• Journal of Electrical Engineering and Technology
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• v.1 no.3
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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.

• Journal of the Microelectronics and Packaging Society
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• v.22 no.4
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• pp.31-36
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• 2015

The demand for multi-functionality, high density, high performance, and miniaturization of IC devices has caused the technology paradigm shift for electronic packaging. So, thermal management of new packaged chips becomes a bottleneck for the performance of next generation devices. Among various thermal solutions such as heat sink, heat spreader, TIM, thermoelectric cooler, etc. on-chip liquid cooling module was investigated in this study. Micro-channel was fabricated on Si wafer using a deep reactive ion etching, and 3 different micro-channel designs (straight MC, serpentine MC, zigzag MC) were formed to evalute the effectiveness of liquid cooling. At the heating temperature of $200^{\circ}C$ and coolant flow rate of 150ml/min, straight MC showed the high temperature differential of ${\sim}44^{\circ}C$ after liquid cooling. The shape of liquid flowing through micro-channel was observed by fluorescence microscope, and the temperarue differential of liquid cooling module was measuremd by IR microscope.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.426-441
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• 2003

A reactor inlet header break experiment, B9401, performed in the RD-14M multi channel test facility was analyzed using RELAP5/MOD3.2 and RELAP5/CANDU[1]. The RELAP5 has been developed for the use in the analysis of the transient behavior of the pressurized water reactor. A recent study showed that the RELAP5 could be feasible even for the simulation of the thermal hydraulic behavior of CANDU reactors. However, some deficiencies in the prediction of fuel sheath temperature and transient behavior in athe headers were identified in the RELAP5 assessments. The RELAP5/CANDU has been developing to resolve the deficiencies in the RELAP5 and to improve the predictability of the thermal-hydraulic behaviors of the CANDU reactors. In the RELAP5/CANDU, critical heat flux model, horizontal flow regime map, heat transfer model in horizontal channel, etc. were modified or added to the RELAP5/MOD3.2. This study aims to identify the applicability of both codes, in particular, in the multi-channel simulation of the CANDU reactors. The RELAP5/MOD3.2 and the RELAP5/CANDU analyses demonstrate the code's capability to predict reasonably the major phenomena occurred during the transient. The thermal-hydraulic behaviors of both codes are almost identical, however, the RELAP5/CANDU predicts better the heater sheath temperature than the RELAP5/MOD3.2. Pressure differences between headers govern the flow characteristics through the heated sections, particularly after the ECI. In determining header pressure, there are many uncertainties arisen from the complicated effects including steady state pressure distribution. Therefore, it would be concluded that further works are required to reduce these uncertainties, and consequently predict appropriately thermal-hydraulic behaviors in the reactor coolant system during LOCA analyses.

• Journal of the Korean Society of Safety
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• v.20 no.2 s.70
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.17 no.7
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• pp.31-41
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• 2016

A computer program that simulates electric utility steam condensers was developed, and used to investigate the effects of enhanced tubes in steam condensers. The replacement of smooth tubes with enhanced tubes reduces the steam condensing temperature, and increases the efficiency of the electric utility. Therefore, a significant amount of power may be reserved without any modification of the utility. Three enhanced tubes, corrugated, low fin with internal ribs, and low fin with internal 3-D roughness, were considered. The results showed that there is an optimal internal roughness height. Low fin tubes with a 3-D roughness were superior to the other enhanced geometries. This was attributed to longitudinal vortices generated between the circumferential dimples. An additional 0.5 MW~1.3 MW was possible when smooth tubes were replaced with enhanced tubes in the 600 MW electric utility condenser. The additional power increased with increasing coolant temperature. More investigations on fouling, corrosion, and mechanical properties will be necessary for actual applications of enhanced tubes in electric utility condensers.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.597-609
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• 2020

Accident Tolerant Fuels have been widely studied since the Fukushima-Daiichi accident in 2011 as one of the options on how to further enhance the safety of nuclear power plants. Deposition of protective coatings on nuclear fuel claddings has been considered as a near-term concept that will reduce the high-temperature oxidation rate and enhance accidental tolerance of the cladding while providing additional benefits during normal operation and transients. This study focuses on experimental testing of Zr-based alloys coated with Cr-based coatings using Physical Vapour Deposition. The results of long-term corrosion tests, as well as tests simulating postulated accidents, are presented. Zr-1%Nb alloy used as nuclear fuel cladding serves as a substrate and Cr, CrN, Cr_xN_y layers are deposited by unbalanced magnetron sputtering and reactive magnetron sputtering. The deposition procedures are optimized in order to improve coating properties. Coated as well as reference uncoated samples were experimentally tested. The presented results include standard long-term corrosion tests at 360℃ in WWER water chemistry, burst (creep) tests and mainly single and double-sided high-temperature steam oxidation tests between 1000 and 1400℃ related to postulated Loss-of-coolant accident and Design extension conditions. Coated and reference samples were characterized pre- and post-testing using mechanical testing (microhardness, ring compression test), Thermal Evolved Gas Analysis analysis (hydrogen, oxygen concentration), optical microscopy, scanning electron microscopy (EDS, WDS, EBSD) and X-ray diffraction.

• Progress in Superconductivity and Cryogenics
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• v.19 no.2
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• pp.19-24
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• 2017

Cryogenic milling which is a combined process of low-temperature treatment and mechanical milling was applied to fabricate high critical current density $(J_c)MgB_2$ bulk superconductors. Liquid nitrogen was used as a coolant, and no solvent or lubricant was used. Spherical Mg ($6-12{\mu}m$, 99.9 % purity) and plate-like B powder (${\sim}1{\mu}m$, 97 % purity) were milled simultaneously for various time periods (0, 2, 4, 6 h) at a rotating speed of 500 rpm using $ZrO_2$ balls. The (Mg+2B) powders milled were pressed into pellets and heat-treated at $700^{\circ}C$ for 1 h in flowing argon. The use of cryomilled powders as raw materials promoted the formation reaction of superconducting $MgB_2$, reduced the grain size of $MgB_2$, and suppressed the formation of impurity MgO. The superconducting critical temperature ($T_c$) of $MgB_2$ was not influenced as the milling time (t) increased up to 6 h. Meanwhile, the critical current density ($J_c$) of $MgB_2$ increased significantly when t increased to 4 h. When t increased further to 6 h, however, $J_c$ decreased. The $J_c$ enhancement of $MgB_2$ by cryogenic milling is attributed to the formation of the fine grain $MgB_2$ and a suppression of the MgO formation.

• Korean Journal of Materials Research
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• v.17 no.7
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• pp.366-370
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• 2007

Silicon nitride ($Si_3N_4$) ceramics have been considered for various components of nuclear power plants such as the mechanical seal of a reactor coolant pump (RCP), the guide roller for a control rod drive mechanism (CRDM), and a seal support, etc. Corrosion behavior of $Si_3N_4$ ceramics in a high-temperature and high-pressure water must be elucidated before they can be considered as components for nuclear power plants. In this study, the corrosion behaviors of $Si_3N_4$ ceramics containing MgO and $Al_2O_3$ as sintering aids were investigated at a hydrothermal condition ($300^{\circ}C$, 9.0 MPa) in pure water and 35 ppm LiOH solution. The corrosion reactions were controlled by a diffusion of the reactive species and/or products through the corroded layer. The grain-boundary phase was preferentially corroded in pure water whereas the $Si_3N_4$ grain seemed to be corroded at a similar rate to the grain-boundary phase in LiOH solution. Flexural strengths of the $Si_3N_4$ ceramics were significantly degraded due to the corrosion reaction. Results of this study imply that a variation of the sintering aids and/or a control (e.g., crystallization) of the grain-boundary phase are necessary to increase the corrosion resistance of $Si_3N_4$ ceramics in a high-temperature water.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.1
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• pp.188-195
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• 2002

To investigate the degradation of mechanical properties induced mainly by neutron irradiation, the tensile tests were conducted from room temperature to 300\\`c using the irradiated and the unirradiated Zr-2.5Nb pressure tube materials. The irradiated longitudinal and transverse specimens were collected from the coolant inlet, middle, and outlet parts of M-11 tube which had been operated in Wolsung CANDU Unit-1 and exposed to different operating temperatures and irradiation fluences. The different tensile behavior was characterized not by the fluences of irradiation but by the tensile loading direction. The transverse specimen showed the higher strength and lower elongation than those of the longitudinal one. It was believed that these phenomena resulted from the microstructure anisotropy caused by the extrusion process. The increased strength hardening and decreased elongation embrittlement of the irradiated material were compard to those of the unirradiated one. While the tensile strength of the inlet was higher than that of the outlet, the elongation of the inlet was lower than that of outlet. Considering the operation condition, it was proposed that the operating temperature could be a more effective parameter than the irradiation fluence for long-time life. Through the TEM observation, it was found that while the a-type dislocation density was increased, the c-type dislocation was not changed in the irradiated. The fact that the higher dislocation density was sequentially distributed over the inlet, the middle, and the outlet parts was consistent with the distribution of the tensile strength.

• Transactions of the Korean Society of Automotive Engineers
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• v.20 no.2
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• pp.123-134
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• 2012

A typical cooling system of an engine relies on a water pump that circulates the coolant through the system. The pump is typically driven by the crankshaft through a mechanical link with engine starting. In order to reduce the friction and warm-up time of an engine, the clutch-type water pump (CWP) was applied in 2.0 liter diesel vehicle. The clutch-type water pump can force cooling water to supply into an engine by the operation of an electromagnetic clutch equipped as the inner part of pump system. The onset of CWP is decided by temperature of cooling water and engine oil. And, the control logic for an optimal operation of the clutch-type water pump was developed and applied in engine and vehicle tests. In this study, the warm-up time was measured with the conventional water pump and clutch-type water pump in engine tests. And the emission and the fuel consumption were evaluated under NEDC mode in vehicle tests. Also, tests were carried out for the various temperature conditions starting the operation of CWP. From the results of the study, the application of CWP can improve the fuel consumption and $CO_2$ reduction by about 3%.