ML20094M980

Forwards Summary Evaluation of Main Steam Line Break Analysis for Both Inside Containment & in Doghouse.Plant Safety Would Not Be Adversely Affected in Event of Design Basis Main Line Break in Doghouse or Inside Containment

ML20094M980
Person / Time
Site:	Mcguire, McGuire
Issue date:	08/08/1984
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
TAC-59005, TAC-59006, NUDOCS 8408160050
Download: ML20094M980 (7)

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DUKE POWER GOMPANY .

P.O. nox 33180 CHARLOTTE, N.O. 28242 HAL B. TUCKER TELEPHONE vaca rassionwr (704) 373-4531

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August 8, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: McGuire Nuclear Station Docket Nos. 50-369, -370

Dear Mr. Denton:

On July 20, 1984, a meeting was held at the request of the NRC Staff to discuss the status of the main steamline break analysis for McGuire in light of recent activities relative to the Catawba licensing. As a result of this meeting, the NRC Staff requested that Duke submit a l

summary of the evaluation that had been conducted by Duke on McGuire on this issue.

Accordingly, please find attached a summary evaluation of the main steam line break analysis, both inside containment and in the doghouse, for McGuire. Also please note that an analysis is in progress with West-inghouse to confirm the preliminary information used in this evaluation for an outside containment main steam line break. This additional analysis will be completed on a schedule consistent with and in support of Catawba licensing.

Very truly yours, 4.W '

Hal B. Tucker RLG: sib Attachment cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 W. T. Orders NRC Resident Inspector McGuire Nuclear Station 8408160050 840808 geol DR ADOCK 05000369 u

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Mr. Harold R.-Dent:n, Diract r

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cc: . Ralph Birkel _

Division of Project Management Office of Nuclear' Reactor-Regulation U.TS. Nuclear Regulatory Commission.

-Washington, D. C. -20555

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Duke Power. Company j,

7, McGuire Nuclear Station 4

Summaky Evaluatio'n of Main' Steam Line Break' Analysis s

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I. fHISTORY > -

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'  : Revised information has'been' received from Westinghouse giving-

. mass / energy release rates for a Main Steam Line Break (MSLB)'

inside containment. The original, Westinghouse information indicated a saturated steam condition from the steam generators;-however, revised information identifies steam generator tube uncovery and the

'. formation of superheated ste'am.

Duke Power had'previously assumed the same saturated steam condition for a MSLB in the-doghouses located outside containment. ' Consequently, environmental qualification parameters for the doghouses were based on original analysis results of 3300F. Utilizing the new data from Westinghouse . revised Duke Power environmental analysis with super-heated steam conditions indicates a potential increase in doghouse temp-erature from the present 3300F parameter to approximately 440 F doghouse temperature. ' The potential existed that safety related components could be subjectedato temperatures. higher than the qualification basis of 3300F, and could possibly preclude components from performing their' intended safety functions following a postulated MSLB in either doghouse. Detailed engineering evaluations were initiated to determine operability.

Duke Power Company received notification of the MSLB problem outside containment by letter from Westinghouse Electric Corporation on June 6, 1984. Duke Power and Westinghouse are currently performing analyses in order to resolve this issue.

' II. CONTAINMENT ANALYSIS Westinghouse has performed a study to determine the impact on the containment temperature transient of producing superheated steam in the steam generator once the steam generator tube bundle uncovers.

A modified version of the LOTIC-3 code was used for this analysis.

The modifications were made to the wall heat transfer model and the ice condenser drain model. These changes were made to better model the steamline break transient inside containment.

This model used is currently under review by the NRC Containment

' Systems Branch.

A small steamline split rupture transient . typical of the limiting temperature transient for Sequoyah and McGuire, was analyzed. The results'are shown in Figure 1. These results show that the peak temperature is approximately 300F below the FSAR transients. This shows that the FSAR MSLB containment model has more than adequate conservatisms to balance the additional superheat energy released.

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III. PRELIMINARY EVALUATION OF MSLB IN THE DOCHOUSE Westinghouse is performing a steamline break analysis for a spectrum of break sizes and power levels to determine a matrix of protection system actuation times and predicted superheat initiation times. The

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analysis is being performed using safety analysis assumptions of initial conditions and protection system time responses to provide a conservatively early prediction of steam generator tube bundle uncovery and, therefore, the earliest superheat initiation time. This will be used to verify that required safety system actuations (in the doghouse) occur prior to the initiation of superheat generation and determine if further doghouse analyses are required to assure functional operation of safety related equipment in the doghouse.

Anticipated results of the Westinghouse analysis are as described in the following paragraphs:

Large Break Typical response to relatively large breaks, such as the .86 ft 2 break analyzed by Westinghouse for the Catawba Containment analysis, will be as follows:

Reactor trip will occur on overpower AT.

. Main feedwater isolation valve (MFIV) closure will occur due to reactor trip and low Tavg.

. Safety injection will be initiated upon receipt of a low pressurizer pressure signal.

. The motor driven auxiliary feedwater pumps (MDAFP's) will start upon receipt of the safety injection signal.

. If a MDAFP fails to start, the turbine driven auxiliary feedwater pump (TDAFP) will start upon occurrence of a low-low level in an intact steam generator very shortly into the accident.

. Low steam pressure in the faulted steam generator causes main

, steam isolation valve (MSIV) closure.

All the above functions will occur before the temperature in the Doghouse exceeds qualification temperatures. Isolation of auxiliary feedwater to the faulted steam generator may be accomplished by the operator by closing the motor operated valves located in the doghouse, or by closing the control valves or manual isolation valves located in the auxiliary feedwater pump room.

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' Inter $nediate Break' i

Response?to'an intermediate break ~of, for. example,:0.5 ft2 . ,117 s be,similar to.the response to the large break previously described.

~T here'is-the possibility that MSIV; closure will not occur until.

shortly after tube-uncovery. However, the present evaluation is that the temperature-effects on the equipment from the expected profile for the time period during which the. equipment is required to function are no more severe.than those of the original equipment qualification.

Therefore, no consequential impairment of the MSIV will occur.

i . Faulted steam generator. isolation will be accomplished as described for the large break, i

Small Break Small Break response will be as follows:

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. . Reactor trip will occur on low-low steam generator level. Note

!- that all steam generators will approach low-low level at approx-

! instely the same time.

. .MDAFP.'s will start upon receipt of first low-low level signal.

TDAFPwillstartuponreceiptof;the"secondlow-lowjlevelsignal.

.. MFIV closure will occur due to reactor trip and low Tavg.

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. Safety injection and main steam isolation will occur due to low steam pressure. This may happen after tube uncovery. However, the present evaluation is that, due to the short duration of the transient, the temperature effects on the equipment from the expected profile are no more severe than those used for'the original equipment l qualification.

Faulted steam generator isolation is accomplished as described for the large break.

l Duke has analyzed the potential consequences if the temperature in l the doghouses' exceeded the qualification temperature after the equip-l ment has performed its intended safety function described above. The l only potential failures that could compromise any safety functions (repositioning of valves) are valve operator heater circuits. These have been disconnected.

l The only post-c.ccident monitoring instrument located in the doghouse are the auxiliary feedwater flow transmitters. Although these may fail under this environment, steam generator. level transmitters o which will not be affected by a steamline break in the doghouse will j be used to monitor auxiliary feedwater.

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- IV. iLOW PROR&BILITY OF A PIPE BREAK EVENT-

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The phenomona of high energy pipe break'is known to be a low' probability event.- Branch Technical Position MEB 3-1 states'that'."It is

. recognized that pipe rupture is a rare event which may only occur under-unanticipated conditions." Studies cited in attachments to the recent Generic Letter.84-04 from the NRC also' support the preceding statement.

'V. PRELIMINARY PRACTURE MECHANICS EVALUATION Currently,'the' environmental analysis for the MSLB in the McGuire doghouses:is' performed by postulating non-mechanistic breaks in which

, catastrophic pipe failure is assumed. More realistic estimates of l l.  : crack opening area and the resulting thermal and mechanical loads f can be obtained through application of fracture mechanics techniques.

.A scoping study has been carried out by Westinghouse for in-containment MSLB's and preliminary results obtained indicate that a non-mechanistic pipe break will not occur in the main steam line.

The purpose of this scoping study was to show that a circumferential

. flaw larger than any that would be present in the McGuire main steam lines will remain stable when subjected to the worst combination of plant loadings. The flaw stability criteria for the analysis examined

both the global and local stability. The global analysis was carried out using the plastic instability method, based on traditional plastic l limit load concepts but accounting for strain hardening and taking

! into account the presence of a flaw. The local stability analysis l was carried out for a postulated 10 inch long through-wall circumferential flaw. The objective of the local analysis was to show that unstable crack extension will not result for the postulated flaw. The crack opening area resulting from faulted load was calculated for the 10 inch tiaw using simplified analysis techniques.

The foiliwing results were obtained from the above evaluations

a. Limit moment calculations indicated that the critical flaw size

' (beyond which the flaw is unstable) would be greater than the pipe diameter.

b. A postulated 10 inch long through-wall circumferential flaw will remain stable when subjected to maximum faulted load of

! less than 20 kai.

c. The crack opening area is estimated to be about 0.2 in 2. If a

safety factor of 10 is used, the area would be about 2 square i inches.

d. Available fatigue crack growth results for the main. steam line of typical PWR plants' indicate no significant crack growth due to the design transients.

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'From these results-it is judged that it could be demonstrated by fracture mechanics analysis that catastrophic pipe breaks in the

-McGuire 1 and 2 main steam line would not occur.

VI.

SUMMARY

.This evaluation has shown that!for a MSLB inside containment, the present. FSAR containment model has more than adequate conservatisms to balance the additional superheat energy released.

Further, this evaluation has shown that for a' MSLB outsid'e containment all essential safety functions are expected to be completed before adverse temperature effects due to increased doghouse temperatures would occur. Spurious actuation of components following initial positioning has also been evaluated. Thus Duke Power has concluded that plant safety would not be adversely affected in the event of a design basis main steam line break in the doghouse or inside containment.

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