Neutronic Analysis Code for Fuel Assembly Using a Vectorized Monte Carlo Method
1989; Taylor & Francis; Volume: 103; Issue: 4 Linguagem: Inglês
10.13182/nse89-a23688
ISSN1943-748X
AutoresYuuichi Morimoto, Hiromi Maruyama, Kazuya Ishii, Motoo Aoyama,
Tópico(s)Nuclear Physics and Applications
ResumoAbstractAbstractA fuel assembly analysis code, VMONT, in which a multigroup neutron transport calculation is combined with a burnup calculation, has been developed for comprehensive design work use. The neutron transport calculation is performed with a vectorized Monte Carlo method that can realize speeds >10 times faster than those of a scalar Monte Carlo method. The validity of the VMONT code is shown through test calculations against continuous energy Monte Carlo calculations and the PROTEUS tight lattice experiment.
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