Monte-Carlo Method

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Statistical calculation procedure, e.g. to calculate the neutron flux distribution in burnup and shielding calculations. In this context, the history of individual, statistically sampled neutrons is calculated until sufficient individual simulations (individual destinies) result in numerical average values for the neutron flux at the points considered. The calculation, which is simple in itself, requires a very large number of individual calculations, since a high degree of accuracy is achieved only with a large number of individual destiny calculations.

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11 - 15 March 2018
Munich, Germany

30 September - 04 October 2018
Prague, Czech Republic