ML20101J768

Proposed Tech Specs,Modifying Statements in TS & Bases to Correctly Reflect Ref Parameter for Anticipatory Scram Signal Bypass

ML20101J768
Person / Time
Site:	Oyster Creek
Issue date:	03/28/1996
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20101J765	List: ML20101J762 ML20101J768
References
NUDOCS 9604020135
Download: ML20101J768 (2)

Text

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I High drywell pressure provides a second means of initiating the core spray to mitigate the l consequences of loss-of-coolant accident. Its trip setting of13.5 psig initiates the core spray in time to provide adequate core cooling. The break size coverage of high drywell pressure was discussed above. Low-low water level and high drywell pressure in addition to initiating core spray also causes isolation valve closure. These settings are adequate to cause isolation to minimize the offsite dose within required limits.

It is permissible to make the drywell pressure instrument channels inoperable during performance.

of the integrated primary containment leakage rate test provided the reactor is in the COLD SHUTDOWN condition. The reason for this is that the Engineered Safety Features, which are effective in case of a LOCA under these conditions, will still be effective i because they will be activated (when the Engineered Safety Features system is required as  !

identified in the technical specification of the system) by low-low reactor water level.*

The scram discharge volume has two separate instrument volumes utilized to detect water accumulation. The high water level is based on the design that the water in the SDIV's, as l i

detected by either set of level instruments, shall not be allowed to exceed 29.0 gallons; thereby, permitting 137 control rods to scram. To provide further margin, an accumulation of not more than 14.0 gallons of water, as detected by either instrument volume, will result in a rod block and an alarm. The accumulation of not more than 7.0 gallons of water, as detected in either instrument volume will result in an alarm.

Detailed analyses of transients have shown that sufficient protection is provided by other scrams j below 45% power to permit bypassing of the turbine trip and generator load rejection scrams.

~ However, for operational convenience,40% of rated reactor thermal power has been chosen as l the setpoint below which these trips are bypassed. This setpoint is coincident with bypass valve capacity.

A low condenser vacuum scram trip of 20 inches Hg has been provided to protect the main condenser in the event that vacuum is lost. A loss of condenser vacuum would cause the turbine stop valves to close, resulting in a turbine trip transient.

The low condenser vacuum trip provides a reliable backup to the turbine trip. Thus, if there is j a failure of the turbine trip on low vacuum, the reactor would automatically scram at 20 inches j Hg. The condenser is capable of receiving bypass steam until 7 inches Hg vacuum thereby mitigating the transient and providing a margin.

The settings to isolate the isolation condenser in the event of a break in the steam or condensate lines are based on the predicted maximum flows that these systems would experience during operation, thus permitting operation while affording protection in the event of a break. The settings correspond to a flow rate of less than three times the normal flow rate of 3.2X10' lb/hr.

Upon initiation of the alternate shutdown panel, this function is bypassed to prevent spurious isolation due to fire induced circuit faults.

OYSTER CREEK 3.1-5 Amendment No.: 73, 79, 112, 149, 171, 176

• Correction 11/30/87 9604020135 960329 PDR ADOCK 05000219 P PDR

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TABLE 3.1.1 (CONT'D)

Action required when minimum conditions for operation are not satisfied.

3 Also permissible to trip inoperable trip system. A channel may be placed '

in an inoperable status for up to six hours for required surveillance without i placing the trip system in the tripped condition provided at least one l OPERABLE instrument channel in the same trip system is monitoring that parameter. l l

• • 1 See Specification 2.3 for Limiting Safety System Settings. I I

Notes: l

a. Permissible to bypass, with control rod block, for reactor protection system reset in REFUEL MODE.

b. Permissible to bypass below 800 psia in REFUEL and STARTUP MODES.  ;

c. One (1) APRM in each OPERABLE trip system may be bypassed or inoperable provided  !

the requirements of Specification 3.1.C and 3.10.C are satisfied. Two APRM's in the i same quadrant shall not be concurrently bypassed except as noted below or permitted by note.

Any one APRM may be removed from service for up to six hours for test or calibration without inserting trips in its trip system only if the remaining OPERABLE APRM's meet the requirements of Specification 3.1.B.1 and no control rods are moved outward during the calibration or test. During this short period,' the requirements of Specifications 3.1.B.2, 3.1.C and 3.10.C need not be met.

d. The IRMs shall be inserted and OPERABLE until the APRMs are OPERABLE and reading at least 2/150 full scale. _ _

e. Offgas system isolation trip set at _< 2.1/E Ci/sec where E = average gamma energy from noble gas in offgas after holdup line (Mev). Air ejector isolation valve closure time delay shall not exceed 15 minutes.

f. Unless SRM chambers are fully inserted.

g. Not applicable when IRM on lowest range.

h. One instrument channel in each trip system may be inoperable provided the circuit which it operates in the trip system is placed in a rimulated tripped condition. If repairs cannot be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the reactor shall be PLACED IN Tile COLD SHUTDOWN CONDITION. If more than one instrument channel in any trip system becomes inoperable, the reactor shall be PLACED IN TIIE COLD SHUTDOWN CONDITION.

Relief valve controllers shall not be bypassed for more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (total time for all controllers) in any 30-day period and only one relief valve controller may be bypassed at a time.

i. The interlock is not required during the start-up te!.t program and demonstration of plant electrical output but shall be provided following these actions.

j. Not required below 40% of rated reactor thermal power. j OYSTER CREEK 3.1-16 Amendment No.: 75,108,110,171