モンテカルロ法によるコンクリート中間貯蔵キャスクの遮蔽解析(NP4 核燃料サイクル・バックエンド技術)

Abstract

This paper describes a neutron and gamma-ray shielding analysis of a concrete interim storage cask with the Monte Carlo code MCNP 5, and evaluated the dose-equivalent date distributions around the cask. The following two remarks are obtained by the Monte Carlo analysis. One is that the radiation streaming from the exhaust ducts is not so noticeable as expected for the ducts are tow-legged ones and the outlets are located away from the upper part of spent fuels and also from the lower part of them, respectively, ant the other is that the secondary gamma-ray dose-equivalent rates on the cask surface are rather large than the neutrons, because a large number of secondary gamma rays are produced by H(n, γ) reactions of thermal neutrons in the thick concrete shield.