• Nuclear Engineering and Technology
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• 제40권1호
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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.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.100-108
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• 2023

In this study, a neural network model inspired by a one-dimensional convolution U-net is developed to automatically accelerate edge localized mode (ELM) detection from big diagnostic data of fusion devices and increase the detection accuracy regardless of the hyperparameter setting. This model recognizes the input signal patterns and overcomes the problems of existing detection algorithms, such as the prominence algorithm and those of differential methods with high sensitivity for the threshold and signal intensity. To train the model, 10 sets of discharge radiation data from the KSTAR are used and sliced into 11091 inputs of length 12 ms, of which 20% are used for validation. According to the receiver operating characteristic curves, our model shows a positive prediction rate and a true prediction rate of approximately 90% each, which is comparable to the best detection performance afforded by other algorithms using their optimized hyperparameters. The accurate and automatic ELM-burst detection methodology used in our model can be beneficial for determining plasma properties, such as the ELM frequency from big data measured in multiple experiments using machines from the KSTAR device and ITER. Additionally, it is applicable to feature detection in the time-series data of other engineering fields.

• 한국초전도ㆍ저온공학회논문지
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• 제19권4호
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• pp.31-39
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• 2017

European R&D on designing their version of a DEMO fusion tokamak has recently resulted in the testing of a prototype $Nb_3Sn$ Cable-in-Conduit Conductor (CICC) for the DEMO TF coil. The characteristics and reported results of low temperature performance tests with the prototype CICC sample are compared with those from CICC samples incorporating other recent $Nb_3Sn$ cable designs. The EU-DEMO TF CICC prototype shows performance characteristics similar to that of the ITER CS CICC with short twist pitch. This is a first for a CICC sample that does not have a circular cross section. Assessment of its internal magnetostatic self-field suggests that a reduction in the internal self-field due to the rectangular geometry of the EU-DEMO TF CICC prototype compared to one with a circular geometry may have contributed to the performance characteristics showing current sharing temperature ($T_{cs}$) initially increase then stabilize with repeated electromagnetic loading, similarly to ITER CS CICC results. However, constraints on the internal self-field are not a sufficient condition for this $T_{cs}$ characteristic to occur.

• Korean Chemical Engineering Research
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• 제58권4호
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• pp.596-603
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• 2020

본 연구에서는 화학공정 최적화에 사용되는 생산계획최적기법을 도입하여 핵융합 공정에서의 삼중수소 재고량을 최소로 유지하는 수학적 모델을 구축하여 최적 운전 시나리오를 도출하였다. 핵융합 발전을 위한 공정 중 연료주기 공정(fuel cycle)은 반응연료인 중수소와 삼중수소를 저장 하고 공급하는 시스템과 핵융합 반응 배가스로부터 이를 회수 및 분리하는 세부 공정들로 구성되어 있다. 이들 공정들은 삼중수소가 방사성 물질이라는 것을 제외하면 대부분 촉매반응과 분리공정으로 이루어져 화공플랜트에 적용된 기술과 유사한 특성이 많아 화학공정에 사용되는 스케줄링 기법을 통해 최적 운전 시나리오를 도출 가능하다. 본 연구에서는 핵융합로의 다양한 장치의 특성을 반영해서, 펌프내부의 삼중수소량을 최소로 하는 최적 재생주기를 구하고, 구해진 최적 재생주기 결과를 반영하여 후단의 트리튬 플랜트에서의 최적 운전 시나리오를 확인해 보았다. 구축된 모델은 실제 토카막 시나리오에 적용되어 ITER 연료주기 내 공정의 연료흐름 및 밸런스 분석에 활용되었다.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2013년도 제44회 동계 정기학술대회 초록집
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• pp.670-670
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• 2013

The concept of a fusion-driven transmutation reactor based on LAR (Low Aspect Ratio) tokamak as a neutron source is studied based on ITER physics and technology. The radial build of transmutation reactor components are self-consistently determined by coupling the systems analysis with radiation transport analysis and an optimal configuration of a transmutation reactor for aspect ratio, A in the range of 1.5 to 2.0 is found. The performance of a transmutation reactor is investigated and shows that a transmutation reactor with a neutron source producing fusion power less than 150 MW can destroy the transuranic actinides contained in the spent fuels produced from more than two 1 GWe PWRs with production of the fission power being greater than 2 GW.

• Nuclear Engineering and Technology
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• 제41권4호
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• pp.455-476
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• 2009

The history of nuclear fusion research in Korea is rather short compared to that of advanced countries. However, since the mid-1990s, at which time the construction of KSTAR was about to commence, fusion research in Korea has been actively carried out in a wide range of areas, from basic plasma physics to fusion reactor design. The flourishing of fusion research partly owes to the fact that industrial technologies in Korea including those related to the nuclear field have been fully matured, with their quality being highly ranked in the world. Successive pivotal programs such as KSTAR and ITER have provided diverse opportunities to address new scientific and technological problems in fusion as well as to draw young researchers into related fields. The frame of the Korean nuclear fusion program is now changing from a small laboratory scale to a large national agenda. Coordinated strategies from different views and a holistic approach are necessary in order to achieve optimal efficiency and effectiveness. Upon this background, the present paper reflects upon the road taken to arrive at this point and looks ahead at the coming future in nuclear fusion research activities in Korea.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2013년도 제45회 하계 정기학술대회 초록집
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• pp.179.2-179.2
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• 2013

A large-area RF-driven ion source is being developed at Germany for the heating and current drive of ITER device. Negative hydrogen ion sources are major components of neutral beam injection (NBI) systems in future large-scale fusion experiments such as ITER and DEMO. The RF sources for the production of positive hydrogen ions have been successfully developed at IPP (Max-Planck-Institute for Plasma Physics), Garching, for the ASDEX-U and W7-AS neutral beam heating systems. Ion sources of the first NBI system (NBI-1) for the KSTAR tokamak have been developed successfully with a bucket plasma generator based on the filament arc discharge, which have contributed to achieve a good plasma performance such as 15 sec H-mode operation with an injection of 3.5 MW NB power. There is a development plan of RF ion source at the KAERI to extract the positive ions, which can be used for the second NBI system (NBI-2) of the KSTAR and to extract the negative ions for future fusion devices such as Fusion Neutron Source and Korea-DEMO. The development progresses of RF ion source at the KAERI are described in this presentation.

• 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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• 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.

• 한국진공학회:학술대회논문집
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• 한국진공학회 2014년도 제46회 동계 정기학술대회 초록집
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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.