• Journal of Biomedical Engineering Research
• /
• v.19 no.3
• /
• pp.225-232
• /
• 1998

Results of Monte Carlo simulations on amorphous silicon based x-ray imaging arrays are described. In order to investigate the characteristics of amorphous silicon x-ray imaging devices and to provide the optimum design parameter, Monte Carlo simulations were performed. Monte Carlo simulation codes for our purpose were developed and various combinations of x-ray peak voltages, aluminum filter thicknesses, CsI(TI) thicknesses, and amorphous silicon photodiode pixel sizes were tested in connection with detection efficiency and spatial resolution of the amorphous silicon based x-ray imager. With usual Csl(TI) thickness of 300${\mu}{\textrm}{m}$-500${\mu}{\textrm}{m}$, detection efficiency was in the range of 70%-95% and energy absorption efficiency was in the range of 40%-70% for 60kVp-120kVp x-ray. From the simulations it was found that amorphous silicon pixel size and Csl(TI) thickness were the most important parameters which determine the resolution of the imager. By use of our simulation results we could provide proper combinations of Csl(TI) thicknesses and pixels sizes for optimum sensitivity and resolution.

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

• Journal of the Korean Institute of Telematics and Electronics D
• /
• v.35D no.3
• /
• pp.28-38
• /
• 1998

In our paper is reported a new 3D(dimensional) trajectory split approach with greatly improved efficiency for the Monte Carlo simulation of the 3D profiles of implanted ionand point defect concentrations in single-crystal silicon. This approach has been successfully implemented in our TRICSI Monte Carlo code. Combined with the previously developed model for damage accumalation in our TRICSI code, this model allows phasically based dynamic simulation of 3D profiles over an subsequent process simulation such as diffusion modeling and simulation. A typical time saving of over 10 timeshas been achieved for 3D simulation. Our method ensures much better region aground the implanted area. For 1-D simulation, the optimized condition for trajectory split has set to 3,000 pseudoparticles with 2 split branches.

• The Transactions of The Korean Institute of Electrical Engineers
• /
• v.61 no.12
• /
• pp.1880-1885
• /
• 2012

Nonlinear dynamical behaviors in thermionic low pressure discharge are investigated using a particle-in-cell(PIC) simulation. An electrostatic PIC code is developed to model the plasma discharge system including the kinetic effects. The elastic collision, excitation collision, ionization collision, and electron-ion recombination collision are considered in this code. The generated electrons and ions are traced to analyze physical characteristics of the plasma. The simulation results show that the nonlinear oscillation structures are observed for cold plasma in the system and the similar structures are observed for warm plasma with a shift in values of the bifurcation parameter. The detailed oscillation process can be subdivided into three distinct mode; anode-glow, temperature-limited, and double-layer modes.

• Nuclear Engineering and Technology
• /
• v.51 no.5
• /
• pp.1209-1217
• /
• 2019

Due to the limitation of the computers, the conventional homogenization method is based on many assumptions and approximations, and some tough problems such as energy spectrum and boundary condition are faced. To deal with those problems, the Monte Carlo global homogenization is adopted. The Reactor Monte Carlo code RMC is used to study the global homogenization method, and the one-step global homogenization method is proposed. The superimposed mesh geometry is also used to divide the physical models, leading to better geometric flexibility. A set of multigroup homogenization cross sections is online generated for each mesh under the real neutron energy spectrum and boundary condition, the cross sections are adjusted by the superhomogenization method, and no leakage correction is required. During the process of superhomogenization, the author-developed reactor core program NLSP3 is used for global calculation, so the global flux distribution and equivalent homogenization cross sections could be solved simultaneously. Meanwhile, the calculated homogenization cross section could accurately reconstruct the non-homogenization flux distribution and could also be used for fine calculation. This one-step global homogenization method was tested by a PWR assembly and a small reactor model, and the results show the validity.

• Nuclear Engineering and Technology
• /
• v.53 no.1
• /
• pp.1-18
• /
• 2021

The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main purpose of this work is to evaluate the newest ENDF/B-VIII.0 nuclear data library against the VVER-1000 mock-up integral experiments and to validate the criticality analysis capability of MCS for light water reactors with hexagonal fuel lattices. A preliminary code/code comparison between MCS and MCNP6 is first conducted to verify the suitability of MCS for the benchmark interpretation, then the validation against experimental data is performed with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The investigated experimental data comprises six experimental critical configurations and four experimental pin-by-pin power maps. The MCS and MCNP6 inputs used for the criticality analysis of the VVER-1000 mock-up are available as supplementary material of this article.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1120-1125
• /
• 2016

The neutron decay constant, ${\alpha}$, and effective delayed neutron fraction, ${\beta}_{eff}$, are important parameters for the control of the dynamic behavior of nuclear reactors. For the heavy water zero power reactor (HWZPR), this document describes the measurements of the neutron decay constant by noise analysis methods, including variance to mean (VTM) ratio and endogenous pulse source (EPS) methods. The measured ${\alpha}$ is successively used to determine the experimental value of the effective delayed neutron fraction as well. According to the experimental results, ${\beta}_{eff}$ of the HWZPR reactor under study is equal to 7.84e-3. This value is finally used to validate the calculation of the effective delayed neutron fraction by the Monte Carlo methods that are discussed in the document. Using the Monte Carlo N-Particle (MCNP)-4C code, a ${\beta}_{eff}$ value of 7.58e-3 was obtained for the reactor under study. Thus, the relative difference between the ${\beta}_{eff}$ values determined experimentally and by Monte Carlo methods was estimated to be < 4%.

• Nuclear Engineering and Technology
• /
• v.53 no.2
• /
• pp.430-438
• /
• 2021

The spatial distribution of neutron flux or reaction rate was calculated by cell or mesh tally in traditional Monte Carlo simulation. However, either cell or mesh tally leads to the increase of memory consumption and simulation time. In this paper, the function expansion tally (FET) method was developed in Reactor Monte Carlo code RMC to solve this problem. The FET method was applied to the tallies of neutron flux distributions of uranium block and PWR fuel rod models. Legendre polynomials were used in the axial direction, while Zernike polynomials were used in the radial direction. The results of flux, calculation time and memory consumption of different expansion orders were investigated, and compared with the mesh tally. Results showed that the continuous distribution of flux can be obtained by FET method. The flux distributions were consistent with that of mesh tally, while the memory consumption and simulation time can be effectively reduced. Finally, the convergence analysis of coefficients of polynomials were performed, and the selection strategy of FET order was proposed based on the statistics uncertainty of the coefficients. The proposed method can help to determine the order of FET, which was meaningful for the efficiency and accuracy of FET method.

• Nuclear Engineering and Technology
• /
• v.52 no.12
• /
• pp.2852-2859
• /
• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Nuclear Engineering and Technology
• /
• v.54 no.9
• /
• pp.3317-3323
• /
• 2022

Half Value Layer calculations theoretically need prior specification of linear attenuation calculations, since the HVL value is derived by dividing ln(2) by the linear attenuation coefficient. The purpose of this study was to establish a direct computational model for determining HVL, a vital parameter in nuclear radiation safety studies and shielding material design. Accordingly, a typical gamma-ray transmission setup has been modeled using MCNPX (version 2.4.0) general-purpose Monte Carlo code. The MCNPX code's INPUT file was designed with two detection locations for primary and secondary gamma-rays, as well as attenuator material between those detectors. Next, Half Value Layer values of some well-known gamma-ray shielding materials such as lead and ordinary concrete have been calculated throughout a broad gamma-ray energy range. The outcomes were then compared to data from the National Institute of Standards and Technology. The Half Value Layer values obtained from MCNPX were reported to be highly compatible with the HVL values obtained from the NIST standard database. Our results indicate that the developed INPUT file may be utilized for direct computations of Half Value Layer values for nuclear safety assessments as well as medical radiation applications. In conclusion, advanced simulation methods such as the Monte Carlo code are very powerful and useful instruments that should be considered for daily radiation safety measures. The modeled MCNPX input file will be provided to the scientific community upon reasonable request.