ML20063F344

Proposed Tech Specs 3.3.1 & 3.3.4,extending Surveillance Test Interval from Monthly to Quarterly to Verify Proper Position of MSIV Limit Switches for RPS Instrumentation Operability

ML20063F344
Person / Time
Site:	LaSalle
Issue date:	01/28/1994
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML19303F853	List: ML20063F308 ML20063F344
References
NUDOCS 9402140253
Download: ML20063F344 (24)

Text

. . . _ ..

1 ATTACHMENT B

. PROPOSED CHANGES.TO THE TECHNICAL ,

SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND  !

NPF-18 NPF-11  !

NPF-18 14 3 3/4 3-7 3/4 3-8 3/4 3-8 .

3/4 3-44 3/4 3-44 ,

B 3/4 3-1  !! 3/4 3-1  ;

Insert A Insert A B 3/4 3-3 B 3/4 3-3 .;

Insert B Insert B  ;

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I TABLE 4.c. .1-1  ;

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l REACTOR PROTECTION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS p I OPERATIONAL h CHANNEL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH e

~ CHECK TEST CALIBRATIOM(a) SURVEILLANCE REQUIRED g FUNCTIONAL UNIT .

1. Intermediate Range Monitors

• - a. Neutron Flux - High S/U(b) 5 S/U(c),W R 2 S W R 3,4,5 Inoperative NA W NA 2,3,4,5 b.

Average rower Range Monitor:

UI 2.

a. _ Neutron Flux - High' 5/U(b) 5 S/U(c),W 1, 2

• SA '

'Setdown ,

3, 5 S W SA

? b. Flow'Blased Steulated Thermal W(d)(e),33,g(h) j Power-Upscale, 4, DI8)

• S/UIC) ,W 1

c.. FImed Neutron Flux'- W Id) . SA w

' High' 5~ S/U IC) ,W 1 W NA 1, 2, 3, 5

d. Inoperative NA

3. Reactor Vessel Steam Dome 1, 2 Pressure - High MA M Q

't

4. Reactor Vessel Water Level - R 1, 2 Low, Level.3 NA M l

5. Main Steam Line Isolation R 1-

- Valve - Closure NA

6. Main Steam Line Radiation - R 1, 2 High 5 M

.k iW 7. ;rri.ar,Contain.entPressore-High NA M Q 1, 2.

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' TABLE 4.3.1.1-l (Continued)'

REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS .

5 CHANNEL OPERAT10NAL ge . CHANNEL FUNCT10NAL- CHANNEL CON 01TIONS FOR WHICH F- FUNCTIONAL UNIT CHECK TEST CAllBRATION SURVEILLANCE REQUIRED .

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, 8. Scram Olscharge Volume Water i level .High NA- M R 1, 2, 5 M

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9. Turbine Stop Valve - Closure NA R 1 w 10. Turbine Control Valve Fast Closure Valve Trip System 011 Pressure - Low MA k 1

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

12. Manual Scrue NA K NA 1,2,3,4,5

13. Control Rod Drive 4

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

A b. Delay Tleer NA M R 2, 5 W Neutron detectors may be excluded from CHANNEL CAllBRATION.

(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 > 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.-

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, adjust any APRM : % nel with a GAF > 1.02. In addition, adjust any APRM channel within r ,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 905 of RATED THERMAL POWER and the APRM channel GAF is' 5 4 0.96 - or (2) lf power is less than 90% of. RATED THERMAL- POWER and the' APRM reading exceeds the power

$ value determined by the heat balance by more then 10% of RATEO-THERMAL POWER. Until any required APRM 5 adjustment has been accomplished, notification shall' be posted on the reactor control panel.

?. (e) This calibration shall consist of the adjustment of the APRM flow biased channe1~ to conform to a-g calibrated flow signal.. *

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

L

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

c . .. .

(h) This calibration.shall consist of verifying the 6 1 1 second simulated thermal power time constant. ,

l "The specified 18-month interval may be waived for Cycle 1 providst the surveillancb performed during Refuel 1, ~

which is to commence no later than October. 27, 1985.

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TABLE 4.3.4.2.1-1 9

m [ND-OF-CYCLE RECIRCULA110N PtiMP TRIP SYSTEM SURVEILLANCE REQUIREMENTS

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• p, CllAtlNEL FUliCIl0NAL CilANNEL c TRIP IllllCTION 1EST CALIBRATION -

1. luebine Stop Valve-Closure f[. h R

2. lurbine Control Valve-Fast Closure ,# h R

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o 3/4.3 INSTRUMENTATION l

BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:

a. Preserve the integrity of the fuel cladding. -

b. Preserve the integrity of the reactor coolant system.

c. Minimize the energy which must be adsorbed following a loss-of-coolant accident, and

d. Prevent inadvertent criticality.

This specification provides the l'imiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of main-tenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.  ;

The reactor protection system is made up of two independent trip systems.

There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the_intant of IEEE-279,1971, for nuclear power plant protection systems. Masies the trip settings of the RPS are discussed in the bases for Specification 2.2.1.

The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are com-pleted within the time limit assumed in the accident analysis. No credit was taken for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, previoed :uch tests demonstrate the total channel response time as defined, Sensor response time verification may be demonstrated by either (1) inpirce, onsite or offsite test measurements, o (2) utilizing replacement sensors with certified response times.

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LA SALLE - UNIT 1 8 3/4 3-1

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't 4 4 ATTACIBIENT B ,

PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT A .

Specified surveillance intervals for MSIV- Closure, TSV-Closcrc, TCV-Closure, and the Manual Scram have been determined in accordance with NEDC-30851P-A, *

" Technical Specification Improvement Analyses for BWR Reac'c. Protection ,

System", March 1988.

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• l INSTRUMENT TION -

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BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION i The anticipated transient without scram (ATWS) recirculation pump trip system provides a means of limiting the consequences of the unlikely occurrence of a failure to scram during an anticipated transient. The response of the plant to this postulated event falls within the envelope of. study events in General Electric Company Topical Report NEDO-10349, dated March 1971 and NEDO-24222, dated December, 1979, and Appendix G of the FSAR.

The end-of-cycle recirculation pump trip (EOC-RPT) system is a part of

~

the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of therral margin which occurs at the end-of-cycle. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity to the reactor system at a faster rate than the control rods add negative scram reactivity. Each EOC-RPT system trips both recircula-tion pumps, reducing coolant flow in order to reduce the void collapse in the core during two of the most limiting pressurization events. The two events fc;r which the EOC-RPT protective feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.

A generic analysis, which provides for continued operation with one or both trip systems of the EOC-RPT system inoperable, has been performed. The analysis determined bounding cycle independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCO) values which must be used if the EOC-RPT system is -

inoperable. These values ensure that adequate reactivity margin to the MCPR safety limit exists in the event of the analyzed transient with the RPT function inoperable. The analysis results are further discussed in the cases for Speci-fication 3.2.3.

A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valv1!s provides input to the second EOC-RPT system. Similarly, a position switch for each of two turbine stop valves provides input to one EOC-RPT system; a position switch from each of the other two stop valves provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves. The operation of either logic will actuate the EOC-RPT system and -

trip both recirculation pumps.

Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administrative 1y controlled. The manual bypasses and the automatic Operating Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room.

The EOC-RPT system response time is the time assumed in the analysis' between initiation of valve motion and complete suppression of the electric arc, i.e. ,190 ms, less the time allotted for sensor response, i.e. ,10 ms, and less the time allotted for breaker arc suppression determined hy test, as correlated to manufacturer's test results, i.e. , 83 ms, and plant pre-operational test results.

LA SALLE - UNIT 1 h B 3/4 3-3 Amendment No. 58

ATTACHMENT B i

PROPOSED CHANGES TO THE TECHNICAL SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 INSERT B Specified surveillance intervals have been determined in accordance with the Ib110 wing:

1. NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System", March 1988.  ;

2. GENE-770-06-1-A, " Bases for Changes to Surveillance Test Intervals and l Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications", December 1992.

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REACTOR Pa0TECTiest SYSTEM IN515t9ENTATION $URVEILLANCE REQUIREMENTS E

eu -

C Et OWelEL 0,E.Tio t CONDITIcels FOR WICH - -

CHAleIEL FisICTIONAL CHECK TEST CALIBRAT1014*II $UINEILLANCE REQUtBES FtseCTI0ft.L telli ,

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• 1. Intermedlete Range Moe 5/11(b) . 5 54(c) if 2

a. liestree Flan iM

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q.* ...21.;p.t

.~ 5 W R 3, 4, 5

h. Insperative IIA W NA 2,3,4,5 ,

2. Aver power Range Moniter: III '

a. ren Flus - Nigh, Setdanni 54(b). 5 54(c) W , SA 1, 2 5 W SA ' 3, 5 y b. Flow Riesed Simulated Thermalg ,3 F'd)(e),SA,a(h) -

• Power-14 scale- 5. B 54(c) W , 1 c.

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Fixed leaetron F1mm -

Nigh 5 54ge) W , W,43 , 54 1

d. Insperative IIA W 11 4 ' 1, 2, 3, 5

3. ReacterVesellSteamDame 1, 2  ;

M

' Pressure - Nigh M 4

4. Reacter vessel Water Level -

• 1, 2 Low, Level 3- 16 4 M R l

5. Main Steam Line Isolatten I -

A Valve - Closure 11 4

6. Main Steam Line Redletten - 1, 2 Nigh 5 N R .

l 3 7. Primary Centalement Pressure -

N4 M q 1. 2 gr High - -

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r TABLE 4.3.1.1-1 (Continued) i REACTOR PROTECTION SYSTEM ~lNSTRUNENTAil0N SURVEILLANCE REQUIREMENTS

[# CHANNEL OPERATIONAL .

1 E' CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH e TUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED I~ 8. Scrae Olscharge Volume Water '

7- Level - High ggfp 4 NA M R 1,2,5 ,

' 9. Turbine Stop Valve - Clepare, MA M R 1

16. Turbine Control Valve ~Fe94 . . .

closure Valve Trip Sys,$se pil:.

Pressure - Low- q.

,..a.lt ..

M N R 1

11. Reactor Mode Switch- .g g 9,<jh .  ;

M M 1,2,3,4,5 Shutdeun Position R

12. Manual Scram M hN M 1,2,3,4,5

13. Centrol Rod Drive-

a. Charging Water Meader .

E Pressure - Low M R 2, 5

b. Delay lleer' M M 2, 5 R

'if (a) Neutron detectors may be excluded from CHANNEL CAllBRAil0N.

(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 previews 7 days.

(c) .Within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-prior to startup..lf not performed within the previous 7 days.

(d) This calibration shall consist of the adjustment of'the APflM channel to conform to the power levels calculated byi a heat balance during OPERATIONAL CON 0lil0N 1 when THERMAL POWER > 25% of RATED THERMAL POWEft. The APRM Gain-Adjustment Factor (GAF) for any channel shall be equal to the power value determined by.the heat balance divided by the APlWI reading for:that channel.~ .

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,l adjust any APlWI channel with a GAF > 1.02. In addition, adjust any APflM channel within 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 then or equal .to 901 of RATED THElWIAL POWER and the APlWI channel GAF is < 0.98, or (2)

'F 1f power is less.than 905.ef' RATED THERMAL POWER:and the APflM reading exceeds the power value determined by the

• , heat- balance by more then 105 of RATEO.THEllMAL POWER. , Untti any required APflM adjustment has been accomplished.

• l notification shall be posted on the reactor control panel..

N .(e) This calibration shall consist of the adjustment of the APlWI. flow biased channel to conform to a calibrated flow

- signal.

! .(f) The LPflMs'shall be tbilbrated at least once per 1000 ef fective full: power hours (EFPH). .

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

M :(h) This calibration shall consist'of verifying the 6 1 1 second slaulated th'rnal e power time constant.

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F4.3 INST 1tuMENTATION .

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The reactor protection system automatically initiates a reactor screa to

.ar.! Preserve the fstogrity of' the fuel cladding

- ' . ..: A., .. Preserve the'intagdty'of the reactor coolant systas.

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t cr <c Ninistze the energy which mustche.adso'rbed foT1owing a Toss-of-coofant _  ;

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. .:c durfag periods when instrument channels may be out'of service because of main-  :

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' * .:. M cThor reactor , .Mian systas is made pof tua independent trfpL systans.

.There.are usually four channels.to monitor.each parameter with we channels in

'ench trip systat., The outputs of the channals in a- trip system are combined  ;

' in a logic se that wither channel will trip that trip systan. The tMpping of .  :

- ' f both trip bystans wili produca a. reactor scras. The systas mosts the intant  :

, of IEEE-279,157L, for nuclear posse piant pretsetion systans. The bases for.

RPE are discuased in the bases for fication Z 2.1.

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. assuranca that the ,...Mive functions associated with each channel are cor i plated withfn the ties limit. assumed in the accident analysis. No credit was '

- taker for those channs1s wf ter response tiens indicated as not applicable. '- -

. . a.itaspense tfas ait be.demonstratad' by any series of sequentiat, overlapping or tatal chassel tant measumment, provided such tests demonstrate the total

channot response time-as defined. Sensoe response time veMfication any be,

• dammastrated ty either(1) inplace, onsita or offsita test measurements,.or. .

c.%.P .

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ATTACHMENT B '

PROPOSED CHANGES TO THE TECHNICAL t SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 ,

INSERT A Specified surveillance intervals for MSIV- Closure, TSV-Closure, TCV-Closure, and ;

the Manual Scram havc been determined in accordance with NEDC-30851P-A,

" Technical Specification Improvement Analyses for BWR Reactor Protection System", March 1988.

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' INSTRIMENTATION .. . i

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BASES 3/4.3.4 RECIRCULATION PtMP TRIP ACTUATION INSTRUMENTATION

{

The anticipated transient without scras (ATWS) recirculation pump trip  ;

system provides a means of limiting the consequences of the unlikely occurrence  !

of a failure to scram during an anticipated transient. The response of the- :1

' plant to this postulated ev ent falls within the envelope of stu@ events in .

l General- Electric Casparty Topical Report NED0-10349, dated March 1971 and i 4

NEDD-24222, dated December,1979, and Appendix G of the FSAR. - -

a The end-of-cycle recirculation pump trip '(EDC-RPT) .systas is a part of l the Reactor Protection System and is an essential safety supplement to the '

reactor trip. The purpose of the EDC-RPT is to recover. the loss of thermal i margin which occurs at the end-of-cycle. The physical phenomenon involved is  !

that the void reactivity feetack due to a pressurization transient can add. .-i positive reactivity to the reactar system at a faster rate than the control i rods add negative scras reactivity. Each E0C-RPT system trips both recircule- l tion pimps, reducing coolant flow in order to reduce the void collapse in the  ;

core during two of the most limiting pressurization events. The two events - i for which the E0C-RPT protective feature will function are closure of the.  ;

tudine stop valves and fast closure of the turbine control valves.  ;

A generic analysis,' which provides for continued operation'with one.or both i trip systans of the EOC-RPT systas inoperable,.har seen perfomed. The ~

analysis determined bounding cycle independent MINIMUM CRITICAL POWER-RATIO - -

\ (MCPR) Limiting condition for Operation ~(LCD) values which must be used if.de 1 EDC-RPT system is inoperable. These values ensure that adequata reactivity 1 margin to the MCPR safety limit exists in the event of the analyzed transient  ;

with the RPT function inoperable. The analysis results are further discussed i in the bases' for Specification 3.2.3. -

A fast closure sensor free each of two turbine control valves provides  !

input to the EDC-RPT systas; a fast closure sensor from each of the other two -i turbine control valves provides input to the second E0C-RPT system. Similarly,- .  !

a position switch for each of two turbine stop valves provides input to one -

l E0C-RPT systas; a position switch from each of.the other two stop valves  !

provides input to the. other E0C-RPT systas.. For each EOC-RPT system, the j

sensor relay contacts /arearranged to form a 2-out-of-2 logic for the fast :. I closure of tuttine control valves and a 2-out-of-2 logic for the turbine stop  !

valves. The operation.of either logic will actuate the EDC-RPT system and '

trip both reci,rculattan.pamps.

Each EDC-RPT'... ,. msystaar any be manually bypassed by use of a keyswitch which - ii; is administrative 1y controlled. The annual asses and the automatic Operating Bypass at less than 305 of RATED THERMAL are annunciated in the control- ,'

( room. .

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L . The EDC-RPT systar response time is the time assumed in the analysis ' l between initiation af valve action and complete suppression of the electric I

arc, i.e. ,190 as, less the time allotted for sensor response, i.e. ,10 as,  :

and less the time allotted for breaker are suppression detemined by test, , a j as correlated to manufacturer's test results, i.e., 83 as, and plant l L

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pre-operational test results. i L

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LA SALLE - UNIT 2 B 3/4 3-3 Amendment No. 41 ~ ,

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3 ATTACIBENT B }

PROPOSED CHANGES TO THE TECHNICAL [

SPECIFICATIONS FOR OPERATING LICENSES NPF-11 AND NPF-18 l i INSERT B Specified surveillance intervals have been detennined in accordance with the  :

fbilowing: l-

1. NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR - -

Reactor Protection System", March 1988,

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2. GENE-770-06-1-A, " Bases for Changes to Surveillance Test Intervals and j Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications", December 1992.

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. . ATTACIBIENT C SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated the proposed Technical Specification Amendment which extends the Surveillance Test Interval (STI) for certain instruments in the Reactor Protection System (RPS) and the End-of-Cycle Recirculation Pump Trip (EOC-RPT) System for LaSalle County Station Units 1 and 2, and determined that they do not constitute a Significant Hazards Consideration. Based on the criteria for defining a significant hazards consideration established in 10CFR50.92, operation of LaSalle County Station Units 1 and 2 in accordance with the proposed amendment will not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated because:

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed changes increase the STI fhr actuation instrumentation supporting RPS and EOC-RPT trip functions. There are no changes in any of the affected systems themselves. Because of this there is no change in the probability of occurrence of an accident or the consequences of an accident or the consequences of malfunction of equipment. With respect to the malfunction of equipment, topical reports prepared by GE demonstrated that there is a reduction in scram frequency for the RPS. This offsets the slight increase in trip function unavailability determined by GE. This was judged acceptable by GE. The NRC concurred with this conclusion in its review of the topical reports (NEDC-30851P-A). For EOC-RPT GE demonstrated that the trip function unavailability when the surveillance interval is extended from 1 to  ;

3 months is lower for the turbine stop valve trip function and slightly higher for the turbine control valve trip function than the same trip '

functions for RPS-scram. However, GE concluded that the small increase in EOC-RPT unavailability (represented by small increased risk of a MCPR violation) is offset by the benefits associated with the similar approved STI !

and AOT changes for the RPS-scram function. Therefore, GE concluded <

that the STI changes for EOC-RPT trip function are bounded by the approved RPS analysis (Reference 5). The NRC accepted the conclusions of i GE by a SER included in Reference 9. The proposed changes are consistent with the Safety Evaluation R.eports issued in these topical reports. The l proposed changes therefore do not involve a significant increase in the  ;

probability or consequences of an accident previously evaluated. l l

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. . ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION

2) Create the possibility of a new or different kind of accident from any accident previously evaluated because:

The proposed changes do not create the possibility for an accident or malfunction of a different type than any evaluated previously in the UFSAR. The proposed changes increase the STI for the RPS and EOC-RPT Instrumentation. There are no changes in the instrumentation of these systems. Since there are no such changes there is no possibility for an accident or malfunction of a different type than any previously evaluated.

3) Involve a significant reduction in the margin of safety because:

The proposed changes do not reduce the margin of safety as defined in the basis fbr any Technical Specification. The proposed changes do not change any setpoints in the above mentioned systems or their levels of redundancy.

Setpoints are based upon the drift occurring during an 18 month calibration interval. The bases in the Technical Specifications either do not discuss STI, or state " ..one channel may be inoperable for brief intervals to conduct required surveillance." The proposed changes are bounded by the analyses of References 5 and 9. These analyses, which were prepared by GE and approved by the NRC, examined the effects of extending STI and found that the proposed changes would not involve a significant reduction in a margin of safety. LaSalle Station Units 1 and 2 RPS and EOC-RPT systems have been compared to the generic analyses and verified to be bounded.

Guidance has been provided in " Final Procedures and Standards on No Significant-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 and are not considered likely to involve significant hazards considerations. These proposed amendments most closely fit 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 l component specified in the applicable Standard Review Plan.

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. . ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION 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 I change does not constitute a significant hazards consideration.

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n ATTACHMENT D - -!

ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW l Commonwealth Edison has evaluated the proposed amendment against the  ;

criteria for identification oflicensing and regulatory action requiring environmental assessment in accordance with 10 CFR 51.21. It has been i 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 hazards considerations or do  ;

not involve a significant increase in the amounts, and no significant changes in the types, of any effluents that may be released off-site. Additionally, this request  ;

does not involve a significant increase in individual or cumulative occupational- '

radiation exposure.

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ATTACHMENT E-GENERAL ELECTRIC TOPICAL REPORT: +

TECHNICAL SPECIFICATION IMPROVEMENT ANALYSIS FOR  ;

THE REACTOR PROTECTION SYSTEM FOR LASALLE COUNTY  ;

STATION, UNITS 1 AND 2 MDE-83-0485 Rev. 3, DRF C71-00072-1, April 1991 i l

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s ATI'ACHMENT F WITIIHOLDING AFFIDAVIT FOR THE GENEIIAL ELECTRIC TECHNICAL SPECIFICATION IMPROVEMENT ANALYSIS h

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~9.' 4 General Electric Company AFFIDAVIT I, Robert C. Mitchell, being duly sworn, depose and state as follows:

(1)I am Project Manager, Safety and Communications, 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 MDE 0485, Rev. 3, Technical Specification Improvement Analysis for the Reactor Protection Systemfor IaSalle County Station, Units 1 and 2, dated April 1991. This information is delineated by brackets around the speciSc 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 financialinformation 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 Enerav Project v. Nuclear Regulatorv .

Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Grouo v. FDA. ~

704F2dl280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

Information that discloses a process, method, or apparatus, including supporting data and .j a.

analyses, where prevention ofits 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; 1

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

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; l

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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.b and 4.d, 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. Its initial designation as proprietary information, and the subsequent steps - taken to prevent its unauthorized disclosure, are as set forth in (6) and (7) following. The information sought to be withheld has, to the 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. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or propdetary agreements which provide for maintenance of the information in confidence. ,

(6) Initial approval of proprietary treatment of a document is made by the 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 j manager of the~cogruzant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary l 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 identi5ed in paragraph (2) is classified as proprietary because it would provide other parties, including competitors, with valuable information regarding the .

application of reliability based methodology to BWR instrumentation. A substantial effort has been expended by General Electric to develop this information in support of the BWR Owners' Group Technical Specifications Improvement Program (9) Public disclosure of the inforneien mught to be withheld is likely to cause substantial luum -

to GE's competitive positicin and foreclose or reduce the availability of profit-making 'I opportunities. The information is part of GE's comprehensive BWR technology base, and its - .;

commercial value extends beyond the original development cost. The value of the technology j base goes beyond the extensive physical database and analytical methodology and includes  :

development of the expertise to determine end apply the appropriate evaluation process. In  ;

addition, the technology base includes the value derived from providing analyses done with l NRC-approved methods.

Affidavit Page 2

p e- , t The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GE.

The precise value of the. expenise to devise an evaluation' precess and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be 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 can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the .

public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a

> windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

STATE OF CALIFORNIA )

COUNTY OF SANTA CLARA ) SS:

Robert C. Mitchell, being duly sworn, deposes and says:

That he has read ti2e foregoing affidavit and the matters stated therein are tme and co: rect to the best ofhis knowledge, information, and belief.

Executed at San Jose, California, this Ibday of M MC H , 1993.

O.

Robert C. Mitchell General Electric Company Subscribed and sworn before me this f Nday of Y c4uk 1993.

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Notary Public, State of Califpa PA11tA F. HUSSEY gxj NOW poetic . CAUFORNIA dM SAMrA Ct. ARA COUNTY dCy'filt'Z".5,,21 Afradavn Page 3 l