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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.
Yukio Oyama, Hiroshi Maekawa
Nuclear Science and Engineering | Volume 97 | Number 3 | November 1987 | Pages 220-234
Technical Paper | doi.org/10.13182/NSE87-A23504
Articles are hosted by Taylor and Francis Online.
Angular neutron fluxes leaking from the surface of beryllium slab assemblies have been measured with irradiation of deuterium-tritium neutrons. The experiment was performed using the time-of-flight technique with an NE-213 scintillation detector. The measured neutron energy range was from 50 keV to 15 MeV. The thicknesses of the slabs were 50.8 and 152.4 mm, and the measured angles of the angular fluxes were 0.0, 12.2, 24.9, 41.8, and 66.8 deg. The experimental results have been compared with the results calculated by the Monte Carlo codes, MORSE-DD and MCNP, using the data of beryllium in the JENDL-3PR1, ENDF/B-IV, and Los Alamos National Laboratory nuclear data files. The results calculated with these files showed discrepancies of 20 to 30% from the experimental results. It was pointed out that the angular distributions of an elastic cross section and the total cross section of an inelastic reaction for 14.8-MeV neutrons in the files were insufficient to reproduce the measured spectra.