ML20153E577

Amends 59 & 39 to Licenses NPF-11 & NPF-18,respectively, Revising Tech Spec to Correct Inconsistency Between Requirements Re Suppression Pool High Level Alarm

ML20153E577
Person / Time
Site:	LaSalle
Issue date:	08/31/1988
From:	Muller D Office of Nuclear Reactor Regulation
To:	Commonwealth Edison Co
Shared Package
ML20153E580	List: ML20153E577 ML20153E592
References
NPF-11-A-059, NPF-18-A-039 NUDOCS 8809060353
Download: ML20153E577 (12)

Text

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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION n

y, t

W ASHING TON, D. C. 20655 y.....)

COMMONWEAI.TH EDISON COMPANY i

DOCKET NO. 50-373 LASALLE COUNTY STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amenduient N. 59 j

9 License No. NPF-11 1.

The Nuclear Regulatory Comission (the Comission or the NRC) has found that:

A.

The application for amendment filed by the Comonwealth Edison Company (thelicensee),datedApril 29, 1987 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) th6t such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR I

Chapter I; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the %miasion's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the enclosure to this license amei. ament and paragraph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.59, and the Er,vironmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accorc'ance with the Technical Specifications and the Environmental Protection Plan.

\\

eso9060353 000031 DR ADOCK 000 3

3.

This amendment is effective upon date of issuance.

FOR THE HUCLEAR REGULATORY COPNISSION Daniel R. Muller, Director Prefect Directorate III 2 Divisfor, of Reactor Projects - III, IV, Y and Special Projects

Enclosure:

Changes to the Technical Specifications Date of Issuance:

August 31, 1988 9

l

ENCLOSURE TO LICENSE AMENDMENT NO. 59 FACILITY OPERATING LICENSE NO. NPF-11 DOCKET N0. 50-373 9

Replica the following pages of the Appendix "A" Technical Specifications with the enclosed pages. T.se revised pages are identified by amendrent number and contain a vertical line indicating the area of change.

REMOVE INSERT i

ytX XIX 3/4 3-30 3/4 3-30 2

3/4 5-8 3/4 5-8 3/4 5-9 3/4 5-9 3/4 6-16 3/4 6-16 3/4 6-13 3/4 0-18 B 3/4 5-2 B k 4 5-2 B 3/4 6-3 8 3/4 6-3 b 3/4 6-3a (new pane) 4

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4 il 1

4 i

l j

1

f 4

A

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INDEX LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIREMENTS........................

3/4 1-21 3.1.5 2 SODIUM PENTABORATE (Na2B o0:s 10 H O) i 2

VOLUME / CONCENTRATION REQUIREMENTS.................

3/4 1-22 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT CENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPC,SURE, JNITIAL CORE FUEL TYPES 8CRB176 8CRB2.19 and 8CRB071........................,.........,..........

3/4 2-2 j

3.2.1-2 MAXIMJM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FUEL TYPE BP8CRB299L.........................,.....

3/4 2-2a 3.2.3-1 MINIMUM CRITICAL POWER RATIO (MC?R) VERSUS

]

t AT RATED FLOW.................................

3/4 2-5 3.2.3-2 K FACTOR.........................................

3/4 2-6 l

f 3.4.1.1-1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE FLOW (% OF RATED)............................

3/4 4-lb 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS, REACICR VESSEL PRESSURE.......................

3/4 4-18 l

4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST........

3/4 7-32 B 3 '; 3-1 REACTOR VESSEL WATER LEVEL........................

B 3/4 3-7 B 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV) at 1/4 T AS A FUNCTION OF SERVICE LIFE.....................

B 3/4 4-7 B 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS..................

B 3/4 6-3a 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS AND LIQUID EFFLUENTS..............................

5-2 5.1.2-1 LOW POPULATION ZONE....

5-3 j

6.1-1 CORPORATE MANAGEMENT......

6-11 4

l 6.1-2 UNIT ORGANIZATION.................................

6-12 6.1-3 MINIMUM SHIFT CREW COMPOSITION....................

6-13 a

i i

}

LA 3ALLE - UNIT 1 X.A Am nde nt No. 59

TABLE 3.3.3-2 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS E

ALLOWABLE e

TRIF FUNCTION TRIP SETPOINT VALUE C

h C.

DIVISION 3 TRIP SYSTEM 1.

HPCS SYSTEM a.

Reactor Vessel ~14ter Level - Low Low, Level 2 1-50 inches

• 1-57 inches

• b.

Drywell Pressure - High i 1.69 psig i 1.89 psig c.

Reactnr Vessel % ter Level - High, Level 8 1 55.5 inches

• i 56 inches

• d.

Condensate Storage Tank Level - Low 1 715'7" 1 715'3" e.

Suppression Pool Water Level - High 1 2 inches **

1 3 inches **

f.

Pump Discharge Pressnre - High 1 120 psig 1 110 psig g.

HPCS System Flow Rate - Low 1 1000 gpm 2 900 gpa

,y h.

Manual Intiation MA NA

,T 0.

LOSS OF POWER o

1.

4.15 kV Emergency Bus Underroltage (Loss of Voltage)#

a.

4.16 kV Basis

1) Divisions 1 and 2 2625 i 131 volts with 2625 1 262 volts with 5 10 seconds time delay 1 11 seconds time delay 24% i 125 volts with 24% i 250 volts with 1 4 seconds time delay 1 3 seconds time delay

2) Division 3 75170 1 143 volts with 2870 1 287 volts with g

5 10 seconds time delay 5 11 seconds time delay A

P

• See Bases Figure B 3/4 3-1.

1. These are inverse time delay voltage relays or instantaneous voltage relays with a time delay. The i

u.

voltages shown are the maximum that will not result in a trip.

Lower voltage conditions will result in decreased trip times.

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

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 SUPPRESSION CHAMBER #

LIMITING CONDITION FOR OPERATION 3.5.3 The suppression chamber shall be OPERABLE:

a.

In OPERATIONAL CONDITION 1, 2, or 3 with a contained water volume of ft, equivalent to a level of -4 1/2 inches.**

l 8

at least 128,800 b.

In OPERATIONAL CONDITION 4. 5* with a contained water volume of at least 70,000 ft3, equivalent to a level of -12 feet 7 inches,** except l that the suppression chamber level may be less than the limit or may be drained in OPERATIONAL CONDITION 4 or 5* provided that:

1.

No operations are performed that have a potential for draining the reactor vessel, 2.

The reactor mode switch is locked in the Shutdown or Refuel

position, 3.

The condensate storage tank contains at least 135,000 available gallons of water, equivalent to a level of 14.5 feet, and 4.

The HPCS system is OPERABLE per Specification 3.5.2 with an OPERABLE flow path capable of taking suction from the condensate storage tank and transferring the water through the spray sparger to the reactor vessel.

i

, APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5*.

,A1 TION:

In OPERATIONAL CONDITION 1, 2, or 3 with the suppression chamber a.

water level less than the above limit, restore the water level to within the limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least. HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the 'ollowing 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

In OPERATIONAL CONDITION 4 or 5* with the suppressior chamber water level less than the above limit or drained and the above required conditions not satisfied, suspend CORE ALTERATIONS en'.' all operations that have a potential for draining the reactor vesst' n d lock the reactor mode switch in the Shutdown position.

Establish SECONDARY CONTAINMENT INTEGRITY withia 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

1. See Specification 3.6.2.1 for presr,ure suppression requirements.

• The suppression chamber is not required to be OPERABLE provi o i that the reactor vessel head is removed, the cavity is flooded or being flooded 3

from the suppression pool, the spent fuel pool gatet are removed when the i

cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

• • Leve,1 is referenced to a plant elevation of 699 feet 11 inches (See i

Figure 8 3/4.6.2-1).

LA SALLE - UNIT 1 3/4 5-8 Amendment No. 59

/

4 EMERGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

ACTION: (Continued) c.

With one suppression chamber water level instrumentation channel inoperable, restore the inoperable channel to OPERABLE status within 7 days or verify the suppression chamber water level to be greater than or equal to -41/2 inches ** or -12 feet 7 inches **, as applicable, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by local indication.

d.

With both suppression chamber water level instrumentation channels inoperable, restore at least one inoperable channel to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and verify the suppression chamber water level to be greater than or equal to -41/2 inches ** or -12 feet 7 inches **, as applicable, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by local indication.

SURVEILLANCE REQUIREMENTL 4.5.3.1 The suppression chamber shall be determined OPERABLE by verifying:

The water level to be greater than or equal to, as applicable:

a.

1.

-4 1/2 inches ** at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

-12 feet 7 inches ** at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.

Two suppression chasher water level instrumentation channels OPERABLE by performance of a:

1.

CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2.

CHANNEL FUNCTIONAL TEST at least once per 31 days, and 3.

CHANNEL CALIBRATION at least once per 18 months, with the los water level alarm setpoint at greater than or equal to

-? inches.**

l j

4.5.3.2 With the suppression chamber level less tha..

5e above limit or drained i

in OPERATIONAL CONDITION 4 or 5*, at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:

a.

Verify the required conditions of Specification 3.3.3.b. to be satisfied, or b.

Verify footnote conditions

• to b satisfied.

• The suppression chamber is not required to be OPERAB!E provided that the reactor vessel head is removed, the cavity is flooded or being flooded from the suppres-sion pool, the spent fuel pool gates are removed when the cavity is flooded, and the water level is maintained within the limits of Specifications 3.9.8 and 3.9.9.

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

LA SALLE - UNIT 1 3/4 5-9 Amendment No. 59 w

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION SYSTEMS SUPPRESSION CHAMBER LIMITING CONDITION FOR OPERATIO'A 3.6.2.1 The suppression chamber shall be OPERABLE with:

a.

The pool water:

1.

Volume between 131,900 ft8 8

and 128,800 ft, e level between +3 inches ** and -4 1/2 inches

• quivalent to a

, and a 2.

Maximum average temperature of 100'F* during OPERATIONAL CONDITION 1 or 2, except that the maximum average temperature may be permitted to increase to:

a) 105'F,## during testing which adds heat to the suppression

chamber, b) 110*F with THERMAL POWER less than or equal to 1% of RATED THERMAL POWER.

c) 120*F with the main steam line isolation valves closed following a scram.

b.

Drywell-to-suppression charmer bypass leakage less than or equal to 10%oftheacceptableA//Edesignvalueof0.03ft.

2 APPLICABILITY:

OPERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

With the suppression chamber water level outside the above limits, a.

restore the water level to within the limits within I hour or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

In OPERATIONAL CONDITION 1 or 2 with the suppression chamber average water temperature greater than or equal to 100'F, restore the average temperature to less than or equal to 100'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, except, as permitted above:

1.

With the suppression chamber average water temperature greater i

than 105*F during testing which adds heat to the suppression charber, stop all testing which adds heat to the suppression chamber and restore the average temperature to less than or equal to 100'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least H3T SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2.

With the suppression chamber average water temperature greater than 110'F place the reactor mode switch in the Shutdown position and operate at least one residual heat removal loop in the suppression pool cooling mode.

3.

With the suppression chamber average water temperature greater than 120'F, depressurize the reacto c 'ssure vessel to less than 200 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1. 5ee Specification 3.5.3 for ECCS requirements.

1. 1. See Special Test Exception 3.10.8.

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

LA SALLE - UNIT 1 3/4 6-16 Amendment No. 59 i

CONTAINMENT SYSTEMS i

SURVEILLANCE REQUIREMENTS (Continued) c.

By verifying at least two suppression chamber water level instru-mentation channels and at least 14 suppression pool water teniperature instrumentation channels, 7 in each of two divisions, OPERABLE by performance of a:

1.

CHA'4NEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2.

CHANNEL FUNCTIONAL TEST at least once per 31 deys, and 3.

CHANNEL CALIBRATION at least once per 18 months.

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

a)

High water level 1 +2 inches

• b)

Low water level > -3 inches

• c)

High temperature 1 100 F 4

d.

By conducting drywell-to-suppression chamber bypass leak tests and verifying that the A/# calculated from the measured leakage is within the specified limit when drywell-to-suppression chamber 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/8 within the specified lianit, and 2.

The A/ 8 calculated from the leak tests at 1.5 psi is < 20% of the specified limit.

then the leak tests at 5 psi may be discontinued.

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

i i

LA SALLF. - UNIT 1 3/4 6-18 Amendment No. 59 i

EMERGENCY CORE COOLING SYSTEMS 4

BASES ECC5-OPERATING and SHUT 00WN (Continued) the suppression pool into the reactor, but no credit is taken in the hazards analyses for the condensate storage tank water.

With the HPCS system inoperable, adequate core cooling is assured by the i

OPERABILITY of the redu'1 dant and diversified automatic depressurization system and both the LPCS and LPCI systems.

In addition, the reactor core isolation cooling (RCIC) system, a system for which no credit is taken in the hazards analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition.

The HPCS out-of-service period of 14 days l

is based on the demonstrated OPERABILITY of rerfundant and diversified low i

pressure core cooling systems.

The surveillance requirements provide adequate assurance that the HPCS system will be OPERABLE when required.

Although all active components are test-able and full flow can be demonstrated by recirculation through a test loop 4

t 4

during reactor operation, a complete functional test with reactor vessel injec-tion requires reactor shutdown.

The pump dischargo piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment.

Upon failure of the HPCS system to function properly, if required, the i

automatic depressurization system (ADS) automatically causes selected safety-relief valves to open, depressurizing the reactor so that flow from the low I

pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200'F. ADS is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 122 psig even though low pressure core

[

cooling systems provide adequate core cooling up to 350 psig.

i ADS automatically controls seven selected safety-relief valves.

Six j

valves are required to be OPERABLE since the LOCA analysis assumes 6 ADS valves in addition to a single failure.

It is therefore appropriate to permit i

I one of the required valves to be out-of service for up to 14 days without materially reducing system reliability.

3/4.5.3 SUPPRESSION CHAMBER l

The suppression chamber is also required to be OPERABLE as part of the i

ECCS to ensure that a sufficient supply of water is available to the HPCS, i

j LPCS and LPCI systems in the event of a LOCA.

This limit on suppression chamber minimum water volume ensures that sufficient water is available to permit 2

re:irculation cooling flow to the core.

The OPERABILITY of the suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is required by Specification 3.6.2.1.

I Repair work might require making the suppression chamber inoperable, i

\\

This specification will permit those repairs to be made and at the same time l

give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chamber must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5.

In OPERATIONAL CODDITION 4 and 5 the suppression chamber minimum required 1'

water volume is reduced because the reactcr coolant is maintained at or below 1

200*F.

Since pressure suppression is not required below E12'F, the minimum i

water volume is based on NPSH, recirculation volume, vortex prevention plus a 1

2'-4" safety margin for conservatism.

i LA SALLE - UNIT 1 B 3/4 5-2 Amendment No. 59 i

i i

CONTAINMENT SYSTEMS BASES 3/4.6.2.

DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 45 psig during primary system blowdown from full operating pressure.

The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.

The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1020 psig.

Since all of the gases in the drywell are purged into the suppression chamber air space during a loss of coolant accident, the pressure of the liquid must not exceed 45 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.

(See Figure B 3/4.6.2-1)

Using the minimum or maximum water volumes given in this specification, containment pressure during the design basis accident is approximately 39.6 psig which is below the design pressure of 45 psig. Maximum water volume of 131,900 ft3 results in a downcomer submergence of 12.4 ft and the minimum volume of 128,800 ft3 results in a submergence approximately 8 inches less.

The majority of the Bogeda tests were run with a submerged length of four f it J

and with complete condensation.

Thus, with respect to the downcomer submergence, this specification is adequate.

Should it be necessary to make the suppression chamber inoperable, this 4

shall only be done as specified in Specification 3.5.3.

under full power operating conditions, blowdown from an initial suppression chamber water temperature of 90'F results in a water temperature of approximately 135'F immediately following blowdown which is below the 200*F used for complete condensation via T quencher devices.

At this temperature and atmospheric pressure.

i the available NPSH exceeds that required by both the RHR and core spray pumps,

)

thus there is no dependency on containment overpressure during the accident j

injection phase.

Experimental data indicates that excessive steam condensing loads can be avoided if the peak bulk temperature of the suppression pool is maintained 1

below 200*F during any period of relief valve operation with sonic conditions at the discharge exit for T quencher devices.

Specifications have been placed i

on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

t LA SALLE - UNIT 1 B 3/4 6-3 Amendment No. 59 i

Control Suppression Plant Room / Local Chamber Level Elevation Indication 26' 10" 700' 2"

+3" High Level LCO (Volume 8

131,900 ft )

26' 9" 700' 1"

+2" High Level Alarm HPCS Suction Valve Transfer 26' 7" 699' 11" 0" Instrument Zero 26' 4" 699' 8"

-3" Low Level Alare 26' 2 1/2" 699' 6 1/2"

-4 1/2" Low Level LCO Opera-tional Condition 1, 2, or 3 (Volume 128,800 ft )

8 14' 687' 4"

-12' 7" Low Level LCO Opera-tional Condition 4 or 5 (Volume 70,000 ft8) l l

l l

l I

SUPPRESSION POOL LEVEL SETPOINTS 1

BASES FIC'.dE B 3/4.6.2-1 l

LA SALLE - UNIT 1 B 4/4 6-3a Amendmerit No. 59 3

o,,

UNITED STATES

[

g NUCLEAR REGULATORY COMMISSION 5

rR WASHINGTON, D. C. 20655

\\,.....j COMMONWEALTH EDISDN COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendinent No. 39 License No. NPF-18 1.

The Nuclear Regulatory Comission (the Comission or the NRC) has found that:

A.

The application for amendment filed by the Comonwealth Edison Company (thelicensee),datedApril 29, 1987 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Comission's regulation: set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.

There is reasonable assurence: (1) that the activities authorized by this amendment can be ccnducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment vill not be inimical to the comon defense and security or to the health and safcty of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by cht.1ges to the Technical Specifica-tions as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating Licensa No. NPF-18 is hereby aranded to read as follows:

(2) Technical Specificctions and EnvironmentiProtectiori Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.39, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

s 3.

This amendenent is effective upon date of issuance.

FOR THE NUCLEAR REGULATORY COM4ISSION W

Daniel R. Muller, Director Project Directorate III-2 Division of Reactor Projects - III, IV, Y and Special Projects

Enclosure:

Changes tc the Technical Spc:ifications Date of Issuance: August 31, 1988 9

l ENCLOSURE TO LICENSE AMENDMENT NO. 39 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 f

i Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change.

REMOVE INSERT XIX XIX 3/4 3-30 3/4 3-30 j

3/4 5-8 3/4 5-8 j

3/4 5-9 3/4 5-9

)

3/4 6-19 3/4 6-19 3/4 6-21 3/4 6-21 8 3/4 5-2 8 3/4 5-2 B 3/4 6-3 B 3/4 6-3 B 3/4 6-3a (new page) 1 l

4 i

l 1

1 J

l 1

p i

I 1

I i

1 d

1 i

i j

J

O LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIRFMENTS........................

3/4 1-21 3.1.5-2 SODIUM PENTABORATE (Na B 2 10 16 10 H O) 0 2

VOLUME /CONCENTRATIONREQUIREMENTS.................

3/4 1-22 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPES 8CRB176 8CRB219 and 8CRB071........................,.........,..........

3/4 2-2 3.2.1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE 4

(MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FUEL TYPE BP8CRB299L..............................,..........

3/4 2-2(a)

[

^

3.2.3-1 MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS t AT RATED FLOW..................................

3/4 2-5 j

3.2.3-2 K FACTOR.........................................

3/4 2-6 f

3.4.1.1-1 CORE THERMAL POWER (% GF RATED) VERSUS TOTAL CORE FLOW (% OF RATED)..................................

3/4 4-2a 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE j

VS. REACTOR VESSEL PRESSURE.......................

3/4 4-19 4.7-1 SAMPLE PLAN 2) FOR SNVBBER FUNCTIONAL TEST........

3/4 7-33

-l B 3/4 3-1 REACTOR VESSEL WATER LEVEL........................

B 3/4 3-7 1

B 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV) at 1/4 T AS A FUNCTION OF SERVICE LIFE.....................

B 3/4 4-7 l

B 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS..................

B 3/4 6-3a 1

5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS AND LIQUID EFFLUENTS..............................

5-2 i

5.1.2-1 LOW POPULATION ZONE...............................

5-3 2

6.1-1 CORPORATE MANAGEMENT..............................

6-11 1

6.1-2 UNIT ORGANIZATION.................................

6-12 3

6.1-3 MINIMUM SHIFT CREW COMPOSITION....................

6-13 LA SALLE - UNIT 2 XIX Arnendment No. 39 i

. _=

l s

l TABLE 3.3.3-2 (Continued)

~

l EMERGENCY CORE COOLING SYSIEM ACIUATION INSTRUMENTATION SETPOINTS 5

l ALLOWABLE y,

l jt TRIP FUPCTION TRIP SETPOINT VALUE E

C.

DIVISION 3 TRIP SYSTEM l

E 1.

HPCS SYSTEM h

a.

Reactor Vessel Water Level - Low Low, level 2 2-50 inches

• 2-57 inches

• b.

Drywell Pressure - High 1 1.69 psig i 1.89 psig i

m I

c.

Reactor Vessel Water Level - High, Level 8

< 55.5 inches *

< Sf, inches

• d.

Condensate Storage Tank Level - Low i 715'7" i 715'3" e.

Suppression Pool Water Level - High 32 inches **

33 inches **

f.

Pump Discharge Pressure - High 2 120 psig 1 110 psig g.

HPCS System Flow Rate - Low

> 1000 gpm 1 900 gpa h.

Manual Intiation N.A.

N.A.

D.

LOSS 6F POWER 1.

4.16 kV Emergency Bus Undervoltage

}

(Loss of Voltage)#

-3 inches

• c)

High temperature 1 100'F 4 d. By conducting drywell-to suppression chamber bypass leak tests and verifying that the A/4 calculated from the measured leakage is within the specified limit when drywell-to-suppression chamber bypass leak tests are conducted: 1. At least once per 18 months at an initial differential pressure of 1.5 psl, and 2. At the first refueling outage and then ur, the schedule required for Typo A Overall Integrated Containment Leakage Rate tests by Speci-fication 4.6.1.2.n., at an initial differential pressure of 5 psi, except ' hat, if the first two 5 psi leak tests performed up to that time result in: 1. A calculated A/8 within the specified limit, and 2. The A/5 calculated from the leak tests at 1.5 psi is < 20% of j the specified limit, t ~ i i then the leak tests at 5 psi may be discontinued. i )

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

i LA SALLE - UNIT 2 3/4 6-21 Amendnent No. 39 ( J EMERGENCY CORE COOLING SYSTEMS BASES ECCS-OPERATING and SHUTDOWN (Continued) the suppression pool into the reactor, but no credit is taken in the hazards l analyses for the cundensete :tnrage tank water. } With the HPCS system inoperable, adequate core cooling is assured by the ( ) OPERABILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems. In addition, the reactor core isolation i cooling (RCIC) system, a system for which no credit is taken in the hazards analysis, will automatically provide makeup at reactor operating pressures or. l a reactor low water level condition. The HPCS out-of-service period of i 14 days is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems. l The surveillance requirements provide adequate assurance that the HPCS l system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test ] loop during reactor operation, a complete functional test with reactor vessel i injection requires reactor shutdown. The pump discharge piping is maintained full to prevent water hammer damage and to provide cooling at the earliest moment. Upon failure of the HPCS system to function properly, if required, the automatic depressurization system (ADS) automatically causes selected safety-I relief valves to open, depressurizing the reactor so that flow from the low i j pressure core cooling systems can enter the core in time to limit fuel cladding i temperature to less than 2200'F. ADS is conservatively required to be OPERABLE l whenever reactor vessel pressure exceeds 122 psig even though low pressure core cooling systems provide adequate core cooling up to 350 psig. l } ADS automatically controls seven selected safety-relief valves. Six j valves are required to be OPERABLE since the LOCA analysis assumes 6 ADS valves in addition to a single failure. It is therefore appropriate to permit one of the required valves to be out-of-service for up to 14 days without i materially reducing system reliability. l 3/4.S.3 SUPPRESSION CHAMBER ) The suppression chamber is also required to be OPERABLE as part of the ECCS a to ensure that a sufficient supply of water is available to the HPCS, LPCS and LPCI systems in the event of a LOCA. This limit on suppression chamber airsirmum water volume ensures that sufficient water is available to permit recirculation cooling flow to the cord (See Figure B 3/4.6.2-1). The OPERABILITY of the l 4 suppression chamber in OPERATIONAL CONDITIONS 1, 2 or 3 is required by j 4 r Specification 3.6.2.1. Repair work might require making the suppression chamber inoperable. I This specification will permit those repairs to be made and at the same time i J give assurance that the irradiated fuel has an adequate cooling water supply I when the suppression chamber must be made inoperable, including draining, in l OPERATIONAL CONDITION 4 or 5. j ] In OPERATIONAL CONDITION 4 and 5 the suppression chamber minimum required i water volume is reduced because the reactor coolant is maintained at or below l 200*F. Since pressure suppression is not required below 212'F the minimum water volume is based on NPSH, recirculation volume, vortex prevention plus a j 2'-4" safety margin for conservatism. I LA SALLE - UNIT 2 B 3/4 5-2 Amendment No. 39 I l CONTAINMENT SYSTEMS BASES I 3/4.6.2 DEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 45 psig during primary system blowdown from full operating pressure. l The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system. The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1020 psig. Since all of the gases in the drywell are purged into the i suppression chamber air space during a loss ef coolant accident, the pressure j of the liquid must not exceed 45 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant and to be considered is discharged to the suppression chamber and that the drywell i volume is purged to the suppression chamber. (See Figure B 3/4.6.2-1) Using the minimum or maximum water volumes given in this specification, containment pressure during the design basis accident is approximately 39.6 psig which is below the design pressure of 45 psig. Maximum water volume of 3 131,900 ft resultsjnadowncomersubmergenceof12.4ftandtheminimum l volume of 128,800 ft results in a submergence approximately 8 inches less. The majority of the Bogeda tests were run with a submerged length of four feet ard with complete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3. i l Under full power operating conditions, blowdown from an initial suppression j chamber water temperature of 90*F results in a water temperature of approximately 135'F immediately following blowdown which is below the 200'F used for complete condensation via T quencher devices. At this temperature and atmospheric pressure, the available NpSH exceeds that required by both the RHR and core spray pumps, thus there is no dependency on containment overpressure during the accident injection phase. l Experimental data indicates that excessive steam condensing loads can be avoided if the peak bulk temperature of the suppression pool is maintained 1 below 200*F during any period of relief valve operation with sonic conditions at the discharge exit for T quencher devices. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can ce depressurized in a timely manner to avoid the regime of potentially high i } suppression chamber loadings, j 1 l i LA SAltE UNIT 2 B 3/4 6-3 Amendment No. 39 i I il,e l Control t l Suppression Plant Room / Local Chamber Level Elevation Indication 26' 10" 700' 2" +3" High Level LCO (Volume 8 131,900 ft ) 26' 9" 700' 1" +2" High Level Alarm HPCS I Suction Valve Transfer l t 26' 7" 699' 11" 0" Instrument Zero 1 26' 4" 699' 8" -3" Low Level Alarm 26' 2 1/2" 699' 6 1/2" -4 1/2" Low Level LCO Opera-tional Condition 1, 2, i 8 or 3 (Volume 128,800 ft ) 14' 687' 4" -12' 7" Low Level LCO Opera-r tional Condition 4 or 5 8 (Volume 70,000 ft ) r l 4 [ i l 1 I P SUPPRESSION POOL LEVEL SETPOINTS BASES FIGURE B 3/4.6.2-1 i LA SALLE UNIT 2 B 3/4 6-3a Amendment No. 39 ) t