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During a severe accident, the lower head of the reactor pressure vessel (RPV) can be subjected to significant thermal and pressure loads. It is possible that the lower head will fail, releasing large amounts of molten corium into containment. The Three Mile Island accident involved the melting of about 20 tonnes of corium, which collapsed into the lower head of the RPV. Despite the presence of water, the lower head reached temperatures of ~1300 K for 30 minutes in an area with an equivalent diameter of 1m. During this period the reactor cooling system was at 10MPa. The objective of this project was to investigate the timing and size of lower head failure under conditions of low reactor coolant system pressure and large differential temperatures across the lower head wall. This objective was achieved through a series of experiments at the Sandia National Laboratory, USA completed in June 2002.
Although the Three Mile Island vessel did not fail, code analyses conducted in the course of an OECD/NEA TMI-II Vessel Investigation Project (VIP) predicted creep rupture in the prevailing conditions. This implies that the then state-of-the art modelling of the lower head failure was not mature because it did not take full account of the effect of the thermal loading. These methodologies have been further developed since the TMI-VIP project to analyse existing and next generation reactors from the perspective of accident assessment, management, and mitigation. In order to improve and validate structural analysis codes, there is a need for experimental data on lower head deformation and failure phenomena.
The Sandia National Laboratory has completed eight USNRC-sponsored tests on lower head failure (LHF). These tests were specifically designed to address lower head faillure issues with prototypic material and geometry. The present OECD project extends the USNRC SNL/LHF programme to address issues such as lower RCS pressures (representative of depressurised or partially depressurised conditions) and pressure transients. These tests also represent an improvement over previous tests by simulating a large temperature gradient across RPV of lower head wall. The temperature gradients addressed in these tests are representative of conditions without ex-vessel cooling.
Participating countries: Belgium, Czech Republic, Finland, France, Germany, Spain, Sweden, United States.
Programme period: September 1998-June 2002.
The distribution of this package is restricted and subject to prior approval.
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Keywords: corium, lower head failure, reactor pressure vessel.