• Nuclear Engineering and Technology
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• v.50 no.8
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• pp.1339-1348
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• 2018

The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.

• Carbon letters
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• v.4 no.1
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• pp.36-39
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• 2003

Carbon and carbon-based materials are used in nuclear reactors and there has recently been growing interest to develop graphite and carbon based materials for high temperature nuclear and fusion reactors. Efforts are underway to develop high density carbon materials as well as amorphous isotropic carbon for the application in thermal reactors. There has been research on coated nuclear fuel for high temperature reactor and research and development on coated fuels are now focused on fuel particles with high endurance during normal lifetime of the reactor. Since graphite as a moderator as well as structural material in high temperature reactors is one of the most favored choices, it is now felt to develop high density isotropic graphite with suitable coating for safe application of carbon based materials even in oxidizing or water vapor environment. Carboncarbon composite materials compared to conventional graphite materials are now being looked into as the promising materials for the fusion reactor due their ability to have high thermal conductivity and high thermal shock resistance. This paper deals with the application of carbon materials on various nuclear reactors related issues and addresses the current need for focused research on novel carbon materials for future new generation nuclear reactors.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.12 s.255
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• pp.1196-1202
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• 2006

A MEMS methanol reformer was fabricated and its performance was evaluated in the present study. Catalytic steam reforming of methanol was selected because the process had been widely applied in macro scale reformers. Conventional Cu/ZnO catalyst that was prepared by co-precipitation method to give the highest coating quality was used. The reactor structure was made by bonding three layers of glass wafers. The internal structure of the wafer was fabricated by the wet-etching process that resulted in a high aspect ratio. The internal surface of the reactor was coated by catalyst and individual wafers were fusion-bonded to form the reactor structure. The internal volume of the microfabricated reactor was $0.3cm^3$ and the reactor produced exhaust gas with hydrogen concentration at 73%. The production rate of hydrogen was 4.16 ml/hr that could generate power of 350 mW in a typical PEM fuel cell.

• Proceedings of the Korean Vacuum Society Conference
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• 2011.02a
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• pp.397-397
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• 2011

In a low aspect ratio (LAR) tokamak reactor with a superconducting toroidal field (TF) coil, the radial build of TF coil and the shield play a key role in determining the size of a reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with one-dimensional radiation transport code. Conceptual design study of a compact superconducting LAR tokamak reactor with aspect ratio less than 2.5 was conducted and the optimum radial build was identified. It is shown that the use of an improved shielding material and high temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at low aspect ratio, and that by using an inboard neutron reflector instead of breeding blanket, tritium self-sufficiency is possible with outboard blanket only and thus compact sized reactor is viable.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2460-2470
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• 2020

To study the thermophysical and neutronic properties of thorium-plutonium fuel, a conceptual design of a hybrid facility consisting of a subcritical Th-Pu reactor core and a source of additional D-D neutrons that places on the axis of the core is proposed. The source of such neutrons is a column of high-temperature plasma held in a long magnetic trap for D-D fusionreactions. This article presents computer simulation results of generation of thermonuclear neutrons in the plasma, facility neutronic properties and the evolution of a fuel nuclide composition in the reactor core. Simulations were performed for an axis-symmetric radially profiled reactor core consisting of zones with various nuclear fuel composition. Such reactor core containing a continuously operating stationary D-D neutron source with a yield intensity of Y = 2 × 10¹⁶ neutrons per second can operate as a nuclear hybrid system at its effective coefficient of neutron multiplication 0.95-0.99. Options are proposed for optimizing plasma parameters to increase the neutron yield in order to compensate the effective multiplication factor decreasing and plant power in a long operating cycle (3000-day duration). The obtained simulation results demonstrate the possibility of organizing the stable operation of the proposed hybrid 'fusion-fission' facility.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2334-2337
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• 2022

Deuterium-tritium reaction is the most promising one in term of the highest nuclear fusion cross-section for the reactor. So it is one of urgent issues to develop materials and components that are simultaneously resistant to high heat flux and high energy neutron flux in realization of the fusion energy. 2.45 MeV neutron production was reported in D-D reaction in KSTAR and regarded as beam-target is the dominant process. The feasibility study of KSTAR to wide area neutron source facility is done in term of D-D and D-T reactions from the empirical scaling law from the mixed fast and thermal stored energy and its projection to cases of heating power upgrade and DT reaction is done.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.87-92
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• 2008

Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.

• Proceedings of the KWS Conference
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• 2002.10a
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• pp.418-424
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• 2002

Korean Fusion Reactor(KSTAR) system consists of a vacuum vessel, in-vessel components, cryostat, thermal shield, super-conducting magnets and magnet supporting structures. These systems are in the final stage of engineering design with the involvement of industrial manufacturers. The overall configuration and the detailed dimensions of the KSTAR structure have been determined and the first stage of manufacturing is progressing now. In this study, the fabrication and assembly sequence were evaluated in viewpoint of high strengthening joints and very high accuracy. Especially for this purpose, the special cleaning process and welding process were proposed for high strengthening austenitic stainless steel which shall be used at cryogenic temperature. The draft procedure qualification data for welding process are presented with precise welding data including special narrow groove design. For the cooling line attachment on the surface of inside wall of magnet structure case, Induction brazing technology is introduced with some special jigging system and some consumables.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.323-327
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• 2020

In this study, the neutronic calculation to obtain tritium breeding ratio (TBR) in a deuterium-tritium (D-T) fusion power reactor using Monte Carlo MCNPX is done. In addition, by using COMSOL software, an efficient cooling system is designed. In the proposed design, it is adequate to enrich up to 40% ⁶Li. Total tritium breeding ratio of 1.12 is achieved. The temperature of helium as coolant gas never exceed 687℃. As regards the tolerable temperature of beryllium (650℃), the design of blanket module is done in the way that beryllium temperature never exceed 600℃. The main feature of this design indicates the temperature of helium coolant is higher than other proposed models for blanket module, therefore power of electricity generation will increase.

• Journal of Welding and Joining
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• v.16 no.3
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• pp.121-128
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• 1998

The degradation of mechanical properties in the high nitrogen steel HN3 developed for nuclear fusion reactor has been evaluated quantitatively using the small punch(SP) test, X-ray diffraction (XRD) analysis has also been conducted to identify carbides or nitrides precipitated on grain boundaries of the heat treated samples. Mechanical properties of the steel HN3 significantly decreased with increasing heat treatment time and temperature or with decreasing testing temperature. Combination of XRD and metallurgical observation, revealed that the material degradation in the thermally aged steel was caused by precipitation of carbides on the grain boundaries. While the weld metal showed the lowest mechanical properties among various microstructures in GTA weldments. By combining SP test and XRD analysis, cryogenic fracture behaviors and aging degradation for high nitrogen steel could be successfully evaluated in nondestructive manner.