• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.715-716
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• 2004

• Proceedings of the Korean Institute of IIIuminating and Electrical Installation Engineers Conference
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• 2005.11a
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• pp.151-155
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• 2005

Though the nuclear fusion system has been fused into hydro-nuclear based on thermodynamics by tokamak system, there has been no success story. Because it's impossible to confine high temperatured plasma long Time actually. New nuclear-fusion-system using this nuclear-fusion-method will gather toroidal-magnetic-field by putting Core Block(C shaped torus iron) and toroidal-aluminium coil into toroidal magnetic-field-aluminium. That will arrange the nuclear-fusion-route on a gap fallen out by a part of cut torus-core and mkee the toroidal-an electric-current flow and electrolyze the fusioned-material (an electrolyte) into troidal-electrocity. That consists of troidal-magnetic-fild coil, toroidal-coial fusioned- material, series circuit. So toroidal-electocity will be changed into filament-electrocity and be introjected into fusioned-material. In a sapce on filament-electrocity, the magnetic inhaling-powr will enlarge to input-electrocity outside. This will exceed the Coulomb force and reache the nuclear-fusion. By this phenomenon there be quantity-loss. By this process I could confirmed that Einstein euation$(E=mC^2)$ releases into thermal energy.

• Journal of the Korean Vacuum Society
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• v.19 no.1
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• pp.72-80
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• 2010

It is very important to predict and analyze the change of voltage and current distribution of current strap, abnormal voltage distribution of transmission line and resonance phenomenon by coupling between current straps for more stable operation of ICRF system. In this study, to understand those phenomena by coupling, 8-port coupled transmission line model is completed by appling S-parameter measured in the prototype KSTAR ICRF antenna to the model. The determined self-inductance, mutual-inductance and capacitance of antenna straps are shown to be lower than that calculated from 2D approximate model due to finite length of strap. The coupled transmission line model of current strap will be utilized to the operation of ICRF system of KSTAR in the future.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1346-1353
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• 2017

The FISPACT-II inventory simulation platform is a modern computational tool with advanced and unique capabilities. It is sufficiently flexible and efficient to make it an ideal basis around which to perform extensive simulation studies to scope a variety of responses of many materials (elements) to several different neutron irradiation scenarios. This paper briefly presents the typical outputs from these scoping studies, which have been used to compile a suite of nuclear physics materials handbooks, providing a useful and vital resource for material selection and design studies. Several different global responses are extracted from these reports, allowing for comparisons between materials and between different irradiation conditions. A new graphical output format has been developed for the FISPACT-II platform to display these "global summaries"; results for different elements are shown in a periodic table layout, allowing side-by-side comparisons. Several examples of such plots are presented and discussed.

• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.4274-4281
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• 2023

Blanket is of vital importance for engineering application of the fusion reactor. Nuclear heat deposition in materials is the main heat source in blanket structure. In this paper, the three-dimensional method for thermal-hydraulics/neutronics coupling analysis is developed and applied for the full-scale module of the helium-cooled ceramic breeder tritium breeding blanket (HCCB TBB) designed for China Fusion Engineering Test Reactor (CFETR). The explicit coupling scheme is used to support data transfer for coupling analysis based on cell-to-cell mapping method. The coupling algorithm is realized by the user-defined function compiled in Fluent. The three-dimensional model is established, and then the coupling analysis is performed using the paralleled Coupling Analysis of Thermal-hydraulics and Neutronics Interface Code (CATNIC). The results reveal the relatively small influence of the coupling analysis compared to the traditional method using the radial fitting function of internal heat source. However, the coupling analysis method is quite important considering the nonuniform distribution of the neutron wall loading (NWL) along the poloidal direction. Finally, the structure optimization of the blanket is carried out using the coupling method to satisfy the thermal requirement of all materials. The nonlinear effect between thermal-hydraulics and neutronics is found during the blanket structure optimization, and the tritium production performance is slightly reduced after optimization. Such an adverse effect should be thoroughly evaluated in the future work.

• Journal of the Korean Society of Safety
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• v.27 no.6
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• pp.14-19
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• 2012

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied $1.5S_m$ or $3S_m$ of the allowable stress($S_m$) in RCC-MR code at the maximum stress region in the FW.

• Nuclear Engineering and Technology
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• v.44 no.6
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• pp.673-682
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• 2012

Three dimensional atom probe tomography (3D APT) is applied to characterize the dissimilar metal joint which was welded between the Ni-based alloy, Alloy 690 and the low alloy steel, A533 Gr. B, with Alloy 152 filler metal. While there is some difficulty in preparing the specimen for the analysis, the 3D APT has a truly quantitative analytical capability to characterize nanometer scale particles in metallic materials, thus its application to the microstructural analysis in multi-component metallic materials provides critical information on the mechanism of nanoscale microstructural evolution. In this study, the procedure for 3D APT specimen preparation was established, and those for dissimilar metal weld interface were prepared near the fusion boundary by a focused ion beam. The result of the analysis in this study showed the precipitation of chromium carbides near the fusion boundary between A533 Gr. B and Alloy 152.

• Journal of Radiation Protection and Research
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• v.37 no.2
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• pp.63-69
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• 2012

The second phase of the national program for fusion energy development in Korea starts from 2012 for design and construction of the fusion DEMO reactor. Radiological assessment for the fusion reactor is one of the key tasks to assure its licensability and the starting point of the assessment is determination of the source terms. As the first effort, the activities of the coolant due to activated corrosion product (ACP) were estimated. Data and experiences from fission reactors were used, in part, in the calculations of the ACP concentrations because of lack of operating experience for fusion reactors. The MCNPX code was used to determine neutron spectra and intensities at the coolant locations and the FISPACT code was used to estimate the ACP activities in the coolant of the fusion DEMO reactor. The calculated specific activities of the most nuclides in the fusion DEMO reactor coolant were 2-15 times lower than those in the PWR coolant, but the specific activities of $^{57}Co$ and $^{57}Ni$ were expected to be much higher than in the PWR coolant. The preliminary results of this study can be used to figure out the approximate radiological conditions and to establish a tentative set of radiological design criteria for the systems carrying coolant in the design phase of the fusion DEMO reactor.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1100-1101
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• 2005

• Analytical Science and Technology
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• v.16 no.4
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• pp.261-268
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• 2003

To investigate the matrix effect of sample for hydrogen analysis by inert gas fusion-thermal conductivity detection, calibration factor for the hydrogen analyser of the inert gas fusion-thermal conductivity detection method was measured with hydrogen standard materials in Ti, Zr-2.5Nb and by hydrogen gas dosing method. Also the hydrogen extraction efficiency for the different sample matrix, Ti and Zr-2.5Nb, was evaluated without adding tin flux. The calibration factor of the hydrogen analyser which was calibrated by hydrogen standard material in Zr-2.5Nb and Ti was 2~3% and 14% higher than that by hydrogen gas dosing method, respectively. Based on the results of calibration factor measurement, it could be concluded that the hydrogen extraction efficiency of the Ti matrix sample will be 12% lower than that of the Zr-2.5Nb. And according to the experimental results of hydrogen recovery test by no tin flux, the hydrogen recovery percentage of the Ti and Zr-2.5Nb matrix sample was about 70% but the recovery rate of Ti matrix sample was slightly lower than that of Zr-2.5Nb.