title: Shielding calculation and criticality safety analysis of spent fuel transportation cask in research reactors.
abstract: In this study, shielding calculation and criticality safety analysis were carried out for general material testing reactor (MTR) research reactors interim storage and relevant transportation cask. During these processes, three major terms were considered: source term, shielding, and criticality calculations. The Monte Carlo transport code MCNP5 was used for shielding calculation and criticality safety analysis and ORIGEN2.1 code for source term calculation. According to the results obtained, a cylindrical cask with body, top, and bottom thicknesses of 18, 13, and 13 cm, respectively, was accepted as the dual-purpose cask. Furthermore, it is shown that the total dose rates are below the normal transport criteria that meet the standards specified.
authors.affiliation: Iran Radioactive Waste Management Company, Tehran, Iran.
authors.forename: A
authors.initials: A
authors.lastname: Mohammadi
authors.affiliation: Nuclear Science and Technology Research Institute, Tehran, Iran. Electronic address: m_hassanzadeh2003@yahoo.com.
authors.forename: M
authors.initials: M
authors.lastname: Hassanzadeh
authors.affiliation: Islamic Azad University, Science and Research Branch, Hesarak, Punak, Tehran, Iran.
authors.forename: M
authors.initials: M
authors.lastname: Gharib
keywords: Criticality safety; MCNP5 code; ORIGEN2.1 code; Research Reactors; Shielding; Spent fuel; Transportation cask