Text

Proposed Tech Specs Associated W/Exemption Requests from App J for LaSalle Unit 2 to Return to Cilrt Type a Test Schedule of Three Times in 10 Yrs & for LaSalle Units 1 & 2 to Decouple Cilrt Type a Test Schedule from ISI Schedule
	ML20078B957
Person / Time
Site:	LaSalle
Issue date:	10/24/1994
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20078B953	List: ML20078B949; ML20078B957;
References
	NUDOCS 9410270306
	Download: ML20078B957 (156)
	v • d • e

.

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS NPF-11 NPF-18 Index page I Index page I Index page II Index page II Index page VII Index page VII Index page XXIII Index page XXII 1-3 1-3 Insert page B-3 Insert page B-3 1-4 1-4 .

1-5 1-5 1-6 1-Sa 1-7 1-6 1-8 1-7 3/4 3-11 3/4 3-11 3/4 3-12 3/4 3-12 3/4 3-13 3/4 3-13 i- 3/4 3-14 3/4 3-14 3/4 3-19 3/4 3-19 3/46-1 3/4 6-1 Insert page B-4 Insert page B-4 3/46-2 3/4 6-2 3/46-3* 3/46-3*

3/46-4* 3/46-4*

3/4 6-18 3/4 6-21 Insert page B-5 Insert page B-5 3/4 6-19 3/4 6-22 3/4 6-22 3/4 6-25 In: ert page B-6 Insert page B-6 2/4 6-23 3/4 6-26 Insert page B-7 Insert page B-7 3/4 6-24 3/4 6-27 3/4 6-25* 3/4 6-28*

3/4 6-26* 3/4 6-29*

3/4 6-27* 3/4 6-30*

3/4 6-28* 3/4 6-31*

3/4 6-29* 3/4 6-32*

k: s a' a s..la salle tilrt fina .wref 4 3 9410270306 941024 FOR ADOCK 05000373 P PDR

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE /I'ECHNICAL SPECIFICATIONS NPF-11 NPF-18 3/4 6-30* 3/4 6-33*

3/4 6-31* 3/4 6-34*

3/4 6-31a* 3/4 6-35*

3/4 6-32* 3/4 6-36*

3/4 6-33* 3/4 6-37*

3/4 6-34* 3/4 6-37a*

3/4 6-34a* 3/4 6-38**

3/4 6-35** 3/4 6-39 I 3/4 6-36 3/4.6-46 3/4 6-43 Bases page B 3/4 6-1 Bases page B 3/4 6-1 Insert page B-8 Insert page B-8 Bases page B 3/4 6-4 Bases page B 3/4 6-4 Insert page B-9 Insert page B-9 Insert page B-10 Insert page B-10 License No. NPF-18, License No. NPF-11, page 9 page 16 Insert page B-12 Insert page B-11

Deleted page.

Page included for information only, no changes.

i l

cs nlaVanallestirtfina.wpf44

t INDEX

DEFINITIONS ;

SECTION 1.0 DEFINITIONS PAGE 1.1 ACTI0N............................................................ 1-1

1. 2 AVE RAGE P LANAR EXP05URE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.4 CHANNE L CALI B RATI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.5 CHANN E L CHE C K. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.6 CHANNEL FUNCTI ONAL TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.7 CO R E ALT E RAT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.8 CORE OPERATING LIMITS REP 0RT. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 1.9 C RITI CAL POWER RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-0 1.10 DOS E EQUI VALENT I-131. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2

1. n E- AVERAGE DISINTEGRATION ENERGY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 I 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME. . . . . . . . . . . . . . . . 1-2 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME......... 1-2 1.14 FRACTION OF LIMITING POWER DENSITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.15 FRACTION OF RATED THERMAL P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.16 FREQUENCY N0TATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.17 GASEOUS RADWASTE TREATMENT SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.18 I DE NTI FI ED LEAKAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1.19 ISOLATION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3

1. 2M LIMITING CONTROL ROD PATTERN. . . . . . . . . . . . . . . . . . . . . . . . . .1-3 ............

.2

1. 2/ LINEAR HEAT GENERATION RATE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-4 ....

1. 2/ LOGIC SYSTEM FUNCTIONAL TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.2 MAXIMUM FRACTION OF LIMITING POWER DENSITY. . . . . . . . . . . . . . . . . . . . . . . . 1-4

1. 2/ EMB E R ( 5 ) 0 F THE P UB LI C . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 !

1. I N I MUM C RI TI C AL POWE R RATI 0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 !

-~

. . . . . . . . . . . . . . . . . 1-4

'N1. 2/- UNIT OFFSITE DOSE * ~ ' f ' CALCULATION AmendmentMANUAL. .

LA SALLE 1 ~/,20 m/_.a "*I No..es i

4 .

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INDEX O'

DEFINITIONS SECTION DEFINITIONS (Continued) PAGE 1.2 OPERABLE - OPERABILITY............................................ 1-5 0

1.2MOPERATIONALCONDITION-CONDITION................................. 1-5 30 1.hl PHYSICS TESTS..................................................... 1-5 3

1.3(PRESSUREBOUNDARYLEAKAGE......................................... 1-5 S.

1. PRIMARY CONTAINMENT INTEGRITY..................................... 1-5 1.3 PROCESS CONTROL PR0 GRAM........................................... 1-6 4

1.3% PURGE - PURGING................................................... 6 6

1-6 1.3\RATEDTHERMALP0WER...............................................

1.3 REACTOR PROTECTION SYSTEM RESPONSE TIME........................... 1-6 1-6 1.3fREPORTABLEEVENT..................................................

1.3fRODDENSITY....................................................... 1-6 1.3)SECONDARYCONTAINMENTINTEGRITY................................... 1-7

1. SHUTDOWN MARGIN................................................... 1-7 1-T 1..(SITE.00NoARY.....................................................

1.4fSOURCECHECK...................................................... 1-8 .

1.4 STAGGERED TEST BASIS.............................................. 1-8 1.4 THERMAL P0WER..................................................... 1-8 !

5 1.4kTURBINE'VPASSRESPONSETIME....................................., 1-8

1. 4 UNIDENTIFIED LEAKAGE.............................................. 1-8 i 1-8

1. .)7 VENTILATION EXHAUST TREATMENT SYSTEM..............................

1.4fvENTING........................................................... 1-8 LA SALLE - UNIT 1 11 Amendment No. 85 l

a 1RQEl LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 4

SECTION EAGE 3/4.6 CONTAINMENT SYSTitil 3/4.6.1 PRIMARY. CONTAINMENT Integrity............................. 3/4 6-1

, g Primary Containment (PrimaryContainmentLeakage...............................

3/46) !

Primary Containment Ai r Locks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-5 >

MSIV Leakage Control. System............................... -3/4 6-7 i

Drywell and Suppression Chamber Internal Pressure......... 3/4 6-13 Drywell Average Air Temperature........................... 3/4 6-14 Drywell and Suppression Chamber Purge System.............. 3/4 6-15 3/4.6.2 DEPRESSURIZATION SYSTEMS

~

Suppression Chamber....................................... 3/4 6-16 S up p re s s i on Pool S p ray . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-20 Suppression Pool Cooling.................................. 3/4 6-21 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES...................... 3/4 6-22 3/4.6.4 VACUUM RELIEF............................................. 3/4 6-35 3/4.6.5 SECONDARY CONTAINMENT Secondary Containment Integrity........................... 3/4 6-37 l Secondary Containment Automatic Isolation Dampers. . . . . . . . . 3/4 6 Standby Gas Treatment System.............................. 3/4 6-40 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL l t

Drywell and Suppression Chamber Hydrogen Recombiner Systems................................................. 3/4 6-43 ;

Drywell and Suppression Chamber Oxygen Concentration...... 3/4 6-44 LA SALLE - UNIT 1 VII Amendment No. 100 t

8 g -. c. 7

I INDEX !

LIST OF TABLES (Continued) j EAE TABLE 4

(3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES

....T.~........ -=_

3/4 6-24 3.6.5.2-1 SECONDARY CONTAINMENT VENTILATION SYSTEM AUTOMATIC ISOLATION DAMPERS ....................... 3/4 6-39 3.7.5.2-1 DELUGE AND SPRINKLER SYSTEMS ...................... 3/4 7-16 3.7.5.4-1 FIRE HOSE STATIONS ................................ 3/4 7-19 3.7.7-1 AREA TEMPERATURE MONITORING ....................... 3/4 7-25 4.8.1.1.2-1 DIESEL GENERATOR TEST SCHEDULE .................... 3/4 8-7b 4.8.2.3.2-1 BATTERY SURVEILLANCE REQUIREMENTS ................. 3/4 8-18 3.8.3.3-1 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION ........................................ 3/4 8-27 83/4.4.6-1 REACTOR VESSEL TOUGHNESS .......................... B 3/4 4-6 5.7.1-1 COMPONENT CYCLIC OR TRANSIENT LIMITS .............. 5-6 1

LA SALLE - UNIT 1 XXIII Amendment No. 100 i

k

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DEFINITIONS ,

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME (Continued) breaker trip coil from when the monitored parameter exceeds.its' trip setpoint at the channel sensor of the associated:

a. Turbine stop valves, and

b. Turbine control valves.

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

FRACTION OF' LIMITING POWER DENSITY 1.14 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LNGR existing 'l -

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

FRACTION OF RATED THERMAL POWER 1.15 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the seasured THERMAL l POWER divided by the RATED THERMAL POWER. -

FREQUENCY NOTATION 1.16 The FREQUENCY NOTATION specified for the performance of Surveillance l Requirements shall correspond to the intervals defined in Table 1.1.

GASEOUS RADWASTE TREATMENT SYSTEM .

1.17 A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and I installed to reduce radioactive gaseous effluents by collecting primary '

coolant system offgases from the primary system and providing for delay or holdup for the purpose,of reducing the total radioactivity prior to '

release to the environment. '

IDENTIFIED LEAKAGE 1.18 IDENTIFIED LEAKAGE shall be: .

I

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or

b. Leakage into the containment atmosphere froe sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to ,

be PRESSURE B0UNDARY LEAKAGE. ,

ISOLATION SYSTEM RESPONSE TIME 1.19 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when l the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times i shall include diesel generator starting and sequence loading delays where applicable. The response time may be seasured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LIMITING CONTROL R00 PATTERN 1.2k A LIMITING CONTROL R0D PATTERN shall'be a pattern which results in the core being on a themal hydraulic limit, i.e., operating on a limiting l value for APLHGR, LHGR, or MCPR.

Add

INSERT A" LA SALLE UNI 1-3 Amendment No. 70 e

.m.. _.-.. -.

l ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSEfI'ECHNICAL SPECIFICATIONS page B-3 INSERT A i La 1.20 The maximum allowable primary containment leakage rate, L,, shall be 0.635 % of primary containment air weight per day at the calculated peak containment pressure (P, = 39.6 psig).

k :' nla tlasallet Irt.fina .wp f 45

1

.. )

DEFINITIONS .

.. l LINEAR HEAT GENERATION RATE neration per unit 1.2fLINEAR length of fuel rod. It is the integral of the Mt f um over theshall HEAT GENERATION RATE (LMGR) heat be the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST 1.2f A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e., all relays and contacts, all trip units, solid state logic elements, etc. of a logic circuit, from sensor through and including the actuated device to verify OPERABILITY. THE LOGIC SYSTEM FUNCTIONAL TEST may be .

performed by any series of sequential, overlapping or total system steps ,

such that the entire logic system is tested.

MAXIMUM FRACTION 0F LIMITING POWER DENSITY 1.2fThe MAXINUN FRACTION OF LIMITING POWER DENSITY (WLPD) shall be the highest value of the FLPD which exists % the core.

MEMBEk(S) 0F THE PUBLIC 1.2fMEMIER(S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include -

persons who use portions of the site for rrereational, occupational, or other purposes not associated with the plant.

MINIMUM CRITICAL POWFR RATIO 1.2kl0The MINIMUN CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFFSITE DOSE CALCULATION MANUAL 1.2fThe OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contal methodology and parameters used in the calculation of offsite doses i resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct of the Environmental Radiological-Monitoring Program. The ODCM shall also contain (1) the Radioactive l Effluent Controls and Radiological Environmental Monitaring Programs required by Technical Specification Section 6.2.F.4 and (2) descrip- .

tions of the information that should be included in the Annual )

Radiological Environmental Operating and Semi-Annual Radioactive 1 Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

4 LA SALLE UNIT 1 1-4 Amendment No. 85 l

i

' DEFINITIONS OPERABLE - OPERABILITY 1.2 A system, subsystem, train, component or device shall'be OPERABLE or have l l OPERABILITY when it is capable of performing its specified function (s), i and when all necessary attendant instrumentation, controls, a normal and '

an emergency electrical power source, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION - CONDITION

1. 2 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive l combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS

1. PHYSICS TESTS shall be thros tests performed to measure the fundamental l nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

for valveStb4 are open unc er,-

PRESSURE BOUNDARY LEAKAGE adenjnj pp4 [

1.3 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable au t l in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY

?

1.3k~PRIMARYCONTAINMENTINTEGRITYshallexistwhen: l

a. All primary containment penetrations required to be closed during accident conditions are either.:

1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, rovided in Table 3.6.3-1 o pecificatio p

b. All primary containment equipment hatches are closed and sealed,

c. Each primary containment air lock is OPERABLE ur uant to Specification 3.6.1.3.

gfgjf N

d. The primary containment leaka e rates are within tne limits pecification 3.6.1.2) gg gpgg

% , t. t.b, LA SALLE UNIT 1 1-5 Amendment No. 85

. DEFINITIONS

e. The suppression chamber is OPERABLE pursuant to Specifir.ation 3.6.2.1.

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

g. Primary containment structural integrity has been verified in

.accordance with Surveillance Requirement 4.6.1.1.e.

PROCESS CONTROL PROGRAM l

1.3/khePROCESSCONTROLPROGRAM.(PCP)shallcontainthecur formulas, . sampling, analyses, test, and determinations to be made to l ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual 'ur simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 61, and 71, State regulations, burial- ground requirements, and other requirements governing the disposal of solid radioactive waste. .

-PURGE - PURGING 1.3 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replace-ment air or gas is required to purify the confinement. ;

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RATED THERMAL POWER l 1.3 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3323 MWT.

REACTOR PROTECTION SYSTEM RESPONSE TIME t

1.3/ EACTOR when the PROTECTION SYSTEM monitored parameter RESPONSE exceeds TIME shall its trip setpoint be channel at the the time interval from sensor until de-energization of the scram pilot valve solenoids. The 1 response time may be measured by any series of sequential, overlapping or j total steps such that the entire response time is measured. l l

REPORTABLE EVENT ;

1.3/ A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. ,

l R0D DENSITY l 1.3)8 ROD DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod. notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

1 LA SALLE UNIT 1 1-6 Amendment No. 100 )

)

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4 DEFINITIONS SECONDARY CONTAlmENT INTEGRITY-

1. SECONDARY CONTAIMENT INTEGRITY shall exist when:

a. All secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one' manual valve, blind flange, or -

deactivated automatic damper secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

b. All secondary containment hatches and blowout panels are closed and sealed.

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c. The standby gas treatment system is OPERABLE pursuant to Specification 3.6.5.3.

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

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

penetration, e.g. , welds, bellows or 0-rings, is OPERABLE.

f. The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.

SHUTDOWN MARGIN L.

1.E) SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be suberitical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.

SITE BOUNDARY 1.4 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

l LA SALLE UNIT 1 1-7 Amendment No. B5 I

DEFINITIONS

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SOURCE CHECK 1.4\ASOURCECHECKshallbethequalitativeassessmentof.channelresponse l when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.4fA STAGGERED TEST BASIS shall consist of: l

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

i

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

THERMAL POWER 1.4 THERMAL POWER shall be the total reactor core heat transfer rate to the l reactor coo 1 ant.

TURBINE BYPASS SYSTEM RESPONSE TIME S r 1.4\TheTUR8INEBYPASSSYSTEMRESPONSETIMEshallbetimeintervalfromwhen l the turbine bypass control unit generates a turbine bypass valve flow signal until the. turbine bypass valves trani to their required positions. -

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

UNIDENTIFIED LEAKAGE 1.4 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

VENTILATION EXHAUST TREATMENT SYSTEM 1.4fAVENTILATIONEXHAUSTTREATMENTSYSTEMshallbeanysystemdesignedand l installed to reduce gaseous radiciodine or radioactive material in particu-late form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any '

, effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING 1.4 VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

LA SALLE UNIT 1 1-8 Amendment No. 85 1

f ~- . .

5 TABLE 3.3.2-1 l- ISOLATION ACTUATION INSTRUMENTATION GROUPS MINIMUM OPERABLE APPLICABLE E OPERATED BY '

q TRIP FUNCTION

SIGNAL a CHANNELS PER OPERATIONAL' l TRIP SYSTEM (b) CONDITION ACTION A. AUTOMATIC INITIATION '

1. PRIMARY CONTAINMENT ISOLATION

a. Reactor Vessel Water Level (1) Low, level'3 7 2 1,2,3 20 (2) Low Low, Level 2- 2, 3 2 (3) Low Low Low, Level 1 1,2,3 20 1, 10 2 1,2,3 20

b. Drywell Pressure - High 2, 7, 10

' 2 '1, 2, - 3 20 R c. Main Steam Line

+ 1) Radiation - High 1 2 1,2,3 y 3 21

- 2 1,2,3 22

" 2) Pressure - Low 1 2 1 23

3) Flow - High 1 2/line(d) 1, 2, 3 21

d. Main Steam Line Tunnel Temperature - High 1 fIII5) 2 I IIIII) , 2 ,

e.

3(1)(j) 21 Main Steam Line Tunnel ATemperagre-High 1 2 1 0)(3s#, 2 I III) 3(1)(j) 21

f. Condenser' Vacuum - Low 1 2 1, 2 * , 3

21 -

2. SECONDARY CONTAINMENT ISOLATION

a. Reactor Building Vent Exhaust h Plenum Radiation - High 4(c)(e) 2 1, 2, 3 and ** 24 j b. Drywell Pressure - High 4(c)(e) 2 1,2,3 24 x c. Reactor Vessel Water Level - Low Low, Level 2 4 ICII")- #

2 1, 2, 3, and 24

% d. Fuel Pool Vent Exhaust Radiation - High 4(c)(e) 2 I, 2, 3, and ** 24

I ,

te l 9 TABLE 3.3.2-1_ (Continued) '

y =

'~ ISOLATION y~ ACTUATION INSTRUMENTATION VALVE GROUPS

' HINIMUH' OPERABLE APPLICABLE TRIP FUNCTION OPERATED e CHANNELS PER

_ SIGNAL a) OPERATIONAL h 3.

TRIP SYSTEM (b)_ _ CONDITION ACTIDH

" REACTOR WATER CLEANUP SYSTEM ISOLATION

a. A Flow - High 5 1

b. Heat Exchanger Area 1, 2, 3 22 Temperature - High 5 1/ heat 1, 2, 3

c. exchanger 22 Heat Exchanger Area l Ventilation AT - High 5 1/ heat 1, 2, 3 22

d. exchanger [

SLCS Initiation S N

!:* NA 1, '2,

e. Reactor Vessel Water 3 22 Level - Low Low, Level 2 5

4. 2 1,2,3 22 REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION

a. RCIC Steam Line Flow - High 8

b. RCIC Steam Supply 1

1, 2, 3 22 Pressure - Low 8, 9 g) g 2

c. RCIC' Turbine Exhaust ,1,2,3 22 Diaphragm Pressure - High 8 2 1,2,3

d. RCIC Equipment Room 22 Temperature - High 8 .

1

s. 1,2,3 22 RCIC Steam Line Tunnel Temperature - High 8 k

=

f.

RCIC Steam Line Tunnel 1

1, 2, 3 22 A Temperature - High 8 i 1 1, 2, 3 E 22

g. Drywell Pressure - High E 9(8 2 1, 2, 3 22

b. RCIC Equipment Room -

o' A Temperature - High 8 1 1,2,3 22 l

___2__-___ _ _ _ _ _ _ m ____ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _

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t- ;

TABLE 3.3.2-1 (Continued) g . ISOLATION ACTUATION INSTRUMENTATION F

m h

VALVE GROUPS MINIMUM GPERA8tE APPLICA8LE a

OPERATED f CHANNELS PER OPERATIONAL .

c z TRIP FUNCTICM SIGNAL a TRIP SYSTEM (b) CONDITION ACTION

5. RHR SYSTEN STEAM CONDENSING MDDE ISOLATI ,

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

b. RHR Area Temperature -

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

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

6. RHR SYSTEM SHUTOOWN COOLING N00E ISOLATION R

a. Reactor Vessel Water Level - Low, Level 3 6 2 1, 2, 3 25 g b. Reactor Vessel (RHR Cut-in Permissive) ,

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

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

d. RHR Area Temperature -

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

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

8. MANUAL INITIATION I

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

2. Outboard Valves 1,2,5,6,7 1/ group 1, 2, 3 26

3. Inboard Valves 4 (c) (e) 1/ group 1, 2, 3 and **,# 26

4. Outboard Valves 4 ICI k'I 1/ group 1, 2, 3 and **,# 26 g 5. Inboard Valves. 3,8,9 1/ valve 1, 2 - 3 26

% 6. Outboard Valves 3,8,9 1/ valve 1, 2, 3 26 F 7.- Outboard Valve 8(h) 1/ group 1, 2, 3 26

. , - - - -, -.,% v -- ~ < , , .. .

9 TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION ACTION 20 -

Be in at least NOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN with the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 21 -

Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within ACTION 22 - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SNUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Close the affected system isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected. system inoperable.

ACTION 23 -

ACTION 24 -

Se in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 25 -

Lock the affected system isolation valves closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

ACTION 26 -

Provided that the manual initiation function is OPERABLE for each other group valve, inboard or outboard, as tpplicable, in each line, restore tne manual initiation function to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; othentise, restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise:

a.

Be in at least H0T SHUTDOWN within the 'next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

b. Close the affected systaa isolation valves within the next hour and declare the affected systas in operable.

i befefe NOTES May be bypassed with reactor steam pressure 5,1043 psig and all turbine stop valves closed.

When handling irradiated fuel in the secondary containment and during CORE

Al.TERATIONS and operations with a potential for draining the reactor vessel.

During CORE ALTERATIONS and operations with a potential .for draining the reactor ves_sel.

(a) 2

. (b) Aee spectfication 3.6.3, Table 3.6.3-1 for _ valves _ in each valve group.)

thannel may be placed in an inoperable status for up to Z hours for required surveillance without placing the channel in the tripped condition provided at least one other OPERABLE channel in the same trip l )

system is monitoring that parameter. In addition for those trip systems with a design providing only one channel per trip system, the channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for required surveillance testing without placing the channel in the tripped condition provided that the redundant isolation valve, inboard or outboard, as applicable, in each line'is operable and all required actuation instrumentation for that redun-dant valve is OPERABLE, or place the trip system in the tripped condition.

(c) Also actuates the standby gas treatment systan.

(d) A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE.

(e) Also actuates secondary containment ventilation isolation dampers per

Table 3.6.5.2-1.

t (f) Closes only RWCU systae inlet outboard valve.

1 LA SALLE - UNIT 1 3/4 3-14 Amendment No. 26 i

_ ____.__._____--_l_- - - - - - - - - - - ~ --" - - - - - - - - - - ^ ^ -

TABLE 3.3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME RESPONSE TIME (Seconds / -l TRIP FUNCTION

6. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION N/A

a. Reactor Vessel Water Level - Low, level 3

b. Reactor. Vessel (RHR Cut-In Permissive) Pressure - High

c. RHR Pump Suction Flow - High

d. RHR Area Cooler Temperature High

e. RHR Equipment Area AT High B. MANUAL INITIATION N/A Inboard Valves 1.

2. Outboard Valves

3. Inboard Valves

4. Outboard Valves '

5. Inboard Valves

6. Outboard Valves !

7. Outboard Valve i

TABLE NOTATIONS

Isolation system instrumentation response time for MSIVs only. No diesel generator delays assumed.

- Radiation detectors are exempt from response time testing. Response' time shall be measured from detector output or the input of the first ectronic component in the channel. gg i Isolation system 'nstrumentation esponse time specif i for the Trip l Function actuatin Vacn71]ve aroup'shall be added to ' isolation time i

(slown in_ Table 3.6.3-1 and 3.6.5,Z-L for valves in each va've groug7to obtain ISDLATION SYSTEM RESPONSE TIME for each valve.

the MSlu i

l t

N/A Not Applicable.

LA SALLE - UNIT 1 3/4 3-19 AMEN 0 MENT NO. 98

3/4.6 CONTAINMENT SYSTEMS %

W VNVe5hafare 3 /4. 6.1 PRIMARY CONTAINMENT 1 *f*!) under ada;,; gf. '

PRIMARY CONTAINMENT INTEGRITY Cor;fr o/ 4S perm f /fs,4 LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2,* and 3.

ACTION: ,

' Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY within I hour 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 SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated:

a. After each closing of each penetration subject' to Type B testing, except the primary containment air locks, if opened following Type A or B test,-by leak rate testing the seal with gas at Pa, 39.6 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Surveillance i

Requirement 4.6.1.2.d for all other Type B .and C penetrations, the kombined leakage rate is less than or equal to 0.60 La.

At least once per 31 days by verifying that all primary containmen /

penetrations ** not capable of being closed by OPERABLE containment f Abbb automatic isolation valves and required to be closed during accident conditions are closed-by valves, blind flangesmor deactivated &

Mg g+ AutomaticLvalves secured in position, except Q$rovided in Table) ~

{

[3.6.3-1 of) Specification 3.6.3. T

s b.

c. By verifying each primary containment air lock OPERABLE per Specification 3.6.1.3.

d. By verifying the suppression chamber OPERABLE per Specification 3.6.2.1. l

e. Verify primary containment structural integrity in accordance with f

the Inservice Inspection Program for Post Tensioning Tendons. The C,i frequency shall be in accordance with the Inservice Inspection l s

N%. Program for Post Tensioning Tendons.

See Special Test Exception 3.10.1

- Except valves, blind flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often than once per 92 days.

LA SALLE - UNIT 1 3/4 6-1 Amendment No. 100

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS pace B-4 INSERT B 4.6.1.1.b. Perform required visual examinations and leakage rate testing except for primary containment air lock testing and main steam lines through the isolation valves, in accordance with and at the frequency' specified by 10 CFR 50, Appendix J, as modified by approved exemptions.

The overall integrated leakage rate acceptance criterion is s 1.0 L,. The Type B and C combined leakage rate acceptance criterion is s 0.60 L,.

However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, the leakage rate acceptance criteria are < 0.60 L, for the combined Type B and Type C tests, and < 0.75 L, for the Type A test.

INSERT C1 The provisions of Specification 4.0.2 are not applicable to the frequencies specified by 10 CFR 50, Appendix J.

i l

k;\nlailasalleiilrtfina.wpf46

Can s 6$

,TN7ENTION LL'[ LEp? g[g h0NTAINMENTSYSTEMS b M6h PRIMARY CONTAINMENT LEAKAGE

}

LIMITING CONDITION FOR OPERATION 3.6.1.2 Primary containment leakage rates shall be limited to:

a. An overall integrated leakage rate of less than or. equal to L,,

0.635 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P,,

39.6 psig,

b. A combined leakage rate of less than or equal to 0.60 L, for all penetrations and all valves listed in Table 3.6.3-1, except for main steam isolation valves and valves which are hydrostatically leak tested per Table 3.6.3-1, subject to Type E and C tests when pressurized to P,, 39.6 psig.

c. *Less than or equal to 100 scf per hour for all four main steam lines through the isolation valvss when tested at 25.0 psig.

d. A combined leakage rate of less than or equal to 1 gpa times the total number of ECCS and RCIC containment isolation valves in hydro-statically tested lines which penetrate the primary containment, I when tested at 1.10 P,, 43.6 psig. ,

APPLICABILITY: When PRIMARY CONTAINHENT INTEGRITY is required per Specification 3.6.1.1.

ACTION:

With:

a. The measured overall integrated primary containment leakage rate exceeding 0.75 L,, or

b. The measured combined leakage rate for all penetrations and all l valves listed in Table 3.6.3-1, except for main steam isolation !

valves and valves which are hydrostatically leak tested per Table 3.6.3-1, subject to Type 8 and C tests exceeding 0.60 L,, or I

c. The measured leakage rate exceeding 100 scf per hour for all four main steam lines through the isolation valves, or l

d. The measured combined leakage n :e for all ECCS and RCIC containment l isolation valves in hydrostatic 4 fly tested lines which penetrate the primary containment exceeding 1 gpa times the total number of such valves, 4

Exemption to Appendix "J" of 10 CFR 50. .

LA SALLE - UNIT 1 3/4 6-2 MP$eIS3h6-5 ndiment No. IB

hNTAINMENTSYSTEMS LETE PAGE' LI TING CONDITION FOR OPERATION (Continued)

~

ACTI k (Continued) resto -

a. The overall integrated leakage rate (s) to less than or equal to 0.75

,, and

b. T4 combined leakage rate for all penetrations and all valves listed in able 3.6.3-1, except for main steam isolation valves and valves whi are hydrostatically leak tested per Table 3.6.3-1, subject to Type and C tests to less than or equal to 0.60 L,, and

c. The le ge rate to less than or equal to 100 sef per hour for all I four sa steam lines through the isolation valves, and I

d. The combi ed leakage rate for all ECCS and RCIC containment isolation ,

valves in drostatically tested lines which penetrate the primary containment to less than or equal to 1 gpa times the total number of such valves, prior to increasing rea r coolant system tamperature above 200*F.

$URVEILLANCE REQUIREMENTS \

4.6.1.2 The primary contai nt leakage rates shall be demonstrated at the following test schedule and s 1 be determined in conformance with the criteria specified'in Appendix J of 10 C Part 50 using the methods and provisions of ANSI l

. N45.4-1972:

a. Three Type A Overall Inte rated Containment Leakage Rate tests shall ibe conducted at 40 1 10 no h intervals during shutdown at P,,

39.6 psig, during each 10-ye service period. The third test of each set shall be conducted d ing the shutdown for the 10 year.

, plant inservice inspection.

b. If any periodic Type A test fails to meet 0.75 L,, the test schedule l for subsequent Type A tests shall reviewed and approved by the Commission. If two consecutive Type tests fail to meet 0.75 L,, a l Type A test shall be iwrformed at less every 18 months until two consecutive Type A tests meet 0.75 L,, which time the above test l schedule may be restmed. .

c. The accuracy of each Type A test shall be rified by a supplemental test which:

1. Confirms the accuracy of the test by veri ng that the difference between the supplemental data a he Type A test data is within 0.25 L,.

2. Has duration sufficient to establish accurate 1 change in

1eakage rate between the Type A test and the sup esental test.

3. Requires the quantity of gas injected into the con inment or bled from the containment during the supplemental t to be equivalent to at least 25% of the total measured lea ge - -l at P,, 39.6 psig. ,

LA SALLE - UNIT 1 3/4 6-3 Amendment . 18

. ~- - . .. . - - -

e

~~

NTAINMENT* SYSTEMS -

.SUR ILLANCE REQUIREMENTS (Continued)

d. Type B and C tests shall be conducted with gas at' P,, 39.6 psig*, at ntervals no greater than 24 s'o nths except for tests involving:

1. Air locks,

2. Mai.n steam line isolation valves,

3. V ves pressurized with fluid from a seal system, and-

4. E and RCIC containment isolation valves in hydrostatically teste lines which penetrate the primary containment. >

e. Air locks sh 1 be tested and demonstrated OPERABLE per Surveillance -

Requirement 4. 1. 3.

f. Main steaa line olation valves shall be leak tasted at least once per 18 months.

g. Leakage from isolatio valves that are sealed with fluid from a seal system may be excluded, subject to the provisions of Appendix J,Section III.C.3, when d eraining the combined leakage rate provided the seal system and valve are pressurized to at least 1.10 P,, ,

43.6 psig, and the seal sy ;ea capacity is adequate to maintain systes pressure for at leas' 30 days.

h. ECCS and RCIC containment iso tion valves in hydrostatically tested lines which penetrate the pria containment shall be leak tested at least once per 18 months.

i. The provisions of Specification 4. 2 are not applicable to 24 month or 40 ! 10 month surveillance inte is.

l "Unless a nyaraulic test is required per Table 3.6.3-1. l

. 1 LA SALLE - UNIT 1 3/4 6-4 m..... ,-_ _ _ , , . _ _ , . . ~.

, ~ CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

c. By verifying at least two suppression chamber water level instru- .

mentation channels and at least 14 suppression pool water temperature instrumentation channels, 7 in esth of two divisions, OPERABLE by performance of a:

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

~

2. CHANNEL FUNCTIONAL TEST at least once.per 31 days, and

3. CHANNEL CALIBRATION at least once per 18 months.

The suppression chamber water level and suppression pool temperature alars setpoint shall be:

a) High water level i +2 inches * !

i b) Low water level > -3 inches * ,

d' c) High temperature i 105'F f*

Y By conducting drywell-to-suppression chamber bypass leak tests and 1 verifying that the A/4 calculated from the measured leakage is within the specified limit when drywell-to-suppression cha:aberi bypass leak tests are conducted:

1. At least once per 18 months at an initial differential pressure of 1.5 psi, and

2. At the first refueling outage and then on the schedule required for Type A Overall Integrated Containment Leakage Rate tests by Speci-

, fication 4.6.1.2.a; at an initial differential pressure of 5 psi, except that, if the first two 5 psi leak tests performed up to that time result in: ,

1. A calculated A/4 within the specified limit, and

2. The A/4 calculated from the leak tests at 1.5 psi is < 20% of the specified limit, ~ i l

then the leak tests at 5 psi may be discontinued.,

vw~ '

. Ak m^INSERTC2" L

Level is referenced to a plant elevation of 699 feet 11 inches (5'ee Figure B 3/4.6.2-1).

4 LA SALLE - UNIT 1 , 3/4 6-18 Amendment No. 67 l l

l .

~I

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS page B-5 INSERT C2

d. By conducting drywell-to-suppression chamber bypass leak tests at least once per 18 months at an initial differential pressure of 1.5 psi and verifying that the NVk calculated from the measured leakage is within the specified limit.

If any 1.5 psi leak test results in a calculated NVk > 20% of the specified limit, then the test schedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 1.5 psi leak tests result in a calculated NVk greater than the specified limit, then:

1. A 1.5 psi leak test shall be performed at least once per 9 months until two consecutive 1.5 psi leak tests result in the calculated NVk within the specified limits, and

2. A 5 psi leak test, performed with the second consecutive successful 1.5 psi leak test, results in a calculated NVk within the specified limit, after which the above schedule of once per 18 months for only 1.5 psi leak tests may be resumed.

If any required 5 psi leak test results in a calculated NVk greater than the specified limit, then the test schedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 5 psi leak tests result in a calculated NVk greater than the specified limit, then a 5 psi leak test shall be performed at least once per 9 months until two consecutive 5 psi leak tests result in a calculated NVk within the specified limit, after which the above schedule of once per 18 months for only 1.5 psi leak tests may be resumed.

k:inleslarallesi1rtfina.wpf47

e CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

[ If any 1.5 psi or 5 psi leak test results in:

1. A calculated V4 greater than the specified limit, or

2. A calculated V4 from a 1.5 psi leak test > 20% of the specified limit, then the test schedule for subsequent tests shall be reviewed by the Comeission. .

'If two consecutive 1.5 psi leak tests result in a calculated V4 greater than the specified limit, then:

1. A 1.5 psi leak test shall be 3:erformed at least once per 9 months until two consecutive 1.5 psi leak tests result in the calculated V8 within the specified limits, and

2. A 5 psi leak test, performed with the second consecutive successful 1.5 psi leak test, results in a calculated V 8 within the specified limit, after which the above schedule for only 1.5 psi leak tests may be resimied.

If two consecutive 5 psi leak tests result in a calculated V4 greater than the specified limit, then a 5 psi leak test shall be perfomed at least once per 9 months until two consecutive 5 psi leak tests res51t in {

a calculated A4k within the specified limit, after which the above t

schedule for only 1.5 psi leak tests may be resumed. -

J

~

l LA SALLE - UNIT 1 3/4 6-19 Amendment No. $

1

\

i CONTAINMENT SYSTEMS 3/4.6,3 PRIMARY CONTAINMENT ISOLATION VALVES PepIace with i

WSbkT 0" h ,

LIMITING CONDITION FOR OPERATION Y J

%ne li 3

excess flow check valves shown in Table 3.6.3-1 shall be OPE !

isolation times less than or equal to those shown in Table 3.6.3-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

h ACTION:

a. With one or more of the primary containment isolation valves sho in Table 3.6.3-1 inoperable: _

1. Maintain 3 least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either; a) Restore the inoperable valve (s) to OPERABLE status, nr b) Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolated position," or c) Isolate each affected penetration by use of at least one closed manual valve or blind flange.* l

2. 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one or more of the reactor i trumentation line excess flow check valves (shown in Table 3.6.3- inoperable:

't

1. Operation may continue and the provisions of Specification 3.0.3 are not applicable provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a) The inoperable valve is returned to OPERABLE status, or b) The instrument line is isolated and the associated instrument is declared inoperable.

2. 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

M$ INSERT E"

" Isolation valves closed to satisfy these requirements may be reopened on an l intermittent basis under administrative control.

3/4 6-22 Amendment No. 94 LA SALLE - UNIT 1

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS page B-G INSERT D 3.6.3 Each primary containment isolation valve and reactor instrumentation line excess flow check valve shall be OPERABLE **

APPLICABILITY: OPERATIONAL CONDITIONS 1,2, and 3.

ACTION:

a. With one or more of the primary containment isolation valves, except the reactor instrumentation line excess flow check valves, inoperable:

INSERT E

- Locked or sealed closed valves may be opened on an intermittent basis under administrative control.

I l

i I

k:%1a\lasalle\iirtiina.wyf48

l

. i CONTAINMENT SYSTDt5 ,

SURVEILLANCE REQUIREMENTS ~

P 4.6.3.1 Each primary containment isolation valve 6hown in Table 3.6.3-1)shall be demonstrated OPERABLE prior to returning the valve to service after mainte-nance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of. full travel and verifing the specified isolation time.

A4_6.3.2 Fach primary containment automatic isolation valve (shown in .

~

UsbTe 3.6.3-3 shall be demonstrated OPERABLE during COLD SHUTDOWN or REFUELING at least once per 18 months by verifying that on a containment isolation test signal amatic. isolation valve actuates to its isolation position.

4. 6. 3.7 n lation tBg j f ==eh'ori ry containment power operated or automatic " valve (shown in Table 3.6.3-37shall be determined to be within its - '

limit when tested pursuant to Specification 4.0.5.

4.6.3.4 ab: Each reactor instrumentation line excess flow check valve 6iiown in e 3.5.3-D shall be demonstrated OPERABLE at least once per 18 monT.ns ny vef fying that the valve checks flow.

m 4.6.3.M Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERA 8LE:

a. At least once per 31 days by verifying the continuity of the explosive charge.

b. At least once per 18 months by removing the explosive squib fros' at least one explosive valve such that the explosive squib in each explosive valve will be tested at least once per 90 months, and initiating the explosive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the ,

one fired or from another batch which has been certified by having at least one of that batch successfully fired. No explosive squib shall remain in use beyond the expiration of its shelf-life and

, operating-life.

0 //

k JSERTF A& "IUSERT G' .

P i

I LA SALLE - UNIT 1 3/4 6-23

~

l

ATTACHMENT B PROPOSED AMENDMENTS TO THE i LICENSE / TECHNICAL SPECIFICATIONS pace _ B-7 INSERT F 4.6.3.6 At least once per 18 months:

a. Verify leakage rate through all four main steam lines through the isolation valves is s 100 scfh when tested at 2 25.0 psig.*

b. Verify combined leakage rate of s 1 gpm times the total number of primary containment isolation valves through hydrostatically tested lines that penetrate the primary containment is not exceeded when these isolation valves are tested at 1.1 P,,2 43.6 psig.*

INSERT G (Footnote)

Results shall be excluded from the combined leakage for all penetrations and seals subject to Type B and C tests.

I I

l

inlaxlavalles21rtfina.vpf49 i

l

M .

TABLE 3.6.3-1 e-PRIMARY CONTAIPO4ENT ISOLATION VALVES E MAXINUM i

ISOLATION TIME-I "I g VALVE FUNCTION AND NUPRER VALVE GROUP (Seconds) !

a. ~ Automatic Isolatt'on Valves

1. Main Steam Isolation Valves 1 5*

. l 1821-F022A, 8 I C, D(DI 1821-F028A, 8, C, D b) 3H o

2. Main Steam Line Drain Valves 1 E 1821-F016 < 15 I' w i 1821-F019 I 15 1821-F067A, 8, C, D(b)

[77 7 23 g 3. Reactor Coolant System Sample Q

l Line Valves ICI -

3 <5 l p O 'q i 1833-F019 ~

'D h

m 1833-F020 y 4. Drywell Equipment Drain Valves 2 f S 1RE024 7 5 20 1RE025 1RE026 5 20 ly 0g s pq

< 15

$ - 5.

1RE029 Drywell Floor Drain Valves 2 315 dg }

W, IRF012 5 20 3- m

[

IRF013

6. Reactor Water Cleanup suction Valves 5 < 30

-ID 1G33-F001(d)

[ 7.

1G33-F004 RCIC Steam Line Valves 8 j 4=

W$l 1E51-F008I *I

< 20

{([X l

1E51-F063 i 15 1E51-F064 III - < 15 M,

- g

,E 1E51-F076 {15 g

= _ _

A

- _ _ _ _ _ _ ____.___-__________________--_______-_=n_ - - _ . ry w -- - or- --w ~ w , _

g TABLE 3.6.3-1 (Continued)

$ PRIMARY CONTAINMENT ISOLATION VALVES r-

"' MAXIMUN ISOLATION TIME VALVE GROUP I *I (Seconds)

VALVE FUNCTION AND NUMBER

,E

e. Automatic Isolation Valves (Continued)

8. Containment Vent and Pur0e Valves 4

10 IVQO26 IVQO27 7 10 IVQO29 7 10 IVQO30 7 10 IVQO31 7 10 IVQO32 75 IVQO34 7 10 R IVQO35 75 /

IVQD36 7 10 T IVQ040 7 10 -Q s DI IVQ042 {10 y\

< 10 \

h 1%

t IVQ043 IVQ047 {5

<5 IVQ048 IVQ050 75 IVQ051 75 IVQ068 75 { %

9. RCIC Turbin xhaust Vacuum Breaker 9 R.A.

(

Line y ves IE5J.fD80 M51-F086 10/LPCS,HPCS,RCIC,RHRInjection k/

o Testable Check Bypass Valves N.A. N.A.

R c.

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

( .

i l .

( -

TABLE 3.6.3-1 (Continued) .s . .

5

PRIMARY CONTAIMENT IEGLATION VALVES

!C r-MAXIMUM b IS0tATION TIME

' VALVE FUNCTION Me NUPGER VALVE GROUP I8 (Seconds) i E .

q Automatic Isolation Valves (Continued) '

11. Containment Monitoring Valves 2 -

15 ICM017A,8 1CM0184,8 1CM019A,8 1CM020A,8 I

1CM0218(hI ICM022A(h)

ICM025A h)

, ICM0268(h)

N .1CM027 u

e

, ICM028 V ,

4 m

1CM029 ICM030 .

ICM031 1CM032 p .

ICM033 l ICM034 y

12. Drywell Pneumatic Valves llN001A and 8 10 1 30 IIN017 10 1 22 ygi;l l.

11N074 10 1 22 11N075 10 1 22 11NO3 2 15

13. RHRd'hutdown Coollag Mode Valves 6

{s

/ 1E12-F008

/ IE12-F009 i 40 1 40

{a 1E12-F023 1E12-F053 A and 8 1 90 1 29 ,

x IE12-F099A and B IU}(O 1 30

.-~-

e, . .

g TABLE 3.6.3-1 (Continued)

PRIMARY CONTAll#ENT-ISOLATION VALVES -

g MAXIMUM

, VALVE FUNCTION Ale Nt9eER ' VALVE GROUPIp#

ISOLATION TlHE (Seconds)

E M Automatic Isolation Valves (Continued) e 14. Tip Guide Tube Ball Valves (Five Valves)

'7 N.A. l 1C51-J004

15. Reactor Building Clused Cooling Water

System Valves '

2 IWR029

< 30 1WR040 IWR179 IWR180 ' +

16. Primary Containment Chilled R

Water Inlet Valves D

i IVP113 A and B '

2

< 90 l %

U IVP063 A and B

17. Primary Containment Chille 7 to D Water Outlet Valves i hy '

IVP053 A and B IVP114 A and B 2

< 40 l

%C j i 90

18. Recirc. Hydr lic Flow Control -

Line Vai esI8) 2 18334338 A and 8 -<5

' 833-F339 A and 8 1833-F340 A and 8 1933-F341 A and 8 Ir:33-F342 A and B '

] s 1G3FF343 A and 8 1933-F344 A and 8 g 1833-F345 A and B

=

" 19. Feedwater Testable Check Valves 2 N.A.

iB21-F032 A and B ,

o Ot

i 1

l

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

4

e. .

TABLE 3.6.3-1 (Continued) g . PRIMARY CONTAINMENT ISOLATION VALVES -

k MAXIMUM i ISOLATI e VALVE FUNCTION AND NUMBER VALVE GROUP I *) (S nds) z -

b. Manual Isolation Valves l

1. IFC006 N.A.

2. IFC113 N.A.

i 3. 1FC114 -

N.A.

4. 1FC115 N.A.

5. I N.A.

1MCO27(II 1MC033 II

6. N.A.

7. I N.A.

8. ISA042(II ISA046 I) N.A.

l' R . TD h D

= s tn i

~

U1 t

a

.F

_ _ ._ _. _ ___ _ _ . _ _ . _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . #~ i- .c i - __v -e . _.+.-r- _ _ _ _ _ _

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

.l.

s) kl k. / , ,

TA8tE 3.6.3-1 (Continued) , ,

g PSIM48V CONIAllelENT IS01ATIGIl VALVES -

k VALVE FtelCI1010 AND BluteE8 I

c. Excess Flow Check Valves I8I

1 1. IB21-F374

" 2. 1821-F376

3. 1821-f 359

4. IB21-f355

5. 1821-f361

6. 1821-f378 .

7. IB21-F372 h 8. Ill21-f 370 T 3. lu21-F363

~ 10. 1821-F353 l 1 II. lu21-f4154, 8 IB21-F357

12. .

13. lH21-F382 ,.

14. .lH21-f328A, 5. C B ,

15. Ill21-F327A, 8. C. 0 11

16. lu21-F413A, 8 i l .,

17. lu21-f344

18. lu21-F365

19. IB21-f443

20. Ill21-F439 _

li

21. 1821-F437 -

l-

22. IH21-f441 -

4 I .

m. - - -. ~ ,_m _________.___._.._---____m._ m-m__

. . . ~ ,

f 1

() ',

. i ,

taste 3.6.3-1 (Continued)

E PA!MAAV CONTAllelENT 15044T1001 VALVES -

k ~

, VALVf flNICil0N ANS NUNSER E

4 ~Encess Flow Check YalvesI8I (Centinised) . -

g 23. IB21-f445A, S ,

24. 1821-F453

25. IR21-F447 -

26. 1821-F455A, b

27. 1821-F451 .

28. IS21-f449 'J -

4:* 29. 1821-f367 V '

30. 1021-f3264, 8, C, S . .

g 31.

3?.

Bil21-f 3254, S. C S til21-f 350

@D .

,u

l. ,

33. til21-F3'46 - '3'

34. 1821-f348 -

35. 1821-F471

36. til21-F4 3 , .

37. 1-F469 1821-F475A, B e-

[39. In21-F4654. B l, 40 ID21-F467 /

, 48 1821-f463 .'..

' ^

j/ 47. 1821-F380 f.

43. Iti33-f312A, S -

. I

44. iG33-F309 g

45. IE12-f315 '.

t

TABLE 3.6.3-l'(Continued) !

PRIMARY CONTAINMENT ISOLATION VALVES

?

i;;

, VALVE FUNCTION AND NUMBER g Excess Flow Check ValvesI9) (Continued)

Z 46. IE12-F359A, B

47. IE12-F319

48. IE12-F317 '

49. 1E12-F360A, B

50. 1E21-F304

51. 1E22-F304 [ I

52. 1E22-F341 w

g 53. 1E22-F342 D 7 54. 1833-F319A, 8 d 55. 1833-F317A, B l

56. IB33-F313A, B, C, D

57. 1933-F311A, B ,D N

58. IB33-F315A M , C, D '

59. 1933J-361A, B

60. 1833-F307A, B, C, D 61!1933-F305A,B,C,D g 62. ICM004

$' 63. ICM002

64. ICM012

[ 65. ICM010

66. IVQ061

67. IB21-F457

' 68. 1921-F459

.t l .

t TABLE 3.6.3-1 (Continued)

PRIMARY CONTAIPMENT ISOLATION VALVES s

l-VALVE FUNCTION AND NUMBER h

x Excess Flow Check ValvesI9) (Continued)

Z 69. 1821-F461 '

70. ICM102 ;

71. 1821-F570

72. 1821-F571

e. T1 I

n -

1

- m.*

. - l WP a

.o t

_ _ . - - _ _ _ _ _ _ _ _ _ _ _ m-- - _ _ _ ___ _ _ _ . _ _ _ _ _ . . . - - - , ,. --.-c ---- -, s,. . . ...,4, , , - . - , , . . .- , ._m _ _ .

? .

4

. v. .

g TABLE 3.6.3-1 (Continued) g -

PRIMARY CONTAINMENT ISOLATION VALVES E VALVE FUNCTION AND NUMBER g d. Other Isolation Valves

1. MSIV Leakage Control System 1E32-F001A, E, J. M(b)

2. Reactor Feedwater and RWCU System Return 1821-F010A, 8 1821-F065A, B ,

1G33-F040 1

3. Residual Heat Removal / Low Jr ssure Coolant Infection System N

$ 1E12-F042A, B, C /-

e J.

1E12-F016A, 1E12-F017A 1E12-F 8g ISI i

h -!

1E12 0 A,B)g .

1El-F024g))8gy)

E12-F021 1E12-F302 III '

5 IE12-F064A, B III IE12-F011A,8g)g gj g IE12-F088A, 8, C g )))

1E12-F025py)8,C 1E12-F030

1E12-F005(3I 1E12-F073A, 8 III k" IE12-F074A, B IJ}

1E12-F055A, B III k 1E12-F036A, B II)

P. 1E12-F311A, B Il} -

m 1E12-F041A, B

IE12-F050A,B(kh(k) ; '

M

A TABLE 3.6.3-1 (Continued) ^

PRINARY CONTAIMENT ISOLATION VALVES b VALVE FUNCTION AND NUM ER g Other Isolation Valves (Continued) p

4. Low Pressure Core Spray System 1E21-F005 1E21-F00l b -)

IE21-F012 b )

1E21-F011 b )

b 1E21-F031 )

IE21-F018(()

1E21-F006 I)

5. High Pressure Core Spray System .

R

1E22-F004 . b

i 'IE22-F015.f 1E22-F023 - I ti

) 4 1E22-F012-)

IE22-F014h) l 1E22-F005 y $

6. Reactor Core J olation Cooling System l

1E51-F013 /

1E51-F067 1E51-F028 IE5ttF068

(

s 1E[1-F040:)

E51-F031 -

)

g IE51-F019 a 1E51-F065

" 1E51-F066 F'

1E51-F059 g IE51-F022(")

IE51-F362 1E51-F363 I) 2 .

- .-. .. . . ~ , .

JABL .6.3-1 (Centinued)

, h PRIMARY CONTAINMENT ISOLATION VALVES l l-M l VALVE FUNCTION AND NUMBER Other Isolation Valves (Continued) l ?

! 7. Post LOCA Hydrocen Control

] '

l 1HG001A, B i lHG002A, B 1HG005A, 8 lHG006A, B

8. Standby Liould Control Sysing ,

IC41-F004A, B ~

j e y w IC41-F007

9. Reactor Recirculation Seal In.iection T IB33-F013A, 8 03 @

IB33-F017A, 8 08 / g M

10. Drywell Pneumatic System

(

llN018 ,

11. Reference Lea Backfill ICll-F422B ICll-F422D ICll-F422F ICll-F422G ICll-F4238 ICll-F4230

g. ICll-F423F g ICll-F423G ,

Y a

E.

DEL 5TE TABLE 3.6.3-1 (Continued)

PRIMARY CONTAINMENT ISOLATION VALVES N

But 2 3 seconds.

(a) See Spec:fication 3.3.2, Table 3.3.2-1, for isolation signal (s) that erates each valve group.

(b) No included in total sum of Type B and C tests.

(c) May opened on an intermittent basis under administrative control.

(d) Not c1 ed by SLCS actuation.

(e) Not clos by Trip Functions Sa, b or c, Specification 3.3.2, Table 3.3.2-1.

(f) Not closed y Trip Functions 4a, c, d, e or f of Specification 3.3.2, Table 3.3.2-1.

(g) Not subject t Type C leakage test.

(h) Opens on an iso tion signal. Valves will be open during Type A test. No Type C test requ ed.

(i) Also closed by d 11 pressure-high signal.

(j) Hydraulic leak test t 43.6 psig.

(k) Not subject to Type C akage test - leakage rate tested per Specifica-tion 4.4.3.2.2.

(1) These penetrations are pr vided with removable spools outboard of the outboard isolation valve. uring operation, these lines will be blind flanged using a double 0-ri and a type B leak test. In addition, the packing of these isolation v ves will be soap-bubble tested to ensure insignificant or no leakage at the containment test pressure each refueling outage.

(m) If valves 1E51-F362 and 1E51-F363 re locked closed and acceptably leak rate tested, then valves 1E51-F059 nd 1E51-F022 are not considered to be primary containment isolation valves nd are not required to be leak rate tested.

(n) Either the 1E51-F362 or the 1E51-F363 va ve may be open when the RCIC system is in the standby mode of operatio and both valves may be open during operation of the RCIC system in the ull flow test mode, providing that:

1) valve IE51-F022 is acceptably leak rate te +ed, and

2) valve IE51-F059 is deactivated, locked closed nd acceptably leak rate tested, and

3) the spectacle flange, installed immediately downst eam of the IE51-F059 valve, is closed and acceptably leak rate ested. ,

1 1

1 LASALLE - UNIT 1 3/4 6-34a Amendment No. 81

~

CONTA!NMENT SYSTEMS h%ation Oni 3/4.6.4 VACUUN RELIEF .

LIMITING CONDITION FOR OPERATION 3.6.4 All suppression chamber - drywell vacuum breakers shall be OPERABLE and closed.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

l ACTION:

a.- With one suppression chamber - drywell vacuum breaker inoperable and/or open, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close the manual isolation valves on both sides of the inoperable and/or open vacuum breaker. Restore the inoperable and/or open vacuum breaker to OPERABLE and closed status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HDT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one position indicator of any OPERABLE suppression chamber -

drywell vacum breaker inoperable, restore the inoperable position indicator to OPERABLE status within 14 days or visually verify the vacuum breaker to be closed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise, declare the vacuum breaker inoperable.

SURVEILLANCE REQUIREMENTS 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be:

a. Verified closed at least once per 7 days.

b. Demonstrated OPERABLE:

1. At least once per 31 days and within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after any discharge of steam to the suppression chamber from the safety-relief valves, by cycling each vacuum breaker through at least one complete cycle of full travel. -

2. At least once per 31 days by verifying both position indicators OPERABLE by performance of a CHANNEL. FUNCTIONAL TEST.

3. At least once per 18 months by; a) Verifying the force required to open the vacuum breaker, from the closed position, to be less than or equal to 0.5 psid, and b) Verifying both position indicators OPERABLE by performance of a CHANNEL CALIBRATION.

LA SALLE - UNIT 1 3/4 6-35 Amendment No. 38

1 i .

CONTAINMENT SYSTEMS

~

SUR'.'EILLANCE REQUIREMENTS (Continued) ,

l 4.6.4.2 The manual isolation valves on both sides of an ino,wrable and/or open suppression chamber-drywell vacuum breaker shall be verJfied to be closed at least once per 7 days.

4.6.4.3 Vacuum breaker header flanges which have been broken shall be leak sted after re-making by Type 8 test at 39.6 psig per Specification 4.6.1.2.d.

h LA SALLE - UNIT 1 3/4 6-36 Amendment No. 18

CONTAINMENT 5Varpi5 3/4.6.5 ' PRIMARY CONTAINMENT ATMDSPNERF CONTROL !

' DRYWELL AND SUPPRESSION CHAMBER HYDROGEN RECOMBINER SYSTEMS LIMITING CONDITION FOR OPERATION

3.6.6.1 Two independent drywell and suppression chamber hdrogsn recombiner systems shall be OPERABLE.

I-APPLICABILITY: OPERATIONAL CONDITIONS I and 2. .

ACTION:

With one drywell and/or suppression chamber hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 d4ys or be in at least HDT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

$URVEILLANCE REQUIREMENTS 4.6.6.1 Each 'drywell and suppression chamber hydrogen recombiner system shall be demonstrated OPERABLE:

a. At least once per 92 days by cycling each flow control valve and recirculation valve through at least one complete cycle of full

travel. ,

b. At least once per nths by verifying, during a recombiner system functional test:- ,. ,

1. That'td heaters are OPERABLE by determining that the current in each phase differs by less than or equal to SE from the other phases and is within 5% of the value observed in the .

original acceptance test, corrected for line volthge differences. ;

2. That the maction chamber gas temperature increases to 1200 2 25'F i within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

c.

At least once per 18 months by: ' "* *

1. Perfoming a CHANNEL CALIBRATION of all recombiner operating instrumentation and control circuits.

2. Verifying the integrity of all heater electrical circuits by performing a resistance to ground test within 30 minutes following the above required functional test. The resistance to ground for any heater phase shall be greater than or equal to 100,000 ohes.

, y measuring the leakage rate:

1. As a part of the overall integrated leakage rate test require by Specification 3.6.1.2, or

2. By measuring the leakage rate of the system outside of the containment isolation valves at P,. 39.6 psig, on the schedule required by Specification 4.6.1.2 and including the measured leaLage as a part of the leakage determined in accordance wit Specification 4.6.1.2. i

~

LA SALLE - UNIT 1 3/4 6- j N Amendment / I

. . . . . . . . - ~ . . . - . . . . . . . - - - - ~ ~ ~ ~ ~ ~ nr "' " * *

i

(

'4 m

3/4.6 CONTAINMENT SYSTEMS BASES

.3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY '

PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the

. site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

The structural integrity of the primary containment is ensured by the successful completion of the Inservice Inspection Program for Post Tensioning Tendons and by associated visual inspections of the steel liner and penetrations for evidence of deterioration or breach of integrity. This ensures that the structural integrity of the primary containment will be maintained in accordance with the provisions of the Primary Containment Tendon surveillance Program. Testing and Frequency are consistent with the recommendations of Regulatory Guide 1.35, Revision 3, except that the Unit I and 2 primary containments shall be treated as twin containments even though the Initial Structural Integrity Tests were not within 2 years of each other.

9 3/4.6.1.2 PRIMARY CONTAINMENT LEAKAGE.

The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the ,

accident analyses at the peak accident pressure of 39.6 psig, P . As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic tests to '

account for possible degradation of the containment leakage barriers between leakage tests.

Operating experience with the main steam line isolation valves has -

indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.

The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J to 10 CFR 50 with the exception of exemption (s) granted for main steam isolation valve leak testing and testing the airlocks after each opening.

3/4.6.1.3 PRIMARY CONTAINMENT AIR LOCKS h The limitation on closure and leak rate for the primary containment air (

locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY o '

and the primary containment leakage rate given in Specificatiord 3.6.1.1 and> '

f3.6.1.h The specification makes allowances for the fact that tTere may be Elong periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock is re ire to aintain the_ integrity of the containment.

3h 6,L2 DE~LET$Dh 3l O U LAIALLE B 3/4 6-1 Amendment No. 100

- ATTACHMENT B PROPOSED AMENDMENTS TO THE e LICENSE / TECHNICAL SPECIFICATIONS page Bdl >

INSERT H PRIMARY CONTAINMENT INTEGRITY is maintained by limiting overall integrated leakage to s 1.0 L, and the Type B and C combined leakage rate acceptance criterion is s 0.60 L,. Prior to the first startup after performing a required 10 CFR 50, '

Appendix J, leakage test, the combined Type B and C leakage must be < 0.60 L,, and .

the overall Type A leakage must be < 0.75 L, when a Type A test is scheduled.' !

Compliance with this LCO will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses, q

The maximum allowable leakage rate for the primary containment (L,) is 0.635%

by weight of the containment atmosphere per day at the calculated maximum peak ,

containment pressure (P,) of 39.6 psig. .

i Individual leakage rates specified for the primary containment air lock, main steam lines through the isolation valves, and valves in hydrostatically tested lines are j addressed in LCO 3.6.1.3, and Surveillance Requirement 4.6.3.6. t Surveillance Requirement 4.6.1.1.b maintains PRIMARY CONTAINMENT INTEGRITY by requiring compliance with the visual examinations and leakage rate test requirements of 10 CFR 50, Appendix J, as modified by approved exemptions.

Failure to meet air lock leakage testing (4.6.1.3) or main steam isolation valve leakage (4.6.3.6.a) does not necessarily result in a failure of this Surveillance Requirement, 4.6.1.1.b. The impact of the failure to meet these Surveillance Requirements 4.6.1.3 ;

and 4.6.1.1.b must be evaluated against the Type A, B, and C acceptance criteria of 1 10 CFR 50, Appendix J, as modified by approved exemptions. The leakage limits for main steam lines through the isolation valves and leakage test results of Seveillance Requirement 4.6.3.6.a are not included in the total sum of Type B and C tests (approved exemption). As-left leakage prior to the first startup after performing a ,

required 10 CFR 50, Appendix J, leakage test is required to be < 0.60 L, for combined Type B and C leakage, and < 0.75 L, for overall Type A leakage. At all other times between required Type A tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At s 1.0 L, the offsite dose consequences are bounded by the assumptions of the safety analysis. The combined Type B and C leakage remains as s 0.60 L, between scheduled tests, in accordance with Appendix J.

The Frequency is required by 10 CFR 50, Appendix J, as modified by approved exemptions. Thus,4.0.2 (which allows Frequency extensions) does not apply to Surveillance Requirement 4.6.1.1.b.

k r \nla\1asalle\ ilrt ! sna.wpf 50

t CONTAINMDff SYSTBt$

LARS_

JEMt$33URIZATf0N SYSTEMS (Cantinued)

Because of the'1erge value and theresi capacity of the suppression pool.

the volume and temperature normally changes very slowly and monitoring these permeters daily is sufficient to establish any temperature trends. By requid ng the suppression pool temperature to be frequently recorded during peMods of significant heat addition, the tamperature trends will be closely followed so that apprepM ata action con be taken. The requirement for an external visual examina-tion following any event where potentially high loadings could occur prwides assurance that ne significant damage was encountered.

In addition to the limits on taperature of the sgpression chamber pool wter, operating precedures define the action ta be taken in the event a safety-relief valve inadvertantly opens or sticks open. As a sinimum this action shall includar.(1) use of all available asent to close the valve, (2) initista suppree-sion pool water coeling, (3) initista reactor shutdown, and (4) if other safety-relief valves are used to depressuMze the reactor, their discharge shalf be ~

separated from that of the stuck-open safety relief valve to assure sizing and wifonrity of energy insertion'ta the peal.

4 3/4.f.3 PRDWlY CONTADeetT ISOLATION VALVET p The OpstASILITY of the pMeary contairment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive estarial to the containment atmosphere or pressvHzation of the containment. Cantainment isolation within the time limits specified ensures that the release of radioactive meterial to the environment will be consistant with the assuptions used in the analyses for a LOCA.

M INSERT m...a vacuuN - m~" .

Vacuer relief breakers are provided to equalize the pressure between the '

suppression chenber and dryme11. This systan will enintain the structural integ-rity of the primary containment under conditions of large differential pressures.

The vacuum breakers between the suppression chamber and the drywell eust not be inoperable in the open position since this would allow bypassing of the .

suppression pool in case of an accident. There are four valves to provide redundancy so that operation may continue for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with one vacum breaker inoperable prwided that the manual isolation valves on each side are in the closed position.

. l l

9 1

l LA SALLE - UNIT 2 8 3/4 6-4 l

l

ATTAC.HMENT B PROPOSED AMENDMENTS TO THE LICENSEfI'ECHNICAL SPECIFICATIONS page B-Q INSERT J Primary Containment Isolation Valves (PCIVs) form a part of the primary containment boundary. The PCIV safety function is related to control of primary containment leakage rates during accidents or other conditions to limit the untreated release of radioactive materials from the containment in excess of the design limits.

The automatic isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal. The valves covered by this specification are listed with their associated stroke times, and other design information for lines penetrating the Primary Containment, in UFSAR Section 6.2.

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact.

Main steam lines through the isolation valves and hydrostatically tested valves must meet alternative leakage rate requirements. Other PCIV leakage rates are addressed by specification 3/4.6.1.1, " PRIMARY CONTAINMENT INTEGRITY". UFSAR Section 6.2 also describes special leakage test requirements and exemptions.

This specification provides assurance that the PCIVs will perform their designed safety functions to control leakage from the primary containment during accidents.

The opening oflocked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the primary containment.

k:inlas1atalletiltttina.wyt51

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE /I'ECHNICAL SPECIFICATIONS page B-10 l

INSERT J (Continued)

Surveillance Requirement 4.6.3.6.a verifies leakage through all four main steam

' lines is s 100 scfh when tested at 2 P (25.0 psig). The transient and accident !

analyses are based on leakage at the specified leakage rate. The leakage rate for l main steam lines through the isolation valves must be verified to be in accordance

. with the leakage test requirements of 10 CFR 50, Appendix J, as modified by approved exemptions. A Note has been added to this Surveillance Requirement requiring the results to be excluded from the total of Type B and Type C tests.

This ensures that leakage rate for main steam lines through the isolation valves is properly accounted for in accordance with an approved exemption. The frequency is "at least once per 18 months" in accordance with an approved exemption.

Surveillance Requirement 4.6.3.6.b test of hydrostatically tested lines provides assurance that the assumptions of UFSAR Section 6.2 are met. The combined leakage rates must be demonstrated in accordance with the leakage rate test at a frequency of"at least once per 18 months". A Note has been added to this Surveillance Requirement requiring the results to be excluded the total of Type B and Type C tests. This is in accordance with 10 CFR 50, Appendix J, and ,

approved exemptions. -I k \nla\lasalle ulrt fina.wpf 52

'16 - 6/23/88 e

Lic3:33 Co. NPF-11 (d) The maximum average planar linear heat generation (MAPLEGR) limit will be reduced by 0.85.

(e) Technical Specification Setpoints shall read as follows:

T.S.2.2.1 S 0.66W + 45.7 (Trip Setpoint) '

S 0.66W + 48.7 (Allowable)

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

Sgg (0.66W + 36.7) T*

To as defined in T.S.3.2.2 T.S.3.3.6 APEN Upscale 0.66W i 36.7 (Trip Setpoint) -

APRM Upscale 0.66W + 39.7 (Allowable)

REN Upscale 0.66W + 34.7 (Trip Setpoint)

REN Upscale- 0.66W + 37.7 (Allowable)

(f) The average power range monitor (APEN) fluz moise will be measured once per shift; and the recirculation loop flow 1

1 will be reduced if the.fluz moise averaged over 1/2 hour 4 01 exceeds 5 percent peak to peak, as measured by the APEN Ie # chart recorder.

/g ] g) The core plate delta P noise will be measured once per shift, and the recirculation loop flow will be reduced if the noise exceeds one (1) pai ak-to-peak.

f Appendices G, E and Am. 12 D. Esemptions from certain requiremen 12/20/82 J and 10 CFR Part 73 are describe a the Safety Evaluation !

Report and Supplement No.1, No. No. 3 to the Safety Evaluation Report. g n addition) esemption was requested "until the completion of the first refueli_ng frpm the require-monts of 10 CTR 70.24 andh a esemption frost 10 CFR Part w ,'

Appendia E from performing a full scale exercise within one year before issuance of an operating license, both esemptions are described in upplement No. 2 of the Safety Evaluation i

' Repo Inally esemption was requested from the h requ rements of 10 CFR 50.44 until either the required 100 percent rated thermal power trip startup test has been ample _ted or_the reactor has operated for 120 effective full

~

r if as specified by the Technical Specifications.

at sempt(ion-isdescribedinthesafetyevaluationof License Amenchment No.12m These esemptions are authorised by law and will not endanger life or property or the common defense and security and are otherwise in the public interest.

Therefore, these esemptions are hereby granted. The facility will operate, to the extent authorised herein, in conformity with the application, as amended, and the rules and regulations of the Comunission (eacept as hereinaf ter enemyted therefrom),

and the provisions of the Act.

er ain 1 Y'*%"trements of 10 CFA Pa 50pocrg Mg 10cFR kt73. These include: (a)

l l

ATTACHMENT B l PROPOSED AMENDMENTS TO THE LICENSEfI'ECHNICAL SPECIFICATIONS i page B-11 INSEItT K (Unit 1, NPF-11)

(c) An exemption from the 1 ;quirement of paragraph III.D of Appendix J to conduct the third Type A test of each ten-year service period when the plant is shutdown for the 10-year plant inservice inspections. Exemption (e) is described in the safety evaluation accompanying Amendment No. to this license.

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INDEX

DEFINITIONS -

SECTION 1.0 DEFINITIONS PAGE -

1.1 ACTI0N............................................... ............ 1-1 1.2 AVERAGE PLANAR EXP05URE........................................... 1-1

1. 3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE........................ 1-1 1.4 CHANNEL CALIBRATION................................................ 1-1 1.5 CHANNEL CHECK.....................................................

1-1 1.6 CHANNEL FUNCTIONAL TEST.............................. ............ 1-1 1.7 CORE ALTERATION................................................... 2 1.8 CORE OPERATING LIMITS REP 0RT...................................... 1-2 1.9 CRITICAL POWER RATI0.............................................. 1-2 1.10 DOSE EQUIVALENT I-131........................................ .... 1-2 1.11 I-AVERAGE DISINTEGRATION ENERGY................................... 1-2 1.12 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSETIME................ 1-2 1.13 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME......... 1-2 1.14 FRACTION OF LIMITING POWER DENSITY................................ 1-3 1.15 FRACTION OF RATED THERMAL P0WER................................... 1-3 1.16 FREQUENCY NOTATION................................................. 1-3 1.17 GASE005'RADWASTE TREATMENT SYSTEM................................. 1-3 1.18 IDENTIFIED LEAKAGE................................................ 1-3 r-f I 1.19 ISOLATION SYSTEM RESPONSE TIME.................................... 1-3 1.2hLIMITINGCONTROLRODPATTERN......................................-

1. 1-3 1.2/LINEARHEATGENERATIONRATE....................................... 1-4 1.2/*LOGICSYSTEMFUNCTIONALTEST......................................

4 1-4 1.2/MAXIMUMFRACTIONOFLIMITINGPOWERDENSITY........................ 1-4 5

1.2/ MEMBER (s)0FTHEPUBLIC............................................ .

_._ 1-4 . . _ _

1.2/b1NIMUMCRITICALPOWERRATI0......................................

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

1-4 LA SALLE - UNIT 2 gj _G araorer osaor, l-

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DEFINITIONS SECTION DEFINITIONS (Continued) .

PAGE.

1.2 OPERABLE - OPERABILITY............................. ~ .-........... 1-4 1.2)9 OPERATIONAL CONDITION - C0NDITION. . . . . 1-5 ..........

30 1.29. PHYSICS TESTS................................... ~............... 1-5 ,,

1 4ff PRESSURE BOUNDARY LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .1-5 ...

1 .

1.3/PRIMARYCONTAINMENTINTEGRITY..........................,........... 1-5 9

1.3/PROCESSCONTROLPR0 GRAM........................................... 1-5 1.3 URGE - PURGING................................................... 1-5 5

1.3/nATEDTHERMALP0WER............................................... 1-6 1.3 REACTOR PROTECTION SYSTEM RESPONSE 7

TIME........................... 1-6 1.3fREPORTABLEEVENT.................................................. 1-6 -

1.3/gROD DENSITY....................................................... 1-6 1.3p'SECONDARYCONTAINMENTINTEGRITY.................................. . 1-6 1.YSHUTDOWNMARGIN...................................................

1 1-6 1.4/ SITE 2

800NDARY..................................................... 1-7 1.4/souRCECNECK...................................................... 2-7 1.4/'STAGGEREDTESTsASIS.............................................. 1-7 2.4/%uERM4tP0wER.....................................................

5 1-7 1.4/TURBINEBYPASSRESPONSETIME....................................... 1-7 b

1.hUNIDENTIFIEDLEAKAGE..............................................

7 1-7

1. 4/ VENTI LATION EXHAUST TREATMENT SYSTEM. . . . . . . . .1-7 . . . . . . . . . . . .i 1.4/sVENTING........................................................... 1-7 l

LA SALLE - UNIT 2 II Amendment No. 69 _

e l

s. ,

E gglIINGCONDITIONSFOROPERATiONANDSURVEILLANCEREQUIREMENTS SECTION E8fai 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity............................ 3/4 6-1 PrimaryContainmentLeakage.............................. 3/46p ,

Primary Containment Air Locks............................ 3/4 6-5 MSIV Leakage Control System.............................. 3'/4 6-7 1

Drywell and Suppression Chamber Internal Pressure........ 3/4 6-16 Drywell Average Air Temperature.......................... 3/4 6-17 ,

Drywell and Suppression Chamber Purge System. . . . . . . . . . . . 3/4 6-18 3/4.6.2 DEPRESSURIZATION SYSTEMS .

Suppression Chamber......................'................ 3/4 6-19 Suppression Pool Spray................................... 3/4 6-23 Suppression Pool Cooling................................. 3/4 6-24 ;

3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES.....................- 3/4 6-25 3/4.6.4 VACUUM RELIEF............................................ 3/4 6-38 3/4.6.5 SECONDARY CONTAINMENT ;

Secondary Containment Integrity. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-40 Secondary Containment Automatic Isolation Dampers........ 3/4 6-41 Standby Gas Treatment System............................. 3/4 6-43 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Drywell and Suppression Chamber Hydrogen Recombiner l Systems................................................ 3/4 6-46 l l

Drywell and Suppression Chamber Oxygen Concentration..... 3/4 6-47 i l

LA SALLE - UNIT 2 VII Amendment No. 84 l

4 INDEX LIST OF TABLES (Continued)

A P_ASE TABLE REMOTE SHUTDOWN MONITORING INSTRUMENTATION ........ 3/4 3-67 3.3.7.4-1 4.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION 3/4 3-68 SURVEILLANCE REQUIREMENTS .........................

ACCIDENT MONITORING INSTRUMENTATION ............... 3/4 3-70 3.3.7.5-1 I 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION 3/4 3-71 SURVEILLANCE REQUIREMENTS .........................

FIRE DETECTION INSTRUMENTFTION .................... 3/4 3-76 3.3.7.9-1 3,3.7.11-1 EXPLOSIVE GAS MONITORING 3/4 3-83 I N ST RUMENT AT I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4.3.7,11-1 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ......... 3/4 3-84 3.3.8-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM 3/4 3-87 ACTUATION INSTRUMENTATION .........................

3.3.8-2 FEEDWATER/ MAIN TURBINE TRIP SYSTEM 3/4 3-88 ACTUATION INSTRUMENTATION SETPOINTS ...............

4.3,8.1-1 FEEDWATER/ MAIN TURBINE TRIP SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQtlIREMENTS ......... 3/4 3-89 ;

3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES3/4 .. 4-10 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS ........... 3/4 4-13 l 3.4.4-1

+

4.4.S-1 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM .................................. 3/4 4-16 4.4.6.1.3-1 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM 3/4 4-20 WITHDRAWAL SCHEDULE ............................... _

PRIMARY CONTAINMENT ISOLATION VALVES .............. 3/4 6

[6.3-1 XXII Amendment No. 84 LA SALLE - UNIT 2

. G.. .=M i .. v,. . ~. :. . . l -

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.m. ; . ,

s DEFIN m 0NS ,

END-OF9 CYCLE RECIRCULATION PUMP TRIP SYSTEM FiSPONSE ( TIME ' Continued)

FRACTION OF LIMITING POWER DENSITY 1.14 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LNGk existing l at a given location divided by the specified LNGR limit for that bundle type.

FRACTION OF RATED THEMAL POWER -

L15 The FRACTION OF RATED THERMAL POWER (FITP) shall be the seasured THERMAL l POWER divided by the RATED TERMAL POWER.

FREDUENCY NOTATION L15 The FREQUENCY WTATION specified for the performance of Surveillance l',

Requirements shall correspond to the intervals defined in Table LL

'GASE0US RADWASTE TREATMENT SYSTEM ,

L17 A GA5EDUS RADWASTE TREATENT SYSTEM shall be any system designed and- l installed to reduce radioactive gaseous affluents by collecting primary .

coolant system offgases from the primary system and providing for delay' or holdup for the purpose of reducing, the total radioactivity prior ta ,

release to the environment. '

IDENTIFIED LEAKAGE .

~

1.18 IDENTIFIED LEARAGE'shall be:

a.

Laakaprinta. collection systems, such as pue seal or valve

'M packing leaks,' that is captured and conducted to a susp or collecting tank, er

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systans or not to be PRESSURE 300 E ARY LEAKAGE. .

ISOLATION SYSTEM RESPONSE TIE I.19 The ISOLATION SYSTEM RESPONSE T7E shall be that time interval from when i- the l monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. Times shall incivde diesel generator starting and seguanca leading delays where applicabler The response time may be asasured by any series of sequential, overlapping,er tstal staps such that the entire response time is sensured.

MITINC CONTEDL ROD PATTEM -

A LIMITING CONTEDL RtB PATTERN shall be a pattern tAich results'in the ~

1.?g/ core being on value for APLNGR, UER, or EPL a thermal heraulic limit, i.e., aperating l on a limit d //

AM INSERTA _ _ _ . _. . .

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

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

page B 3 INSERT A L.

1.20 The maximum allowable primary containment leakage rate, L , shall be 0.635 % of primary containment air weight per day at the calculated peak containment pressure (P, = 39.6 psig).

k : e la\lasalloulct fina.wpf 45 l

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e DEFINITIONS LINEAR HEAT GENERATION RATE 1.fft.INEARHEATGENERATIONRATE(LHGR)shallbetheheatgenerationperunit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LDGIC SYSTEM FUNCTIONAL TEST 1.2YALOGICSYSTEMFUNCTIONALTESTshallbeatestofalllogiccomponents, ,

i.e., all relays and contacts, all trip units, solid state logic elements, ;

etc. of a logic circuit from sensor through and including the actuated device to verify DPERABILITY. THE LOGIC SYSTEM FUNCTIONA. TEST may be

.perfomed by any series of sequential, overlapping or total system steps such that the entire logic systas is tested. ..

MAXIMUM FRACTION OF LIMITING POWER DENSITY 1.2%heMAXIMUMFRACTIONOFLIMITINGPOWERDENSITY(MFLPD)shallbethehighest value of the FLPD which exists in the core.

MEMBER (S) 0F THE PUBLIC 1.2[ MEMBER (S)DFTHEPUBLICshallincludeallpersonswhoarenotoccupation-ally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This cate site for recreational, gory does include occupational, persons or other purposes who use notportions of the associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.2hheMINIMUMCRITICALPOWERRATIO(MCPR)shallbethesmallestCPRwhich exists in the core. [

OFFSITE DOSE CALCULATION MANUAL '

1.2pTheOFFSITEDOSECALCULATIONMANUAL(ODCM)shallcontainthemethodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alam/ Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.2.F.4 and (2) descriptions of the infomation that should be included in the Annual Radiological Environmental Operating and -

Semi-Annual Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4. >

OPERABLE - OPERABILITY '

1.2)Asystem, subsystem, train,componentordeviceshallbeOPERABLEorhave I !

$ OPERABILITY when it is capable of perfoming.its specified function (s),

and when all necessary attendant instrumentation controls, a normal and an emergency electrical power source, cooling or s,eal water, lubrication .

or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).

'l LA SALLE - UNIT 2 1-4 Amendment No. 69 i

pEFINITIONS OPERATIONAL CONDITION - CONDITION 1.2/ An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS 90 1.M PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE B0UNDARY LEAKAGE I

1.3/ PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY L

1.?/ PRIMARY CONTAINMENT INTEGRITY shall exist when:

a. All primary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE primary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except 3.6.3)fas~ provided in' Table g 3.6.3-1 opSpecifi All primary containment equipment hatches are closed and sealed.

Each primary containment air lock is PERABL ursuant

c. '

\

Specification 3.6.1.3. -

MQIhYattled cef' }

I i hin the limits (of d.hTheprimarycontainmentle aer s

$pecification 3.6.1 8 Surves/ nce gqu[pemgf;f gjg

e. The suppression chamber is CPERABLE pursuant to Specification 3.6.2.1.

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

g. Primary containment structural integrity has been verified in accordance with Surveillance Requirement 4.6.1.1.e.

for va ves .+lna are open anae.n 40ministrative con %lasPermitted by 1-5 Amendment No. 84 LA SALLE - UNIT 2 !

l

a. DEFINITIONS

j. .

I PROCESS CONTROL PROGRAM he PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, 1.3/ sampling, analyses, test,anddeterminationstobemadetoensurethat t

- processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in l

t such a way as to assure compliance with 10 CFR 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing .

the disposal of solid radioactive waste.

PURGE - PURGING

. 1. FURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replace-ment air or gas is required to purify the confinement.

)

1 i

i i

LA SALLE - UNIT 2 1-5 a Amendment No. 84

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DEFINITIONS RATED THERMAL POWER 1.3/RATEDTHERMALPOWERshallbeatotalreactorcoreheattransferrateto l

.fthe reactor coolant of 3323 WT.

REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from ! l 1.3/6 when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

REPORTABLE EVENT

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

ROD DENSITY 1.3iRODDENSITYshallbethenumberofcontrolrodnotchesinsertedasa

$ fraction of the total number of control rod notches. All rods fully inserted is equivalent to 100% ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY 1.3) SECONDARY CONTAINMENT INTEGRITY shall exist when:

7 a. All secondary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or '

2. Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

b. All secondary contsinment hatches and blowout panels are closed and sealed.

c. The standby gas treatment system is OPERABLE pursuant to Specification 3.6.5.3.

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

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

l f. The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.5.1.a.

SHUTDOWN MARGIN

1. ASHUTDOWN MARGIN shall be the nount of reactivity by which the reactor is l l l$ suberitical or would be subtritical assuming all control rods are fully l

inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68*F; and xenon free.

l l

LA SALLE - UNIT 2 1-6 Amendment No. 69 l

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. ~ o ..:. : : ~ - 4 DEFINITIONS ,

l SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither 1.4/j owned, nor leased, nor otherwise controlled by the licensee.

SOURCE CNECK 1.4[Awhenthechannelsensorisexposedtoaradioactivesource.A SOURCEl CHECK l I

STAGGERED TEST BASIS 1.4 A STAGGERED TEST BASIS shall consist of: ,l

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

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

THERMAL POWER 1.4 THERMAL POWER shall be the total reactor core heat transfer rate to the l reactor coolant.

TURBINE BYPASS SYSTEM RESPONSE TIME 1.4/TheTURBINEBYPASSSYSTEMRESPONSETIMEshallbetimeintervalfromwhen 5 the turbine bypass control unit generates a turbine bypass valve flow -

l' signal until the turbine bypass valves travel to their required positions.

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

UNIDENTIFIED LEAKAGE 1.4/UNIDENTIFIEDLEAKAGEshallbeallleakagewhichisnotIDENTIFIEDLEAKAGE. I VENTILATION EXHAUST TREATMENT SYSTEM 1.fAVENTILATIONEXHAUSTTREATMENTSYSTEMshallbeanysystemdesignedand 1 7 installed to reduce gaseous radiciodine or radioactive material in particu-late form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

VENTING l 1.4hVENTINGshallbethecontrolledprocessofdischargingairorgasfroma l l 8 confinement to maintain temperature, pressure, humidity, concentration or i other operating condition, in such a manner that replacement air or gas is l not provided or required during VENTING. Vent, used in system names, does i not imply a VENTING process.

LA SALLE - UNIT 2 1-7 Amendment No. 69 l

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r- ISOLATION ACTUATION INSTRUDENTATION E -

VALVE GROUPS MININUM OPERABLE APPLICA8tE OPERATED g TRIP FUNCTION SIGNAL a CHANNELS PER OPERATIONAL TRIP SYSTEM (b) CotR)ITION ACTION

,o A. AUTOMATIC IMITIATION -

1. PRIMARY CONTAlfetENT ISOLATION

a. Reactor Vessel Water Level (1) Low, Level 3 7 2 1,2,3 (2) Low Low, Level 2 20 2, 3 2' 1, 2, 3 20 (3) Low Low Low, Level 1 1, 10 2 1,2,3 20 *

b. Drywell Pressure - High 2, 7, 10 2 1,2,3 20

c. Main Steam Line R

1) Radtatton - High 1 2 1, 2, 3' 21 3 2 1,2,3

';* 2) Pressure - Low 22 1 2 p 3) Flow - High 1 2/11ne(d) 1 1,2,3 23 21

d. Main Steam Line Tunnel Temperature - High 1 2 I IIIII) ,2IIIIII, 3(1)(j) 21

e. Main Steam Line Tunnel ATempera{ure-High 1 2 I IIIIII 2(IIIII, 3(1)(}) 21

f. Condenser Vacuum - Low 1 2 1, 2*, 3* 21

2. SECONDARY CONTAlletENT ISOLATION

a. Reactor Building Vent Exhaust Plenum Radiation - High 4 ICII*) 2 1, 2, 3 and ** 24

[ b. Drywell Pressure - High 2 1,2,3 4(c)(e) 24 i? c. Reactor Vessel Water g Level - Low Low, Level 2 4(CII'I 2 1, 2, 3, and #~

24

d. Fuel Pool Vent Exhaust S Radiation - High ,

4(c)(e) 2 1, 2, 3, and ,, 24 l

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LA SALLE - UKIT 1 3/4 3-13 Amendment No. 14 l

e e t l

l

TABLE 3.3.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION l

ACTION STATEMENTS ACTION 20 -

Be in at least HOT SHIRDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN with the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 21 -

Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within ACTION 22 - 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 .

L Close the affected system isolation valves within I hour and declare the affected system inoperable.

ACTION 23 -

' Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24 -

Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within I hour.

ACTION 25 - Lock the affected system isolation valves closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

ACTION 26 -

Provided that the manual initiation function is 0PERABLE for each other group valve, inboard or outboard, as applicable, in each line, restore the manual initiation function to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; otherwise, restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ; otherwise:

a.

Be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , or

b. Close the affected system isolation valves within the next gg hour and declare the affected system in oparable.

TABLE NOTATIONS May be bypassed with reactor steam pressure < 1043 psig and all turbine stop valves closed.

When handling irradiated fuel in the secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

(a) ee Spec' fication 3.6.3, Table 2.6.3-1 for valves in each valve grouoJ A (b) A channe' may be placea in an inoperable stat ~us for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the channel in the tripped condi-tion provided at least one other OPERA 8LE channel in the same trip system

-is monitoring that parameter. In addition for those trip systems with a design providing only one channel per trip system, the channel may be placed in an inoperable status for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for required surveillance testing without placing the channel in the tripped condition provided that the redundant isolation valve, inboard or outboard, as applicable, in each line is operable and all re uired actuation instrumentation for that re-dundant valve is OPERABLE, .. lace the trip system in the tripped condition. ,

(c) Also actuates the standby gas treatment system.

(d) A channel is OPERABLE if 2 of 4 instruments in that channel are OPERABLE.

(e) Also actuates secondary containment ventilation isolation dampers per Table 3.6.5.I 1.

(f) Closes only RWCU system inlet outboard valve. 1 4

LA SALLE - UNIT 2 3/4 3-14 Amendment No. 61

'a TABLE 3.3.2-3 (Continu:d)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

6. RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION N/A

a. Reactor Vessel Water Level - Low, Level 3

b. Reactor Vessel (RHR Cut-In Permissive) Pressure - High

c. RHR Pump Suction Flow - High

d. RHR Area Cooler Temperature High

e. RHR Equipment Area AT High B. MANUAL INITIATION N/A

1. Inboard Valves

2. Outboard Valves

3. Inboard Valves

4. Outboard Valves

5. Inboard Valves

6. Outboard Valves

7. Outboard Valve TABLE NOTATIONS

Isolation system instrumentation response time for MSIVs only. No diesel generator delays assumed.

- Radiation detectors are exempt from response time testing. Response time shall be measured from detector output or the input of the first electronic component in 'che channel.

k Isolation syst]emn_strumentationresponsetimespecifidortheTrip Function actuatinWeach valve group)shall be added to(isolation time hown I

(in Table 3.6.3- 1 and 3.6.5.2- 1 for valves in each vzTve greyto obtain ISOLATION SYSTEM RESPONSE TIME for each valve. t

& Msn/s usiv l

N/A Not Applicable.

!a SAttF - UNIT 2 3/4 3-19 AMENDMENT NO. 8?

r g 3/4.6 CONTAINMENT SYSTDtS 3 3/4.6.1 PRIMARY CONTAINMENT U BM tntS ty~djve Coritrof PRIMARY CONTAINMENT INTEGRITY QS)D6r'm/[fgo LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAllMENT INTEGRITY shall be maintained.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2,* and 3. !

ACTION:

Without PRIMARY CORTAINMENT~ INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY ;

within I hour 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 SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY s',iall be demonstrated: P,

( After each closing of each r,enetration subject to Type B testing, ,

except the primary contairraent air locks, if opened following Type A or B test, by leak rate testing the seal with gas at Pa, 39.6 psig, and verifying that when the measured leakage rate for these seals is !

added to the leakage rates detemined pursuant to Surveillance Requirement 4.6.1.2.d for all other Type B and C penetrations, the ,

Q combined leakage rate is less than or equal to 0.60 La. ,

At least once per 31 days by verifyin that all primary containment penetrations ** not capable of being c osed by OPERABLE containment 1 automatic isolation valves and required to be closed during accident 4Wew . conditions are ciosed by valves, biind fianges or deactivatnd

, f rom automatic valves secured in position, exceptdsgyroTrided in "ablQj lIg3{gp (3.6.3-1 oDSpecification 3.6.3. T

c. By verifying each primary containment air lock OPERABLE per Specification 3.6.1.3.

d. By verifying the suppression chamber OPERABLE per Specification 3.6.2.1.

e. Verify primary containment' structural integrity in accordance with the Inservice Inspection Program for Post Tensioning Tendons. The frequency shall be in accordance with the Inservice Inspection Q o TNSERTc(' gram for Post Tensioning Tendons.

See Special Test Exception 3.10.1

- Except valves, blind flanges, and deactivated automatic valves which are located inside the containment, and are locked, sealed or otherwise secured in the close' position. These penetrations shall be verified closed during '

each COLD SHUfDOWN except such verification need not be performed when the primary containment has not been deinerted since the last verification or more often than once per 92 days.

9 3/4 6-1 Amendment No. 84 ;

LA SALLE - UNIT 2 1

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSEffECHNICAL SPECIFICATIONS page B-4 INSERTH 4.6.1.1.b. Perform required visual examinations and leakage rate testing except for primary containment air lock testing and main steam lines through the isolation valves,in accordance with and at the frequency' specified by 10 CFR 50, Appendix J, as modified by approved exemptions.

The overall integrated leakage rate acceptance criterion is s 1.0 L,. The Type B and C combined leakage rate acceptance criterion is s 0.60 L,.

However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, the leakage rate acceptance criteria are < 0.60 L, for the combined Type B and Type C tests, and < 0.75 L, for the Type A test.

INSERT C1 The provisions of Specification 4.0.2 are not applicable to the frequencies j specified by 10 CFR 50, Appendix J.

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Limwai COWITION POR OpstATION 3.5.1.2 Primary containmast leakage ratas shall he 11mitad to: .

a. An everall integretad leakage rate of less than or equal to L,,

8.8N percent by weight of the containment air per 34 hours3.935185e-4 days 0.00944 hours 5.621693e-5 weeks 1.2937e-5 months at P,,

5.1 psig.

h. A emeiant leakage ruta sf less than er' equal to 0.00 L, for all penetrations and aT1 valves listad is Tele 3.4.3-1 except for main steen isolaties valves and valves which are Iqrdrestatically leak

' tasted per Tels 3.8.3-1, adject to Type 5 and.C tests when '

~

pressartand to p., 3s.s pois.

c. 8 tass then er equal to 100 ser per hour for all four main staan lines through the feelaties valves.when testad at 25.0 pois. .

d A camined Tankage ruta of less than or equal to 1.gis times the -

total number of ECCS and ACIC containment isolation. valves in hydre- i statically tastad lines ufrich penetrate the primary containment. l when testad at L18 P,, 48.6 psig. ,

APPLICA81LITY: idhan PRDMY CONTADOWIT affBIRITY 'is required per Specification 3.6.1.1.

M .

with:

i

a. .Tha measured evers11 integrated primary containment leasage rata i

.e edine,a:7s L , .r .

l e cr.

h. ,The amasured cambined leakage reta for all penetrations and all

?..valves valvesamt fistad in Tele valves s ich3.5.3-1, except for main are Iqrdrestatically leak steam testad isolation per .

' Tele 3.8.3-1, subject to Type 5 and C tests exceeding 0.50 L,, or

c. The amasured leakage ruta exceeding 100 scf per hour for all four 2 meio stamm 11 ass through the isolation valves, or i

d. The measured camined leakage rate for all ECC$ and.RCIC containment 1selstian valves in Iqyerestatically tasted lines which penetrata the primary containment succeding 1 gym times the total number of such ,

valves, l

Exasgition ta Appendix J ef 10 CFR Part 50. -

w_ .m <v CNextpa e 15 (g-LA sAtts - uwIT 2 3/4 6-2 A

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DELETE PAG.E

.i ADOENT SYSTB6 '

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Lemi.CDSTTION POR opt 1tATION (Continued)

M (Continued) ,

. ._. .m -

a. overall fatagrated leakage rate (s) to less than er equal to 0.75

's- '

k. The ined 1essage rata for all penetrations and all valves listed in Tel 3.8.3-1, except for main staes isolation valves and valves l dich ' Iqydrostatically leak tasted per Tele 3.5.3-1, seject tu !

Type & C testa te less thee er equal to 0.00 L,, ans

, l c The lealwgo to less then er equal 'a 200 scf per hour for all .

four meia lines through the isol.ction valves, and

. . d. The eneined I reta for all.ECCS mut RCIC ' containment isolation- ,

valves la cally tasted liner dich penetrate the primary l containment to 1 then er equal to 1 gis times .the total number of ;

such valves, .

Prior ta increeSing rautar a t syst s t e move 2 N 'F. !

SUltVEILLANCE'ItEDUIRTi 4.5.1.2 The primary containment. I rates shall be dommstented at the following test schedule and shall be ned in confomance witt the critaria specified in Appendia J af la CFR Part using the methods and previsions of ANSI M .4-1972- , ,

s. Three Type A Overall Integrated inment Laakage Rate tests shall be emneucted at 40 s 30 seeth 1 1s during shutdown at P,,

39.8 peig, dering" each 10 year period. The third test of each set shall be conducted during shutdown for the 10 year plant inservice inspection.

h. If arqr periodic Type A test fails to 0.75 L,, the test schedule for seneguent Type A tests shall be revi and approved try the ..

Cammissioe If tuo consecutive Type A tas fail ta most 0.75 L,, a The A test shall be perfereed at least every months until two consecutive Type A tests meet 0.75 L,, at 21 '

time the'above test schedule ear be resumed.

c. The accarecy of each Type A test shall be.verifi try a supplemental test dich:

2. Canfims the accuracy of the test try verifying t the difference between the supplemental data and the A test ;

data is within 0.25 L,.

L Has duration sufficient ta establish accurateM the in 1 1eakage reta'between the Type A tast and the s'uppl test.

3. Requires the quantity of gas injected ints the contai er bled free the containment 1uring the supplemental tast to be equivalent to at least 255 of the total asasured leakage a P,, 39.5 psig.'

LA SALLE - UNIT 2 3/4 6-3 . l

-n- .::

. .- .n s . ..

. y,5 9; ., .: . .. - 4 ....h..__., .

DELETE PAGE AINGif 5Vmer6 REDUIREMENf3 (Continued) - -

d. I and C tests shall be conducted with gas at P,, 39.5 psig*, at i als no grestar then 24 months sucept for tests involving:

L 1r locks, .

2. steam line isolation valves, .

3. Valv pressurtzed with fluid from a seal systas, and

4. ECCS ECIC containment isolation valves in hydrostatically testad 1 d ich penetrate the primary containment.

~

e." Air locks shall tastad and demonstrated OPSABLE per surveillance Requirement 4.5. ,

~

f. Main stans line iso en valves shall be . leak testad at least once per 18 months. ,

g. Lankage from isolation Ives that are sealed with fluid from a seal system he excluded, jact to the provisions of Appendix J,

! Section C.3, when data ning the camined leakage rate ded l

.the seal system and valves pressurized to at least 1.10 ,,

43.5 psig, and the seal sys capacity is. adequate to maintain systas pressure for at least days.

h. ECCS and E!C containment isolati valves in hydrostatically testad lines dich penetrate the primary inment shall be leak tasted at least once per la months. .

1. The provisions of specification 4.0.2 not applicable to 24 month er 40.2 la month surveillance intervals.

'vntess .a styaraulic tast is required per T'able 3.6.3-1. -

e f LA 5ALLE - UNIT 2 3/4 6-4 .

~

CDMAINMENT SYST' EMS

<v.

SUWEILLANCE REQUIREMENTS (Continued) l

c. By verifying at least 2 suppression chamber water level instru- ,

mentation channels and at least 14 suppression pool water tamperature l instrumentation channels, 7 in each of two divisions OPERA 8LE by performar;e of a:

L CHMNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and

3. CHANNEL CALIBRATION at least once per 18 months.

The suppression chamber water level and suppression pool -

temperature alara setpoint shall be:

a) High totar level 1 +2 inches

b) Law water levei 3, -3 inches * . .

,, c) High tamperatum 120$*F

d. I By conducting drywell-to-suppression chamber bypass leak tests and g

. 4 ver.ifying that the A/4 calculated from the measured leakage is .

within the specified limit when drywell-to-sgpression chamber bypass leak tests are conducted:

L At least once'per 28 months at an initial differential pressure of L5 psi, and

2. At the first refueling outage and then on the schedule required for Type A Overe11 Integrated Containment Leakage Rate tests by Speci-fication 4.5.L2.a., at an initial differential pressure of 5 psi, except that, if the first two 5 psi leak tests performed g to that ties result in:. .

b y >:- -

2. A. calculated A/4 within the specified limit, and

. .; e ;. . . .

22 The A/8 calculated from the' leak tests at L5 psi is <~ 205 of

.,., the specified limit, then the leak tests at 5 psi may be discontinued.

j had *INSERTC2" '

" Level is referenced to a plant elevation of 699 feet 11 inches (See g Figure.B 3/4.6.2-1).

LA SALLE - UNIT 2 -

3/4 6-21 Amendment No. 49

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSEffECHNICAL SPECIFICATIONS pace B-5 INSERT C2

d. By conducting drywell-to-suppression chamber bypass leak tests at least once per 18 months at an initial differential pressure of 1.5 psi and verifying that the A/Vk calculated from the measured leakage is within the specified limit.

If any 1.5 psi leak test results in a calculated NVk > 20% of the specified limit, then the test schedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 1.5 psi leak tests result in a calculated A/4k greater than the specified limit, then:

1. A 1.5 psi leak test shall be performed at least once per 9 months until two consecutive 1.5 psi leak tests result in the calculated NVk within the specified limits, and

2. A 5 psi leak test, performed with the second consecutive successful 1.5 psi leak test, results in a calculated NVk within the specified limit, after which the above schedule of once per 18 months for only 1.5 psi leak tests may be resumed.

If any required 5 psi leak test results in a calculated NVk greater than the specified limit, then the test schedule for subsequent tests shall be reviewed by the Commission.

If two consecutive 5 psi leak tests result in a calculated NVk greater than the specified limit, then a 5 psi leak test shall be performed at least once per 9 months until two consecutive 5 psi leak tests result in a calculated NVk within the specified limit, after which the above schedule of once per 18 months for only 1.5 psi leak tests may be resumed.

~

4:sniaNiasallei21rttina wTf47 i

CONTADeeff 5YST1M5 SURVEILLaCE REOL11REMENTS (Contie_H)

[.. .I 1

If any 1.5 psi er 5 psi leek test results ta:

L , A calculated V4 grestar them the'specified lief t, or .

1 L

A calculated V4 from a L5 poi leak test > 205 of the specified limit. . - J then the test schedule for seseguent tasts shall be reytened by the ,

cassission.

If tus ceasecutive LE pai leak tasts result is a calculstad Al4 grestar -

them the spectflee iisit, th :

L A.LS psi Teek test shall be performed at least once per ,

5 masths, estil tus ceasecutive Lf psi . leak tests. result is ther 1:alamisted V4 witMs the speciftet limits, and ,

L A E psi leek test, performed with the second consecutive .

eennessful La psi leek test, results is a calculated V4 '

witkie ther specified Itsit..after which the aheve schedule for esly'LE pst Teek, tasta say be resumed. .

~

r . . .. ,.;., -

If tus canaecutive 5 pst leak testa rese1t le a es1calstad V4 greater thee the specified finit, them a 5 poi leak test shall be perfemed at least esce per 5 months estil the consecutive 5 psi leak tests result in '

a calculatat #4 withis the specified limit, after which the above schedule for sely La poi leak tests any be resumed. -

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%s LA SALLE - UNIT 2 3/4 6-22 l

1

CONTAINMENT SYSTEMS DW 3 /4. 6.3 PRIMARY CONTAINMENT ISOLATION VALVES "INSERTD" LIMITING CONDITION FOR OPERATION /

k v

T 3.6.3 The primary containment isolation valves and the reactor instrumentation line excess flow check valves shown in Table 3.6.3-1 shall be OPERABLE with isolation times less than or equal to those shown in Table 3.6.3-1.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a. With one or more of the primary containment isolation valves shown in Table 3.6.3-1 inoperable:

1. Maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either; a) Restore the inoperable valve (s) to OPERABLE status, or b) Isolate each affected penetration by use of at least one deactivated automatic valve secured in the isolated position," or c) Isolate each affected penetration by use of at least one closed manual valve or blind flange.* l

2. 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one or more of the reactor instrumentation line excess flow check valves (shown in Table 3.6.3-1) inoperable:

D

1. Operation may continue and the provisions of Specification 3.0.3 are not applicable provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a) The inoperable valve is returned.to OPERABLE status, or b) The instrument line is isolated and the associated instrument is declared inoperable.

2. 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 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Ma *rNssarg"

lsolation valves closed to satisfy these requirements may be reopened on an l ntermittent basis under administrative control.

LA SALLE - UNIT 2 3/4 6-25 Amendment No. 78

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSEffECHNICAL SPECIFICATIONS patre B-G INSERT D 3.6.3 Each primary containment isolation valve and reactor instrumentation line excess flow check valve shall be OPERABLE **

APPLICABILITY: OPERATIONAL CONDITIONS 1,2, and 3.

ACTION:

a. With one or more of the primary containment isolation valves, except the reactor instrumentation line excess flow check valves, inoperable:

INSERT E

- Locked or sealed closed valves may be opened on an intermittent basis under administrative control.

k;\nlasissalle'11rttina.wpf48

SURVEILLANCE 8L8ru11RDENT5 Y -- ~ - . ~ :'. :* . .:.:~ ~.

"* -~

~'4.5'T.T

. Eacit prfr#y containment [ isolation valve @ in Table 3.6.3-)shall ~

be demonstrated trJtABLE prfor to returning the valve to service after mainte-nonce, repair er replacement work is performed en the valve or its associated actuator, control er power circuit by cycling the valve through at least one completa cycle of full travel and verifing.the specified isolation time.

1.E.3.2 Each primary containment autaustic isolation valvs d (Tahia Ls.3-Dshall be emmenstrated OPERABLE during COLD SHLT1MM er REFUELING at least once per 28 aanths by verifying that en a containment isolation test Mc isolaties valve actustas to its isolation position.

signal =? '

(ISOjotfa

4.5.3.3 ghe es ties of sary containment power operstad er autaustfCvalve to Tame a.aa- shall be determined to be within its ,

limit when testad pursuant to spectfication 4.0.5.

~

4.5.3.4 Eacir reactor fastrementation 11'ne escass flow check valve -

Table 3.5.3-Jsha11 he emmenstrated OPERABLE at least once per 18 months by vertfying that tha valve checks fles. .

4.5.3.5 Each traversing in-core probe system explosive isolation valve shall be demonstrated OPERASLE: .-

l a; At Taast once per 31. dayer by verifying the continuity of the l aglesive charga.

l

k. At least once per 28 months by removing the explosive squib free at least ans emplosive valve such that the explosive squib in each emplosive valve will be tastad at least once per 90 months, and irttisting the emplesive squib. The replacement charge for the exploded squib shall be from the same manufactured batch as the one fired er free another batch which has been certified by having at least one of that batch successfully fired. No explosive squib

- , shalf russis ?g use buyend the expiration of its shelf-life and

. .W sperstiag-} ifs. - . .

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- Aog ".INSERTG LA SALLE - latIT 2 3/4 6-25

1 ATTACHMEN'I' B PROPOSED AMENDMENTS TO THE LICENSEffECHNICAL SPECIFICATIONS page B-7 INSERT F 4.0.3.6 At least once per 18 months:

a. Verify leakage rate through all four main steam lines through the isolation valves is s 100 scfh when tested at 2 25.0 psig.*

b. Verify combined leakage rate of s 1 gpm times the total number of primary containment isolati9n valves through hydrostatically tested lines that penetrate the primary containment is not exceeded when these isolation 1 valves are tested at 1.1 P,,2 43.6 psig.* l I

l INSERT G (Footnote) 1 Results shall be excluded from the combined leakage for all '

penetrations and seals subject to Type B and C tests.

I s . 5.n la giata 11e s 21 r t f ina . sy f 49

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_h R 1 Ed j .. B.: .

e M5" I ; Ie -

! ! \'. S' 138 'i -

! 4j 11 :

W II=RI==RRRf!IllIII R I Is==n E!! ~ -~--

$ i i 3

' LA SALLE - UNIT 2 3/4 6-28 Amen ,nt N- 25

-e - - - - - - - - _ ._ - - - - - . - _ _ -

~ '

TABLE 3.6.3-1 (Continued) i

.t'

~

5- PRIMARY C00lTAll0 G T ISOLATION VALVES - .

VALVE GROUP I ,I It4MIMuM ISOLATI0ll TI . -

VALVE FUIICTIGIl AIS 9889ER. (Seconds V g -

i a Automatic Isoletten Valves (Continued)

11. Containmentpesqflerj Valyes 2 i l l 2CM017A.B ~.'J.,

~

2CM0184,5 i &*A .'

20W194,5 3W '@ .J.I' '

2002M,8 A l* ' V'?, - '

2CfE021B(h) , .{

r_ i '

20*ttAh

,C,.m 2040264 I "I l T

2CM027 m i 20ers -

\\) '

I i 2 00029 %

II 204030 20ED31' I

f

(!

2CM032 ,

L 2CM033 i i

2CM034 i

12. Drywell Pnematic Valves - '

2IN001A an,a le $ 30 ,

2111017 10 $ 22 2IN014 le S 22 '

2In015 10 . 1 22 2INO31 2 15

{g 13. RHR She 2E12-F Coeling Mode Valves 5 -

i 48 - '

2E1 009 5 40

[

= 2-F023 i 90 i 2E12-F053 A and 8 1 29 ,,

r

- ,,y.,,, , ,,,g(.)(i) 3m i s

e e

b I

^

J

?hGE w _ _ _

j s d

a vi se vivi es vivi

=

v1 s

E

~

h '

W l' =:it .

e o 5 -

l S

l

- * * = = = w 4 E =

W -

r. 5 E g m a 3 5 I 1IkE s"

.E A 7

.I21j 5:

Ir 1 1 j

E A i

l 3 : : u =

e

%*1a 2

" 5 1 nr . "3 f u

G t "2 g g "gg gg y enemmem=. c 2 . !,1 ,

a 4; ._...8-s 3 --

= s. 22 1111111 s s 3 2113 - a ~- -

4I8 _ *~EE:Eku 3 4 3:s.e. e t--

a.s..-8.11Im:. ......:.

a 3. . g :: A

e e M .I. ,

_ AA 4 i:Ma v.s

2 =EA E I e ,=_,32222 I AAAAl:~ ---

2222R W -

I dd i d i d m

7. . , . -- . . , .

o.

TABLE 3.6.3-1 (Continued) g .n

, PRIMARY CONTAINNENT 150 TAT 10N VALVES .

MAXI g 150 TINE VALVE ruMCY10N AND NUMBER VALVE GNOUP ,) ands) h ,

U b. Maaned Isoletted Valwe,s I

1. 2FC006 '

N.A.

2. 2FC113 N.A.

3. 2FC114 N.A. -

4. 2FC115 N.A.

5. 2MCO27 III N.A.

4. I N.A. -

7. 2MC033(

25A042 N.A.

'j

s. 2sA046(I) N.A. '

s.*

k T '

M .

- ~

l i-

. (T,T t - -

E G

TABLE 3.6,3-1 (Contisimed)~

PRIMARY CONTAIIstENT 150LAT38N VALVE 5

,m yALVE FUNCTICII NW IR8SER ,

Excess Flow Check ValvesI 8I c.

1. 2821-F374 '

r

" 2. 321-F376 -

3. 321-F359 . .

4. 321-F355 '

3/ 5. 2821-F M 1

. [ __

ji 5. 321-F378' ' - a is 7. 2921-F372 -1 .

g 6 8. 2B21-F379 w

9. 2B21-FM 3

.. 2.n;M. .

1. . -

11. N 21-F415A, 8 -

12. 2821-F357 ,

13. 2821-F382 .

I -

14. 2821-F3284, 5, C, 8 %

f

15. 2821-F327A, S. C, 8

16. 2921-F4134, 8 , -

. 17. 2021-F344 '

18. 2821-F l

~

19. -F443 -

2321-F439 -

s .

21. 2821-F437 .

22. 2821-F441 -

23. 2921-F4454, 4 .

l 24. 2821-F453 B 5

.. - _ _____m__

l * .

IABLE3.6.3-1(Centinmed) -

P83MAltf CONTAIMMENT 1501.4110N val.W $ .

/ .

. VALW Fl8ICTieN MS laseEq ' , ., ,

l EncessFlowCheckVelvesf,N.(Continued) .

25. 2s21-F447 .-

. .. r M. 321-f4554,B ,.,,-

27, 321-F451 i '*

i

~

28. 2821-F449 '

29. 2821-FM7 .

M. 3 21-F3264, 8, C, N '

31, 2821-F3254, 8, C, e 'e

$ 32. 321-F350 .

33. 2821-F344 r -

T -

0 34. 2821-F348 N

35. 321-F471 - b 36, 2821-f473 - t

37. 2921-F469 -

l

30. 2821-F4754. B '

39, 2821-F4454, 8 . l I

44. 2821-F467 & -

41. -N21-F463

- I I

42. 2921-F ,

43, 2G33- 12A, E

44. 3-F309

. 2E12-F315 44, 2E12-F3594, 8 .

47. 2E12-F319 ,

e

, . - n - - -

g . .

~.

g TAsLE 3.s.3-1 (cents ned) '

PRIlWlY CONTAl W NT ISOLATION VALVE 5

, VALVE nasc7Ian am meieEn .

g Excess Flow Check ValvesI8I (Contimmed)

Of .l U ' R' ; y . 7

48. M 12-F317

49. M12-F3604,p ,

f l

a

',)*. '

50. M21-F304

- Y

51. M22-F304

52. N22-F341

53. M tt-F342 i

~

54. 3 33-F3194. 8

$ 55. 2333-F317A. t

54. 2333-F3134. B C, B y i

57. 2933-F311A, S. C, 3 I '

54. 2933-F315A, B, C, 9

59. 2933-F301A, 3

-- I i -

68. 2813-F307A 8, C, 3 ,

51. 2033-F305A, 0, 6h 29804 '

63. 20e0-M. U .

6 2aste ' "

66. 2V4061

" $7. 2921-F457 .

. 68. 2821-F459

$; 69. 2821-Fj61 ro. 2cnnot .

71. 2821-F570 l

72. 2821-F571

TABLE 3.6.3-1 (Continued) ,

PRIMARY CONTAIPMENT ISOLATION VALVES E VALVE FUNCTION AND NUMBER g d. Other Isolation Valves

1. MSIV Leakage Control Systes_

2E32-F001A, E, J N(b)

2. Reactor Feedvater and RWCU System Return 2821-F010A, 8 2821-F065A B j

2G33-F040*I __

3. Residual Heat Removal / Low Pressure Coolant injection stem w 2E12-F042A, 8, C y 1 2E12-F016A, 8 g e 2E12-F017A, 8 g a 2E12-F004A, 8 g -

2E12-F027A,8(jj(j) j i

2E12-F024 )B j) 2E12-F021 2E12-F302 II) gy)

- For the remainder o ycle 5, or until the first tage in i 2E12-F064A, Bg 2E12-F011A,8jj which the unit 1 in Cold ,

g Shutdown for t weeks or 2E12-F088A, 8 C(j) greater dura , the Type k. y 2E12-F025gy)B,C C test is required to be 2E12-F030(3 2E12-F005 current f the 2G33-F040 valve its leakage is not

{ 2E12-F0 2E12- 74A, B

,8 requi the to be included in tal Type 8 & C leakage 2 -F055A,Bgj) a E12-F036A, Bg j) s fled by Specification

2E12-F311A, B y'gg) 3 .1.2.b.

F 2E12-F041A, Bggj l

2E12-F050A, 8 l .T 1

l . .

l

TABLE 3.6.3-1 (Continued) p

- PRIMARY CONTAllMENT ISOLATION VALVES >

r-G VALVE FUNCTION AND NUpBER ;

g Other Isolation Valves (Continued) .

4. Low Pressure Core Spray System .

2E21-F005 '

2E21-F001 IO) 2E21-F032 b) 2E21-F011 U) 2E21-F018 I2 } ~~

2E21-F031h)-

2E21-F006

}

5. High Pressure Core Spray Systen jf R 2E22-F004 2E22-F015 b I T 2E22-F023 UI M 2E22-F012 U) b .

2E22-F014 U) 2E22-F005 0)

6. Reactor Core isolation ling System 2E51-F013 2E51-F069 h'

2E51-F028 2E51-F06 2E51- :

[

a 2E 031g) g 51-F019h;))

a 2E51-F065 f 2E51-F066 0) o

2E51-F059(("))

2E51-F022 "

E 2E51-F362(I")

2E51-F363 ")

IABLL .6.3-1 (Continued) h PRIMARY CONTAlWENT ISOLATION VALVES h VALVE FUNCTION AND NUMBER Other Isolation Valves (Continued)

M 7. Post LOCA Hydronen Control N

2HG001A, B -

2HG002A, B 2HG005A, B 2HG006A, B

8. Standby Llould Control System 2C41-F004A, B -

2C41-F007

, g i

9. Reactor Recirculation Seal Injection j

U 2B33-F013A, B'I' 2B33-F017A, B'I' hh W,

10. Drywell Pneumatic System - A Y

ZIN018 g

11. ReferenceLeoBackftI 2Cll-F4228

()

2Cll-F422 2Cll-F F 201 422G ;

d

. 1-F4238 2Cll-F423D k 2Cll-F423F 2Cll-F423G A

.F

l TABLE 3.6.3-1 (Continued)

E g PRINARY CONTAllt9ENT ISOLATION VALVES .

' . E TABLE NOTATIONS E *But 2 3 seconds.

Q (a) See Specification 3.3.2, Table 3.3.2-1, for isolation signal (s) that operates each valve gro .

m b) Not included in total sum of Type B and C tests.

c) May be opened on an intermittent basis under administrative control.

d) Not closed by SLCS actuation.

e) Not closed by Trip Functions Se, b, or c Specification 3.3.2, Table 3.3.2-1.

f Not closed by Trip Functions 4a, c, d, e, or f of specification 3.3.2, Tabl .3.2-1.

g Not subject to Type C leakage test.

(h Opens on an isolation signal. Valves wl11 be open during Type A tes No Type C test required.

1) Also closed by drywell pressure-high signal.

) Hydraulic leat test at 43.6 psig.

Not subject to Type C leakage test - leakage rate tested Specification 4.4.3.2.2.

w 3Thesepenetrationsareprovidedwithremovablespools rd of the outboard isolation valve.

1 During operation, these lines will be blind fla ing a double 0-ring and a type B leak ,

e test. In addition, the packing of these isolat valves will be soap-bubble tested to ensure J. Insignificant or no leakage at-the containes test pressure each refueling outage. -

ii;' (m) If valves 2E51-F362 and 2E51-F363 are I lclosed and acceptably leak ,

rate tested, then valves 2E51-F059 a 51-F022 are not considered to be primary containment isolation val and are not required to be leak rate tested. ,

(n) Either the 2E51-F362 or. 2E51-F363 valve may be open when the RCIC g system is in the sta mode of operation, and both valves may be open during operation o

- that:

RCIC system in the full flow test a6de, providing (7 i

f5N,NI ku 1) vs 2E51-F022 is acceptably leak rate tested, and .

P7 %

I valve 2E51-F059 is deactivated, locked closed and acceptably leak y a rate tested, and e

3) the s downstream of the 2E51hectacleflange,installedimmediate1 059 valve, is closed and acceptably eak rate tested.

m l

I nfOrrMQif0M M 3/4.6.4- VACUUN RELIEF gg (b4Mgg3 LINITING CONDITION FOR OPERATION .

3.6.4 All suppression chamber - drym11' vacuum breakers shall be OPERABLd and closed.'

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.

t ACTION: ,

s. With one suppression chamber - drywell vacuum breaker inoperable -

and/or open, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months close the manual isolation valves on both sides of the inoperable and/or open vacuum breaker. Restore the inoperable and/or open vacuum breaker to OPERA 8LE and closed status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HDT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. With one position indicator of arty OPERABLE suppression chamber -

drywell vacuum breaker inoperable, restore the inoperable position indicator to OPERABLE status within 14 days or visually verify the vacuum breaker to be closed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Otherwise, declare-the vacuum breaker inoperable. ,

SURVEILLANCE REQUIREMENTS 4.6.4.1 Each suppression chamber - drywell vacuum breaker shall be: ,

a. Verified closed at least once per 7 days.

b. Demonstrated OPERABLE: ,

1. At least once per 31 days and within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after arty discharge of steam to the suppression chamber from the safety-relief ;

valves, by cycling each vacuum breaker through at least one complete cycle of full travel.

2. At least once per 31 days by verifying both position indicators OPERABLE by performance of a CHANNEL FUNCTIONAL TEST.

3. At least once per 18 months by; a) Verifying the force required to open the vacuum breaker, from the closed position, to be less than or equal to 0.5 psid, and b) Verifying both position indicators OPERABLE by performance of a CHANNEL CALIBRATION.

k I

i

- LA SALLE - UNIT 2 3/4 6-38 3

j i

i

. l 1

l CONTAIMMDir sysTDt3 I l

SURVEII. LANCE REQUIREMENT 5 (C'ontinued) 4.E.4.2 .The annual isolation valves on both sides of an inoperable and/or '

open suppression e ' _- t 7_;11 vacuus breaker shall be verified to be closed at least once per.7 days.

[4.5.4.3 Vacuum breaker header flanges which have been broken shall be leak -

( tasted after remaking by Type E test at 39.5 psig per Specification 4.5.1.2.d. j l

b .

e 8

4

. i

g 4

e 4

.O es LA SALLE - UNIT 2 3/4 6-39 l

. i:, *

- ~ -

. 1 I

ColfTAIWWff SYSTeti 3/4.5.5 PRIMARY CDefTAIIsett ATW5MtERE C0ffTit0L

. _ DemdELL Ale SUPMtESSI0ff CHAfsp NYDitoGEN ItEC0pSINER SYSTDIS -

J Luuuni CDWITIoll POR OPetATICII

~

3.5.5.1 Two independent drywell' and suppression chamber lqrdregon recombiner systems shall be OPBASLL ,

! A9 U CAgIUTY: OPDATIINIAL COISITIofts 1 and 2. -

ACTTOII: .

tHth one drywell and/or sgpression chamber hydrogen recombiner system inoperable, restare the inoperable system to OPDASLE. status within 30 days or be in at least WT 56EffDOWI within the next 12'heurs., ,

SURVEILLAf1CE -_-RRTi

4. 5. 5.1 Each dryuell and.Auppression chamber tqyttregen recombiner systas shall be esmanstrated OPDASLE: .

a. At leeirt once per SE days luy cycling each flow control valve and rec 1resistian val at least one campista cycle of full trevel. -

i b. At laest once per- by verifying, during a recombiner system i functional tast: 1 L That the hosters are OPDASLE try determining that the current l l

in each phase differs try less than or equal to 5 from the ether phases and is within 5 of the value observed in the -i original acceptance test, corrected for line weltage differences.

L That the reaction chaser gas tamperature increases to 1200 a 25'F within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . .

c. At least ones per 38 months try:

L Performing a CHAf01EL CALIBRATION of all recediner operating instrumentation and control circuits.

2. Verifyfng the integrity of all hoster electrical circuits by

~

performing a resistance to ground test within 30 minutes following the above requimd functional tast. The resistance to ground for asqr heater phase shall be greater than or equal to 100,000 ohms.

d. By measuring the leakage reta:

L As a part of tM evere11 integrated leakage rate test required try Specification 3.5.L2, or _{

L Dy measuring the leakage reta of the system outside of the containment isolation valves at P ,, 39.5 psig, on the schedule required try Specification 4.5.L2 ad including the measured

- leakage as a part of ths leakage dotardned in accordance with Specification 4.5.L2.

LA SALLE - tmIT 2 3/4 6-45

l l

- Ih "INSE'RT H" l

l 3/4.6 CONTAINMENT SYSTEMS l BASES ;

3/4.6.1 PRIMARY LONTAINMENT 3/4.6,1.1 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

The structural integrity of the primary containment is ensured by the successful completion of the Inservice Inspection Program for Post Tensioning Tendons and by associated visual inspections of the steel liner and penetrations for evidence of deterioration or breach of integrity. This ensures that the structural integrity of the primary containment will be maintained in accordance with the provisions of the Primary Containment Tendon Surveillance Program. Testing and frequency are consistent with the recomendations of Regulatory Guide 1.35, Revision 3, except that the Unit I and 2 primary containments shall be treated as twin containments even though the Initial Structural Integrity Tests were not within 2 years of each other.

" 3/4.6.1.2 PRIMERYCONTAINMENTLEAKAGE The limitations on primary containment leakage rates ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 39.6 psig, P . As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 L, during performance of the periodic tests to account for possible degradation of the containment leakage barriers between leakage tests.

Operating experience with the main steam line isolation valves has indicated that degradation has occasionally occurred in the leak tightness of the valves; therefore the special requirement for testing these valves.

The surveillance testing for measuring leakage rates is consi: tent with the requirements of Appendix J to 10 CFR 50 with the exception of exemption (s) granted for main steam isolation valve leak testing and testing the airlocks after each opaning.

3/4.6.1.3 PR1B RY CONTAINMENT AIR LOCKS [

The limitat?on on closure and leak rate for the primary containment air {

locks are required to meet the restrictions on PRIMARY CONTAINMENT _INI Aan a rimary containment leakage rate given in Specification 6 3.6.1.1 an G.6.1 .

The specification makes allowances for the fact that there may be I long periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each air lock is re u re o maint in e integrity of the containment.

3/% bJ,2 DELEVD hAI'b T B 3/4 6-1 Amendment No. 84 xl

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS p_ age B-8 INSERT H PRIMARY CONTAINMENT INTEGRITY is maintained by limiting overall integrated leakage to s 1.0 L, and the Type B and C combined leakage rate acceptance criterion is s 0.60 L,. Prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test, the combined Type B and C leakage must be < 0.60 L , and the overall Type A leakage must be < 0.75 L, when a ' a pe A test is scheduled.

Compliance with this LCO will ensure a primary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leal age rates assumed in the safety analyses.

The maximum allowable leakage rate for the primary containment (L ) is 0.635%

by weight of the containment atmosphere per day at the calculated maximum peak containment pressure (P,) of 39.6 psig.

Individual leakage rates specified for the primary containment air lock, main steam lines through the isolation valves, and valves in hydrostatically tested lines are addressed in LCO 3.6.1.3, and Surveillance Requirement 4.6.3.6.

Surveillance Requirement 4.6.1.1.b maintains PRIMARY CONTAINMENT INTEGRITY by requiring compliance with the visual examinations and leakage rate test requirements of 10 CFR 50, Appendix J, as modified by approved exemptions.

Failure to meet air lock leakage testing (4.6.1.3) or main steam isolation valve leakage (4.6.3.6.a) does not necessarily result in a failure of this Surveillance Requirement, 4m.1.1.b. The impact of the failure to meet these Surveillance Requirements 4.6.1.3 and 4.6.1.1.b must be evaluated against the Type A, B, and C acceptance criteria of 10 CFR 50, Appendix J, as modified by approved exemptions. The leakage limits for main steam lines through the isolation valves and leakage test results of Surveillance Requirement 4.6.3.6.a are not included in the total sum of Type B and C tests (approved exemption). As-left leakage prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test is required to be < 0.60 L, for combined Type B and C leakage, and ,< 0.75 L, for overall Type A leakage. At all other times between required Type A tests, the acceptance criteria is based on an overall Type A leakage limit of s 1.0 L,. At s 1.0 L, the offsite dose consequences are bounded by the assumptions of the safety analysis. The combined Type B and C leakage remains as s 0.60 L, between scheduled tests, in accordance with Appendix J.

The Frequency is required by 10 CFR 50, Appendix J, as modified by approved exemptions. Thus,4.0.2 (which allows Frequency extensions) does not apply to Surveillance Requirement 4.6.1.1.b.

l i

k : snia \lasalle\ 11rr fir a .wpf 50

t COWfAI M Dff SYST95-

- 5EE DEPIttpWTIZATION SYSTB5 (Continued) .- l Secause af the'large value and thermal capact1y of the sopression pool.

the volume and tagersture nemally changes very slowly and monitoring these parameters daily is sufficient to esta$1ish any tamperature trends. By requiring .

the suppression pool temperature ta be frequently recorded during periods of significant host addities, the tamperature tmnes will be closely followed so that appropriata action can be taken. The requirement for an external visual examine-time fo*1 ewing any event where potantially high loadings could accur provides assurance that no significant desage was encountered.

Is addition ta the limita en tamperaturs of the sworession chenhor pool water, operating procedures define the action ta he taken in the event a safety-relief valve inadvertantly opens or sticks open. As a sinimum this action shall includar.(1) use of all available seene to close the valve, (2) initista sppree-sien pool meter cooling, (3) initista reactor shutdeun, and (4) if other safety-relief yalves are used to depressurize the raector, their discharge shall be saperstad from that of the stuck-spen safety relief valve to assure sixing and uniformity of enery inserties'ta the poet. .

3/4.g.3 papens' CONTADpqBrr IptATION VALVEF f

The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Cantainment isolation within the time limits specified ensures that the release of radioactive material to the environment u

will be consistant with the assumptinns used in the analyses for a LOCA. j 3/4.6.4 'VACUUN RELIEF 0C'" INSERT .

Vacuus rslief breakers are provided to equalize the pressurs between the "

suppression chamber and drywell. This syntas will maintain the structural integ-rity of the primary containment under conditions of large diffemntial pressures.

The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the -

suppmssion pool in case of an accident. There are four valves to provide ,

redundancy so that operation may continue fer w to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with one vacum i breaker inoperable provided that the manual isolation valves on each side are in the closed position. .

l LA SALLE - LMIT 2 8 3/4 6-4 b

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSEffECHNICAL SPECIFICATIONS page B 9 INSERT _d Primary Containment Isolation Valves (PCIVs) form a part of the primary containment boundary. The PCIV safety function is related to control of primary containment leakage rates during accidents or other conditions to limit the untreated release of radioactive materials from the containment in excess of the design limits.

The automatic isolation valves are required to have isolation times within limits and actuate on an automatic isolation signal. The valves covered by this specification are listed with their associated stroke times, and other design information for lines penetrating the Primary Containment, in UFSAR Section 6.2.

The normally closed isolation valves are considered OPERABLE when manual valves are closed, automatic valves are de-activated and secured in their closed position, blind flanges are in place, and closed systems are intact.

Main steam lines through the isolation valves and hydrostatically tested valves must .neet alternative leakage rate requirements. Other PCIV leakage rates are addressed by specification 3/4.6.1.1, " PRIMARY CONTAINMENT INTEGRITY". UFSAR Section 6.2 also describes special leakage test requirements and exemptions.

This specification provides assurance that the PCIVs will perform their designed safety functions to control leakage from the primary containment during accidents.

The opening oflocked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the primary containment.

x :inla dualles 11rt fina.wpf 51

i e

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS page B-10 INSERT J (Continued)

Surveillance Regtu.ement 4.6.3.6.a verifies leakage through all four main steam lines is s 100 scfh when tested at 2 P,(25.0 psig). The transient and accident analyses are based on leakage at the specified leakage rate. The leakage rate for main steam lines through the isolation valves must be verified to be in accordance with the leakage test requirements of10 CFR 50, Appendix J, as modified by approved exemptions. A Note has been added to this Surveillance Requirement requiring the results to be excluded from the total of Type B and Type C tests.

This ensures that leakage rate for main steam lines through the isolation valves is properly accounted for in accordance with an approved exemption. The frequency

, is "at least once per 18 months" in accordance with an approved exemption.

Surveillance Requirement 4.6.3.6.b test of hydrostatically tested lines provides assurance that the assumptions .-f UFSAR Section 6.2 are met. The combined leakage rates must be demonstrated in accordance with the leakage rate test at a frequency of"at least once per 18 months". A Note has been added to this Surveillance Requirement requiring the results to be excluded the total of Type B and Type C tests. This is in accordance with 10 CFR 50, Appendix J, and approved exemptions.

l

'l k i t rdaUasallet11rt fina .wrf 52 l

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. ,- a c 12/16/83

!' ., Liceano Do. CPF-18

- 9-the BNR Owners Group Report SLI-8211,and SLI-8218 and the recomumendations of the BNR Owners Group reports. Any required modifications shall be completed on a scheduled acceptable to the NRC staff. *

^

d. Modiflemeinn of Aue A le nenreasurimmelon ses*== f.nele -

Famalh(11tv far fuerammad D1wermity for E- Ewant seen==n.m f n .r. 3. ia . men, men ai. mere as, mere an P'ior r to startup after the first refuellag outage, the licensee shall:

., (1) Install modif*. cations to the Automatic Depressurise ,

l tion system described la the licensee's letter dated July 1, 1983. The flaal circuit diagrams and an analysis of the bypass timer time delay shall be submitted for NRC staff review and approval prior to installation.

,, (11) . Incorporate into the Plant Ahaormal Procedures the V ==* a ' ** i=*** l * -i"* > ="

I#SERTL gitach6 (iii) Modify the Technical Specifications to provide the I bypass timer mammal inhibit switch.

. 6 D. Exemptions from certain regu., ts of Appendices G, E and J to 10 CFR Part 50, and to 10 CFR Part 7 are described la the Safety Evaluation Report and Supolement rsj l, 2, 3, and 5 to the Safety Evaluation Report. Qa addition, esemption was requested until ~

l comoletion of the first refuell'ag from the requirements of_10 CFR 70.2O( These exemptions are authorised by law and will.not endanger j life or property or the cosmon defense and security and are otherwise i

in the public interest. Therefore, these esemptions are granted pursuant to 10 CFR 50.12. With the granting of semptica$

the facility will operate to the extent authorised herein, a

%nformity with the application, as amended, the provisions of the

Act, and the rules and regulations of the Casumission.

E. Before engaging in additional construction er operational activities 1 I

which may result in a significant adverse environmental impact that k was not evaluated or that is significantly greater than that evaluated in the Final Environmental Statement and its Addendian, the licensee shall provide a written actification to the Director of the Office of Nuclear Reactor Regulation and receive written approval from that office before prer.wding with such activities.

0011k -

1

%e ?ac;hty regdees exedphns-Prom ce&

regulrernets d hcFR Partfo lo cFR Part70, 10CFR fd73: The;e Inclu]e: (a) -

ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS pace B-12 INSERT L (Unit 2, NPF-18)

(c) An exemption from the requirement of paragraph III.D of Appendix J to ,

conduct the third Type A test of each ten-year service period when the plant is shutdown for the 10-year plant inservice inspections. (d) A one-time exemption from the requirement of paragraph III.A.6(b) of Appendix J to resume a Type A test schedule of three times in ten years. Exemptions (c) and (d) are described in the safety evaluation accompanying Amendment No, to this license.

l 1

l ir s nlaV asalle\ilttfina.wrf54

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION 1

Commonwealth Edison has evaluated the proposed Technical Specification '

Amendment and determined that it does not represent a significant hazards consderation. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Units 1 and 2 in accordance with the proposed amendment will not:

1

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

a. The relocation of Technical Specification 3/4.6.1.2, Primary Containment Leakage, and Surveillance Requirements 4.6.1.1.a, 4.6.4.3, and 4.6.6.1.d to specification 3/4.6.1.1, Primary Containment Integrity, as Surveillance Requirement 4.6.1.1.b continues to assure that Primary Containment leakage is maintained within the analyzed limit assumed for accident analysis by testing in accordance with 10 CFR part 50, Appendix J as modified by approved exemptions.

The requirement to be less than 0.75 L, for as-left Type A test and less than 0.60 L, for Type B and C tests prior to first unit startup following testing performed in accordance with 10 CFR part 50, Appendix J, as modified by approved exemptions, provides margin for degradation between tests and thus primary containment integrity is maintained during the time period between required leakage testing. The current Limiting Condition for Operation 3.6.1.2 in conjunction with Surveillance Requirements 4.6.1.2 basically require the same leakage limits as proposed Surveillance Requirement 4.6.1.1.b. The Limiting Condition for Operation (LCO) is required to be less than 1.0 L, and is applicable during a fuel cycle for the Type A test. The LCO for Type B and C combined leakage total is currently required to be less than 0.60 L,. The proposed Surveillance Requirement maintains the following:

1. The current LCO for Overall Containment leakage (as determined by a Type A test) and for the Type B and C combined leakage during the cycle by requiring overall containment leakage to be less than 1.0 L, and Type B and C leakage ti tal less than 0.60 L,.

x :inlas la sa llw\11 rt f ina.wyf %

ATTACHMENT C SIGNIFICANT IIAZARDS CONSIDERATION

2. The associated 6 nits specified in the current Action Statements are maintained by verifying Overall Containment leakage to be less than 0.75 L, and Type B and C leakage total less than 0.60 L, prior to startup from an outage in which the applicable leakage testing is conducted, t

Therefore, there is no change to the consequences of an accident previously evaluated, because maintaining leakage within the analyzed limit assumed for accident analysis does not change either the onsite or offsite dose consequences resulting from an accident. In addition to this, containment leakage is not an accident initiator, so there is no effect on the probability of accident initiators. Thus there is no increase in the probability of an accident previously analyzed,

b. Helocation of Technical Specification table of Primary Containment Isolation Valves, Table 3.6.3-1, to the LaSalle UFSAR is an administrative change to remove the component list of Primary Containment Isolation Valves, Table 3.6.3-1, from the Technical Specifications. The Limiting Condition for Operation (LCO),3.6.3, is being revised to define which components the LCO applies to. The wording of the revised LCO encompasses all of the components listed in the current Technical Specification Table 3.6.3. Removal of this component list does not change the probability of any accident initiators or change any other relevant initial assumptions. Also, there il no change to the consequences of an accident previously evaluated, because removing this list from Technical Specifications does not change either the onsite or offsite dose consequences resulting from the event. The component list will be controlled by an Administrative Procedure and can only be changed by the 10 CFR 50.59 change process with review and approval per the Onsite Review and Investigative Function.

Therefbre, there is no increase in either the probability or consequences of an accident previously evaluated.

k:* in.tIesalledilttf2na w fte

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION

c. The change in the functional test interval for the Drywell and Suppression Chamber Hydrogen Recombiner systems from "once per 6 months" to "once per 18 months" was determined by the NRC in NUREG 13G6 and Generic Letter 93-05 to be acceptable by evaluation of the industry Licensing Event Reports (LERs) to assess the reliability of

. hydrogen recombiners. The conclusion was that the interval should bc ,

changed, because of the redundancy and apparent high reliability. A '

review 'of LaSalle LERa hu shown only one LER that involved the operability of the hydrogen recombiner system and that was due to a Part 21 issue regarding circuit breaker environmental qualification. The breakers were replaced with qualified breakers. Therefore, the LaSalle Hydrogen Recombiner reliability is consistent with or better than that found by the NRC in determining this surveillance interval extension based on all LERs. Also, redundancy is the same as that assumed by the NRC; because, LaSalle has two hydrogen recombiner subsystems that are shared by Unit 1 and Unit 2. Both hydrogen recombiners subsystems are required to be Operable for either or both units in Operational Conditions 1 and 2. Based on LaSalle operating experience, the hydrogen recombiner subsystems are expected to continue to be demonstrated operable when the functional test is performed at an 18 month frequency.

Therefore, there is minimal or no change to the consequences of an accident previously evaluated, because at least one of the hydrogen recombiner subsystems is expect.ed to be available to meet its design ,

function to reduce the potential for hydrogen explosion or hydrogen burn in the primary containment. By preserving the integrity of the primary containment, there is no change to either the onsite or offsite dose consequences resulting from an accident. In addition to this, control of hydrogen concentration by use of a hydrogen recombiner subsystem is not an accident initiator, so there is no effect on the probability of l accident initiators. Thus there is no significant increase in the probability of an accident previously analyzed.

I k : \nla s iasalle u lrt f ina .wpf S7

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION

d. The first exemption request is from the requirements of paragraph III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to return to or resume a Type A test schedule of three times in ten years (40 i10 months). Due to consecutive failures,10 CFR 50 Appendix J requires that Type A tests be performed every refueling outage on Unit Two until two consecutive Type A tests are satisfactory.10 CFR Part 50 ;

has an exemption process and is specified in 10 CFR Part 50.12(a),

which states:

"The Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of this part,..."

The exemption process requires showing that the granting of the exemption is authorized by law, will not present an undue risle to the public health and safety, and is consistent with the common deiense and security. Also, special circumstances are required to be present for the granting of an exemption. One of the special circumstances that would apply in ibis instance is 10 CFR part 50.12(a)(2)(ii) which statt

" Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule" This requires that it be shown that unacceptable containment leakage will be identified and corrected, by alternative methods. The alternative method is specifically Type B and C tests, which willidentify any local penetration leakage. This is acceptable, because Type C test failures i have been the cause for failures of as-found Type A tests in the LaSalle Unit 2 first, third, and fburth refueling outages.

Exceeding the allowable leakage rate during the performance of the Type I

A test is indicative of either a passive or a structural component that is leaking or that there is an inadequacy in the Local Leak Rate Test (Type B and C tests) program. When the failure of a Type A test is due to a passive or structural component, the only test for adequate repair would 1 be the Type A test. For a Local Leak Rate Test program inadequacy, the k \nla;1asalicillrtfina.wif58 1

1

l ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Type A test would serve as a means of verification of the results of the test program. The Type A tests have not found new significant Type B or C tested local penetration leakage that has not been identified by Type B or C testing alone. Therefore, the LaSalle Local Leak Rate Test program is adequate to find and correct Type B and C containment

, penetration leakage. 2 When it is determined that Type A tests failed as a direct result of as-found Type B and C minimum path leakage penalty additions and not due to a non Type B or C tested components or structures, then performance of the Type A test more frequently as required by 10 CFR Part 50, Appendix J, due only to Type B and C test failures is redundant to the performance of Type B and C tests. Therefore, Type B or C tested penetration leakage that can be determined by Type B or C tests is evaluated and corrected, as applicable, to maintain overall containment leakage within limits, without an additional Type A test.

Primary Containment leakage which includes the minimum path Primary Containment Isolation Valve leakage is an assumption in any analyzed accident which could involve an offsite radioactive release.

Because performance of Type B and C tests will find and allow correction / repair ofleaking valves / penetrations, verification of as-found and as-left local leakage assures that Primary Containment leakage will be within the analyzed limit assumed for accident analysis.

Therefore, for this one-time exemption for LaSalle Unit 2, there is little or no increase in the consequences of an accident previously evaluated involving the dose previously calculated either onsite or offsite at the site boundary due to any analyzed accident. In addition to this, containment leakage is not an accident initiator, so there is no effect on the probability of accident initiators. Thus there is no significant increase in the probability of an accident previously analyzed, un aumun u n n ne m

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SIGNIFICANT HAZARDS CONNDERATION

e. The request for a partial exemption from paragraph III.D of Appendix J to 10 CFR 50 involves a deletion of the requirement to perform the third ;

Type A test for each 10-rmr service period during the shutdown for tne 10-year plant inservice i ;oections. There is no significant benefit in coupling these two sun mees (i.e., the Type A test and the'10-year. !

ISI program). . Each of . ..two surveillances.is independent of the other i and provides assurance ef different plant characteristics. The Type A l' test assures the required leak-tightness for the reactor containment building be less than Appendix J acceptance criteria. This demonstrates compliance with the guidelines of 10 CFR Part 100 based on the assumptions used in the UFSAR which conform to NRC Safety Guide 4. ,

The 10-year ISI program provides aesurance of the integrity of the plant structures, systems, and components in compliance with 10 CFR 50.55(a). There is no safety-related concern necessitating their ,

coupling to the same refueling outage. As a result, this change cannot l increase the consequences (i.e., offsite dose) of any accident previously '

evaluated. Furthermore, since the decoupling of the test schedules has j no affect on the test's effectiveness, decoupling their schedules will not ;

increase the probability of an accident. l i

?

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

i

a. Technical Specification 3/4.6.1.2, Primary Containment Leakage, and .;

Surveillance Requirements 4.6.1.1.a,4.6.4.3, and 4.6.' ..d are being relocated to specification 3.4.6.1.1, Primary Containment Integrity, as ;

Surveillance Requirement 4.6.1.1.b. The proposed Surveillance l Requirement 4.6.1.1.b assures that Primary Containment leakage is maintained within the analyzed limit assumed for accident analysis by .

testing in accordance with 10 CFR part 50, Appendix J as modified by I approved exemptions. Primary containment leakage is an assumption in accident analyses, and is maintained by both the current specifications i snd the proposed specification. The leakage does not cause an accident ;

and no new failure modes are created. Therefore this request for :

exemption does not create the possibility of a new or differe'nt kind of accident from any accident previously evaluated. !

I h

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i

l l

ATTACHMENT C -

SIGNIFICANT HAZARDS CONSIDERATION l

b. This is an administrative change to control the list of Primary Containment Isolation Valves outside the LaSalle Unit 1 and Unit 2 !

Technical Specifications. The administrative controls provided to control i this component list assure that the design and operation of the plant will continue to be in accordance with the UFSAR, Facility License and the

. associated Technical Specifications. Therefore, the possibility of a new or :

different kind of accident from any previously evaluated is not created. .

c. The change in the functional test interval for the Drywell and Suppression Chamber 1lydrogen Recombiner systems from "once per 6 months" to "once per 18 months" is based on good equipment i performance on a 6 month frequency. The expected outcome of the 18 month surveillances, based on the low failure rate at a six month frequency, is to show the hydrogen recombiner subsystems Operable.

This system is for mitigating the consequences of an accident that causes generation of hydrogen and oxygen in the primary containment. No new failure modes are created by this change in surveillance frequency.

Therefore, the possibility of a new or different kind of accident from any previously evaluated is not created.

d. The first exemption is from the requirements of paragraph III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to return to or resume a Type A test schedule of three times in ten years (40 210 !

months). Containment leakage testing, including both Type B and C testing and Type A testing as specified in the LaSalle County Station Safety Analysis Report were evaluated in Section 6.2.6 of Safety Evaluation Report, NUREG-0519, and found to be acceptable. Since Type B and C testing will find and verify correction of penetration leakage when Type B and C test as-found penalties are specifically what caused the failure of the as-found Type A tests, then Type B and C testing will provide adequate assurance of the continued integrity of the Primary Containment without increasing the frequency of Type A tests.

As a result, the Primary Containment will continue be maintained as designed and previously evaluated.

i k: Mla\ lan.alle\ilrt fina.wrf 61

.=.

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION Based on this, the requirement of two acceptable as-found Type A tests !

prior to returning to the Appendix J paragraph III.D frequency of three times in ten years (40 10 months) is not necessary to assure that the primary containment remains within the analyzed leakage limits. !

Containment leakage is an assumption for the dose consequences of

- accident analyses, and not an accident initiator. . Also, no new failure modes are created by this exemption. Therefore this Amendment does I not create the possibility of a new or different kind of accident. l l

e. The request for a partial exemption from paragraph III.D of Appendix J l to 10 CFR 50 involves a deletion of the requirement to perform the third l Type A test for each 10-year service period during the shutdown for the ;

10-year plant inservice inspections. The proposed exemption does not ,

involve any change to the plant design or operation. As discussed above, j this change cannot increase the consequences of any accident previously i evaluated. As a result, no new failure modes are created. Therefore, !

this proposed change cannot create the possibility of any new of different i kind of accident from any accident previously evaluated. l

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

a. Technical Specification 3/4.6.1.2, Primary Containment Leakage, and Surveillance Requirements 4.6.1.1.a, 4.6.4.3, and 4.6.6.1.d are being relocated to specification 3/4.6.1.1, Primary Containment Integrity, as proposed Surveillance Requirement 4.6.1.1.b. The proposed Survillance Requirement 4.6.1.1.b continues to assure that Primary Containment leakage is maintained within the analyzed limit assumed for accident analysis by testing in accordance with 10 CFR part 50, Appendix J as modified by approved exemptions.

As stated. in 1)a. above, the proposed Surveillance Requirement 4.6.1.1.b maintains the acceptance criteria and limits for continued operation of the current specification for primary containment leakage. Therefore, the margin of safety is not reduced by this change. Also, the proposed k : i rila s lasalle s il rt f ina .wr f 6 2

1 ATTACHMENT C ,

SIGNIFICANT HAZARDS CONSIDERATION I l

addition of a definition for the maximum allowable primary containment l leakage rate assures that the margin of safety is maintained. I The leakage limits for MSIVs and hydrostatically tested valves are l maintained by relocating the current surveillance requirements to specification 3/4.6.3, with the acceptance criteria of the current.

specification retained. Thus preserving the current margin of safety by maintaining the leakage rates as assumed in the accident analyses.

I

b. The Limiting Condition for Operation for Technical Specification 3.6.3, Primary Containment Isolation Valves, is revised by this Technical Specification change to specifically define the components to which the LCO applies. Therefore, removal of Technical Specification Table 3.6.3-1, which lists the specific components to which the LCO applies does not change the scope or applicability of the specification. The component list will be controlled administratively with any changes to the list made in accordance with the 10 CFR 50.59 change process. Therefore, this is an administrative change only and there is no reduction in the margin of j safety.

c. The change in the functional test interval for the Drywell and .

Suppression Chamber Hydrogen Recombiner systems from "once per 6 !

months" to "once per 18 months" is based on good equipment ,

performance on a 6 month frequency. The expected outcome of the 18 i month surveillances, based on the low failure rate at a six month i frequency, is to show the hydrogen recombiner subsystems Operable.

The change in frequency has no affect on the hydrogen or oxygen generation assumptions or the recombination rate of the hydrogen recombiner subsystems. Therefore, the margin of safety is not reduced or changed by this surveillance interval change.

d. The first exemption is from the requirements of paragraph III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to return to or resume a Type A test schedule of three times in ten years (40 10 months). The limit of total leakage determined from Type B and C tests will remain the same, providing a margin of 40 percent to the maximum k:\nlaslasalle\ilrtfIna.mTff3 l

1 1

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION allowable containment leakage rate (L,) at the design basis accident pressure specified in proposed Technical Specification definition of L,.

This 40 percent is as specified by 10 CFR Part 50, Appendix J. In ;

addition to this, administrative guidelines have been set for eachpenetration/ valve, so that any abnormal leakage will be corrected

. by adjustment or. repair as needed. Any postponement of repairs is !

based on a technical evaluation and then only if the total Type B and Type C leakage is maintained at less than 0.60 L,. Repairs will be ;

required to restore the leakage rate to less than the administrative limit at the next refueling outage.

This request for exemption is based the fact that Type B and C testing minimum path leakage rate penalties are the direct cause of the failure of as-found Type A tests. The leakage through Type B and C tested !

penetrations is best measured and corrected via a local leak test.

Therefore, verification of an adequate margin of safety is assured by conducting Type B and C tests, and not another increased frequency Type A test.

c. The request for a partial exemption from paragraph III.D of Appendix J to 10 CFR 50 involves a deletion of the requirement to perform the third Type A test for each 10-year service period during the shutdown for the 10-year plant inservice inspections. The proposed exemption does not change the acceptance criteria that must be met for inservice ,

inspections, does not relax the condition of containment that must be met prior to plant restart, and does not change the requirements that must be met between plant refueling outages. Therefore, the proposed change does not result in a significant reduction in the margin of safety.

Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule,51 FR 7744, for the application of standards j to license change requests for determination of the existence of significant hazards i considerations. This document provides examples of amendments which are and are not considered likely to involve significant hazards considerations.

k : \nia\inealle\ ri rt f a na.wpf 04 l

l

E ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION

a. The relocation of Technical Specification 3/4.6.1.2, Primary Containment ;

Leakage, and Surveillance Requirements 4.6.1.1.a,4.6.4.3, and 4.6.6.1.d to specification 3/4.6.1.1, Primary Containment Integrity, as Surveillance Requirement 4.6.1.1.b most closely fits the example of a change which may either result in some increase to the probability or consequences of a previously. analyzed accident or may reduce in some way a safety margin, but I where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in Standard Review Plan 6.2.6, .

Containment Leakage Testing. The proposed surveillance requirements and !

definition retain the current specification limits and acceptance criteria, thus preserving the safety margin and will not increase the consequences of an accident. ,

b. These proposed amendment for relocation of Table 3.6.3-1 outside of Technical Specifications most closely fits the example of an administrative change to the Technical Specifications. The Limiting Condition for Operation for the affected specification is being revised to clearly define the components '

to which the specification applies, so the Technical Specificati n Table which lists the components is redundant to the revised LCO and can be removed without changing the application or interpretation of the LCO. Likewise, other references in Technical Specifications to Table 3.6.3-1 are being ,

changed to clearly define the components to which the specifications apply.

Any changes to the component list after removal from Technical Specifications will be performed under Administrative control and in accordance with the 50.59 change process, which will assure compliance with '

the Technical Specification.

c. The change in the functional test interval fo" the Drywell and Suppression l Chamber Hydrogen Recombiner systems fro:n "once per 6 months" to "once l per 18 months" most closely fits the exampl 3 of a change which may either l result in some increase to the probability or ensequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan. The system design is not changed, and therefore the acceptability of the subsystem design is not changed. The frequency extension is consistent with the reliability of the equipment, and thus the hydrogen recombination subsystems will be verified to function as designed by functional testing on an l 18 month frequency. Therefore, this functional test interval increase is consistent with Standard review Plan 6.2.5, Combustible Gas Control in Containment.

k:\nlailasalle\ilrtfina.wpf65

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION

d. The first exemption is from the requirements of paragraph III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to return to or resume a Type A test schedule of three times in ten years (40 10 months). This proposed exemption most closely fits the example of a change which may either result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan.

When as-found Type A tests fail as a direct result of Type B and C test minimum path leakage penalties, then Type B and C testing in conjunction with the corrective and preventative maintenance for penetrations determined to be poor performers through a comprehensive corrective action plan will assure that the Primary Containment integrity will be maintained without additional Type A tests. Based on this, allowing exemption to the requirement of two acceptable as-found Type A tests prior to returning to the

. Appendix J paragraph III.D frequency of three times in ten years (40 10 months)is not necessary to assure that the primary containment remains within the analyzed leakage limits. The Standard Review Plan section 6.2.6 basically verifies Containment leak rate testing is in conformance to 10 CFR 50 Appendix J. The above provides confidence that the requested exemption assures adequate Containment leak rate testing will be conducted to maintain overall containment leakage within all acceptable criteria.

e. The request for a partial exemption from paragraph III.D of Appendix J to 10 CFR 50 involving a deletion of the requirement to perform the third Type A test for each 10-year service period during the shutdown for the 10-year plant inservice inspections most closely fits the example of a change which I may either result in some increase to the probability or consequences of a previously analyzed accident or may reduce in some way a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in Standard Review Plan 6.2.6, Contaimnent Leakage Testing. The conduct of Type A tests in accordance k: ula\lasalle\ilttfins.wpf66 i

J

l ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION l

with Appendix J at a frequency of three times in ten years (40 10 months) provides assurance of ongoing acceptable overall integrated primary containment leakage without a test coinciding with the last outage of the ten year Inservice Inspection program.

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.

l i

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ATTACHMENT D EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE l' TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES ;

BACKGROUND The primary objective of the regulations is to assure continuance of primary containment leak-tight integrity by.the conduct of Type A, B, and C tests on a i periodic basis. 10 CFR 50 Appendix J establishes two major categories of tests with separate criteria. The Type B and C tests, Local Leak Rate Tests (LLRTs),

are performed during each refueling outage while the Type A test, Containment Integrated Leak Rate Test (CILRT), is only performed every 40110 months to achieve three Type A tests in a ten year period. The Type B and C tests provide periodic surveillances of components such as isolation valves and resilient seals.

The Type A test provides a measure of the overall integrated leakage rate of the containment, including passive and structural components.

Exceeding the allowable leakage rate during the performance of the Type A test is indicative of either a passive or a structural component that is leaking, or an inadequacy in the local leak rate test program. When the failure of a Type A test is due to a passive or structural component, the only test for adequate repair would be the Type A test. If Type B and C leakage rates constitute an identified contributor to a Type A test failure, then as-left leakage iates are best determined by the associated Type B or Type C test.

However,10 CFR 50 Appendix J, Section IILA 6.(b), requires an increase in the frequency of Type A tests irregardless of the cause of two or more consecutive failures of as-found Type A tests.

LaSalle County Station (LaSalle) Unit 2 has experienced Containment Integrated Leak Rate Test (CILRT) (Type A test) failures for the "as-found" condition at the first, third and fourth refueling outages as a result of penalties from Local Leak Rate Test (LLRT)(Type B and C) failures. Therefore the requirements of 10 CFR 50 Appendix J, Section III.A.6(b) are applicable, necessitating that a Type A test be performed at both the fifth refueling outage (L2R05) and the sixth refueling outage (L2R06) due to consecutive "as-found" Type A test failures. The Type A test performed at the end of L2R05 was acceptable in the as-found condition.

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l ATTACHMENT D l EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE i TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES BACKGROUND (continued)

Appendix J establishes two types of tests with separate criteria. The Local Leak Rate Tests (LLRT)(Type B and C) are performed during each refueling outage while the frequency of the Type A test is as follows per 10 CFR 50, Appendix J section IILD.1.(a) Periodic retest schedule:

"(a) After the preoperational leakage rate tests, a set of three Type A tests shall be performed, at approximately equal intervals during each 10-year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year plant inservice inspections."

Likewise, the current LaSalle Unit 2 Technical Specification Surveillance Requirement 4.6.1.2.a restates the Appendix J requirement:

Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40 10 month intervals during shutdown at P,,39 6 psig, during each 10-year service period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection.

Therefore, Type A tests are performed every three or four years (40 10 months) to achieve three Type A tests in ten years, with the third Type A test coinciding with the tenth year of each ten year inservice inspection interval. The Type B and C tests provide periodic surveillances of components such as isolation valves and resilient seals. The Type A test provides a measure of the overall integrated leakage rate of the containment including testing of passive and structural components, thereby verifying the overall leakage integrity of the primary ;

containment. !

Based on IE Information Notice 85-71 dated August 22,1985 and an exemption granted to Iowa Electric Light and Power Company for the Duane Arnold Energy Center, the following discussion and attachments provide justification for the granting of a similar exemption to Commonwealth Edison Company for LaSalle County Station Unit Two. Upon approval, this exemption will allow LaSalle Unit 2 to return to a frequency of three times in ten years (40 *10 months).

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ATTACHMENT D EXEMPTION REQUEST FROM APPENDIX J RF,QUIREMENTS FOR THE TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES BACKGROUND (continued)

The granting of an exemption is contingent upon the Commission's approval of proposed Technical Specification amendments for LaSalle County Station Unit 1 and Unit 2. The proposed amendments are being submitted in conjunction with this exemption request.

DISCUSSION Pursuant to 10 CFR Part 50.12(a), Commonwealth Edison Company (CECO),

requests a one-time exemption from the requirement to conduct an additional Type A test) at LaSalle County Station, Unit Two as required by 10 CFR Part 50, Appendix J, Section III.A.6(b) which states, "If two consecutive periodic Type A tests fail to meet the applicable acceptance criteria in III.A.5.(b), not withstanding the periodic retest schedule of III.D., a Type A test shall be performed at each shutdown for refueling or approximately every 18 months, whichever occurs first, until two consecutive Type A tests meet the acceptance criteria in III.A.5.(b), after which time the retest schedule specified in III.D may be resumed."

Likewise, LaSalle Unit 1 and 2 Technical Specifications 4.6.1.2.b require the following:

"If any periodic Type A test fails to meet 0.75 L,, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission.

If tw, consecutive Type A tests fail to meet 0.75 L,, a Type A test shall be performed at least every 18 months until two consecutive Type A tests meet 0.75 L,, at which time the above test schedule may be resumed."

This exemption is requested to allow LaSalle to return to or resume a Type A test schedule of three times in ten years (40 10 months), as specified in 10 CFR Part 50, Appendix J, Section III.D.

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l ATTAC.HMENT D EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES DISCUSSION (continued)

Appendix J establishes two types of tests with separate criteria. The Local Leak Rate Tests (LLRT) (Type B and C) are performed during each refueling optage :

while the Type A test is only performed every three to four years to achieve three Type A tests in a ten year period. The Type B and C tests provide periodic surveillances of components such as isolation valves and resilient seals. The Type A test provides a measure of the overall integrated leakage rate of the containment, including passive and structural components.

Exceeding the allowable leakage rate during the performance of the Type A test is indicative of either a passive or a structural component that is leaking, or an inadequacy in the local leak rate test program. When the failure of a Type A test is due to a passive or structural component, the only test for adequate copair would be the Type A test.

The LaSalle Unit 2 Type A tests for the "as-found" condition at the first, third and fourth refueling outages failed as a result of as-found penalties from Type B and C test minimum path leakage penalty additions. The failures were not due to a non-local leak rate tested component or structure. As a result of the as-found Type A test failures,10 CFR 50 Appendix J, Section III.A.6(b) requires a Type A test at 18 month frequencies until two consecutive tests are less than 0.75 L,.

The first of the increased frequency Type A tests was performed during the LaSalle Unit 2 fourth refueling outage (L2R04) and was unacceptable. The second of the increased frequency Type A tests was performed during the LaSalle Unit 2 ;

fifth refueling outage (L2R05) and was acceptable. The third of the increased frequency Type A tests is due to be performed during the upcoming sixth refueling outage (L2R06).

The third LaSalle Unit 2 Type A test, performed in the fourth refueling outage, L2R04, exceeded the 0.75 L, limit, yet remained below the 1.0 L, limit. This !

failure was overwhelmingly due to a local failure of one penetration with two containment isolation valves in the reactor water cleanup system. The leakage of these valves accounted for more than 72% of the Type B and C test penalties which were applied to the "as-found" Type A test results.

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ATTACHMENT D EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES DISCUSSION (continued)

The LaSalle Station L2R04 as-found Type A test did not exceed the 1.0 L, criteria.

This was a non-reportable event even though the 0.75 L, criteria was exceeded, i due to the current LaSalle Technical Specification Limiting Condition for ,

Operation 3.6.1.2. ofleakage less than L,, not less than 0.75 L,. Thus, the design leakage for the primary containment was not exceeded during the previous 18 month cycle.

The following table references the Licensee Event Reports (LER's) and " Reactor Containment Building Integrated Leak Rate Test" for L2R04 (References v) documenting these failures.

H.EFUELING OUTAGE DATE REPORT NO.

1,L2R01 January,1987 LER 50-374-87-002, Revision 1 3,L2R03 March,1990 LER 50-374-90-004, Revision 1 4,L2R04 March,1992 See Reference v.

LaSalle is conducting an aggressive Corrective Action Plan to directly address and eliminate the local sources ofleakage as an alternative to conducting an increased frequency of Type A testing. In lieu of this, continuing to perform a Type A test at a frequency greater than 3 times in 10 years, due solely to Type B and C test failures, is only redundant to the performance of Type B and C tests. There is little or no benefit to be gained from performing a Type A test at an increased frequency.

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ATTACHMENT D l

EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES DISCUSSION (continued)

In support of this are the following:

1. IE Information Notice 85-71 dated August 22,1985 states in part,

" .. if Type B and C leakage rates constitute an identified contributor !

to this failure of the "as-found" condition for the Type A test, the l general purpose of maintaining a high degree of containment ;

integrity might be better served through an improved maintenance and testing program for containment penetration boundaries and isolation valves. In this situation, the licensee may submit a Corrective Action Plan with an alternative leakage test program proposal as an exemption request for NRC staff review. If this submittal is approved by the NRC staff, the licensee may implement the corrective action and alternative leakage test program in lieu of the required increase in Type A test frequency incurred after the failure of two successive Type A tests."

2. 10 CFR Part 50.12(a) indicates that the Commission may grant exemptions if the exemption will not present undue risk to public health and safety, is consistent with the common defense and security and special circumstances are present. l An exemption is authorized by law because only the regulation requires the test and the NRC is authorized to grant exemptions from its regulations.

No undue risk to public health and safety would result from this exemption j because local leakage will be corrected in accordance with the Corrective l Action Plan which is described in Attachment B. The common defense and l security is not afTected by this exemption. Therefore the exemption :

conditions in 10 CFR 50.12(a)(1) are completely satisfied.

One of the special circumstances presented in Part 50.12(a)(2)(ii)is:

" Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule."

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1 ATTACHMENT D :

EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE l TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE l FAILURES !

p_ISCUSSION (continued)

The underlying purpose of 10 CFR 50 Appendix J, Section III.A.6(b) is to ensure that unacceptable containment leakage is identified and corrected, i As described herein, performance of a Type A test at the next refueling ,

outage is not necessary to satisfy this purpose. Therefore :

10 CFR 50.12(a (2)(ii) is satisfied.

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3. Iowa Electric Light and Power Company requested an exemption to the same retest requirements of 10 CFR 50 Appendix J in a letter dated April 2, 1990 fbr the Duane Arnold Energy Center. Their Corrective Action Plan involved the repair, modification, maintenance, and special testing of the Main Steam Isolation and Feedwater Check Valves, which were the significant contributors to the as-found Type A test failures due to excess leakage through these valves. The initial problems experienced with LaSalle's Main Steam Isolation and Feedwater Check Valves were corrected through applicable procedure changes and/or modifications. Attachment F,

" Corrective Action Plan for Type C Test Failures Contributing to "As-Found" Type A Test Failures", includes similar successful or planned corrective actions for specific valves that have been major contributors to as-found Type A test fhilures. The Corrective Action Plan is directed at improving the overall Type B and C test program based upon enhanced trending and engineering evaluation.

4. 10 CFR Part 50.12(aX2)(iv) identifies as another special circumstance:

"The exemption would result in a benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemption."

Exemption from the requirements to perform a Type A test at the next refueling outage will prevent a radiological exposure of approximately 3 manrem to plant personnel associated with performing a Type A test. This health and safety benefit helps compensate for any perceived decrease in safety that may result from granting the requested exemption.

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ATTACHMENT D EXEMPTION . REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES DISCUSSION (continued)

5. There is also a time and cost benefit to not performing a Type A test more frequently. The test would add two to four days (scheduled for 3.5 days) to i the end of the sixth LaSalle Unit Two refueling outage (L2R06). The Engineering Staff, Operating, Maintenance, and Radiation Protection departments would need to devote approximately 1000 man-hours of work to prepare for, perform, and recover from this major test.

6. This request for a one-time exemption from Appendix J also merits consideration, because:

a. Type B and C tests are better at detecting both penetration seal leakage and containment isolation valve leakage.

b. The third Type A test failure was non-reportable since the 1.0 L, criteria was not exceeded. This failure was shown to be almost wholly attributable to the local failure of the reactor water cleanup suction isolation valves. The Reactor Water Cleanup Suction containment isolation valves were replaced with double-disc gate valves prior to startup from L2R04.

LaSalle is addressing the concern of excessive leakage found during Type C Local Leak Rate Testing through an aggressive Corrective Action Plan. This plan includes an update of the corrective actions taken or scheduled to be done for the valves specified in Attachment F, using the guidance given in IE Information )

Notice 85-71. The Corrective Action Plan in summary addresses. ;

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1. Specific corrective actions completed or planned on specific valves identified j as problems thus far. .

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2. An Alternative Leakage Test Program.

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ATTACHMENT D EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES i DISCUSSION (continued)

3. Development and implementation of an improved trending program for leakage rate. performance, comparing valve type, service, and manufacturer i for leakage rate performance comparisons.

4. Recommendation and implementation based on engineering evaluation for ;

test method improvement or repair, and modification or replacement of l problem Primary Containment Isolation valves identified in 3. above. l

SUMMARY

LaSalle is presently required by 10 CFR 50, Appendix J, Section HI.A.6.(b)(and the current Technical Specification Surveillance Requirement 4.6.1.2.b.) to conduct a Type A test on Unit Two during the next refueling outage (L2R06) scheduled to begin in February,1995. This Type A test will be performed unless an exemption from this requirement (and approval of the associated proposed amendment to LaSalle Unit 1 and 2 Technical Specifications) is granted. The granting of an exemption to Iowa Electric Light and Power Company's Duane Arnold Energy Center (June 29, 1990) has established precedence in this matter.

LaSalle's extensive Corrective Action Plan proposed in lieu of the additional Type A testing includes an augmented Local Leak Rate Test program similar to that of Duane Arnold's. The attached Corrective Action Plan follows the guidance in Information Notice 85-71, assuring that Primary Containment integrity can be maintained without increasing the frequency of Type A tests. The plan not only better implements the safety purpose of the rule, but it also avoids unnecessary occupational radiation exposure to station personnel.

The third Unit 2 Type A test failure remained below the reportable limit of 1.0 L, !

even though 0.75 L, was exceeded. This was overwhelmingly due to a local failure of two containment isolation valves in the same penetration in the Reactor Water Cleanup System. The leakage of these valves accounted for the majority of the Type B and C test penalties which were applied to the "as-found" Type A test results.

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1 ATTACHMENT D EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A INCREASED FREQUENCY AS A RESULT OF CONSECUTIVE FAILURES

SUMMARY

(continued)

IE Information Notice 85-71 dated August 22,1985 addresses alternatives to increased frequency Type A testing such as correctivo action plans as well as an C increased frequency of Type B and C testing. The replacement of the two troublesome Reactor Water Cleanup Suction containment isolation valves combined with the increased frequency of Type B and C testing is consistent with the Commission's philosophy regarding alternatives to Type A testing.

CONCLUSION LaSalle Station's Type A test one-time exemption request merits consideration, since the as-found Type A test was less than 1.0 L,, the current LCO, for L2R04 and the as-found Type A test was less than 0.75 L, for L2R05. Granting LaSalle a one-time exemption from this testing is consistent with NRC requirements and practice. LaSalle meets the requirements for an exemption from the need for an additional Type A test as previously demonstrated. If the Type A test scheduled to be performed in the LaSalle Unit 2 seventh refueling outage, L2R07, is unacceptable in the as-found condition per 10 CFR 50, Appendix J, then the frequency will be adjusted in accordance with Appendix J as if there were 2 consecutive failures. A Type A test summary and proposed Type A test schedule is included as Attachment G.

It is requested that this exemption request be reviewed and approved for implementation by March 15,1995. This would serve to avoid delays and additional work during the LaSalle Unit Two sixth refueling outage scheduled in February of 1995.

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ATTACHMENT E EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A NORMAL TEST SCHEDULE, SEPARATING THE TYPE A TEST SCHEDULE FROM THE INSERVICE INSPECTION SCHEDULE BACKGROUND In order to ensure offsite doses remain below those previously evaluated in the event of a design basis accident, leakage from the primary containment must be limited. To ensure that containment leakage remains within these limits, periodic leakage rate tests are performed. Specifically,10 CFR 50.54(o) requires primary reactor containments for water cooled power reactors to be subject to the leakage rate testing requirements set forth in Appendix J to 10 CFR 50. LaSalle County Station Units 1 and 2 (LaSalle) Technical Specifications 3/4.6.1.2, " Primary Containment Leakage," provides additional requirements for performing leakage rate testing and specifies the associated limits. The leakage rate testing program for LaSalle is described in Section 6 of the LaSalle Updated Final Safety Analysis Report (UFSAR).

LaSalle is requesting an exemption from paragraph III.A.6.(b) to return to a Type A test frequency of three times in ten years, and thus not perform a second increased frequency Type A test in the upcoming Unit 2 sixth refueling outage, L2R06. The exemption request, discussed in Attachment D of this submittal, necessitates that LaSalle request partial exemption from 10 CFR 50 Appendix J in accordance with 10 CFR 50.12, because L2R06 is the last refueling outage in the first ten year Inservice Inspection (ISI) interval for Unit 2.

The attached request, for a one time exemption from the increased Type A test frequency requirements of Appendix J section III.A.6(b), requires a change to the Technical Specification Surveillance Requirements of specification 3/4.6.1.2, l Primary Containment Leakage. Likewise, the permanent exemption to decouple the normal frequency of 3 times in 10 years (40 10 months) from the Inservice Inspection Schedule (ISI) requires a change to the Technical Specification Surveillance Requirements of specification 3/4.6.1.2, Primary Containment Leakage.

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ATTACHMENT E EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A NORMAL TEST SCHEDULE, SEPARATING THE TYPE A TEST SCHEDULE FROM THE INSERVICE INSPECTION SCHEDULE BACKGROUND (continued)

Both exemption requests require a change to Technical Specifications 3/4.6.1.2, because the 4.6.1.2 Surveillance Requirements for Type A tests basically duplicate the requirements of 10 CFR 50 Appendix J with respect to Type A testing and frequency of testing. Therefore, any time a rule change occurs or an exemption is required, a Technical Specification amendment is also required. The Technical Specification change, requested in Attachment A of this submittal, is proposed to relocate Primary Containment leakage specification 3/4.6.1.2 to specification 3/4.6.1.1, " Primary Containment Integrity," as new Surveillance Requirement 4.6.1.1.b.

DISCUSSION In accordance with 10 CFR 50.12, LaSalle is requesting a partial exemption from the 10 CFR 50 Appendix J, Section III.D.1(a) requirement to perform the third Type A test of each 10-year service period when the plant is shut down for the 10-year plant inservice inspections.

LaSalle also requests the following change to LaSalle Technical Specification 3/4.6.1.2, " Primary Containment Leakage." The change is being proposed in accordance with 10 CFR 50.90 and is reflected in Attachment A. The change to Technical Specification 3/4.6.1.2 and its associated bases is to relocate the requirements regarding primary containment leakage to specification 3/4.6.1.1,

" Primary Containment Integrity," as new Surveillance Requirement 4.6.1.1.b. This will delete the specifics regarding Type A testing frequency. The proposed change to Technical Specifications will remove the Technical Specification requirement of performing the third Type A test of each 10-year service period during the shutdown for the 10-year plant inservice inspections and is therefore consistent with the proposed partial exemption.

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ATTACHMENT E EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A NORMAL TEST SCHEDULE, SEPARATING THE TYPE A TEST SCHEDULE FROM THE INSERVICE INSPECTION SCHEDULE DISCUSSION (continued)

The frequency of the Type A test is as follows per 10 CFR 50, Appendix J section III.D.1.(a) Periodic retest schedule:

"(a) After the preoperational leakage rate tests, a set of three Type A tests shall be performed, at approximately equal intervals during each 10-year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year plant inservice inspections."

Likewise, LaSalle Unit 1 and 2 Technical Specification Surveillance Requirements 4.6.1.2.a restate the Appendix J requirement:

"Three Type A Overall Integrated Containment Leakage Rate tests shall be conducted at 40 10 month intervals during shutdown at P,,39.6 psig, during each 10-year service period. The third test of each set shall be conducted during the shutdown for the 10-year plant inservice inspection. "

LaSalle proposes to perform the three Type A tests at approximately equal intervals within each 10-year period, with the third test of the set conducted as close as practical to the end of the 10-year period without exceeding the 40 10 month band currently specified in specification 4.6.1.2.a. However, there would be no required connection between the Appendix J 10-year interval and the inservice inspection 10-year interval. LaSalle Unit 2's first 10-year Appendix J interval ends in 1994. LaSalle is therefore currently required to perform a Type A test during the spring 1995 refueling outage.

The 10-year plant inservice inspection (ISI) is the series of inspections performed every 10 years in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. LaSalle performs the ISI volumetric, surface and visual examinations of components and system pressure tests in accordance with 10 CFR 50.55a(gX4) throughout the 10-year inspection interval. The major portion of this effbrt is presently being performed every 18 months during the refueling outages. LaSalle's FIRST 10-year ISI i program ends in October,1994. LaSalle is scheduled to complete the first 10-year k sniadasallelilrtIina.wpfPO

ATTACHMENT E EXEMPTION REQUEST FROM APPENDIX J REQUIREMENTS FOR THE TYPE A NORMAL TEST SCHEDULE, SEPARATING THE TYPE A TEST SCHEDULE FROM THE INSERVICE INSPECTION SCHEDULE DISCUSSION (continued) program during the spring of 1995 as allowed by Section XI IWA 2400(c). The inservice inspections scheduled during the 1995 refueling-outage will complete the first 10-year ISI program. The LaSalle Unit 2 first 10-year Appendix J interval will end in October 1994. The fourth Unit 2 Type A test was performed in the fifth refueling outage in December 1993. Approval of this proposed exemption, in conjunction with the approval of the exemption requested by Attachment D of this submittal, will allow scheduling of the next LaSalle Unit 2 Type A test for the seventh refueling out. age to start the three-times-in-ten-years frequency for the second ten year interval. This will maintain approximately equal intervals as required by paragraph III.D of Appendix J.

SUMMARY

AND CONCLUSION Each of these two surveillance tests (i.e., the Type A and the 10-year ISI program) is independent of the other and provides assurances of different plant characteristics. The Type A test assures the required leak-tightness to demonstrate compliance with the guidelines of 10 CFR Part 100. The 10-year ISI program provides assurance of the integrity of the structures, systems, and components in compliance with 10 CFR 50.55a. There is no benefit to be gained by coupling these requirements to the same refueling outage in that elements of the LaSalle ISI program are conducted throughout each 10-year cycle rather than during a refueling outage at the end of the 10-year cycle. Consequently, the subject coupling requirement offers no benefit either to safety or to economical '

operation of the facility. Accordingly, the subject exemption request meets the underlying purpose of the rule [10 CFR 50.12(a)(2)(ii)].

Consistent with this exemption request, a proposed change to LaSalle Technical Specification 3/4.6.1.2 is also being requested, as described in Attachment A.

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l ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES l l

A. Problems:

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1. The LaSalle Unit Two Type B and C test leakage exceeded the !

0.60 L, maximum path leakage, for Type B and C components, during

.each of the first five refueling outages .(1987,1988,.1990,1992 and 1993, respectively).

2. During the first, third and fourth refueling outages, the LaSalle Unit Two Type A test failed the as-found condition due to the penalty additions from Type C tests (minimum path leakage). The minimum path leakage would have been acceptable if a Type A test had been required during the second refueling outage. The as-found 0.60 L, minimum path leakage was not exceeded during either the Unit Two fourth refueling outage, L2R04 or the fifth refueling outage, L2R05.

B. Root Causes of the Problems:

1. Containment Isolation Valve leakage,

2. Inappropriate local leak rate test method for some types of valves, and

3. Weak trending and comparison of valve / penetration performance.

C. General Local Leak Rate Test program improvements:

This is achieved through the development of an ongoing quality Type B and .

C test program coupled with proper maintenance, root cause investigation, and engineering ana?ysis to determine short and long term corrective actions.

LaSalle County Station has a policy to maintain the primary containment leakage as low as possible. Administrative limits are set for each penetration / component. A component is repaired ifits limit is exceeded. A technical evaluation is performed on a case-by-case basis to allow an administrative limit to be exceeded only if the overall containment leakage 1

I J

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ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES is exceptionally low. This policy is established by LaSalle Technical Surveillance LTS-300-5, " Local Leak Rate Test,0.60 L Accountability Program".

The. total maximum-pathway leakage rate is repo.rted to the start-up on-site review committee for approval prior to the end of each refueling outage.

Problem Identification Reports (PIFs) (investigative reports) are initiated as required to document any Type B and/or C failures and describe corrective actions. Type B and C testing is considered an important and effective program where actions are taken to address problem areas independent of the Type A test.

LaSalle County Station has established motor operated valve preventative maintenance on Containment Isolation valves which includes:

1. The valve operator will be scheduled for refurbishment, ifits ,

associated motor operator gear box grease sample is unacceptable.

2. Diagnostic testing of motor operated valve thrust capabilities is underway in accordance with LaSalle responses to Generic Letter 89-10 and it's supplements. Static testing will be completed at least once on all applicable valves by the end of the sixth refueling outage on each LaSalle Unit. Dynamic testing will be completed at least once on all applicable valves by the end of the sixth refueling outage for LaSalle Unit 2 and the seventh refueling outage for LaSalle Unit 1.

D. Specific corrective actions completed or planned on specific valves identified as problems thus far which also impact the Type A test:

1. 2RE024/2RE025, Drywell Equipment Drain Sump and 2RF012/2RF013, Drywell Floor Drain Sump Penetrations:

The drywell equipment and floor drain sumps penetration isolation J valves have repeatedly failed Type C tests. The cause of the failures has been attributed to the introduction of dirt / foreign objects into the drywell equipment and/or floor drain sumps during extended or i refueling outages. The dirt or foreign material would subsequently be !

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ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES pumped through the valves causing this material to settle on valve ,

seats. The result of this caused improper / incomplete valve disc seating and/or irregularities to the valve seat /dise during subsequent valve operation. ,

An evaluation was performed to resolve the recurring failures. It was determined that the best solution to this problem was to install screens at the bottom of the sumps during extended or refueling outages. This would prevent foreign material from entering into the piping / isolation valve seats, thus mitigating valve leakage. The equipment and floor drain sumps have additionally been placed on a periodic cleaning schedule which will also minimize foreign material buildup / transfer.

Sump screens were installed and cleaned during each Unit's fourth refueling outage (Unit 1, L1R04, in the Spring of 1991, and Unit 2, L2R04, in the Spring of 1992). Since the fourth refueling outage for each unit, the sumps have been cleaned and have had screens / standpipes installed to prevent / minimize the intrusion of foreign material into the system piping. This corrective action has demonstrated an improvement in the valves' leakage performance.

Since the implementation of the corrective action, only one valve Type C test failure has occurred. This failure was during L2R05 when it was determined that 2RF012 had excessive leakage. The failure was attributed to dirty seating surface coupled with inadequate spring tension on the valve actuator. the valve seating surfaces were cleaned and the actuator was rebuilt. The minimum pathway leakage was not affected by this failure, as the redundant containment isolation valve 2RF013 wa3 determined to have satisfactory performance.

Ongoing monitoring of valve leakage performance continues to demonstrate that the resolution is in fact appropriate and does not i require any further action, at this time.

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ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES ,

i

2. 2G33-F001/2G33-F004, Reactor Water Cleanup Suction Penetration:

Recurrent Type C test failures of the Reactor Water Cleanup (RWCU)

Suction Isolation Valves have been a contributor to the "as-found" Type A test failures. The causes of the RWCU Isolation Valve i failures vary, but were mainly attributed to dirty or scratched seating surfaces. Table 2 shows the as-found Local Leak Rate Test results for both Unit I and Unit 2 Reactor Water Cleanup Suction Isolation Valves, starting with each unit's first refueling outage.

Starting with the Unit 2 third refueling outage (L2R03), LaSalle revised the testing methodology for the RWCU isolation valves.

Previous isolation valve testing simultaneously pressurized the volume between the two valves. This method created much difficulty in troubleshooting to determine leakye through each valve and ,

identifying the minimum path leakage which would be added to the as-found Type A test results. The new method tests the valves individually in the normal direction (from inside the containment).

This simpler method allows the test engineer to identify the exact leakage through each valve.

An engineering evaluation determined that the single flex wedge gate valve design was inappropriate for the RWCU suction application.

The evaluation concluded that a double disc gate valve would greatly improve leakage performance for this particular application. Each valve of the new design has two discs and associated seating surfaces, thus doubling the number of valve leakage barriers.

LaSalle surveyed other utilities to obtain information on the performance of double disc gate valves in the RWCU suction application. Double disc gate valves are in use in the same application at Clinton Power Station with no valve leakage problems.

The Hope Creek Nuclear Station has not had any recurring problems with double disc gate valves installed since 1985. The Fitzpatrick plant has recently installed double disc gate valves for the same application due to failures with the single flex wedge gate design.

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T' ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES Fitzpatrick experienced Type C test problems after initial installation, believed to be due to improper installation. The valvcs were retested satisfactorily after maintenance.

The L2R04. primary-containment.as-found Type.A test result of i 0.6155% per day exceeded the 0.75 L, criteria of 0.476% per day due to the Calculated Adjusted of 0.2632% per day and was therefore considered a failure. The maximum allowable containment leak rate (L,) of 0.635% per day was not exceeded. Note: The Calculated

. Adjusted leakage rate is found by adding any improvements in local leakage rates due to repair and adjustments to the Type A test results using minimum pathway leakage methodology. The Calculated Adjusted is then used to determine the total as-found containment minimum pathway leakage rate which applies to the as-found Type A test.

The Reactor Water Cleanup Suction i enetration contributed 115.7 scfh to the total Calculated Adjusted of 159.765 scfh and accounted for 72.4% of the total. The sum of the Calculated Adjusted and the as-left Type A test, less the RWCU contribution, would have been 44.1 scfh, or 0.0726% per day. This would have resulted in an as-found Type A test of 0.4249% per day, which is below the 0.75 L, criteria of 0.476% per day.

The Reactor Water Cleanup Suction penetration has historically been a major contributor to and cause of as-found Type A test failures at LaSalle Station. During L2R04 the Reactor Water Cleanup Suction Containment Isolation valves 2G33-F001 and 2G33-F004 were replaced with updated double disc design gate valves per planned modification. They were both successfully Local Leak Rate Tested with zero leakage following replacement. Approximately one month later the 2G33-F001 and 2G33-F004 valves were retested due to motor operator maintenance with Type C test results of 1.5 scfh and 1.44 scfh, respectively.

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ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES After on full operating cycle, the 2G33-F001 and 2G33-F004 valves were tested during the Unit 2 fifth refueling outage (.L2R05). As found T,vpe C tests resulted in outstanding performance with 0.37 scfh for both valves. Following VOTES testing activities, the 2G33-F001 and_2G33-F004 valves were retested in the as-left {

condition with zero leakage for both valves. !

LaSalle Station is confident that the leakage performance of the Reactor Water Cleanup penetration will greatly enhance the Type B and C test / Type A test program.

3. 2IN031, Transverse Incore Probe (TIP) Air Purge Supply Penetration:

The TIP Purge Air Supply Isolation Valve,2IN031, is a single containment isolation valve. This component failed its Type C test during the Unit 2 first refueling outage and greatly contributed to the "as-found" Type A test failure. The cause of the failure was attributed to a dirty seating surface along with corrosion. The valve was disassembled, cleaned and the internals were replaced.

The 2IN031 valve was originally tested in the " reverse" direction. l The test procedure, LTS-100-22, "Drywell Pneumatic System l Discharge Isolation Valves Local Leak Rate Tests 1(2)IN017, l 1(2)IN031, and 1(2)IN018", was revised to test the component in the

" normal" direction which is from inside the containment.

The 2INO31 valve has performed satisfactorily with no leakage since the initial failure during the first refueling outage, demonstrating that the corrective actions taken in the Unit 2 first refueling outage (L2R01) were adequate.

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. .. . . . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ . ________________________________.____a

ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES .

4. 2E12-F053A, "A" Residual Heat Removal Shutdown Cooling (RHR ;

SDC) R-eturn Penetration:

The "A" RHR SDC Return Isolation Valve,2E12-F053A, failed its Type C test during the. Unit 2 first refueling outage and greatly contributed to the "as-found" Type A test failure. The cause of the leakage was identified to be a defective valve disc. A new valve disc was installed and the nlve seat was refaced.

The repair action was very successful based on subsequent testing '

performed during the following four refueling outages. Testing conducted since the initial Type C test failure yielded minimal or no leakage.

5. 2E12-F053B, "B" RHR Shutdown Cooling (RHR SDC) Return Penetration:

The "B" RHR SDC Return Isolation Valve,2E12-F053B, failed its Type C test during the Unit 2 third refueling outage and greatly contributed to the "as-found" Type A test failure. The cause of the excessive leakage was determined to be the failure of the motor operator to properly drive the disc far enough into its seat. The motor operator was refurbished and the valve was retested with satisfactory results. A torque switch that may have contributed to the failure was the sole item replaced during the refurbishment. It was replaced only because it was made of the wrong material (melamine torque switches have been systematically replaced due to a ,

Part 21 notification) and no specific problem with the torque switch was noted in the work request documentation.

The root cause of the valve failing to fully close was undetermined.

LaSalle Administrative Procedure LAP-300-31, " Motor Operated Valve Program", has since been revised to require root cause determination for any Motor Operated Valve failure to prevent recurrence. This failure is deemed an isolated occurrence based upon a review of motor operated valve work request history.

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ATTACHMENT F CORRECTWE ACTION PLAN FOR TYPE C TEST FAILURES ,

CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES ,

i The testing and refurbishment conducted in accordance with the Motor Operated Valve preventative maintenance program yielded significant valve leakage performance, as was demonstrated in L2R04 and L2R05. This valve was satisfactorily tested with zero leakage during the Unit 2 fourth and fifth refueling. outages.(L2R04 and L2R05). Prior to the Type C test failure the 2E12-F053B valve was tested with satisfactory results during the first and second refueling outages.

6. 2HG005A/2HG006A, Unit 2 Hydrogen Recombiner 2HG01A Unit 2 Suppression Pool Discharge Penetration:

The Unit 2 Hydrogen Recombiner 2HG01A Unit 2 Suppression Pool Discharge Isolation Valves,2HG005A and 2HG006A, have had recurring Type C test failures. During the LaSalle Unit 2 first refueling outage (L2R01), the excessive leakage greatly contributed to the "as-found" Type A test failure. The cause of the Type C test failure was determined to be valve seat irregularities. Valve seating surfaces were lapped and the valves were retested satisfactorily.

Only the 2HG006A isolation valve continued to experience leakage during the second, third, and fifth refueling outages (L2R02, L2R03 and L2R05). In each of these instances the valve was determined to have a dirty and irregular seating surface. Retests were performed satisfactorily following cleaning and lapping of the seating surfaces.

Type A test "as-found" results would not have been affected by the single valve Type C test failure since the first refueling outage.

Because of the 2HG005A/2HG006A Isolation valve Type C test failures, all Hydrogen Recombiner Primary Containment Isolation Valves are under an ongoing evaluation to determine what long term corrective action is required to prevent recurring Type C test failures.

There are six other isolation valves of the same size and function as 2HG005A and 2IIG006A, associated with the Unit 1 and Unit 2 Combustible Gas Control Systems. The only other valve failing in a similar manner is the Unit 2 Hydrogen Recombiner 2HG01A Unit 1 Suppression Pool Discharge Valve,1HG005B, which failed the Type C test administrative limit.

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ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES b

There have been a total of seven Type C test failures in 24 Type C tests for these four penetrations and four of these are associated with ;

the 2HG005A/2HG006A penetration. The three remaining failures were associated with the 1HG005B/1HG006B penetration, which were .

corrected by cleaning and. lapping the valve. seat of the.1HG005B. l This Unit 1 penetration was satisfactorily Type C tested three times after the initial failure during the Unit I reactor recirculation pumps forced maintenance and surveillance outage, 5/28/87- 9/14/87 (hereafter designated as "RR/PP"). However, the penetration has failed the last two cefueling outages, L1R05 and L1R06, in the as-found condition.

There are four other penetrations associated with the supply lines to the Hydrogen Recombiners. The Unit 1 Hydrogen Recombiner ,

1HG01A Unit 1 Drywell Suction Valve,1HG001A, and the Unit 2 Hydrogen Recombiner 2HG01A Unit 1 Drywell Suction Valve, 1HG001B, each failed their respective Type C tests during the Unit 1 third refueling outage and were repaired by lapping the valve seats. !

Both of these penetrations were satisfactorily Local Leak Rate Tested during the Unit 1 fourth, fifth and sixth refueling outages. The Unit 1 Type C test failures were shown by test to be due to only one of the two valves in each penetration, so the as-found Type A test was not affected. i l

The Unit 2 2HG005A and 2HG006A containment isolation valves are l to be tested during any non-refueling outage with Unit 2 in cold i shutdown for 14 days or longer, unless tested within the previous 6 months. The Unit 11HG005B and 1HG006B containment isolation valves are to be tested during any non-refueling outage with Unit 1 in cold shutdown for 14 days or longer, unless tested within the previous 6 months.

The overall performance of the Hydrogen Recombiner containment isolation valves has been satisfactory, based upon test results since the first refueling outage of each LaSalle unit.

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A'ITACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES The Unit 2 components have all performed satisfactorily except for the 2HG006A containment isolation valve which has failed four out of '

five Type C tests.

Upon further investigation into the 2HG006A valve Type C test failures, it was discovered that after the L2R01 and L2R02 failures the leakage was reduced to an acceptable level as a result of machine lapping the seating surface of the valve. However, after machine lapping during L2R01, the 2HG006A valve failed its Type C test the following L2R02 outage. After machine lapping during L2R02, the 2HG006A valve failed its Type C test the following outage L2R03.

A different approach was taken to reduce the 2HG006A leakage following the failure during L2R03. T' a seating surface of the 2HG006A valve was lapped by hand, , en blue checked to verify ,

proper seating. Per discussions with members of the Mechanical Maintenance Department, this resulted in a smoother and flatter seating surface. As a result of these actions the 2HG006A valve passed its Type C test during the following L2R04 outage.

During the next Unit 2 refueling outage,12R05, the 2HG006A valve !

once again failed its Type C test. An evaluation is being conducted to determine long term corrective action required to resolve the recurring Type C test failures of the 2HG006A and 1HG005B containment isolation valves. ,

A tabulation of the historical Local Leak Rate Test performance of the above Unit Two 2HG005A and 2HG006A valves is included in Table 1; The performance of the Combustible Gas Control Valves 1(2)HG001A/B,2A/B,5A/B, and GA/B are included in Table 3.

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ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES ,

CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES

7. 2VP053A, Unit 2 Primary Containment Chilled Water Supply An evaluation was performed to address the Type C test failures of the 2VP053A containment isolation valve. All eight Containment

. . Chilled Water Outboard isolation valves of Units 1.and 2 were evaluated. They are 8" 150# single flex wedge gate valves supplied by Anchor / Darling. They are all installed at a 45 degree angle. The 1(2)VP053A/B and the 1(2)VP063A/B are the Containment Chilled.

Water return and supply lines, respectively.

The overall performance of the Unit 1 and 2 Containment Chilled Water Outboard Isolation Valves has been satisfactory, with the exception of the 2VP053A valve.

Unit 1 has experienced two failures over a six outage period which included 24 individual tests. Only one Type C test failure occurred ihr the IVP053A/1VP114A valves (RIUPP). The second Type C test failure occurred for the IVP063A/1VP113A valves (RR/PP). The Unit 1 components have not experienced recurring failures.

The Unit 2 Containment Chilled Water Return Isolation Valve, 2VP053A, has failed its Type C test during all five refueling outages.

The valve was repaired after each failure, then successfully Local Leak Rate Tested prior to Unit startup. The repair actions consisted l of valve disassembly, cleaning, lapping of the disc and seating surface, frost and/or blue checking and reassembly. Although the ,

valve has been successfully retested after each initial outage failure, l the accepted final Type C test results have been substantially higher i than all the other Containment Chilled Water Outboard Isolation ;

Valve Type C test results.

l The 2VP053A valve appears to continuously degrade during each operating cycle, requiring further attention and investigation. The valve was disassembled and inspected during the Unit 2 fifth refueling outage (L2R05) to determine the root cause of the failures, but was indeterminate. A Problem Identification Form (PIF) has k:\nlaslasalle\1lttfina.wpf92 I

ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES been generated for root cause evaluation and a work request has been generated to perform more extensive repairs to 2VP053A during L2R06.

. A tabulation of.the. historical Local Leak Rate Test. performance of the above chilled water valves is included in Table 4.

E. Corrective Actions for the Local Leak Rate Test Program:

1. Development and implementation of an improved trending program to track penetration and valve leakage rate performance. The assembly of an automated plant-specific database will permit identifying valve type, service application, and manufacturer for leakage rate performance comparison of all Primary Containment Isolation Valves that are required to be Local Leak Rate Tested. This will help to determine any patterns or groups of valves that are exceptionally good performers with minimal or no leakage, or poor performers with ceveral cases of high leakage.

2. Alternative Leakage Test Program:

LaSalle will perform Type B and C tests on penetrations identified to be susceptible to excessive leakage in accordance with the above >

stated item 1. These Type B and C tests will be performed during any non-refueling outage with the unit in cold shutdown for 14 days or longer. These tests are in addition to the scope of Type B and C tests performed during refueling outages for as-found maximum path leakage.

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ATTACHMENT F CORRECTIVE ACTION PLAN FOR TYPE C TEST FAILURES CONTRIBUTING TO "AS-FOUND" TYPE A TEST FAILURES

3. Recommendation and implementation of test method improvement, or repair, modification, or replacement of problem Primary Containment Isolation valves identified in 1. above, based upon engineering evaluation.

The test method improvement will include a technical review of Type B and C test procedures associated with poorly performing components to verify the following:

9 Proper type of test for the penetration involved (for example, test direction from the Primary Containment outward versus between a pair ofisolation valves in the same penetration, or, from the outside in),

G Test boundary adequacy, and, 9 Other aspects of test methodology such as clarity and user friendliness.

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ATTACHMENT F TABLE 1 Unit Two Local Leak Rate, Type C, Test Performance Since 1987 VALVE TYPE C LEAKAGE TEST, "AS-FOUND" LEAKAGE CONTRIBUTION TO "AS-IDENTIFI- MAXIMUM PATH LEAKAGE, SCFH FOUND" TYPE A LEAKAGE TEST CATION 1987 1988/89 1990 1992 1993 L2R01 L2R03 L2R04 L2R05 L2R01 L2R02** L2R03 L2R04 L2R05 2RE024 29.1 5.9 5.2 5.05 19.4 0.0 N/A* 0.84 N/A*

2RE025 2RF012 0.51 9.19 225.5 1.02 107.66 N/A* 110.32 N/A* 20.04 2RF013 2G33-F001 17.7 200.1 GROSS 127.7 0.37 8.67 GROSS 115.7 0.37 2G33-F004 2INO31 250.9 0.0 0.0 0.0 0.0 250.88 N/A* N/A* N/A*

2E12-F053A 88.2 0.0 2.65 0.0 4.43 0.0 N/A* N/A* N/A*

2E12-F053B 0.46 0.0 65.7 0.0 0.0 0.0 63.28 N/A* N/A*

2HG005A 121.6 31.9 111.0 4.17 21.16 60.57 0.5 0.0 0.2 2HG006A 2HG005B 0.37 0.32 0.37 0.0 0.0 N/A* N/A* 0.0 N/A*

2HG006B

No work performed on valves **Non-Type A test outage.

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ATTACHMENT F TABLE 2 Local Leak Rate, Type C, Test Performance for Unit 1 and 2 Reactor Water Cleanup Suction Isolation Valves LASALLE VALVE FIRST RR/PP*,*

SECOND THIRD FOURTH FIFTH SIXTH UNIT REFUEL REFUEL

REFUEL REFUEL

REFUEL REFUEL (VALUES SHOWN ARE AS-FOUND/AS LEFT) 1 1G33-F001 1.71/1.72 1.84/1.84 2.7/2.7 GROSS'/37.4 23.382/0.0 25.68/0.0 0.76/0.76 IG33-F004 28.06 /0.0 0.0/0.0 75.83'/0.0 2 2G33-F001 17.74 /0.37 N/A 200.15/3.31 GROSS'i/3.31 115.78/0.0 0.37/0.0 (NEXT 2G33-F004 GROSS7 /4.19 127.78/0.0 OUTAGE)

NOTES:

Non-Type A test outage, L2R04 and L2R05 were Type A test outages due to Type A test failures in L2R01 and L2R03.

- Unit I reactor recirculation pumps forced maintenance and surveillance outage, 5/28/87- 9/14/87.

1 Incomplete seating surfaces on 1G33-F004. The valve seating surfaces were cleaned and lapped. The retest was accepted based on low total maximum combined leakage.

2 Incomplete seating surfaces on 1G33-F001. The valve seating surfaces were cleaned and lapped and the valve was repacked.

k:snia\lasc.11e\tirtfinc.wpf96 i

k

.---.---.---.--._-,---~._-----------=-.---------------a-_-n - - w we -n -- - ,. .ei s-

ATTACHMENT F TABLE 2 Local Leak Rate, Type C, Test Performance for Unit 1 and 2 Reactor Water Cleanup Suction Isolation Valves 3 Incomplete seating surfaces on 1G33-F004. The valve seating surfaces were cleaned and lapped.

4 Incomplete seating surfaces on 2G33-F001. The valve seating surfaces were cleaned and lapped. The seating surface was slightly damaged and a crack was found in the disc of 2G33-F004. Lapped the valve seat, machined and lapped the disc to match the seat, and repacked 2G33-F004.

5 Incomplete seating surfaces on 2G33-F001. The valve seating surfaces were cleaned and lapped. Replaced the disc on 2G33-F004 due to crack found the previous outage. Lapped the valve seat, machined and lapped the new disc to match the seat and repacked 2G33-F004.

6 Incomplete seating surfaces on 2G33-F001. The valve seating surfaces were cleaned and lapped.

7 Incomplete seating surfaces and severe packing leak on 2G33-F004. The valve seating surfaces were cleaned and lapped and the valve was repacked.

8 Single Gex wedge gate valves removed and replaced with double disc gate valves.

9 Valve disk was fbund to be cracked and seats warped. This is believed to be caused as a result ofinitial installation problems where the valve had to be welded / stress relieved several times.

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_ _ _____....__ ___. ___._ _ ____m.._ _ ______ _ __ .

ATTACHMENT F TABLE 3 Local Leak Rate, Type C, Test Performance for Unit I and 2 Combustible Gas Control Isolation Valves LaSalle Unit 1 (Values shown are as found/as left)

Valves L1R01 RR/PP* L1R02 LIR03 LIR04 L1R05 L1.R06 1HG001A/ 1.76/0.89 0.56/0.56 0.6/0.6 125.9/0.0 3.8/0.38 0.0 0.0 1HG002A _

1HG005A/ 1.15/1.15 1.39/1.39 1.75/1.75 0.186/0.186 1.58/1.58 2.9/2.32 1.67/0.37 1HG006A 1HG001B/ 1.72/1.72 1.21/1.21 3.24/3.24 19.3/0.0 0.56/0.0 0.46/0.46 0.37/0.0 1HG002B 1HG005B/ 4.8/4.8 9.86/1.61 1.29/1.29 2.96/1.67 0.28/0.65 13.33/3.52 27.75/4.16 1HG006B

- Unit 1 reactor recirculation pumps forced maintenance and surveillance outage, 5/28/87- 9/14/87.

k:\nlasle alle\ilrtfina.wpf99

ATTACHMENT F TABLE 3 Local Leak Rate, Type C, Test Performance for Unit I and 2 Combustible Gas Control Isolation Valves LaSalle Unit 2 (Values shown are as found/as left)

Valves L2R01 L2R02 L2R03 L2R04 L2R05 2HG001A/ 0.37/0.37 0.28/0.28 0.52/0.52 1.07/1.07 0.83/1.11 2HG002A 2HG005A/ 121.6/0.47 31.9/2.61 111.0/1.0 4.17/6.51 21.16/6.01 2HG006A 2HG001B/ 1.3/1.3 0.28/0.28 0.0/0.0 0.56/0.56 0.74/0.37 2HG002B 2HG005B/ 0.37/0.37 0.32/0.32 0.37/0.0 0.0/0.0 0.0/0.0 2HG006B k:\nla\laralle\ilrtfina.wpf99

ATTACHMENT F TABLE 4 Local Leak Rate, Type C, Test Performance for Unit 1 and 2 Primary Containment Chilled Water (VP) Isolation Valve Performance LaSalle Unit 1 (Values shown are as found/as left)

Valves L1R01 RR/PP* L1R02 L1R03 L1R04 L1R05 L1R06 IVP053A/ 1.15/1.15 GROSS /1.25 0.05/0.05 0.78/0.78 0.0/0.37 0.0/0.37 0.37/0.0 1VP114A IVP053B/ 1.38/1.38 2.39/2.39 1.67/1.67 1.3/1.3 1.71/1.71 1.71/1.71 1.39/1.39 IVP114B IVP063A/ 3.57/3.57 25.3/1.21 1.07/1.07 1.49/1.49 0.0/0.0 0.0/0.0 0.37/0.37 IVP113A IVP063B/ 0.05/0.05 1.1/1.1 1.86/1.2 2.03/2.03 1.35/1.35 1.35/1.3 1.58/1.57 IVP113B

- Unit I reactor recirculation pumps forced maintenance and surveillance outage,5/28/87- 9/14/87.

k : \ rt! s \ lasalle\ ilrt f ina .wpf 100

ATTACHMENT F TABLE 4 Local Leak Rate, Type C, Test Performance for Unit 1 and 2 Primary Containment Chilled Water (VP) Isolation Valve Performance LaSalle Unit 2 (Values shown are as found/as left)

Valves L2R01 L2R02 L2R03 L2R04 L2R05 2VP053A/ 9.38/3.4 32.5/7.4 GROSS /16.66 24.5/5.6 0.83/12.16 2VP114A 2VP053B/ 0.47/0.47 0.55/0.55 0.0/0.46 0.37/0.37 0.0/0.0 2VP114B 2VP063A/ 1.04/1.04 0.0/0.0 0.0/0.0 1.39/1.39 0.0/1.38 2VP113A 2VP063B/ 1.85/1.85 0.83/0.83 0.37/9.34 0.46/0.37 0.69/0.69 2VP113B k:\nla\lasalle\ilrtfina.wpfl01 L _ __ _ _ _ _ _ _ _ _ _ _ . _ _ _ _- _ . . _ . . .- . ._. -- _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

l ATTACHMENT G Previous and Proposed Type A Test Schedule for Unit 2 TYPE A TEST DATF; DESCRIPTION CONDUCTED PASS / FAIL .

I JANUARY 1987 L2R01 YES FAIL OCTOBER 1988 L2R02 NO' ----

MARCH 1990 L2R03 YES FAIL MARCH 1992 L2R04 YES FAIL 8 DECEMBER 1993 L2R05 YES PASS FEBRUARY - APRIL 1995 L2R06 NO 2 . . ,

NOVEMBER 1996 - L2R07 YES2 ___,

JANUARY 1997 NOTE 1: Minimum path leakage for determining leakage penalties for an as-found Type A test would have been acceptable.

NOTE 2: Proposed Type A test Schedule NOTE 3: The L2R04 primary containment as-found Type A test result of 0.6155% per day exceeded the 0.75 L, criteria of 0.476% per day due to the Calculated Adjusted of 0.2632% per day and was therefore considered a failure. The maximum allowable containment leak rate (L ) of 0.635% per day was not exceeded. The Reactor Water Cleanup Suction penetration contributed 115.7 scfh to the total Calculated Adjusted of 159.765 scfh and accounted for 72.4% of the total. The sum of the Calculated Adjusted and the as-left Type A test, less the RWCU contribution, would have been 44.1 scfh, or 0.0726% per day.

This would have resulted in a satisfactory as-found Type A test of ;

0.4249% per day, which is below the 0.75 L, criteria of 0.476% per day.

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l ATTACIIMENT H ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW .

Commonwealth Edison has evaluated the proposed amendment against the criteria for identification oflicmsing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been determined that the '

. proposed changes meet the criteria for a categorical exclusion as provided under 10 CFR 51.22 (c)(9). This conclusion has been determined because the changes :

requested do not pose significant hazards consideration or do n'ot involve a significant increase in the amounts, and no significant changes in the types, of any '

effluents that may be released offsite. Additionally, this request does not involve a -

significant increase in individual or cumulative occupational radiation exposure.

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ML20078B957

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