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Forwards Summary of Planned Initiation of BWR Mark III Plant Studies by Severe Accident Sequence Analysis Program
	ML20204C184
Person / Time
Issue date:	06/22/1986
From:	Curtis R; NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Duraiswamy S; Advisory Committee on Reactor Safeguards
Shared Package
ML20204C177	List: ML20204C180; ML20204C184;
References
	ACRS-2427, NUDOCS 8607310044
	Download: ML20204C184 (6)
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MEMORANDUM FOR: S. Duraiswany, ACRS FROM: Robert T. Curtis Containment Systems Research Branch

SUBJECT:

MAY 7, 1986 MEETING OF THE ACRS SUBCOMMITTEE ON THE SAFETY RESEARCH PROGRAM Please forward the enclosed sumary of the planned initiation of BWR Mark III plant studies by the SASA Program. This work addresses the integrity of the reactor pressure vessel support skirt as well as the integrity of the support pedestal that was questioned in the May 7,1986 subcommittee meeting on the Safety Research Programs.

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Robert T. Curtis, Chief Containment Systems Research Branch cc: M. Silberberg G. Marino 8607310044 e60622 PDR ACRS PDR 2427 ATTAcnnsat. G ,

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MARK III CONTAINMENT ANALYSIS The initial SASA Mark III Containment analysis will investigate a couple of basic considerations of containment integrity before proceeding with the analysis of risk dominant sequences. The first of these is the integrity of the reactor pressure vessel (RPV) support pedestal. The second is the interrelation of the RPV support skirt, radiation loss from the melt below the vessel, various heat sinks within containment, and the eventual maximum size of the melt and final location. Previous studies have ignored these interrelations because of model limitations and have lead to unconvincing resul ts.

ORNL-SASA has been cooperating with SNL in the development of CONTAIN models for BWRs. It is now expected that this work will be completed by September

( 1986. Analysis models of core melt and subsequent failure of the RPV will also be sufficient to define: debris deposition onto the concrete within the pedestal; the extent of concrete penetration (vertical and horizontal);

pedestal and reactor ' skirt and vessel heat-up by radiation; and convection and energy distribution within containment.

With gases and volatile fission products removed from the melt, only about 0.5%

of equivalent full power remains in the melt 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months after shutdown and about 0.2% at 10 days. The CONTAIN analysis will include a VANESA calculation for the best evaluation of the fission product removal as the melt / concrete interaction proceeds. The geometry of the melt and upward heat loss will define the energy that must be dissipated to the basemat and other boundaries.

The continued drop in decay heat assures a reasonably accurate value for total penetration.

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I A schematic of the pedestal area is shown in Figure 1. Appropriate parts of the Grand Gulf FSAR description note the following. The pedestal base mat is 7'9" thick and is 68' in diameter. The surface area (the foundation base mat) directly below the RPV is 6 ft below the top of the pedestal base mat. If all of the core and core support structure melted and dropped into the cavity, the debris bed would be 3 ft deep. MELPRI calculations for BWRs show a relatively rapid melt to about 40% of the core, subsequent melt through, and continued melting for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months until about 96% of the core has melted. Early MELPROG analyses for PWRs indicate also that pressure vessel melt through will occur with discharge of about 40% of the core initially. For the BWR case the melt would only fill the 6 ft cavity to about 14 inches. The melt would then penetrate the remaining l'9" of pedestal base mat and erode on into the containment foundation base mat. Thus the early concrete ablation would take a shape similar to the dotted line in Figure 1.

I The vertical cylinder of the pedestal is rigidly connected to the reinforced concrete pedestal base mat. The reactor vessel is attached to the concrete pedestal by 120 3.5 in diameter bolts. A more detailed overall view with the re-bar indicated is given in Figure 2. The apparent extensive shear support area from the 68' diameter pedestal base mat to the pedestal would be unlikely to fail as the melt penetrates the containment foundation. The pressure vessel support skirt is possibly more vulnerable, thus the interactions of the melt by radiation and convection with the vessel and skirt as noted previously should be very important.

The application of the CONTAIN-BWR code to this problem should provide numerical evaluation for the various estimates of upward vs downward heat transmission and interactions with other containment features and is scheduled for completion by February 1987.

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ML20204C184

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