Neutronic Analysis Code for Fuel Assembly Using a Vectorized Monte Carlo Method

Yuuichi Morimoto; Hiromi Maruyama; Kazuya Ishii; Motoo Aoyama

doi:dx.doi.org/10.13182/NSE89-A23688

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Home / Publications / Journals / Nuclear Science and Engineering / Volume 103 / Number 4

Neutronic Analysis Code for Fuel Assembly Using a Vectorized Monte Carlo Method

Yuuichi Morimoto, Hiromi Maruyama, Kazuya Ishii, Motoo Aoyama

Nuclear Science and Engineering / Volume 103 / Number 4 / December 1989 / Pages 351-358

Technical Paper / dx.doi.org/10.13182/NSE89-A23688

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A 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.