Year: 2011
Type: article
Abstract: The spent fuel transport cask has structure based mostly on the multi-layer for the radiation shielding.However, the structure for neutron shielding is very weak around the trunnion.It is important to evaluate realistic dose-equivalent rate in shielding design of the spent fuel transport cask, therefore the three dimensional continuous-energy Monte Carlo radiation transport code that exactly treating the complicated geometry was applied.In the fixed source problem such as a neutron deep penetration calculation with the Monte Carlo method, the application of the variance reduction method is the most important for a high figure of merit and the most reliable calculation.The concerned items are setting method for the variance reduction.The validation of dose-equivalent rate evaluation for the spent fuel transport cask by Monte Carlo code were performed by two kinds of neutron shielding benchmark experiments, deep penetration experiment for the side surface of the cask and streaming experiment around the trunnion.Dose-equivalent rate distributions at each benchmark were measured and compared with the calculated results.The comparison showed a good consistency between calculation and experiment results.
DOI: 10.15669/pnst.2.855
Authors: Mitsufumi ASAMI, Kenichi SAWADA, Akiko KONNAI, Naoteru ODANO
Institution: National Maritime Research Institute
Language: English
Subfield: Safety, Risk, Reliability and Quality
Field: Engineering
Domain: Physical Sciences
Open Access status: bronze
OpenAlex ID: W656607094