ML18143B362

Responds to Request for Estimate of Probability That Break Outside Containment During V Sequence at Facility Will Be Submerged.Best Estimate Probability Lies Between 50% & 90%, Depending on Operator Reaction.Lpsi Sys Diagrams Encl

ML18143B362
Person / Time
Site:	Surry
Issue date:	07/17/1985
From:	Benjamin A SANDIA NATIONAL LABORATORIES
To:	Jeffrey Mitchell NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML18143B363	List: ML18143B362 ML18143B364
References
RTR-NUREG-0956, RTR-NUREG-956 NUDOCS 8508120811
Download: ML18143B362 (10)

Text

f"'* *(. July 17, 1985 Ms. Jocelyn Mitchell u. S. Nuclear Regulatory Commission Accident Source Term Project Office 7915 Eastern Avenue, MS 1130SS Willste Building Silver,spring, Maryland 20910

Dear Jocelyn:

Sandia t9tional Laboratories A I bu q u er q u e , N-e w M e xi co 8 7 18 5 Allan So Benjamin This letter responds to your request for an estimate of the probability that a break outside containment during a V sequence at Surry will be submerged.

The question posed is difficult to answer without a systematic thermalhydraulic/structural analysis of the situation pertaining during the postulated accident.

No such analysis exists. Further, I am not a structural engineer and do not claim any expertise in this. area. Therefore, I relied upon the opinions of others more expert than I to formulate a more-or-less subjective*estimate of the required probability.

Toward this end, I specifically talked with the following people: James E. Metcalf, Stone and*Webster Company Walter A. von Riesemann, Sandia National Laboratories Peter R. Davis, Intermountain Technology Company Robert L. Ritzman, Electric Power Research Institute The Surry V s*equence involves the failure of a check valve in one of the 6-inch cold leg emergency core cooling system injection lines following a preexisting condition in the same line where the other check valve was stuck open following a test/maintenance operation (See Figure 1). Water from the reactor coolant system at operating pressure and temperature (2250 psi, 550 F) rapidly flows through a locked-open motor-operated valve at the outside containment boundary, whereupon it enters a 10-inch diameter ASME class 2 pipe of lower pressure capacity (600 psi. design pressure for temperatures under 200 F). The low pressure piping is postulated to break, causing a loss of coolant accident outside containment.

The four people I talked with all expressed uncertainty as to whether the low pressure piping will in fact rupture. If the pipe does not rupture, the path of flow would be (~18~~~;-;;;~~--

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• . J. A. Mitchell July 19, 1;985 through a 1-inch diameter relief valve to a liquid wc;1ste collection tank. This would comprise a significantly less severe accident than the one postulated to occur, since there are emergency procedures which the operator could implement to provide continued core cooling. A double-ended guillotine.break of the 10-inch line outside containment would cause the contents of the refuelingwater storage tank and primary system inventory to discharge into the safeguards building.

The discharge would rapidly .fill the pump shafts and flood-the floor of the safeguards-*

building up to the level of the pipe blackout (Figure 2), where it would spill over to the auxiliary building~_

The location of the pipe break determines the depth of. submergence

• . Messrs. Metcalf, Davis, and Ritzman, who have examined the piping at the site and have reviewed*

the drawings in detail, claim that about 80% of the piping run in the safeguards building is within 2 feet of the floor

• and would* be covered by 2 to 3 feet of water. The ; . remaining 20% is at least 2 feet higher in elevation and so would be covered by less than 1/2 foot of water.

• Mr. Metcalf of Stone and Webster, the architect engineering company for the Surry plant, indicated in our conversations that the most likely point of rupture would be at the first elbow following the transition to low pressure piping (see Figures 2* and 3). His rationale was that (a) this elbow would be the first to see high stress levels and (b) the elbow is restrained by an integral attachment, which would increase the stress levels.-The elbow in question is within the area of piping that would be submerged by 2*to 3 feet. Based on these observations, Mr. Metcalf stated that* he would attribute a 90% probability to the break being submerged.

• Mr. von Riesemann, supervisor of the Containment Integrity Division at Sandia, corroborated Mr *. Metcalf's observations qualitatively but felt that the 90% probability of sub-mergence might be on the high side. Without any structural analyses, Mr. von Riesemann could offer no quantitative estimate of the probability of submergence.

• In a follow-up letter (Attachment A), Mr. Metcalf added that stone and Webster had earlier*performed an analysis of normal operating stresses for the system in question (i.e., the stresses occurring following a design-basis loss of coolant accident in containment).

The situation analyzed is quite different from the V sequence, in that the . direction, velocity, and thermodynamic conditions of ** the flow are different.

However, the results can provide limited insights into the likely locations of high stress.

J. A. Mitchell July 19, 1985 In that analysis, the elbow _in question was the location of second highest stress. Maximum stress occurred at an elbow which was similarly restrained by an integral attachment and which would similarly be submerged by 2 to 3 feet. This led Mr. Metcalf-to conjecture that the.probability that the break would be submerged was "at least".90%.

Most of Stone and Webster's views on the V sequence are summariz~i)in a paper presented to the American Nuclear Society, ( for which Mr. Metcalf is a co-author (see. Attachment B). I reviewed this paper as well. Mr. Davis of Intermountain Technology performed his own assessment of the Surry V sequence, based on a plant visit, review of drawings, and discussions with personnel at Virginia Electric Power Company and.Stone and Webster. In my conversation with Mr. Davis, he indicated general ment with the idea that the break occurring during a V sequence would likely*be submerged.

Asked for-.his own subjective estimate of the probability that the break would be submerged, he tentatively took a figure of 80%. Mr. Ritzman of E.P.R.I. analyzed the Surry V sequence while at Science Applications International Corporation under a c:=ontract with E.P.R.I. The re 9~lts of this analysis are now part of an E.P.R.I. report( > *for which he is the principal author. In my discussion with him, he indicated that he and the other authors of the report could not justify any conclusions about where the break_ would be located. Based on piping lengths alone, however, he felt that. there would be at least an 80% probability that the break, if one occurred, would be in a location where it would be submerged.

A key question among the authors and the reviewers of the E.P.R.I. report, he indicated, was what he termed the morphology of the break. If a 10-inch diameter ended guillotine break were to occur at a submergence depth of 2 to 3 feet, there would be a real question as to whether bubble breakup could occur rapidly enough, at the* gas velocities involved, to assure any significant amount of scrubbing of the fission products.

Data from General Electric Company, he noted, indicates that a rise height of 1 2 A. Drozd, F. A. Elia, Jr., and J.E. Metcalf, "The V sequence:

An Engineering Viewpoint," presented at ANS Topical Meeting on Fission Product Behavior and Source Term Research, Snowbird, Utah, July 15-19, 1984. R. L. Ritzman, et al., "Surry Source Term and consequence Analysis," EPRI NP-4096, June 1985.

I~ .e e J. A. Mi_tchell July 19, 1985 one to two pipe diameters is required for a gas discharge to break up into a swarm of bubbles. Based on this, he and the other authors had concluded that it was "unlikely" that. one would get classic pool scrubbing.

He felt that the confined nature of the safeguards building made it more likely that one would get scrubbing as a result of two-phase mixing and turbulence,*

a situation for which there are no applicable fission product scrubbing models. Based on his own judgment, he felt that it was "fairly probable" that a large fraction of the fission products would end up in the.water as a result of this confined turbulence.

On further questioning, he indicated that "fairly probable" meant about equally likely as not and "large fraction of fission products" meant a decontamination factor of 10. Considering deposition on the walls as well as in the water, he felt that it was "highly likely" that the* overal.l fraction of fission products remaining in the safeguards building would be large. Let me return now to the original question.

Based on piping lengths and the assessment of likely rupture locations, I would estimate the probability of the break being at least 2 feet below the overflow level (i.e., the pipe blockout) to be at least 80% and probably more like 90%. However, a possibility not examined in previous

• analyses is whether the operator might take action to isolate the refueling water-*storage tank within the 10 minutes or so that it takes-to flood the safeguards building. (The valves between the refueling water storage tank and the .reactor coolant system injection points are normally open.) Further, there may not be an adequate water head above the rupture to assure breakup of the gas bubble if the rupture is the full size of the pipe diameter.

I th~refore redefined the question to ask whether the rupture location is sufficiently submerged to assure significant fission product scrubbing.

For this question, I felt that the best-estimate probability lies between 50% and 90%; therefore, I subjectively took a more-or-less average value of 75%. ASB:6411:cg Enclosures S i_ncerely, I L. {" L .c.,;.. *--~ ~-_1 *, r. r. -. , . *i A. s. Benjamin, Supervisor Reactor Safety Technology Division 6411 e J. A. Mitchell July 19, 1985 Copy to: NRC J. c. Glynn NRC D. Pyatt NRC M. Ernst BCL R. Denning BNL T. Pratt 6410 J. w. Hickman 6412 A. L. Camp 6415 F. E. Haskin 6449 D. c. Williams-I -~--

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I I I I I I ' ' I ACCESS DOOR POSTULATED BREAK e _________________

FIGURE 2 e PIPE BLOCKOUT7 I / I-MAT I I I ! I LPSl PUMPS .,* ------'---------------

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( ATTACHMENT A . . e . * . STONE 8 WEBSTER ENGINEERING CORPORATION 245 SUMMER STREET, BOSTON, MASSACHUSETTS ADDRESS ALL CORRESPONDENCE TO P.O. BOX 232!5, BOSTON, MASS. 02107 W. U. TELEX: 94-0001 BOSTON NEW YORK CHERRY HILL, N . .J. DENVER CHICAGO HOUSTON* PORTLAND, OREGON WASHINGTON.

D.C

• 94-0977 DESIGN CONSTRUCTION REPORTS EXAMINATIONS CONSULTING ENGINEERING . Dr. Allan Benjamin May 22, 1985 Sandia National Laboratories 1515 Eubank

• SE Albuquerque, NM 87125 BREAK LOCATION FOR.SURRY V SEQUENCE As a followup to your visit to our Boston office on May 21, 1985 to discuss details of our analysis of the V sequence for the Surry plant, this letter transmits the results of our investigation of the normal operating stresses in the portion of the low pressure safety injection system piping exposed to RCS .pressure.

Although the accident induced stresses would be different from the normal operating stresses, the latter are an indication of relative stress intensities at various points of interest throughout the system. The findings of our.review are as follows: (1) Maximum normal operating stress between MOV1890C and RV1845B (that portion of piping exposed to high temperature flow through the relief valve) is confirmed to be at the currently postulated break location (900 elbow with hanger /ll27C2.ll).

(2) Maximum normal operating system stress occurs at the 900 elbow where the piping turns upward towards the pump discharge at a 450 angle to the floor. An .integral attachment is also placed on this elbow (hanger #127C2.13).

A break at this location would be submerged to the same depth as the currently postulated break location.

(3) Normal operating stresses in the 8" line running above the flooded water level were found to be a minimum factor of three below the maximum normal operating system stress *. (4) Normal operating stresses*

in the portion of the 10" line running slightly below the flooded water level were found to be a minimmn factor of two below the maximmn normal operating system stress.

(. We conclude that the currently postulated break location remains the most likely. The alternative break location is also a 90o elbow with integral attachment, located at the same elevation with respect to the flooded water level as the currently postulated break location.

Breaks slightly below, at, or above the flooded water level are extremely unlikely, estimated to be at least a factor of ten less likely than either the currently postulated break location or the alternate location discussed above. Very truly yours, .-t.Mt> ~\..i__Jes E. Metcalf Lead Engineer Source Term Project STONE & WEBSTER A