• Journal of Electrical Engineering and Technology
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• 제1권3호
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• pp.359-365
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• 2006

Since the start of the KSTAR (Korea Superconducting Tokamak Advanced Research) project, Instrumentation and Control (I&C) of the Neutral Beam Test Facility (NB-TF) has been striving to answer diverse requests arising from various facets during the project's development and construction phases. Hard-wired electrical circuits have been designed, tested, fabricated, and finally installed to the relevant parts of the system. In relation to the vacuum system I&C, controlling functions for the rotary pumps, a Roots pump, two turbomolecular pumps, and four cryosorption pumps have been constructed. I&C for the ion source operation are the temperature and flow rate signal monitoring, Langmuir probe signal measurements, gradient grid current measurements, and arc detector circuit. For the huge power system to be monitored or safely operated, many temperature measurement functions have also been implemented for the beam line components like the neutralizer, bending magnet, ion dump, and calorimeter. Nearly all of the control and probe signals between the NB test stand and the control room were made to be transmitted through the optical cables. Failures of coolant flow or beam line vacuum pressure were made to be safely blocked from influencing the system by an appropriate interlock circuit that will shut down the extraction voltage application to the system or prevent damages to the vacuum components. Preliminary estimation of the beam power through the calorimetric measurement shows that 87.9% of the total power of the 60kV/18A beam with 200 seconds duration is absorbed by the calorimeter surface. Most of these I&C results would be highly appropriate for the construction of the main NBI facility for the KSTAR national fusion research project.

• 한국소음진동공학회논문집
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• 제25권12호
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• pp.807-814
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• 2015

A control rod drive mechanism(CRDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod containing a neutron absorbing material within a reactor core to control the reactivity of the core. The top-mounted CRDM for Jordan Research and Training Reactor(JRTR) with 5 MW power has been designed and fabricated based on the HANARO's experience through KAERI and DAEWOO consortium project. This paper describes the performance qualification test results to demonstrate the operability of a prototype and four production CRDMs during the reactor lifetime. The driving performance, the drop performance and the endurance tests for CRDM are carried out at a test rig simulating the actual reactor conditions. A vibration of internal components due to the coolant flow is also measured using a laser vibrometer. As a result, the CRDMs are driven having a good driving performance without a malfunction between command and output signals for the stepping motor. Also, the pure drop time and the impact acceleration are within 0.72 s and 4.2 g to meet the design requirements, and the vibrational displacement of control rod is measured as maximum $5.2{\mu}m$.

• Nuclear Engineering and Technology
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• 제38권2호
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• pp.185-194
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• 2006

Transient operations for an integral type reactor, SMART-P, have been experimentally investigated using a thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), in order to verify the system design and performance of the SMART-P, a pilot plant of SMART. The VISTA facility was subjected to various accident conditions such as feedwater increase and decrease, loss of coolant flow, and control rod withdrawal accidents in order to elucidate the thermal-hydraulic responses following such accidents and finally to verify the system design of the SMARTP. Full functional control logics have been implemented in the VISTA facility in order to control the required control action for an accident simulation. As one of the sensitivity tests to verify the PRHRS performance, the effects of the initial water level in the compensation tank are experimentally investigated. When the initial water level is 16%, the water is quickly drained and nitrogen gas is then introduced into the PRHR system, resulting in deterioration of the PRHRS performance. It is thus found that nitrogen ingression should be prevented to ensure stable PRHRS operation.

• 한국재료학회지
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• 제10권6호
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• pp.445-449
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• 2000

가압 경수형 원자로 (pressurized light water reactor) 냉각재 계통 내의 주된 분식 생성물로 알려져 있는 nickel ferrite의 거동에 대해 고찰하기 위해 모의 nickel ferrite($Ni_{0.75}Fe_{2.25}O_4$)를 공침법으로 제조하였다. 수용액-pH-조절로는 am-monia 또는 potassium carbonate를, 공침물-세척제는 ammonia 수용액이나 potassium carbonate 수용액 또는 2차 증류수를 사용하였다. Nickel ferrite의 생성 및 수용액-pH-조절제와 공치물-세척제가 최종 생성물의 Ni-Fe 몰 비에 따른 수율 및 특성에 미치는 영향은 EDX, XPS, XRD 및 SEM으로 고찰하였다. 반응 전.후 Ni/Fe 몰 비에 따른 수율은, pH를 potassium carbon-ate로 조절한 후 2차 증류수로 공침물을 세척한 경우가 0.994로 가장 높이 나왔으며, pH-조절제로 potassium carbonate를 사용한 경우가 ammonia를 사용한 경우에 비해 높은 수율을 나타냈다. 이러한 차이는 공침 시에 수용액 내에서 ammonia가 보여주는 상대적으로 큰 $Na_{2+}{\leftarrow}NH_3$ 착화 효과와 더불어 공침물-세척제의 pH에 기인하는 것으로 해석하였다.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• 한국자기학회지
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• 제26권5호
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• pp.154-158
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• 2016

원자력발전소의 원자로에는 노심 반응 속도를 제어하기 위하여 제어봉구동장치가 사용된다. 한국원자력연구원의 SMART 원자로는 원자로 가동 중 제어봉집합체의 위치를 확인하기 위하여 제어봉구동장치에 영구자석과 리드스위치로 구성되는 위치지시기가 설치된다. 원자로 가동 온도는 최대 $350^{\circ}C$로 고려되어 설계되며, 영구자석은 원자로 내에 설치된다. 반면에 리드스위치와 전기회로는 원자로 외부에 설치된다. Test coil은 리드스위치의 품질 검증을 위한 장비로서, 코일과 철심으로 구성되어 있다. 본 연구는 리드스위치에 미치는 Test coil과 영구자석의 자기 특성을 비교하고자 수행되었으며, 유한요소 전자기 시뮬레이션을 활용하였다.

• 대한전자공학회논문지
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• 제17권6호
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• pp.33-37
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• 1980

자동차의 패널에 전자식 표시장치를 도입하고, 마이크로프로셋서를 사용하여 운전자에게 차의 동작상태 및 여행자료와 같은 정보를 표시하는 정보 표시시스템(information display system)을 설계 개발하였다. 본 시스템의 하드웨어로는 기능 선택 keyboard, 중앙 처리장치 표시공(displays)등이 있으며, 소프트웨어로는 여러 가지 감지기(Sensors)의 입력으로 부터, 주행속도, 사용 가능한 연료량, 냉긱수 은도, 바테리전압, 목적지까지 남은 거리, 현재의 시긱등 12가지의 여행자료등 운전자가 원하는 정보로 바꾸어 주는 main routine을 비롯하여, keyboard 및 연splay를 위한 interrupt service routine으로 구성하였다. 마지막으로, 본 시스템을 실제로 실장시험한 결과와 문제점을 논의하였다.

• 대한기계학회논문집B
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• 제37권1호
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• pp.35-41
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• 2013

금형온도는 사출성형시 수지의 유동특성이나 성형품의 변형에 영향을 미치는 중요한 변수로서, 고온의 수지 주입과 냉각회로에 주입되는 냉각수의 영향을 받아 사출 사이클이 반복될수록 온도의 상승과 하강이 반복되는 주기적인 변화특성을 가지고 있다. 본 연구에서는 금형 냉각회로에 저온과 고온의 유체를 번갈아 주입하는 가변 금형온도 제어기법을 적용하여 성형전에는 금형온도를 높게 유지하고 성형후에는 낮게 유지함으로써 사출성형시 품질과 생산성을 동시에 높일 수 있는 연구를 수행하였다. 특히 열전달-유동해석을 연계한 다중사이클 사출성형 과도해석을 수행하여 수지와 금형, 냉각수간의 과도적인 온도변화를 수치적으로 고찰하였고, 기존 냉각방법과의 해석결과를 비교하여 제안된 가변 금형온도 제어기법의 가열 및 냉각과정에서의 효율성을 비교하였다.

• 한국수소및신에너지학회논문집
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• 제23권4호
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• pp.323-329
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• 2012

A thermal management system of a proton exchange membrane fuel cell is taken charge of controlling the temperature of fuel cell stack by rejection of electrochemically reacted heat. Two major components of thermal management system are heat exchanger and pump which determines required amount of heat. Since the performance and durability of PEMFC system is sensitive to the operating temperature and temperature distribution inside the stack, it is necessary to control the thermal management system properly under guidance of operating strategy. The control study of the thermal management system is able to be boosted up with hardware in the loop simulation which directly connects the plant simulation with real hardware components. In this study, the plant simulation of fuel cell stack has been developed and the simulation model is connected with virtual data acquisition system. And HIL simulator has been developed to control the coolant supply system for the study of PEMFC thermal management system. The virtual data acquisition system and the HIL simulator are developed under LabVIEWTM Platform and the Simulation interface toolkit integrates the fuel cell plant simulator with the virtual DAQ display and HIL simulator.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2725-2732
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• 2020

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of ¹⁶O, ⁹Be, ²³⁵U, ²³⁸U, and S(𝛼,𝛽) of ¹H in H₂O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.