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INL makes first fuel for Molten Chloride Reactor Experiment
Idaho National Laboratory has announced the creation of the first batch of enriched uranium chloride fuel salt for the Molten Chloride Reactor Experiment (MCRE). INL said that its fuel production team delivered the first fuel salt batch at the end of September, and it intends to produce four additional batches by March 2026. MCRE will require a total of 72–75 batches of fuel salt for the reactor to go critical.
Pedro A. Landeyro, Tadeusz Zoltowski
Nuclear Science and Engineering | Volume 104 | Number 2 | February 1990 | Pages 183-187
Technical Note | doi.org/10.13182/NSE90-A23714
Articles are hosted by Taylor and Francis Online.
The aim of the research was to study the monitoring of keff in some steps of fissile material processing. Two sets of experimental tests were performed. In the first one, variable gradients of concentration were modeled, while in the second one (uniform) the tank was filled at different levels with ura-nyl-nitrate solution having 45 g U/ (80 wt% enriched in 235U). Another uniform configuration series was calculated to cover a range of keff not far from criti-cality. In the last computations, all the calculational models were used to represent experiments in which only the source position changes from the axis of the setup to the position 0.5 cm out of the external surface of the tank. The keff calculations were performed with the KENO-IV Monte Carlo code. The flux computations were carried out with the MORSE C. G. code. The dependence between k∞ and Mepi (the external source epithermal multiplication) previously established was used for the analysis of experimental and calculational tests. Excellent agreement between neutron multiplication or epithermal neutron multiplication measured and calculated values was found. The agreement between keff values obtained from the measured and calculated epithermal neutron multiplication data and KENO-IV results improves with the increase of keff values. The relative differences are within the Monte Carlo calculation error range (∼10%) for keff values >0.65 for the axially located source configurations, >0.37 for configurations when neutron source was located externally and fissile material distributed uniformly, and >0.66 when neutron source was located externally and for configurations with fissile material concentration gradients.
