Abstract
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.
Highlights
The Monte Carlo (MC) particle transport analysis for a complex system such as an accelerator or a fusion facility may require accurate modeling of the complicated geometry
We examine the performance of McCARD/computer aided design (CAD) for the neutronics analysis of Korea Superconducting Tokamak Advanced Research (KSTAR) device [12]
For efficient distance to surface (DTS) computations, cells bounded by elementary surfaces such as planar, cylindrical, spherical, conical, and toroidal surfaces in the CAD geometrical data file are converted into the internal geometrical data structure in McCARD
Summary
The Monte Carlo (MC) particle transport analysis for a complex system such as an accelerator or a fusion facility may require accurate modeling of the complicated geometry. There have been two kinds of approaches to develop MC code systems utilizing the CAD data: the external format conversion and the CAD kernel imbedded MC simulation. Which were developed to automatically convert the CAD data into the MCNP [7] geometrical input data. This approach makes the most of the existing MC codes without any modifications, but implies latent data inconsistency due to the difference of the geometry modeling system. A MC code directly utilizes the CAD data for the MC particle tracking or an internal conversion to an internal data structure of the constructive solid geometry (CSG) and/or boundary representation (B-rep) modeling with help of an imbedded CAD kernel. We examine the performance of McCARD/CAD for the neutronics analysis of Korea Superconducting Tokamak Advanced Research (KSTAR) device [12]
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