| Issue |
EPJ Web Conf.
Volume 153, 2017
ICRS-13 & RPSD-2016, 13th International Conference on Radiation Shielding & 19th Topical Meeting of the Radiation Protection and Shielding Division of the American Nuclear Society - 2016
|
|
|---|---|---|
| Article Number | 06019 | |
| Number of page(s) | 3 | |
| Section | 6. Calculation Methods Monte Carlo & Deterministic | |
| DOI | https://doi.org/10.1051/epjconf/201715306019 | |
| Published online | 25 September 2017 | |
https://doi.org/10.1051/epjconf/201715306019
CAD-Based Monte Carlo Neutron Transport KSTAR Analysis for KSTAR
Nuclear Engineering Department, Seoul National University, 08826 Seoul, Korea
* Corresponding author: shimhj@snu.ac.kr
Published online: 25 September 2017
The Monte Carlo (MC) neutron transport analysis for a complex nuclear system such as fusion facility may require accurate modeling of its complicated geometry. In order to take advantage of modeling capability of the computer aided design (CAD) system for the MC neutronics analysis, the Seoul National University MC code, McCARD, has been augmented with a CAD-based geometry processing module by imbedding the OpenCASCADE CAD kernel. In the developed module, the CAD geometry data are internally converted to the constructive solid geometry model with help of the CAD kernel. An efficient cell-searching algorithm is devised for the void space treatment. The performance of the CAD-based McCARD calculations are tested for the Korea Superconducting Tokamak Advanced Research device by comparing with results of the conventional MC calculations using a text-based geometry input.
© The Authors, published by EDP Sciences, 2017
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