• Nuclear Engineering and Technology
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• v.38 no.3
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• pp.225-230
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• 2006

With the advent of inexpensive computing power over the past two decades, applications of Monte Carlo radiation transport techniques have proliferated dramatically. At Los Alamos, the Monte Carlo codes MCNP5 and MCNPX are used routinely on personal computer platforms for radiation shielding analysis and dosimetry calculations. These codes feature a rich palette of variance reduction (VR) techniques. The motivation of VR is to exchange user efficiency for computational efficiency. It has been said that a few hours of user time often reduces computational time by several orders of magnitude. Unfortunately, user time can stretch into the many hours as most VR techniques require significant user experience and intervention for proper optimization. It is the purpose of this paper to outline VR strategies, tested in practice, optimized for several common radiation shielding tasks, with the hope of reducing user setup time for similar problems. A strategy is defined in this context to mean a collection of MCNP radiation transport physics options and VR techniques that work synergistically to optimize a particular shielding task. Examples are offered in the areas of source definition, skyshine, streaming, and transmission.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.323-327
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• 2020

In this study, the neutronic calculation to obtain tritium breeding ratio (TBR) in a deuterium-tritium (D-T) fusion power reactor using Monte Carlo MCNPX is done. In addition, by using COMSOL software, an efficient cooling system is designed. In the proposed design, it is adequate to enrich up to 40% ⁶Li. Total tritium breeding ratio of 1.12 is achieved. The temperature of helium as coolant gas never exceed 687℃. As regards the tolerable temperature of beryllium (650℃), the design of blanket module is done in the way that beryllium temperature never exceed 600℃. The main feature of this design indicates the temperature of helium coolant is higher than other proposed models for blanket module, therefore power of electricity generation will increase.

• Proceedings of the Korea Contents Association Conference
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• 2014.11a
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• pp.229-230
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• 2014

본 연구는 소아백혈병의 치료 방법 중 하나인 조혈모세포이식법의 전처치로서 사용되고 있는 전신방사선조사법에 대하여 선량분포에 대한 연구를 진행한 것으로, MCNPX 프로그램을 이용하여 모의실험을 하였다. 결과 피부선량은 평균 112.43 mGy/min, 심부장기선량은 평균 47.52 mGy/min으로 나타났으며, 조직보상체의 재질과 거리에 따라 다르게 나타나는 경향성을 볼 수 있었으며, 결과를 바탕으로 전신방사선조사를 임하기 전에 정량적인 선량평가를 할 수 있을 것으로 생각된다.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1075-1080
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• 2019

This paper presents a study on thermal neutron reflection properties of neutron guide for cylinder, spindle, elliptic and parabolic geometries using $^{241}Am-Be$ neutron source (5.2 Ci) and $BF_3$ detector, whereas neutron guide is important instrument for transportation of neutrons. To this goal, the required inner and outer radii of neutron guide have been calculated to achieve the highest guided thermal neutron flux based on MCNPX Monte Carlo code. The maximum flux of cylinder geometry with a length 50 cm has been obtained at an inner radius 9 cm and an outer radius 21 cm. Also, the maximum value of thermal neutron albedo is $0.46{\pm}0.001$ at 12 cm thickness of parabolic guide.

• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1349-1354
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• 2018

Inductively Coupled Plasma Atomic Emission Spectroscopy (ICP-AES) is widely used for the determination of trace elements in geological samples in Iran. In this paper, we have calculated the detection limits of neutron activation analysis (NAA) for some of the common elements in such samples utilizing the ORIGEN and MCNP codes and verified the simulations using the experimental results of three soil standard reference materials, namely, G02.SRM, G18.SRM, and G28.SRM. The results show that while the detection limit of ICP-AES method is usually in the mg/kg range, it is represented to the ${\mu}g/kg$ range for most of the elements of interest using the NAA method, and the simulations can be verified in a tolerance range of 20%.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2564-2569
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• 2021

The neutronic analysis of Helium Cooled Ceramic Reflector (HCCR) breeding blankets has been performed using the 3D Monte Carlo code MCNPX and ENDF nuclear data library. This study aims to reduce ⁶Li percentage in the breeder zones as much as possible ensuring tritium self-sufficiency. This work is devoted to investigating the effect of ⁶Li percentage on the HCCR breeding blanket's neutronic parameters, such as neutron flux and spectrum, Tritium Breeding Ratio (TBR), nuclear power density, and energy multiplication factor. In the ceramic breeders at the saturated thickness, increasing the enrichment of ⁶Li reduces its share in the tritium production. Therefore, ceramic breeders typically use lower enriched Li from 30% to 60%. The investigation of neutronic analysis in the suggested geometry shows that using 60% ⁶Li in Li₂TiO₃ can yield acceptable TBR and energy deposition results, which would be economically feasible.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3390-3397
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• 2022

Among the different techniques available, nuclear methods, including gamma-gamma logging tools, are of special importance. Though the real environment which surrounds the drilled borehole is a complex fractured medium which the fluid can flow through the porosities, simulation studies generally use the traditional model of a homogeneous mixture of formation and the liquid. Considering a previously published study, which shows that modeling of fluid flow in fractured reservoirs and simulating the formation as an inhomogeneous fractured medium leads to different results compared with those of homogeneous mixture, here we study the effect of the presence of drilling fluid (mudcake) on the response of the detectors in both the models. To study this effect, a typical gamma-gamma logging tool was modeled by using the MCNPX Monte Carlo code. The results show that the responses of the detectors in the mixture model in the presence of various thicknesses of mudcake are sensitive to the density of the formation material. However, this effect is not notable in the inhomogeneous fractured medium. These results emphasize the importance of the model employed for simulation of the medium in gamma-gamma well-logging.

• Journal of Radiation Protection and Research
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• v.33 no.4
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• pp.167-172
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• 2008

Recently a high-quality voxel model of a Korean adult male was constructed at Hanyang University by using very high resolution serially-sectioned anatomical images of a cadaver, which was provided by the Korean Institute of Science and Technology Information (KISTI). Most existing voxel phantoms are developed based on an individual in the supine posture. This study converted the HDRK-Man voxel model into surface model and adjusted the flattened back of the HDRK-Man to a normal shape in the upright posture using 3D graphic softwares such as $3D-DOCTOR^{TM}$, $Rapidform^{(R)}$2006, $Rhinoceros^{(R)}$4.0, $MAYA^{(R)}$8.5. The effective doses of adjusted model were compared with those of unadjusted model for some standard irradiation geometries (i.e., AP, PA, LLAT, RLAT). In general, the differences were not very large and, among those, the largest difference was found for the PA radiation geometry, as expected. These methodologies can be used for the development of various deformed posture models of HDRK-Man in the later stage of this project.

• Journal of the Korean Vacuum Society
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• v.15 no.1
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• pp.103-109
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• 2006

A target for a high-brightness microfocus x-ray tube, which is based on carbon nanotubes (CNT) as electron source, is designed. The x-ray tube has the following specifications: brightness of $1\times10^{11}phs/s.mm^2. mrad^2$, spot size $\~5{\mu}m$, and average x-ray energy of $20\~40 keV$. In order to meet the specifications, the design parameters of the target, such as configuration, material, thickness of the target as well as the required beam current, were optimized using computer code MCNPX. The design parameters were determined from the calculation of both x-ray spectrum and intensity distribution for a transmission type configuration. For the thin transmission type target to withstand vacuum pressure and localized thermal loading, the structural stability and temperature distribution were also considered. The material of the target was selected as molybdenum(Mo) and the optimized thickness was $7.2{\mu}m$ to be backed by $150{\mu}m$ beryllium (Be). In addition, the calculations revealed that the maximum temperature of the transmission target can be maintained within the limits of stable operation.

• Progress in Medical Physics
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• v.23 no.3
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• pp.162-168
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• 2012

In proton therapy, in vivo dose verification is one of the most important parts to fully utilize characteristics of proton dose distribution concentrating high dose with steep gradient and guarantee the patient safety. Currently, in order to image the proton dose distribution, a prompt gamma distribution detection system, which consists of an array of multiple CsI(Tl) scintillation detectors in the vertical direction, a collimator, and a multi-channel DAQ system is under development. In the present study, the optimal design of prompt gamma distribution detection system was studied by Monte Carlo simulations using the MCNPX code. For effective measurement of high-energy prompt gammas with enough imaging resolution, the dimensions of the CsI(Tl) scintillator was determined to be $6{\times}6{\times}50mm^3$. In order to maximize the detection efficiency for prompt gammas while minimizing the contribution of background gammas generated by neutron captures, the hole size and the length of the collimator were optimized as $6{\times}6mm^2$ and 150 mm, respectively. Finally, the performance of the detection system optimized in the present study was predicted by Monte Carlo simulations for a 150 MeV proton beam. Our result shows that the detection system in the optimal dimensions can effectively measure the 2D prompt gamma distribution and determine the beam range within 1 mm errors for 150 MeV proton beam.