Description
We present some examples of preparing CAD-geometries for use by particle transport codes. To be precise we have prepared geometries for OpenMC[1], but the underlying libraries are able to use the same geometry in both MCNP[2] and fluka[3].
Our vehicle for preparing geometries is the newly developed tool CAD_to_OpenMC[4], which is able to ingest a geometry defined in a step-file (or a set set thereof), create a surface-discretized version of it, and assemble it in a binary HDF5-based format which is readable by particle transport codes - if compiled with support for the DAGMC[5] extension.
Below is one of the biggest and most advanced geometry models we have worked on, namely the Molten Salt Reactor Experiment[6].
References.
[1] P. K. Romano and B. Forget,
“The OpenMC Monte Carlo particle transport code,
” Ann. Nucl. Energy, vol.
51, pp. 274–281, Jan. 2013, doi: 10.1016/j.anucene.2012.06.040.
[2] J. A. Kulesza et al.,
“MCNP Code Version 6.3.0 Theory & User Manual,
” Los Alamos National Laboratory,
Los Alamos, NM, USA, LA-UR-22-30006, Rev. 1, Sep. 2022. doi: 10.2172/1889957.
[3] T. T. Böhlen et al.,
“The FLUKA Code: Developments and Challenges for High Energy and Medical
Applications,
” Nucl. Data Sheets, vol. 120, pp. 211–214, Jun. 2014, doi: 10.1016/j.nds.2014.07.049.
[4] E. B. Knudsen and L. Chierici,
“CAD_to_OpenMC: from CAD design to particle transport,
” J. Open Source
Softw., p. to appear.
[5] P. P. H. Wilson, T. J. Tautges, J. A. Kraftcheck, B. M. Smith, and D. L. Henderson,
“Acceleration techniques
for the direct use of CAD-based geometry in fusion neutronics analysis,
” Fusion Eng. Des., vol. 85, no.
10–12, pp. 1759–1765, Dec. 2010, doi: 10.1016/j.fusengdes.2010.05.030.
[6] D. Shen, G. Ilas, J. J. Powers, and M. Fratoni,
“Reactor Physics Benchmark of the First Criticality in the
Molten Salt Reactor Experiment,
” Nucl. Sci. Eng., vol. 195, no. 8, pp. 825–837, Aug. 2021, doi:
10.1080/00295639.2021.1880850.
