Development of CAD-MCNP Interface Program GEOMIT and Its Applicability for ITER Neutronics Design Calculations

Nasif, Hesham; Masuda, Fukuzo; Morota, Hidetsugu; Iida, Hitomasa; Sato, Satoshi; Konno, Chikara

doi:10.13182/nt12-a14521

Cited by 6 publications

(4 citation statements)

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“…(2) Simplification of input data is important in order to eliminate the error in calculation. The computer-aided design (CAD) / MCNP conversion interface code, GEOMIT [13], will be used for this purpose. (3) Visualization of calculation results is important in order to be judged the validity timely and accurately.…”

Section: Future Problem For Mcnp Calculationmentioning

confidence: 99%

Radioactivity evaluation of the secondary sodium in DRACS of the Japan Sodium-cooled Fast Reactor

Sasaki¹,

Naito²,

Ohki

et al. 2014

Progress in Nuclear Science and Technology

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This paper evaluates the secondary sodium activation in Direct Heat Exchanger (DHX), the radioactivity density and the dose rate around the secondary sodium pipes in Direct Reactor Auxiliary Cooling System (DRACS) of the Japan Sodium-cooled Fast Reactor (JSFR), by improving the exactness of a shielding calculation model with the Monte Carlo methodology. From this result, the prospects to meet the shielding design requirements for radioactivity free areas are obtained.

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Section: Future Problem For Mcnp Calculationmentioning

confidence: 99%

Radioactivity evaluation of the secondary sodium in DRACS of the Japan Sodium-cooled Fast Reactor

Sasaki¹,

Naito²,

Ohki

et al. 2014

Progress in Nuclear Science and Technology

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“…The first approach involves converting the internal representation of a CAD model (boundary representation) to the native CSG representation used in MC codes. This approach has been around since at least 2005 (Tsige- Tamirat and Fischer, 2005); specific software packages for performing CAD-to-CSG conversion include McCAD (Harb et al, 2023), MCAM (Wang et al, 2014), GEOMIT (Nasif et al, 2012), CMGC (Wang et al, 2020), SuperMC (Wu et al, 2015), and GEOUNED (García et al, 2021). Of these packages, McCAD 1 and GEOUNED 2 are available under an open-source license.…”

Section: Introductionmentioning

confidence: 99%

CAD and constructive solid geometry modeling of the Molten Salt Reactor Experiment with OpenMC

Yilmaz,

Romano,

Chierici

et al. 2024

Front. Nucl. Eng.

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In this study, we present a detailed comparison of two independently developed models of the Molten Salt Reactor Experiment (MSRE) for Monte Carlo particle transport simulations: the constructive solid geometry (CSG) model that was developed in support of the MSRE benchmark in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and a CAD model that was developed by Copenhagen Atomics. The original Serpent reference CSG model was first converted to OpenMC’s input format so that it could be systematically compared to the CAD model, which was already available as an OpenMC model, using the same Monte Carlo code. Results from simulations using the Serpent and OpenMC CSG models showed that keff agreed within 10 pcm while the flux distribution in space and energy generally agreed within 0.1%. Larger differences were observed between the OpenMC CAD and CSG models; notably, the keff computed for the CAD model was 1.00872, which is more than 1% lower than the value for the CSG model and much closer to experiment. Several areas of the reactor that were modeled differently in the CSG and CAD models were discussed and, in several cases, their impact on keff was quantified. Lastly, we compared the computational performance and memory usage between the CAD and CSG models. Simulation of the CSG model was found to be 1.4–2.3× faster than simulation of the CAD model based on the Embree ray tracer while using 4× less memory, highlighting the need for continued improvements in the CAD-based particle transport ecosystem. Finally, major performance degradation was observed for CAD-based simulations when using the MOAB ray tracer.

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“…For reusing the abundant existed CAD model and convenient modeling with CAD modeling tools, an automatic modeling tool for accurate prompt modeling from CAD model is needed. Some previous researches (Grosse and Tsige-Tamirat, 2009;Nasif et al, 2012) of automatic modeling for Monte Carlo simulation had been studied, but most of them focused on half space model. In this paper a newly automatic modeling method for primitive solids was researched for the need of rapid development and various applications of Monte Carlo simulation.…”

Section: Introductionmentioning

confidence: 99%

CAD-based Monte Carlo automatic modeling method based on primitive solid

Wang

Song

Shen

et al. 2016

Annals of Nuclear Energy

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Development of CAD-MCNP Interface Program GEOMIT and Its Applicability for ITER Neutronics Design Calculations

Cited by 6 publications

References 0 publications

Radioactivity evaluation of the secondary sodium in DRACS of the Japan Sodium-cooled Fast Reactor

Radioactivity evaluation of the secondary sodium in DRACS of the Japan Sodium-cooled Fast Reactor

CAD and constructive solid geometry modeling of the Molten Salt Reactor Experiment with OpenMC

CAD-based Monte Carlo automatic modeling method based on primitive solid

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