ML20082E618

Proposed Tech Spec Changes,Section 4.5.2.d & Table 3.3-10 Re DHR Isolation Valves

ML20082E618
Person / Time
Site:	Davis Besse
Issue date:	11/21/1983
From:	TOLEDO EDISON CO.
To:
Shared Package
ML20082E602	List: ML20082E599 ML20082E614 ML20082E618
References
TAC-53297, TAC-53298, NUDOCS 8311280257
Download: ML20082E618 (5)

Text

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T umm M Wta TABLE 3.3-10 (Continued)

> E "_

ota c POST-ACCIDENT _HONITORING INSTR 0HENTATION 80 5 H o ~

lilNIMUM CHANNELS 88 INSTRUMENT OPERABLE ,-

8: .

g 15. Low Pressure Injection (DHR) Flow 1/ Channel m e-

16. IIPI System Pump and Valve Status 1/ System

17. LPI System Pump and Valve Status 1/ System

$ 18. Containment Spray Pump and Valve Status 1/ System .

19. Core Flood' Valve Status 1/ System .

20. BWST Valve Status 1/ System

21. Containment Emergency Sump Valve Status 1/ Valve ,

22. Containment Air Recirculation Fan Status 1/ Fan

23. Containment Air Cooling Fan Status 1/ Fan

24. EVS Fan and Damper Status 1/ System

25. Auxiliary Feedwater Flow Rate _

SMSteamGenerator .

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ko 26. RC System Subcooling Margin tionitor i -

27. PORV [losition Indicator j

28. PORV Block Valve Position Indicator 1 u 29. Safety Valve Position Indicator 1/ Valve . ..

30. BUST Ievel 3

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Docket No. 50-346 License No. NPF-3 Serial No. 997 November 21, 1983 Attachment 11 I. Changes to Davis-Besse Nuclear Power Station Unit 1, Appendix A Technical Specifications Table 3.3-10.

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

B. Reason for Change (Facility Change Request 83-123).

Toledo Edison was requested on September 2, 1983 (per telecon) to add the non-safety grade auxiliary feedwater flow rate post-accident monitoring instrumentation channel per steam generator. This request also is in compliance with requested instrumentation per NUREG 0737 Item II.E.1.2.

C. Safety Evaluation (See Attached)

D. Significant Hazard Consideration (See Attached)

SAFETY EVALUATION This amendment request is to revise Table 3.3-10 " Post-Accident Monitoring Instrument" to include two (2) Auxiliary Feedwater Flow (AFW) rate instruments per steam generator.

There is, at present, one safety grade AFW flow indicator for each steam generator. The safety grade indicators are powered from redundant lE sources and are included in the technical specifications. In addition, there is a second non-safety grade AFW flow indicator on each steam generator. These indicators are powered by redundant non-1E uninterruptable power sources.

The only safety function of these AFW flow indicators is to provide the operators with a back up indication that the AFW system flow condition is normal. The primary indicators for the operator to make this determina-tion are the safety grade steam generator level indicators. The level indicators are also used by the operators when the AFW system is on manual control. At no time does the operator use AFW flow indicators as the primary means to control the AFW system. The fact that one AFW flow indication on each steam generator is not safety grade will not degrade the safety of the AFW system or the station. The sensors for the non-safety grade flow indicators are installed externally to the pipe in a manner that does not degrade the piping classification or qualification established for the system.

The safety function of the AFW flow indicator technical specification is to assure the operability of the indicators. This change to the technical specification will assure that both the safety grade and non-safety grade indicators on each steam generator are tested periodically and are in operation in Modes 1, 2 and 3.

Therefore, this is not an unreviewed safety question.

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SIGNIFICANT HAZARD CONSIDERATION The proposed amendment request does not contain a significant hazard. The l

request is to revise the Post-Accident Monitoring Instrumention table to include two (2) Auxiliary Feedwater (AFW) Flow rate indication channels per rteam generator. This is an additional flow rate indication per steam ,

I generator.

Davis-Besse has installed and included in the technical specification one safety grade AFW flow indication per steam generator. The proposed Revision to the License provides one non-safety grade AFW flow indicator per steam generator. This would provide one safety and one non-safety grade AFW flow indication in the control room per steam generator.

The Commission has provided guidance concerning the application of the standards in 10 CFR 50.92 by providing certain examples (48 FR 14870).

One of the examples of actions involving no significant hazards considera-tions relates to a change that constitutes an additional limitation, restriction or control not presently included in the technical specifications.

The above amendment request is an additional restriction by requiring two AFW flow rate monitoring instrumentation channels (one cafety grade and one non-safety grade) per steam generator to be operable in Modes 1, 2 and 3.

Therefore based on the above and the attached safety evaluation this is not a significant hazard.

cj c/11

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ISURVEILLANCE REOUIREMENTS l ,

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'b. At least once per 18 months, or prior to operation after ECCS piping has been drained by verifying that the ECCS piping is full of water by venting the ECCS pump casings and. discharge piping high points. -

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c. By a visual inspection which verifies that no loose hebris

(rags, trash, clothing, etc.) is present in the containment which could be transported to the containment emergency sump and cause restricticn of the pump suction during LOCA conditions.

This visual inspection shall be perfqrmed:

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1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2. Of the areas affected within containment at the completion of each containment entry when CCNTAINMENT INTEGRITY is established.

d. At least once per 18 months by:

1.' Verifying that the interlocks:

a) Close OH-11 and DH-12 and deenergize the ' pressurizer heaters, if either DH-ll or DH-12 is open and a simulated reactor coolant 7 tem pressure which is greater than the trip sc int (<438 esig) is applied.

l'Ke (mhdeia to CQese DN-il Mcf 47 f DH-12. is met-

.,.e.yed Q M valae is c.Lcsed a"wl 480 V A c pcsa<v is itiscc+veerect ]rcm C no % CWOC- ~~

b) Prevent the opening of DN-il and OH-12 when a simulate'c or actual reactor coolant system presure which is greater than the trip. setpoint (<43B psig) is applied.

2. a) A visual inspection of the containment emergency sump l l i which verifies that the subsystem suction inlets are not

! i restricted. by debris and that the sump components (trash

! I racks, screens, etc.) show no evidence of structural

!, distress or corrosion.

j b) Verifying that on a Borated Water Storage Tank (bWST)

Low-Low Level interlock trip, the BWST Outlet Valve

! HV-CH7A (HV-CH7B) automatically close in <75 seconds j

after the operator manually pushes the coiitrol switen to

, i open the Containment Emergency Sump Valve HV-DH9A (HV-CH93) l which should be verified to open in 175 seconds.

3. Verifying a total leak rate 1 02 gallons per hour for the LPI system at:

a) Normal operating pressure or hydrostatic test pressure of 3,150 psig for those parts of the system dcwnstream o .

of the pump suction isolation valve, ano

, b) 145 psig for the piping frem the centainment emergency sump isolation valve to the pump suction isolaticn valve.

L CAVIS-BESSE, UNIT 1 3/4 5-4 Amencment .No. 2, 25, 25,C I

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