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Guidelines for radiation shielding evaluation by Monte Carlo method was designed and developed for practical radiation shielding for radioactive material transport casks. The guidelines consist of current technical advancement on estimating the dose equivalent rates around the surface of a transport cask. The guidelines could be useful to validate the procedure and the results of radiation shielding analyses for transport casks.

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For the purpose of performing the reasonable shielding calculation of transport/storage spent fuel cask, the discussion of the cause of the discrepancy between the measured and calculated dose rate of a spent fuel cask is important. This paper shows the several items that may have large discrepancy between the measured and calculated dose rate of a spent fuel cask with respect to the calculation models and methods. The items include I) the expression of source terms such as burnup profile, distribution of activated gamma-ray source, and gamma-ray source spectra, 2) the usage of a detailed geometrical model that is able to calculate the distribution of fission reaction precisely, and 3) the difference between detector response and dose conversion factor. The consideration of these effects may improve the agreement between the measurement and the calculation of a spent fuel cask.