• Nuclear Engineering and Technology
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• 제54권10호
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• pp.3717-3723
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• 2022

For the first time, two-dimensional temporal behavior of the edge localized mode (ELM) filament is measured in the edge tokamak plasma with a multi-channel electrical probe array (MCEP). MCEP, which has 16 floating probes (4 × 4), is mounted at the far scrape-off layer (SOL) region in the KSTAR. An electron temperature and an ion flux are measured by sideband method (SBM), which can achieve two-dimensional measurements with high time resolution. Furthermore, temporal evolutions of the electron temperature and the ion flux are obtained during the ELM occurrence. In the H-mode period, short spikes from ELM bursts are observed in measured plasma parameters, and the trend is similar to that of typical Hα signal. Interestingly, when blob-like ELM filaments crash the probe, the heat flux is significantly higher in a local region of the probe array. The results show that our probe array using the SBM can measure the ELM behavior and the plasma parameters without the effect of the stray current caused by the huge device. This study can provide valuable data needed to understand the interaction between the SOL plasma and the plasma facing components (PFCs).

• 분석과학
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• 제23권6호
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• pp.511-517
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• 2010

황산 제조 및 그 외에 많은 유기화합물을 산화시키는 촉매제로 널리 이용되고 있는 공업용 오산화바나듐의 용해방법을 연구하였다. 오산화바나듐 시약은 왕수-$H_2O_2$-HF로 완전히 용해되었으나 판상 모양의 오산화바나듐 시료는 혼합산에 완전히 용해되지 않아 시료 전처리 방법을 확립할 필요가 있다. 따라서 본 연구에서는 산 처리법과 용융법을 혼용한 방법 및 마이크로파 용해 장치를 이용한 용해법을 비교 분석하여 바나듐화합물의 용해성을 조사하였다. 바나듐 화합물은 왕수, 플루오르화수소산 및 과산화수소의 혼합산으로 녹이는 것이 최적이었으며, 두 용해 방법 중 마이크로파 용해 장치를 이용한 용해법이 분석의 신속성을 고려하면 보다 유용한 방법으로 판단되었다. 마이크로파 용해 장치로 용해한 다음 유도 결합 플라스마 원자방출분광법으로 측정한 결과 바나듐산화물($V_2O_5$) 함량은 $97.9{\pm}0.9%$이었다.

• 기술혁신연구
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• 제22권4호
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• pp.145-164
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• 2014

미래의 궁극에너지로 인식되고 있는 핵융합에너지 개발을 위해서는 DEMO라는 최종 실증 단계를 거쳐야만 한다. 특히 중국, EU, 일본 등의 주요 국가는 DEMO 건설에 대한 구체적 계획을 수립하고 이를 실행 중에 있다. 한국도 1995년부터 KSTAR 사업을 시작으로 핵융합 연구개발에 착수한 점을 감안하면, 핵융합에너지 상용화라는 최종 목표달성 뿐만 아니라, 주요 국가와 DEMO 경쟁 상황에서 주도권을 확보하기 위한 본격적 연구개발이 필요하다. 이에 본 논문에서는 DEMO 개발을 위한 핵심기술을 파악하기 위하여 준정량적 방법론을 적용, 해당 분야의 핵심기술을 도출함으로써 우선적으로 연구개발이 필요한 기술을 식별하여 향후 연구개발 추진시 기술별 우선순위를 제안하고자 한다. 이를 위한 핵융합 에너지 개발과 관련하여 핵융합의 과학적 원리, 주요국가의 DEMO 개발 동향 등을 파악한다. 다음으로 핵융합 실증로와 관련된 기술분류 체계를 검토하여 분석할 기술분류 체계를 선정한다. 선정된 기술체계에 준정량적 방법론으로 기술수준(TRL)을 파악하고 이를 보완하기 위하여 분석적 계층화 과정(AHP)을 적용한다. TRL과 AHP의 결과를 종합하여 우선적으로 확보해야 할 핵융합 실증로의 핵심기술은 실증로용 연소 플라즈마 기술, 대면재료기술, 구조재기술, 고주파 가열장치 기술, 중성입자빔 장치기술, 안전기술, 연소플라즈마 진단장치기술, 핵융합로 시뮬레이터기술 등으로 나타났다.

• 한국수소및신에너지학회논문집
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• 제22권3호
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• pp.372-379
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• 2011

A nuclear fusion fuel cycle plant is composed of various subsystems such as a hydrogen isotope storage and delivery system, a tokamak exhaust processing system, and a hydrogen isotope separation system. Korea shares in the construction of its ITER fuel cycle plant with the EU, Japan, and the US, and is responsible for the development and supply of the storage and delivery system. The authors thus present details on the development status of hydrogen isotope storage technologies for nuclear fusion fuel cycle plants. We have developed various hydride beds of different size. We have realized a hydrogen delivery rate of 12.5 $Pam^3/s$ with a typical 1242g-ZrCo bed.

• 한국수소및신에너지학회논문집
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• 제27권2호
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• pp.163-168
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• 2016

Korea has been developing nuclear fusion fuel storage and delivery system (SDS) technologies including a basic scientific study on hydrogen storage. To develop nuclear fusion technology, it is necessary to store and supply hydrogen isotopes needed for Tokamak operation. SDS is used for storing hydrogen isotopes as a metal hydride form. The rapid hydriding of tritium is very important not only for safety reasons but also for the economic design and operation of the SDS. In this study, we designed and fabricated a medium-scale getter bed of depleted uranium (DU). The hydriding of DU has been measured by varying the initial temperature ($100-300^{\circ}C$) of the DU getter bed to investigate the influence of the cooling temperature. Furthermore, we analyzed the effect of a helium blanket on the hydriding performance with 0 - 12% helium content in hydrogen.

• 한국세라믹학회지
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• 제51권5호
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

• 한국분말재료학회지
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• 제21권4호
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• pp.271-276
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• 2014

15Cr-1Mo base oxide dispersion strengthened (ODS) steel which is considered to be as a promising candidate for high- temperature components in nuclear fusion and fission systems because of its excellent high temperature strength, corrosion and radiation resistance was fabricated by using mechanical alloying, hot isostatic pressing and hot rolling. Torsion tests were performed at room temperature, leading to two different shear strain routes in the forward and reverse directions. In this study, microstructure evolution of the ODS steel during simple shearing was investigated. Fine grained microstructure and a cell structure of dislocation with low angle boundaries were characterized with shear strain in the shear deformed region by electron backscattered diffraction (EBSD). Grain refinement with shear strain resulted in an increase in hardness. After the forward-reverse torsion, the hardness value was measured to be higher than that of the forward torsion only with an identical shear strain amount, suggesting that new dislocation cell structures inside the grain were generated, thus resulting in a larger strengthening of the steel.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.311-322
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• 2013

High-Cr martensitic steel HT-9 is one of the candidate materials for advanced nuclear energy systems. Thanks to its excellent thermal conductivity and irradiation resistance, ferritic/martensitic steels such as HT-9 are considered for in-core applications of advanced nuclear reactors. The harsh neutron irradiation environments at the reactor core region pose a unique challenge for structural and cladding materials. Microstructural and microchemical changes resulting from displacement damage are anticipated for structural materials after prolonged neutron exposure. Consequently, various irradiation effects on the service performance of in-core materials need to be understood. In this work, the fundamentals of radiation damage and irradiation effects of the HT-9 martensitic steel are reviewed. The objective of this paper is to provide a background introduction of displacement damage, microstructural evolution, and subsequent effects on mechanical properties of the HT-9 martensitic steel under neutron irradiations. Mechanical test results of the irradiated HT-9 steel obtained from previous fast reactor and fusion programs are summarized along with the information of irradiated microstructure. This review can serve as a starting point for additional investigations on the in-core applications of ferritic/martensitic steels in advanced nuclear reactors.

• 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.

• Nuclear Engineering and Technology
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• 제55권3호
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• pp.981-988
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• 2023

A seismic event caused an accident at the Fukushima Nuclear Power Plant, which further resulted in simultaneous accidents at several units. Consequently, this incident has aroused great interest in the safety of nuclear power plants worldwide. A reasonable safety evaluation of such an external event should appropriately consider the correlation between SSCs (structures, systems, and components) and the probability of failure. However, a probabilistic safety assessment in current nuclear industries is performed conservatively, assuming that the failure correlation between SSCs is independent or completely dependent. This is an extreme assumption; a reasonable risk can be calculated, or risk-based decision-making can be conducted only when the appropriate failure correlation between SSCs is considered. Thus, this study analyzed the effect of the failure correlation of SSCs on the safety of the system to realize rational safety assessment and decision-making. Consequently, the impact on the system differs according to the size of the failure probability of the SSCs and the AND and OR conditions.