• Journal of computational fluids engineering
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• v.4 no.3
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• pp.63-70
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• 1999

A Direct simulation Monte-Carlo (DSMC) code is developed, which employs the Monte-Carlo statistical sampling technique to investigate hypersonic rarefied gas flows accompanying chemical reactions. The DSMC method is a numerical simulation technique for analyzing the Boltzmann equation by modeling a real gas flow using a representative set of molecules. Due to the limitations in computational requirements. the present method is applied to a flow around a simple two-dimensional object in exit velocity of 7.6 km/sec at an altitude of 90 km. For the calculation of chemical reactions an air model with five species (O₂, N₂, O, N, NO) and 19 chemical reactions is employed. The simulated result showed various rarefaction effects in the hypersonic flow with chemical reactions.

• Journal of Korean Society for Quality Management
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• v.43 no.3
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• pp.239-252
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• 2015

Purpose: Lead has been widely used in radiation shielding for its low price and high workability. Recently in several europe countries, use of lead was banned for environmental issues. Also lead can cause health problems like alergies. Alternative materials for lead are highly required. The purpose of this study was to propose lead free radiation shielding material. Methods: Research of radiation shielding in Korea is not easy for certain limits such as radiation materials, experimental facilities and places. The collected data through the research were simulated using MCNPX. The simulation tools used for this study were utilized Monte Carlo method. Results: we suggest new design of lead free radiation shielding material using MCNPX code comparing shielding performance of new composite materials to lead. Conclusion: This newly introduced nano-scale composite of metal and polymer makes new chance for highly lightened radiation protective garments with endurable shielding performance.

• Journal of Radiation Protection and Research
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• v.42 no.3
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• pp.154-157
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• 2017

Background: Shielding properties of compound or mixture is presented in terms of mass attenuation coefficients using Monte Carlo simulation. Mass attenuation coefficients of cement, gypsum and the mixture of gypsum and $PbCO_3$ has been investigated using monte carlo MCNPX. Materials and Methods: The mass attenuation coefficients of cement, gypsum and the mixture of gypsum and $PbCO_3$ were calculated for photon energies 365.5, 661.6, 1,173.2, and 1,332.5 keV energies. Results and Discussion: The simulated values of mass attenuation coefficients were compared avaialable experimental results, theoretical values by XCOM and found good comparability of the results. Conclusion: Standard simulation geometry used in the present investigation would be very useful for various types of sample for shielding and dosimetry applications.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2004.05b
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• pp.321-323
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• 2004

We developed the monte-carlo simulation code for analysis of the On signal in dental medium. In calculation, we obtain the two different propagation signals as a function of the probing depth. Signal 2 begins to exceed the signal 1 at a very small probing depth(=60${\mu}{\textrm}{m}$). For reduce the signal, detection area is limited to radius and detection angle. As numerical result, probing depth becomes appoximately 500${\mu}{\textrm}{m}$.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.920-927
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• 2017

With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.161-176
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• 2012

McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.439-445
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• 2018

Experimental measurements of the response of $BF_3$ detector to a 3 Ci Am-Be neutron source for three different reflecting materials, i.e., aluminum, wood, and Perspex of varying thicknesses have been carried out. The varying contribution of wall effect to the response due to change in active volume of the detector has also been determined experimentally. Then, a Monte Carlo code has been developed for the calculation of the neutron response function of the $BF_3$ detector using source biasing and importance sampling. This code simulates the $BF_3$ detector response exposed to the neutron field in a three-dimensional source, detector, and reflecting medium configurations. The results of simulation have been compared with the corresponding experimental measurements and are found to be in good agreement. The experimental neutron albedo measurements for various values of Perspex thickness show saturating behavior, and results agree very well with the data obtained by Monte Carlo simulation.

• Proceedings of the Korean Society of Medical Physics Conference
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• 2004.11a
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• pp.36-38
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• 2004

The detector size effect due to the spatial response of defectors is one critical source of inaccuracy in clinical dosimetry and has been a subject of numerous studies. Conventionally, the detector response kernel contains all of the influence that the detector size has on the measured beam profile. Various analytic models for this kernel have been proposed and studied in theoretical and experimental works. Here, we use a method to determine detector response kernel simply by using Monte Carlo simulation and convolution theory. Based on this numerical method and DOSIMETER, an EGS4 Monte Carlo code, the detector response for a Farmer type ion chamber embedded in water phantom is obtained. There exists characteristic difference in the simulated chamber readings between one with carbon graphite wall and the other with Acrylic wail. Using the obtained response and the convolution theory, we are planning to derive the detector response kernel numerically and remove detector size effect from measurements for 6MV, 10${\times}$l0cm2 and 0.5${\times}$10 cm2 photon beam.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.916-921
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• 2024

The control rod worth sensitivity to the neutron detector position in the rod drop experiment is studied by the time-dependent Monte Carlo (TDMC) neutron transport calculations for AGN-201K educational reactor and the Kyoto University Critical Assembly. The TDMC simulations of the rod drop experiments are conducted by the Seoul National University Monte Carlo (MC) code, McCARD, yielding time-dependent neutron densities at detector positions. The detector-position-dependent results of the total control rod worth calculated by the extrapolation, the integral counting, and the inverse methods are compared with the numerical reference using the MC eigenvalue calculations and the experimental results. From these comparisons, it is observed that the total control rod worth can be estimated with a considerable difference depending on the detector position through the rod drop experiment. The proposed TDMC simulation of the rod drop experiment can be applied for searching a better detector position or quantifying a bias for the control rod worth measurement.

• Proceedings of the Korean Society of Medical Physics Conference
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• 2002.09a
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• pp.291-293
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• 2002

Accurate dose calculation in radiation treatment planning is most important for successful treatment. Since human body is composed of various materials and not an ideal shape, it is not easy to calculate the accurate effective dose in the patients. Many methods have been proposed to solve the inhomogeneity and surface contour problems. Monte Carlo simulations are regarded as the most accurate method, but it is not appropriate for routine planning because it takes so much time. Pencil beam kernel based convolution/superposition methods were also proposed to correct those effects. Nowadays, many commercial treatment planning systems, including Pinnacle and Helax-TMS, have adopted this algorithm as a dose calculation engine. The purpose of this study is to verify the accuracy of the dose calculated from pencil beam kernel based treatment planning system Helax-TMS comparing to Monte Carlo simulations and measurements especially in inhomogeneous region. Home-made inhomogeneous phantom, Helax-TMS ver. 6.0 and Monte Carlo code BEAMnrc and DOSXYZnrc were used in this study. Dose calculation results from TPS and Monte Carlo simulation were verified by measurements. In homogeneous media, the accuracy was acceptable but in inhomogeneous media, the errors were more significant.