ML20073R678

Proposed Changes to App a to Tech Specs Revising Tables 4.3-6 & 4.3-10 Re Extension of Surveillance Interval for Steam Generator Outlet Pressure Instrumentation

ML20073R678
Person / Time
Site:	Davis Besse
Issue date:	05/02/1983
From:	TOLEDO EDISON CO.
To:
Shared Package
ML20073R670	List: ML20073R667 ML20073R678
References
NUDOCS 8305040480
Download: ML20073R678 (4)

Text

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Docket No. 50-346 License No. NPF-3 Serial No. 941 May 2, 1983 Attachment I I. Changes to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications Tables 4.3-6 and 4.3-10.

A. Time required to implement. This change is to be effective upon NRC approval.

B. Reason for Change (Facility Change Request 83-037). To allow power operation until the schedule refueling outage in September, 1983. The Surveillance Test requires containment entry and cannot be accomplished during power operation.

C. Safety Evaluation. (See Attached) 8305040400 830502 PDR ADOCK 05000346 p PM

Safety Evaluation This amendment request is for a one-time extension of the Tech. Spec.

(Section 4.3.3.5, Table 4.3-6 and Section 4.3.3.6, Table 4.3-10)

Surveillance Testing requirements on the 3.25 criteria for the steam generator outlet pressure instrumentation from 5/17/83 to 9/17/83.

The safety function of the steam generator outlet pressure instrumentation is to allow the operators to monitor shutdown and post accident conditions both in and out of the control room.

The last Surveillance Testing for the steam generator outlet pressure instruments was performed on 11/19/81. Per Tech. Spec. requirements, the next Surveillance Testing for these transmitters shall be performed by 5/17/83 which is before the projected 1983 refueling outage (presently scheduled to start 7/29/83). The extension is needed to allow continued plant operation until refueling.

These pressure instrument strings were calibrated on 7/13/77, 6/25/78, 10/1/80, and 11/19/81. No calibration was performed between 6/25/78 and 10/1/80 (the station had a 7 month refueling outage during that period of time).

The "as found" and "as left" condition for these instruments from all previous testings / calibrations were carefully examined and found to be less than the FSAR tolerances allowed (12%). This means the drifts on these instruments are not excessive during 18 months of operations, and can be expected to perform satisfactorily for four more months of operation. These instruments are only used to provide indications without automatic initiation of any safety equipment.

The Surveillance Test due date for the entire instrument strings is delayed to 9/17/83. Partial implementation of the Surveillance Testing before 5/17/83 by performing calibration of those instrument modules located outside of the containment will not be necessary, since the risk and consequence of losing these transmitters far outweights the I benefit of the limited reduction on the inaccuracies.

l l Pursuant to the above, it is concluded that the plant can continue l operation until 9/17/83 to perform the Surveillance Testing of these i pressure instruments and there is no unreviewed safety questions

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TABLE 4.3-10 .

o POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

D* CHANNEL CHANNEL

&. INSTRUMENT CHECK CALIBRATION

1. SG Outlet Steam Pressure M R

2. RC Loop Outlet Temperature M R E

- 3. RC Loop Pressure M R

4. Pressurizer Level M R

5. SG Startup Range Level M R Auxiliary Feedwater Status ' M NA 6.

7. Containment Vessel Hydrogen M R R

• R

8. Containment Vessel Post-Accident. Radiation M T NA

$ 9. Containment Vessel Isolation Status M .

SFAS Status M NA

10. .

Safety Features Equipment Status M NA 11.

M NA

12. RPS Status

13. SFRCS Status- H NA ,

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14. High Pressure Injection Flow M R <

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TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS T CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

."' l. Reactor Trip Breaker Indication M N.A.

2. Reactor Coolant Temperature-Hot Legs M R

3. Reactor Coolant System Pressure M R

4. Pressurizer Level M R

5. Steam Generator Outlet Steam Presstfre M

6. Steam Generator Startup Range Level M R

7. Control Rod Position Limit Switches M R ggg .

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