Proceedings of the Korean Nuclear Society Conference (한국원자력학회:학술대회논문집)
- 1995.05a
- /
- Pages.847-852
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- 1995
Neutron Dose Rate Analysis of PWR Spent Fuel Transport Cask Using Monte Carlo Method
- Do, Mahnsuck (Hanyang University) ;
-
Kim, Jong-Kyung
(Hanyang University)
;
- Yoon, Jeong-Hyoun (Korea Atomic Energy Research Institute)
- Published : 1995.05.01
Abstract
A shielding analysis for KSC-7, the shipping cask for transporting the 7 PWR spent fuel assemblies, has been carried out. Radiation source term has been calculated on spent fuel with burnup of 50,000 MWD/MTU and 1.5 years cooling time by ORIGEN2 code. The shielding calculation for the cask has been made by using MCNP4A code with continuous cross section data library from ENDF/B-V. As a result of neutron dose rate analysis, another shielding calculational model on spent fuel shipping cask was provided which is using the Monte Carlo method.
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