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Each test consisted of PWR-type fuel elements, one meter in length, using inconel grid spacers. Tests SFD1-ST and SFD1-1 utilized 32 unirradiated fuel elements, and tests SFD1-3 and SFD1-4 utilized 26 fuel elements irradiated to nominally 36,000 MWD/t and two instrumented fresh fuel elements. The prjssure was 7 MPa. The fuel was fission heated at decay heat level by the PBF reactor. 1 K/s) characteristic of small-break LOCA conditions. Steam flow was provided by water boiloff in the channels. The first test in the series had a simulated reflood transient.

All comments are expected to be received by the middle of October, 1995. The necessity of an additional round of comments will be evaluated in the near future. As mentioned previously, this specific work was not dependent on details of core-melt progression. Work described above pertaining to core melt progression is used for vessel ablation evaluation and is a part of the in-vessel sterjn explosion evaluation headed by Professor Theofanous of the University of California at Santa Barbara. FUTURE NEEDS In-vessel retention (TVR) via external cavity flooding is a severe accident mitigation approach adopted by Westinghouse for use in their AP600 design.

2, _v————————i-______-exp(—————) } FV dt 60 T F-l K L (10) In the SCDAP/RELAP5, there are two models for BWR geometry. 7, according to the manual. 35 are coded. It is interesting to note that this model calculates a decreasing reaction rate (M B4C /M|4C ) with increasing temperature T. In the newer BWR blade/box component model, developed at ORNL, the B4C oxidation is derived from an advanced B4C/H2/H2O chemistry package, based on the SOLGASMIX code [19]. SOLGASMIX calculates equilibrium compositions in systems containing one gaseous phase, condensed mixtures, and condensed phases of invariant or variable stoichiometry.

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Behavior of LWR Core Materials Under Accident Conditions (IAEA TECDOC-921)


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