• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.25 no.12
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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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• v.38 no.2
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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.

• Korean Journal of Materials Research
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• v.10 no.6
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• pp.445-449
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• 2000

Nickel ferrite $(Ni_{0.75}Fe_{2.25}O_4$ was synthesized by co-precipitation method in order to investigate its behavior under conditions of the reactor coolant system in pressurized light water nuclear power plants. Ammonia or potassium carbonate was used as a solution pH control agent, and aqueous ammonia or potassium carbonate solution or secondary distilled water was used as a co-precipitate washing agent. The effects of the pH control agent and the co-precipitate washing agent on the production yield on the basis of the Ni/Fe molar ratio and the particle characteristics of final products were investigated by XRD, SEM, EDX and XPS. The production yield was almost congruent with that of the initial aqueous mixture in case of using potassium carbonate as a pH control agent, while in case of using ammonia, it was quite changed. The difference seemed to be due to the effects of $Ni^{2+}{\leftarrow}NH_3$complexation in the aqueous solution and of the pH of co-precipitate washing agent.

• Nuclear Engineering and Technology
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• v.55 no.10
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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.

• Journal of the Korean Magnetics Society
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• v.26 no.5
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• pp.154-158
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• 2016

A CRDM (Control Rod Drive Mechanism) is an electromagnetic device which drives a control rod assembly linearly to regulate the reactivity of a nuclear core. An RPIS (Rod Position Indication System) is used as a position indicator for a control rod assembly of a CRDM of SMART, and an RPIS consists of permanent magnets and reed switches. SMART is designed for the maximum coolant temperature of $350^{\circ}C$, and the permanent magnets are installed inside of the reactor. The reed switches and electrical circuit are installed outside of the reactor on the other hand. Test coil for a reed switch is test equipment for quality verification of a reed switch, and a test coil consists of a coil and core. In this study, magnetic property of test coil and permanent magnet on a reed switch is compared by using finite element electromagnetic simulation.

• Journal of the Korean Institute of Telematics and Electronics
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• v.17 no.6
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• pp.33-37
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• 1980

The information display control system is designed and implemented on an automobile in which the conventional panel displays are replaced by electronic ones. The system hardware consists of three main parts, i. e., (i) the function select keyboard (ii) the central processing unit (iii) the displays, The system software consists of main routine and several interrupt service routine such as keyboaiuand display interrupt service routine:. The main routine handles various sensor inputs to generate the appropriate information for the driver such as running speed, available fuel quantity. coolant temperature, battery voltage, remaining distance to the destination , time of day, and so on. Finally the results of the field test of the system and some associated difficulties of realization problems are discussed.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.37 no.1
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• pp.35-41
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• 2013

In injection molding, the mold temperature is one of most important process parameters that affect the flow characteristics and part deformation. The mold temperature usually varies periodically owing to the effects of the hot polymer melt and the cold coolant as the molding cycle repeats. In this study, a pulsed mold temperature control was proposed to improve the part quality as well as the productivity by alternatively circulating hot water and cold water before and after the molding stage, respectively. Transient thermal-fluid coupled analyses were performed to investigate the heat transfer characteristics of the proposed pulsed mold heating and cooling system. The simulation results were then compared with those of the conventional mold cooling system in terms of the heating and cooling efficiencies of the proposed pulsed mold temperature control system.

• Transactions of the Korean hydrogen and new energy society
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• v.23 no.4
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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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• v.52 no.12
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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.

• Transactions of Materials Processing
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• v.14 no.4 s.76
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• pp.368-374
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• 2005

In the present study we used a diagrammatic analysis of 6 sigma quality control and Taguchi method for injection molding of monitor back-cover, evaluated the influence on the cycle time with part design, mold design, molding process and standardization activity involving design and molding, adopted analysis of sensitivity and effective factors of the part design and molding process conditions for productivity, identified main design molding factors. The contributing factors for the final cycle time could be enumerated as follows; the thickness of hot spot, main nominal part thickness, coolant inlet temperature, melt temperature and cooling line layout, etc.. As a first step, all the critical factors of design process applied to the current monitor housing were investigated through 6 sigma process. Thereafter, the optimal and better critical factors found in the first step were applied to new product design to prove that our process was correct. The Moldflow was used for injection molding simulation, and Minitab software for the statistical analysis, respectively. Finally, the productivity of new design was increased about 33 percents for our specific case.