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Severe Accident Related Research and Development at Forschungszentrum Karlsruhe for Present and Future Needs

Werner Scholtyssek, Gerhard Heusener, Fritz Hofmann, Helmut Plitz

Nuclear Technology

Volume 139 / Number 1 / July 2002 / Pages 10-20


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The research and development program at the Forschungszentrum Karlsruhe, performed within the Program Nuclear Safety Research, is centered around phenomena and processes that could possibly endanger the containment integrity of a large pressurized water reactor after a severe accident. The program includes three activities.

The first activity is in-vessel steam explosion. Premixing phenomena are studied in the QUEOS and PREMIX test series. The efficiency of energy conversion is the subject of ECO tests. The BERDA experimental program investigates the load capacity of a reactor pressure vessel (RPV) in steam explosion events.

The second activity is hydrogen behavior and mitigation. Advanced models and numerical tools are developed to describe hydrogen sources, distribution of gases in containment, the various modes of hydrogen combustion, and corresponding structural loads.

The third activity is ex-vessel melt behavior. The release behavior of melt after RPV failure is studied in DISCO and KAJET tests. In support of core catcher development, interaction with sacrificial and refractory materials, further melt spreading and cooling phenomena are investigated in KAPOOL, KATS, and COMET tests.

The goal is to describe and quantify the governing mechanisms and to develop verified models and numerical tools that are able to predict maximum possible loads for severe accident scenarios on full plant scale. The work supported the development and assessment of the safety design of the French-German European Pressurized Water Reactor (EPR). It led to a broader understanding of severe accident phenomena and of controlling and mitigating measures that can also be of benefit for existing plants.

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