• Transactions on Electrical and Electronic Materials
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• v.17 no.4
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• pp.189-195
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• 2016

Micro Hotplate (MHP) is the key component in micro-sensors, particularly gas sensors. Indeed, in metal oxide gas sensors MOX, micro-heater is used as a hotplate in order to control the temperature of the sensing layer which should be in the requisite temperature range over the heater area, so as to detect the resistive changes as a function of varying concentration of different gases. Hence, their design is a very important aspect. In this paper, we have presented the design and simulation results of a meander micro heater based on three different materials - platinum, titanium and tungsten. The dielectric membrane size is 1.4 mm × 1.6 mm with a thickness of 1.4 μm. Above the membrane, a meander heating film was deposed with a thickness of 100 nm. In order to optimize the geometry, a comparative study by simulating two different heater thicknesses, then two inter track widths has also been presented. Power consumption and temperature distribution were determined in the micro heater´s structure over a supply voltage of 5, 6, and 7 V.

• Korean Journal of Materials Research
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• v.27 no.12
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• pp.679-687
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• 2017

A strain-gradient crystal plasticity finite element method(SGCP-FEM) was utilized to simulate the compressive deformation behaviors of single-slip, (111)[$10{\bar{1}}$], oriented FCC single-crystal micro-pillars with two different slip-plane inclination angles, $36.3^{\circ}$ and $48.7^{\circ}$, and the simulation results were compared with those from conventional crystal plasticity finite element method(CP-FEM) simulations. For the low slip-plane inclination angle, a macroscopic diagonal shear band formed along the primary slip direction in both the CP- and SGCP-FEM simulations. However, this shear deformation was limited in the SGCP-FEM, mainly due to the increased slip resistance caused by local strain gradients, which also resulted in strain hardening in the simulated flow curves. The development of a secondly active slip system was altered in the SGCP-FEM, compared to the CP-FEM, for the low slip-plane inclination angle. The shear deformation controlled by the SGCP-FEM reduced the overall crystal rotation of the micro-pillar and limited the evolution of the primary slip system, even at 10 % compression.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1269-1286
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• 2017

This paper presents a newly developed resonance self-shielding method based on Tone's method in $APOLLO3^{(R)}$ for fast and thermal reactor calculations. The new method is based on simplified models, the narrow resonance approximation for the slowing down source and Tone's approximation for group collision probability matrix. It utilizes mathematical probability tables as quadrature formulas in calculating effective cross-sections. Numerical results for the ZPPR drawer calculations in 1,968 groups show that, in the case of the double-column fuel drawer, Tone's method gives equivalent precision to the subgroup method while markedly reducing the total number of collision probability matrix calculations and hence the central processing unit time. In the case of a single-column fuel drawer with the presence of a uranium metal material, Tone's method obtains less precise results than those of the subgroup method due to less precise heterogeneous-homogeneous equivalence. The same options are also applied to PWR UOX, MOX, and Gd cells using the SHEM 361-group library, with the objective of analyzing whether this energy mesh might be suitable for the application of this methodology to thermal systems. The numerical results show that comparable precision is reached with both Tone's and the subgroup methods, with the satisfactory representation of intrapellet spatial effects.

• Journal of the Korean institute of surface engineering
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• v.52 no.4
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• pp.227-232
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• 2019

Mo-N based coatings have been studied for enhancing mechanical characteristics of thin films. In the case of Mo-X-N coatings, the microstructure and mechanical properties can be affected by the addition of the third element. In this work, Mo-Cu-N coatings were successfully fabricated with varying the Cu content from 4.5 at% to 31 at% by the co-sputtering method. Thus, properties of the coatings were analyzed by EDS, SEM, XRD, AFM, nano indentation and scratch test techniques. From observed results, MoxN bonds were made in a nitrogen atmosphere and Cu elements were present at grain boundaries. In addition, coatings with the Cu content above 14 at% had a Cu3N peak in the XRD results. Thus, it is suggested that the formation of Cu3N phase affected the microstructure and mechanical properties of Mo-Cu-N coatings. Mechanical properties of Mo-Cu-N coatings were found to be relatively better at Cu content of about 12 at%.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.25-78
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• 2005

As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming. After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows: (1) Derivation of the multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (now almost three decades old) diffusion nodal methods. (3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues. (4) Description of the analytic function expansion nodal (AFEN) method. (5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods. (6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis. (7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2788-2802
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• 2021

Multigroup cross section (MG XS) generation by the UNIST in-house Monte Carlo (MC) code MCS for fast reactor analysis using nodal diffusion codes is reported. The feasibility of the approach is quantified for two sodium fast reactors (SFRs) specified in the OECD/NEA SFR benchmark: a 1000 MW_th metal-fueled SFR (MET-1000) and a 3600 MW_th oxide-fueled SFR (MOX-3600). The accuracy of a few-group XSs generated by MCS is verified using another MC code, Serpent 2. The neutronic steady-state whole-core problem is analyzed using MCS/RAST-K with a 24-group XS set. Various core parameters of interest (core k_eff, power profiles, and reactivity feedback coefficients) are obtained using both MCS/RAST-K and MCS. A code-to-code comparison indicates excellent agreement between the nodal diffusion solution and stochastic solution; the error in the core k_eff is less than 110 pcm, the root-mean-square error of the power profiles is within 1.0%, and the error of the reactivity feedback coefficients is within three standard deviations. Furthermore, using the super-homogenization-corrected XSs improves the prediction accuracy of the control rod worth and power profiles with all rods in. Therefore, the results demonstrate that employing the MCS MG XSs for the nodal diffusion code is feasible for high-fidelity analyses of fast reactors.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3653-3664
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• 2021

In PWR three-dimensional transient coupling calculation software CORCA-K, the nodal Green's function method and diagonal implicit Runge Kutta method are used to solve the spatiotemporal neutron dynamic diffusion equation, and the single-phase closed channel model and one-dimensional cylindrical heat conduction transient model are used to calculate the coolant temperature and fuel temperature. The LMW, NEACRP and PWR MOX/UO₂ benchmarks and FangJiaShan (FJS) nuclear power plant (NPP) transient control rod move cases are used to verify the CORCA-K. The effects of burnup, fuel effective temperature and ejection rate on the control rod ejection process of PWR are analyzed. The conclusions are as follows: (1) core relative power and fuel Doppler temperature are in good agreement with the results of benchmark and ADPRES, and the deviation between with the reference results is within 3.0% in LMW and NEACRP benchmarks; 2) the variation trend of FJS NPP core transient parameters is consistent with the results of SMART and ADPRES. And the core relative power is in better agreement with the SMART when weighting coefficient is 0.7. Compared with SMART, the maximum deviation is -5.08% in the rod ejection condition and while -5.09% in the control rod complex movement condition.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1079-1087
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• 2021

NTP-ERSN is a package developed for solving the multigroup form of the discrete ordinates, characteristics and collision probability of the Boltzmann transport equation in one-dimensional cartesian geometry, by combining pin cells. In this work, C5G7 MOX benchmark is used to verify the accuracy and efficiency of NTP-ERSN package, by treating reactor core problems without spatial homogenization. This benchmark requires solutions in the form of normalized pin powers as well as the vectors and the eigenvalue. All NTP-ERSN simulations are carried out with appropriate spatial and angular approximations. A good agreement between NTP-ERSN results with those obtained with OpenMC calculation code for seven energy groups. In addition, our studies about angular and mesh refinements are carried out to produce better quality solution. Moreover, NTP-ERSN GUI has also been updated and adapted to python 3 programming language.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3085-3094
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• 2022

The design concept of a Small High-flux Multipurpose LBE(Lead Bismuth Eutectic) cooled Fast Reactor (SHMLFR) was proposed in the paper. The primary cooling system of the reactor is forced circulation, and the fuel element form is arc-plate loaded high enrichment MOX fuel. The core is cylindrical with a flux trap set in the center of the core, which can be used as an irradiation channel. According to the requirements of the core physical design, a series of physical design criteria and constraints were given, and the steady and transient parameters of the reactor were calculated and analyzed. Regarding the thermal and hydraulic phenomena of the reactor, a simplified model was used to conduct a preliminary analysis of the fuel plates at special positions, and the temperature field distribution of the fuel plate with the highest power density under different coolant flow rates was simulated. The results show that the various parameters of SHMLFR meet the requirements and design criteria of the physical design of the core and the thermal design of the reactor. This implies that the conceptual design of SHMLFR is feasible.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3930-3943
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• 2021

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper (Kawada and Suzuki, 2020) [1]. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, based on the analysis of the experimental data, the behaviors related to the stub motion were evaluated and quantified by the author from scratch. A simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.