• Journal of Ceramic Processing Research
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• v.13 no.spc2
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• pp.328-331
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• 2012

With the recent development of super-precision optical instruments, camera modules for devices, such as portable terminals and digital camera lenses, are increasingly being used. Since an optical lens is usually produced by high-temperature compression molding methods using tungsten carbide (WC) alloy molding cores, it is necessary to develop and study technology for super-precision processing of molding cores and coatings for the core surface. In this study, Rhenium-Iridium (Re-Ir) thin films were deposited onto a WC molding core using a sputtering system. The Re-Ir thin films were prepared by a multi-target sputtering technique, using iridium, rhenium, and chromium as the sources. Argon and nitrogen were introduced through an inlet into the chamber to be the plasma and reactive gases. The Re-Ir thin films were prepared with targets having a composition ratio of 30 : 70, and the Re-Ir thin films were formed with a 240 nm thickness. Re-Ir thin films on WC molding core were analyzed by scanning electron microscope (SEM), atomic force microscope (AFM), and Ra (the arithmetical average surface roughness). Also, adhesion strength and coefficient friction of Re-Ir thin films were examined. The Re-Ir coating technique has received intensive attention in the coating processes field because of promising features, such as hardness, high elasticity, abrasion resistance and mechanical stability that result from the process. Re-Ir coating technique has also been applied widely in industrial and biomedical applications. In this study, WC molding core was manufactured, using high-performance precision machining and the effects of the Re-Ir coating on the surface roughness.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.3
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• pp.638-644
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• 2019

In this study, the fluid flow and heat transfer characteristics within sandwich panels are investigated using computational fluid dynamics. Within the sandwich panels having periodic cellular cores, air can freely move inside the core section so that the structure is able to perform multi-functional roles such as simultaneous load bearing and heat dissipation. Thus, there needs to examine the thermal and flow analysis with respect to design variables and various conditions. In this regard, ANSYS Fluent was utilized to explore the flow and heat transfer within the pyramidal truss sandwich structures by varying the truss angle and inlet velocity. Without the entry effect in the first unitcell, the constant rate of pressure and the constant rate of Nusselt number was observed. As a result, it was demonstrated that Nusselt number increases and friction factor decreases as the inlet velocity increases. Moreover, the rate of Nusselt number and friction factor was appreciable in the range of V=1-5m/s due to the transition from laminar to turbulent flow. Regarding the effect of design variable, the variation of truss angle did not significantly influence the characteristics.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.68-79
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• 1999

The YGN Units 3&4 plant conditions during shutdown operation were reviewed to identify the possible event scenarios following the loss of shutdown cooling (SDC) event. For the five cases of typical reactor coolant system (RCS) configurations under the worst event sequence, such as unavailable secondary cooling and no RCS inventory makeup, the thermal hydraulic analyses were performed using the RELAP5/MOD3.2 code to investigate the plant behavior following the event. The thermal hydraulic analyses include the estimation of time to boil, time to core uncovery, and time to core heat up to determine the containment closure time to prevent the uncontrolled release of fission products to atmosphere. The result indicates that the containment closure is recommended to be achieved within 42 minutes after the loss of SDC for the steam generator (SG) inlet plenum manway open case or the large cold leg open case under the worst event sequence. The containment closure time is significantly dependent on the elevation and size of the opening and the SG secondary water level condition. It is also found that the containment closure needs to be initiated before the boiling time to ensure the survivability of the workers in the containment. These results will provide useful information to operators to cope with the loss of SDC event.

• Nuclear Engineering and Technology
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• v.45 no.5
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• pp.581-588
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• 2013

This paper deals with the Safety Analysis for $CANDU^{(R)}$ 6 nuclear reactors as affected by main Heat Transport System (HTS) aging. Operational and aging related changes of the HTS throughout its lifetime may lead to restrictions in certain safety system settings and hence some restriction in performance under certain conditions. A step in confirming safe reactor operation is the tracking of relevant data and their corresponding interpretation by the use of appropriate thermal-hydraulic analytic models. Safety analyses ranging from the assessment of safety limits associated with the prevention of intermittent fuel sheath dryout for a slow Loss of Regulation (LOR) analysis and fission gas release after a fuel failure are summarized. Specifically for fission gas release, the thermal-hydraulic analysis for a fresh core and an 11 Effective Full Power Years (EFPY) aged core was summarized, leading to the most severe stagnation break sizes for the inlet feeder break and the channel failure time. Associated coolant conditions provide the input data for fuel analyses. Based on the thermal-hydraulic data, the fission product inventory under normal operating conditions may be calculated for both fresh and aged cores, and the fission gas release may be evaluated during the transient. This analysis plays a major role in determining possible radiation doses to the public after postulated accidents have occurred.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.204-214
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• 1993

Work on fracture mechanics has provided a technical basis for elimination of main coolant loop double ended guillotine breaks from the structural design basis of reactor coolant system. Without main coolant loop pipe breaks, the tributary pipe breaks must be considered as design bases until further fracture mechanics work could eliminate some of these breaks from design consideration. This paper determines the core support barrel shell responses for the 3 inch pressurizer spray line nozzle break which is expected to be the only inlet break remaining in the primary side after leak-before-break evaluation is extended to smaller size pipes in the near future. The responses are compared with those due to 14 inch safety injection nozzle break and main coolant loop pipe break. The results show that, when the leak-before-break concept is applied to the primary side piping systems with a diameter of 10 inches or over, the core support barrel shell responses due to pipe breaks in the primary side are negligible for the faulted condition design.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1584-1602
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• 2024

It is vital to output the required electrical power following various task requirements when the space reactor power supply is operating in orbit. The dynamic performance of the closed Brayton cycle thermoelectric conversion system is initially studied and analyzed. Based on this, a load tracking power regulation method is developed for the liquid metal cooled space reactor power system, which takes into account the inlet temperature of the lithium on the hot side of the intermediate heat exchanger, the filling quantity of helium and xenon, and the input amount of the heat pipe radiator module. After comparing several methods, a power regulation method with fast response speed and strong system stability is obtained. Under various changes in power output, the dynamic response characteristics of the ultra-small liquid metal lithium-cooled space reactor concept scheme are analyzed. The transient operation process of 70 % load power shows that core power variation is within 30 % and core coolant temperature can operate at the set safety temperature. The second loop's helium-xenon working fluid has a 65K temperature change range and a 25 % filling quantity. The lithium at the radiator loop outlet changes by less than ±7 K, and the system's main key parameters change as expected, indicating safety. The core system uses less power during 30 % load power transient operation. According to the response characteristics of various system parameters, under low power operation conditions, the lithium working fluid temperature of the radiator circuit and the high-temperature heat pipe operation temperature are limiting conditions for low-power operation, and multiple system parameters must be coordinated to ensure that the radiator system does not condense the lithium working fluid and the heat pipe.

• Journal of Advanced Marine Engineering and Technology
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• v.35 no.7
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• pp.907-913
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• 2011

Swirling flows are related to the spiral motion in the tangential direction in addition to the axial and radial direction using several swirl generators. These type of flows are used in combustion chambers to improve flame stability, heat exchanger to enhance heat transfer coefficients, agricultural spraying machines and some vertical pipes to move slurries or transport of materials. However, only a few studies three dimensional velocity profiles in a vertical pipe have been reported. In this present paper, 3 dimension particle image velocimetry(PIV) technique was employed to measure the velocity profiles in water along a vertical circular pipe with Reynolds number from 6000 to 13,000. A tangential inlet condition was used as the swirl generator to produce the required flow. The velocities were measured with swirling flow in the water along the test section using the PIV technique.

• Journal of computational fluids engineering
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• v.7 no.3
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• pp.44-52
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• 2002

An analytic methodology was developed for free surface motions between liquid metal coolant and cover gas in order to calculate the phenomena of gas entrainment in hot pool surface through IHX EMP and reactor core. The methodology was setup by applying the first order VOF convection model to CFX4 general purpose fluid dynamics analysis code. The methodology was validated by applying it to an experimental apparatus designed for free surface motions of KALIMER reactor. The distributions of free surface calculated by the present methodology were almost coincident with the experimental data. The developed methodology was applied to the KALIMER reactor of full power operating condition. The shapes of the free surface were nearly uniform. From the results, it was found that the altitude of the free surface from the IHX inlet nozzle of KALIMER reactor is high enough not to affect to free surface motions of generating gas bubbles from the turbulent shear flows such as hydraulic jump and water falls.

• Proceedings of the KIEE Conference
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• 1998.11c
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• pp.1054-1060
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• 1998

Previous counter-current flow limitation (CCFL) and critical heat flux (CHF) studies included investigations on the inlet entrance, inclined channel and gap effects for the most part. In this study, the local CHF correlation was presented to be used in the numerical analysis for the 3 dimensional hemispherical geometry. Also, first-principle analyses were performed to determine the maximum heat removal capability from the debris through the gap that may be formed during a core melt accident. The maximum heat removal capability by gap cooling can be applied in quantitatively assessing the severe accident management measures.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.254-259
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• 1996

The hydraulic tests for 16$\times$16 JDFA were performed to obtain the pressure loss coefficients using the FACTS. The pressure loss coefficients are calculated by converting the each properties of experimental values for inlet region, mixing vane grid, outlet region and core region by performing a power fit of the pressure loss coefficient values to the corresponding Reynolds number. The test results are compared with the existing calculated values and evaluated by using the CALOPR code in terms of pressure drop. It is turned out that the differences between the test results and the calculated values are about by 3.8% for the pressure loss coefficients and by 8.5% for the pressure drop.