RESEARCH ON THE RADIATION SHIELDING CALCULATIONS

摘要

This paper is devoted to the research on thernradiation shielding calculations. Previously there arernmany methods for radiation shielding calculations.rnThe representative methods are point-kernel methodsrnand Monte Carlo methods. QAD code is one of thernpoint-kernel codes widely used in the radiationrnshielding calculations. MCNP is a general-purpose,rncontinuous-energy, generalized-geometry,rntime-dependent, coupled neutron / photon/ electronrnMonte Carlo transport code. When QAD is used inrnthe simple geometry shielding calculation, the resultrnis nice. But if the geometry is complex, it will berndifficult for QAD code to describe and the error ofrnthe results will be greater. An experiment is designedrnto research the shielding of the source of Cs-137 inrnthe laboratory, followed with the calculation of thernshielding with the above two codes. Then the resultsrnis compared to analyze the advantages andrndisadvantages of the two codes. It is shown thatrnMCNP code describes the geometry more exactlyrnthan QAD code. The radiation shielding of the STC isrnalso calculated with MCNP. The STC is designed tornsafely transport spent fuel assemblies and it canrntransport 26 directly loaded PWR fuel assemblies.rnThe key parameters of the design basis on directlyrnloaded spent fuel assemblies in a domestic NPP. Atrnlast the results are compared with the survey results.