ML20148U259

Proposed Tech Specs to Change Definition of Channel Calibration & Correcting Miscellaneous Errors in TS & Bases. TS Affected Are TS Definition 1.4,TS Tables 3/4.3.2-1 & 3/4.3.6-1 & Bases for TS 3/4.3.1

ML20148U259
Person / Time
Site:	LaSalle
Issue date:	07/01/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20148U258	List: ML20148U255 ML20148U259
References
NUDOCS 9707100132
Download: ML20148U259 (17)

Text

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, ATTACHMENT B ,

PROPOSED AMENDMENTS TO THE i j

LICENSEITECHNICAL SPECIFICATIONS

. NPF-11 NPF-18 3/41-1 3/4 1-1 3/4 3-6* 3/4 3-6*

3/4 3-13 3/4 3-13 i' 4 3/4 3-51 3/4 3-51 3

3/4 3-52* 3/4 3-52*

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

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• These pages do not have changes; they are included for information only.

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B-1 9707100132 970701 PDR ADOCK 05000373 P PDR j l

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1 1.0 DEFINITIONS S

The following terms are defined so that uniform interpretation of these speci-i fications may be achieved. The defined terms appear in capitalized type and  ;

shall be applicable throughout these Technical Specifications.

ACTION i i

' 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.

i AVERAGE PLANAR EXPOSURE 1.2 The AVERAGE PLANAR EXPOSURE sha'll be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in -

the specified bundle at the specified height divided by tha number of i fuel rods in the fuel bundle.  !

1 I AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable '

to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the  ;

specified height divided by the number of fuel rods in the fuel bundle. I CHANNEL CALIBRATION 1 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary of the P

y% 4 channel output such that it responds with the necessary rang,e and accuracy to known values of the parameter which the channel monitors. The CHANNEL  !

% CALIBRATION shall enccmpass the entire channel including the sensor and

alarm and/or trip functions. and shall include the CHANNEL FUNCTICNAL TEST.

i The CHANNEL CALIBRATION may be performed by any series of segwr.Wim ,

verlapping or total channel steps such that the entire channel is calibMr::

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CHANNEL CHECK ~

4 1.5 A CHANNEL CHECK shall be the cualitative assessment of channel behavior during operation by observatien. This determination shall include, where possible, comc~arison of the channel indication and/or status with other indications and/or status derived from independent instrument channels j measuring the same parameter.

i CHANNEL PJMCTICNAL TEST i

1. 6 A CHANNEL FI;NCTICNAL TEST shall be:

4 s.

Analog

hanrels - the injection of a simulated signal into the ,

channel as close to the sensor as practicable to verify CPERABILITY including alarm and/or trip fun:tions and channel failure j trips.

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b. Bistable channels - the injection of a simulated signal into i tne sensor to verify OPERABILITY in:1uding alarm and/or trip fur:tions.

Tu :F":.i'. f a.:T!: sat T:.7 ce ce r . /te: . j c , scH es af se ue.tia;,

overia;pr.g, sr total enannd; 4:eps sucn tnat tne er. Lire caannei is tested.

LA SALLE - UNIT 1 1-1 i

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i , ATTACHMENT B

PROPOSED AMENDMENTS TO THE 1

LICENSEITECHNICAL SPECIFICATIONS

INSERT A i

i A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel

output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION l shall encompass the entire channel, including the required sensor, alarm, display, l l and trip functions, and shallinclude the CHANNEL FUNCTIONAL TEST. Calibration i

of instrument channels with resistance temperature detector (RTD) or thermocouple L

sensors may consist of an inplace qualitative assessment of sensor behavior and

, normal calibration of the remaining adjustable devices in the channel. The l

CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

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

4 TABLE 3.3.1-2 i

REACTOR PROTECTION SYSTEN RESPONSE TIMES i

RESPONSE TINE .!

FUNCTIONAL UNIT fSeconds1  !

1. Intermediate Range Monttgrs:

a. Neutron Flux - High NA

b. Inoperative NA i

2. Average Power Range Monitor *

a. Neutron Flux - High, Setdown MA

b. Flow Blased Simulated Thermal Power-Upscale

c. s 0.09**

Fixed Neutron Flux - High s 0.09 '

d. Inoperative NA

3. Reactor Vessel Steam Dome Pressure - High

4. s 0.55" .

Reactor Vessel Witer Level - Low, Level 3 s 1.05"

5. Main Steam Line Isolation Valve - Closure t

6. Deleted s 0.06

7. Primary Containnent Pressure - High NA

8. Scram Discharge Volume Water Level - High NA

9. Turbine Stop Valve - Closure 5 0.06 'T ol

10. Turbine Control Valve Fast Closure.

Trip 011 Pressure - Low

$ 0.08# Q q

3

11. Reactor Mode Switch Shuttown Position

12. Manual Scram

13. Control Rod Drive MA NA 'h3*

~

%h

a. Charging Water Header Pressure - Low NA

b. Delay Timer '7 3

NA

% 3 h o i

r.

. 3

, Neutron detectors are exempt from response time testing. Response time shall be measured from the detector output or from the !?put of the first electronic component in the channel. '

• • Not including simulated thermal power time constant . '

1. Measured from start of turbine control valve fast closure,

1. 1. sensor is eliminated from response time testing for the RPS circuits. Response time testing and conformance t to the administrative limits for the remaining channel including trip unit and relay logic are required.

LA SALLE - UNIT 1 3/4 3-6 Amendment No.115

TABLE 3.3.t-1 (Centinued)' .

ISOLATION ACTUATION' INSTRUMENTATION -

VALVE GROUPS MINIMUM OPERABLE APPLICABLE --

OPERATED BY CHANNELS PER OPERATIONAL -

TRIP FUNCTION SIGNAL TRIP SYSTEM (b) CONDITION ACTIGN l

5. RHR SYSTEM STEAM CONDENSING MODE ISOLATION

a. RHR Equipment Area A Temperature - High 8 1/RHR area 1, 2, 3 22

b. RHR Area Temperature -

High 8 1/RHR ara I,'2, 3 22

c. RHR Heat Exchanger Steam Supply Flow - High 8 1 1, 2, 3 22

6. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION *

a. Reactor vessel Water Level - Low, level 3 6 2 1, 2, 3 25

b. Reactor Vessel (RHR Cut-in Permissive)

Pressure - High 6 1 1, 2, 3 25 i

c. RHR Pump Suction F'dow - High 6 1 1, 2, 3 25

d. RHR Area Temperature -

High 6 1/RHR area 1, 2, 3 25

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

B. MANUAL INITIATION

1. Inboard Valves 1, 2, 5, 6, 1/ group 1, 2, 3

2. Outboard Valves 26 l 1/ group 1,.2, 3

3. Inboard Valves 4g *)g' 5, 1/ group 26 1 **,# 26

4. Outboard Valves 4(*H'3 1/ group 1,, 2, 2, 33 and and **,# 26

5. Inboard Valves 3,8,9 1/ valve 1, 2, 3

6. 26 Outboard Valves

7. . Outboard Valve 3

8 3 ,8, 9 1/ valve 1, 2, 3 26 1/ group 1, 2, 3 26

LA SALLE - UNIT 1 3/4 3-13 Amendment No. 102

TABLE 3.3.6-1 -

9 CONTROL R00 Wlill0HAWAL Bl.0CK INSTRUMENTATION D

r-MINiHIIM OPERADLE APPLICABLE

~

E CllANNELS PER OPERAil0NAL

• ' TRIP fill!CIION TRil' I UNCfl0N CONDITIONS ACTION

$" 1. ijff!) ULOCK MONITOR (a) . "

~~ ... lipscale 2 l' 60 I.. Inoperative 2 1* 60

c. 1:ownscale 2 1* '

60

2. . .I I :i t

.i. I low Blased Simulated Thermal -

Power-Upscale 4 1 61

h. Inoperative 4 1, 2, 5 61

c. llownscale 4 1 61 il. lieutron flux-Iligh 4 2, 5 61 U 3. 'AlllI!CE RAIME M0!!!IURS I i Detector not full in(b) 3

h. Upscale (c) 3 2 61 i 2 5 61

c. Inoperative IC) 3 '

f

• el. Downscale(d)

4. INI[l: MEDIATE RAtEE_MONITOIS .

a lietector not l'ull in 6 2, 5 61 li. lipscale 6 ' 2, 5 61

c. Inoperall 6 2, 5 61 it. liownstale{g) 6 2, 5 61

5. stP.All DiscilARGE VOLUME

.. tlater Level-High 2 1, 2, 5** 62

1. 9: ram Discharge Volume Switch in Bypass 1 5** 62

6. RtC.IRCULATION FLOW UNIT

.i . lipscale 2 1 62 1 b inoperative 2 1 62

.. Comparator 2 1 62 '

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• , _ _ _ ,. - - - - , e.- -- . . -- ,e.,, -r- -

1 TAPtr 1.1.6-1 (Centinued)

. CONTROL ROM WT?HBRAWAL BLOCK TNETRUMPMTATTON ACTION ACTION 60 -

Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3. 3 ACTION 61 - With the number of OPERABLE channels:

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped condition within the next hour.  !

b. >

Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.

ACTION 62 -

With the number of OPERABLE Channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

EtIE With THERMAL POWER t 30% of RATED THERMAL POWER.

With more than one control rod withdrawn. Not applicable to control rods removed per specification 3.9.10.1 or 3.9.10.2.

a.

The RBM shall be automatically bypassed when a peripheral control rod is selected.

b. This function shall be automatically bypassed if detector count rate is t 100 cps or the IRM channels are on range 3 or higher.

c.

This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.

d.  !

This function shall be automatically bypassed when the IRM channels are on range 3 or higher.

e.

This function shall be automatically bypassed when the IRM channels are on range 1.

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LA SALLE - UNIT 1 3/4 3-52 Amendment No.104

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l 3 /4. 3 INSTRUMENTATION I l 1

' BASES i

3/A.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTFION l l 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-j coolant accident, and s

d. Prevent inadvertent criticality.

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This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended i

function even during periods when instrument channels may be out of service

because of maintenance. When necessary, one channel may t,e 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 I 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 intent of IEEE-279, 1971, for nuclear power plant protection systems. Specified  ;

surveillance intervals and surveillance and maintenance outage times have been 1 determined in accordance with NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System", March 1988, and

! MDE-83-0485 Revision 3, " Technical Specification Improvement Analysis for the Reactor Protection System for LaSalle County Station, Units I and 2", April i 1991. the bases for the trip settings of the PRS are discussed in the bases t for Specification 2.2.1. When a channel is placed in an inoperable status solely for performance of rquired surveillances, entry into LC0 and required ACTIONS may be delayed, provided the associated function maintains RPS trip

! capability.  !

i The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are  ;

i completed within the time limit assumed in the accident analysis. The RPS j RESPONSE TIME acceptance criteria are included in plant Surveillance procedures. Only those functions with times assumed in the accident analysis

{ are required to be response time tested.

1 As stated in Note

• of Table 3.3.1-2, Neutron detectors are exempt from N

i response time testing. In addition, for Functional Units 3 and 4, per Note

{ the associated sensors are not required to be response time tested. For these LA SALLE - UNIT I B 3/4 3-1 Amendment No.114 a

i or lnfhematim INSTRUMENTATION l Only.

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f ' BASES Ndames j 3/4.3.1 l

REACTOR PROTECTION SYSTEM INSTRUMENTATION (Continued) +

Functional Units, response time testing for the remaining channel components, I including any analog trip units, is required. This allowance is supported by

{ NEDO-32291-A, " System Analyses for the Elimination'of Selected Response Time j Testing Requirements," October 1995, i

j Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification '

may be demonstrated by either (1) inplace, onsite or offsite test  ;

} measurements, or (2) utilizing replacement sensors with certified response '

times.

t

} 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION i This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip

! setpoints and response times for isolation of the reactor systems. When i

necessary, one channel may be inoperable for brief intervals to conduct required surveillance. Both channels of each trip system for the main steam i

tunnel ventilation system differential temperature may be placed in an inoperable status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for required reactor building ventilation j system maintenance and testing and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to loss of reactor building 1

ventilation or for the required secondary containment Leak Rate test without placing the trip system in the tripped condition. This will allow for

! maintaining the reliability of the ventilation system and secondary 1 j  ;

containment. Specified surveillance intervals and surveillance and I i

maintenance outage times have been determined in accordance with NEDC-i )

30851P-A, Supplement 2, " Technical Specification Improvement Analyses for BWR i

i Isolation Instrumentation Common to RPS and ECCS Instrumentation", March 1989, and with NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation", July 1990. When a channel is placed in t

' an inoperable status solely for performance of required surveillances, entry into LC0 and required ACTIONS may be delayed, provided the associated function maintains primary containment isolation capability. Some of the trip settings 3

i may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other i instrumentation, where only the high or low end of the setting have a direct

bearing on safety, are established at a level away from the nomal operating i

range to prevent inadvertent actuation of the systems involved, Except for the MSIVs, the safety analysis does not address individual i sensor response times or the response times of the logic systems to which the sensors are connected. For A.C. operated valves, it is assumed that the A.C.

i LA SALLE - UNIT 1 8 3/4 3-2 Amendment No. 134 e

g = q % s g w Q a- e ._ _ _ . _ _ _n._ _ _ .

.w;n: -

i 4 . .

i 1.0 hannmuse 1

, The following terms a=a defined so that unifors interpretation of these speci-

! fications ser be achieved. Ths, defined terus appear in capitalized type and

, shall he applicable throughout these Technical. Specifications.

5 i

M -

l L1 ACTZW shelf he that part of a SpecfHestion which prescribes remedial

, maseures required under designated conditions.

AVBtAGE PuMAR EXPOSURE ,

LZ The AVERAGE PuMAR EXP05 TEE s%11 hp applicable to a specific planar  !

height and la aquel to the sue of the exposure af all the fuel rods in

! tha specified bundle at the specified height. divided by the number of j fuel rods in the fuel bandia. .

l AV0tAGE PLANAR LINEAR NEAT GENERATION RATE L1 The AVERAGE PuMAR LDEAR WAT GBERATION RATE (APLHER) shall be applicabla to a specific planer height and is equal to the se of the LINEAR lEAT -

)

j .

GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.

Q WeIEL CALIBRATT W ,,

. 'y  ;

{ 4 A OlAlllEL CALIBRATIM shall be the adjustaant, as necessary, of the l 7 .g channel output sucir that it responds wi^.tr the necessary range and accuracy

/t to kneun values of the parametar which the channel monitors. The QWelEL CALIRRATIM shall encespass the entire channel including the sensor and l

alars and/or trip functions, and shall include the CHAfsIEL PWCTIONAL l I

i TEST. The OWOIEL CALIRRATI0ll may be performed try ag series of sequential, l lappihg er total channel staps such that the entire channel is calibratedj i QWs EL oscE l

, L5 A OWBEL DECK shall be the qualitative assessment of channel behavior l during operation by observation. This dotarvination shall include, where a

possible, cooperisen of the channel indication and/or status with other f adications-and/or status ' derived from independent instruent channels asseuring the same parameter. - -.

OWelEL PtsecTICII4L TEST 3 . . .

l L6 A DIAISEL PWC21GIAL TEST shall be:

I a. -maias ehennels - the injection of a simulated signal into the l channel as close to the sensor as practicable te verify OPDtABILITY i including alars and/or trip functions and channel failure l MM.

1 h. Sistalile channels - the in.jection of a simulated signal into i the senser to verify OPERARILITY including alars and/or trip j functions. -

) The OWOIEL PWCTIONAL TEST may be performed by ag series of sequential, j Jveriapping, or total channel steps such that the entire channel is testad.

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, ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS l

l lNSERT A A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel )

output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CAllBRATION shall encompass the entire channel, including the requirsi sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple  ;

sensors may consist of an inplace qualitative assessment of sensor behavior and l normal calibration of the remaining adjustable devices in the channel. The CHANNEL CAllBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated, i

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TABLE 3.3.1-2 '

REACTOR PROTECTION SYSTEN RESPONSE TIMES RESPONSE TIME FUNCTIONAL UNIT -

fSeconds1

1. Intemediate Range Monitors:

a. Neutron Flux - High* NA

b. Inoperative NA

2. Average Power Range Monitor *

a. Neutron Flux - High, Setdown MA

b. Flow Blased Simulated Thermal Power Upscale

c. s 0.09**

Fixed Neutron Flux - High 5 0.09

d. Inoperative NA

3. Reactor Vessel Steam Dome Pressure - High

4. Reactor Vessel Water Level - Low, level 3 s 0.55"

5. s 1.05" Main Steam Line Isolation Valve - Closure s 0.06

6. DELETED l

7. I Primary Containment Pressure - High NA

8. Scram Discharge Volume Water Level - High

9. NA Turbine Stop Valve - Closure l

10. Turbine Control Valve Fast Closure, s 0.06 Trip 011 Pressure - Low s 0.08'

11. Reactor Mode Switch Shutdown Position MA

12. Manual Scram '

NA

13. Control Rod Drive a.

b.

Charging Water Header Pressure - Low Delay Timer NA NA N

D k i

3

• Neutron detectors are exempt from response time testing. Response time shall be measured

> $h 'I from the detector output or from the input of the first electronic component in the. 3 channel.

• • Not including simulated thermal power time constant. kW A NI!

1. Measured from start of turbine control valve fast closure. ~

D

1. 1. Sensor is eliminated from response time' testing for the RPS circuits. Response time testing D and conformance relay to the administrative limits for the remaining channel including trip unit and logic are required.

LA SALLE - UNIT 2 3/4 3-6' Amendment No.100

i e s

TABLE 3.3.2-1 (Continued)

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

5. RHR SYSTEM STEAM CONDENSING MODE ISOLATION

a. RHR Equipment Area A Tempsrature - High 8 1/RHR area 1, 2, 3 22

b. RHR Area Temperature -

High 8 1/RHR area 1, 2, 3 22

c. RHR Heat Exchanger Steam Supply Flow - High 8 , 1 1, 2, 3 22

6. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION

a. Reactor Vessel Water Level - Low, Level 3 6 2 1, 2, 3 25

b. Reactor Vessel (RHR Cut-in Permissive)

Pressure - High 6 1 1, 2, 3 25

c. RHR Pump Suction Flow - High 6 1 2, 1, 3 25

d. RHR Area Temperature -

High 6 1/RHR area 3 1, 2, 25

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

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

2. Outboard valves 1, 2, 5, , 7 1/ group 1, 2, 3 26

3. Inboard Valves 4 '* " * ' 1/ group 1, 2, 3 and ** ' 26

4. Outboard Valves d ' * "* ' 1/ group 1, 2, 3 and ** ' 26

5. Inboard valves 3, 8, 9 1/ valve 1, 2, 3 26

6. Outboard valves 3, 8, 9 1/ valve 1, 2, 3 26

7. Outboard Valve 8 '"' 1/ group 1, 2, 3 26 LA SALLE - UNIT 2 3/4 3-13 AMENDMENT NO. 87

s l

TABLE 3.3.5-1 .

g Callipet 300 iffilDRAM4L StoCE Ill5Ipl81ENTATIell i -

I MINIORSI OPERABLE APPLICASLE ,

CimielELS PER SPERATIGIIAL

, IRIP FINICTIcel . Trip resectices coeggigests . gigg

1. pos gleCE tellIIeR I 'I .

a. $ scale 2 le 6e ce b. Insperative 2 -

18 '

60 -

c. Baumscale . I la 64

2. 823l ,.

a. Flew glased Sloulated Theres)

T r : ' cale 4 I 61

b. InsperatlIre 4 1 61

c. Sounscale 4 1, 8, 5 61 - .

d. Ilestren Flum-High ,

4 3, 5 61

3. soupCE RAIIGE 110111 1 0 8 5 .

b a. Detector not tell inI "I l ,

-l*-

b. $ scaleI *I 3 '8I 5-e 2 5 51 3 61

c. Insperative(c) 2 - .' 5t 61

.. scaisa>

i g -

4. IIITENSIATE RAIIGE Imilliams '

'a Betector not full i

• 6 2, 5 41-

b. $ scale 4 2, 5 51

c. 6 2, 5 61 -

Insperatig) 6 2, 5 61

d. Daunscale -

5.- SCRAN DISCitAAGE VOL W

a. Water Level-High 2 . l. 2, 5** 62 ,

h. Scram Discherge Volume '

Switch in typass 1 5** 62

6. RECIRCULATICII FLOW 411111

a. Ilpscale 2 1 62

b. Insperative 2 I .

62 l

c. Ceeparator . 2 i 1 62 1

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• I

. - . . ~ . _ _ - - . .._ - m ._. ___-._._...~m. _ . _ _ . - _ - . . _ . _ . . . . ._m . . _ _ . _ _ - . . . . - .

1 r .

-s TaBl? 3.1.5-1 (Centinucd) j CONTROL ROD WITHMR&,WA' ELOCK TMETRUMEN*ATION ACTION ACTION 60 - Specification Declare the RBM 3.1.4.3.

inoperable and take the ACTION required by ACTION 61 - With the number of OPERABLE channels: *

a.

One less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 7 days or place the inoperable channel in the tripped

}

i condition within the next hour.

b.

i Two or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in j the tripped condition within one hour.

I ACTION 62 - With the number of OPERABLE Channels less than required by the I

Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

[

EQ21 1

• With THERMAL POWER 130% of RATED THERMAL POWER.

With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

a.

The RBM shall be automatically bypassed when a peripheral control rod is selected.

b.

This function shall be automatically bypassed if detector count rate is I 1 100 cps or the IRM channels are on range 3 or higher.

c.

This function shall be automatically bypassed when the associated IRM channels are on range 8 or higher.

d.

This function shall be automatically bypassed when the IRM channels are on range 3 or higher.

a.

This function shall be automatically bypassed when the IRM channels are on range 1.

1

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}f ihlIQ(!!10 Vl0f1  !

Or1I)!<

I Alo <J m e.s LA SALLE - UNIT 2 3/4 3-52 Amendment No. 90

i I 3/4.3 INSTRUMENTATION J

i l BASES i 3/4.3.1 I REACTOR PROTECTION SYSTEM INSTRUMENTATION i i

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 2

l d. Prevent inadvertent criticality, t i

.This specification provides the limiting 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 hecause of maintenance. >

i for brief intervals to conduct required surveillance.When necessary, one channel i The reactor protection system is made up of two independent trip systems.

g There each trip aresystem. usually four channels to monitor each parameter with two channels in

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 intent i

of IEEE-279,1971, for nuclear power plant protection systems. Specified i surveillance intervals and surveillance and maintenance outage times have been j determined in accordance with NEDC-30851P-A, "Techr.ical Specification Improvement Analyses for BWR Reactor Protection System", March 1988, and MDE-83-0485 Revision 3, " Technical Specification Improvement Analysis for the Reactor 1991. Protection System for LaSalle County Station, Units I and 2", April for Specification 2.2.1.The bases for the trip settings of the RPS are discussed in th When a channel is placed in an inoperable status solely for performance of r6 quired surveillances, entry into LCO and required ACTIONS capability. may be delayed, provided the associated function maintains RPS trip The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the accident analysis. The RPS procedures. TIME acceptance criteria are included in plant Surveillance

RESPONSE

are required to Only those functions be response time tested. with times assumed in the accident analysis response As stated in Note

• of Table 3.3.1-2, Neutron detectors are exempt from time testing.

the associated sensors are not required to be responseFor time these tested.In addi LA stile - UNIT 2 B 3/4 3-1 Amendment No. 99

- - - , . e -,s = - +g---.ep.ww-,+yg-. - ps-p --.+e --ge--eg 4w ,

a

• f' I hf0s^rna& ,4 INSTRUMENTATION .' O /l

' \

O&b67rty25 j BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION (Continued) i Functional Units, response time testing for the remaining channel components, including any analog trip units, is required. This allowance is supported by j NEDO-32291-A, " System Analyses for the Elimination of Selected Response Time Testing Requirements," October 1995.

i .

i Response time may,be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined. Sensor response time verification i

may be demonstrated by either (1 inplace, onsite or offsite test measurements, times.

or (2) utilizing re) placement sensors with certified respon

~

l 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION

[

j This specification ensures the effectiveness of the instrumentation used

to mitigate the consequences of accidents by prescribing the OPERABILITY trip j setpoints and response times for isolation of the reactor systems. When 5 necessary, required surveillance. one channel may be inoperable for brief intervals to conduct Both channels of each trip system for the main steam j tunnel ventilation system differential temperature may be placed in an '

i inoperable status for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for required reactor building ventilation

- system maintenance and testing and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> due to loss of reactor building

! ventilation or for the required secondary containment Leak Rate test without

placing the trip system in the tripped condition. This will allow for i j

maintaining the reliability of the ventilation system and secondary containment. Specified surveillance intervals and surveillance and i maintenance outage times have been determined in accordance with NEDC-i 30851P-A, Supplement 2, " Technical Specification Improvement Analyses for BWR i Isolation Instrumentation Common to RPS and ECCS Instrumentation', March 1989, and with NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR i

Isolation Actuation Instrumentation", July 1990. When a channel is placed in  !

j an inoperable status solely for performance of required surve j maintains primary ~ containment isolation capability. Some of the trip settings may have tolerances explicitly stated where both the high and low values are 3

critical and may have a substantial effect on safety. The setpoints of other

instrumentation, where only the high or low and of the setting have a direct l bearing on safety, are established at a level away from the normal operating range to ptevent inadvertent actuation of the systems involved.

Except for the MSIVs, the safety analysis does not address individual i

sensor response times or the response times of the logic systems to which the sensors are connected. For A.C. operated valves, it is assumed that the A.C.

IA SALLE - UNIT 2 B 3/4 3-2

. Amendment No. 99 i

. ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Summary of the Proposed Technical Specification Chances:

a. The definition of Channel Calibration does not include an exemption for the calibration of thermocouple and resistance temperature detector (RTD) sensors, which can not be calibrated. Therefore the definition is being ,

changed to perform an in place qualitative assessment of thermocouple  !

and RTD sensors.

b.

4 TS Table 3.3.2-1 Isolation Actuation instrumentation, item B.2, Outboard Valve isolation incorrectly lists isolation group 7 as receiving a l containment manual isolation signal. Containment isolation group 7 only contains inboard automatic isolation valves for the Transversing in-core

{

Probe (TIP) system. Group 7 is proposed to be removed from the i outboard manual isolation function, since there are no automatic outboard isolation valves for the TIP system.

c. TS Table 3.3.6-1, Control Rod Withdrawal Block Instrumentation Trip Function 4.a., intermediate Range Monitors (IRM) Detector not full-in rod block is modified by Table Note e. Table note e only applies to the IRM Downscale rod block, Trip Function 4.d. Note e is proposed to be deleted from Trip Function 4.a.

d. TS Bases section 3/4.3.1, Reactor Protection System Instrumentation, has a typographical error. The last paragraph on page 3/4 3-1, should refer to Note ## instead of Note #.

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. ATTACHMENT C

, SIGNIFICANT HAZARDS CONSIDERATION i

1

, Commonwealth Edison (Comed) has evaluated the proposed Technical j Specification Amendment and determined that it does not represent a significant

! hazards consideration. Based on the criteria for defining a significant hazards j consideration established in 10 CFR 50.92, operation of LaSalle County Station

] Units 1 and 2 in accordance with the proposed amendment will not:

i

1) Involve a significant increase in the probability or consequences of an

accident previously evaluatsd because:

i 4

a. The change in the definition of a Channel Calibration is to make the wording more clear and to require an inplace qualitative

assessment in place of the calibration of thermocouple and l resistance temperature detector (RTD) sensors. The thermocouple

and RTD sensors are not adjustable and are not subject to drift due i to their design. The inplace qualitative assessments will assure

proper functioning of the sensors, due to the nature of these li sensors and the associated failure modes, and thus will verify that the sensors will be able to fulfill their intended function (s).

l Therefore the change to the definition will not change the probability or consequences of an accident previously evaluated.

l b. Manual initiation of isolation actuation instrumentation trip systems l l for inboard and outboard valves is required to be operable per TS l Table 3.3.2-1, Trip Functions B.1 and B.2, respectively. Trip Funtion i B.2, outboard valves, lists valve group 7, TIP system isolation

valves. Valve group 7 consists of an automatic inboard isolation

! valve for each TIP guide tube penetrating the primary containment (correctly listed under B.1), and a manual outboard isolation valve j on each guide tube, that is an explosive squib valve. Each

! explosive squib valve is manually actuated with a keylock switch

! from the main control room per design. Each is a positive control backup upon failure of an inboard valve in the open position. The squib valves are not actuated from isolation actuation channel 4

logic. This configuration meets the current design and licensing  !

basis. Therefore, deletion of valve group 7 from TS Table 3.3.2-1  ;

will not change the probability or consequences of an accident '

4 previously evaluated.

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4 i i

a

__ __ _ _ _ _ __.. _ _ _ _ _ _._ _ . _ _ - ~._ _ __ _. _ _ ___ _,

s l

ATTACHMENT C l SIGNIFICANT HAZARDS CONSIDERATION

c. The proposed change to TS Table 3.3.6-1, Control Rod Withdrawal i

Block Instrumentation, deletes Note (e) from Trip Function 4.a, IRM i- detector-not-full-in rod block. This rod withdrawal block functions during Operational Conditions 2, Startup, and 5, Refuel, to assure j that IRMs are operable during control rod withdrawal in these plant Operational Conditions. The rod block is not bypassed when the

IRMs are on range 1. Thus Note (e) does not apply to this trip

function and is being deleted. Therefore, the correction of this l error will not change the probability or consequences of an

accident previously evaluated.

t i d. The change to TS Bases 3/4.3.1 to correct a typographical error

}' referencing TS Table 3.3.1-2, Note #, instead of Note ## is an

administrative change and thus will not change the probability or j consequences of an accident.

2) Create the possibility of a new or different kind of accident from any ,

j accident previously evaluated because: I t

l The changes to the definition of Channel Calibration and correction of the i

other miscellaneous errors in the TS and TS Bases will not create the

! possibility of a new or different kind of accident, because the changes will

! not affect the design or operation of any structure, system, or component.

j in the plant

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

a. The definition of Channel Calibration is being changed to be like the definition in NUREG 1434, Standard Technical Specifications General Electric Plants, BWR/6, Revision 1. The primary changes involve requiring only a inplace qualitative assessment of thermocouple and RTD sensors. These sensors are not adjustable and not susceptible to setpoint drift. Thus the appropriate check of

.the sensors is a qualitative assessment only. The inplace qualitative assessment assures operability of the sensors. I Therefore there is no reduction in the margin of safety.

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4 ATTACHMENT C

SIGNIFICANT HAZARDS CONSIDERATION

b. The remaining miscellaneous changes are corrections due to errors l t

in the TS. The corrections will make the associated TS consistent with the design and licensing basis of LaSalle or correct typographical errors. Therefore, there is no reduction in the margin of safety.

1 Guidance has been provided in " Final Procedures and Standards on No l 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 l hazards considerations. These proposed amendments most closely fit the  !

example of a change which 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 the acceptance criteria with respect to the system or component specified in the l Standard Review Plan. l 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.

I 1

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c

. ATTACHMENT D t

ENVIRONMENTAL ASSESSMENT GTATEMENT APPLICABILITY REVIEW Commonwealth Edison (Comed) has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR Part 51.21. It has been determined that the proposed changes meet the criteria for categorical exclusion as provided for under 10 CFR Part 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.

I 1

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