• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.57-61
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• 1997

A Canadian Deuterium Uranium (CANDU) reactor core calculation was performed using lattice parameters generated by HELIOS. The HELIOS-based lattice parameters were processed by TABGEN in a form suitable for the core analysis code RFSP. The core calculation was performed and the results were compared to those of the reference calculation which uses POWDERPUFS-V (PPV) for the lattice parameter generation. The characteristics of the core calculated based on the PPV and HELIOS lattice parameters match within 0.4%$\Delta$k and 7% for the excess reactivity and the channel power distribution, respectively.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.495-501
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• 1997

A preliminary and parametric sensitivity study on recriticality risk of fast reactor cores after a hypothetical total core meltdown accident was performed. Only neutronic aspects of the accident were considered, independent of the accident scenario, and efforts were made to estimate the quantity of molten fuel which must be ejected out of the core to assure a sub-critical state after the accident. Two types of parameters were examined : characteristic parameters of molten core such as geometry, molten pool type (homogenized or stratified), fuel temperature, environment, and relative parameters to normal core such as core size(small or large), and fuel type (oxide, nitride, metal). The first type of parameters was found to intervene more directly in the recriticality risk than the second type of parameters.

• Interaction and multiscale mechanics
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• v.1 no.2
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• pp.231-250
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• 2008

The statistical two-order and two-scale method is developed for predicting the mechanics parameters, such as stiffness and strength of core-shell particle-filled polymer composites. The representation and simulation on meso-configuration of random particle-filled polymers are stated. And the major statistical two-order and two-scale analysis formulation is briefly given. The two-order and two-scale expressions for the strains and stresses of conventionally strength experimental components, including the tensional or compressive column, the twist bar and the bending beam, are developed by means of their classical solutions with orthogonal-anisotropic coefficients. Then a new effective mesh generation algorithm is presented. The mechanics parameters of core-shell particle-filled polymer composites, including the expected stiffness parameters, minimum stiffness parameters, and the expected elasticity limit strength and the minimum elasticity limit strength, are defined by means of the stiffness coefficients and elasticity strength criterions for core, shell and matrix. Finally, the numerical results for predicting both stiffness and elasticity limit strength parameters are compared with the experimental data.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.291-298
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• 2014

This paper concerns estimating uncertainties of the core neutronics design parameters of power reactors by direct sampling method (DSM) calculations based on the two-step McCARD/MASTER design system in which McCARD is used to generate the fuel assembly (FA) homogenized few group constants (FGCs) while MASTER is used to conduct the core neutronics design computation. It presents an extended application of the uncertainty propagation analysis method originally designed for uncertainty quantification of the FA FGCs as a way to produce the covariances between the FGCs of any pair of FAs comprising the core, or the covariance matrix of the FA FGCs required for random sampling of the FA FGCs input sets into direct sampling core calculations by MASTER. For illustrative purposes, the uncertainties of core design parameters such as the effective multiplication factor ($k_{eff}$), normalized FA power densities, power peaking factors, etc. for the beginning of life (BOL) core of Yonggwang nuclear unit 4 (YGN4) at the hot zero power and all rods out are estimated by the McCARD/MASTER-based DSM computations. The results are compared with those from the uncertainty propagation analysis method based on the McCARD-predicted sensitivity coefficients of nuclear design parameters and the cross section covariance data.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.677-690
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• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4583-4590
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• 2023

This study aims at modeling the beryllium reflector poisoning under neutron irradiation conditions and calculating the impact of beryllium poisoning on the core parameters of ETRR-2 research reactor. The CITVAP code was used to calculate the neutron flux and parameters of ETRR-2 core with beryllium reflector elements. The neutron flux in each reflector element was calculated to solve the modeling equations for the atomic densities of lithium-6 (⁶Li), tritium-3 (³H), and helium-3 (³He) using the BERYL program. The results are discussed based on CITVAP calculations of the core excess reactivity and cycle length Full Power Days (FPD). Possible solutions to minimize the degradation due to beryllium poisoning are also discussed and compared based on calculations.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1017-1023
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• 2018

The measurement and calculation of neutronic parameters in nuclear research reactors has an important influence on control and safety of the nuclear reactor. The power peaking factors, reactivity coefficients and kinetic parameters are the most important neutronic parameter for determining the state of the reactor. The position of the control shim safety rods in the core configuration affects these parameters. The main purpose of this work is to use the MTR_PC package to evaluate the effect of the partially insertion of the control rod on the neutronic parameters at the operating core of the Tehran Research Reactor. The simulation results show that by increasing the insertion of control rods (bank) in the core, the absolute values of power peaking factor, reactivity coefficients and effective delayed neutron fraction increased and only prompt neutron life time decreased. In addition, the results show that the changes of moderator temperature coefficients value versus the control rods positions are very significant. The average value of moderator temperature coefficients increase about 98% in the range of 0-70% insertion of control rods.

• Nuclear Engineering and Technology
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• v.46 no.4
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• pp.513-520
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• 2014

In this paper, three types of operational and industrial absorbers used at research reactors, including Ag-In-Cd alloy, $B_4C$, and Hf are selected for sensitivity analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, thermal neutron flux, power density distribution, and Power Peaking Factor (PPF). The IAEA 10 MW benchmark core is selected as the case study to verify calculations. A two-dimensional, three-group diffusion model is selected for core calculations. The well-known WIMS-D4 and CITATION reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the $B_4C$; also the lowest PPF is gained using the Ag-In-Cd alloy. The maximum point power densities belong to the inside fuel regions surrounding the central flux trap (irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the graphite reflectors. The greatest and least fluctuation of the point power densities are gained by using $B_4C$ and Ag-In-Cd alloy, respectively.

• Proceedings of the KSME Conference
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• 2001.11a
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• pp.616-622
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• 2001

In this paper, we analyzed the laminated composite sandwich plate structure of honeycomb 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. The laminated angle has the maximum value when the plate of honeycomb core is join to opposite direction. This paper shows that the natural frequency of CFRP is higher than that of GFRP, and also impact strength marks maximum value in case of antisymmetry than symmetry of core. Also it shows that the mode shapes are various along with the angle-ply of laminated composite plate.

• 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.