Text

Proposed Tech Spec Changes Modifying Opening LPCI & LPCS Injection Valves in Initiation Condition & Revising Reactor Vessel Matl Surveillance Removal Schedule
	ML20063A187
Person / Time
Site:	LaSalle
Issue date:	08/19/1982
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20063A178	List: ML20063A176; ML20063A187;
References
	NUDOCS 8208240227
	Download: ML20063A187 (18)
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ATT AC HMEN T L ASALLE COUNTY STATION UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST NPF -l l/8 2 -10

Subject:

Reactor Pressure Interlock on LPCI and LPCS Injection Valves References (a): License NPF-ll , Condition 2.C.17 (b): Supplement 2, Safety Evaluation Report, Section 6.3.4.

Background

Modification No. M-1-1-82-049 changes the permissive for opening the LPCI and LPCS injection valves in an initiation condition, to the condition that Rx pressure is less than 500 psig. The interlock prevents the automatic opening the injection valve when reactor pressure is above the design value o f the low pressure piping. This modification is being made to satisfy Reference (a) interfacing at Low and High Pressure as explained in Reference (b) . The concern o f the above license condition is to prevent the inadvertent overpres-surization of the low pressure piping in the event of an initiation condition concurrent with the failure o f an ECCS injection line check valve.

Discussion The modification consists o f 4 pressure switches sensing Rx Pressure in a "one out of two taken twice" configuration for each division, giving the permissive for opening the LPCS and LPCI injection valves during an initiation condition. This relay logic replaces the permissive signal o f dif ferential pressure across the injection valve being less than 722 psid. The changes made by the modification satisfy the concerns o f the NRC o f overpressurizing the low pressure piping, but the changes also have impact on the ECCS accident analysis is performed by GE f or LaSalle.

8208xaupa]

4 GE has performed an additional ECCS accident analysis to determine the impact of the modification on the predicted peak cladding temperatures. The modification of the injection valve logic causes

, an additional delay to the opening of the ECCS injection valves.

The results of the analysis show that for a valve opening time of 20 seconds, the delay in the ECCCS injection causes only a 1.6- second delay in core reflooding and a corresponding increase in the peak cladding temperature o f 10 0 F. There is sufficient margin in the LaSalle FSAR analysis to absorb the temperature increase and remain within the PCT limit o f 2200 F. 0 The modification is therefore acceptable from an ECCS viewpoint.

Conclusion Commonwealth Edison finds no unreviewed safety questions in this Technical Specification Change Request.

NOTE: THIS TECH SPEC CHANGE MUST BE EFFECTIVE ONLY UPON

{ COMPLETION OF THE INST ALLATION OF THE MODIFICATION.

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INSTRUMENTATION '

3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ,

LIMITING CONDITION FOR OPERATION !

3.3.3 The emergency core cooling system (ECCS) actuation instrumentat' ion

/

channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoir.t column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in/ Table 3.3.3-3.

APPLICABILITY: As shown in Table 3.3.3-1.

ACTION:

a. With an ECCS actuation instrumentation channel f i. rip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.3-2, declare the channel inoperab)6 until the channel is restored to OPERABLE status with its tri setpoint, adjusted consistent with the Trip Setpoint value,

b. With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3(3-1.

c. With either ADS trip system "A" or B" inoperable, restore the inoperable trip system to OPERABL, status within:

1. 7 days, provided that the HPCS and RCIC systems are OPERABLE.

2. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reducereactorsteamdome/pressureto less than or equal to 122 psig within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.3.3.1

/

Each ECCS actuat' ion instrumentation channel shall be demonstrated OPERABLE by the perform $nce of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION op'erations for the OPERATIONAL CONDITIONS and at the frequencies shown in' Table 4.3.3.1-1.

4.3.3.2 LOGIC Sy EM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.3.3 The CS RESPONSE TIME of each ECCS trip function shown in Table 3.3.3-3 shall be dem,/onstrated to be within the limit at least once per 18 months.

Each tes hall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ECCS trip system.

LA SALLE - UNIT 1 3/4 3-23 i

TABLE 3.3.3-1 5-m EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION

? MINIMUM OPERABLE APPLICABLE I

, CHANNELS PER TRIP OPERATIONAL TRIP FUNCTION FUNCTION (a) CONDITIONS ACTION E

Z

~ A. DIVISION I TRIP SYSTEM

1. RHR-A (LPCI MODE) & LPCS SYSTEM

a. Reactor Vessel Water Level - Low Low Low, Level 1 2(b) 1, 2, 3, 4*, 5* 30

b. Drywell Pressure - High 2(b) 1, 2, 3 30 i c. LPCS Pump Discharge Flow-Low (Bypass) 1 1, 2, 3, 4*, 5* 31 1 & LPa A qcm LPCSpInjection Valve [v...erential Pressure-Low @ 1,2,3 32 38 d.

(Pe r.T.i s;i vc ) LTeswcK 4 * , 5* 33 39

$ e. LPCI Injeet4en-Valve Differer.tial Pressure-Lcw -

1 1,2,3 4*, 5*

32 33

m. De(wtitPemiccive) 1, 2, 3, 4* , 5* 32 T f. LPCI Pump A Start Time Delay Relay 1

g. LPCI Pump A Discharge Flow-Low (Bypass) 1 1, 2, 3, 4*, 5* 31

h. Manual Initiation 1/ division 1, 2, 3, 4*, 5* 34

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#

a. Reactor Vessel Water Level - Low Low Low, Level 1 2(b) 1, 2, 3 30 coincident with

b. Drywell Pressure - High 2(b) 1, 2, 3 30

c. ADS Timer 1 1,2,3 32

d. Reactor Vessel Water Letel - Low, Level 3 (Permissive) 1 1,2,3 32

e. LPCS Pump Discharge Pressure-High (Permissive) 2 1,2,3 32

f. LPCI Pump A Discharge Pressure-High (Permissive) 2 1,2,3 32

g. Manual Initiation 1/ division 1,2,3 34 O O O -

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TABLE 3.3.3-1 (Continued) .

$ EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION N MINIMUM OPERABLE APPLICABLE 9;

CHANNELS PER TRIP OPERATIONAL FUNCTION (a) CONDITIONS ACTION E TRIP FUNCTION p

~ B. DIVISION 2 TRIP SYSTEM

1. RHR B & C (LPCI MODE)

a. Reactor Vessel Water Level - Low, Low Low, Level 1 2(b) 1, 2, 3, 4*, 5* 30

b. 2(b) 1, 2, 3 30 D

DellPressure-Highg c 38 .

c. LPCI3 Injection Valve D ppca..erentiel Pressure-Low if -1/vsive 1,2,3 32 (Pemiccive) _T.maa t.c cl< 4*, 5* -B- 39

d. LPCI Pump B Start Time Delay Relay 1 1, 2, 3, 4* , 5* 32 w

1 e. LPCI Pump Discharge Flow - Low (Bypass) 1/ pump 1, 2, 3, 4* , 5* 31 Manual Initiation 1/ division 1, 2, 3, 4*, 5* 34 Y f.

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#

l a. Reactor Vessel Water Level - Low Low Low, Level 1 2(b) 1, 2, 3 30 coincident with

b. Drywell Pressure - High 2(b) 1, 2, 3 30

c. ADS Timer 1 1,2,3 32

d. Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 32

e. LPCI Pump B and C Discharge Pressure - High 2/ pump 1,2,3 32 (Permissive) 1/ division 1, 2, 3 34

f. Manual Initiation e

TABLE 3.3.3-1 (Continued) 5 v, N EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION R '. MINIMUM OPERABLE APPLICABLE i

CHANNELS PE RIP OPERATIONAL CONDITIONS ACTION E TRIP FUNCTION 'N FUNCTION

+ x e C. DIVISION 3 TRIP SYSTEM

1. HPCS SYSTEM '

a. Reactor Vessel Water Leve'ly Low, Low, Level 2 4 1, 2, 3, 4*, 5* 35

b. Drywell Pressure - High ' 4 1,2,3 35 s

c. Reactor Vessel Water Level-High,xLevel 8 1, 2, 3, 4*, 5* 32 2(d) 2 1, 2, 3, 4*, 5* 36

d. Condensate Storage Tank Level-Low' '

e. Suppression Pool Water Level-High ' 2(d) 1, 2, 3, 4*, 5* 36

f. Pump Discharge Pressure-High (Bypass) 'sN 1 1, 2, 3, 4*, 5* 31 w g. HPCS System Flow Rate-Low (Permissive) 1 1, 2, 3, 4*, 5* 31 1 h. Manual Initiation s 1/ division 1, 2, 3, 4*, 5* 34 w

4*

D. LOSS OF POWER TOTAL NO.

\CHANNELS OPERABLE MINIMUM APPLICABLE OPERATIONAL OF CHANNELS TO TRIP CHANNELS CONDITIONS ACTION

1. 4.16 kv Emergency Bus Undervoltage 1/ bus 1/ bus x1/ bus 1, 2, 3, 4**, 5** 37 (Loss of Voltage) '

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

(b) Also actuates the associated division diesel generator.

(c) Provides signal to close HPCS pump discharge valve only on 2-out-of-2 logic.

l (d) Provides signal to HPCS pump suction valves only.

!

Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.

Required when ESF equipment is required to be OPERABLE.

Not required to be OPERABLE when reactor steam dome pressure is 5 122 psig.

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TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:

a. With one channel inoperable, place the inoperable channel in the tripped condition within one hour

or declare the associated system inoperable,

b. With more than one channel inoperable, declare the associated system inoperable.

ACTION 31 - With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip Function, place the inoperable channel in the tripped condition within one hour; restore the inoperable channel to OPERABLE status within 7 days or declare the associated system inoperable.

ACTION 32 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, declare the associated ADS trip system or ECCS inoperable.

ACTION 33 - With the nuinber-of-OPERABLE-channels-less-than-the-Minimur 3 -OPERABLE-Ehannels per-Trip-Function requirement; place-the V inoperable-channel-in-the-tripped-condi-tion within one-hour.

DeletecL ACTION 34 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or declare the associated ADS valve or ECCS inoperable.

ACTION 35 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement

a. For one trip system, place that trip system in the tripped condition within one hour

or declare the HPCS system inoperable.

b. For both trip systems, declare the HPCS system inoperable.

ACTION 36 - With the number of OPERABLE channels 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

or declare the HPCS system inoperable.

ACTION 37 - With the number of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 or 3.8.1.2, as appropriate.

V , "The provisions of Specification 3.0.4 are not applicable.

i LA SALLE - UNIT 1 3/4 3-27

TABLE 3.3.3-2 5-m EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETP0INTS

?

ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE C

5 A. DIVISION 1 TRIP SYSTEM

~ 1. RHR-A (LPCI MODE) AND LPCS SYSTEM

a. Reactor Vessel Water Level - Low Low Low, Level 1 >- 129 inches * >- 136 inches *

b. Drywell Pressure - High 7 1.69 psig 7 1.89 psig

c. LPCS Pump Discharge Flow-Low I 750 gpm I 640 gpm (d. 4PC-S-Injection Valve Dif ferential Pressurc--Low 709 psid, decreasing

- i:. e . LPCI Injection Valve Differential Pressure-Law II 720 psid, decreasing 720 psid, decreasing"I 709-ps-id,-decreasing

f. LPCI Pump A Start Time Delay Relay 7 5 seconds 7 6 seconds

/ g. LPCI Pump A Discharge Flow-Low I 1000 gpm I 550 gpm

h. Manual Initiation RA RA

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A" w

Reactor Vessel Water Level - Low Low Low, Level 1 >- 129 inches * >- 136 inches

h a.

b. Drywell Pressure - High 7 1.69 psig 7 1.89 psig

c. ADS Timer 7 105 seconds 7 117 seconds

d. Reactor Vessel Water Level-Low, Level 3 7 12.5 inches

I 11 inches *

e. LPCS Pump Discharge Pressure-High I 146 psig, increasing i 136 psig, increasing

f. LPCI Pump A Discharge Pressure-High 5 119 psig, increasing i 106 psig, increasing

g. Manual Initiation RA NA OLUG

[d.. LPCS R e n c.dr m LPG I"a' pAs w L>Js msI ~rm u ex c - Lc.w gg ggq 5 00 p>s #3 1 ZO psij C- Delete 1 9 9 e

TABLE 3.3.3-2 (Continued) g EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETP0INTS c-R ALLOWABLE

, TRIP FUNCTION TRIP SETPOINT VALUE C

2

8. DIVISION 2 TRIP SYSTEM

~

i 1. RHR B AND C (LPCI MODE)

a. Reactor Vessel Water Level - Low Low Low, Level 1 >- 129 inches * >- 136 inches *

b. Drywell Pressure - High 7 1.69 psig 7 1.89 psig g -f c. LPCI Injection Velve Differential Pressure--Lcw [729psid, decreasing ; 709 psid, decreasing

d. LPCI Pump B Start Time Delay Relay < 5 seconds < 6 seconds i e. LPCI Pump Discharge Flow-Low I 1000 gpm I 550 gpm

f. Manual Initiation NA NA

/ 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B" x*

! a. Reactor Vessel Water Level - Low Low Low, Level 1 >- 129 inches * >- 136 inches

Y b. Drywell Pressure - High 7 1.69 psig 7 1.89 psig D! c. ADS Timer 7 105 seconds 7 117 seconds i

d. Reactor Vessel Water Level-Low, Level 3 i 12.5 inches

I 11 inches *

e. LPCI Pump B amd C Discharge Pressure-High 5 119 psig, increasing i 106 psig, increasing

f. Manual Initiation NA NA i

. -c. Lf'cI B 4 C. I~scctn~ kve soogs,3 Sco essa 20 g3,3 Rnu f'ks5aa - Lo.o %rm w<x l

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TABLE 3.3.3-2 (Continued)

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y, EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS 1

E ALLOWABLE 9; VALUE TRIP FUNCTION TRIP SETPOINT C

5 C. DIVISION 3 TRIP SYSTEM

1. HPCS SYSTEM

\ J

a. Reactor Vessel Water Le 1 - Low Low, Level 2 1- 50 inches

3- 57 inches *

b. Drywell Pressure - High 5 1.69 psig 5 1.89 psig

c. Reactor Vessel Water Level - ligh, Level 8 5 55.5 inches

5 56 inches *

d. Condensate Storage Tank Level Low 3 715'7" 3 715'3"

e. Suppression Pool Water Level - Hi 5 700'1" $ 700'2"

f. Pump Discharge Pressure - High 1 120 psig 3 110 psig HPCS System Flow Rate - Low 1 1000 gpm 2 900 gpm

g. NA

h. Manual Intiation NA w

1 w D. LOSS OF POWER d2 o 1. 4.16 kv Emergency Bus Undervoltage (Loss of Voltage)#

a. 4.16 Kv Basis 2625 131 lts with 2625 262 volts with

1) Divisions 1 and 2 5 10 second (ti e delay 5 11 second time delay 2496 125 volts w h 2496 250 volts with 1 4 second time dela 1 3 second time delay 2870 143 volts with 870 1 287 volts with

2) Division 3 5 10 second time delay 51 second time delay

See Bases Figure B 3/4 3-1. %

These are inverse time delay voltage relays or instantaneous voltage relays with a time delay. in voltages shown are the maximum that will not result in a trip. Lower voltage conditions will resul decreased trip times.

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l TABLE 3.3.3-3 i -

0 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES ECCS RESPONSE TIME (Seconds)

! 1. LOW PRESSURE CORE SPRAY SYSTEM < 40

1

2. LOW PRESSURE C00}. ANT INJECTION MODE 6% #

1 0F RHR SYSTEM L Pomes A , B, l. e )

c. P=p:; A ar.d B

-b - Pump-C i 45 3 4 0---

3. AUTOMATIC DEPRESSURIZATION SYSTEM NA

4. HIGH PRESSURE CORE SPRAY SYSTEM < 27

5. LOSS OF POWER NA l ray neove r>+es C,meteo,. e o v m s o. a. m

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G TABLE 4.3.3.1-1 r- EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS "m OPERATIONAL CHANNEL c'- CONDITIONS FOR WHICH CHANNEL 5 CHANNEL FUNCTIONAL TEST CALIBRATION SURVEILLANCE REQUIRED TRIP FUNCTIO _'i CHECK

]

A. DIVISION I TRIP SYSTEM

1. RHR-A (LPCI MODE) AND LPCS SYSTEM

a. Reactor Vessel Water Level - 1, 2, 3, 4* , 5*

Low Low Low, Level 1 5 M R

b. Drywell Pressure - High NA M Q 1,2,3 LPCS Pump Discharge Flow-Low M 1, 2, 3, 4*, 5*

s c. NA Q bWo LFCIJ d. LPWInjection Valve Dif fcccatici Rom.rt>a.

M R 1, 2, 3, 4*, 5*

Pressure-Low t oncn u<<.- N A

e. -LCPI Injectica Vehe Dif fcrcatic!

t'

P mssurc-Lc.: S M l? 1, 2, 3, 4* , 5*

';' f. LPCI Pump A Start Time Delay M 1, 2, 3, 4*, 5*

M Relay NA Q M 1, 2, 3, 4*, 5*

g. LPCI Pump A Flow-Low NA Q NA 1, 2, 3, 4*, 5*

n. Manual Initiation NA R

2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#

a. Reactor Vessel Water Level -

Low Low Low, Level 1 5 M R 1, 2, 3

b. Drywell Pressure-High NA M Q 1,2,3

c. ADS Timer NA M Q 1,2,3

d. Reactor Vessel Water Level -

Low, Level 3 S M R 1,2,3

e. LPCS Pump Discharge Pressure-High NA M Q 1,2,3

f. LPCI Pump A Discharge Pressure-High NA M Q 1, 2, 3

g. Manual Initiation NA R NA 1,2,3 i

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5 TABLE 4.3.3.1-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS

'i' c:

CHANNEL OPERATIONAL 5 CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH

] TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED B. DIVISION 2 TRIP SYSTEM

1. RHR B AND C (LPCI MODE)

a. Reactor Vessel Water Level -

Low Low Low, Level 1 S M R 1, 2, 3, 4*, 5*

b. Drywell Pressure - High NA M 1,2,3 (B y c. LPCIFInjection Valve Differe-tial Remcroc Pressure-Low Turtoeux Q

-S- W A 4*, 5*

i M R 1, 2, 3,

d. LPCI Pump B Start Time Delay Relay NA M Q 1, 2, 3, 4*, 5*

e. LPCI Pump Discharge Flow-Low NA M 1, 2, 3, 4*, 5*

Q R f. Manual Initiation NA R NA 1, 2, 3, 4* , 5*

i' 2. AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#

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a. Reactor Vessel Water Level -

Low Low Low, Level 1 S M R 1,2,3

b. Drywell Pressure-High NA M 1, 2, 3 Q

c. ADS Timer NA M l , 2, 3 Q

d. Reactor Vessel Water Level -

Low, Level 3 S M R 1,2,3

e. LPCI Pump B and C Discharge Pressure-High NA M Q 1, 2, 3

f. Manual Initiation NA R NA 1,2,3

)

g TABLE 4.3.3.1-1 (Continued) h EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS m OPERATIONAL CHANNEL J

= CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH

-4 TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED C. DIVISION 3 TRIP SYS EM

1. HPCS SYSTEM

a. Reactor Vessel Water Level -

Low Low, Level 2 S M R 1, 2, 3, 4*, 5*

b. Drywell Pressure-High NA M Q 1,2,3

c. Reactor Vessel Water Level-High Level 8 $ M R 1, 2, 3, 4*, 5*

d. Condensate Storage Tank Level - \

1, 2, 3, 4* , 5*

u e.

Low Suppression Pool Water NA

\ x M Q S Level - High NA \M Q 1, 2, 3, 4*, 5*

w f. Pump Discharge Pressure-High NA Mx Q 1, 2, 3, 4* , 5*

M 'N 1, 2, 3, 4*, 5*

g g. HPCS System Flow Rate-Low NA

'N Q

1, 2, 3, 4*, 5*

h. Manual Initiation NA R NA D. LOSS OF POWER 5**

1. 4.16 kv Emergency Bus Under- NA NA R 1, 2, 3, 4**,

voltage (Loss of Voltage)

Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 122 psig.

When the system is required to be OPERABLE after being manually realigned, a's 'applicable, per Specification 3.5.2. s

- Required when ESF equipment is required to be OPERABLE.

4 f

(

1 _ - _ - _ _ _ _ _ - _ - _ _

REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE O)~ LIMITING CONDITION FOR OPERATION

> /

/

3.4.3.2 Reactor coolant system leakage shall be limited to: /

a. No PRESSURE B0UNDARY LEAKAGE.

b. 5 gpm UNIDENTIFIED LEAKAGE. /

c. 25 gpm total leakage averaged over any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period. [

d. 1gpmleakageatareactorcoolantsystempressureat100di10psig from any reactor coolant system pressure isolation valve'specified in Table 3.4.3.2-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, be in at ieastf HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within/the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

/

b. reater than the limits in b With and/orany reactorreduce c, above, coolantthe system leakage leakage rate j;to within the limits within 4hoursorbeinatleastHOTSHUTDOWp'withinthenext12hoursand in COLD SHUTDOWN within the followin

c. With any reactor coolant system pre /g valve ssure isolation 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

leakage h

Q 1 greaterthantheabovelimit,is,oiatethehighpressureportionof the affected system from the lpw pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two closed valves, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months an 'in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. With one or more high/ low pressure interface valve leakage pressure monitors inoperable, r,estore the inoperable monitor (s) to OPERABLE status within 7 days jor verify the pressure to be less than the alarm setpoint at 1 east once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by local indication; restore the inope76ble monitor (s) to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN with 'the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.4.3.2.1 Thereac/ tor coolant system leakage shall be demonstrated to be withineachof)tfeabovelimitsby:

a. Monitoring the primary containment atmospheric particulate and g eous radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. Monitoring the primary containment sump flow rate at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

. Monitoring the primary containment air coolers condensate flow rate at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

v LA SALLE - UNIT 1 3/4 4-7

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE:

a. Pursuant to Specification 4.0.5, except that in lieu of any leakage testing required by Specification 4.0.5, each valve shall be demonstrated OPERABLE by verifying leakage to be within its limit:

1. At least once per 18 months, and

2. Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.

dition, until the LPCS system and the LPCI system injectio v ferential pressure-low permissive is modified dujr'ng or before the 'est refueling outage, the LPCS systemJc eervalve (jj 1E21-F006 and tbRCI system check valves 1Ep1 -41 A, B, and C y shall also be demonsthted OPERABLE by ve '-ffing leakage to be O within its limit: \

E 1. Whenevertheunithpa en in LD' SHUTDOWN or REFUELING, after C thelastvalvejisturbancepriortoreactorcoolantsystem 14 temperature-exceeding 200 F M) /

2. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve disturbance except when 'nsCOLD y SHUTDOWN or REFUELING. N _

The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.

b. By demonstrating OPERABILITY of the high/ low pressure interface valve leakage pressure monitors by performance of a:

l 1. CHANNEL FUNCTIONAL TEST at least once per 31 days, and l 2. CHANNEL CALIBRATION at least once per 18 months, With the alarm setpoint for the:

1. HPCS system 5 100 psig.

2. LPCS system 5 500 psig.

3. LPCI/ shutdown cooling system 5 400 psig.

x

4. RHR shutdown cooling 5 190 psig.

5. RCIC 1 60 psig.

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

1 AT T AC HMEN T L ASALLE COUNTY STATION UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST NPF -11/82 -12 Subj ec t : Revised Reactor Vessel Material Surveillance Removal Schedule References (a): C. W. Schroeder letter to A. Schwencer dated April 8, 1982, submitting revised Reactor Vessel Material Surveillance Removal Schedule.

(b): A. Schwencer letter to L. O. DelCeorge dated May 17, 1982, Accepting Change in Withdrawal Schedule.

Background and Discussion The original Reactor Vessel Material Surveillance Withdrawal Schedule specified removal of samples from each of the sample locations o f 300, 120, and 30 degrees at each withdrawal period (10 and 30 EFPY) . The design of the surveillance specimen holder is welded, containing 6 capsules within, representing tensile and impact of base, welc and HAZ (heat affected zone) metals. To facilitate removal of the samples for examination, the table has been revised to remove one surveillance holder at the designated time period from one location instead of samples from each location since each holder is welded closed. This has been accepted by the NRC in Reference (b).

Each sample holder contains representative vessel material samples and will receive essentially the same neutron flux in any of the l three locations.

l

Conclusion:

1 i Commonwealth Edison finds no unreviewed safety questions in this Technical Specification Change Request. The NRC has already reviewed and approved this change in Reference (b) .

l 4799N

4:m

'o UNITED STATES 8\

n NUCLEAR REGULATORY COMMISSION C h WASHINGTON, D. C. 20555 ht.J J, u Q

%*****/ MAY 7 1962 Docket No.,50-373 V-Mr. Louis 0. DelGeorge Director of Nuclear Licensing Commonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690

Dear Mr. DelGeorge:

Subject:

La Salle County Station, Unit 1 Revised Reactor Vessel Materials Surveillance Program Withdrawal Schedule s

In Commonwealth Edison's letter dated April 8,1982, it was indicated ,

that the specimen holders containing the surveillance specimens 'for the reactor vessel materials are welded-closed enclosures. Therefore, you stated that it was not realistic to remove samples from each holder to ob,tain the number. of surveillance specimens required, but instead proposed to remove an entire holder at a time to obtain the required specimens.

We find the requested change acceptable.

' Sincerely.

0 %

A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing cc: See next page, u

m ']

g - :s >

s 1

... . .~ .. .. . - . . . . . . . ,. .. . . . .. .

~

. Commonwealth Edison

.* one First National Plaza. Ch.cago. Ilknois

' Address Reply to: Post Office Box 767

. Chicago. tilinois 60690 i April 8, 1982 l

Mr. A. Schwence r, . Chie f Licensing Branch #2 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

LaSalle County Station Unit 1 Revised Reactor Vessel Material Surveillance Program Withdrawal Schedule NRC Docket No. 50-373

Dear Mr. Schwencer:

The purpose of this letter is to transmit a revised Tech 5pec Table 4.4.6.1.3-1. This table is being revised because it has come to our attention that the specimen holders are welded closed enclosures. Therefore, it is not realistic to remove samples from the holders. Instead, as shown on the attached revised table, we now propose to remove an entire specimen holder at a time.

If there are any further questions in this matter, please contact this office.

Very truly yours, l

% f8)fI%

C. W. Schroeder Nuclear Licensing Administrator i

l Im 1 .

cc: NRC Resident Inspector - LSCS 3850N ,

(

s CT -

4 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITH0RAWAL SCHEDULE TABLE 4.4.6.1.3-1 SPECIMEN HOLDER VESSEL LOCATION LEAD FACTOR WITH0RAWAL TIME (EFPY) 117C4936G010 3000 0.6 10 117C4936G011 1200 0.6 30 117C4936G012 300 0.6 Spare Neutron Desimeter 300 1st Refuel Outage 1

3850N

l

n r

9 TABLE 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM-WITHDRAWAL SCH'EDULE R

CAPSULE SOLDER VESSEL LEAD WITHDRAWAL TIME LOCATION FACTOR (EFPY)

$ NUMBER

~

i G0 10 3009 1.61IC4936_00[u2~-

117C4936 120.-

-1Mf393500Q3 -0309 - 0.6 10 1HE4936G00f- - (- 300" -

-1RC49366065 -15LO"

-4WC4936G006 =.-- - 030N.

2. 117C4030C007 300

-117C4030G000 120"-

w -117040300000 -030*- 0.6 30 l 1 7 00"-

A 1 (lilC43360010_._-

117C4936G011 150'?"300012 120 M

e 117C -seep 03 o *)

l

3. (g4936G01/2. 9g,ggg=. .___ _1ggc.;- -

-1HO'4936GQ15 - .. ' 0309- 0.6 Spare

-1MC403500pQ 300'-

li,3u ouuvi, -12'0 -

417C40350018- 03US-030 t% l4L. QluS VQL l

l

s -

REACTOR C0OLANT SYSTEM , .\_

u REACTOR STEAM DOME

<h LIMITING CONDITION FOR OPERATION 6 y\

/

3.4.6.2 The pressure in the reactor steam dome shall be less than 10 0'psig.

APPLICABILITY: OPERATIONAL CONDITION 1* and 2*. /

ACTION:

/

With the reactor steam dome pressure exceeding 1020 psig, reduce the pressure to less than 1020 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

/

SURVEILLANCE REQUIREMENTS [

/ 1 4.4.6.2 The reactor steam dome pressur'e shall be verified to be less than 1020 psig at least once per 12 bours.,

[

/

,/

/

A !

Notapplicable,duringanticipatedtransients.

/

l

/

j!

/

',/

/

./ O

/ LA SALLE - UNIT 1 3/4 4-20 i

[

/

oe - -

f J

ATTACHMENT B i

j Status of Tech Spec Change Requests i

NRC

Topic Submitted Ac tio n NPF-ll/ 82-7 SRM Countrate 6/14/82 Issued Am. 2

(Required prior to source 7/09/82 decay below 3 cps).

NPF-ll/82-8 Revise RCIC suction delta p 7/02/82 Issued Am. 2 alarm setpoint (Required 7/09/82 prior to pressurization).

NPF-11/82-9 Add commitment to complete 7/14/82 Is sue d Am . 3 torque checks on bolting on 7/15/82 S/R valves outside containment (NRC Requested to be submitted

, ASAP) .

NPF-11/ 82-10 Add ECCS delta p specification 8/19/82 to reflect modification (License

, Condition 2.C.17) l NPF -11/ 82-ll Revise RCIC surveillance to 8/11/82 Verbal

account for flow differences approval between test flow and normal 8/13/82 flow paths (Required prior to restart a f ter next shutdown /

upgraded to emergency change on 8/12/82).

1

NPF -11/ 82-12 Revise Reactor Vessel Materials 8/19/82 Surveillance Program Withdrawal i

Schedule (Previously approved by NRC on 5/07/82) .

1 i

1 1

4730N

ML20063A187

Contents

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Subject:

Background

Conclusion:

Dear Mr. DelGeorge:

Subject:

Subject:

Dear Mr. Schwencer:

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ML20063A187

Text

Subject:

Background

Conclusion:

Dear Mr. DelGeorge:

Subject:

Subject:

Dear Mr. Schwencer:

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