Text

Amends 70 & 54 to Licenses NPF-11 & NPF-18,respectively. Revising Tech Spec 3/4.2 Re Power Distribution Limits to Replace Values of cycle-specific Parameter Limits W/Ref to Core Operating Limits Rept
	ML20033H018
Person / Time
Site:	LaSalle ; (NPF-11-A-070, NPF-11-A-70, NPF-18-A-054, NPF-18-A-54)
Issue date:	12/18/1989
From:	Jocelyn Craig; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20033H019	List: ML20033H018;
References
	NUDOCS 9004160060
	Download: ML20033H018 (60)
	v • d • e

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WITIOSTATIS

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4 NUCLEAR REGULATORY COMMISSION

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COMMONWEALTH EDISON COMPANY DOCKET N0. 50-373 l

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tA$ALLE COUNTY STATION. UNIT 1 l

l AMENDMENT TO FACILITY OPERATING LICENSE i

t Amendment No. 70 l

License No. NPF.11 t

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1.

The Nuclear Rtgulatory Comission (the Colvission or the NRC) has found j

that.

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l A.

The application for amenonent filed by the Comonwealth Edison Company l

(the licensee), dated August 18, 1989 and supplemented September 13, 1989 con. plies with the standards and requirements of the Atoric Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter 1; B.

The facility will operate in conforn.ity with the application, the provisions of the Act, and the regulations of the Comission; l

C.

There is ressorable assurance: (i)thattheactivitiesauthorizedby this amencment can be conducted without endangering the health and safety of the public, and (11) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR j

Chapter 1; D.

The issuance of this amendment will not be inimical to the comon ceferse and security or to the health and safety of the puts 11c; and E.

The issuance of this anendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been j

satisfied.

2, Accordingly, the license is amended by changes to the Technical specifica-tions as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:

(2) Technical Specifications and_ Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendnent No. 70, 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 Environrental Protection Plan.

9004160060 893230 i

hDR ADOCK 05000373 l

PDC

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3.

This amenoment is effec'tive upon date of issuance.

f FOR THE NUCLEAR REGULATORY COMMIS$10N f

f l

r John W. Craig, Director i

llroject Directorate !!!-2 Division of Reactor Projects i

IV, Y and Special Projects j

i

Enclosure:

Changes to the Technica'1 Specifications i

Date of Issuance: December 18, 1989 i

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[WCLOSURE TO LICth5E AWIWt*ENT WO,70

_rACILITY.0f tRATING LICtW5E NO. kPF-11 00 Crit wo, !0 373 Replace the following pages of the Appendix 'A' Technical 5pecifications with the enclosed pages. The revised pages are identified by amendment nuntaer and contain a vertical line indicating the area of chappe.

REMOVE INSERT I

IV lY X11 X11 XIX XIX XIXn 12 12 13 13 1-4 1-4 1-5 15 1-6 16 17 17 3/4 2 1 3/4 2-1 3/4 2 2 3/4 2 2a 3/4 2 2b 3/4 2-3 3/4 2 2 3/4 2-4 3/4 2 3 3/4 2 Aa 3/4 2 4 3/4 2 5 3/4 245a 3/4 2-6 3/4 2 7 3/4 2-5 3/4 3 53 3/4 3 53 3/4 3 53a 3/4 3 53a B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2 3 8 3/4 2-2 8 3/4 2 36 8 3/4 2 3 B 3/4 2 3b 8 3/4 2 4 B 3/4 2 4 B 3/4 2 5 8 3/4 2 5 03/426 8 3/4 2-6 54 5-4 6-24 6 24 6-25 6 25

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 ACT10N............................................................

11 1.2 AVERAGE PLANAR EXP0$VRE...........................................

11 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE........................

11 1.4 CHANNEL CAtlBr.AT10N...............................................

11 1.5 CHANNEL CHECK.....................................................

11 1.6 CHANNEL FUNCTIONAL TE$1...........................................

11 1.7 CORE ALTERA110N.....s.............................................

12

1. 6 CORE OPERATING LIMITS REP 0RT......................................

12 1.9 CRITICAL POWER RAT10..............................................

12 1.10 00$E EQUIVALENT l 131.............................................

12 1.11 E AVERAGE DI$1NTEGRAT10N ENERGY...................................

12 1.12 EMERGENCY CORE COOLING $Y$ TEM (ECCS) RESPONSI TIME................

12 1.13 END OF CYCLE RECIRCULATION PUMP TRIP $Y$ TEM RESPON$E TIME.........

12 1.14 FRACTION OF LIMITING POWER DEN $1TY................................

13 1.15 FRACTION OF RATED THERMAL P0WER...................................

13 1.16 FREQUENCY NOTAT10N.................................................

13 1.17 GA$E005 RADWA$1E TREATMENT SYSTEM.................................

13 1.18 IDENTIFIED LEAKAGE................................................

13 1.19 !$0LAT10N $YSTEM RESPONSE TIME....................................

13 1.20 LIMITING CONTROL R0D PATTERN......................................

13.

1.21 LINEAR HEAT GENERATION RATE.......................................

14 1.22 LOGIC $YSTEM FUNCTIONAL TEST......................................

Ir4 1.23 MAXIMUM FRACTION OF LIMITING POWER DEN 51TY.......................,

14 1.24 MINIMUM CDITICAL POWER RAT10......................................

1-4 1.25 0FFSITE OOSE CALCULATION MANUAL...................................

1-4 LA SALLE UNIT 1 1

Amendment No.7D L

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INDEX DEFINITIONS l

1(CT10N i

DEFINITIONS (Continued)

PAGE

1. 2 6 O P E R AB L E O PE R AB I L I T Y............................................

14 1.27 OPERATIONAL CONDITION CONDIT10N.................................

14 1.2B PHYSIC $ TESTS.....................................................

14 1,29 PRES $URE BOUNDARY LEAKAGE.........................................

15 1.30 PRIMARY CONTAINMENT INTEGRITY.....................................

15 1.31 PROCESS CONTROL PROGRAM...........................................

15 1.32 PURGE PURGING...................................................

15 1.33 RATED THERMAL P0WER...............................................

15 1.34 REACTOR PROTECTION SYSTEM RESPONSETIME...........................

15 1.35 REPORTABLE EVENT..................................................

16 1.36 R00 0ENSITY.......................................................

16 1.37 SECONDARY CONTAINMENT INTEGRITY...................................

1-6 1.3B SHUTDOWN MARGIN...................................................

16 2.39 SOLIDIFICATION....................................................

16 1.40 SOURCE CHECK......................................................

17 1,41 STAGGERED TEST BAS15..............................................

1-7 1.42 THERMAL P0WER.....................................................

17 1.43 TURBINE BYPASS RESPONSE T1ME......................................

17 1,44 UNIDENTIFIED LEAKAGE..............................................

17 1.45 VENTILATION EXHAUST TREATMENT SYSTEM..............................

1-7 r

1.46 VENTING................,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,

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i LA SALLE - UNIT 1 11 Amendment No.70

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...................................................

3/4 0 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN..............................................

3/4 1 1 3/4.1.2 REACTIVITY ANDMALIES.........................................

3/4 1 2 3/4.1.3 CONTROL RODS Control Rod Operability......................................

3/4 1 3 t

Control Rod Maximum Scram Insertion Times.................'...

3/4 1 6 Control Rod Average Scram Insertion Times....................

3/4 1-7 Four Control Rod Group Scram Insertion Times.................

3/4 1 8 Control Rod Scram Accumulators...............................

3/4 1 9 Control Rod Drive Coupling...................................

3/4 1 11 Control Rod Position

.,dication..............................

3/4 1-13 Control Rod Drive Housing Support............................

3/4 1 15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer..........................................

3/4 1-16 Rod Sequence Control System..................................

3/4 1 17 Rod Block Monitor............................................

3/4 1 1B 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM................................

3/4 1-19 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM...........................

3/4 1 23 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERAi!ONRATE...................

3/4 2-1 3/4.2.2 A P RM S E T P0 l NT S...............................................

3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RAT10.................................

3/4 2-3 j

3/4.2.4 LINEAR HEAT GENERATION RATE..................................

3/4 2-S I

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LA SALLE'- UNIT 1 IV Amendment No. 70 l

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i INDEX 1

r BASES I

t SECTION PAGE 2

l 3 /4. 0 APPLICABILITY...................................................

B 3/4 0 1 t

t 3/4.1 REACT!v!TY CONTROL SYSTEMS 3/4.1.1 SHUT DOWN MARG I N.........................................

8 3/4 1 1 l

t 3/4.1.2 REACTIVITY AN0MALIES....................................

8 3/4 1 1 i

3/4.1.3 CONTROL R0DS............................................

B 3/4 1 2 l

3/4.1.4 CONTROL ROD PROGRAM CONTR0LS............................

B 3/4 1 3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM...........................

B 3/4 1 4 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM......................

B 3/4 1 5 i

3/4.2 power O!STRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR NEAT GENERATION RATE..............

B 3/4 2-3 3/4.2.2 AFRM SETP0lNTS..........................................

8 3/4 2 2 t

3/4.2.3 MINIMUM CRITICAL POWER RAT10............................

B 3/4 2 ?

4 3/4.2.4 LINEAR HEAT GENERATION RATE.............................

B 3/4 2 0 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...............B 3/4 3 3 3/4.3.2 ISOLATION ACTUATION 1NSTRUMENTATION.....................

B 3/4 3 2 I

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.........................................

8 3/4 3 2 t

3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.......

B 3/4 3 3

'i 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION l

INSTRUMENTAT10N.........................................

B 3/4 3 3 i

3/4.3.6 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION............

B 3/4 3 4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.....................

B 3/4 3-4 t

Seismic Monitoring Instrumentation......................

B 3/4 3 4 I

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LA SALLE - UNIT 1 XII Amendment No. 70 f

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E LIST OF FIGURES

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flGDRE PAGE

}

3.1.5 1

$0DlVM PENTABORATE $0LUT10N TEMPERATURE /

CONCENTRATION REQUIREMENTS.........................

3/4 1 21 i

l 3.1.5 2 SODIUMPENTABORATE(Na:Bn0a*10H0) l VOLUME / CONCENTRATION REQUIREMENTS..................

3/4 1 22 f

d 3.4.1.5 1 CORE TNERMAL POWER (% OF RATED) VER5US TOTAL CORE F LOW (E OF RATED)............................

3/4 4 4c 3.4.6.1 1 MINIMUM REACTOR VESSEL METAL TEMPERATURE V5. REACTOR VESSEL PRESSURE........................

3/4 4 18 4.7 1 SAMPLE PLAN 2) FOR $NUBBER FUNCTl'ONAL TEST.........3/4 7 32 B 3/4 3 1 REACTOR VESSEL WATER LEVEL.........................

B 3/4 3 7 B 3/4.4.6 1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV at 1/4 T AS A FUNCTION OF SERVICE LIFE..........)............B 3/4 4 7 B 3/4.6.2 1 SUPPRESSION POOL LEVEL SETPOINTS...................

B 3/4 f 3a 5.1. P 1 EXCLU510N AREA AND SITE BOUNDARY FOR GAS AND LIQUID E F FLUENTS................... EOUS 52 5.1.2 1 LOW POPULATION ZONE................................

53 6.1 1 DELETED............................................

6 11 6.12 DELETED............................................

6 12 6.1*3 MINIMUM SHIFT CREW COMP 051T!0N.....................

6 13

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LA SALLE

UNIT 1 XIX Amendment No.70

pylWITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of

' fuel, sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.8 The CORE OPERATING LIMITS REPORT is the unit specific document that provides core operating limits for the. current operating reload cycle.

These cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.A.6.

Plant operation within these operating limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the I

assembly which is calculated by application of the GEXL correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I 131 1.10 DOSE EQUlvALENT 1 131 shall be that concentration of I-131 l

microcuries/ gram,whichalonewouldproducethesamethyrolddoseasthe quantity and isotopic mixture of I 131, I-132, 1-133, 1-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of T10 14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

E-AVERAGE DISINTEGRATION ENERGY 1.11 I shall be the average, weighted in proportion to the concentration of i

each radionuclide in the reactor coolant at the time of sampling, MeV of the sum of the average beta and gamma energies per disintegration, in for isotopes, with half lives greater than 15 minutes, making up at least 9% of the total non iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.12 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be interval from when the monitored parameter exceeds its ECCS actuation j

setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values etc.

TimesshallincludedieselgeneratorstartingandsequenceloadIngdelays w, Tere applicable.

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

END OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.13 The END OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME sha l

that time interval to energitation of the recirculation pump circuit LA SALLE UNIT 1 1-2 Amendment No. 70 i

_a

o OEFINITIONS (WD OF CYCLE RECIRCULATION PUMP TRIP $YSTEM 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 DEN 51TY 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 OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL l

POWER divided by the RATED THERMAL POWER.

FREQUENCY NOTATION 1.16 The FREQUENCY NOTATION specified for the performance of Surveillance l

Requirements shall correspond to the intervals defined in Table 1.1.

GASE0US RADWASTE TREATMENT $YSTEM 1.17 A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and I

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:

I Leakage into collection systems, such as pump seal or valve a.

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 specifically located and known either not to interfere with the 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 l

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 applicable.

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

LIMITING CONTROL ROD PATTERN 1.20 A LIMITING CONTROL ROD 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.

LA SALLE UNIT 1 1-3 Amendment No. 70

o DEFINITf0NS LINEAR HEAT GENERATION RATE 1.21 LINEAR HEAT GENERAT]ON RATE (LHGR) shall be the heat generation per unit l

1ength of fuel rod.

It is the integral of the heat flux over the heat l

i transfer area associated with the unit length.

l LOGIC SYSTEM FUNCTIONAL TEST 1.22 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components.

l 1.e., all relays a etc. of a logic ci.nd contacts, all trip units, solid state logic elements, rcuit, from senscar 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.

MAXIMUMFRACTIONpFllMITINGPOWERDENSITY 1.25 The MAXIMU" f M CTION OF LIMIT]NG POWER DENSITY (MFLPD) shall be the highest l

value of the TLPD which exist! in the core.

MINIMUM CRITICAL POWER RATIO 1.24 The M]NIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l

exists in the core.

0FFSITE DOSE CALCULATION MANUAL 1.25 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology

[

i and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid vffluent monitoring alarm / trip setpoints.

OoERABLE - OPERABILITY 1.26 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 l

an emergency electrical power source, cooling or s,eal 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 performing their related support function (s).

OPERATIONAL CONDITION - CONDITION 1.27 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive l

combination of mode switch position and average reactor coolant temperature l

as specified in Table 1.2.

l PHYSICS TESTS l

1.28 PHYSICS TESTS shall be those tests performed to measure the fundamental 1

nuclear characteristics of the reactor core and related instrumentation I

and 1) described in Chapter 14'of the FSAR, 2) authorized under the j

provisiens of 10 CFR 50.59, or 3) otherwise approved by the Commission.

LA SALLE UNIT 1 1-4 Amendment No. 70

DEFINITIONS 1

PRESSURE BOUNDARY LEAKAGE 1,29 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 1.30 PRIMARY CONTAINMENT INTEGRITY shall exist when:

l All primary containment penetrations required to be closed a.

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 as provided in Table 3.6.3-1 of Specification, 3.6.3.

b.

All primary containment equipment hatches are closed and sealed.

Each primary containment air lock is OPERABLE pursuant to c.

Specification 3.6.1.3.

d.

Tha primary containment leak &ge rates are within the limits of Specification 3.6.1.2.

The suppression chamber is OPERABLE pursuant to Specification e.

3.6.2.1.

f.

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

PROCESS CONTROL PROGRAM 1.31 The PROCESS CONTROL PROGRAM (PCP) shall contain the sampling, analysis, l

and formulation determination by which SOLIDIFICATION of radioactive wastes from liquid systems is assured.

PURGE - PURGING 1,32 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 replacem-ent air or gas is required to purify the confinement.

RATED THERMAL POWER 1.33 RATED THERMAL POWER shall be a total reac'or core heat transfer rate tol t

the reactor coolant of 3323 MWT.

REACTOR PROTECTION SYSTEM RESPONSE TIME I

1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval f 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 reries of sequential, overlapping or total steps such that the entire response time is measured, t

LA SALLE UNIT 1 1-5 Amendment No.70 j

I

DEFINITIONS

_ REPORTABLE EVENT 1.35 A REPORTABLE EVENT shall be any of those conditions specified in l

Section 50.73 to 10 CFR Part 50.

ROD DENSITY 1l36 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% ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.37 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, The standby gas treatment system is OPERABLE pursuant to c.

Specification 3.6.5.3.

d.

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

The sealing mechanism associated with each secondary containment e.

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 1.38 SHUTDOWN MARGIN shall be the amount of reac'tivity by which the resctor is l

subcritical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assuined to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.

SOLIDIFICATION 1.39 SOLIDIFICATIOi shall be the conversion of radioactive wastes from liquid l

systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

LA SALLE UNIT 1 1-6 Amendment No. 70

i DEFINITIONS SOURCE CHECK 1.40 A SOURCE CHECK shall be the qualitative assessment of channel response l

when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.41 A STAGGERED TEST BASIS shall consist of:

l a..

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

b.

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

THERMAL POWER 1.42 THERMAL POWER shall be the total reactor core heat transfer rate to the l

reactor coolant.

TURBINE BYPASS SYSTEM RESPONSE TIME 1.43 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 mehsured by any series of sequential, overlapping or total steps such that the entire response time is measured.

UNIDENTIFIED LEAKAGE 1,44 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

I VENTILATION EXHAUST TREATMENT SYSTEM 1.45 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and l

installed to reduce gaseous radiciodine or radioactive material in particu-late 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 release to the environment (such a system is not considered to have any effect on noble gas effluents).

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

VENTING, 1.46 VENTING shall be the controlled process of discharging air or gas from a l

I confinement to maintain temperature, pressure, humidity, concentration or 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 SALLE UNIT 1 1-7 Amendment No. 70

e 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION I

3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT.

l APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 2 n of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits specified in the CORE OPERATING LIMITS REPORT.

l a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is c.

operating with a LIMITING CONTROL R0D PATTERN for APLHGR.

i i

LA SALLE - UNIT 1 3/4 P-1 Amendment No. 70 i

l POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power-upscale scram trip setpoint (S) and flow biased simulated thermal power-upscale control rod block trip setpoint (Sgg) shall be established according to the following relationships:

a.

Two Recirculation Loop Operation S less than or equal to (0.58W + 59%)T S

less than or equal to (0.58W + 47%)T RB b.

Single Recirculation Loop Operation S less than or equal to (0.58W + 54.3%)T S

less than or equal to (0.58W + 42.3%)T RB where:

5 and Seg are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 108.5 million 1bs/hr.

T = Lowest value of the ratio of FRACT10N OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY or the value 1.0.

T is always less than or equal to 1.0.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 2 W of RATED THERMAL POWER.

ACTION:

With the APRM flow biased simulated thermal power-upscale scram trip setpoint and/or the flow biased simulated thermal power upscale control rod block trip setpoint set less conservatively than 5 or S as above determined, initiate correctiveactionwithin15minutesandrestNe,5and/ ors to within the 4

required limits

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THEuMAL POWER to Tess than 25's of RD RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1 SURVEILLANCE RE0VIREMENTS 4.2.2 The FRTP and the MFLPD for each class of fuel shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power upscale scram and control rod block trip setpoint verified to be within the above limits or adjusted, as r,equired:

t 1

a.

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

. b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.

with MFLPD greater than or equal to FRTP.

"With MFLPD greater than the FRTP up to 90% of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100% times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10% of RATED THERMAL POWER, and a notice of the adjustment is posted on the reactor control panel.

LA SALLE - UNIT 1 3/4 2-2 Amendment No,70 l

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit specified in the CORE OPERATING LIMITS REPORT, 9

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERHAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION With MCPR less than the applicable MCPR limit as determined for one of the a.

conditions specified in the CORE OPERATING LIMITS REPORT.

l 1.

Initiate corrective action within 15 minutes, and 2.

Restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 3.

Otherwise, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

When operating in a condition not specified in the CORE OPERATING LIMITS REPORT, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

LA SALLE - UNIT 1 3/4 2-3 Amendment No. 70

e t

POWER DISTRIBUTION LIMITS (Continued) 3/4.2.3 MINIMUM CRITICAL POWER RATIO SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

t,y,

0.86 prior to performance of the initial scram time measurements a.

for the cycle in accordance with Specification 4.1.3.2, or b.

1,y, determined within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT.

l a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is oper-c.

ating with a LIMITING CONTROL ROD PATTERN for MCPR.

J i

j LA SALLE - UNIT 1 3/4 2-4 Amendment No. 70 k

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT.

l APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

t With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS i

4.2.4 LHGR's shall be determined to be equal to or less than the limit:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.

on a LIMITING CONTROL ROD PATTERN for LHGR.

4 i

e LA SALLE - UNIT 1 3/4 2-5 Amendment No. 70

TABLE 3.3.6-2 2,

CONTROL ROD WilHURAWAL BLOCK INSTRUMENTATION SEIPOINTS E

1 RIP FUNCTION 1 RIP SEIPOINT ALLOWA8LE VALUE l.

ROD BLOCK MONITOR E

a.

Upscale

~*

The Rod Block Monitor Upscale Setpoints shall be established according to the relationships specified in the CORE OPERATING LIMITS REPORT.

b.

Inoperative N.A.

c.

Downscale M.A.

3 5% of RATED THERMAL POWER

> 3% of RATED THERMAL POWER 2.

APRM a.

Flow Biased Simulated Thermal Power-Upscale 1)

Two Recirculation

< 0.58 W + 47%*

< 0.58 W + 50%"

Loop Operation R

2)

Single Recirculation

~

Loop Operation

< 0.58 W + 42.3%

< 0.58 W + 45.3%*

T b.

Inoperative N.A.

N.A.

O c.

Downscale

> 5% of RATED THERMAL POWER

.> 3% of RATED THERMAL POWER d.

Neutron Flux-High

[12%ofRATEDTHERMALPOWER

{14%ofRATEDTHERMALPOWER 3.

SOURCE RANGE MONITORS a.

Detector not full in N. A.

0 N.A.

b.

Upscale

< 2 x 10 cps

< 5 x 10 cps c.

Inoperative N.A.

N.A.

d.

Downscale 2 0.7 cps 1 0.5 cps 4.

INTERMEDIATE' RANGE MONITORS a.

Detector not full in N. A.

M.A.

b.

Upscale

< 108/125 of full scale

< 110/125 of full scale

-c.

Inoperative N.A.

N.A.

d.

Downscale 1 5/125 of full scale 1 3/125 of full scale 5

E mm.,

._w

,,__--n

_-.-----=,.-m..

am-_

c.

-p eu_p

.a

.~

TABLE 3.3.6-2 (Continued) 2 N

CONTROL R00 WITH0RAWAL BLOCK INSTRUNENTATION SETPOINTS u,

l, TRIP FUNCTION TRIP SEIPOINT ALLOWABLE VALUE E

5.

SCRAM DISCHARGE VOLUME a.

Water Level-High b.

Scram Discharge Volume

-< 765' 5's"

< 765' 5%"

Switch in Bypass N.A.

N.A.

6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale

< 108/125 of full scale

< 111/125 of full scale b.

Inoperative N.A.

N.A.

c.

Comparator,

< 10% flow deviation

< 11% flow deviation R

s T

E, 1

E:

E

$n

-N y

The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W).

The trip setting of this function must be maintained in accordance with Specification 3.2.2.

t 3/4.2 POWER DISTRIBUTION LIMITS i

i BASES The specifications of this section assure that the peak cladding temperature followin the2200gFlimitspecifiedin10CFR50.46.the postulated design basis loss-of-coolant acc t

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.

The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor.

However, the current General Electric (GE) ca.lculational models (SAFER /GESTR described in Reference 3), which are consistent with the requirements of Appendix K to 10 CFR 50, have established that APLHGR values are not expected to be limited by LOCA/ECCS considerations.

APLHGR limits are still required, however, to assure that fuel rod mechanical integrity is maintained.

They are specified for all resident fuel types in the Core Operating Limit Report based on the fuel thermal-mechanical design analysis.

LA SALLE UNIT I B 3/4 2-1 Amendment No. 70

-.,r.

r

POWER DISTRIBUTION SYSTEMS BASES 3/4.2.2 APRM SETPOINT5 The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The flow biased simulated thermal power-upscale scram setting and control rod block functions of the APRM instruments for both two recirculation loop operation and single recirculation loop operation must be adjusted to ensure that the MCPR does not become less than the fuel cladding safety limit or that > 1% plastic strain does not occur in the degraded situation.

The scram settings and rod block settings are adjusted in accordance with the for-mula in this specification when the combination of THERMAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

3/4.2.3 MINIMUM CRITICAL POWER RATIO j

The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the l

initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in i

Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduc-tien in CRITICAL POWER RATIO (CPR).

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in the CORE I

OPERATING LIMITS REPORT.

i Analyses have been performed to determine the effects on CRITI' CAL POWER kATIO (CPR) during a transient assuming that certain equipment is out of service.

A detailed description of the analyses is provided in Reference 5.

The anal-yses performed assumed a single failure only and established the licensing i

bases to allow continuous plant operation with the analyzed equipment out of l

service.

The following single equipment failures are included as part of the transient analyses input assumptions:

1) main turbine bypass system out of service, l

2) recirculation pump trip system out of service, LA SALLE UNIT 1 B 3/4 2-2 Amendment No. 70 l

l

,e POWER DISTRIBUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued) s

3) safety / relief valve (S/RV) out of service and

4) feedwater heater out of service (correspon, ding to a 100 degree F reduction in feedwater temperature).

For the main turbine bypass and recirculation pump trip systems, specific cycle-independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for i

Operation (LCO) values are established to allow continuous plant operation with these systems out of service.

A bounding end-of-cycle exposure condition was used to develop nuclear input to the transient analysis model.- The bounding exposure condition assumes a more top peaked axial power dir,tribution than the nominal power shape, thus yielding a bounding scram response with reasonable conservatisms for the MCPR LCO values in future cycles.

The MCPR LCO values shown in the CORE OPERATING LIMITS REPORT for the main turbine bypass and recirculation pump trip systems out of service are valid provided that these r

limits bound the cycle specific results.

The analysis for main turbine bypass and recirculation pump trip systems inoperable allows operation with either system inoperable, but not both at the same time.

For operation with the feedwater' heater out of' service,' a cycle specific analysis will be performed.

With reduced feedwater temperature, the Load Reject Without Bypass event will be less severe because of the reduced core steaming rate and lower initial void fraction.

Consequently, no further analysis is needed for that event.

However, the feedwater controller failure event becomes more severe with a feedwater heater out of service and could become the limiting transient for a specific cycle.

Consequently, the cycle specific analysis for the feedwater controller failure event will be performed with a 100 degree F feedwater temperature reduction.

The calculated change in CPR for that event will then be used in determing the cycle specific MCPR LCO value.

In the case of a single S/RV out of service, transient analysis results showed that there is no impact on the calculated MCPR LCO value.

The change in CPR for this operating condition will be bounded by reload licensing calcu-lations, and no further analyses are required.

The analysis for.a single S/RV out of service is valid in conjunction with dual and single recirculation loop i

operation.

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-1 that are input to a GE-core dynamic behavior transient computer program.

The codes used to evaluate events are described in i

LA SALLE UNIT 1 B 3/4 2-3 Amendment No. 70

[

POWER DISTRIBUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued)

NEDE-24011-P-A-US (Reference 4).

The outputs of these programs along with the initial MCPR form the input for further analyses of the thermally limiting bundle (Reference 4).

The principal result of this evaluation is the reduction in MCPR caused by the transient.

The need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the ODYN computer code for analyzing rapid pressurization events.

Generic statistical analyses were performed for plant groupings of similar design which considered the statistical variation in several parameters, i.e., initial power level, CRD scram insertion time, and model uncertainty.

These analyses, which are described further in Reference 2, produced generic Statistical Adjustment Factors which have been applied to plant and cycle specific ODYN results to yield operating limits which provide a 95% probability with 95% confidence that the limiting pressurization event will not cause MCPR to fall below the fuel cladding integrity Safety Limit, i

As a result of this 95/95 approach, the average 20% insertion scram time must be monitored to assure compliance with the assumed statistical distribu-tion.

If the mean value on a cycle cumulative, running average basis were to exceed a 5% significance level compared to the distribution assu,med in the ODYN statistical analyses, the MCPR limit must be increased linearly, h as a function of the mean 20% scram time, to a more conservative value whic reflects an NRC determined uncertainty penalty of 4.4%.

This penalty is applied to the plant specific ODYN results, i.e. without statistical adjust-ment, for the limiting single failure pressurization event occurring at the limiting point in the cycle.

It is not applied in full until the mean of all current cycle 20% scram times reaches the 0.86 seconds value of Specifica-l tion 3.1.3.3.

In practice however, the requirements of 3.1.3.3 would most likelybereached,i.e.,i[1dividualdatasetaverage>0.86 secs,andthe required actions taken well before the running average exceeds 0.86 secs.

The 5% significance level is defined in Reference 4 as:

B=p+1.65(N/$N)1/2, T

3 j

i=1 where p

=

mean value for statistical scram time distribution to 20% inserted = 0.672 standard deviation of above distribution = 0.016 o

=

N

=

number of rods tested at BOC, i.e., all operable 2

rods n

INg= total number of operable rods tested in the i=1 current cycle LA SALLE UNIT 1 B 3/4 2-4 Amendment No. 70 l

I POWER DISTRIBUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued)

{

The value for tg used in Specification 3.2.3 is 0.687 seconds which is

(

4 conservative for the following reason:

For simplicity in formulating and implementing the LCO, a conservative n

value for INg of 598 was used.

This represents one full core data set i=1 at BOC plus one full core data set following a 120 day outage plus twelve 10% of core, 19 rods, data sets.

The 12 data sets are equivalent to 24 operating months of surveillance at the increased surveillance frequency of one set per 60 days required by the action statements of Specifications 3.1.3.2 and 3.1.3.4.

That is, a cycle length was assumed which is longer than any past or contemplated refueling interval and the number of rods tested was maximized in order to simplify and conservatively reduce the criteria for the scram time

.at which MCPR penalization is necessary.

The purpose of the K, factor specified in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow conditions.

At less than 100% of rated flow, the required MCPR is the product of the MCPR and the K factor.

The K factor assures that the Safety Limit MCPR will not be violate,d.

Methodology for establishing the K factor is described in f

Reference 4.

f At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.

During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed.

The MCPR margin will thus be demonstrated'such that future MCPR evaluation below this power level will be shown to be unnecessary.

The daily requirement for. calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The requirement for calculating MCPR'when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

LA SALLE UNIT 1 B 3/4 2-5 Amendm.ent No. 70

POWER DISTRIBUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued) i

References:

1.

General Electric Company Analytical Model for Loss of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K. NEDE-20566 November 1975.

2.

" Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," General Electric Company Licensing Topical Report NEDO 24154 Vols. I and II and NEDE-24154 Vol. III as sup-plemented by letter dated September 5, 1980, from R. H. Buchholz (GE) to P. S. Check (NRC).

3.

"LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-r Coolant Accident Analyses," General Electric Company Report NEDC-31510P, December 1987.

4.

" General Electric Standard Application for Reactor Fuel,"

NEDE-24011-P-A, (latest approved revision).

5.

" Extended Operating Domain.and Equipment Out-of-Service for LaSalle County Nuclear Station Units I and 2," NEDC-31455, November 1987.

3/4.2.4 LINEAR HEAT GENERATION RATE The specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

The power spike penalty specified is based on the analysis presented in Section 3.2.1 of the GE topical report NEDM-10735 Supplement 6, and assumes a linearly increasing variation in axial gaps between core bottom and top and assures with a 95% confidence that no more than one fuel rod exceeds the design LINEAR HEAT GENERATION RATE due to power spiking.

k LA SALLE UNIT 1 B 3/4 2-6 Amendment No. 70

DESIGN FEATURES 5.3 REACTOR CORE t

FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 764 fuel assemblies.

Each assembly consists of a matrix of Zircalloy clad fuel rods with an initial composition of slightly enriched uranium dioxide, UO to those fuel designs approved for use ik.

Fuel assemblies shall be limited BWR's, CONTROL ROD ASSEMBLIES t

5. 3. 2 The reactor core shall contain 185 cruciform shaped control rod assemblies.

The control material shall be boron carbide power (B,C) and/or hafnium metal.

The control rod assembly shall have a nominal axial Ibsorber length of 143 inches.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE

5. 4.1 The reactor coolant system is designed and shall be maintained:

F In accordance with the code requirements specified in Section 5.2 a.

of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

1.

1250 psig on the suction side of the recirculation pumps.

2, l

1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

t 3,

1500 psig from the discharge shutoff valve to the jet pumps.

c.

For a temperature of 575'F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is s 21,000 cubic feet at a nominal T,y, of 533'F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

i LA SALLE - UNIT 1 5-4 Amendment No. 70

O ADMIN]STRATIVE CONTROLS Semiannual Radioactive Effluent Release Report (Continued)

The radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report period:

a.

Container volume, b.

Total curie quantity (specify whether determined by measurement or estimate),

Principal radionuclides (specify whether determined by c.

measurement or estimate),

l d.

Type of waste (e.g)., spent resin, compacted dry waste, l

evaporator bottoms,

Type of container (e.g., LSA, Type A, Type B, Large Quantity),

e.

and a

f.

Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.

i 1

5.

Monthly Operating Report 1

j Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, shall be submitted on a monthly basis to the Director, Office of Nuclear Reactor Regulation, Mail Station P1-137, US Nuclear Regulatory l

Commission Washington, DC 20555, with a copy of the appropriate l

Regional Office, to arrive no later than the 15th of each month.

following the calendar month covered by the report.

Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) i was made effective.

In addition, a report of any major changes to l

the radioactive waste treatment systems shall'be submitted with the Monthly Operating Report for the period in which the evaluation was l

reviewed and accepted by Onsite Review and Investigative Function, j

j 6.

CORE OPERATING LIMITS REPORT i

I s

j a.

Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or I

any remaining part of a reload cycle for the following:

1 LA SALLE UNIT 1 6-24 Amendment No. 70

l

]

i ADMINISTRATIVE CONTROLS Semiannual Radioactive Effluent Release Report (Continued)

(1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.

(2) The minimum Critical Power Ratio (MCPR) (including 20%

scram time, tau (t), dependent MCPR limits, and K, core flow MCPR adjustment factors) for Technical Specification 3.2.3.

(3) The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.4.

(4) The Rod Block Monitor Upscale Instrumentation Setpoints for Technical Specification Table 3.3.6-2.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel (latest approved revision).

The core operating (e.mits shall be determined so that all c.

li applicable limits g., fuel thermal-mechanical limits, core thermal-hydraulic limits,. ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

The CORE OPERATING LIMITS REPORT, including any mid-cycle d.

revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the U.S. Nuclear Regulatory Commission Document Control Desk with copies to the Regional Administrator and Resident Inspector.

B.

Deleted LA SALLE UNIT 1 6-25 Amendment No. 70

o.aeeg A

umito starts

[

NUCLEAR REGULATCRY COMMISSION n

s

[

wAnnmorow, o, c. seems

g...+

t COMMONWEALTH EDISON COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION.. UNIT 2 AMENDMENT TO FACILITY OPERATING-LICENSE Amoument No. 54 License No. NPF-18 1.

The Nuclear Regulatory Comission (the Comission or the NRC) has found that:

A.

The application for amendment filed by tne Commonwealth Edison Company (the licensee), dated August 18, 1989 and supplemented September 13, 1989 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.

There is reasonable assurance: (1) 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 Comission'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.

l l

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to

(

read as follows:

(2) Technical Specifications and. Environmental. Protection. Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 54, and the Environmental Protection Plan contained in

)

Appendix B, are hereby incor'porated in the license.

The licensee shall l

operate the facility in accordance with the Technical Specifications and the Envirormental Protection Plan.

9

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P 3.

This amendment is effective upon criticality following the third refueling outage or on 30 June 1990 whichever is earlier.

FOR THE NUCLEAR REGULATORY COMMISSION P

/ John W. Craig, Director Project Directorate 111-2 Division of Reactor Projects - III, IV, Y and Special Projects

Enclosure:

Changes to the Technical Specifications Date of Issuance: December 18, 1989 i

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ENCLOSURE TO LICENSE AMENDNENT N0. 54 FACILITY OPERATING LICENSE NO. NPF-1B,,

i DOCKET NO. 50-374 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change.

REMOVE INSERT I

I II II IV IV XII XII XIX XIX 1-2 1-2 1-3 1-3 1-4 1-4 1-5 1-5 1-6 1-6 1-7 1-7 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2a 3/4 2-2b 3/4 2-3 3/4 2-2 3/4 2-4 3/4 2-3 3/4 2-4a 3/4 2-4 3/4 2-5 3/4 2-5a 3/4 2-6 3/4 2-7 3/4 2-5 3/4 3-53 3/4 3-53 i

3/4 3-54 3/4 3-54 B 3/4 2-1 B 3/4 2-1 l

B 3/4 2-2 l

B 3/4 2-3 8 3/4 2-2 l

B 3/4 2-4 8 3/4 2-3 B 3/4 2-5 B 3/4 2-4 i

B 3/4 2-6 8 3/4 2-5 B 3/4 2-7' B 3/4 2-6 5-4 5-4 6-24 6-24 6-25 6-25 i

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1 i

INDEX i

DEFINITIONS SECTION

1. 0 DEFINITIONS PAGE 1.1 ACTION............................................................

1-1 1.2 AVERAGE PLANAR EXP05URE.........................................'..

1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATIONRATE........................

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

1-1 1.5 CHANNEL CHECK.....................................................

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

1-1 1.7 CORE ALTERATION...................................................

1-2

1. 8 CORE OPERATING LIMITS REP 0RT......................................

1-2

1. 9 CRITICAL POWER RAT10..............................................

1-2 1.10DOSEEQUIVALENTI131..........l..................................

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

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

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

1-2 1.24 FRACTION OF LIMITING POWER DENSITY................................

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

1-3 1.16 FREQUENCY N0TATION.................................................

1-3 4

1.17 GASEOUS RADWASTE TREATMENT SY5 TEM........................'......... 1-3 1.18 IDENTIFIED LEAKAGE................................................

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

a 1-3 1.20 LIMITING CONTROL ROD PATTERN......................................

1-3 1.21 LINEAR HEAT GENERATION RATE.......................................

1-4 1.22 LOGIC SYSTEM FUNCTIONAL TEST......................................

1-4 1.23 MAXIMUM FRACTION OF LIMITING POWER DENSITY........................

1-4 1.24 MINIMUM CRITICAL POWER RAT10......................................

1-4 1.25 0FFSITE DOSE CALCULATION MANUAL...........'........................

1-4 e

LA SALLE - UNIT 2

.I Amendment No. 54

1 u

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j INDEX DEFINITIONS SECTION DEFINITIONS (Continued)

PAGE 1.26 OPERABLE - OPERABILITY............................................

1-4 1.27 OPERATIONAL CONDITION - CONDITION.................................

1-4 1.28 PHYSICS TESTS.....................................................

14 1,29 PRESSURE BOUNDARY LEAKAGE.........................................

1-5 1.30 PRIMARY CONTAINMENT INTEGRITY.....................................

1-5 1.31 PROCESS CONTROL PR0 GRAM...........................................

1-5 1.32 PURGE - PURGING...................................................

1-5 1.33 RATED THERMAL PGWER...............................................

1-5 1.34 REACTOR PROTECTION SYSTEM RESPONSETIME...........................

1-5 1.35 REPORTABLE EVENT..................................................

1-6 1.36 ROD DENSITY.......................................................

1-6 1.37 SECONDARY CONTAINMENT INTEGRITY...................................

16 1.38 SHUTDOWN MARGIN...................................................

1-6 1.39 SOLIDIFICATION....................................................

1-6

1. 4 0 S O U R C E C H E C K...................................................

1-7 1,41 STAGGERED TEST BASIS..............................................

1-7 l

1.42 THERMAL P0WER.....................................................

1-7 1.43 TURBINE BYPASS RESPONSE TIME......................................

1-7 1,44 UNIDENTIFIED LEAKAGE..............................................

1-7 1.45 VENTILATION EXHAUST TREATMENT SYSTEM..............................

1-7 1.46 VENTING...........................................................

1-7 LA SALLE - UNIT 2 II Amendmant No. 54

to I

INDEX 1

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

9 SECTION PAGE 3/4.0 APPLICABILITY...................................................

3/4 0 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.......................'.......................

3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.........................................

3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability......................................

3/4 1-3 Control Rod Maximum Scram Insertion Times....................

3/4 1-6 Control Rod Average Scram Insertion Times....................

3/4 1-7 Four Control Rod Group Scram Insertion Times.................

3/4 1-8 Control Rod Scram Accumulators...............................

3/4 1-9 Control Rod Drive Coupling...................................

3/4 1-11 Control Rod Position Indication..............................

3/4 1-13 Control Rod Drive Housing Support............................

3/4 1 15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer..........................................

3/4 1-16 Rod Sequence Control System..................................

3/4 1-17 Rod Block Monitor.......,.......'..............................

3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM................................

3/4 1-19 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM...........................

3/4 1-23 1

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE...................

3/4 2-1 3/4.2.2 APRM SETP0lNTS...............................................

3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RAT10.................................

3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE..................................

3/4 2-5 LA SALLE - UNIT 2 IV Amendment No. 54

i.

INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY...................................................

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.........................................

B 3/4 1-)

3/4.1.2 REACTIVITY AN0MALIES....................................

B 3/4 1-)

t 3/4.1.3 CONTROL R0DS............................................

B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS............................

B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL S'YSTEM...........................

B 3/4 1-4 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM......................

B 3/4 1-L 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..............

B 3/4 2-)

3/4.2.2 APRM SETP01NTS..........................................

B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0............................

B 3/4 2-P 3/4.2.4 LINEAR HEAT GENERATION RATE.............................

B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............... B 3/4 3-)

3/4. 3. 2 ISOLATION ACTUATION INSTRUMENTATION.....................

B 3/4 3-2 i

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.........................................

B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.......

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

B 3/4 3-3 3/4.3.6 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION.,..........

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

B 3/4 3-4 Seismic Monitoring Instrumentation.......................

B 3/4 3-4 LA SALLE - UNIT 2 X))

Amendment No. 54

to LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIREMENTS........................

3/4.1 21 3.1.5-2 SODIUM PENTABORATE (Na B 0 2 10 16 10 H O) 2 VOLUME / CONCENTRATION REQUIREMENTS.................

3/4 1-22 3.4.1.5-1 CORE THERMAL POWER (% OF RATE 0) VERSUS TOTAL CORE i

FLOW (% OF RATED)..................................

3/4 4-5c 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS REACTOR VESSEL PRESSURE.......................

3/4 4-19 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST........

3/4 7-33 B 3/4 3-1 REACTOR VESSEL WATER 1.EVEL........................

B 3/4 3-7 B 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV at 1/4 T AS A FUNCTION OF SERVICE LIFE..........)........... B 3/4 4-7 B 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS..................

B 3/4 6-3a 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GAS AND LIQUID EFFLUENTS...................EOUS 52 5.1.2-1 LOW POPULATION 20NE...............................

53 I

6.1-1 DELETED...........................................

6-11 j

6.1-2 DELETED...........................................

6 12 i

6.1-3 MINIMUM SHIFT CREW COMPOSITION.....................

6-13 r

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l LA SALLE - UNIT 2 XIX Amendment No. 54 1

i.

pEr!NITIONS i

i CORE ALTERATMON D CORE A CERATION shall be the addition, removal, relocation or movement of I

fuel, sources, incore instruments or resr.tivity controls within the l

reactor pressure vessel with the vessel head removed and fuel in the L

vessel.

Suspension of CORE ALTERATION $ shall not preclude com'.letion of i

j W movement of a component to a safe conservative position.

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CORE OPERATING LIMITS REPORT I

1. 8 The CURE OPERATING LIMITS REPORT is the unit specific document that provides core operating limits for the current operating reload cycle.

j These cycle specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.6.A.6.

Plant operation j

1 within these operating limits is addressed in individual specifications.

i CRITICAL POWER RATIO i

1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the l

assembly which is calculated by application of the GEXL correlation to j

i cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

i 00$E EQUIVALEN1 1 131 l

1.10 D05E EQUIVALENT l 131 shall be that concentration of I 131 l

j micracuries/ gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I 131, b 132, 1-133, 1 134, and I 135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table !!! of T10 14644, " Calculation i

j of Distance Factors for Power and Test Reactor Sites."

i

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IAVERAGEDISJ,NTEGRATIONENERGY i

j 1.11 I shall be the average, weighted in proportion to the concentration of I

l each radionuclide in the reactor coolant at the time of samplin, of the 1

sum of the average beta and gamma energies per disintegration, n MeV, for isotopes, with half lives greater than 15 minutes, making up at least j

9W of the total non iodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME i

1.12 The EMERGENCY CORE COOLING SYSTEM (ECC$) RESPONSE TIME shall b l

interval from when the monitored parameter exceeds its ECC8 actuation i

setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e.

the valves travel to their required positions, pump discharge pressures r,each their required values, etc.

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 response time is measured.

t END OF CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.13 The END OF CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall b l

that time interval to energitation of the recirculation pump circuit 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 i

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

LA SALLE - UN!T 2 '

1-2 Amendment No. Si l

i.

DEFINITIONS

[ND 0F' CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPON5E TIME (Continu FRACTION OF LIMITING POWER DEN 517Y 1.14 The FRAETION OF LIMITING POWER DENSITY (FLPD) sha11 be the LHGR existing l

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

FRACTION OF RATED THERMAL POWER 1.15 The FRACTION Of RATED THERMAL POWER (FRTP) shall be the measured T l

POWER divided by the RATED THERMAL POWER, fRQLENCYNOTATION 1.16 The FREQUENCY NOTATION specified for the performance of Surveillance l

Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM 1.17 A GA$E0V5 RADWASTE TREATMENT SYSTEM shall be any system designed and I

installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and provicing for celay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

IDENT1F]ED LEAKAGE 1.18 IDENllFIED LEAKAGE shall be:

I Leakage into collection cystems, such as pump seal or valse a.

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 specifically located and known either not to interfere with the operation of the leakage detection systens or not to be PRESSURE BOUNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME 1.19 The ISOLATION SYSTEM RE$PONSE TIME shall be that time interval from w l

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 leading delays where applicable.

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

LIM 111NC CONTROL R0D FATTERN 1.20 A LIMITING CONTROL ROD 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.

LA $ALLE UNIT 2 13 Amendment No. 54 1

s.

DEFINITIONS LINEAR HEA1 GENERATION RAlt 1.21 LINEAR HEAT GENERATION RAT [ (LHGR) shall be the heat generation per unit l

1ength of fuel rod.

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

LOGIC $YSTEMFUNCTIONALTEST 1.22 A LOGIC $YSTEM FUNCTIONAL TEST shall be a test of all logic components, I

i.e., all re19ys and contacts, all trip units, solid state logic elements, etc. of a logic circuit from sensor through and including the actuated devicetoverifyOPERABILITY, THE LOGIC $Y$1[M FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

MAX 1 MUM FRACTION OF LIMITING POWER OtN51TY 1.25 The MAXIMUM FRACTION OF LIMITING POWER DEN 51TY (MFLPD) shall be the highest l

value of the FLPD which exists in the core.

MINIMUM CRITICAL POWER RATIO 1.24 Ths MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l

exists in the core.

OFF$1TE 00$E CALCULATION MANUAL 1.25 The OFFSITE 00$E CALCULATION MANUAL (00CM) shall contain the methodology I

and parameters used in the calculation of offsite doses due to radioactive gaseous and licuie ef fluents and in the calculation of gaseous and liquid ef fluent monitoring alarm / trip setpoints.

OPERABLE OPERABILITY 1.26 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),

when all necessary attendant instrumentation, controls, a normal and 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 capabit of ptrforming their related support function (s).

OPERATIONAL CONDITION CONDITION 1.27AnOPERATIONALCONDITION,i.e.ltionandaveragereactorcoolantt CONDITION, shall be any one inclusive l

combination of mode switch pos as specified in Table 1.2.

PHY$1CS TEST $

1.28 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation l

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.

LA SALLE - UNIT 2 14 Amendment No. 54

I' O

D[FINIT10h5 PR[5$URI BDUNDARY LEAKAGE 1,29 PRIS$URE BOUNDARY LEAKAGE shall be leakage through a non* isolable faultl in a reactor coolant System component body, pipe well or vessel well.

PRIMARY CONTA}NM[NT INTEGRITY 1.30 PRIMARY CONTAINMENT INTEGRITY shall exist when:

l All primary containw.ent penetrations required to be closed a.

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 as provided in Table 3.6.31 of Specification, 3.6.3.

b.

All primary containment equipment hatches are closed and sealed.

Each primary Containment air lock is OPERABLE pursuant to c.

5pecification 3.6.1.3.

The primary containment leakage rates are within the limits of d.

Specification 3.6.1.2.

The suppression chameer is OPERABLE pursuant to Specification e.

3.6.2.1.

f.

The sealing mechanism associated with each primary containtrent penetration; e.g., weles, bellows or 0 rings, is OPERAELE.

PROCESS CONTROL PROGRAM 1.31 The PROCESS CONTROL PROGRAM (PCP) shall contain the sampling, analysis.

l and formulation determination by which SOLIDIFICATION of racioactive wastes from liquid systems is assured.

PURGE PURGING 1,32 PURGE or PURGING shall be the controlled process of discharging air or 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 1.33 RATED THERMAL POWER shall be a total reactor core heat transfer rate to l

the reactor coolant of 3323 M.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interv when the monitored parameter exceeds its trip setpoint at the channel l

sensor until de energitation 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.

LA SALLE UNIT 2 1-5 Amendment No. 54 I

i.

OEFIN1110NS REPORTABLE EVEN1 1.35 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

l ROD DEN $1TY 1.36 ROD DEN 51TY shall be the number of control rod notches inserted as a j

fraction of the total number of control rod notches.

All rods fully inserted is equivalent to 100% ROD DENSITY,

$ECONDARY CONTAINMENT INTEGR1TY 1.37 SECONDARY CONTAINMENT INTEGRITY shall exist when:

l All secondary containment penetrations required to be closed a.

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.

The standby gas treatment system is OPERABLE pursuant to c.

Specification 3.6.5.3.

t At least one door in each access to the secondary containment d.

is closed The sealing mechanism associated with each secondary containment e.

penetration, e.g., welds, bellons 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.

$HUTDOWN MARGIN 1.38 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is l

subcritical 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.

SOLIDIFICATION 1.39 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid l

systems to a homogeneous (uniformly distributed), monolithic, immobilized solid with definite volume and shape, bounded b distinct outline on all sides (free-standing). y a stable surface of LA Salt.E UNIT 2 16 Amendment No.54

i..

DEFINITIONS

$0VRCE CHECK 1.4D A SOURCE CHECK shall be the qualitative assessment of channel response l

when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BA515 1,41 A STAGGERED TEST 6A515 shall consist of:

I A test schedule for n systems, subsystems, trains or other a.

designated components obtained by dividing the specified test interval into n equal subintervals, b.

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

THERMAL POWER 1,42 THERMAL POWER shall be the total reactor core heat transfer rate to the l

reactor coolant.

TUREINE BYPAS$ SYSTEM RESPONSE TIME 1.43 The TURBINE BYPAS$ SYSTEM RESPONSE TIME shall be time interval from when I

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

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

UNIDENTIFIED LEAKAGE 1.44 UNIDEN11FIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

l VEN71LA110N EKHAUST TP.EATMENT SYSTEM 1.45 A VEN11LAT10N EXHAUST TREATMENT SYSTEM shall be any system designed and I installed to reduce gaseous radiciodine or radioactive material in particu-late 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 release to the environment (such a system is not considered to have any effect on noble gas effluents).

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

VENTING 1,46 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or l

other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING.

not imply a VENTING process.

Vent, used in system names, does LA SALLE - UNIT 2 17 Amendment No.54

so 3/4.2' POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT.

(

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25'4 of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the ner.t 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2,1 All APLHGRs shall be verified to be equal to or less than the limits specified in the CORE OPERATING LIMITS REPORT.

l a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is c.

operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

LA SALLE - UNIT 2 3/4 2-1 Amendment No. 64

Y.

s.

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETP01NTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased simulated thermal power upscale scram trip setpoint

($) and flow biased simuisted thersal power-upscale control red block trip setpoint ($gg) shall be established according to the following relationships:

a.

.Two Recirculation Loop Operation

$ less than or equal to (0.58W + 59%)T S

less than or equal to (0.58W + 47%)T gg b.

Single Recirculation Loop Operation 5 less than or equal to (0.58W + 54.3%)T

$gg less than or equal to (0.58W + 42.3%)T where:

5 and $RB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 108.5 million 1bs/hr, T = Lowest value of the ratio of TRACTION OF RATED THERMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY or the value 1.0.

T is always less than or equal to 1.

APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to m of RATED THERMAL POWER.

ACTION:

With the APRM flow biased simulated thermal power upscale scram trip setpoint and/or the flow biased simulated thermal pomer upscale control rod block trip setpoint set less conservatively than S or $

as above determined, initiate correctiveactionwithin15minutesandrestNe,5and/ ors,towithinthe recuired limits

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to Tess than 25% of g

RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVE1LLANCE RE001 REPENTS 4.2.2 The FRTP and the MFLPD for each class of fuel shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power upscale scram and control rod block trip setpoint verified to be within the above limits or adjusted, as required:

At taast once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a.

b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.

with MFLPD greater than or equal to FRTP,

'witn MFLPD greater than the FRTP up to 90% of RATED THERMAL POWER, rather than i

adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM read-ings are greater than or equal to 100% times MFLPO, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, the required gain adjustment increment does not exceed 10% of RATED THERMAL POWER, and a notice 1

of the adjustment is posted on the reactor control panel.

1 LA SALLE - UNIT 2 3/4 2 2 Amendment No.54

[

l

r.

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO l

LIMITING CONDITION FOR OPERATION I

I

3. 2. 3 The MINIMUM CRITICAL POWER RA110 (MCPR) shall be equal to or greater than the MCPR limit specified in the CORE OPERATING LIMITS REPORT.

l APPLICABILITY:

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or eaual to 25% of RATED THERMAL POWER.

ACTION I

a.

With MCPR less than the applicable MCPR limit as determined for one j

of the conditions specified in the CORE OPERATING LIMITS REPORT:

l t

1 1.

Initiate corrective action within 15 minutes, and l

2.

Restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

3.

Otherwise, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

When operating in a condition not specified in the CORE OPERATING LIMITS REPORT, reduce THERMAL POWER to less than 25% of RATED THERMAL i

POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

LA SALLE UNIT 2 3/4 2 3 Amendment No. 54

so P0hTR DISTRIEUTION LIMITS 3/4.2.3 MINIMUM CRITICAL P0hTR RATIO SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

a.

t

= 0.86 prior to performance of the initial scram time measurements 8V' for the cycle in accordance with Specification 4.1.3.2, or b.

1,y, determined within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit specified in the CORE OPERATING LIMITS REPORT.

l a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.

with a LIMITING CONTROL ROD PATTERN for MCPR.

LA SALLE - UNil 2 3/4 2-4 Amendment No. 54

s.

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed the limits specified in the CORE OPERATING LIMITS REPORT.

I APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 2 n of RATED THERMAL POWER.

ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than the limit:

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a.

b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Intially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.

on a LIMITING CONTROL ROD PATTERN for LHGR.

1 LA SALLE UNIT 2 3/4 2-5' Amendment No. 54 1

l

1 r-g TABLE 3.3.6-2

?g CONIROL ROD WITHDRAWAL 810CK INSTRtlR NTATION SETPOINTS

[

iRIP FUNCTION IRIP SETPOINT ALLOWASLE VALUE 1.

ROD BLOCK MONITOR

=

a.

Upscale The Rod Block Monitor Upscale Setpoints shall be established according to the relationships specified in the CORE OPERATING LIMITS REPORT.

b.

Inoperative N.A.

N.A.

c.

Downscale 3 5% of RAl[D THERMAL POWER 3 3% of RAi[0 TNElW44L POWER 2.

APRN a.

Flow Biased Simulated Thermal Power-Upscale 1)

Two Recirculation U

Loop Operation

~< 0.58 W + 47%*

< 0.58 W + SOE*

~

2)

Single Recirculation T

toop Operatior.

< 0.58 W + 42.3%

< 0.58 W + 45.3%*

C b.

Inoperative N.A.

N.A.

c.

Downscale

> 5% of RATED THEIW44L POWER

> 3% of RATED THEW 14L POWER d.

Neutron flux-High 512% of RATED THElW44L POWER 514% of IIATED INElsent POWER 3.

500RC'E RAIIGE MONITORS a.

Detector. net full in M.A.

N.A.

b b.

Upscale

< 2 x 10 cps

< $ x 10 cps c.

Inoperative R.A.

N.A.

d.

Downscale 3 0. 7 cps 3 0.5 cps 4.

INTERMEDIATE RAIIGE PWIIITORS

~

Y a.

Detector not full in M. A.

N.A.

3 b.

Upscale

< 106/125 of full scale

< 110/125 of full scale E

c.

Inoperative' N.A.

N.A.

E, d.

Downscale 1 5/125 of full scale 3 3/125 of full. scale

.E

~

O

7 g

TA81E 3.3.6-2 (Continued)

~

Ep CONTRot ROD WITHDRAWAt BIOCK INSTRUMENTATION SETPOINTS TRIP IUNCTION TRIP SEIPOINI AttotestE VALUE

_E 5.

SCRAM DISCHARGE VOLUME m

a.

Water tevel-High 1 165' 5's" 1 765' % "

b.

Scras Discharge Volume Switch in Bypass N.A.

N.A.

6.

REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.

Upscale

< 108/125 of full scale

< 111/125 of full scale b.

Inoperative N.A.

R.A.

c.

Comparator i 10% flow deviation i 11% flow deviation

1 m

h N

3 9:

e O

E

The Average Power Range Monitor rod block f tenction is varied as a function of recirculation loop flow (W).

The trip setting of this function must te aintaid in accordance with Specification 3.2.2.

4

so 3/4.2 POWER DISTRIBUTION LIMITS

_ BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss of coolant accident will not exceed the 2200'F limit specified in 10 CFR 50.46.

3/4.2.1 AVER' AGE PLANAR LINEAR HEAT GENERAT!0N RATE This t,pecification assures that the peak cladding temperature following the postulated design basis loss of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

This specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations.

The peak cladding temperature (PCT) following a postulated loss of-coolant accident it, primarily a function of the average heat generction rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady-state gap conductance and rod-to-rod local peaking factor.

The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor.

However, the current General Electric (GE) calculational models (SAFER /GESTR described in Reference 3), which are consistent with the I

requirements of Appendix K to 10 CFR 50, have established that APLHGR values are not expected to be limited by LOCA/ECCS considerations.

APLHGR limits are still required, however, to assure that fuel rod mechanical integrity is maintained.

They are specified for all resident fuel types in the Core Operating Limit Report based on the fuel thermal mechanical design analysis.

4 LA SALLE UNIT 2 B 3/4 2 1 Amendment No. 54

i.

POWERDISTRIBUTIONSYSTiM5 BASES 3/4.2.2 APRM $E'TPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LNGR at RATED THERMAL POWER.

The flow biased simulated thermal power upscale scram setting and con-trol red block functions of the APRM instruments for both two recirculation loopoperationandsinglerecirculationloopoperationmustbeadjustedtoensure that the MCPR does not become less than the fuel cladding safety limit or that 3, 3 plastic strain does not occur in the degraded situation.

The scram settings and rod block settings are adjusted in accordance with the formula in this speci-fication when the combination of THERMAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

3 /4. 2. 3 MINIMUM CRIT] CAL POWER RATIO The required operating limit MCPRS at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Stfety Limit MCPR and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming, instrument trip setting,given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest recuc-tion in CRITICAL POWER RATIO (CPR).

The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest delta MCFR.

When added to the Safety Limit MCPR, the required minimum operating limit MCFR of Specification 3.2.3 is obtained and presented in the CORE OP[ RATING LIM]TS REPORT.

Ahalyses have been performed to determine the effects on CRITICAL POWER RATIO (CPR) during a transient assuming that certain equipment is out of service. A detailed description of the analyses is provided in Reference 5.

The analyses performed assumed a single failure only and established the licensing bases to allow continuous plant operation with the analyzed equipment out of service.

The following single equipment failures are included are part of the transient analyses input assumptions:

1.

main turbine bypass system out of service, 2.

recirculation pump trip system out of service.

LA SALLE UNIT 2 8 3/4 2 2 Amendment No,54

\\

l P0m'ER Di$TRIEUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued) 3.

safety / relief valve (S/RV) out of service, and 4.

feedwater heater out of service (corresponding to a 100 degree F reduction in feedwater ten @erature),

i For the main turbine bypass and recirculation pump trip systems specific cycle independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for i

Operation (LCO) values are established to allow continuous plant operation with these systems out of service.

A bounding end of cycle exposure condition was used to develop nuclear input to the transient analysis model.

The bounding exposure condition assumes a more top peaked axial power distribution than the nominal power shape, thus yielding a bounoing scram response with j

reasonable conservatisms for the MCPR LCO values in future cycles.

The i

MCPR LCO values shown in the CORE OPERATING LIMITS REPORT for the main turbine bypass and recirculation pump trip systems out of service are valid provided l

]

that these limits bound the cycle specific results, r

i The analysis for main turbine bypass and recirculation pump trip systems inoperable allows operation with either system inoperable, but not both at the I

same time.

For operation with the feedwater heater out of service, a Cycle specific analysis will be performed.

With reduced feedwater temperature, the Load Reject Without Bypass event will be less severe because of the reduced core i

i steaming rate and lower initial void fraction.

Consequently, no further analysis is needed for that event.

However, the feedwater controller failure event becomes more severe with a feed =ater heater out of service and could i

become the limiting transient for a specific cycle.

Consequently, the cycle specific analysis for the feedwater controller failure event will be performed with a 100 degree F feedmater temperature reduction.

t The calculated change in CPR for that event will then be used in determining the cycle specific MCPR LCO value.

i r

In the case of a single S/RV Out of service, transient anal showed that there is no impact on the calculated MCPR LCO value.ysis results l

The change l

in CPR for this operating condition will be bounded by reload licensing i

calculations and no further analyses are required.

The analysis for a single S/RV out of service is valid in conjunction with dual and single recirculation loop operation.

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.01 that are input to a GE-core dynamic behavior transient computer program.

The codes used to evaluate events are described I

l l

l t

LA SALLE - UNIT 2 B 3/4 2 3 Amendment No.54 l

i

i.

POWER DISTRIBUTION SYSTEMS BASE $

MINIMUM CRITICAL POWER RATIO (Continued) in NEDE 24011 P A US (Reference 4). The outputs of these programs along with the initial MCPR form the input for further analyses of the thermally limiting bundle (Reference 4).

The principal result of this evaluation is the reduction in MCPR caused by the transient.

The need to adjust the MCPR operating limit as a function of scram time arises from the statistical approach used in the implementation of the ODYN computer code for analy2ing rapid pressurization events.

Generic statistical analyses were performed for plant groupings of similar design which considered t,he statistical variat, ion in several parameters, i.e., initial power level, CRD scram insertion time, and model uncertainty.

These analyses, which are described further in Reference 2, produced generic Statistical Adjustment Factors which have been applied to plant and cycle specific ODYN results to yield operating limits which provide a 95% probability with 95% confidence that 1,he limiting pressuritation event will not Cause MCPR to fall below the fuel cladding integrity $afety Limit.

As a result of this 95/95 approach the average 20% insertion scram time must be monitored to assure compliance w,ith the assumed statistical distribu-tion.

If the mean value exceed a 5% significance,on a cycle cumulative, running average, basis were to level compared to the distribution assumed in the ODIN statist,ical analyses, t,he MCPR limit must be increased linearly, as a function of the mean 20% scram time, to a more conservative value which reflects an NRC determined uncertainty penalty of 4.4%.

This penalty is applied to the plant specific ODYN results, i.e. without statistical adjustment, for the limit-ing sin cycle. gle failure pressurization event occurring at the limiting point in the It is not applied in full until the mean of all current cycle 20% scram times reaches the 0.B6 seconds value of Specification 3.1.3.3.

In pract, ice, however, the requirements of 3.1.3.3 would most likely be reached, i.e., indivi-cual cata set average > 0.86 secs, and the required actions taken well before the running average exceeds 0.86 secs.

The 5% significance level is defined in Reference 4 as:

tg = p + 1.65 (N /

N )1/2 e g

g i=1 where p

=

mean value for statistical scram time distribution to 20% inserted =.672 standard deviation of above distribution =.016 o

=

N a

y number of rods tested at 800, i.e., all operable rods n

IN a j

tot,a1 number of operable roos tested in the i=1 current cycle LA SALLE - UNIT 2 8 3/4 2-4 Amendment No.54

(

- g,8 g g.

POWER DISTRIBUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued)

The value for tg used in Specification 3.2.3 is 0.687 seconds which is conservative for the following reason:

For simplicity in forwulating and implementing the LCO, a conservative n

value for iNg of 598 was used.

This represents one full core data set

(=1 at BOC plus,one full core data set following a 120 day outage plus twelve 10% of core, 19 rods, data sets.

The 12 data sets are equivalent to 24 operating months of surveillance at the increased surveillance frequency of one set per 60 days required by the action statements of Specifications 3.1.3.2 and 3.1.3.4.

That is, a cycle length was assumed which is longer than any past or contemplated refueling interval and the number of rods tested was maximized in order to simplify and conservatively reduce the criteria for the scram time at which MCPR penalization is.necessary.

The purpose of the K, factor specified in the CORE OPERATING LIMITS REPORT is to define operating % of rated flowlimits at other than rated core flow conditions.

At less than 100

, the required MCPR is the product of the MCPR and the K, factor.

The K factor assures that the Safety Limit Methodolog,y for establishing the K factor is MCPR will not be vio14ted.

described in Reference 4.

g At THERMAL POWER 1evels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.

For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.

evaluation will be made at 25% of RATED THERMAL POWER level with minim recirculation pump speed.

The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.

The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.

The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

LA SALLE - UNIT 2 B 3/4 2-5 Amendment No, 54

I POWER D151R!tV710N $Y$TEMS BASES MINIMJM CRITICAL POWER RAT 10 (Continued)

References:

1.

General Electric Company Analytical Model for Loss of Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE 20566 November 1975, 2.

" Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors" General Electric Co. Licensing Topical Report NEDO 24154 Vols. 1 and Il and NEDE 24154 vol. 111 as sup-piemented by letter dated September 5, 1980, from R. H. Buchholz (GE) to P. 5. Check (NRC).

f 3.

"La$sile County Station Units 1 and 2 SAFER /GESTR LOCA Lons of Coolant Accident Analyses", General Electric Co. Report NEDC 31510P, December 1987.

4

" General Electric Standard Application for Reactor Fuel",

NEDE 24011 P A, (latest approved revision).

5.

" Extended Operating Domain and Equipment Out of service for Lassile County Nuclear $tation Units 1 and 2", NEDC 31455, November 1987.

3/4.2.4 LINEAR HEAT GENERATION RATE The specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in any rod is less than the design linear heat generation even if fuel pellet censification is postulated, The power spike penalty specified is based on the analysis presented in Section 3.2.1 of the GE topical report NEDM 10735 Supplement 6, and assumes a linearly increasing variation in axial gaps between core bottom and top and assures with a 95% confidence that no more than one fuel red exceeds the design LINEAR HEAT GENERATION RATE due to power spiking.

LA TALLE - UNIT 2 B 3/4 2 6 Amendment No.54 l

+

Oti!GN FIATURi$

' 5. 3 REACTOR CORE FUEL A$$tMBLl[$

i 5.3.1 The reactor core shall contain 764 fuel assemblies.

Each assembly I

consists of a matrix of Zircelloy clad fuel rods with an initial composition of slightly enriched uranium dioxide, We, Fuel assemblies shall be limited to those fuel designs approved for use in SWR's.

CONTROL ROD A$$tMSLIES 5.3.2 The reactor core shall contain 185 cruicform shaped control rod assemblies.

The control material shall be boron carbide powder (8 C) and/or hafnium metal.

Thecontrolrodassemblyshallhaveanominalaxialabsorber length of 143 inches.

5.4 REACTOR COOLANT SYSTEM 0[51GN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

In accordance with the code requirements specified in Section 5.2 a.

of the FSAR, with allomance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

1.

1250 psig on the suction side of the recirculation pumps.

2.

1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

3.

1500 psig from the discharge shutoff valve to the jet pumps.

c.

For a temperature of 575'F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is s 21,000 cubic feet at a nominal T,y, of 533'F.

5.5 METEOROLOGICAL TOWER LOCATION

5. 5.1 The meteorological tower shall be located as shown on Figure 5.1.1 1.

LA SALLE - UNIT 2 54 Amendment No.54

c.

1 I

i l

A MIN 1$7R&T10N CONTROLS l

Semiannual fadioactive Effluent Release Report (Continued) 1 The radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report pet iodt

)

a.

Container volume, b.

Total curie quantity (specify whether determined by I

l measurement or estimate).

i c.

Principal radionuclides measurementorestimate)(, specify whether determined by Type of waste (e.g)., spent resin, cortpacted dry waste, d.

evaporator bottoms,

e.

Type of container (e.g., LSA, Type A Type 6, Large Quantity),

and f.

Solidification agent (e.g., cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned releasts from the site to unrestrictec areas of radioactive materials in gesteus and liquid effluents on a quarterly basis.

The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reptrting period.

Monthly Operating Report Routine reports of operating statistics and shutdown experience, including docurrentation of all challenges to safety / relief valves, shall be submitted on a monthly basis to the Director Office of Nuclear Reactor Regulation, Mail Station PI 137 US Nuclear Regulatory Cornission, Washington, DC 20555, with a copy of the appropriate Regional Office, to arrive no later than the 15th of each month following the calendar month covered by the report.

?

Any changes to the OFF$1TE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 day's in which the change (s) -

was made effective.

In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.

6.

CORE OPERATING LIMITS REPORT a..

Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

LA SALLE UNIT 2 6-24 Amendment No. $4 mm

- s.

f f

i

\\

ADMINISTRATION CONTROLS Semiannual Radioactive Iff'19ent Release Report (Continued) l (1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.

(2) The minimum Critical Power Ratio (MCPR) (including 20%

scram time, tau (1), dependent MCPR limits, and Kf core flow MCPR adjustment f actors) for Technical Specification 3.2.3.

j (3) The Linear Heat Generation Rate (LHGR) for Technical i

Specification 3.2,4

{

3 (4) The Rod Block Monitor Upscale Instrumentation Setpoints for Technical Specification Table 3.3.6 2.

i 1

The ar,alytical methods used to determine the core operating b.

limits shall be those previously reviewed and approved by NRC in NEDE 24011 P A, General Electric Standard Application for Reactor fuel (latest approved revision).

i c.

The core operating (e.mits shall be determine so that all li applicable limits g., fuel thermal mechanical limits, core I

thermal hydraulic limits ECCS Limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

d.

The CORE OPERATING LIMITS REPORT, including any mid cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the U.S. Nuclear Regulatory Commission Document Control Desk with copies to the Regional i

Administrator and Resident Inspector.

i B.

Deleted, i

f i

I l

l LA SALLE UNIT 2 6-25 Amendment No. 54

_ _

ML20033H018

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ML20033H018
Site:	LaSalle (NPF-11-A-070, NPF-11-A-70, NPF-18-A-054, NPF-18-A-54)
Issue date:	12/18/1989
From:	Jocelyn Craig Office of Nuclear Reactor Regulation
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ML20033H019	List: ML20033H018
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NUDOCS 9004160060
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