• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.21 no.4
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• pp.381-387
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• 2008

Neutron production rate in spherically convergent beam fusion(SCBF) device as a portable neutron source strongly depends on the ion current and the grid cathode structure. In this paper, as the process of design and analysis, Design of Experiment(DOE) based on the results by Finite Element Method-Flux Corrected Transport(FEM-FCT) method is employed to calculate the ion current. This method is very useful to find optimal design conditions in a short time. Number of rings, radius of rings, and distance between the grid cathode and center are selected as control factors. From the results in the optimized model, the higher ion current is calculated and deeper potential well is also observed.

• Proceedings of the Korean Vacuum Society Conference
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• 2014.02a
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• pp.241.1-241.1
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• 2014

The large-area high-power radio-frequency (RF) driven ion sources based on the negative hydrogen (deuterium) ion beam extraction are the major components of neutral beam injection (NBI) systems in future large-scale fusion devices such as an ITER and DEMO. Positive hydrogen (deuterium) RF ion sources were the major components of the second NBI system on ASDEX-U tokamak. A test large-area high-power RF ion source (LAHP-RaFIS) has been developed for steady-state operation at the Korea Atomic Energy Research Institute (KAERI) to extract the positive ions, which can be used for the NBI heating and current drive systems in the present fusion devices, and to extract the negative ions for negative ion-based plasma heating and for future fusion devices such as a Fusion Neutron Source and Korea-DEMO. The test RF ion source consists of a driver region, including a helical antenna and a discharge chamber, and an expansion region. RF power can be transferred at up to 10 kW with a fixed frequency of 2 MHz through an optimized RF matching system. An actively water-cooled Faraday shield is located inside the driver region of the ion source for the stable and steady-state operations of RF discharge. The characteristics and uniformities of the plasma parameter in the RF ion source were measured at the lowest area of the expansion bucket using two RF-compensated electrostatic probes along the direction of the short- and long-dimensions of the expansion region. The plasma parameters in the expansion region were characterized by the variation of loaded RF power (voltage) and filling gas pressure.

• Proceedings of the Korean Vacuum Society Conference
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• 2014.02a
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• pp.456.1-456.1
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• 2014

Transmutation characteristics of transuranics (TRU) in a transmutation reactor based on LAR (Low Aspect Ratio) tokamak as a neutron source are investigated. Optimum radial build of a transmutation reactor is found by coupled analysis of the tokamak systems and the neutron transport. The dependence of the transmutation characteristics on an aspect ratio, A in the range of 1.5 to 2.5, and on a fusion power in the range of 150 MW to 500 MW are investigated. Equilibrium fuel cycle is developed for effective transmutation and it is shown that with one unit of the transmutation reactor based on the LAR tokamak producing fusion power in the range of a few hundred MW, up to 3 PWRs (1.0 GWe capacity) can be supported with the burn-up fraction bigger than 50%.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.20 no.5
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• pp.471-477
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• 2007

SCBF(Spherically Convergent Beam Fusion) device has been studied as a neutron source. Neutron production rate is a most important factor for the application of SCBF device and is proportional to the square of the ion current[1]. It is regarded generally that some correlations between the potential well structure and the ion current exist. In this paper, the ion current and potential distribution were calculated in a variety of grid cathode geometries using FEM-FCT method. Single potential well structure was certified inside the grid cathode. The deeper the potential well became, the higher the ion current due to the high electric field near the grid cathode became.

• Proceedings of the KIEE Conference
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• 2004.07c
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• pp.1823-1825
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• 2004

Spherically convergent beam fusion device accelerate ions, which are generated between outer anode and inner grid cathode, toward the spherical center. The collision of opposite direction ions give rise to fusion reactions. Spherically convergent beam fusion device is very simple and compact, therefore the device has a potential that is applied to a portable neutron source. An experimental device consist of a 20cm-diameter spherical mesh-type anode and 7cm-diameter open spherical grid cathode and was maintained at a constant pressure of about 1333 Pa by feeding argon gas.

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1736-1746
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• 2021

The results of full-scale numerical experiments of a hybrid thorium-containing fuel cell facility operating in a close-to-critical state due to a controlled source of fusion neutrons are discussed in this work. The facility under study was a complex consisting of two blocks. The first block was based on the concept of a high-temperature gas-cooled thorium reactor core. The second block was an axially symmetrical extended plasma generator of additional neutrons that was placed in the near-axial zone of the facility blanket. The calculated models of the blanket and the plasma generator of D-T neutrons created within the work allowed for research of the neutronic parameters of the facility in stationary and pulse-periodic operation modes. This research will make it possible to construct a safe facility and investigate the properties of thorium fuel, which can be continuously used in the epithermal spectrum of the considered hybrid fusion-fission reactor.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1210-1217
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• 2023

The assessment of radiation fields in the lithium loop pipes and dump tank during the operation were performed for International Fusion Materials Irradiation Facility - DEMO-Oriented NEutron Source (IFMIF-DONES) in order to obtain the radiation dose-rate maps in the component surroundings. Variance reduction techniques such as weight window mesh (produced with the ADVANTG code) were applied to bring the statistical uncertainty down to a reasonable level. The biological dose was given in the study, and potential shielding optimization is suggested and more thoroughly evaluated. The MCNP Monte Carlo was used to simulate a gamma particle transport for radiation shielding purposes for the current Li Systems' design. In addition, the shielding efficiency was identified for the Impurity Control System components and the dump tank. The analysis reported in this paper takes into account the radiation decay source from and activated corrosion products (ACPs), which is created by d-Li interaction. As a consequence, the radiation (resulting from ACPs and Be-7) shielding calculations have been carried out for safety considerations.

• Transactions of the Korean hydrogen and new energy society
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• v.24 no.2
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• pp.121-127
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• 2013

Tritium was attracted with high energy source in neutron fusion energy systems. A number of research was performed in tritium storage materials. The Korea was raised storage and delivery systems (SDS) of international thermonuclear experimental reactor (ITER) research. However, bottles of SDS would be important because of stability. The bottles have a welding zone, this zone will be vulnerable to hydrogen embrittlement. This zone have a high thermodynamic energy and heat deterioration. Therefore bottles were studied about hydrogen embrittlement to retain stability. The heat treatment of hydrogen was carried under pressure-composition-temperature (PCT) apparatus because of checking at real time. And then, mechanical properties were evaluated by tensile test and hardness test. In results of this study, hydrogen atmosphere condition is very important by tensile test and kinetics test. The samples were evaluated, that is more weak hydrogen pressure, increasing temperature and time. This results could be useful in SDS bottle designs.

• 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.