• Imaging Science in Dentistry
• /
• v.46 no.2
• /
• pp.69-75
• /
• 2016

Purpose: Sufficient area in the interforaminal region is required for dental implant placement, and the anterior loop of the mandibular canal is located within the limits of this area. The aim of this study was to evaluate the prevalence and extent of the anterior loop in a Brazilian sample population using cone-beam computed tomography (CBCT). Materials and Methods: CBCT images from 250 patients (500 hemimandibles) obtained for various clinical indications were randomly selected and evaluated to determine the presence and length of the anterior loop. The length of the anterior loop was then compared based on gender, age, and the side of the mandible. The data were analyzed using the Pearson chi-square test and linear regression analysis. Results: An anterior loop was identified in 41.6% of the cases, and its length ranged from 0.25 mm to 4.00 mm (mean, $1.1{\pm}0.8mm$). The loop had a greater mean length and was significantly more prevalent in males (p=0.014). No significant differences were found between the right and left sides regarding length (p=0.696) or prevalence (p=0.650). Conclusion: In this study, a high prevalence of the anterior loop of the mandibular canal was found, and although its length varied greatly, in most cases it was less than 1 mm long. Although this is a prevalent anatomical variation, safety limits for the placement of implants in this region cannot be established before an accurate evaluation using imaging techniques in order to identify and preserve the neurovascular bundles.

• Nuclear Engineering and Technology
• /
• v.41 no.1
• /
• pp.53-62
• /
• 2009

A set of experiments has been conducted on the performance sensitivity of the passive residual heat removal system (PRHRS) for an advanced integral type reactor, SMART, by using a high temperature and high pressure thermal-hydraulic test facility, the VISTA facility. In this paper the effects of the opening delay of the PRHRS bypass valves and the closing delay of the secondary system isolation valves, and the initial water level and the initial pressure of the compensating tank (CT) are investigated. During the reference test a stable flow occurs in a natural circulation loop that is composed of a steam generator secondary side, a secondary system, and a PRHRS; this is ascertained by a repetition test. When the PRHRS bypass valves are operated 10 seconds later than the secondary system isolation valves, the primary system is not properly cooled. When the secondary system isolation valves are operated 10 or 30 seconds later than the PRHRS bypass valves, the primary system is effectively cooled but the inventory of the PRHRS CT is drained earlier. As the initial water level of the CT is lowered to 16% of the full water level, the water is quickly drained and then nitrogen gas is introduced into the PRHRS, resulting in the deterioration of the PRHRS performance. When the initial pressure of the PRHRS is at 0.1MPa, the natural circulation is not performed properly. When the initial pressures of the PRHRS are 2.5 or 3.5 MPa, they show better performance than did the reference test.

• Journal of Power Electronics
• /
• v.14 no.6
• /
• pp.1217-1223
• /
• 2014

A novel half load-cycle worked dual input single output (DISO) DC/AC inverter is presented. The basic circuit consists of a dual buck regulator, which works in continuous current mode. The working principle of DISO DC/AC inverter has been used. The control method applied for half load-cycle worked DISO DC/AC inverter has been studied. The control effects of the open-loop proportional control and closed-loop proportional-integral control are compared by using PSIM software. The parameters are adopted in the realistic simulation and experiment test. Moreover, the waveforms, such as voltage of modulation reference signal and output voltage, were given. The simulation and experiment results proved that the half load-cycle worked DISO DC/AC inverter could achieve good performance, gain a line frequency of 50 Hz, and verify the correctness of theoretical analysis.

• Journal of the Korean Society of Fisheries and Ocean Technology
• /
• v.30 no.1
• /
• pp.33-44
• /
• 1994

This work studies for boiling and condensation heat transfer performance of trapezoidally shaped integral-fin tubes having fin densities from 748fpm to 1654fpm. For comparison, tests are made using a plain tube having the same inside and outside diameter as that of the root of fins of finned tubes. Hahne's theoretical model and Webb's theoretical model are used to predict the R-11 boiling heat transfer coefficient and condensing heat transfer coefficient respectively for plain tube and all integral-fin tubes. Experiments are carried out using R-11 as working fluid. This work is limited to film-wise condensation and pool boiling on the outside surface of plain tube and 4 low integral-fin tubes. In case of condensation, the refrigerant condenses at saturation state of 32$^{\circ}C$ on the outside tube surface cooled by coolant and in case of boiling. the refrigerant evaporates at saturation state of 1bar on the outside tube surface. The amount of non-con-densable gases in the test loop is reduced to a negligible value by repeated purging. The actual boiling and condensing processes occur on the outside tube surfaces. Hence the nature of this surface geometry affects the heat transfer performances of condenser and evaporator in refrigerating system. The condensation heat transfer coefficient of integral-fin tube is enhanced by both extended tube surface area and surface tension. The ratio of the condensation heat transfer coefficients of finned to plain tubes is greater than that of surface area of finned to plain tubes, while ratio of the boiling heat transfer coefficient of finned to plain tubes shows reverse result. As a result, low integral-fin tube can be used in condenser more effectively than used in evaporator.

• Proceedings of the KSME Conference
• /
• 2004.04a
• /
• pp.2053-2058
• /
• 2004

The pressurized light water cooled, medium power (330 MWt) SMART (System-integrated Modular Advanced ReacTor) has been under development at KAERI for a dual purpose : seawater desalination and electricity generation. The SMART design verification phase was followed to conduct various separate effects tests and comprehensive integral effect tests. The high temperature / high pressure thermal-hydraulic test facility, VISTA(Experimental Verification by Integral Simulation of Transient and Accidents) has been constructed to simulate the SMART-P (the one fifth scaled pilot plant) by KAERI. Experimental tests have been performed to investigate the thermal-hydraulic dynamic characteristics of the primary and the secondary systems. Heat transfer characteristics and natural circulation performance of the PRHRS (Passive Residual Heat Removal System) of SMART-P were also investigated using the VISTA facility. The coolant flows steadily in the natural circulation loop which is composed of the steam generator (SG) primary side, the secondary system, and the PRHRS. The heat transfers through the PRHRS heat exchanger and ECT are sufficient enough to enable the natural circulation of the coolant.

• Journal of the Institute of Electronics Engineers of Korea SC
• /
• v.45 no.6
• /
• pp.102-107
• /
• 2008

We propose an autotuning algorithm for PI controller with unknown plant. The proposed algorithm uses a saturation function and time delay element as a test signal. Since the integral element of PI controller reduces a phase margin in the closed loop system, the closed loop system could be resulted in unstable with PI controller. To avoid unstable in the closed loop system with PI controller, the proposed algorithm identifies one point information in the 3rd quadrant of Nyquist plot with a time delay element. The proposed method improves an accuracy of one point identified information with one saturation function. We demonstrate a performance of the proposed method via a simulation.

• The Journal of Korea Robotics Society
• /
• v.15 no.3
• /
• pp.212-220
• /
• 2020

This paper deals with a strategy of gain optimization for the kinematic control algorithm of a wire-driven surgical robot. The proposed controller consists of the closed-loop inverse kinematics with the back-calculation method. The closed-loop inverse kinematics has 18 PID control gains, and the back-calculation method has 6 gains. An efficient strategy is designed to optimize 18 values first and then the remaining 6 values. The optimal gain sets are searched under the step input with performance indices. In this gain optimization, the objective function is defined as the minimum value of signal-to-noise ratio of the performance indices for 6 DoF (Degree-of-Freedom) motion that is based on the Taguchi method, and the constraints are applied to obtain stable responses for each motion evenly. The gain sets obtained are verified by simulations using the test trajectories. In comparative results, the optimal gain value based on the performance index combined with ISE (integral of square error) and settling time showed the best control performance.

• Nuclear Engineering and Technology
• /
• v.48 no.5
• /
• pp.1083-1095
• /
• 2016

The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V&V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodiumshowed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.1066-1080
• /
• 2024

SMART100 has a containment pressure and radioactivity suppression system (CPRSS) for passive containment cooling system (PCCS). This prevents overheating and over-pressurization of a containment through direct contact condensation in an in-containment refueling water storage tank (IRWST) and wall condensation in a CPRSS heat exchanger (CHX) in an emergency cool-down tank (ECT). The Korea Atomic Energy Research Institute (KAERI) constructed scaled-down test facilities, SISTA1 and SISTA2, for the thermal-hydraulic validation of the SMART100 CPRSS. Three separate effect tests were performed using SISTA1 to confirm the heat removal characteristics of SMART100 CPRSS. When the low mass flux steam with or without non-condensable gas is released into an IRWST, the conditions for mitigation of the chugging phenomenon were identified, and the physical variables were quantified by the 3D reconstruction method. The local behavior of the non-condensable gas was measured after condensation inside heat exchanger using a traverse system. Stratification of non-condensable gas occurred in large tank of the natural circulation loop. SISTA2 was used to simulate a small break loss-of-coolant accident (SBLCOA) transient. Since the test apparatus was a metal tank, compensations of initial heat transfer to the material and effect of heat loss during long-term operation were important for simulating cooling performance of SMART100 CPRSS. The pressure of SMART100 CPRSS was maintained below the design limit for 3 days even under sufficiently conservative conditions of an SBLOCA transient.

• Nuclear Engineering and Technology
• /
• v.42 no.6
• /
• pp.656-661
• /
• 2010

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER$^2$) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.