Development and optimization of reactor depletion methodology in Monte Carlo iMC code

Abstract

In this research, a study on the depletion analysis in the Monte Carlo neutron transport analysis code iMC was conducted. The iMC code is continuous energy Monte Carlo code developed in the KAIST. Monte Carlo-based depletion approach is considered as a best option for depletion analysis, since the method provides the solution with greater accuracy and precision compared to other depletion methodologies. Conventional Monte Carlo codes therefore developed and validated their own depletion module. Likewise, this study is conducted to implement the Monte Carlo-based depletion capability to the iMC code. In addition, efforts have also been made to optimize the additional computational requirements in the Monte Carlo transport which are originated from the depletion calculations. The depletion module of the iMC code was intended to validate by calculating various benchmark or model problems from the simplest single fuel pin problem to a standard depletion benchmark fuel assembly problem VERA. The validation was attempted by comparing the solutions from the iMC code with the Serpent, which contains the pre-validated depletion module. From this research, the accuracy and the efficiency of the depletion module of the iMC were studied. Also, the study can serve as a foundation for presenting and researching various application plans based on the depletion module.