• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.715-716
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• 2004

• 한국조명전기설비학회:학술대회논문집
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• 한국조명전기설비학회 2005년도 학술대회 논문집
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• pp.151-155
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• 2005

지금까지의 핵융합 장치는 자기우물과 열역학을 바탕으로 한 토카막 장치가 수소핵융합을 하였으나 고온의 플라스마를 장시간 동안 가둠이란 현실적으로 불가능하여 성공사례가 없었다. 본 핵융합 촉진 방법을 이용한 새로운 핵융합 장치는 토카막 장치의 토로이달 자기장용 알루미늄코일 내에 토러스 철심 C형 Core Block과 토로이달 알루미늄코일을 넣어 토로이달 자기장을 모으고 토로이달 전류 흐르도록 하여 토러스 Core의 일부가 절개된 간극사이에 핵융합로를 배치하고 융합원료(전해액)를 토로이달 전류로 전기분해 한다. 전원과 토로이달 자계용 코일, 토로이달 코일 융합원료로 직렬회로로 이루어져 있어 토로이달 전류는 필라멘트 전류로 되어 융합원료에 투입된다. 필라멘트 전류사이에는 자기흡입력이 외부의 입력전력으로 계속 증대되어 쿨롱에 힘을 넘어 핵융합에 이르고 그로인해 질량결손이 생겨 아인슈타인의 질량에너지$(E=mC^2)$가 빛에너지와 열에너지로 방출됨을 확인했다.

• 한국진공학회지
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• 제19권1호
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• pp.72-80
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• 2010

KSTAR ICRF 시스템에서 안테나 전류띠 간 커플링에 의한 전류띠의 전압, 전류 분포 변화와 전류띠 간 전력 전달에 의한 공명현상 그리고 전송선상의 이상 전압 분포 등을 예측하거나 분석하는 것은 그것의 안정적이고 신뢰성 있는 운전을 위해 매우 중요하다. 본 연구에서는 이러한 전류띠 간 커플링에 의한 현상들을 이해하기 위해 ICRF 안테나에서 측정한 S-parameter를 커플링이 고려된 8포트 전송선회로 모델에 적용하여 전류띠의 전송선 회로 모델을 완성하였다. 완성된 전송선 회로모델의 자체유도계수, 상호유도계수, 전기용량성 등은 전류띠의 유한한 길이로 인하여 2D 모델의 값보다 작은 것으로 나타났다. 커플링이 고려된 전류띠의 전송선 회로모델은 공명루프와 결합되어 있는 KSTAR ICRF 시스템의 운전에 활용될 것이다.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1346-1353
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• 2017

The FISPACT-II inventory simulation platform is a modern computational tool with advanced and unique capabilities. It is sufficiently flexible and efficient to make it an ideal basis around which to perform extensive simulation studies to scope a variety of responses of many materials (elements) to several different neutron irradiation scenarios. This paper briefly presents the typical outputs from these scoping studies, which have been used to compile a suite of nuclear physics materials handbooks, providing a useful and vital resource for material selection and design studies. Several different global responses are extracted from these reports, allowing for comparisons between materials and between different irradiation conditions. A new graphical output format has been developed for the FISPACT-II platform to display these "global summaries"; results for different elements are shown in a periodic table layout, allowing side-by-side comparisons. Several examples of such plots are presented and discussed.

• Nuclear Engineering and Technology
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• 제55권11호
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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.

• 한국안전학회지
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• 제27권6호
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• pp.14-19
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• 2012

The Helium cooled ceramic reflector(HCCR) test blanket module(TBM) has been designed and developed to participate the ITER(International Thermonuclear Experimental Reactor) test blanket program in Korea. The TBM was one of the main objectives for developing ITER for proving the tritium self-sufficiency and the heat transfers to produce the electricity with the breeding blanket concept. Among the TBM components, the first wall(FW) was the most important component in safety since it was directly faced a high level of a heat and fast neutrons from the plasma side and could protect the others components inside TBM. In this paper, the FW has been designed through the thermo-mechanical analysis considering ITER operation conditions. With the developed simple models, the stress limit analysis based on RCC-MR code which is the nuclear power plant design codes in France was evaluated for the allowable design criteria. The results showed that the designed FW model satisfied $1.5S_m$ or $3S_m$ of the allowable stress($S_m$) in RCC-MR code at the maximum stress region in the FW.

• Nuclear Engineering and Technology
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• 제44권6호
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• pp.673-682
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• 2012

Three dimensional atom probe tomography (3D APT) is applied to characterize the dissimilar metal joint which was welded between the Ni-based alloy, Alloy 690 and the low alloy steel, A533 Gr. B, with Alloy 152 filler metal. While there is some difficulty in preparing the specimen for the analysis, the 3D APT has a truly quantitative analytical capability to characterize nanometer scale particles in metallic materials, thus its application to the microstructural analysis in multi-component metallic materials provides critical information on the mechanism of nanoscale microstructural evolution. In this study, the procedure for 3D APT specimen preparation was established, and those for dissimilar metal weld interface were prepared near the fusion boundary by a focused ion beam. The result of the analysis in this study showed the precipitation of chromium carbides near the fusion boundary between A533 Gr. B and Alloy 152.

• Journal of Radiation Protection and Research
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• 제37권2호
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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.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.1100-1101
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• 2005

• 분석과학
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• 제16권4호
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• pp.261-268
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• 2003

불활성기체용해-열전도도검출법에 의한 수소분석시 매질효과를 조사하기 위해 티타늄 및 지르코늄-2.5니오븀 매질의 수소 표준물질 및 수소기체 주입에 의한 교정계수를 측정하였다. 또한 주석 조연제를 사용하지 않고 티타늄 및 지르코늄-2.5니오븀 매질 수소 표준물질의 수소 추출효율을 조사하였다. 수소기체 주입에 의한 수소분석기의 보정에 대해 지르코늄-2.5니오븀 매질 수소표준물질의 그것은 2~3% 높았으며, 티타늄 매질의 수소 표준물질은 약 14% 높은 값을 주었다. 교정계수 측정결과에 의하면 티타늄 매질 시료의 수소추출 효율이 지르코늄-2.5니오븀 매질 시료에 비해 약 12% 낮을 것으로 예상된다. 주석을 사용하지 않았을 때 티타늄 및 지르코늄-2.5니오븀매질 수소 표준물질의 수소 회수율은 약 70% 이었으며, 티타늄의 수소 회수율이 지르코늄-2.5니오븀 보다 낮았다.