistance to l

ground for any heater phase shall be greater than or equal to 100,000 ohms.

i s

LA SALLE - UNIT 1 3/4 6-43 Amendment No. 102 I

CONTAINMENT SYSTEMS DRYWELL AND SUPPRESSION CHAMBER OXYGEN CONCENTRATION LIMITING CONDITION FOR OPERATION 3.6.6.2 The drywell and suppression chamber atomosphere oxygen concentration shall be less than 4% by volume.

APPLICABILITY:

OPERATIONAL CONDITION 1*, during the time period:

Withim 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is greater than 15% of RATED a.

THERMAL POWER, following startup, to b.

Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing THERMAL POWER to less than 15% of RATED THERMAL POWER, preliminary to a scheduled reactor shutdown.

ACTION:

With the oxygen concentration in the drywell and/or suppression chamber exceeding the limit, restore the oxygen concentration to within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least STARTUP within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

1 SURVEILLANCE REQUIREMENTS P

4.6.6.2 The oxygen concentration in the drywell and suppression chamber shall be verified to be within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is greater than 15% of RATED THERMAL POWER and at least once per 7 days thereafter.

t "See Special Test Exception 3.10.5.

LA SALLE - UNIT 1 3/4 6-44

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and arsociated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

The structural integrity of the primary containment is ensured by the successful completion of the Inservice Inspection Program for Post Tensioning Tendons and by associated visual inspections of the steel liner and penetrations for evidence of deterioration or breach of integrity. This ensures that the structural integrity of the primary containment will be maintained in accordance with the provisions of the Primary Containment Tendon Surveillance Prcgram. Testing and Frequency are consistent with the recommendations of Regulatory Guide 1.35, Revision 3, except that the Unit I and 2 primary containments shall be treated as twin containments even though the Initial Structural Integrity Tests were not within 2 years of each other.

PRIMARY CONTAINMENT INTEGRITY is maintained by limiting overall integrated leakage to s 1.0 L and the Type B and C combined leakage rate acceptancecriterioniss0.66L,. Prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test, the combined Type B and C leakage must be < 0.60 L, and the overall Type A leakage must be < 0.75 L, when a Type A test is scheduled. Compliance with this LC0 will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

The maximum allowable leakage rate for the primary containment (L 0.635% by weight of the containment atmosphere per day at the calculate 8) is maximum peak containment pressure (P,) of 39.6 psig.

Individual leakage rates specified for the primary containment air lock, main steam lines through the isolation valves, and valves in hydrostatically tested lines are addressed in LC0 3.6.1.3, and Surveillance Requirement 4.6.3.6.

Surveillance Requirement 4.6.1.1.b maintains PRIMARY CONTAINMENT INTEGRITY by requiring compliance with the visual examinations and leakage rate test requirements of 10 CFR 50, Appendix J, as modified by approved exemptions.

Failure to meet air lock leakage testing (4.6.1.3) or main steam isolation valve leakage (4.6.3.6.a) does not necessarily result in a failure of this Surveillance Requirement, 4.6.1.1.b.

The impact of the failure to meet these Surveillance Requirements 4.6.1.3 and 4.6.1.1.b must be evaluated against the Type A, B, and C acceptance criteria of 10 CFR 50, Appendix J, as modified by approved exemptions. The leakage limits for main steam lines through the isolation valves and leakage test results of Surveillance Requirement 4.6.3.6.a are not included in the total sum of Type B and C tests (approved exemption). As-left leakage prior to the first startup after LA SALLE - UNIT 1 B 3/4 6-1 Amendment No.102

l 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY (Continued) performing a required 10 CFR 50, Appendix J, leakage test'is required to be

< 0.60 L, for combined Type B and C leakage, and < 0.75 L, for overall Type A leakage. At all other times between required Type A tests, the acceptance.

criteria is based on an overall Type A leakage limit of s 1.0 L,.

At s 1.0 L, the offsite dose consequences are bounded by the assumptions of the safety analysis. The combined Type B and C leakage remains as s 0.60 L, between scheduled tests, in accordance with Appendix J.

The Frequency is required by 10 CFR 50, Appendix J, as modified by approved exemptions. Th;;, 8.0.2 (which allows Frequency extensions) does not i

apply to Surveillance Requirement 4.6.1.1.b.

I i

3/4.6.1.2 DELETED 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS.

The limitation on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY i

and the primary containment leakage rate given in Specification 3/4.6.1.1.

l r

The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during l

reactor operation. Only one closed door in each air lock is required to maintain the integrity of the containment.

i e

i i

l i

i LA SALLE - UNIT 1 B 3/4 6-la Amendment No. 102 i

~

t CONTAINMENT SYSTEMS BASES DEPRESSURIZATION SYSTEMS (Continued)

Because of the large volume and thermal capability of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.

By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for the external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event of i

safety-relief valve inadvertently opens or sticks open. As a minimum this action shall include: (1) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES 1

Primary Containment Isolation Valves (PCIVs) form a part of the primary containment boundary. The PCIV safety function is related to control primary containment leakage rates during accidents or other conditions to limit the untreated release of radioactive materials from the containment in excess of the design limits.

The automatic isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal.

The valves covered by this specification are listed with their associated stoke times, and other design information for lines penetrating the Primary Containment, in i

UFSAR Section 6.2.

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact.

Main steam lines through the isolation valves and hydrostatically tested valves must meet alternative leakage rate requirements. Other PCIV leakage rates are addressed by specification 3/4.6.1.1, " PRIMARY CONTAINMENT INTEGRITY". UFSAR Section 6.2 also describes special leakage test requirements and exemptions.

This specification'provides assurance that the PCIVs will perform their f

designed safety functions te control leakage from the primary containment during accidents.

t The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication F

LA SALLE - UNIT 1 B 3/4 6-4 Amendment No. 102

CONTAINMENT SYSTEMS

)

BASES PRIMARY CONTAINMENT ISOLATION VALVES (Continued) with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude assess to close the valves and that this action will prevent the release of radioactivity outside the primary containment.

Surveillance Requirement 4.6.3.6.a verifies leakage through all four main steam lines is s 100 scfh when tested at 2 P, (25.0 psig). The transient and accident analyses are based on leakage at the specified leakage rate.

The leakage rate for main steam lines through the isolation valves must be verified to be in accordance with the leakage test requirements of 10 CFR 50, Appendix J, as modified by approved exemptions. A Note has been added to this Surveillance Requirement requiring the results to be excluded from the total of Type B and Type C tests. This ensures that leakage rate for main steam lines through the isolation valves is properly accounted for in accordance with an approved exemption. The frequency is "at least once per 18 months" in accordance with an approved exemption.

Surveillance Requirement 4.6.3.6.b test of hydrostatically tested lines provides assurance that the assumptions of UFSAR Section 6.2 are met. The combined leakage rates must be demonstrated in accordance with the leakage rate test at a frequency of "at least once per 18 months". A Note has been added to this Surveillance Requirement requiring the results to be excluded the total of Type B and Type C tests. This is in accordance with 10 CFR 50, Appendix J, and approved exemptions.

3/4.6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell. This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident. There are four valves to provide redundancy so that operation may continue for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with one vacuum breaker inoperable provided that the manual isolation valves on each side are in the closed position.

LA SALLE - UNIT 1 B 3/4 6-4a Amendment No. 102

~

pMk p-4 UNITED STATES g

g NUCLEAR REGULATORY COMMISSION "t

WASHINGTON, D.C. 20565 4001

%...../

COMMONWEALTH EDISON COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.87 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 October 24, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and t!.e Commission's regulations set forth in 10 CFR Chapter I; 8.

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 Comission'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.D* of Facility Operating License No.

NPF-18 are hereby amended to read as follows:

Page 10 is attached, for convenience, for the composite license to reflect this change.

. 2.C.(2)

Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 87, 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.

D.

The facility requires exemptions from certain requirements of 10 CFR Part 50, 10 CFR Part 70, and 10 CFR Part 73. These include:

(a)

Exemptions from certain requirements of Appendices G, H and J to 10 CFR Part 50, and to 10 CFR Part 73 are described in the Safety Evaluation Report and Supplement Numbers 1, 2, 3, and 5 to the Safety Evaluation Report.

(b)

An exemption was requested until completion of the first refueling from the requirements of 10 CFR 70.24.

(c)

An exemption from the requirement of paragraph III.D of Appendix J to conduct the third Type A test of each ten-year service period when the plant is shutdown for the 10-year plant inservice inspections.

(d)

A one-time exemption from the requirement of paragraph III.A.6(b) of Appendix J to resume a Type A test schedule of three times in ten years.

Exemptions (c) and (d) are described in the Safety Evaluation accompanying Amendment No.

to this license. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. Therefore, these exemptions are hereby granted pursuant to 10 CFR 50.12. With the granting of these exemptions the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

'e

! t 3.

This amendment is effective upon date of issuance and 'shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMMISSION D

b William D. Reckley, Pro anager Project Directorate III.

Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Attachments:

1.

License page 10 2.

Changes to the Technical' Specifications Date of Issuance:

March 16, 1995 P

N I

s i

o hg 1

l&

q

- 10.-

1 i

' D.' The' facility requires exemptions from'certain requirements of 10 CFR-l.

)

U Part 50, 10 CFR Part 70..and 10 CFR Part 73. These' include:

J to 10 CFR Part 50.and to 10 CFR Part 7$pendices G H and l-Exemptions'from certain requirements of A

-(a) aredescrlbedin 1

the Safety Evaluation Report and Supplement Numbers 1, 2, 3, and 5 to the Safety Evaluation Report.

-(b)

Anexemptionwasrequesteduntilcombetionofthe'first' l~

l refueling from the requirements of I CFR 70.24.

(c)

An exemption from the requirement of paragraph III.D of..

Appendix J to conduct the third Type A test of each ten-year service period when the plant'is shutdown for the 10-year plant inservice inspections.

l (d)

A one-time exemption from the requirement of paragraph.

III.A.6Jb three time)s in ten years.of Appendix J to resume a Type A test schedule.of Exemptions. c -and=

l described in the Safety Evaluation acco(mp)anying(d) are Amendment

'l No.

to this license. These exemptions are authorized by i

law and will not endanger life or property or the common defense and security.and are otherwise in the public i

interest. Therefore pursuantto10CFR56.theseexemptions.areherebygranted

12. With the granting of these.

exemptions the facility will operate to.the extent i'

inconformitywitbtheapplication authorized herein the provlsions of the Act, and the rules and, as-amended regulatlonsoftheCommission.

E.

Before engaging in additional construction or operational' activities which may result in a significant adverse environmental impact'that was not evaluated or that is significantly greater than that evaluated in the Final Environmental Statement and its Addendum the licensee shall provide a written notification to the Director of the Office of Nuclear Reactor Regulation and receive written approval from that office before proceeding with such activities.

F.

With the exception of Section 2, Item C(2), the licensee shall report any violations of the requirements contained in Section 2.C' and 2.E of this license within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone and confirm by Administrator, gram, or facsimile transmission to the NRC Regional telegram, mail Region III or that administrator's designee, no later than the first working day following the violation, with a written followup report within 14 days.-

G.

The licensee shall notify the Commission, as soon as possible but not later than one hour, of any accident at this facility which could result in an unplanned release of quantities of fission -

products in excess of allowable limits' for normal operation -

established by the Commission.

H.

The licensee shall have.and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

Amendment No. 87

ATTACHMENT TO LICENSE AMENDMENT NO. 87 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO.50-074 Replace the following pages of the Appedix "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 1

I I

l II II 3

VII VII XXII XXII l-3 1-3 1-4 1-4 1 -5 1-5 1-Sa 1-Sa 1-6 1-6 1-7 1-7 l

3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 3/4 3-14 3/4 3-14 j

3/4 3-19 3/4 3-19 3/4 6-1 3/4 6-1 3/4 6-2 3/4 6-2 i

3/4 6-3 3/4 6-3 3/4 6-21 3/4 6-21 3/4 6-22 3/4 6-22 3/4 6-25 3/4 6-25 3/4 6-26 3/4 6-26 3/4 6-27 3/4 5-27 3/4 6-39 3/4 6-39 l

3/4 6-45

3/4 6-45 l

3/4 6-46 3/4 6-46 B 3/4 6-1 B 3/4 6-1 B 3/4 6-2 B 3/4 6-2 B 3/4 6-2c B 3/4 6-2a l

B 3/4 6-4 B 3/4 6-4 8 3/4 6-4a B 3/4 6-4a i

l 1

l l

i

a A

I i

lEQEl i

DEFINITIONS SECTION 1.0 DEFINITI0ftS E6SE 1.1 ACTI0N............................................................'l-1 1.2 AV E RAG E PLANAR EX P0SURE..........................................

1-1 l

1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.......................

1-1 i

1.4 CHANN E L C AL I B RAT I ON..............................................

1 - 1 i

1.5 CHANNEL CHECK....................................................

1-1 1.6 CHANNEL FUNCTIONAL TEST..........................................

1-1 1.7 C OR E A L T E RA f l 0N..................................................

1 - 2 1.8 CORE OPERATING LIMITS REP 0RT.....................................

1-2 1.9 CRITICAL POWER RATI0.............................................

1-2 1.10 DOSE EQUIVALENT I-131............................................

1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY..................................

1-2 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME...............

1-2 2.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME........

1-2 1.14 FRACTION OF LIMITING POWER DENSITY...............................

1-3 a

1.15 FRACTION OF RATED THERMAL P0WER..................................

1-3

- l 1.16 FREQUENCY N0TATION...............................................

1-3 1.17 GASE0US RADWASTE TREATMENT SYSTEM................................

1-3 1.18 IDENTIFIED LEAKAGE...............................................

1-3 1.19 ISOLATION SYSTEM RESPONSE TIME..................

1-3 1.2D L,...............................................................

1-3

[

1.21 LIMITING CONTROL R0D PATTERN.....................................

1-4 1.22 LINEAR HEAT GENERATION RATE......................................

1-4 1.23 LOGIC SYSTEM FUNCTIONAL TEST.....................................

1-4 t

1.?4 MAXIMUM FRACTION OF LIMITING POWER DENSITY.......................

1-4 1.25 MEMBER (S) 0F THE PUBLIC..........................................

1 :

1.26 MINIMUM CRITICAL POWER RATI0.....................................

1-4 LA SALLE - UNIT 2 I

Amendment No. 87 h

t i

1.

m

l O

l INDEX DEFINITIONS SECTION ii DEFINITIONS (Continued)

EASE i

1.27 0FFSITE DOSE CALCULATION MANUAL..................................

1-4 1

i 1.28 ' OPERABLE - OPERABILITY...........................................

1-5 1.29 OPERATIONAL CONDITION - CONDITION................................

1-5

~

1.30 PH Y S I C S T E S T S....................................................

1 - 5 1.31 PRESSURE BOUNDARY LEAKAGE........................................

1-5 I

1.32 PRIMARY CONTAINMENT INTEGRITY....................................

1-5 1.33 PROCESS CONTROL PR0 GRAM..........................................

1-Sa j

1.34 PU R G E - P U R G I N G..................................................

1 - S a 1.35 RAT E D T H E RMAL P0W E R..............................................

1 - S a 1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME..........................

1-Sa 1.37 RE PORT AB L E E V E NT.................................................

1 - S a 1.38 R00 DENSITY.......................................................

1-Sa 1.39 SECONDARY CONTAINMENT INTEGRITY..................................

1-6 1.40 SHUTDONN MARG I N..................................................

1 - 6 i

1.41 S I T E B0VND ARY....................................................

1 - 6 1.42 SOURCE CHECK.......

1-6 1.43 STAGGERED TEST BASIS.............................................

1-6 l

1.44 THERMAL P0WER....................................................

1-7 I

1.45 TURB I N E B Y PAS S RE S PON S E T I M E......................................

1 - 7 1.46 UN I DENT I F I ED L E AKAG E.............................................

1-7 1.47 VENTIiATION EXHAUST TREATMENT SYSTEM.............................

1-7 1.48 VENTING..........................................................

1-7 l

i r

LA SALLE - UNIT 2 II Amendment No. 87

i r

t INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS a

~SECTION pgE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT i

Primary Containment Integrity............................

3/4 6-1 l

i Primary. Containment Air Locks............................

3/4 6-5 MSIV Leakage Control System..............................

3/4_6-7 Drywell and Suppression Chamber Internal Pressure........-

3/4 6-16 j

Drywell Average Air Temperature..........................

3/4 6-17 j

Drywell and Suppression Chamber Purge System............

3/4 6-18 j

3/4.6.2 DEPRESSURIZATION SYSTEMS l

4 Suppression Chamber......................................

3/4 6-19 Suppression Pool Spray...................................

3/4 6-23 Suppression Pool Cooling.................................

3/4 6-24 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES.....................

3/4 6-25 i

3/4.6.4 VACUUM RELIEF............................................

3/4 6-38 3/4.6.S SECONDARY CONTAINMENT Secondary Containment Integrity..........................

3/4 6-40 j

Secondary Containment Automatic Isolation Dampers........

3/4 6-41 l

r Standby Gas Treatment System.............................

3/4 6-43 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Drywell and Suppression Chamber Hydrogen Recombiner 5ystems................................................

3/4 6-46 Drywell and Suppression Chamber Oxygen Con entration.....

3/4 6-47 b

I LA SALLE - UNIT 2 VII Amendment No. 87 f

5

1.@fl

'i LIST OF TABLES (Continued) i TABLE fAGI 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 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION................ 3/4 3-70 l

4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS......................... 3/4 3-71 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION..................... 3/4 3-76 3.3.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION................................... 3/4 3-83 4.,.7.11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS......... 3/4 3-84 i

3.3.8-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION......................... 3/4 3-87 3.3.8-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM l

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 COOLANT SYSTEM PRESSURE ISOLATION VALVES... '3/4 4-10 i

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

3/4 4-13 l

4.4.5-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM.................................. 3/4 4-16 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE............................... 3/4 4-20 l

l 9

i LA SALLE - UNIT 2 XXII Amendment No. 87

j l

' DEFINITIONS 1

j.

FRACTION OF' LIMITING POWER DENSITY

^!

q 1.14 The-FRACTION 0F~ LIMITING POWER; DENSITY-(FLPD) shall'be the LHGR existingL i

at a given location divided by the specified LHGR limit for that bundle type. '

]

FRACTION OF RATED THERMAL POWER f

1.15 The FRACTION OF RATED THERMAL POWER (FRTP) shall' be the measured THERMAL j

. POWER divided by the RATED THERMAL POWER.

FREQUENCY NOTATION-1.16 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

GASE0US RADWASTE TREATMENT SYSTEM 1.17' A GASE0US RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or i

holdup for the purpose of reducing the total radioactivity prior to j

release to the environment.

IDENTIFIED' LEAKAGE j

1.18 IDENTIFIED LEAKAGE shall be:

a.

Leakage into collection systems, such as pump seal or valve 1

packing leaks, that is captured and conducted to a sump or collecting tank, or

-r b.

Leakage into the containment atmosphere from sources that.are-

{

both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE.

l j

ISOLATION SYSTEM RESPONSE TIME 1.19 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when.

I the monitored parameter exceeds its isolation _ actuation setpoint at the channel sensor until the isolation valves travel to their. required l

positions. Times shall include diesel generator starting and sequence loading delays where applicable.

The response! time may be measured by any-series of sequential, overlapping or total steps such that the entire i

response time is measured.

Le 1.20 The maximum allowable primary containment leakage rate, L, shall be l

0.635 % of. primary containmenet air weight per day at the calculated peak i

containment pressure (Pa - 39.6 psig).

1 l

i l

LA SALLE - UNIT 2 1-3 Amendment No. 87 j

i i

DEFINITIONS LIMITING CONTROL R0D PATTERN 1.21 A LIMITING CONTROL R0D PATTERN shall be a pattern which results in the l

core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.22 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod.

It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.23 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, l

1.e., all relays and contacts, all trip units, solid state logic elements, etc. of a logic circuit, from sensor through and including the actuated device to verify OPERABILITY. THE LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.24 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) snall be the l

highest value of the FLPD which exists in the core.

MEMBER (S) 0F THE PUBLIC 1.25 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupation-ally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.26 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l

exists in the core.

OFFSITE DOSE CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATION MANUAL (ODCH) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological fionitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.2.F.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Semi-Annual Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

LA SALLE - UNIT 2 1-4 Amendment Nc. 87

l eo i

DEFINITIONS a

OPERABLE - OPERABILITY I

1.28 A system, subsystem, train, component or device shall be OPERABLE or have l

OPERABILITY when it is capable of performing its specified function (s),

i and when all necessary attendant instrumentation, controls,- a normal and an emergency electrical power source, cooling or seal water, lubrication j

or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of-l performing their related support function (s).

OPERATIONAL CONDITION - CONDITION 1.29 An OPERATIONAL CONDITION,. i.e., CONDITION, shall be any one inclusive combination of mode switch positicn and average reactor coolant temperature as specified in Table 1.2.

i PHYSICS TESTS i

1.30 PHYSICS -TESTS shall be those tests performed to measura the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.of the FSAR, 2) authorizec onder.the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE B0UNDARY LEAKAGE l

1.31 PRESSURE B0UNDARY LEAKAGE shall be leakage through a non-isolable fault.

in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY

[

1.32 PRIMARY CONTAINMENT INTEGRITY shall exist when:

a.

All primary containment penetrations required to be closed during accident conditions'are either-1.

Capable of being closed by an OPERABLE primary containment automatic isolation system, or i

2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except for valves that are open under administrative control as permitted by Specification 3.6.3.

7 i

b.

All primary containment equipment hatches are closed and sealed.

j c.

Each primary containment air lock is OPERABLE pursuant to Spe'cification 3.6.1.3.

d.

The primary containment leakage rates are maintained within the limits per Surveillance Requirement 4.6.1.1.b.

j e.

The suppression chamber is OPERABLE pursuant to specification 3.6.2.1.

LA SALLE - UNIT 2

.1-5 Amendment No. 87 l

~

DEFINITIONS' i

PRIMARY CONTAINMENT INTEGRITY (Continued)

{

.f.

The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

g.

Primary containment rtructural integrity'has been verified in

[

accordance with Surveillance Requirement 4.6.1.1.e.

PROCESS CONTROL PROGRAM 1.33 The' PROCESS CONTROL PROGRAM (FCP) shall contain the current formulas, l

sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing r

the disposal of solid radioactive waste.

PURGE - PURGING l.34 PURGE or PURGING shall be the controlled process of discharging air or l

gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replace-ment air or gas is required to purify the confinement.

RATED THERMAL POWER 1.35 RATED THERMAL POWER shall be a total reactor core heat transfer rate to l

the reactor coolant of 3323 MWT.

t REACTOR PROTECTION SYSTEM RESPONSE TIE 1.36 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from l

when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scran pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

REPORTABLE EVENT 1.37 A REPORTABLE EVENT, hall be any of those conditions specified in l

Section 50.73 to 10 CFR Part 50.

R0D DENSITY l.38 R0D DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

l t

f LA SALLE - UNIT 2 1-Sa Amendment No. 87

i DEFINITIONS l

SECONDARY CONTAINMENT INTEGRITY l

.l.39 SECONDARY CONTAINMENT INTEGRITY shall exist when:

l a.

All secondary containment penetrations required to be closed.

during accident conditions are either:

1.

Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed position, except as provided in Table 3.6.5.2-1 of-Specification 3.6.5.2.

b.

All secondary containment hatches and blowout panels are closed and sealed.

c.

The standby gas treatment system is OPERABLE pursuant to Specification 3.6.5.3.

d.

At least one door in each access to the secondary containment is closed.

e.

The sealing mechanism as:,ociated with each secondary containment penetration, e.g., welds, bellows or 0-rings, is OPERABLE.

f.

The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.

SHUTDOWN MARGIN 1.40 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is subtritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of h:9 est reactivity worth h

which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68*F; and xenon free.

SITE BOUNDARY i

1.41 The SITE BOUNDARY shall be that line beyond which the land is neither l

owned, nor leased, nor otherwise controlled by the licensee.

SOURCE CHECK 1.42 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.43 A STAGGERED TEST BASIS shall consist of:

l a.

A test schedule for n systems, subsystems, trains or other designated components obtained oy dividing the specified test interval into n equal subintervals.

i LA SALLE - UNIT 2 1-6 Amendment No. 87

,,1 DEFINITIONS

~ STAGGERED TEST-BASIS (Continued) b.

The testing of one system, subsystem, train or other designated i

component at the beginning of each subinterval.

THERMAL POWER i

1.44 THERMAL POWER shall be the total reactor core heat transfer rate to the-l I

reactor coolant.

TURBINE BYPASS SYSTEM RESPONSE TIME 1.45 The TURBINE BYPASS SYSTEM RESPONSE TIME shall be time interval from when' l

the turbine bypass control unit generates a turbine bypass valve flow signal until the turbine bypass valves travel to their required positions.

r The response time may be measured by any series of sequential, overlapping i

or total steps such that the entire response time is measured.

UNIDENTIFIED LEAKAGE 1.46 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

VENTILATION EXHAUST TREATMENT SYSTEM 1.47 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l

installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases-through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the l

release to the environment (such a system is not considered to have any effect on noble gas effluents).

Engineered Safety Feature (ESF) i atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.48 VENTING shall be the controlled process of discharging air or gas from a l

confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is i

not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

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l LA SALLE - UNIT 2 1-7 Amendment No. 87

TABLE 3.3.2-1 ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICABLE OPERATED BY C!!ANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL TRIP SYSTEM (b)

CMDITION-ACTION l-A.

AUTOMATIC INITIATION 1.

PRIMARY' CONTAINMENT ISOLATION a.

Reactor Vessel Water level level 3 7

2 1,2,3 20

) Low, Low, level 2 t low 2, 3 2

- 1, 2, 3 20.

h low Low Low, Level 1 1, 10 2

1,2,3 20 b.

Drywell Pressure - High 2, 7, 10 2

1, 2, 3 20 c.

Main Steam Line 1)

Radiation - High 1

2 1,2,3 21 3

2 1, 2, 3 22 2)

Pressure - Low I

2 1

23 3)

Flow - High 1

2/line(d) 1, 2, 3 21 d.

Main Steam Line Tunnel 1"),(l3f;32""I)

)

Temperature - High 1

2 e.

Main Steam Line Tunnel 1")ll3g;32""3) 3 ATemperature - High 1

2 f.

Condenser Vacuum - Low 1

2 1, 2*, 3*

21 2.

SECONDARY CONTAINMENT ISOLATION a.

Reactor Building Vent Exhaust Plenum Radiation - High 4(*"')

2 1, 2, 3 and,,

24 b.

Drywall Pressure - High 4(*"')

2 1, '2, 3

24

'c.

Reactor Vessel Water Level - Low Low, Level 2

' 4 '*"')

2 1, 2, 3, and *-

24

~ d.

Fuel Pool Vent Exhaust Radiation - High-4 ' * " ')

2 1, 2,- 3, and **

'24 1.A SALLE - UNIT 2 3/4 3-11 Amendment No. ?7

TABLE 3.3.2-1 (Continue 1)

ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICABLE OPERATED BY CHANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL TRIP SYSTEM (b)

CONDITION

~ ACTION l

3.

REACTOR WATER CLEANUP SYSTEM ISOLATION a.

A Flow - High 5

1 1,2,3 22 b.

Heat Exchanger Area 5

1/ heat 1, 2, 3 22 Temperature - High exchanger c.

Heat Exchanger Area 5

1/ heat 1, 2, 3 22 Ventilation AT - High exchanger d.

SLCS Initiation 5' "

NA 1, 2, 3 22 e.

Reactor Vessel Water Level - Low Low, level 2 5

2 1,2,3 22 4.

REAC'iOR CORE ISOLATION COOLING SYSTEM ISOLATION a.

RCIC Steam Line Flow - High 8

1 1,2,3 22 b.

RCIC Steam Supply Pressure - Low

8. 9(O 2

1,2,3 22 c.

RCIC Turbine Exhaust Diaphragm Pressure - High 8

2 1, 2, 3 22 d.

RCIC Equipment Room Temperature - High 8

1 1,2,3 22 e.

RCIC Steam Line Tunnel Temperature - High 8

1 1, 2, 3 22 f.

RCIC Steam Line Tunnel A Temperature - High 8

1 1, 2,. 3 22 g.

Drywell Pressure - High 9(*

2 1,2,3 22 h.

RCIC Equipment Room A Temperature - High 8

1 1,2,3 22 LA SALLE - UNIT 2 3/4 3-12 Amendment No. 87

+

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E TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICABLE OPERATED BY CHANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL TRIP' SYSTEM (b)

CONDITION ACTION l

5.

RHR SYSTEM STEAM CONDENSING MODE ISOLATION a.

RHR Equipment Area A Temperature - High 8

1/RHR area 1, 2, 3 22 b.

RHR Area Temperature -

High 8'

1/RHR area 1, 2, 3 22 c.

RHR Heat Exchanger Steam Supply Flow - High 8

1 1,2,3 22 6.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION f

a.

Reactor Vessel Water Level - Low, level 3 6'

2 1,2,3 25 b.

Reactor Vessel (RHR Cut-in Permissive)

Pressure - High 6

1 1,2,3 25 c.

RHR Pump Suction Flow - High 6

-. 1 1, 2, 3 25 d.

RHR Area Temperature -

High 6

1/RHR area 1, 2, 3 25 i

e.

RHR Equipment Area AT - High 6 1/RHR area 1, 2, 3 25 B.

MANUAL INITIATION 1.

Inboard Valves 1,2,5,6,7 1 group 1, 2, 3 26 2.

Outboard Valves l

1 group 1, 2, 3 26 4(**)g)5,.6,7 g

I group 1,.2, 3 and..,

26 3.

Inboard Valves 4.

Outboard Valves 4

1/ group 1, 2, 3 and..,,

26

)")

5.

Inboard Valves 3,

8,- 9 1/ valve 1, 2, 3 26 6.

Outboard Valves 3 )8, 9 1/ valve 1, 2,- 3 26 3

7.

Outboard Valve 8

1/ group 1, 2, ' 3 26 i

LA SALLE UNIT 2 3/4 3-13 Amendment No. 87l

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TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION STATEMENTS Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN ACTION 20 within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l ACTION 21 Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least H0T SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Close the affected system isolation valves within I hour and ACTION 22 declare the affected system inoperable.

ACTION 23 Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24 Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within I hour.

ACTION 25 Lock the affected system isolation valves closed within I hour and declare the affected system inoperable.

ACTION 26 -

Provided that the manual initiation function is OPERABLE for each other group valve, inboard or outboard, as applicable, in each line, restore the manual initiation function to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise:

a.

Be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or b.

Close the affected system isolation valves within the next hour and declare the affected system in operable.

IABLE NOTATIONS May be bypassed with reactor steam pressure < 1043 psig and all turbine stop valves closed.

When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

(a) Deleted.

l (b) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the channel in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

In addition for those trip systems with a design providing only one channel per trip system, the channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for required surveillance testing without placing the channel in the tripped condition provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is operable and all required actuation instrumentation for that redundant valve is OPERABLE, or place the trip system in the tripped condition.

(c) Also actuates the standby gas treatment system.

(d) A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE.

(e) Also actuates secondary containment ventilation isolation dampers per Tabl e 3.6.5.2-1.

(f) Closes only RWCU system inlet outboard valve.

LA SALLE - UNIT 2 3/4 3-14 Amendment No. 87

7 L.

TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds 1#

-6.

RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION N/A a.

Reactor Vessel Water Level - Low, Level 3 b.

Reactor Vessel (RHR Cut-in Permissive RHR Pump Suction Flow -) HighPressure - High c.

d.

RHR Area Cooler Temoerature High e.

RHR Equipment Area AT High B.

MANUAL INITIATION N/A 1.

Inboard Valves 2.

Outboard Valves 3.-

Inboard Valves 4

Outboard Valves 5.

Inboard Valves 6.

Outboard Valves 7.

Outboard Valve TABLE NOTATIONS Isolation system instrumentation response time for MSIVs only. No diesel generator delays assumed.

Radiation detectors are exempt from response time testing.

Response time shall be measured from detector output or the input of the first electronic component in the channel.

isolation system instrumentation response time specified for the Trip function actuatino the MSIVs shall be added to MSiv isolation time to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.

N/A Not Applicable.

LA SALLE - UNIT 2 3/4 3-19 Amendment No. 87

3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRI'Y shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2,* and 3.

ACTION:

Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within I hour or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE0VIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a.

At least once, per 31 days by verifying that all primary containment penetrations not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind ilanges, or deactivated automatic valves secured in position, except for valves that are open under administrative control as permitted by Specification 3.6.3.

b.

Perform required visual examinations and leakage rate testing except for primary containment air lock testing and main steam lines through the isolation valves, in accordance with and at the frequency" specified by 10 CFR 50, Appendix J, as modified by approved exemptions.

The overall integrated leakage rate acceptance criterion is s 1.0 L,.

The Type B and C combined leakage rate acceptance criterion is s 0.60 L.

However, during the first unit startup following testis.3 performe$inaccordancewith10CFR50,AppendixJ,asmodifiedby approved exemptions, the leakage rate acceptance criteria are < 0.60 L for the combined Type B and Type C tests, and < 0.75 L, for the Ty,pe A test.

See Special Test Exception 3.10.1

- Except valves, blind flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often than once per 92 days.

The provisions of Specification 4.0.2 are not applicable to the frequencies specified by 10 CFR 50, Appendix J.

LA SALLE - UNIT 2 3/4 6-1 Amendment No. 87

CONTAINMENT SYSTEMS'

i PRIMARY CONTAINMENT LEAKAGE SURVEILLANCE REOUIREMENTS (Continued) c.

By verifying each primary containment air lock OPERABLE per Specification 3.6.1.3.

d.

By verifying the suppression chamber OPERABLE per Specification 3.6.2.1.

e.

Verify primary containment structural integrity in accordance with the Inservice Inspection Program for Post Tensioning Tendons. The J

frequency shall be in accordance with the Inservice Inspection Program for Post Tensioning Tendons.

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i LA SALLE - UNIT 2 3/4 6-2 Amendment No. 87

l CONTAINMENT SYSTEMS i

PRIMARY CONTAINMENT LEAKAGE f

.f 3.6.1.2' Deleted.

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t Pages 3/4 6-3 and 3/4 6-4 Deleted i

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LA SALLE - UNIT 2 3/4 6-3 (next page is 3/4 6-5)_

Amendment No. 87 i

1

CONTAINMENT' SYSTEMS

+

SURVEILLANCE REOUIREMENTS (Continued) c.

By verifying at least two suppression chamber water level instru-mentation channels and at least 14 suppression pool water temperature instrumentation channels, 7 in each of two divisions, OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 2.

CHANNEL FUNCTIONAL TEST at least once per 31 days,.and 3.

CHANNEL CALIBRATION at least once per 18 months.

4 The suppression chamber water level and suppression pool temperature alarm setpoint shall be:

a)

High water level s +2 inches

b)

Low water level 2 -3 inches

c)

High temperature $105'F d.

By conducting drywell-to-suppression chamber bypass leak tests at least once per 18 months at.an initial differential pressure of 1.5 psi and verifying that the A//k calculated from the measured 1eakage is within the specified limit.

If any 1.5 psi leak test results in a calculated A//k >20% of the specified limit, then the test sschedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 1.5 psi leak tests result in a calculated A//k greater than the specified limit, then:

1.

A 1.5 psi leak test shall be performed at least once per 9 months until two consecutive 1.5 psi leak tests result

+

in the calculated A//k within the specified limits, and 2.

A 5 psi leak test, performed with the second consecutive successful 1.5 psi leak test, results in a calculated A//k within the specified. limit after which the above schedule of once per 18 months for only 1.5 psi leak tests may be resumed.

If any required 5

)si leak test results in a calculated A//k greater than the specified limit, tien the test schedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 5 psi leak tests reult in a calculated A//k greater than the specified limit, then a 5 psi leak test shall be performed at least once i

per 9 months until two consecutive 5 psi leak tests result in a calculated A//k within the specified limit, after which the above schedule of once per 18 months for only 1.5 psi leak tests may be resumed.

l

Level is referenced to a plant elevation of 699 feet 11 inches (See Figure B 3/4.6.2-1).

LA SALLE - UNIT 2 3/4 6-21 Amendment No. 87 r

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INTENTIONALLY LEFT BLANK

(

T LA SALLE - UNIT 2 3/4 6-22 Amendment No. 87

't CONTAINMENT SYSTEMS 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES l

LIMITING CONDITION FOR OPERATION 3.6.3 Each primary containment isolation valve and reactor instrumentation line excess flow check valve shall be OPERABLE".

. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With one or more of the primary containment isolation valves, except the reactor instrumentation line excess flow check valves, inoperable:

1.

Maintain at least one isolation valve OPERABLE in each affected penetration that is open.and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either; a)

Restore the inoperable valve (s) to OPERABLE status, or l

b)

Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolated position,' or l

c)

Isolate each affected penetration by use of at least one

[

closed manual valve or blind flange.*

2.

Otherwise, be in at least H0T SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

[

and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i b.

With one or more of the reactor instrumentation line excess flow check valves inoperable:

l 1.

Operation may continue and the provisions of Specification 3.0.3 are not applicable provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

i a)

The inoperable valve is returned to OPERABLE status, or b)

The instrument line is isolated and the associated instrument is declared inoperable.

c 2.

Otherwise, be in at least HOT' SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i

Isolation valves closed to satify these requirements may be reopened on an intermittent basis under administrative control.

- Locked or sealed closed valves may be opened on an intermittent basis under l

administrative control.

i

[

LA SALLE - UNIT 2 3/4 6-25 Amendment No. 87 f

P CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS 4.6.3.1 Each primary containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its asscciated actuator, control or power.

circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.

4.6.3.2 Each primary containment automatic isolation valve shall be demonstrated OPERABLE during COLD SHUTDOWN or REFUELING at-least once per 18 months by verifying that on a containment isolation test signal each automatic isolation valve actuates to its isolation position.

r 4.6.3.3 The isolation time of each primary containment power operated or automatic isolation valve shall be determined to be within its limit when tested pursuant to Specification 4.0.5.

4.6.3.4 Each reactor instrumentation line excess flow check valve shall be demonstrated OPERABLE at least once per 18 months by verifying that the valve checks flow.

4.6.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERABLE a.

At least once per 31 days by verifying the continuity of the explosive charge.

b.

At least once per 18 months by removing the explosive squib from at least one explosive valve such that the explosive squib in each explosive valve will be tested at least once per 90 months, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the one fired or from another batch which has been certified by having at least one of that batch successfully fired.

No explosive squib shall remain in use beyond the expiration of its shelf-life and operating-life.

4.6.3.6 At least once per 18 months:

a.

Verify leakage rate through all four main steam lines through the isolation valves is s 100 scfh when tested 'at 2 25.0 psig.

b.

Verify combined leakage rate of s 1 gpm times the total number of primary containment isolation valves through hydrostatically tested these isolation valves are tested at 1.1 P, 2 43.6 psig.,ded when lines that penetrate the primary containment is not excee i

Results shall be excluded from the combined leakage for all penetrations and seals subject to Type B and C tests.

LA SALLE - UNIT 2 3/4 6-26 Amendment No. 87

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LA SALLE UNIT 2 3/4 6-27 (next page is 3/4 6-38)

Amendment No. 87 l

j I

s.--

i CONTAINMENT SYSTEMS SURVE1LLANCE REOUIREMENTS (Continued) i 4.6.4.2 The manual isolation valves on both' sides of an inoperable and/or open-'

i suppression chamber-drywell vacuum breaker shall be verified to be closed at least once per 7 days.

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LA SALLE - UNIT 2 3/4 6-39 Amendment No. 87 i

I

CONTAINMENT SYSTEMS i

SURVEILLANCE REQUIREMENTS (Continued)

)

Af ter each complete or partial replacement of a HEPA filter bank by e.

verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm i 10%.

i

(

f.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in place in accordance with ANSI N510-1975 while operating the system at a flow rate of 4000 cfm i 10%.

i i

LA SALLE - UNIT 2 3/4 6-45

L

-CONTAINMENT SYSTEB.i

~

3/4.6.6 ' PRIMARY' CONTAINMENT ATMOSPHERE CONTROL i

DRYWELL AND SUPPRESSION CHAMBER HYDR 0 GEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION

.3.6.6.1.Twoindegendentdrywellandsuppressionchamberhydrogenrecombiner systems'shall be PERABLE APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

'With one drywell and/or suppression chamber hydrogen recombiner system I

inoperable. restore the inoperable system to OPEMBLE status within 30 days or-be in at le,ast HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE0VIREMENTS 4.6.6.1 Each drywell and suppression chamber hydrogen recombiner system shall be demonstrated OPERABLE:

-i

'I a.

- At least once per 92 days by cycling each flow control valve and recirculation valve through at least one complete cycle of full travel.

b.

At least once per 18 months by verifying, during a recombiner system 1

functional test:

1.

That the heaters are OPERABLE by determining % from the other that the current in each phase-differs by % of the value observed in the original less than or equal to 5 phases and is within 5 acceptance test, corrected for line voltage differences.

2.

That the reaction chamber gas temperature increases to 1200 25'F within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , c.

At.least once per 18 months by:

1.

Performing a CHANNEL CALIBRATION of all recombiner opotating instrumentation and control circuits, i

2.

Verifying the integrity of all heater electrical circuits by performing a resistance to ground test within 30 minutes following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to l

100,000 ohms.

t i

LA SALLE - UNIT 2 3/4 6-46 Amendment No. 87 f

e 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINM_ENT 3 /4. 6.1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage path; and associated leak rates assumed in the accident analyses.

This restriction, in conjunction with the leakage rate limitation,-will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

t The structural integrity of the primary containment is ensured by the successful completion of the Inservice Inspection Program for Post Tensioning l

Tendons and by associated visual inspections of the steel liner and penetrations for evidence of deterioration or breach of integrity.

This

[

ensures that the structural integrity of the primary containment will be maintained in accordance with the provisions of the Primary Containment Ter. don Surveillance Program.

Testing and frequency are consistent with the recommendations of Regulatory Guide 1.35, Revision 3, except that the Unit 1 and 2 primary containments shall be treated as twin containments even though the Initial Structural Integrity Tests were not within 2 years of each other.

PRIMARY CONTAINMENT INTEGRITY is maintained by limiting overall i

integrated leakage to s 1.0 L and the Type B and C combined leakage rate acceptancecriterioniss0.66L,.

Prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test, the combined Type B and L leakage must be < 0.60 L, and the overall Type A leakage must be < 0.75 L, when a Type A test is scheduled.

Compliance with this LC0 will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

The maximum allowable leakage rate for the primary containment (L,) is 0.635% by weight of the containment atmosphere per day at the v.lculated maximum peak containment pressure (P,) of 39.6 psig.

Individual leakage rates specified for the primary containment air lock, main steam lines through the isolation valves, and valves in hydrostatically tested lines are addressed in LC0 3.6.1.3, and Surveillance Requirement 4.6.3.6.

LA SALLE - UNIT 2 B 3/4 6-1 Amendment No. 87

?

I 3/4.6 CONTAINMENT SYSTEMS BASES i

3 /4. 6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY (Continued)-

Surveillance Requirement 4.6.1.1.b maintains PRIMARY CONTAINMENT INTEGRITY by requiring compliance with the visual examinations and leakage f

g' rate test requirements of 10 CFR 50, Appendix J, as modified by approved i

exemptions.

Failure to meet air lock leakage testing (4.6.1.3) or main steam isolation valve leakage (4.6.3.6.a) does not necessarily result in a failure of this Surveillance Requirement, 4.6.1.1.b. -The impact of the failure to.

meet these Surveillance Requirements 4.6.1.3 and 4.6.2.1.b must be evaluated against the Type A, B, and C acceptance criteria of_10 CFR 50, Appendix J, as i

mMified by approved exemptions. The leakage limits for main steam lines tbrough the isolation valves and leakage test results of Surveillance Requirement 4.6.3.6.a are not included in the total sum of Type B and C tests (approved exemption). As-left leakage prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test is required to be

< 0.60 L, for combined Type B and C leakage, and < 0.75 L, for overall Type A leakage. At all other times between rec.uired Type A tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,.

At s 1.0 L, the offsite dose consequences are bounded by the assumptions of the' safety analysis.

The combined Type B and C leakage remains as s 0.60 L, between j

scheduled tests, in accordance with Appendix J.

The frequency is requi Td by 10 CFR 50, Appendix J, as modified by l

[

approved exemptions. Thus, s.0.2 (which allows Frequency extensions) does not (I

apply to Surveillance Requirement 4.6.1.1.b.

3/4.6.1.2 DELETED 3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS j

The limitation on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY i

and the primary containment leakage rate given in Specification 3/4.6.1.1.

The specification makes allowances for the fact that there may be long periods of time when the air locks will be in a closed and secured position during j

reactor operation.

Only one closed door in each air lock is required to maintain the integrity of the containment.

3/4.6.1.4 MSIV LEAKAGE CONTROL SYSTEM t

Calculated doses resulting from the maximum leakage allowcnce for the main steamline isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR 100 guidelines provided the main steam line sys+.em from the isolation valves up to and including the turbine condenser remains intact.

Operating experience has indicated that degradation has LA SALLE - UNIT 2 B 3/4 6-2 Amendment No. 87 j

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l 3/4.6 CONTAINMENT SYSTEMS l

BASES

.3/4.6.1 PRIMARY CONTAINMENT-MSIV LEAKAGE CONTROL SYSTEM (Continued) occasionally occurred in the leak tightness of the MSIV's such that the specified leakage requirements have not always been maintained continuously.

The requirement for the leakage control system.will reduce the untreated i

leakage from the isolation valves when isolation of the primary-system and containment is required.

3/4.6.1.5 DELETED f

3/4.6.1.6 DRYWEll AND SUPPRESSION CHAMBER INTERNAL PRESSURE The limitation on drywell and suppression chamber internal pressure ensure that the containment peak pressure of 39.6 psig does not exceed the design pressure of 45 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of 5 psid. The limit of 2.0 psig for initial positive primary 1

containment pressure will-limit the total pressure to 39.6 psig which is less that the design pressure and is consistent with the accident analysis.

3/4.6.1.7 DRYWELL AVERAGE AIR TEMPERAM The limitation on drywell average air temperature ensures that the l

containment peak air temperature does not exceed the design temperature of 340*F during LOCA conditions and is consistent with the accident analysis.

3/4.6.1.8 DRYWELL AND SUPPRESSION CHAMBER PURGE SYSTEM The drywell and suppression chamber purge supply and exhaust isolation valves are required to be closed during plant operation except as required for

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inerting, de-inerting and pressure control. These valves have been i

demonstrated capable of closing during a LOCA or steam line break accider.t from the full open position.

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.LA SALLE - UNIT 2 B 3/4 6-2a Amendment No. 87 i

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L r

r CONTAINMENT SYSTEMS BASES' DEPRESSURIZATION SYSTEMS (Continued)

Because of the large volume and thermal capability of the suppression L

pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.

By-requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely L

followed so that appropriate action can be taken. The requirement for the

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external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered.

In addition to the limits on temperature of the suppression chamber pon1 water, operating procedures define the action to be taken in the event of safety-relief valve inadvertently opens or sticks open. As a minimum this action shall include: (I) use of all available means to close the valve, (2) initiate suppression pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve to assure mixing and uniformity of energy insertion to the pool 3/4.6.3 PRIMARY CONTAINMENT ' SOLATION VALVES Primary Containment Isolation Valves (PCIVs) form a part of the primary containment boundary. The PCIV safety function is related to control primary containment leakage rates during accidents or other conditions to limit the untreated release of radioactive materials from the containment in excess of the design limits.

The automatic isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal.

The valves covered by this specification are listed with their associated stoke times, and other design information for lines penetrating the Primary Containment, in UFSAR Section 6.2.

The normally closed isolatica valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact.

Main steam lines through the isolation valves and hydrostatically tested valves must meet alternative leakage rate requirements. Other PCIV leakage rates are addressed by specification 3/4.6.1.I, "PRIMAPY-CONTAINMENT INTEGRITY".

UFSAR Section 6.2 also describes special 1sakage test requirements and exemptions.

LA SALLE - UNIT 2 B 3/4 6-4 Amendment No. 87

a CONTAINMENT SYST,ggi BASES PRIMARY CONTAINMENT ISOLATION VALVES (Continued)

This specification provides assurance that the PCIVs will perform their designed safety functions to control leakage from the primary containment during accidents.

The opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following i

censiderations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude assess to close the valves and that this action will prevent the release of radioactivity outside the primary containment.

Surveillance Requirement 4.6.3.6.a verifies leakage through all four main steam lines is s 100 scfh when tested at ;t P, (25.0 psig). The transient and accident analyses are based on leakage at the specified leakage rate. The leakage rate for main steam lines through the isolation valves must be verified to be in accordance with the leakage test requirements of 10 CFR 50, Appendix J, as modified by approved exemptions. A Note has been added to this Surveillance Requirement requiring the results to be excluded from the total r

of Type B and Type C tests. This ensures that leakage rate for main steam lines through the isolation valves is properly accounted for in accordance with an approved exemption. The frequency is "at least once per 18 months" in 1

accordance with an approved exemption.

Surveillance Requirement 4.6.3.6 b test of hydrostatically tested lines provides assurance that the assumptions of UFSAR Section 6.2 are met. The combined leakage rates must be demonstrated in accordance with the leakage rate test at a frequency of "at least once per 18 months". A Note has been added to this Surveillance Requirement requiring the results to be excluded the total of Type B and Type C tests. This is in accordance with 10 CFR 50, Appendix J, and approved exemptions.

l 2/L6.4 VACUUM RELIEF Vacuum relief breakers are provided to equalize the pressure between the suppression chamber and drywell. This system will maintain the structural i

integrity of the primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and tha drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident. There are four valves to provide redundancy so that operation may continue for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with one vacuum breaker inoperable provided that the manual isolation valves on each side are in the closed position.

LA SALLE - UNIT 2 B 3/4 6-4 a Amendment No. 87

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ML20081D655

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