Abstract
For the purpose of performing reasonable shielding calculation of a spent fuel transport cask, the use of Monte Carlo methods has been proposed for solving the radiation transport problem on a detailed structure of the transport cask considering fixed neutron sources. A SMIRE (Simplified MCNP-ANISN_W Variance Reduction) system has been developed in the present study, which is possible to generate automatically the lower weight boundary of the weight window for each mesh based on the Consistent Adjoint Driven Importance Sampling (CADIS). Compared with the case of the importance based on the empirical formula, the figure of merit is increased by a factor of 25. In this system, it is possible to calculate the weight suitable for the distantly-positioned detector point from the fuel effective region to introduce the relaxation factor which relaxes the increase of the particle numbers at the boundary of weight window meshes that are generally caused by the large attenuation of adjoint flux. This system is used for a variety of the radiation transport problems as well as the transport cask.
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