• Nuclear Engineering and Technology
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• 제37권5호
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• pp.401-422
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• 2005

We review here research and development progress achieved in US Plasma Chamber technology roughly over the last decade. In particular, we focus on two major programs carried out in the US: the APEX project (1998-2003) and the US ITER TBM activities (2003-present). The APEX project grew out of the US fusion program emphasis in the late 1990s on more fundamental science and innovation. APEX was commissioned to investigate novel technology concepts for achieving high power density and high temperature reactor coolants. In particular, the idea of liquid walls and the related research is described here, with some detailed examples of liquid metal and molten salt magnetohydrodynamic and free surface effects on flow control and heat transfer. The ongoing US ITER Test Blanket Module (TBM) program is also described, where the current first wall/blanket concepts being considered are the dual coolant lead lithium concept and the solid breeder helium cooled concepts, both using ferritic steel structures. The research described for these concepts includes both thermofluid MHD issues for the liquid metal coolant in the DCLL, and thermomechanical issues for ceramic breeder packed pebble beds in the solid breeder concept. Finally, future directions for ongoing research in these areas are described.

• 한국수소및신에너지학회논문집
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• 제26권2호
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• pp.120-126
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• 2015

Recently, the ever-increasing use of fossil fuels for rapid industrial development and population significantly caused an environment pollution and global warming such as climate change. So research and development of sustainable and eco-friendly energy have been performed. Especially the interest in nuclear fusion fuel was significantly increased from the developed countries. The system of fusion energy production was tritium separation, storage and delivery, and purification. Republic of Korea is in charge of Storage and Delivery System (SDS) in the International Thermonuclear Experimental Reactor (ITER). Welding part of the SDS bottles for storing the tritium is known to be susceptible to hydrogen embrittlement. In this study, conducted a study for the relaxation of the stability and hydrogen embrittlement of the weld area. The hydrogen heat treatment was processed through the Pressure-Composition-Temperature (PCT) device according to the time variation. Also mechanical properties such as impact test and hardness test according to using the alkaline cleaning liquid for hydrogen embrittlement relief and the fracture was observed by scanning electron microscopy (SEM) after the mechanical properties evaluation.

• 한국수소및신에너지학회논문집
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• 제24권2호
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• pp.128-135
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• 2013

The nuclear fuel cycle plant is composed of various subsystems such as a fuel storage and delivery system (SDS), a tokamak exhaust processing system, a hydrogen isotope separation system, and a tritium plant analytical system. Korea is sharing in the construction of the International Thermonuclear Experimental Reactor (ITER) fuel cycle plant with the EU, Japan, and the US, and is responsible for the development and supply of the SDS. Hydrogen isotopes are the main fuel for nuclear fusion reactors. Metal hydrides offer a safe and convenient method for hydrogen isotope storage. The storage of hydrogen isotopes is carried out by absorption and desorption in a metal hydride bed. These reactions require heat removal and supply respectively. Accordingly, the rapid storage and delivery of hydrogen isotopes are enabled by a rapid cooling and heating of the metal hydride bed. In this study, we designed and manufactured a vertical-type hydrogen isotope storage bed, which is used to enhance the cooling performance. We present the experimental details of the cooling performances of the bed using various cooling parameters. We also present the modeling results to estimate the heat transport phenomena. We compared the cooling performance of the bed by testing different cooling modes, such as an isolation mode, a natural convection mode, and an outer jacket helium circulation mode. We found that helium circulation mode is the most effective which was confirmed in our model calculations. Thus we can expect a more efficient bed design by employing a forced helium circulation method for new beds.

• 한국세라믹학회지
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• 제50권5호
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• pp.333-340
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• 2013

To fabricate spherical lithium titanate ($Li_2TiO_3$) pebbles which are used for a breeder material in fusion reactor, titanium oxide ($TiO_2$) granules were used as a starting material. The granules were pre-sintered, and then aqueous lithium nitrate solution infiltrated into the granules at vacuum condition. The granules were crystallized to $Li_2TiO_3$ after sintering under the control of process parameters. In this study, the concentration of lithium in the solution, as well as the number of penetration times and sintering temperature affected the final crystallite phase and the microstructure of the pebbles. In particular, the sphericity and size of the pebbles were effectively controlled by a technical rolling process. The useful spherical $Li_2TiO_3$ pebbles which have 10~20% porosity and 60~120 N compressive strength were obtained through the sintering at $1000{\sim}1100^{\circ}C$ in the multi-times infiltration process with 50 wt% solution. The physical properties of pebbles such as density, porosity and strength, can be controlled by a selection of $TiO_2$ powders and control of processing parameters. It can be thought that the lithium penetration method is a useful method for the fabrication of mass product of spherical $Li_2TiO_3$ pebbles.

• 한국해양공학회:학술대회논문집
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• 한국해양공학회 2003년도 춘계학술대회 논문집
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• pp.300-305
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• 2003

Reduced activation ferritic (JLF-1) steel is leading candidates for blanket/first-wall structures of the D-T fusion reactor. In fusion application, structural materials will suffer effects of repeated changes of temperature. Therefore, the data base of tensile strength and fracture toughness at operated temperature $400^{\circ}C$ are very important. Fracture toughness ($J_{IC}$) and tensile tests were carried out at room temperature and elevated temperature ($400^{\circ}C$). Fracture toughness tests were performed with two type size to investigate the relationship between the constraint effect of a size and the fracture toughness resistance curve. As the results, the tensile strength and the fracture toughness values of the JLF-1 steel are slightly decreased with increasing temperature. The fracture resistance curve increased with increasing plane size and decreased with increasing thickness. The fracture toughness values of JLF-1 steel at room temperature and at $400^{\circ}C$ shows an excellent fracture toughness ($J_{IC}$) of about $530kJ/m^2\;and\;340kJ/m^2$, respectively.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.939-946
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• 2023

Comprehensive Research Facility for Fusion Technology (CRAFT) is an integration of different demonstrating or testing facilities, which aim to develop the critical technology or composition system towards the fusion reactor. Due to the importance and challenge of the negative ion based neutral beam injection (NNBI), a NNBI test facility is included in the framework of CRAFT. The initial object of CRAFT NNBI test facility is to obtain a H⁰ beam power of 2 MW at the energy of 200-400 keV for the pulse duration of 100 s. Inside the negative ion accelerator of NNBI system, the interactions of the negative ions with the background gas and electrodes can generate abundant stray electrons. The stray electrons can be further accelerated and dumped on the electrodes or eject from the accelerator. The stray electrons, including the ejecting electrons, cause the unwanted particle and heat flux onto the electrodes and the inner components of beamline (especially the temperature sensitive cryopump). The suppression of the stray electrons from the CRAFT accelerator is carried out via a series of design and simulation works. The paper focuses the influence of different magnetic field configurations on the stray electrons and the character of the ejecting electrons.

• 대한용접접합학회:학술대회논문집
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• 대한용접접합학회 2010년도 춘계학술발표대회 초록집
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• pp.43-46
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• 2010

As the nuclear power plant has been constructed continuously for several decades in Korea, the welding technology for components manufacturing and installation has been improved largely. Standardization for weld test and qualification was also established systematically according to the concerned code. The welding for the main components requires the high reliability to keep the constant quality level, which means the repeatability of weld quality. Therefore the weld process qualified by thorough test and evaluation is able to be applied for manufacturing. Narrow gap SAW and GTAW process are usually applied for girth seam welding of pressure vessel like Reactor vessel, steam generator, and etc. For the surface cladding with stainless steel and Inconel material, strip welding process is mainly used. Inside cladding of nozzles is additionally applied with Hot wire GTAW and semi-auto welding process. Especially the weld joint having elliptical weld line on curved surface needs a specialized weld system which is automatically rotating with adjusting position of the head torch. The small sized pipe, tube, and internal parts of reactor vessel requests precise weld processes like an automatic GTAW and electron beam welding. Welding of dissimilar materials including Inconel690 material has high possibility of weld defects like a lack of fusion, various types of crack. To avoid these kinds of problem, optimum weld parameters and sequence should be set up through the many tests. As the life extension of nuclear power plant is general trend, weld technologies having higher reliability is required gradually. More development of specialized welding systems, weld part analysis and evaluation, and life prediction for main components should be taken into a consideration extensively.

• Korean Chemical Engineering Research
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• 제56권2호
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• pp.169-175
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• 2018

본 연구는 핵융합 배가스 중 삼중수소가 포함된 화합물인 메탄($CQ_4$) 및 물($Q_2O$)로부터 수소동위원소를 회수하기 위한 공정에 관한 것이다(Q는 수소, 중수소, 삼중수소). 수증기-메탄 개질반응과 수성가스 전환반응을 이용하여 $CQ_4$와 $Q_2O$를 $Q_2$로 변환시키고, 후속하는 팔라듐 분리막으로 생성된 $Q_2$를 회수한다. 본 연구에서는 $CQ_4$ 및 $Q_2O$ 중 하나의 물질인 $CH_4$ 및 $H_2O$로부터 수소 회수를 위해 촉매반응기, 팔라듐 분리막, 순환펌프로 구성된 순환루프를 적용하였다. 촉매반응온도 및 순환유량을 변화시켜가며 $CH_4$ 및 $H_2O$의 전환율을 측정하였다. $CH_4$ 중 수소 회수는 촉매반응온도 $650^{\circ}C$, 순환유량 2.0 L/min 조건에서 99% 이상의 $CH_4$ 전환율을확인하였고, $H_2O$ 중수소 회수는촉매반응온도 $375^{\circ}C$, 순환유량 1.8 L/min 조건에서 96% 이상의 $H_2O$ 전환율을 확인하였다. 이와 더불어, 향후 핵융합 실증로(K-DEMO)에서의 $CQ_4$ 발생량을 예측하고, 이에 대한 처리공정을 제안하였으며, HAZOP (Hazard and Operability) 분석을 실시하여 공정의 위험요소와 운전상의 문제점을 도출하고 해결방안을 제시하였다.

• 청정기술
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• 제23권3호
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• pp.331-342
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• 2017

하수 슬러지 고형연료 및 우드 펠렛의 연소 특성을 평가 하기 위하여 열중량 분석(TGA), 회 융점(AFT) 분석, 그리고 회분 성분 분석을 수행하였다. TGA 분석 결과, 하수 슬러지 고형연료의 연소성이 우드 펠렛에 비해 상대적으로 좋지 않았다. 또한 AFT 분석을 통해 하수 슬러지 고형연료의 슬래깅 가능성이 매우 높은 것을 확인하였다. 또한 연소성 평가를 위해 pilot-scale 기포 유동층 반응기를 적용하였으며, 장치는 예열기, 유동층 반응기, 연료 공급장치, 사이클론, 회분 포집 장치, 그리고 가스분석기로 구성되었다. 반응기는 직경 400 mm, 높이 4300 mm이며, 하수 슬러지는 $54.5{\sim}96.5kW_{th}$의 열량으로 실험을 수행하였고 우드 펠렛은 $96.1kW_{th}$ 실험을 수행하였다. 실험 결과, 하수 슬러지 고형연료 연소의 경우 평균적으로 우드 펠렛의 연소 보다 배기가스 중 $NO_x$는 10.1배, CO는 3.5배 높았다. 또한 사이클론에서 포집한 회분을 분석한 결과, 모든 실험 조건에서 연소 효율은 99% 이상이었고, 회분의 성분 분석을 통해 슬래깅/파울링 가능성이 높은 것을 확인하였다.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2139-2146
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• 2023

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 ℃ and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ~268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m² for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.