• Nuclear Engineering and Technology
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• 제3권4호
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• pp.185-197
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• 1971

1950년대의 미국 General Atomic사에서 열출력 100 kw로 설계, 제작하여 1962년 3월에 건조완료한 TRIGA Mark-II원자로는 1969년 7월에 250 kw로 출력 증강되었으나 방사선차폐는 보강되지 않았다. 본 논문에서의 계산에 의하면 출력 증강후 현재의 차폐물로도 중성자에 대하여는 확실히 안전하지만 Gamma선에 대해서는 위험하다는 것이 판명되었다. 원자로의 구조와 출입인 및 실험종사자들의 위치로 보아 차폐물의 안전도 검토는 수평방향에 한하였고, 또 정확을 기하기 위하여 중성자와 Gamma선의 투과문제를 나누어 검토하였다. 이를 근거로 하여 이론적인 측면에서 본 콘크리트의 보강을 요하는 두께도 산출하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.1996-2000
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• 2008

In autothermal reforming reaction, oxygen to carbon ratio (OCR) and steam to carbon ratio (SCR) are significant factors, which control temperature and carbon deposition into the reactor. The OCR is more sensitive than the SCR to affect the temperature distribution and reforming efficiency. In conventional operation, hydrocarbon fuel, steam, and oxygen was homogeneously mixed and injected into the reactor in order to get hydrogen-rich gas. The temperature was abruptly raised due to fast oxidation reaction in the former part of the reactor. Deactivation of packed catalysts can be accelerated there. In the present study, therefore, the effect of the oxygen distribution is introduced and investigated to suppress the carbon deposition and to maintain the reactor in the mild operating temperature (e.g., $700{\sim}800^{\circ}C$). In order to investigate the effect numerically, the following models are adopted; heterogeneous reaction model and two-medium model for heat balance.

• Nuclear Engineering and Technology
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• 제51권4호
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• pp.968-976
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• 2019

This paper presents a number of verification case studies for a recently developed sensitivity/uncertainty code package. The code package, ROMUSE (Reduced Order Modeling based Uncertainty/Sensitivity Estimator) is an effort to provide an analysis tool to be used in conjunction with reactor core simulators, in particular the Virtual Environment for Reactor Applications (VERA) core simulator. ROMUSE has been written in C++ and is currently capable of performing various types of parameter perturbations and associated sensitivity analysis, uncertainty quantification, surrogate model construction and subspace analysis. The current version 2.0 has the capability to interface with the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) code, which gives ROMUSE access to the various algorithms implemented within DAKOTA, most importantly model calibration. The verification study is performed via two basic problems and two reactor physics models. The first problem is used to verify the ROMUSE single physics gradient-based range finding algorithm capability using an abstract quadratic model. The second problem is the Brusselator problem, which is a coupled problem representative of multi-physics problems. This problem is used to test the capability of constructing surrogates via ROMUSE-DAKOTA. Finally, light water reactor pin cell and sodium-cooled fast reactor fuel assembly problems are simulated via SCALE 6.1 to test ROMUSE capability for uncertainty quantification and sensitivity analysis purposes.

• Nuclear Engineering and Technology
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• 제25권3호
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• pp.457-469
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• 1993

NJOY 코드 씨스템으로 JEF-1 평가 핵자료를 처리하여 고속로용 50군 군정수 SET을 생산하였다. 이를 이용하여 스물일곱가지 고속 임계로심 실험에서 얻어진 임계도 및 노심 중앙에서의 반응 율비를 계산하고 측정치와 비교·분석하였다. 아울러 ENDF/B-IV와-V 자료로 해석한 결과와도 비교·검토하였다. 일반적으로, 임계실험의 적분량 추정에서 JEF-1의 결과는 지금까지 사용해온 ENDF/B-IV의 결과보다 개선되었고, ENDF/B-V의 결과에 근접하고 있다.

• JSTS:Journal of Semiconductor Technology and Science
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• 제5권3호
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• pp.149-158
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• 2005

This paper describes design, fabrication, and application of the silicon based temperature controllable micro reactor. In order to achieve fast temperature variation and low energy consumption, reaction chamber of the micro reactor was thermally isolated by etching the highly conductive silicon around the reaction chamber. Compared with the model not having thermally isolated structure, the thermally isolated micro reactor showed enhanced thermal performances such as fast temperature variation and low energy consumption. The performance enhancements of the micro reactor due to etched holes were verified by thermal experiment and numerical analysis. Regarding to 42 percents reduction of the thermal mass achieved by the etched holes, approximately 4 times faster thermal variation and 5 times smaller energy consumption were acquired. The total size of the fabricated micro reactor was $37{\times}30{\times}1mm^{3}$. Microchannel and reaction chamber were formed on the silicon substrate. The openings of channel and chamber were covered by the glass substrate. The Pt electrodes for heater and sensor are fabricated on the backside of silicon substrate below the reaction chamber. The dimension of channel cross section was $200{\times}100{\mu}m^{2}$. The volume of reaction chamber was $4{\mu}l$. The temperature of the micro reactor was controlled and measured simultaneously with NI DAQ PCI-MIO-16E-l board and LabVIEW program. Finally, the fabricated micro reactor and the temperature control system were applied to the thermal denaturation and the trypsin digestion of protein. BSA(bovine serum albumin) was chosen for the test sample. It was successfully shown that BSA was successfully denatured at $75^{\circ}C$ for 1 min and digested by trypsin at $37^{\circ}C$ for 10 min.

• 전자공학회논문지
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• 제51권11호
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• pp.187-192
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• 2014

카메라와 같이 연속적인 영상을 제공하는 환경에서 특징 점들을 추출하기 위해 다양한 알고리즘들이 연구되고 있다. 특히, FAST (Feature from Accelerated Segment Test) 알고리즘은 연산 구조가 간단하고 실시간 특징 점 추출이 용이하여 FPGA 기반 하드웨어 가속기로 구현되어 사용되고 있다. 사용된 FAST 하드웨어 가속기는 특징 점을 추출하기 위해 임계값을 필요로 한다. 임계값은 영상에서 추출되는 특징 점의 기준이 되는 값으로, 값의 크기에 따라 추출되는 특징 점의 개수가 정해질 뿐만 아니라 전체 수행시간에도 영향을 주기 때문에, 일정한 수행시간 동안에 많은 특징 점들을 추출하기 위해서는 적절한 임계값 제어 방법이 요구된다. 본 논문에서는 임계값 제어를 위해 PI 제어기를 제안한다. 제안한 PI 제어기는 시험 영상들을 통해 기능 및 성능을 검증하였고, Xilinx Vertex IV FPGA 기반의 로직으로 구현 비용을 계산하였다. 제안한 PI 제어기는 47개의 Flip Flops, 146개의 LUTs, 그리고 91개의 Slices을 사용해, FAST 하드웨어 가속기 2.1%의 Flip Flop, 4.4%의 LUTs, 그리고 4.6%의 Slice에 해당하는 적은 비용으로 구현되었다.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.

• 대한기계학회논문집
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• 제14권1호
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• pp.112-118
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• 1990

본 연구에서는 ASTM-E24.01.06에서 제안하고 있는 실험방법을 응용하여 균열 정지 파괴인성값을 측정하였다.즉 쐐기와 분리형 부싱(wedge and split bushing)으 로 압축하중을 가함으로 균열선 웨지하중 시편[crack line wedge loaded specimen(CL- WL시편)]에 인장력을 발생시켜서 균열정지 응력확대계수( $K_{1a}$)를 결정하였다. 그리고 균열개시 응력확대계수가 균열정지 응력확대계수에 미치는 영향들을 여러가지 재료들에 대하여 체계적으로 검토하였다.다.

• 한국초전도ㆍ저온공학회논문지
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• 제13권1호
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• pp.41-45
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• 2011

Hybrid fault current limiters (FCL) have been researched at Yonsei University. The hybrid FCL has advantages such as having a rapid response to a sudden fault situation and a fast recovery time from a quench. It consists of an asymmetric HTS coil, a switching module, and a bypass reactor. The asymmetric HTS coil is wound with two different types of HTS wires in an opposite direction so that it has nearly zero inductance at the superconducting state. When the quench occurs at the fault state, a strong magnetic field is generated from the asymmetric coil because of different quench characteristics of two HTS wires, and then a repulsive force is induced in the switching module. The force opens the switch and the fault current is pushed into the bypass reactor. In this research, we analyzed the cause of the repulsive force and confirmed, experimentally and computationally, that the magnitude of a repulsive force is varied by changing the gap distance between the asymmetric coil and the switching module. By using the FEM simulation, we calculated the repulsive force with respect to the gap distance and verified that the effect of the gap distance. Then, short circuit test was carried out to confirm the correct operation of the fast switch.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.