ML20069J780

Proposed Tech Specs for Partial Implementation of GE ARTS Program

ML20069J780
Person / Time
Site:	LaSalle
Issue date:	06/09/1994
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML19312B508	List: ML20069J767 ML20069J780
References
NUDOCS 9406150128
Download: ML20069J780 (44)

Text

- . . _ . . . _ _ . =. ._ . - . - - . . . ... _- - . . _ ..

ATTACHMENT B PROPOSED CHANGES TO THE LICENSE /fECHNICAL SPECIFICATIONS l

FOR OPERATING LICENSES NPF-11 AND NPF-18 I

l l

NPF-11 NPF-18 License page 15 Index page IV License page 16 Index page XII l Index page IV B 2-10 Index page XII 3/4 2-2 B 2-10 3/4 3-8 3/4 2-2 3/4 3-53 3/4 3-8 3/4 3-54 3/4 3-53 3/4 4-1 3/4 3-53a B 3/4 2-1 3/4 4-1 Insert Paragraph #1 B 3/4 2-1 B 3/4 2-2 Insert Paragraph #1 B 3/4 2-5 B 3/4 2-2 Insert Paragraph #2 B 3/4 2-5 B 3/4 2-6 Insert Paragraph #2 B 3/4 3-4 B 3/4 2-6 '6-25 B 3/4 3-4 6-25 l

l l

l Ka\NLA\LASALLE\ARTSRE2:10 9406150128 DR 940609 ADOCK 05000373

_ .. PDR __. . _ . _ . - _ . .

6/23/88

- License No. NPF-ll (d) The maximum average planar linear heat generation ,

(MAPLHGR) limit will be reduced by 0.85.

(e) Technical Specification Setpoints shall read as follows:

T.S.2.2.1 S 0.66W + 45.7 (Trip Setpoint)

S 0.66W + 48.7 (Allowable)

T.S.3.2.2 S (0.66W + 45.7) T*

SRB (0.66W + 36.7) T*

T* as defined in T.S.3.2.2 T.S.3.3.6 APRM Upscale 0.66W + 36.7 (Trip Setpoint)

APRM Upscale 0.66W + 39.7 (Allowable)

RBM Upscale 0.66W + 34.7 (Trip Setpoint)

RBM Upscale 0.66W + 37.7 (Allowable) l (f) The average power range monitor (APRM) flux noise will be measured once per shift; and the recirculation loop flow will be reduced if the flux noise averaged over 1/2 hour exceeds 5 percent peak to peak, as measured by the APRM chart recorder.

(g) The core plate delta P noise will be measured once per shift, and the recirculation loop flow will be reduced if .

the noise exceeds one (1) psi peak-to-peak. ,,f Am. 12 D. Exemptions from certain requirements of Appendices G, H and 12/20/82 J and 10 CFR Part 73 are described in the Safety Evaluation Report and Supplement No. 1, No. 2 and No. 3 to the Safety Evaluation Report. In addition, an exemption was requested until the completion of the first refueling from the require-ments of 10 CFR 70.24 and an exemption from 10 CFR Part 50, Appendix E from performing a full scale exercise within one year before issuance of an operating license, both exemptions are described in Supplement No. 2 of the Safety Evaluation Report. Finally, an exemption was requested from the requirements of 10 CFR 50.44 until either the required 100 percent rated thermal power trip startup test has been completed or the reactor has operated for 120 effective full power days as specified by the Technical Specifications. This latter exemption is described in the safety evaluation of License knendment No. 12. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

Therefore, these exemptions are hereby granted. The faellity will operate, to the extent authorized herein, in conformity with the application, as amended, and the rules and regulations of the Commission (except as hereinafter exempted therefrom),

and the provisions of the Act.

0011k

j 6/23/88 License No. WPF-11 i

fj (iii) Prior to exceeding five percent power, the licensee shall include a description of the dose calculational methodology with a Class A transport and diffusion module, and a description of an i

acceptable meteorological measurement preventative and corrective maintainance program in the 3

radiological emergency plan.

i Am. 3 (31) Rolting of Valves j 7/15/82 Prior to January 15, 1983, the. licensee shall check the torque on all non-pressure boundary bolts (bolts whose failure will 4

l effect the operability of the valve) on each~ safety-related j valve located outside containment.

i Am. 4 (32) Eacuum Breaker valves 1 8/13/82 i

Prior to November 1, 1982, the licensee shall complete a test and shall submit its evaluation of the results which confirm i the capability of the vacuum breaker valves to withstand the i opening and closing forces associated with pool swell.

Am. 4 (33) Heatina-Ventilation and Air Condition Systems 8/13/82 (a) Prior to exceeding five percent power operation, the licensee must provide formal documentation of information regarding HVAC design fabrication and installation, discussed in meetings with the NRC on August 2 and 4,1982.

(b) Prior to exceeding fifty percent' power operation, the licensee shall submit the results of an independent review i acceptable to the NRC staff of the HVAC system, including design changes, fabrication,.and installation. The review shall encompass all safety- related HVAC systems.and the effect of non-safety related HVAC system failures on ,

safety systems. _

Am. 11 (34) IhInunh the First Fuel Cvele of Plant Ooeration. Technical 12/16/82 Specification 3.4.1.1 is Modified for One Recirculation Loop gut of Service._w_1.th Provis1Rna (a) The steady-state thermal power level will not exceed 50 percent of rated power.

(b) The minimum critical power ratio (MCPR) safety limit will be increased by 0.01 to 1.07.

(c) The minimum critical power ratio limiting condition for operation (LCO) will be increased by 0.01.

1

, , - - . . v._,.,#,

INDEX 1 i

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l SECTION PAGE 3/4.0 APPLICABILITY................................................... 3/4 0-1

, 3/4.1 REACTIVIlY CONTROL SYSTEMS

3/4.1.1 SHIRDOWN MARGI N. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4

. . . 1-1 3  !

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

b 3/4 1-2 I l 3/4.1.3 CONTROL R0DS l

i Control Rod 7perability...................................... i 3/4 1-3 l

Control Rod llaximus Scram Insertion Times.................... 3/4 1 6 Control' Rod Average Scram Insertion Times.................... 3/4 1-7 j

Four Control Rod G roup Scram Inse rti on Times . . . . . . . . . . . . . . . . . 3/4 1-8 Control Rod Scram Accumulators............................... 3/4 1-9

Controi Rod Drive Cou,iin..........,......................... 3/4 1-11 l

Control Rod Position Indication.............................. 3/4 1-13 3 Control Rod Drive Housing Support............................ 3/4 1-15 j 3/4.1.4 CDNTROL R00 PROGRAM CONTROLS l

Rod Worth Mi nimi ze r. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 . . . 1-16 4

s Itod B l ock Moni tor. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4 . . . . . I.

. . .1-18

3/4.1.5 4

STAND 8YLIQUIDCONTROLSYSTEM................................ 3/4 1-19 4

3/4.1.6 ECONOMIC GENERATION CONTROLSYSTEM...........................

i 3/4 1-23 j 3/4.2 POWER DISTRI8tlTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEATERATION

_ RATE................... 3/4 2-1 1

3/4.2.2 (?- -~isai.G..4DtLEfEb .......=.....................................

^

3/4 2-2 3/4.2.3 NINIMUM CRITICAL POWER RATI0................................. 3/4 2-3 i 3/4.2.4 LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . .3/4 . . .2-5 1

LA SALLE - UNIT 1 IV 9 Amendment No. 88 4

a - kw e a _ e a ,_a .%a ,a me .m .8%p . n an w .t _.um.,, aA,A e, wes..-.me-m..-- . --.- - - - - - - - - - - - - .,

. INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY................................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SNUTDOWN MARGIh......................................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.................................... B 3/4 1-1 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 SYSTEM........................... B 3/4 1-4 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM...................... B 3/4 1-5 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR NEAT GENERATION RATE.............. B 3/4 2-1 3/4.2.2 (1PRM SETPOI ............................. B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0............................ B 3/4 2-2' 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 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION..................... B 3/4 3-2 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 3-4 3/4.3.6 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION............ B 3/4 3-4 3/4.3.7 MONITC2ING INSTRUMENTATION Radiation Monitoring Instrumentation.................... - B 3/4 3-4 Seismic Monitoring Instrumentation............./........ B 3/4 3-4 LA SALLE - UNIT 1 XII Amendment No. 85

-r- --.,e e ew, y 9 - r

i

.?MITfNG SAFETY SYSTEM SETTfNGS SASES l

2EACTOR DROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) i average power Range Monitor (Continued) the flux distribution associated with uniform Tod withdraw 41s' does not involve '

nign iotal peaks anc because several roos must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod  !

withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than aceouate to assure snutoown oefore the power could exceed the Safety Limit.

The 15% neutron. flux trip remains active until the mode switch is placed. in the Run position.

I The APRM trip system is calibrated using heat balance data taken during l itesoy state conditions. Fission champers provide the basic input to the i system and therefore the monitors respond directly and quickly to changes due l to transient operation for the case of the Fixed Neutron Flux-High 1185 setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than tnat inoicateo by the neutron flux due to the time: constants of the heat transfer associated with the fuel. For the Flow Biased Simulated Thermal i

Power-Upscale setpoint. A time constant of 611 seconds is introduced into the .

flow biased APRM in order to simulate the fuel thermal transient characteristics. ,,,

A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.

~

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet Allow operating marcin that reduces the possibility of _,5L--

ynnecessary shutdown.I The flow references trip setpoint must be acjusted by tni specifieo formula in Specification 3.2.2 in oroer to maintain these margins w'nen MFLPD is gr. eater than or equal to FRTP.

3.

Reactor vessel Steam Dome Pressure-HiQh High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure

'ncrease while operating will also tend to increase the power of the reactor ey comoressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly nigner trips.

than T

the operating pressure to permit normal operation without spurious he setting provides for a wide margin to the maximum allowable design tressure ano takes into account the location of the pressure measurement

omoarea to the nignest pressure that occurs in the system curing a transient.

• nts trio setooint is ef fective at low power / flow conoitions when the turbine

.:c .a ne closure t mp ., oypasseo. For a turoine trip unoer tnese conoitions.

tne

'Imit.

transient analysis inoicateo an aceouate margin to the thermal nyaraulic ,

4 >ALLE - UNIT 1 8 2-10

M 4(v 8 3/y M - A WER C 3TRIBUTION LIMITS N.2.'k Innunonuy wPT DLMK 3/ k 2 ADRM SET 80!NTS " U

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• D

.!w!' ONDITION FOR OPERATION 3.2.2 ine A (5) ano flow e flow ciasec simulated thermal power upscale scram trip seyticint ec simulatee thermal power voscale control roc block tr/o setpoint (5gg) s 11beestaolisneeaccorcingtotnefollowingrelatiefsnios: ,

a. Two Recireglation Loop Operation 5 less than r ecual to (0.58W + 59%)T /

S gg less than r ecual to (0.58W + 47%)T /

,/

{

b. Single Recircula 'on Loop Operation /

5 less than or eau to (0.58W + 54.3%)T f'

S RB 1ess than or eau to (0.58W + 42.3%)T wnere:

/

5 anc Spg are in percent e RATED THERMAL POWER,j /

W = LooD recirculation flow s a percentage of ,the icoe recirculation i: flow wnien procuces a ratec core flow of 108.5 million 1bs/hr, Lowest value of the ratio o'f. FRACTION OF/ RATED THERMAL POWER civiced by the MAXIMUM FRACTIt[N OF LIM 'ING POWER DENSITY or the value 1.0. T is always less than or qual to 1.0. i 1

APPLICAEILITY:

OPERATIONALCONDITION1,wnhT RMAL POWER is greater than or i ecual to 2 n of RATED THERMAL POWER.

ACTION:

Witn tne APRM flow ciased simulated therm power anc/or the flow ciased simulatec thermal power ups$oscale scram egle control rodtrip setpoint block trip setoointsetlessconservativelythanf' ors correctiveactionwithin15minutesaMcrestNe,5analerSas ahove determined, initiate l g to within the  !

RATED THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. within the nexrecuireclimits'with 5URVEIL.' ANCE RE0VIREMENTS 4.2.2 /

value ofIne FRTP andand T calculated, the MFLPD for each class of fuel shall be eterminea, the thermal power upscale scr he most recent actual APRM flow bi ed simulated within the acove limits and control rod block trip setpoint erified to be adjusted, as required:

a. At least one per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 ours after completion of a THERMAL POWER increa of at least 1 of RATED THERMAL POWER, and c.

Initia y and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is with erating

LPD greater than or equal to FRTP.

~

"Witn NFLPD preater than 'the FRTP up to 90% of RATED THERMAL POWER, han rather adjusting /.he APRM setpoints, the APRM gain may be adjusted such that APRM readings /are greater than or equal to 100% times MFLPD, provided that the '

adjustre APRM reading coes not exceed 100% of RATED THERMAL POWER, the reov n tice of the adjustment is posted on the reactor control pane LA SALLE - UNIT 1 3/4 2-2 Amendment No.70 l

= , . .- .

- l

. t TABLE 4.3.1.1-1 (Continued) i g REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS i E CHANNEL OPERATIONAL f FUNCTIONAL UNIT CHANNEL CHECK FUNCTIONAL TEST CHANNEL CAllBRATION CONDITIONS FOR WHICH SURVEILLANCE REQUIRED E 8. Scram Discharge Volume Water

• Q Level - High NA M R 1, 2, 5

- 9. Turbine Stop Valve - Closure NA Q R -

1 I

10. Turbine Control Valve Fast Closure Valve Trip System 011 -

Pressure - Low NA Q R 1 l II. Reactor Mode Switch Shutdown Position NA R NA 1,2,3,4,5

12. Manual Scram NA W

13. Control Rod Drive NA 1,2,3,4,5 l

a. Charging Water Header ,

Pressure - Low NA M R 2, 5 w b. Delay Timer NA M R 2, 5 k

e.

(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM, and SRM channels shall be determined to overlap for at least 1/2 decades during each startup and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days. t (c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior.to startup, if not performed within the previous 7 days. 1 (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power levels .

calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 2 25% of RATED THERMAL '

POWER. The APRM Gain Adjustment Factor (GAF) for any channel shall be equal. to the power value deter-mined by the heat balance divided by the APRM reading for that channel. ,

g Within 2 h adjust any APRM channel with a GAF > 1.02. In addition, adjust any APRM channel within m 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, 1

if power is greater than or equal to 90X of RATED THERMAL  ;

Pmern ana lhe APRM channel GAF is  !

-@ < 0.95~ ort.f tr power is less than zum of RATED THERMAL POWER and the APRM repadtx the p@er d m (value~cetermined by the neat pasance oy more than 10% of RATED THENML PUwt.M.[ Until any requ r

• i W n L nt M M en accomplisnea, notificattion shall be postW un ine reactor control panel.

g (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal. ,

• (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EfPH). .

(g) Measure and compare core flow to rated core flow.

(h) This calibration shall consist of verifying the 61 I second simulated thermal power time constant.

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3 /4. 4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCVLATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION.FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2 ACTION

a. With only one (1) reactor coolant system recirculation loop in operation, comply with Specification 3.4.1.5 and:

1. Within four (4) hours:

~ a) Place the recirculation flow control system in the Master Manual mode or lower, and ,

1 b) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety I Limit by 0.01 to 1.08 per Specification 2.1.2, and j c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation by 0.01 per Specification 3.2.3, j and, l d) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable to single i recirculationloopoperationperSpecifications2.2.; gat and 3.3.6.

2. Otherwise, be in at least HOT SHUTDOWN within the next twelve (12) hours.

b. With no reactor coolant recirculation loops in operation:

1. Take the ACTION required by Specification 3.4.1.5, and

2. Be in at least HOT SHUTDOWN within the next six (6) hours.

C LA SALLE - UNIT 1 3/4 4-1 Amendment No. 94

27: 2 :0='E: :

• 57 4 !!.7 :N t.*W!?!

ia5ii "ne s:ecificat::ns of ints sect :n assure that the Dean claccing temperature

'::!c-in; :ne costulatec casign oasis loss of coolant,acc1 cent will not exceto

e ::::or :mit seecifiec in 13 CFR 50.46.

i 3/4 2.'.

AVERAGE DLANAR LINEAR WEAT GENERATION RATE

.This specification assures that the peak claccing temperature following

the costulateo cesign easts loss of-coolant accident will not exceed the limit soecifiec in 10 CFR 50.46. The specification also assures that fuel roc j

2 meenanical integrity is maintainea curing normal and transient coerations.

i i ace:cen: 'ne ceau c!accing temocrature (8CT' following a costulatec loss of-coolant i :s ortmart y a fune:1on of tr.e average neat generation rate of all i :ne :cs of a fuei assemoly at any axial location anc is cepencent only j sec:ncar ly on tne ree-to-ree power cistrioutton within an assemoly. The pean

iac temoerature is calculateo assuming a LNGR for the hignest poweren roc wnicn is toual to or less than the cesign LMGR correctec for censification .

ints LMGR times 1.02 is usea in the neatuo come along with the exocsure The Tecnnical Soecif' cation AVERAGE PLANAR LINEAR HE

( APLHGR) is inis LMGR of the nignest coweren roc divices by its local peaking fact:r

• i 1

However, tne current General Electric (GE) calculational mooels i ' CAFE /GEITA cescrioec in Reference 3), wnich are consistent with the '

eeovirements are n of Appenc1x K to 10 CFR 50, nave establisned that APLHGR values st111 exoectea to ce limiten Dy LOCA/ECCS considerations.  ;

APLNGR limits 'are 1

atn:a: ecuirec.

ec. newever. to assure tnat fuel roc mecnanical integrity is j ~;:e at : They are soecifiec for all resioent fuel types in the Core

mit Reoert casec on tne fusi thermal-mecnanical cesign analysis.

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e LA SALLE UNIT '

4 B 3/4 2-1 Amenoment No. 70 I

Paragraph #1 The purpose of the power and flow dependent MAPLHCR factors specified in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core now and core power conditions. At less than 100% of rated flow or rated power, the required MAPLHCR is the minimum of either (a) the product of the rated MAPLHCR limit and the power, dependent MAPLHCR factor or (b) the product of the rated MAPLHCR limit and the flow dependent MAPLHCR factor. The power and flow-dependent MAPLHCR factors assure that the fuel remains within the fuel design basis during transients at off rated conditions. Methodology for establishing these factors is described in Reference 6 .

I i

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DOWER DISTRIBU'ICN Sv5T!ws c

BASES k 3/a.2.2 k ~ ~* - ~

on Tne fuel ciaccing integrity Safety Limits of Specification 2.1 sea Inr cistribution wnten would yield the design LHGR POWER. u 4HERMAL w Dia' sea simulatec thermal power up scram setting anc control roc bloc ctions of tne APRM ins loop operation and sing irculati ts for both two recirculation op operation must be adjustec to ensure that the MCPR coes not less than the fuel claccing safety limit or that 1 1% plastic stra' es not oc ne degraded situation. The scram settings and oct settings are adjuste '

mula in this corcance with the for-incica . fication wnen the combination of THERMA nc M LPD

, w nigner peaxed power cistricution to ensure that an LHG nt i

not be increasec in tne cegraced concition.

_3 /4. 2. 3 w!NIMUM CR:~! CAL DOWER RATIO The reouired coerating limit MCPRs at steady state operating concitions as specifiec in Specification 3.2.3 are cerivec from the establisnec fuel tclaccing integrity Safety Limit MCPR, anc an analysis of abnormal operational rans ients.

For any abnormal operating transient analysis evaluation with the initial concition of the reactor Deing at the staaey state operating limit ,

l MCPR at any time curing the tranzient assuming instr Specification 2.2. .

To assure that the fuel claccing integrity Safety Limit is not exceeded during any anticipatec abnormal operational transient, the most limiting transients have caen analyzec to cetermine which result in the largest reduc-tien in CRITICAL POWER RATIO (CPR).

of flow, The type of transients evaluatec were loss coolant temperature cecrease. increase in pressure and power, positive re' act MCPR. The limiting transient yields the largest celta aimit MCPR of Specification 3.2.3 is cotainec and presentee in OPERATING LIMITS REPORT.

Analyses have been perfomed to determine the effects on CRITICAL POWER RATIO (CPR) during a transient assuming that certain equipment is out of servic A detailed description of the analyses is provided in Reference 5. The anal-yses service. performed assumed a single failure only and establishe transient analyses input assumptions:The following single equipment failu

1) main turbine bypass system out of service,

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

i

, ;0WER C

3?AIBUTf0N SYSTEMS iASEI

" N uuu CRIT::AL DOWER RAT *: (Continuto) i{ Ne value

nservat1<e forfor to useo inreason:

ine"following 50ecification 3.2.3 is 0.687 seconos wnien is i

or simolicity in formulating anc implementing the LCO, a conservative i "

i value for IN i=1 ; of 598 was useo. This represents one full core cata set k

1 at B0C plus one full core cata set following a 120 day . outage clus twelve 10% of core, 19 roos, cata sets.

The 12 cata sets are eouivalent to 24 coerating montns of surveillance at the increaseo surveillance 9ecuency of one set oer 60 days repuirec ey tne action statements of J

icec1fications 3.'.3.2 . ano 3.1.3.4 i That is, a cycle lengtn was assumeo which is longer than any past or

! :entemoiateo refueling interval ano the number of roos tested was maxis 12ed

] in.orcer to simolify and conservatively reduce the criteria for the scraa tice ,

at wnien MCPR penalization is necessary. {

The ourpose of the K f i factor specified in the CORE OPERATING LIMITS 7 i REPORT is to define opera (ing limits at other than rated core flow conditions.1 -

At less tnan 100% of rated flow, the required MCPR is the product of the MCPR I

.t and the 4 factor. The K 7 factor assures that the Safety Limit MCPR will not  ;

be violataa.

Q ference a. Methocology for establishing the fK factor is described in l

(< En. serf" farey-joh *3 >> i i At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POW j the reactor will be coerating at minimum recirculation puso speed ano the 1

mocerator void content will be very small. For all designated control rod oatterns wnien may oe employeo at this point, operating plant experience j inotcates that consideraole the resulting MCPR value is in excess of requirements by a margin.

During initial start up testing of the plant, a MCPR

! evaluation will be recirculation pump speed.mace at 25% of RATED THERMAL POWER level with m i The MCPR sargin will thus ce demonstrated such that 1

! The caily reevirement for calculating MCPR when THER

, or eoual to 25% of RATED THERMAL POWER is sufficient since power aistribution

snif ts are very slow when there nave not been significant power or control rod enanges.

4 . The reouirement for calculating MCPR when a limiting control rod a oattern is approacneo ensures inat MCPR will be known following a change in THERMALatDOWER operation a tnermal orlimit.

power snape, regardless of magnituce, that could place i

l .

l LA SALLE UNIT 1 B 3/4 2-5 Amenoment No. 70 4

i l

Paragraph #2 The purpose of the power and flow dependent MCPR limits specified in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow and core power conditions. At a given power and flow operating condition, the l

required MCPR is the maximum of either the power dependent MCPR limit or the l flow-dependent MCPR limit. The required MCPR assures that the Safety Limit MCPR will not be violated. Methodology for establishing the power- and flow dependent MCPR limits is described in Reference 6.

I l

l l

. l e

m

i l , POWER DISTRIBUTION SYSTEMS

). BASES 3/4.2.4 LINEAR HEAT GENERATION RATE

. The specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in j 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 j fuel rod exceeds the design LINEAR HEAT GENERATION RATE due to power spiking.

2

References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NED0-20566A,

! September 1986.

) 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-1 plemented by letter dated September 5, 1980, from R. H. Buchholz (GE) to P. S. Check-(NRC)..

1 3. "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-

{

2 Coolant Accident Analysis," General Electric Company Report NEDC-32258P, October 1993. ~

l 4. " General Electric Standard Application for Reactor Fuel,"

NEDE-240ll-P-A, (latest approved revision).

0

5. " Extended Operating Domain and Equipment Out-of-Service for LaSalle County Nuclear Station Units 1 and 2;" NEDC-31455, November 1987.

{ g.

i "}tTS &mmur %ym thtps le La%Ue Mut Gm7 i

142.," GJ Ekree Co. R ep,n- NEDC - s t s'3/ P, Dee A Rt3 b

4 j '

i

~

1 .

LA$ALLEUNIT1 8 3/4 2-6 hauuhant No. %

j l

1 1 -

i -

INSTRUMENTATION f n

4 N

i sAsEs 3/4. 3. 5 REACTOR CORE ISOLATION C0OLING SYSTEM ACTUATION INSTRUMEKTATION The reactor core isolation cooling system actuation instrumentation is

) provided to initiate actions to assure adequate core cooling in the event of j reactor isolation from its primary heat sink and the loss of feedwater flow to j the reactor vessel without providing actuation of any of the emergency core

{ cooling equipment.

1 .

! 3/4. 3. 6 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION -

4 The control a block functions are provided consistent with the require- -

l ments of na w ,w .inastin Section 3/4.1.4, Control Rod Program Controls i _ ction 3/4.2 Powei _.stribution Limits 3 The trip logic is arranged so that a

trip in any one of sne inputs will result in a control rod block.

k 3/4.3.7 MONITORING INSTRUMENTATION

! 3 /4. 3. 7.1 RADIATION MONITORI Q 'TRUMENTATION i

The OPERABILITY of the raai sion monitoring instrumentation ensures that; (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when

the radiation level trip setp61nt is exceeded.

j 3.4.3.7.2 SEISMIC M3NITORING INSTRUMENTATION

~

4 The OPERABILITY of the saismic monitoring instrumentation ensures that suffic-i isnt capability is availabit a oromptly detemine .the magnitude of a seismic event

and evaluate the response of see features important to safety. This capability j

l is required to pemit comparison of the measured response to that used in the

. design basis for the unit. This instrumentation is consistent with the recommen-

dations of Regulatory Guide 1.12 " Instrumentation for Earthquakes", April 1974.

} 3/4.3.7.3 METEOROLOGICAL ENITORING INSTRUMENTATION i

j The OPERABILITY of the meteorological monitoring instrumentation ensures i

that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release of

}

i radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.

j This instrumentation is consistant with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February,1972.

l 3/4.3.7.4 REMOTE SHUTDOWN MONITORING INSTRUMENTATION 4

The OPERABILITY of the remote shutdown monitoring instrumentation ensures i that sufficient capability is available to pemit shutdown and maintenance of .

i HOT SHUTDOWN of the unit from locations outside of the control room. This i

capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50. -

LA SALLE - UNIT 1 (3/43-4 Amendment No. 58 4

5

l' ADMINISTRATIVE CONTROLS

Semiannual Radioactive Effluent Release Report (Continued) 4 Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective. In addition, a report of any major changes to i

' the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the perico in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.

6. Core Operating Limits Report l 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:

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

j (2) The minimum Critical Power Ratio (MCPR) (includino 205 p

' 3pw2r gas) No scram.t_ime, tau (th dependent MCPR limits, and G G endent

M M,W.CPR n**zr-t 2 "

==)

for Technical Specmcation '

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

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

3 b.

j" The analytical methods used to determinit the core operating limits shall be these previously reviewed and approved by the NRC in the latest approved revision or supplement of the topical reports describing the methodology. For LaSalle County Station Unit 1, the topical reports are:

1 (1) NEDE-240ll-P-A, " General Electric Standard Application for

' Reactor Fuel," (latest approved revision).

1 (2) Commonwealth Edison Topical Report NFSR-0085, " Benchmark of j

I BWR Nuclear Design Methods," (latest approved revision).

1 i (3) Commonwealth Edison Topical Report NFSR-0085, Supplement 1, i " Benchmark of BWR Nuclear Design Methods - Quad Cities Samma Scan Cumparisons," (latest approved revision). l 2

(4) Cosuoonwealth Edison Topical Report NFSR-0085, Supplement 2,

" Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).

i .

j LA SALLE UNIT 1 6-25 l Amendment No. N 86

1

! INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

SECTION PAGE 3/4.0 APPLICABILITY...................................'................ 3/4 0-1

) 3/4.1 REACTIVITY CONTROL SYSTEMS i

j 3/4.1.1 SHUTDOWN MARGIN.............................................. 3/4 1-1 j 3/4.1.2 REACTIVITY AN0MALIES......................................... 3/4 1-2 l 3/4.1.3 CONTROL RODS 1

4

. Control Rod Operability...................................... 3/4 1-3 4

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

Control Rod Average Scram Insertion Times.................... 3/4 1-7 j Four Control Rod Group Scras Insertion Times. . . . . . . . . . . . . . . . . 3/4 1-8 Control Rod Scram Accumulators............................... 3/4 1-9 3 Control Rod Drive Coupling................................... 3/4 1- n

Control Rod Posi ti on Indi cation. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-13 Control Rod Drive Housing Support............................ 3/4 1-15 3/4.1.4 CONTROL ROD PROGRAN CONTROLS Rod Worth Minimizer.......................................... 3/4 1-16 i

Rod Block Monitor............................................ 3/4 1-18 i 3/4.1.5 STAND 8Y 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 d

3/4.2.1 AVERAGE PLANAR LINEAll HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . 3/4 2-1 3/4.2.2 ce .. _Q&. . . . . . [. .E7Ep)

, ... m ................................ 3/4 2,2

, 3/4.2.3 NINIMM CRITICAL POWER RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-3 4 .

3/4.2.4 LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-5 LA SALLE - UNIT 2 IV Amendment No. 73

7j ,w

.g= .

. ~ :' ~ .m f.TL.':i - ?.-

= g__ .-,

r.

1 .

INDEX  :

BASES l

e i SECTION PAGE j 3/4.0 APPLICABILITY................................................... B 3/4 0-1 l 3/4.1 REACTIVITY CONT'ROL SYSTEMS 3/4.1.1 SNuT00WN mRGIN......................................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0M LIES.................................... B 3/4 1-1 3/4.1.3 CONTROL R005............................................ B 3/4 1-2

. 3/4.1.4 CONTROL R00 PROGRAM C0NTROLS............................ B 3/4 1-3 J

j 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 3/4.2 POWER DISTRIBUTION LIMITS

} 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............. B 3/4 2-1 i

1 3/4.2.2($RNSETP0d.[..M.LEf.k.....................

B 3/4 2.-2 ,

3/4.2.3 MINIMUM CRITICAL POWER RATI0............................ B 3/4 2-2 3/4.2.4 LINEAR NEAT GENERATION RATE............................. B 3/4 2-6 4

3/4.3 INSTRUMENTATION .

3/4.3.1 REACTOR PROTECTION SYSTEM, INSTRUMENTATION............... B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION..................... B 3/4 3-2 i 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION i

i INSTRUMENTATION......................................... B 3/4 3-2

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

} B 3/4 3-3 i 3/4.3.5 REACTOR CDRE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION......................................... B 3/4 3-4

! 3/4.3.6 CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION............B 3/4 3-4 1 3/4.3.7 MONITORING INSTRLMENTATION Radiation Monitoring Instroentation.................... B 3/4 3-4 -

, Seismic Monitoring Instr uentation..............t........ B 3/4 3-4

! LA SALLE - UNIT 2 XII Amendment No. 69 ,

i . /:

dMITING 5AFETY SYSTEM SE--' NGS BASE!

REACTOR DROTECT!ON SYSTEM !NSTRUMENTATION SETPOINTS* (Continuec)

Averace Power Rance Monitor (Continuea) the flux cistribution associated with uniform rod withdrawals does not involve .

nign local ceans anc because several rocs must be movea to enange power by a significant amount, the rate of. power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumec uniform rod witnerawal approacn to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER oer minute and the APRM system would be more than aceouate to assure snutcown before the power could exceed the Safety Limit.

The 15% neutron flux trip remains active until the mooe switch is placed in the Run position.

The APRM trio system is calibrated using heat balance cata taken curing steacy state'conaitions. Fission enameers provide the easic input to the system ano therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Fixed Neutron Flux-High 1185 setpoint; i.e. for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow Biased Simulated Thermal

• Power-Upscale setooint, a time constant of 6 21 seconds is introduced into the i flow ciasec APRM in orcer to simulate the fuel thermal transient characteristics.  !

A more conservative maximum value is used for the flow biased setpoint as snown in Table 2.2.1-1.

• The APRM setooints were selected to provide 'aceouate margin for the Safety Limits and yet a)1w na--='ina marcin that reduces the nossibility of  ;

unnecessary snutcown. 1Tht,-flow referencac trip setootnt must be adjusten by 7)E"

the specifimo when MFLPD isformula in Specification greater than or equal to FRTP. 3.2.2 in oroer to maintain these marginsj

3. Reactor vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process carrier resulting in the release of fission products. A pressure increase = nile operating will also tend to increase the power.of the reactor by compressing voics thus anoing reactivity. The trio will' quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly nigner than the operating pressure to permit normal operation without spurious tries.

The setting provices for a wioe margin to tne maximum allowanle oesign Dressure ano takes into account the location of the pressure measurement

omoarea to the nignest oressure that occurs .in the system curing a transient.

his trio setootnt is effective at low power / flow conditions wnen the turnine stoo valve closure trio is oyoassee. r or a turoine trip unoer tnese conditions, tne limit.transient analysis i,cicatec an aceouate margin to tne tnermai nyaraulic

.A 5ALLE - JNIT : 3 2-10

, -,-e e--.-,. , . . . ,, , . , - - . - ye-w.p.g,.r.r,.w,, a

' PAGE % 5~E ~

3/9',2.~,t z/rrEpiomLY t-EFT GL 6'K owe; ::5 a:BUT:CN L:MITS 2h 2.

p APRW SETD0!NTS 4 Jht, c ed'ml ** /*78 7)

.:" IN: 03NDIT!0N FOR OPERA 10N 3.2.2 he APRM flow biaseo simulatec thermal power-upscale scram trip s potnt (5) anc low ciasec simulated thermal power upscale control roc elect set:oin in

.g) snall be estaelisnec accorcing to the following relat' nships:

a. T Recirculation Loop Operation i

5I s than or equal to (0.58W + 59%)T S 1 gg s than or ecual to (0.58W + 47%)T

b. Single circulation Loop Operation 5 less tha or ecual to (0.58W + 54.3%)T 5;g less tna or ecual to (0.58W

• 42.3%)

.ncre: 5'ano SRB are in ercent of RATED THERMAL OWER,

= = Loop recircula 1on flow as a percen age of the loop recirculation flow whien proc es a ratec core 3 ow of 108.5 million lbs/hr, T = Lowest value of t e ratio of F ' ION OF RATED THERMAL POWER civiene by tne MAX UM FRACTIO OF LIMITING POWER DENSITY or the value 1.0. T is al s less an or equal to 1.

APPLICABILITY: OPERATIONAL CONDI ON when THERMAL POWER is greater than or ecual to 2% of RATED THERMAL POWE ,

ACT:0N:

With the APRM flow Diased simulat the al power upscale scram trip setooint anc/or tne flow biased simulatec hermal ower voscale control rod block trip setooint set less conservativel than 5 or as above determined, initiate corrective action witnin 15 m' utes and res Ie,5and/ ors recuirec limits" within 2 ho rs or reduce TH 6(AL POWER to RR t' *IthI" th'

, RATED THERMAL POWER within Tess than 25% of he next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REOUIREMENT s

4.2.2 The FRTP and value of T calculat , eand MFLPD for each class of fuel all be determined, the the most recent actual APRM low biased simulated thermal power upse e scram and control rod block trip s cint verified to be within the above iaits or adjusted, as required:

a. At I st once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, '

b. Wi in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER crease of at I

st 15% of RATED THERMAL POWER, and c.

nitially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reacto is operating with MFLPD greater than or equal to FRTP.

. Witn NLPD greater than the FRTP up to 90% of RATED THERMAL POWER, r than rat adjysting,the APRM setpoints, the APRM gain may be adjusted such that Aread-ings are greater than or equal to 100% times MFLPD, provided that ted the ad; RM reading does not exceed 100% of RATED THERMAL POWER, the repuired ga -

djustmentincrementdoesnotexceed10%ofRATEDTHERMALPOWER,andanotici of the adjustment is posted on the reactor control panel.

LA SALI.E - UNIT 2 3/4 2-2 Amendment No.54 l

, TABLE 4.3.1.1-1 (Continued) g REACTOR PROTECTION SYSTEN INSTRUMENTATION SURVElllANCE REQUIREMENTS CHANNEL OPERATIONAL h"

CHANNEL FUNCTIONAL CHANNEL CON 0lil0NS FOR WHICH FUNCTIONAL UNIT CHECK TEST CAllBRATION SURVEILLANCE REQUIRED E 8. Scram Discharge Volume Water a level - High NA M R 1, 2, 5 m 9. Turbine Stop Valve - Closure NA Q R ,

I I

10. Turbine Control Valve Fast ,

Closure Valve Trip System 011 Pressure - Low NA Q .R 1 I

11. Reactor Mode Switch Shutdown Position NA R NA 1,2,3,4,5

12. Manual Scram NA W NA 1,2,3,4,5 1

13. Control Rod Drive

a. Charging Water Header Pressure - Low NA M R 2, 5

b. Delay Timer NA M R 2, 5 s.

?'

o (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

(b) The IRM, and SRM channels shall be determined to overlap for at least 1/2 decades during each startup and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.

(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power levels '

calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 2 25% of RATED THERMAL POWER. The APRM Gain Adjustment Factor (GAF) for any channel shall.be equal to the power value deter-

mined by the heat balance divided by the APRM reading for that channel.

g Within 2 h r adjust any APRM channel with a GAF > 1.02. In addition, adjust any APRM channel within

' if power it ar==t-r than or eoval to 90% of RATED THERMAL POWER and the APRM cha m 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (1 AF is

< 0.98 or g; ir power is less than 90% of RATED IMtRRAL POWER and the APRM readi s i alue determined by the heat balance by more than 10% of RATED THERMAL POWER. Untti any re d APRM

'I adjustment has been accomplished, notification shall be posted on the reactor control painel.

g (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a i - calibrated flow signal.

g The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH).

4( Measure and compare core flow to rated core flow, j t This calibration shall consist of verifying the e i I second simulated thermal power time constant.

4 .

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f LASALI.E'- UNIT 2 3/4 3-53 Amendment No. 54 9

w ~ -.w ww~,-w-,

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== me

- =

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3T s= s=

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3/4.4 REACTOR COOLANT SYSTEM l 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 ACTION

a. With only one (1) reactor coolant system recirculation loop in operation, comply with Specification 3.4.1.5 and:

l 1. Within four (4) hours:

l a) Place the recirculation flow control system in the Master Manual mode or lower, and l b) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety  ;

l Limit by 0.01 to 1.08 per Specification 2.1.2, and c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation by 0.01 per Specification 3.2.3,

{ and, l

l d) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable to single 4

recirculation loop operation

• per Specifications 2.2.l(,7 -

and 3.3.6.

2. Otherwise, be la at least HOT SHUTDOWN within the next twelve (12) hours,

b. With no reactor coolant recirculation loops in operation:

l

1. Take the ACTION required by Specification 3.4.1.5, and l

2. Be in at least HOT SHUTDOWN within the next six (6) hours.

6 LA SALLE - UNIT 2 3/4 4-1 Amendment No. 78

' f.

3 'a 2 20wE4 O!!TRIBUT!ON L:Mf?S BASE *

'ne specifications of inis section assure that the mean claccing temetrature following the costulatec cesign casis less-of-coolant accident

-ii' not exceto int 2200*F iimit specifiec in 10 CFR.50.46.

3/a 2.1 AVERAGE DLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the costulateo cesign cas1s loss-of coolant accident will not exceed the limit soecifico in 10 CFR 50.46. This specification also assures that fuel rod meenanical integrity is maintaineo curing normal ano transient coerations.

The cean claccing temperature (PCT) following a postulateo loss-of coolant accicent is crimartiy a funciten of the average heat generation rate of all the roos of a fuei assemoly at any axial location and is cepencent only clac temoerature is calculatec assuming a LHGR for the hig wnten is ecual to or less tnan the design LHGR corrected for densification .

This LMGR times 1.02 is usec in the heatup code along with the exoosure oeoencent steacy-state gap conductance and red-to-red local peaking factor.

The Tecnnical Specification AVERAGE PLANAR LINEAR HEAT GENERATIO

,i is tats LHGR of the nignest powered rod divided by its local peaking factor. ~/

However, the current General Electric (GE) calculational mecals '

(SAFER /GESTR oescritto in Reference 3), which are consistent with the recuirements of ADoenoix K to 10 CFR 50, have established that APLHGR values are not exoected to be limited by LOCA/ECCS considerations. .

APLHGR limits a're  :

stillt e.ta ma recuitec.

i nea. newever, to assure that fuel rod mechanics 1 integrity is They are specifieo for all resident fuel types in the Core ,

. Goerating Limit Report caseo on tne fuel thermal meenanical design analysis.

$r) Sed S*fl* lO a

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

e Paragraph #1 The purpose of the power and flow dependent MAPLHGR factors specified in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow and core power conditions. At less than 100% of rated flow or rated power, the required MAPLHGR is the minimum of either (a) the product of the rated MAPLHGR limit and the power. dependent MAPLHGR factor or (b).the product of the rated MAPLHCR limit and the flow dependent MAPLHCR factor. The power and flow-dependent MAPLHGR factors assure that the fuel remains within the fuel design basis during transients at off rated conditions. Methodology for establishing these factors is described in R fe erence 6.

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t Po s C?! afBUT?ON SYSTEMS _ _ _ _

BASE 5 3/4.2.2 C 7 Z' T "' D The fuel claccing integrity Safety Limits of Specification 2.1 were on a

• districution wnich would yield the design LHGR at RAT F L POWER. The iased simulatec tnermal power upscal setting ano con-trol roc block func. of tne APRM instrumen etn two recirculation loop coeration and single r lation operation must be adjusted to ensure that the MCPR does not Decome le ne fuel cladding safety limit or tnat 1 1*. plastic strain coes n cur in the situation. The scram settings and rod block setti re adjusted in accordance wi formula in this spect-fication wne peakee comoination of THERMAL POWER and MFLPD inct higher r distribution to ensure snat an LHGR transient would not c m

1 asec in the cegradec concition.

3/4.2.3 WINIMUM CRIT: CAL POWER RATIO i Tne reouirec operating limit MCPRs at steady-state oper: ting concitions as spec 1fiec in Specification 3.2.3 are derivec from the established fuel claccing integrity Safety Limit MCPR and an analysis of abnormal operational transients. For any annormal operating transient analysis evaluation with the l

initial condition of the reactor being at the steady-state operating limit, it l is reouired that the resulting MCPR does not decrease below the Safety Limit  !

MCPR at any time in Specification curing the transient assuming instrument trip setting given 2.2.

To assure that tne fuel claccing integrity Safety Limit is not excencea l curing any anticipatec abnormal operational transient, the most limiting transients nave been analyzed to determine which result in the largest recuc-  !

tien in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss l of flow, increase in pressure anc power, positive reactivity insertion, and coolant temoerature decrease. The limiting transient yields the largest celta MCPR.

. limit MCPR of Specification 3.2.3 is cotained and presented in the OPERATING LIMITS REPORT.

Analyses have been performed to determine the effects on CRITICAL POWER RATIO service.

(CPR) during a transient assuming that certain equipment is out of A cetailed description of the analyses is provided in Reference 5.

The analyses performed assumed a single failure only and established the licensing bases equipment out ofto allow continuous plant operation with the analyzed service. The following single equipment failures are included are part of the transient analyses input assumptions:

1.

main turbine bypass systes out of service, 2.

recirculation pump trip system out of service, LA SALLE - UNIT 2 3 3/4 2 2 Amendment No. 54

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0Wis -*! RIBUTTON Sv!TEWS iA5!!

u w!Cw w MIT: CAL DOWER RATIO (Continuto)

'te value for , useo in Specification 3.2.2 is 0.687 seconos wnien is

nse-.ative for ine following reason: '

'or simolicityn in formulating anc implementing the LCO, a conservative value for IN i=1 ; of 598 was used. This represents one full core cata set at BOC plus one full core data set following a 120 day outage plus twelve 1C*. of core,19 roos, cata sets.

The 12 data sets are eouivalent to 24 coerating montns of surveillance at the increased surveillance frecuency of one set oer 60 days reouired by tne action statements of Soec:fications 2.1.3.2 ane 3.1.3.4.

~

na; 15, a cycle length was assumed which is longer than any past or

r. 9 tem:iateo refueling interval and the nunoer of roos tested was maximizac or:er to simolify ano conservatively reouce the criteria for tne scram time at unica MCPR penalization is necessary.

(REPORT conottions.

is to cefine operating limits at other than rated c i

of the MCPR ano the K, factor.At The K less than 100% of rated flow, the reovirea MC MCPR will not be violitec. Methocolog,y for estaclishino the X, facter isfactor a .

oescrioeo in Reference a gg gg l.

At THERMAL POWER levels less than or eoual t!25% of RATED tne reactor will be operating at minimum recirculation puso speed anc the ,

mocerat:r void content will be very small. For all designatec control red

atterns nien may ce emoloyeo at this point, operating plant exoerience not:ates that

ns:cersole the resulting MCPR value is in excess of recuirements by a margin.

During initial start uo testing of the plant, a MCPR evaiwatt:n recirculation puso willspeed.

be mace at 25% of RATED THERMAL POWER level The MCPR margin will thus be demonstrated such that The caily reouirement for calculating MCPR when THE

r toual to 25% of RATED THERMAL POWER is sufficient since power distribution snifts are

nanges. very slow when there have not been significant power or control roc The recuirement for calculating MCPR when a limiting control rod pattern is approacneo ersures that MCPR will be known following a change in

'HERMALatPOWER operation a thermalorlimit.power shace, regardless of magnitvoe, that could place

' A SALLE - UNIT 2 8 3/4 2-5 Amenoment No. 54

Paragraph #2 i

The purpose of the power and flow dependent MCPR limits specified in the CORE

' OPERATING LIMITS REPORT is to define operating limits at other than rated core flow and core power conditions. At a given power and flow operating condition, the required MCPR is the maximum of either the power dependent MCPR limit or the flow dependent MCPR limit. The required MCPR assures that the Safety Limit MCPR will not be violated. Methodology for establishing the power and flow dependent MCPR limits is described in Ref trence 6.

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POWER DISTRIBUTION SYSTEMS

BASES l

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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 1

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 i core bottom and top and assures with a 95% confidence that no more than one i

fuel rod exceeds the design LINEAR HEAT GENERATION RATE due to power spiking.

i

References:

1. General Electric Company Analytical Model for Loss-of-Coolant ,

Analysis in Accordance with 10 CFR 50, Appendix K, NED0-20566A, j September 1986.

2. " Qualification of the One-Dimensional Core Transient Model for i

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 I and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," General Electric Company Report l

, NEDC-32258P, October 1993. .

1

! 4. " General Electric Standard Application for Reactor Fuel,"

l NEDE-240ll-P-A, (latest approved revision).

1

, i 1 S. " Extended Operating Domain and Equipment Out-of-Service for L'aSalle County Nuclear Station Units 1 and 2;" NEDC-31455, November 1987.

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LA SALLE - UNIT 2 83I42-6 Amendment No. 80 1 - - -

l

'- l iltSTRUNDffAUON '

SASE5 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEN ACTUATION INSTRUNDfTAMO The reactor core isolation cooling system actuation instroentation is provided to initista actions to assure adequate core cooling in the event of I reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of sty of the emergency core cooling equipment. ,

, 3/4.3.8 CONTROL RDO WITHORAWAL BLOCK IItSTRUNENTATION

' ~

- l The control red black functions sre provided consistant with the requirements of the Wficaticry*in Section 3/4.1.4, Control Rod Program  ;

controls-dsd w- tragpower psstribution LimTED The tHp logic is arranged so that a trip in any one er T,ne inputs w1D1 result in a control red l' block.

F4.3.7 NONITORING INST *W SfTATIOff ' \

W4.3.7.1 RADIATION DONITORING INST *LaSf7ATION

• l The'0PERA81LITY of the radiation monitoHng instroentation ensures that; (1) the radiation levels are continually measured in the areas served by the -

- individual channels, and (2) thei alars or automatic action is initiated when ,

the radiation level trip setpoint is exceeded. .. i 3.4.3.7.2 SEISNIC DONITORING IllSTRUDENTATION *

~

1 I

The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the angnitude of a i seismic event and evalusta the response of those features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for the unit.

wtth the recommendations of Regulatory Guide 1.12 " Instrumentation for This instrumentation is consistant Earthquakes" April 1974.

W4.3.7.3 IET*n'*m3 CAL ICHITORING INSTRUPSITATI0ff The OPERABILITY of the meteorelegical monitoMag instrumentation ensures that sufficiest noteorelegical esta is available for estiasting potential radiation deses to the pelic as a result of routine er accidental release of radianctive mateHals to the atmosphere. This capability is required to evolusta the need for fattisting pretmetive measures to protect the health and safety of the pelic. This instrumentation is consistant with the recesumende- C tions 1972. of Regulatory Guide L23 "Oneita Meteorological Programs," February, W4. 3. 7. 4 RDCTE SWT'"" IONITORIIE IltST*L*SfTATICII The OPERABILITY of sne remote shutdown monitedng instrumentation ensures '

that sufficient capability is available to permit shutdown and maintenance of ICT SWTDOWN of the unit from locations outside of the control room. This capability is required in the event censistant with General Design Criteria 19 of la CFR 50.

control reos habitability is lost arp is .

LA SALLI - UNIT 2 8 3/4 3-4 Amendment lie. 41 r - .- ,- ~ -

) l ADMINISTRATION CONTROLS i

R.-i "W.YdI$ '$:  ;'x x Report (Continued)

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

4

- - x __

(2) Ther minimum Critical Power Ratio (MCPR) (includi

! W.f/g i time, tau h), depender depeoowr ,,,3. " = p ; } = =nt MCPR

• • W== limits, and I

(3) The Linear Heat Gen Rate (LNGR) 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  !

3

limits shall be those previously reviewed and approved by the I

" NRC in the latest approved revision or supplement of the topical reports describing the methodology. For LaSalle County Station Unit 2, the topical reports are:

I (1) NEDE-24011-P-A " General Electric Standard Application for i

j Reactor Fuel,",(latest approved revision).

1 (2) Commonwealth Edison Topical Report NFSR-0085, " Benchmark of

• l BWR Nuclear Design Methods," (latest approved revision).

1

(3) Commonwealth Edison Topical Report NFSR-0085, Supplement 1,

" Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma

{ Scan Comparisons," (latest approved revision).

1 l

I

(4) Consonwealth Edison Topical Report NFSR-0085, Supplement 2,

" Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).

c.

The core operating limits shall be determined so that all j applicable limits (e.g., fuel thereal-sechanical limits, core j

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

The CORE OPERATING LIMITS REPORT, including any mid-cycle 3

4 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.

) 8. Deleted.

1 .

LA SALLE - UNIT 2 6-25 Amendeant No. 68

\

ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10 CFR 50.92 (c), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

The proposed chances do not involve a significant increase in the orobability or conseauences of an accident previously evaluated because:

The probability of an accident previously evaluated will not increase as a result of this change, because no changes to plant systems will occur. All changes are related to core monitoring software, and there will be no physical changes to equipment.

The consequences of an accident previously evaluated will not increase as a result of the proposed changes. The power- and flow-dependent MCPR and MAPLHGR limits incorporate sufficient conservatism so the safety limit MCPR (operating limit MCPR for automatic flow control) and the fuel thermal-mechanical limits will not be violated for any power and flow condition. Because these limits are protected during normal operation, the consequences of any transient will not increase with this change in limit definition.

General Electric has verified in Attachment E that the introduction of Arts will not cause any change in the Licensing Basis PCT resulting from a Loss-Of-Coolant Accident, nor any change in the results satisfying the other LOCA acceptance criteria of 10CFR 50.46 '

and Section 15.6.5 of NUREG-0800 (Standard Review Plan), which are: cladding oxidation, metal-water reaction (hydrogen generation), coolable geometry and long-term cooling.

The proposed changes do not create the possibility of a new or different kind of accident from any accident previousiv evaluated because:

Since no physical changes to any plant system are occurring, there will be no new or different types of accidents created by this change. . No interactions between equipment systems will be changed in any manner.

1he_p_mposed changes do not involve a significant reduction in a margin of safety because:

The power- and flow-dependent MCPR and MAPLHGR limits will sufficiently protect the SLMCPR (OLMCPR for automatic flow control) and the fuel thermal-mechanical limits at all power and flow conditions. The . ARTS limits conservatively assure that all licensing criteria are satisfied without setdown of the flow referenced APRM scram and rod block trips. The limits were developed using NRC approved methods, and satisfy the same NRC approved criteria that the APRM setdown requirement does.

K:\NLA\LASALLE\ARTSRE2:11

ATTACHMENT C  ;

i EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION i l

l CONCLUSION Guidance has been provided in " Final Procedures and Standards on No Significant l Hazards Considerations," Final Rule,51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are not considered likely to involve significant hazards considerations. This proposed amendment most closely fits the example of a change which may either result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan, section 4.4, " Thermal and Hydraulic Design".

This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations.

Therefore, based on the guidance provided in the Federal Register and the criteria .

established in 10 CFR 50.92 (c), the proposed change does not constitute a significant hazards consideration.

1 Ki\NLA\LASALLE\ARTSPE2:12

ATTACHMENT D i

ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW l

Commonwealth Edison has evaluated the proposed amendment against the criteria for l identification oflicensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided under 10 CFR 51.22 (c)(9). This conclusion has been determined because the changes requested do not pose significant i hazards consideration or do not involve a significant increase in the amounts, and no l significant changes in the types, of any effluents that may be released offsite.

l Additionally, this request does not involve a significant increase in individual or l cumulative occupational radiation exposure.

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Y. : \NLA\ LASALLE\ ARTSRE2 : 13 l

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ATTACHMENT E l

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I GE ARTS ANALYSIS i I

FOR LASALLE COUNTY STATION UNITS 1 AND 2 K:\NLA\LASALLE\ARTSRE2:14

ATTACHMENT F WITHOLDING AFFIDAVIT FOR GENERAL ELECTRIC ARTS ANALYSIS REPORT Ki\NLA\LASALLE\ARTSRE2:15

-- - - - . - - - ._ - - ~ . _ . .- .- -

O General Electric Company l l

AFFIDAVIT 1

I, David J. Robare,being duly sworn, depose and state as follows:

(1) I am Project Manager, Plant Licensing / Renewal Projects, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the GE proprietary report NEDC-31531P, ARTS Improvement ProgramAnalysisforLaSalle County Station Units 1 and 2, Class III (GE Company Proprietary Information), dated December 1993. The proprietary information is delineated by bars marked in the margin adjacent to the specific material.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom ofInformation Act ("FOIA"),5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4),

2.790(a)(4), and 2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information", and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Group v.FDA.

704F2dl280 (DC Cir.1983).

(4) Some examples of categories ofinformation which fit into the definition of -)

proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which,if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; Affidavit Page i

... i

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.

(5) The information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by .

GE, and is in fact so held. The information sought to be withheld has, to the l best of my knowledge and belief, consistently been held in confidence by GE,  !

no public disclosure has been made, and it is not available in public sources. l All disclosures to third parties including any required transmittals to NRC, ,

have been made, or must be made, pursuant to regulatory provisions or l proprietary agreements which provide for maintenance of the information in i confidence. Its initial designation as proprietary information, and the i subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the l' manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical models, methods and processes, including computer codes, which GE has developed, obtained NRC approval of, and applied to perform evaluations of plant improvements to Affidavit Page 2

,e p .

increase operational flexibility and efficiency for the BWR.

I

The development and approval of the BWR loss-of-coolant accident analysis

! computer codes used in this analysis was achieved at a significant cost, on the

order of several million dollars, to GE.

The development of the evaluation process along with the interpretation and.

i application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

i

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the e availability of profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commercial value a extends beyond the original development cost. The value of the technology l base goes beyond the extensive physical database and analytical methodology l and includes development of the expertise to determine and . apply the 4

appropriate evaluation process. In addition, the technology base includes the

] value derived from providing analyses done with NRC-approved methods.

i

! The research, development, engineering, analytical and NRC review costs

! comprise a substantial investment of time and money by GE.

1 The precise value of the expertise to devise an evaluation process and apply

the correct analytical methodology is difficult' to quantify,' but it clearly is j substantial.

I GE's competitive advantage willbe lost ifits competitors are able to use the

results of the GE experience to normalize or verify their own process or if

, they are able to claim an equivalent understanding by demonstrating that they-i can arrive at the same or similar conclusions.

a l The value of this information to GE would be lost if the information were j disclosed to the public. Making such information available to competitors I

without their having been required to undertake a similar expenditure of

resources would unfairly provide competitors with a windfall, and deprive GE j of the opportunity to exercise its competitive advantage- to seek an adequate j'

return on its large investment in developing theselvery valuable analytical tools.

i I

Afridavit Page 3 i

,,c

!. i l

l STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

David J. Robare, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

14 Executed at San Jose, California, this 14 day of bEamBER 1993.

kbk$_ k, m' David J. Robare General Electric Company l

Subscribed and sworn before me this /4 y of 44__. 1993.

k m Y Y>ca L 1/

Notary Public, State of California

. /h, MARY L. KENDALL ti i COMM

• 987864 z U ddg+/.. Notcry Pubhc - Cohfomro~I SANTA CLARA COUNTY f-.~ My Corr.m. Expires MA9 26.1997 l

l 12/13/93RTH l

f

, Afridsvit Page 4 I

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