RBG-24405, Forwards Addl Info Re 860917 Application for Amend to License NPF-47,changing RCIC Sys Tech Specs as Result of 860908 Valve Inoperability Event.Fuel Failure Not Expected Due to Provision of Makeup Water from Feedwater or HPCS Sys

Forwards Addl Info Re 860917 Application for Amend to License NPF-47,changing RCIC Sys Tech Specs as Result of 860908 Valve Inoperability Event.Fuel Failure Not Expected Due to Provision of Makeup Water from Feedwater or HPCS Sys

ML20215F712
Person / Time
Site:	River Bend
Issue date:	09/19/1986
From:	Deddens J GULF STATES UTILITIES CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RBG-24405, RBG-24419, NUDOCS 8610160342
Download: ML20215F712 (6)

Text

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G UEaF STATRiS UTILITIES COMPANY RIVER BENO STATION POST OFFICE BOX 220 ST. FhNCISVILLE, LOUISIANA 70775 AREA CODE 504 635 6094 346-8651 September 19, 1986 RBG- 24419 File No. G9.5 Mr. Harold R. Denton, Director office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Denton:

River Bend Station - Unit 1 Docket No. 50-458 As requested by your Staff, Gulf States Utilities (GSU) is providing below supplemental information to our application for amendment to the River Bend Station -

Unit 1 Technical Specifications, Appendix A to Facility Operating License NPF-47 detailed in a letter (J. C. Deddens to H. R. Denton) dated September 17, 1986 (RBG-24405).

The River Bend Station Operations staff performed a functional test on the 1E51*MOVF076 annunciator at approximately 1415 on September 8, 1986 when the MOV76 breaker tripped. The initial investigation determined a' ground in the system. During the following two days, Electrical Maintenance investigated the grounding source and indicated that the motor was apparently wet.

An attempt was made to dry-out the valve motor. The operator then attempted to stroke the valve and indications revealed that the valve was fully opened. An attempt was made to reclose the valve back to its normal position at which time the MOV76 breaker tripped. During the next few days, several other attempts were made to reclose MOV76 and to establish valve operability. On September 15, 1986 Electrical Maintenance indicated that it had exhausted all means of fixing the grounding problem and GSU proceded to investigate a request for Technical Specification change. Between September 15 and September 17 GSU confirmed, by calculations and analysis, the safety significance of the proposed change.

A hand-drawn schematic that features the main components which are discussed in this request is proivded as Attachment B for your use.

The Reactor Core Isolation Cooling (RCIC) system has several small instrument lines ( 3/4") and drain lines ( 1") contained in rooms that have been analyzed for large line breaks. We have

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examined the RCIC line and have identified seven 3/4" instrument lines each with normally opened valves and a drain line with a

' normally opened valve located in the RCIC room. The analyses for these lines include a High Energy Line Break analysit in the Main Steam Tunnel and RCIC cubicle areas.

The RCIC line is common to the Residual Heat' Removal . (RHR) line

in the main steam tunnel. This common line.is designed for the steam condensing mode of operation which is not presently used at River Bend Station. This RHR line is isolated by two valves in the RHR Equipment Removal cubicle. There are two 3/4" instrument lines each with normally opened valves located in the RHR room.

Break analysis have been performed in this area including a j Moderate and High Energy Line Crack. analysis. The remaining RCIC line between MOV64 and MOV45 has been analyzed for High Energy

Line Breaks.

Assuming a break in the RCIC steam line between valves MOV64 and l

MOV45, 10CFR Part 100 limits will not be exceeded. This conclusion is ' based upon the assumptions given in Attachment 1.

Fuel failure is not expected because the make-up water from the fluid loss can be provided by the Feedwater System or the High Pressure Core Spray system. The only activity available for release from the break is that which is present in the reactor j coolant and steam lines prior to the break. Moreover, the large

break LOCA (Double-Ended Recirculation Line Break) with fuel 1

failure assumes a closed loop in the RCIC system and filtration by the Standby Gas Treatment System for any valve packing leaks i and meets the current accident analysis and 10CFR Part 100 2 limits.

I Additionally requested by your Staff was supplemental infor-1 mation concerning Item 2 of the "No Significant Hazards Evaluation":

The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated because line breaks such as may be postulated for the RCIC system have been considered in

the analysis and design basis with many of the previous analysis bounding all the short-term consequences of the j plant for this event. There may,however, be longer term

environmental effects which would only impact the overall lifespan of the equipment and would be evaluated if such an event should occur. This change introduces

, no new mode of operation and no change in the' plant design is being made.

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6 -D Therefore, there is no new or different kind of accident from any accident previously evaluated.

Sincerely, o

. C. Deddens Vice President River Bend Nuclear Group JCD/JEP/je Attachments ,

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t ATTACHMENT 1 Asgymptions For RCIC Steam Line Break

1. Leakage will be picked up by high temperature leak detection based on 125 gpm leakage which provides an isolation signal to F063 and F064. Also high steam flow indication will isolate the RCIC System.

2. Once a signal is received.to close the isolation valves, a 10 second delay for diesels to start plus a 10 second delay for valve closure will be assumed. A total of 20 seconds delay is used.

3. Assume full flow through the 8" line for 20 seconds.

At the end of 20 seconds the isolation valve F063 is assumed to close and the F064 valve fails to close (Single railure). This is conservative in that the 10 second closure time for the valve will actually ramp the flow from full flow to the 30 gpm bypass flow.

4. Moody Critical flow of 2100 lb/sec-ft2 is used for steam flow at fp00 psia will be used to calculate the flow rates.

5. The bypass flow will be held constant (30 gpm) for the first 30 minutes of the transient and then will be ramp to 0.0 in the next 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at which time the reactor will be in cold shutdown.

Calculations Mass released for first 20 seconds of = 13,400 lbm the transient i

Mass released for next 30 minutes = 7,787 lbm of the transient Mans released for next 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of = 27,254 lbm the transient

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4 Total mass from the 48,441 lbm RCIC Break -

r-From River Bend Station (RBS) FSAR, Section 15.6.4.4, the total integrated mass leaving the RPV through the steam line break is 80,562 lb . The radiological analysis for the steam line break is based on conservative assumptions considered to be acceptable to the NRC for meeting 10CFR100 guidelines and is the " Design Basis Accident" for RBS as identified in FSAR Section 15.6.4.5.

Therefore, as can be seen from the caculation, the RCIC worst case piping break with unisolated bypass flow through the F076 valve is bounded by the Main Steam Line break by a considerable margin. As previously stated, the current FSAR thyroid dose for the exclusion area from the Design Basis Accident is 29 Rem (FSAR Ta' ole 15.6-7). Comparing the masses given for each event, a conservative corrolation to the dose would be approximately 17 Rem for the RCIC event which is below the 30 Rem regulatory allowable dose.

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