Research Papers: Nuclear Power

Pressure Load Estimation During Ex-Vessel Steam Explosion

[+] Author and Article Information
Matjaž Leskovar

Reactor Engineering Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana, Sloveniamatjaz.leskovar@ijs.si

J. Eng. Gas Turbines Power 131(3), 032901 (Feb 13, 2009) (7 pages) doi:10.1115/1.3078789 History: Received August 19, 2008; Revised August 21, 2008; Published February 13, 2009

An ex-vessel steam explosion may occur when, during a severe reactor accident, the reactor pressure vessel fails and the molten core pours into the water in the reactor cavity. A steam explosion is a fuel-coolant interaction process where the heat transfer from the melt to water is so intense and rapid that the timescale for heat transfer is shorter than the timescale for pressure relief. This can lead to the formation of shock waves and production of missiles that may endanger surrounding structures. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. In the article, different scenarios of ex-vessel steam explosions in a typical pressurized water reactor cavity are analyzed with the code MC3D , which is being developed for the simulation of fuel-coolant interactions. A comprehensive parametric study was performed by varying the location of the melt release (central and side melt pours), the cavity water subcooling, the primary system overpressure at vessel failure, and the triggering time for explosion calculations. The main purpose of the study was to determine the most challenging ex-vessel steam explosion cases in a typical pressurized water reactor and to estimate the expected pressure loadings on the cavity walls. Special attention was given to melt droplet freezing, which may significantly influence the outcome of the fuel-coolant interaction process. The performed analysis shows that for some ex-vessel steam explosion scenarios much higher pressure loads are predicted than obtained in the OECD program SERENA Phase 1.

Copyright © 2009 by American Society of Mechanical Engineers
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Figure 1

Geometry and mesh of 2D axial symmetric model of reactor cavity for central melt pour. The scales in horizontal and vertical directions are different.

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Figure 2

Geometry and mesh of 2D slice model of reactor cavity for left (top) and right (bottom) side melt pours. The scales in horizontal and vertical directions are different.

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Figure 3

Maximum calculated pressure in cavity (left) and maximum calculated pressure impulse at cavity walls (right) for simulated cases

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Figure 4

Maximum calculated pressure in cavity (left) and maximum calculated pressure impulse at cavity walls (right) for most explosive case C2-60, considering different corium droplet bulk temperatures, below which the fine fragmentation process is suppressed

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Figure 5

Mass of hot corium droplets with the bulk temperature above the selected minimum fine fragmentation temperatures (2700 K - top, 2750 K - middle, 2800 K - bottom) in regions with the vapor fraction less than the specified values (<20–<100%) during premixing (left side large scale, right side small scale). In addition also the total (hot and cold) corium droplet mass is presented (total).




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