• Nuclear Engineering and Technology
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• 제39권3호
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• pp.187-192
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• 2007

The basic research on fast reactor technology was started in the mid-1960's in China. The emphasis was put on fast reactor neutronics, thermohydraulics, sodium technology, materials, fuels, safety, sodium devices and instrumentation. In 1987, the research turned to applied basic research with the conceptual design of a 60 MW experimental fast reactor as a target. The Project of the China Experimental Fast Reactor(CEFR) with a thermal power 65 MW was launched in 1993. The R&D of fast reactor technology then carried out to serve a design demonstration connected with the different phases of the conceptual, preliminary and detailed design of the CEFR. Recently, three directions of fast rector technology R&D activities have been considered, and some research programs have been developed. They are: (1) R&D related to the CEFR, i.e. experiments to be conducted on the CEFR for its safe operation, (2) R&D related to the projects of a prototype and the demonstration of fast reactors, and(3) advanced SFR technology within the framework of the international cooperation of INPRO and GIF.

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.924-933
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• 2015

The correlation of the isotopic composition of uranium, plutonium, neodymium, and cesium with the burnup for high burnup pressurized water reactor fuels irradiated in nuclear power reactors has been experimentally investigated. The total burnup was determined by Nd-148 and the fractional $^{235}U$ burnup was determined by U and Pu mass spectrometric methods. The isotopic compositions of U, Pu, Nd, and Cs after their separation from the irradiated fuel samples were measured using thermal ionization mass spectrometry. The contents of these elements in the irradiated fuel were determined through an isotope dilution mass spectrometric method using $^{233}U$, $^{242}Pu$, $^{150}Nd$, and $^{133}Cs$ as spikes. The activity ratios of Cs isotopes in the fuel samples were determined using gamma-ray spectrometry. The content of each element and its isotopic compositions in the irradiated fuel were expressed by their correlation with the total and fractional burnup, burnup parameters, and the isotopic compositions of different elements. The results obtained from the experimental methods were compared with those calculated using the ORIGEN-S code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.1001-1007
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• 1995

A Comprehensive Vibration Assessment Program (CVAP) has been performed for Yonggwang Nuclear Power Plant Unit 4 (YGN 4) in order to verify the structural integrity of the reactor internals for flow induced vibrations prior to commercial operation. The theoretical evidence for the structural integrity of the reactor internals and the basis for measurement and inspection are provided by the analysis. Flow induced hydraulic loads and reactor internals vibration response data were measured during pre-core hot functional testing in YGN 4 site. Also, the critical areas in the reactor internals were inspected visually to check any existence of structural abnormality before and after the pre-core hot functional testing. Then, the measured data have been analyzed and compared with the predicted data by analysis. The measured stresses are less than the predicted values and the allowable limits. It is concluded that the vibration response of the reactor internals due to the flow induced vibration under normal operation is acceptable for long term operation.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1679-1686
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• 2024

Graphite material plays an important role in nuclear reactors especially the high-temperature gas-cooled reactors (HTGRs) by its outstanding comprehensive nuclear properties. The structural integrity of graphite pebble fuel elements is the first barrier to core safety under any circumstances. The correct knowledge of the stiffness coefficient of the graphite pebble fuel element inside the reactor's core is significant to ensure the valid design and inherent safety. In this research, a vertical extrusion device was set up to measure the stiffness coefficient of the graphite pebble fuel element by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. The stiffness coefficient equations of graphite pebble fuel elements at different temperatures are given (in a helium atmosphere). The result first provides the data on the high-temperature stiffness coefficient of pebbles in helium gas. The result will be helpful for the engineering safety analysis of pebble-bed nuclear reactors.

• 대한기계학회논문집B
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• 제34권5호
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• pp.553-561
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• 2010

한국원자력연구원에서는 초고온가스로를 모사할 수 있는 중형 헬륨 회로를 건설 중에 있다. 이 실험헬륨 회로에서 두 개의 전기 가열기가 헬륨 유체를 1 ~ 9 MPa 의 압력 하에서 $950^{\circ}C$ 까지 가열하게 된다. 이 실험 헬륨 회로의 설계 사양을 최적화하기 위하여, 본 연구에서는 두 개의 가열기 중 하류에 위치한 고온헬륨가열기 안의 복합열전달 현상을 전산유체역학으로 해석하였다. 해석 결과에서 헬륨 가열기 내 최대 온도는 허용 한계를 넘지 않았고, 이로써 선정된 기하구조의 열적 특성은 설계요건을 만족함이 확인되었다.

• 대한기계학회논문집A
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• 제39권1호
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• pp.117-123
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• 2015

핵연료 연소시험 도중 핵연료봉에서 발생하는 열을 효과적으로 제거하기 위해서는 핵연료의 발열량을 정확하게 계산하고 충분한 유속을 갖는 냉각수를 순환시켜야 한다. 하나로는 개방형 수조 형태로서 핵연료 연소시험을 위한 별도의 냉각수 순환 루프를 갖추고 있는데, 여기에 핵연료 조사리그를 장착하고 냉각수를 순환시킴으로써 조사중인 핵연료봉의 온도를 일정온도 이하로 유지시킨다. 특히 순환되는 냉각수의 유속이 매우 높은 상태에서 조사리그 내에 부착된 부품이나 센서들이 유체유발 진동에 의해 파손되거나 기능을 상실하는 경우 매우 큰 기회비용을 야기한다. 본 연구에서는 조사리그 부품의 건전성 사전 검토 및 고속 유동에서의 센서 동작 특성에 대한 사전검토를 위해 냉각수 모의 순환장치를 개발하였다.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.269-274
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• 2019

$^{64}Cu$ is a favorable radionuclide in nuclear medicine applications because of its unique characteristics such as three types of decay (electron capture, ${\beta}^-$ and ${\beta}^+$) and 12.7 h half-life. Production of $^{64}Cu$ by irradiation $^{nat}Cu$ and $^{nat}CuNPs$ in Tehran Research Reactor was investigated. The characteristics of copper nanoparticles were investigated with SEM, TEM and XRD analysis. The cross section of $^{63}Cu(n,{\gamma})^{64}Cu$ reaction was done with TALYS-1.8 code. The activity value of $^{64}Cu$ was calculated with theoretical approach and MCNPX-2.6 code. The results were compared with related experimental results which showed good adaptations between them.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.58-63
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• 1997

This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore defector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.617-626
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• 2007

A centrifugal atomization process for uranium fuel was developed in order to fabricate high uranium density dispersion fuel for advanced research reactors. Spherical powders of $U_3Si$ and U-Mo were successfully fabricated and dispersed in aluminum matrices. Thermal and mechanical properties of dispersion fuel meat were characterized. Irradiation tests at the research reactor HANARO confirm the excellent performance of high uranium density dispersion fuel.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.532-533
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• 2005