• Journal of the Korean Society of Manufacturing Process Engineers
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• v.20 no.9
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• pp.11-19
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• 2021

Linear variable differential transformer (LVDT) is a displacement sensor and is commonly used owing to its wide measurement range, excellent linearity, high sensitivity, and precision. To improve the output characteristics of LVDT, a few studies have been conducted to analyze the output using a theoretical method or a finite element method. However, the material properties of the core and the electromagnetic force acting on the core were not considered in the previous studies. In this study, a finite element analysis model was proposed considering the characteristics of the LVDT composed of coils, core, magnetic shell and electric circuit, and the core displacement. Using the proposed model, changes in sensitivity and linear region of LVDT according to changes in process and material parameters were analyzed. The outputs of the LVDT model were compared with those of the theoretical analysis, and then, the proposed analysis model was validated. When the electrical conductivity of the core was high and the relative magnetic permeability was low, the decrease in sensitivity was large. Additionally, an increase in the frequency of the power led to further decrease in sensitivity. The electromagnetic force applied on the core increased as the voltage increased, the frequency decreased, and the core displacement increased.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.13 no.5
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• pp.70-75
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• 2014

Permanent magnets have recently been considered as device that can be used to control the behavior of mechanical systems. Neodymium magnets, a type of permanent magnet, have been used in numerous mechanical devices. These are permanent magnets made from an alloy of neodymium, iron, and boron to form the Nd2Fe14B tetragonal crystalline structure. The magnetic selection, magnet core design and mechanical errors of the magnetic component can affect the performance of the magnetic force. In this study, the coercive force, residual induction, and the dimensions of the design parameters of the magnet core are optimized. The design parameters of magnet core are defined as the gap between the magnet and the core, the upper contact radius, and the lower thickness of the core. The force exercised on a permanent magnet in a non-uniform field is dependent on the magnetization orientation of the magnet. Non-uniformity of the polarization direction of the magnetic has been assumed to be caused by the angular error in the polarization direction. The variation in the magnetic performance is considered according to the center distance, the tilt of the magnetic components, and the polarization direction. The finite element method is used to analyze the magnetic force of an optimized cylindrical magnet.

• The Proceedings of the Korean Institute of Illuminating and Electrical Installation Engineers
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• v.5 no.3
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• pp.78-83
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• 1991

This study proposes a unified method to design toroidal transformer and its optimal design algorithm. This unified design method was derived from the fundamental equation of power on the basis of electromagnetic energy of a core and the definition of three parameters(K1, K2, KW) that influence the form of a core and the ratio of a core and coil. Accordingly this design method condenses the whole data for design of toroidal transformer to a standard variable which is the inner diameter of a core. The minimal cost, weight and volume values of the transformer were computed by means of the algorithm to search the optimal values of the parameters. Furthermore, through the CAD, the efficiency of this unified design method and optimal algorithm proposed in this paper was confirmed.

• Proceedings of the KSME Conference
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• 2001.11a
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• pp.703-709
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• 2001

In this paper, we analyzed the laminated composite sandwich plate structure of truss core with changing values of the designing parameters. As a result, in designing parameters of that, the more height and thickness of the laminated composite plate's core, the more increase of natural frequency. In this type of structure, in the case of applying core of the laminated composite plate and antisymmetric stacking, natural frequency has high value and we calculated the optimum angle-ply making natural frequency maximum. Natural frequency of CFRP is higher than that of GFRP. Both are materials of the laminated composite plate. The mode shapes are various along with the angle-ply of the laminated composite plate.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1731-1735
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• 2021

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for ²³⁹Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1775-1782
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• 2023

Several reactor physics commissioning experiments were conducted to obtain the neutronic parameters at the beginning of the G.A. Siwabessy Multi-purpose Reactor (RSG-GAS) operation. These parameters are essential for the reactor to safety operate. Leveraging the experimental data, this study evaluated the calculated core reactivity, control rod reactivity worth, integral control rod reactivity curve, and fuel reactivity. Calculations were carried out with Serpent 2 code using the latest neutron cross-section data ENDF/B-VIII.0. The criticality calculations were carried out for the RSG-GAS first core up to the third core configuration, which has been done experimentally during these commissioning periods. The excess reactivity for the second and third cores showed a difference of 510.97 pcm and 253.23 pcm to the experiment data. The calculated integral reactivity of the control rod has an error of less than 1.0% compared to the experimental data. The calculated fuel reactivity value is consistent with the measured data, with a maximum error of 2.12%. Therefore, it can be concluded that the RSG-GAS reactor core model is in good agreement to reproduce excess reactivity, control rod worth, and fuel element reactivity.

• Proceedings of the KIEE Conference
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• 2005.07c
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• pp.2132-2135
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• 2005

Cylindrical type air core pulse transformers capable of passing high voltage and energy pulse waveforms with high efficiency and low distortion require a much more delicate design balance of physical dimensions and electrical parameters than iron or ferrite core units. The structure of an air core high voltage pulse transformer is relatively simple, but considerable attention is needed to prevent breakdown between transformer windings. Since the thickness of the windings in spiral type is on the order of sub-millimeter, field enhancement at the edge of the windings is very high. It is, therefore, important to find proper electrical insulation Parameter to make the system compact. Several shapes of the winding are considered for air core pulse transformer development. In this paper, we are described design procedure, parameters measure and experiment results of air core type HV pulse transformer.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1037-1042
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• 2018

The present study aims to assess the excess induced reactivity in a Miniature Neutron Source Reactor (MNSR) for a Beyond Design Basis Accident (BDBA) scenario. The BDBA scenario as defined in the Safety Analysis Report (SAR) of the reactor involves sticking of the control rod and filling of the inner and outer irradiation sites with water. At the end of the MNSR core life, 10.95 cm of Beryllium is added to the top of the core as a reflector which affects some neutronic parameters such as effective delayed neutrons fraction (${\beta}_{eff}$), the reactivity worth of inner and outer irradiation sites that are filled with water and the reactivity worth of the control rod. Given those influences and changes, new neutronic calculations are required to be able to demonstrate the reactor safety. Therefore, a validated MCNPX model is used to calculate all neutronic parameters at the end of the reactor core life. The calculations show that the induced reactivity in the BDBA scenario increases at the end of core life to $7.90{\pm}0.01mk$ which is significantly higher than the induced reactivity of 6.80 mk given in the SAR of MNSR for the same scenario but at the beginning of the core's life. Also this value is 3.90 mk higher than the maximum allowable operational limit (i.e. 4.00 mk).

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.55-63
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• 2016

Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy ($E_e$) and source multiplication coefficient ($k_s$), has been investigated. A Monte Carlo code (MCNPX_2.6) has been used to calculate neutronic parameters such as effective multiplication coefficient ($k_{eff}$), net neutron multiplication (M), neutron yield ($Y_{n/e}$), energy constant gain ($G_0$), energy gain (G), importance of neutron source (${\varphi}^*$), axial and radial distributions of neutron flux, and power peaking factor ($P_{max}/P_{ave}$) in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current ($I_e$) have been decreased in the highest case of $k_s$, but G and ${\varphi}^*$ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing $E_e$ from 100 MeV up to 1 GeV, $Y_{n/e}$ and G improved by 91.09% and 10.21%, and $I_e$ and $P_{acc}$ decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np-Pu assemblies on the periphery allows for a consistent $k_{eff}$ because the Np-Pu assemblies experience less burn-up.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2725-2732
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• 2020

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of ¹⁶O, ⁹Be, ²³⁵U, ²³⁸U, and S(𝛼,𝛽) of ¹H in H₂O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.