ML20210B530

Forwards Westinghouse Evaluation Demonstrating That Accident Analysis for Damage of Centrifugal Charging Pumps Following Secondary Sys High Energy Line Rupture Remains Valid After Mods.No Further Mods Planned Based on IE Bulletin 80-18

ML20210B530
Person / Time
Site:	Summer
Issue date:	04/29/1987
From:	Nauman D SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
TASK-2.K.3.05, TASK-TM GL-83-10, GL-85-12, IEB-80-18, NUDOCS 8705050349
Download: ML20210B530 (7)

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nucwar opeoons c g a g s ta SCEAG April 29, 1987 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Charging Pump Miniflow Modification

Dear Mr. Denton:

l South Carolina Electric & Gas Company (SCE&G) considers the action described in our letter dated January 22, 1982, in response to IEB 80-18, " Maintenance of Adequate Minimum Flow Through Centrifugal Charging Pumps Following Secondary Side High l

Energy Line Rupture," adequate to address the concerns identified in the Bulletin.

The following actions were taken in response to the IEB:

1. The safety injection automatic closure signal for the Centrifugal Charging l Pumps (CCP) miniflow isolation valves has been removed. Miniflow is now aligned to the Volume Control Tank (VCT) and the VCT relief valve has been verified operable.

2. Emergency Operating Procedures have been changed requiring the operator to:

(a) close the miniflow isolation valves for each pump if pressure decreases to 1380 psig and (b) to reopen the valves if pressure increases to 2000 psig or flow decreases to less than 200 gpm per running pump.

The operator actions described above have been justified by Westinghouse in a .

generic evaluation transmitted to SCE&G by letter CGWS-1047 dated July 16, 1980 l (Attachment 1 enclosed). Since this evaluation, SCE&G has removed the Baron 1 Injection Tank (BIT) and performed an additional analysis for a Secondary System Rupture (page 2. Item B, of the attached) which shows core protection in a

" credible" steamline rupture, even though the reactor may return to criticality after a reactor trip. Operator action required to isolate miniflow during a LOCA is not required until 10 minutes into the event.

1 The operator action is initiated by the Reactor Coolant Pump trip criteria which ensures the event is a LOCA and not a steamline, feedline or steam generator tube l rupture. The initiating criteria for operator action to trip Reactor Coolant pumps has been evaluated and accepted by the NRC. This information was provided in the i SCE&G response to TH! Action Item !!.K.3.5, Generic Letter 83-10 and Generic letter I i 85-12.

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0705050349 070429 PDR G

ADOCK 05000395 PDR f(0 \\

Mr. Harold R. 0:nton April 29, 1987

. Page 2 The valves required to be closed are powered from diesel generator backed motor control centers. The individual pump mintflow valves are powered from (B) train power. A common isolation valve is available as a backup to the individual pump .

miniflow isolation valves. This common valve is powered from the "A" train diesel generator backed motor control centers. The RCS pressure indicators, used to determine when to isclate miniflow, are powered from separate redundant power sources. Each power supply is fed from an inverter which is powered by diesel l

backed AC or DC which allows continuous uninterrupted indication.

The attached Westinghouse evaluation demonstrates that the accident analysis remains valid for the analyzed events. Since the accident analysis criteria is j more conservative than the Technical Specifications bases, the Technical Specifications remain valid.

l Based on the above, SCE&G plans no further modification based on IEB 80-18.

! If you should have any further questions, please advise.

! Very truly yours, 1

, D. A. Nauman RJB: DAN /bjh

Attachment

, c: 0. W. Dixon, Jr./T. C. Nichols, Jr. R. A. Stough

! E. C. Roberts G. O. Percival i

0. S. Bradham K. S. West J. G. Connelly, Jr. R. L. Prevatte D. R. Moore J. B. Knotts, Jr.

W. A. Williams, Jr. I & E Washington Group Managers NPCF W. R. Baehr File C. A. Price C.L.Ligon(NSRC)

R. M. Campbell K. E. Nodland i

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0 AR%achmen2 ? CGWS-1047 July 16, 1980 CEtiTRIFUGAL CHARGING PUMP OPERATION

! FOLLOWING SECCiOARY SIDE HIGH E!1ERGY LINE RUPTURE Reference 1: NS-TMA-2245,5/8/80 Reference 1notifiedtheNRQ,ofaconcernforconsequentialdamageof one or more centrifugal charging pumps (CCP) following a secondary system high energy line rupture. Reference 1 included a calculational method i

and sample calculation to permit evaluation of this concern on a plant specific basis. Should a plant specific problem be identified, Westinghouse

' provided several recommendations for the interim until necessary design modifications'can be implemented to resolve the problem. These recomenda-tions included two proposed interim modifications which included:

1. Remove the safety injection initiation automatic closure signal from the CCP miniflow isolation valves.

2. Modify plant emergency operating procedures to instruct the operator to:

a. Close the CCP miniflow isolation valves when the actual RCS pressure drops to the calculated pressure for manual reactor i coolant pump trip.

b. Reopen the CCP miniflow isolation valves should the wide range RCS pressure subsequently rise to greater than 2000 psig.

Prior to making this recommendation, Westinghouse evaluated the impact of l the recomended operating procedure modifications on the results of the various accidents which initiate safety injection and are sensitive to CCP l

flow delivery. The accidents evaluated in dotati include secondary system ruptures and the spectrum of small loss of coolant accidents. The analytical '

results for steam generator tube rupture and large loss of coolant accident l are not sensitive to a reduction in CCP flow of the magnitude that results from the reconinended modifications.

This letter functions to supplement Reference 1 and identify the sensitivity of the accident analyses to

-l the recommended modifications. This evaluation is generic in nature.

2 A%2achmene 1 Secondary System Rupture Sensitivity analyses have been perforced for secondary high energy line ruptures to evaluate the impact of reduced safety injection flew due to normally open miniflow isolation valves. These analyses indicate an insignificant effect on the plant transient response.

A. Feedline Rupture Following a feedline rupture, the reactor coolant pressure will reach the pressurizar safety valve setpoint within approximately 100 seconds assuming maximum safeguards with the power-operated relief valves inoperable. With minimum safeguards, the reactor coolant pressure will not reach the pressurizer safety valve setpoint until approximately 300 seconds. The time that the reactor coolant system pressure remains at the pressurizer safety valve setpoint is a function of the auxiliary feedwater flow injected into the non-faulted steam generators and the time at which the operator is assumed to take action. With the mini-flow isolation valves open, the peak reactor coolant system pressure and the water discharged via the pressurizer safety valves are insigniff-cantly changed from the FSAR results, s

8. Steamline Rupture The effects of maintaining the miniflow isolation valves in a normally open position was also investigated following a main steamline rupture.

For the condition !! " credible" steamline rupture, the results of the transient with the miniflow valves open showed that the licensing criterion (no return to criticality after reactor trip) continues to i

be met. The condition !!! and IV main steamline ruptures were also reenalyzed assuming the miniflow valves were open. The results of

, the analysis showed that, even with reduced safety injection flow into the core, no ON8 occurred for any rupture.

> 3 Attachment 1 Small less of Ceolant Accidents Sensitivity analyses have been performed to evaluate the impact of reduced safetyinjectionflowons$allbreaklossofcoolantaccidents(LOCAs).

These analyses indicated that miniflow isolation can be delayed, but it

, must occur at some time into the small break LOCA transient in order to ,

limit the peak clad temperature (PCT) penalty.

The proposed modification delays miniflow isolation and reduces $1 flow

' delivered by approximately 45 gpm at 1250 psia during the delay time period.

The impact of this modification was evaluated based on two isolation times

1) The time' equivalent to the RCP trip time, and 2) approximately 10 minutes in the transient, or just prior to system drain to the break for the worst 1 small break sizes. The second time was evaluated to determine the impact if the operator does not isolate miniflow within the proposed prescribed time. The spectrum of small break sizes are considered to encompass all 4

possible small break scenarios. Only cold leg break locations are considered

! since they will continue to be limiting in tems of PCT.

A. Very small breaks that do not drain the RCS or ;.ncover the core, and

! maintain RCS pressure above secondary pressure (< $2" diameter).

For these break sizes, it is quite possible that the operator may l

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i never isolate the mintflow line, since the pressure setpoint will not be reached, and continued pumped SI degradation will persist.

However, this will have no adverse consequences in tems of core uncovery and PCT. No core uncovery will be expected for the degraded

$1 case, similarly to the base comparison case with full St. The only effect would be a slightly lower equilibration pressure for a

! given break size.

B. Small breaks that drain the RCS and result in the maximum cladding f

temperatures (2"< diameter <6").

This range of break sizes represents the worst small break size for 4 ,

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~4* Attachment 1 i.

l most plants as determined utilizing the currently approved October 1975 Evaluation Model version, as shown in WCAP-8970-P-A. If miniflow is f '

isolated at the RCP trip setpoint rather than the "S" signal, a reduc-l tion in safety injection flow of less than 45 gpm results, averaged l

< for the approximately 40 secord period of time separating the two events. ,

i This reduction in RCS liquid inventory results in core uncovery less than one second earlier, and has a negligible impact on PCT. If mini-f flow is. isolated at the time of core uncovery, or approximately 10 j

minutes for break sizes in this range, a greater reduction in RCS liquid

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inventory results in a core uncovery 10 seconds earlier in the transients resulting in less than a 10*F PCT penalty for the worst size small break.

fl This would not result in any present F5AR small break analysis becoming l '

j more limiting than the corresponding large break LOCA F5AR analysis.

If miniflow isolation does not occur at any time into the transient for this categor; of small LOCA, a PCT penalty of 200*F or more could occur.

C. Small break sizes larger than the worst break through the intermediate breaksizes(y,6" diameter). >

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i Break sizes in this range have been determined to be non-limiting for i f small break utilizing the currently approved October 1975 Evaluation f j Model, WCAP-8970 P-A. If miniflow isolation occurs at the RCP trip time for these break sizes, the negligible effect on KT presented l above also applies. 51milarly, if isolation occurs prior to core

! uncovery, the sme11 (< 10*F) PCT penalty will result as well. However, for these larger break sizes, the time of first core uncovery occurs J l

prior to 10 minutes. If miniflow isolation is not performed until l l

l 10 minutes, m duced $! will be delivered during the core uncovery time, j which can have a greater impact on PCT. Studies indicate a potential j PCT penalty of 40*F resulting for these non-limiting bmak sizes if

miniflow is not isolated until 10 minutes. This is not espected to shift the worst break size to larger breaks, since these breaks are typically hundreds of degrees less than smelter limiting small breaks analysed with the currently approved Evaluation Model, i

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Attac eene 1 For all FSAR small LOCA analyses, one complete train failure is assumed. It i is clear that two charging pumps without miniflow isolation provides more flow than one pump with miniflow isolation. The impact presented in this

, evaluation maintains the one train failure and assumes no miniflow isola-tion for the remaining pump. If both pumps were operating, the PCT results would be much lower than p' resent FSAR calculations even if miniflow isola-tion is not assumed to occur for the two pump case. In this situation, the plant FSAR small break calculation's remain censervative.

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i These sensitivity studies form the basis for the recormended interim modifications to the emergency operating procedures. The accidents evalu

ated are relatively insensitive to the recosmonded modifications. Further, I the accidents evaluated will give results that satisfy acceptance criteria l as long as the CCP miniflow is isolated within 10 minutes of event initiation.

j However, small LOCA sensitivity studies with one Si train operating confirm that small LOCA analyses require mintflow isolation within 10 minutes.

To comply with the recormended modifications, the c;:erator can isolate mini-

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i flow at any point in the depressurization transient prior to RC5 pressure reaching the RCP trip setpoint. Should a repressurization transient occur, the operator can open CCP mintflow at any point between the RCP trip set-

, point and 2000 psig. Such operator actions will, ensure that plant accidents satisfy acceptance criteria and protect the CCPs from consequential damage during the repressurization transient that accompanies a secondary system high energy line rupture at high initial power levels.

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