ML20214Q715

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Proposed Tech Spec,Changing Table 3.2.B,Sections 4.5.E.1.a & 3.2 to Reflect Mods Made to Facility to Address NUREG-0737,Item II.K.3.18
ML20214Q715
Person / Time
Site: Pilgrim
Issue date: 05/20/1987
From:
BOSTON EDISON CO.
To:
Shared Package
ML20214Q694 List:
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.18, TASK-TM NUDOCS 8706050127
Download: ML20214Q715 (6)
v • d • e


Text

PNPS TABLE 3.2.8 (Cont'd)

INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE A!10 CONTAINMENT COOLING SYSTEMS Minimum # of Operable Instrument Channels Per Trio System (1) Trio Function Trio Level Settina Remarks 1 Core Spray Pump Start In conjunction with. loss of Timer 0<t<l sec. power initiates sequential 1 LPCI Pump Start Timer 4<t<6 sec. starting of.CSCS pumps 1 LPCI Pump Start Timer 9<t<11 sec.

1 Auto Blowdown Timer 190, 1120 sec. In conjunction with Low Low Reactor Water Level, High Drywell Pressure and LPCI or

  • Core Spray Pump running interlock, initiates Auto Blowdown.

2 ADS Drywell Pressure 11 i 2 min. Permits starting CS 4

Bypass Timer and LPCI pumps and actuating ADS SRV's ir RPV

. water level is low and drywell pressure is not high.

2 RHR (LPCI) Pump Discharge 150 1 10 psig Defers ADS actuation pending Pressure Interlock confirmation of Low Pressure core cooling system operation.

J 2 Core Spray Pump Discharge 150 10 psig (LPCI or Core Spray Pump Pressure Interlock running interlock.)

2 Emergency Bus Voltage 20-25% of rated 1. Permits closure of the Relay voltage resets Diesel Generator to an at less than 50% unloaded emergency bus.

2. Permits starting of CSCS 4 kV motors.

i ~ B706050127 870520 3 i DR ADOCK 05 1

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LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENT

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3.5.0 Reactor Core Isolation Cooling 4.5.0 Reactor Core Isolation on Cooling (RCIC) Subsystem (Cont'd) LRXIf)Ahsystem (Cont'd)

d. Flow Rate at Once/3 months 1000 psig
e. Flow Rate at Once/ operating 150 psig cycle
2. From and after the date that the The RCIC pump shall deliver at RCICS is made or found to be least 400 gpm for a system head inoperable for any reason. corresponding to a reactor continued reactor power operation is permissible only pressure of 1000 to 150 psig.

during the succeeding seven days 2.

provided that during such seven When it is determined that the RCIC subsystem is inoperable, the HPCIS days the HPCIS is operable.

shall be demonstrated to be

3. If the requirements of 3.5.D operable immediately and weekly cannot be met, an orderly thereafter.

N shutdown shall be initiated and the reactor pressure shall be reduced to or below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.5.E Automatic Deoressurization System 4.5.E Automatic Deoressurization System (ADS)

(ADS)

1. The Automatic Depressurization 1. During each operating cycle the Sub. system shall be operable following tests shall be performed when'cVer there is irradiated on the ADS:

fuel in the reactor vessel and the reactor pressure is greater- a. A simulated automatic actuation than 104 psig and prior to a test shall be performed prior to startup from a Cold Condition, startup after each refueling except as specified in 3.5.E.2 outage. The ADS manual inhibit below.

switch will be included in this test.

b. With the reactor at pressure, each relief valve shall be manually opened until a corresponding change in reactor pressure or main turbine bypass valve positions indicate that steam is flowing from the valve.
c. Perform a test from the alternate shutdown panel to verify that the relief valve solenoids actuate. Test shall be performed after each refueling outage prior to startup.

Amendment No. 109

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.S

3.2 . BASES (Cont'd)-

For most parameters monitored, as listed in Table 3.2.F. there are two (2) channels of instrumentation. By comparing readings between these two (2) channels, a near continuous surveillance of instrument performance is available. Meaningful deviation in comparative rea' dings of these instruments will initiate an early recalibration, thereby maintaining the quality of the instrument readings.

The Safety --Safety / Relief Valve position indication instrumentation provides the operator with information on selected plant parameters to monitor and assess these variables during and following an accident.

, . In response to NUREG-0737, modifications were made to the ADS logic to extend automatic ADS operation to a class of transients that involve slowly uncovering the core without depressurizing the vessel or pressurizing the drywell. These transients were analyzed assuming no.

high pressure injection systems (feedwater, HPCI or RCIC) are available. Only ADS is considered'available to depressurize the vessel,

, permitting operation of LPCI. The transients generally involve pipe t

breaks outside containment. Automatic ADS would not occur on low water

] level because high drywell pressure would not be present and ADS logic

) has a high drywell pressure permissive. The modification added a timer i

to the ADS logic which bypasses the high drywell pressure permissive, and a manual inhibit switch which allows the operator to inhibit automatic ADS initiation for events where automatic initiation is not desirable.

An analysis was performed to determine an upper time limit on the bypass timer. The goal was to ensure ADS is automatically initiated in time to prevent peak clad temperature (PCT) from exceeding 1500*F for a limiting

!- break, which was determir.ed to be a Reactor ' Hater Cleanup line break.

The analysis concluded that there are 18 minutes between the low water
level initiation of the timer and the heatup of the cladding to the limit. Since the logic includes a 2 minute delay already, the bypass timer upper limit can not be more than 16 minutes, which provides a conservative margin for PCT and allows sufficient time for operator intervention if required. A minimum time delay is incorporated to allow

! RPV water level to recover, resetting the timer and preventing depressurization. The choice of a timer setting of 11 minutes places

! the setting in the middle and provides maximum tolerance from either

(

Reference:

GE Report " Bypass Timer Calculation for the limit.

ADS /ECCS Modification for Pilgrim Station" December 16, 1986).

The. recirculation pump trip / alternate rod insertion systems are

i. consistent with the "Monticello RPT/ARI" design described in NED0-25016 (Reference 1) as referenced by the NRC as an acceptable design
(Reference 2) for RPT. Reference 1 provides both system descriptions
. and performance analyses. The pump trip is provided to minimize reactor l pressure in the highly unlikely event of a plant transient coincident i~ with the failure of all control rods te scram. The rapid flow reduction
increases core voiding providing a negative reactivity feedback. High pressure sensors and low water level sensors initiate the trip. The Amendment No. 73 i

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3.2 BASES (C@nt'd) recirculation pump trip is only required at high reactor power levels, where the safety / relief valves have insufficient capacity to relieve the steam iiich cethues to be generated in this unlikely postulated

-snt. Requiring the trip to ba operable only when in the RUN mode is tnerefore conservative. The low water level trip function includes a time delay of nine (9) seconds one (1) second to avoid increasing the consequences of a postulated LOCA. This delay has an insignificant effect on ATHS consequences.

Alternate rod insertion utilizes the same initiation logic and functions as RPT and provides a diverse means of initiating a reactor scram. ARI uses sensor diverse from the reactor protection system to depressurize the scram pilot air header, which in turn causes all control rods to be inserted.

References l 1. NLDO-25016, " Evaluation of Anticipated fransients Without Scram for tle Monticello Nuclear Generating Plant," September 1976.

2, NUREG-0460, Volume 3, December 1978.

Amendment No. 73a

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  • 3.2 BASES (Cont'd)

Drvwell Temoerature The drywell temperature limitations of Specification 3.2.H.1 ensure that safety related equipment will not be subjected to excess temperature.

Exposure to excessive temperatures may degrade equipment and can cause loss of its operability.

The temperature elements for monitoring drywell temperature specified in Table 3.2.H were chosen on the basis of their reliability, location, and their redundancy (duel - element RTD's). These temperature elements are the primary elements used for the PCILRT.

The temperature limits specified in 3.2.H.1 are based on the BECo report entitled Drywell Temperature Report, dated January 28, 1982. The limits derived from this report take into consideration the long-term effects of ambient temperature on equipment design limits and material degradation of components required for accident mitigation or plant shutdown. The evaluation process addressed the actual assessment of potential damage and the determination of equipment status from the standpoint of both qualification integrity (for safety-related equipment) and reliability to perform it's intended function.

If the drywell temperature exceeds the limits specified in 3.2.H.1 an engineering evaluation must be initiated in order to determine whether any safety related component has been adversely affected.

The limiting drywell temperature value of 215'F (Section 3.2.H.2) was selected as to guarantee that ECCS trips occur on/or before present Technical Specification values.

The time interval of 30 minttes between successive drywell temperature instrument readings (Section 3.2.H.1) was selected so as to guarantee that ECCS trips occur on/before present Technical Specification values in the event of a drywell temperature excursion in excess of 215*F.

The instrument check interval of once per shift provides adequate assurance of equipment operability based upon engineering judgement.

Amendment No. 73b

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