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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.
Zhang Huanqiao, Liu Zuhua, Ding Shengyue, and Liu Shaoming
Nuclear Science and Engineering | Volume 86 | Number 3 | March 1984 | Pages 315-319
Technical Note | doi.org/10.13182/NSE84-A17560
Articles are hosted by Taylor and Francis Online.
This research was published (in Chinese) in Chin. J. Nucl. Phys., 3, 2, 149 (1981). The average number of prompt neutron and the distributions of prompt neutron number probability P(ν) for spontaneous fission of 240Pu, 242Cm, and 244Cm relative to (252Cf) have been measured using a large gadolinium-loaded liquid scintillation counter with a co-incidence method. The results were (240Pu) = 2.141 ± 0.016, (242Cm) = 2.562 ± 0.020, and (244Cm) = 2.721 ±0.021. The measured distributions of prompt neutron number were fitted with Gaussian curves by a weighted least-squares method. The widths of Gaussian distribution are 1.149 ± 0.047, 1.159 ± 0.074, and 1.175 ± 0.098 for 240Pu, 242Cm, and 244Cm, respectively. These results as well as a previous measurement of spontaneous fission of 252Cf show the linear variation of σ with at the first order of approximation. The data were fitted by a least-squares method, and the result is given by σ = 0.980 + 0.076. This fact demonstrates the trend that the width of the excitation energy distribution of fission fragments increases with the average excitation energy of the fission fragments in the range of nuclides mentioned above.