• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.36 no.1
• /
• pp.117-124
• /
• 2012

In this paper, an analysis of the joint that transmits power from the blades to the generator is performed using the FEM (finite element method). The mode shapes and natural frequencies were extracted using experimental modal analysis in order to establish the FEM model. Then, the model was verified by comparing the mode shapes and natural frequencies to those obtained from the ANSYS modal analysis. Dynamic stress analysis was performed at the rated and limited wind speeds considering the wind load and gravity.

• Transactions of the Korean Society for Noise and Vibration Engineering
• /
• v.20 no.3
• /
• pp.215-222
• /
• 2010

Long and slender body, like a fuel rod, oscillating in axial flow can be unstabilized even by the small cross flow which can be activated by the flow mixer or turbulent generator. It is important to include these effects of flow disturbance in dynamic stability analysis of nuclear fuel rod. This work shows how eigen frequency of a multi-span fuel rod can be changed by the swirl flow, which is discretely generated by a flow mixer. By solving a state-space form of the eigenvalue equation for a multi-span fuel rod system, the critical velocity at which a fuel rod becomes unstable was calculated. Based on the simulation results, we evaluated how stability of a multi-spanned nuclear fuel rod with mixing vanes can be affected by the coolant flow in an operating reactor core.

• Nuclear Engineering and Technology
• /
• v.45 no.4
• /
• pp.439-458
• /
• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.

• Nuclear Engineering and Technology
• /
• v.49 no.5
• /
• pp.928-940
• /
• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• KSII Transactions on Internet and Information Systems (TIIS)
• /
• v.13 no.4
• /
• pp.1795-1811
• /
• 2019

In order to achieve rapid and accurate detection of vehicle objects in complex traffic conditions, we propose a novel vehicle detection method. Firstly, more contextual and small-object vehicle information can be obtained by our Joint Feature Network (JFN). Secondly, our Evolved Region Proposal Network (EPRN) generates initial anchor boxes by adding an improved version of the region proposal network in this network, and at the same time filters out a large number of false vehicle boxes by soft-Non Maximum Suppression (NMS). Then, our Mask Network (MaskN) generates an example that includes the vehicle occlusion, the generator and discriminator can learn from each other in order to further improve the vehicle object detection capability. Finally, these candidate vehicle detection boxes are optimized to obtain the final vehicle detection boxes by the Fine-Tuning Network(FTN). Through the evaluation experiment on the DETRAC benchmark dataset, we find that in terms of mAP, our method exceeds Faster-RCNN by 11.15%, YOLO by 11.88%, and EB by 1.64%. Besides, our algorithm also has achieved top2 comaring with MS-CNN, YOLO-v3, RefineNet, RetinaNet, Faster-rcnn, DSSD and YOLO-v2 of vehicle category in KITTI dataset.

• Nuclear Engineering and Technology
• /
• v.51 no.3
• /
• pp.723-730
• /
• 2019

Acquiring instrumentation signals generated from nuclear power plants (NPPs) is essential to maintain nuclear reactor integrity or to mitigate an abnormal state under normal operating conditions or severe accident circumstances. However, various safety-critical instrumentation signals from NPPs cannot be accurately measured on account of instrument degradation or failure under severe accident circumstances. Reactor vessel (RV) water level, which is an accident monitoring variable directly related to reactor cooling and prevention of core exposure, was predicted by applying a few signals to deep neural networks (DNNs) during severe accidents in NPPs. Signal data were obtained by simulating the postulated loss-of-coolant accidents at hot- and cold-legs, and steam generator tube rupture using modular accident analysis program code as actual NPP accidents rarely happen. To optimize the DNN model for RV water level prediction, a genetic algorithm was used to select the numbers of hidden layers and nodes. The proposed DNN model had a small root mean square error for RV water level prediction, and performed better than the cascaded fuzzy neural network model of the previous study. Consequently, the DNN model is considered to perform well enough to provide supporting information on the RV water level to operators.

• Journal of the Korean Society of Industry Convergence
• /
• v.21 no.6
• /
• pp.301-307
• /
• 2018

This paper deals with the development of the high frequency thawing machine. The fishery products caught over the world are kept frozen to maintain freshness. These fishery products require thawing before they are sold to customers as food. However, the thawing process can cause freshness reduction, drip coming out, quality deterioration, discharging polluted water, as well as a lot of space and time. The high frequency thawing machine developed to solve this problem has a narrow space, a short thawing time and a small drip. The developed high frequency thawing machine can be used in many fields such as fish processing plant, livestock processing plant. This paper describes the design of the high frequency thawing machine by developing the high frequency generator, development of the controller, and the design of mechanism, and shows the superiority of the high frequency thawing machine by the performance evaluation.

• Journal of Power Electronics
• /
• v.19 no.2
• /
• pp.580-590
• /
• 2019

The concept of virtual synchronous generators (VSGs) is a valuable means for improving the frequency stability of microgrids (MGs). However, a great virtual inertia in a VSG's controller may cause power oscillation, thereby deteriorating system stability. In this study, a small-signal model of an MG with two paralleled VSGs is established, and a control strategy for maintaining a constant inertial time with an increasing active-frequency droop coefficient (m) is proposed on the basis of a root locus analysis. The power oscillation is suppressed by adjusting virtual synchronous reactance, damping coefficient, and load frequency coefficient under the same inertial time constant. In addition, the dynamic load distribution is sensitive to the controller parameters, especially under the parallel operation of VSGs with different capacities. Therefore, an active power increment method is introduced to improve the precision of active power sharing in dynamic response. Simulation and experimental is used to verify the theoretical analysis findings.

• Nuclear Engineering and Technology
• /
• v.51 no.2
• /
• pp.402-409
• /
• 2019

When a loss of coolant accident (LOCA) occurs in a nuclear power plant, accident scenarios which can prevent core damage are defined based on break size. Current probabilistic safety assessment evaluates that core damage can be prevented under small-break LOCA (SBLOCA) and steam generator tube rupture (SGTR) with rapid cool down (RCD) strategy when all safety injection systems are unavailable. However, previous research has pointed out a limitation of RCD in terms of initiation time. Therefore, RCD success criteria estimation based on allowable coping time under a SBLOCA or SGTR when all safety injection systems are unavailable was performed based on time-line and thermal-hydraulic analyses. The time line analysis assumed a single emergency operating procedure flow, and the thermal hydraulic analysis utilized MARS-KS code with variables of break size, cooling rate, and operator allowable time. Results show while RCD is possible under SGTR, it is impossible under SBLOCA at the APR1400's current cooling rate limitation of 55 K/hr. A success criteria map for RCD under SBLOCA is suggested without cooling rate limitation.

• Nuclear Engineering and Technology
• /
• v.51 no.1
• /
• pp.310-316
• /
• 2019

Fluidelastic instability is a key issue in steam generator tube bundles subjected in cross-flow. With a low flow velocity, a large amplitude vibration of the tube observed by many researchers. However, the mechanism of this vibration is seldom analyzed. In this paper, the mechanism of cross-flow effects on fluidelastic instability of tube bundles was investigated. Analysis reveals that when the system reaches the critical state, there would be two forms, with two critical velocities, and thus two expressions for the critical velocities were obtained. Fluidelastic instability experiment and numerical analysis were conducted to obtain the critical velocity. And, if system damping is small, with increases of the flow velocity, the stability behavior of tube array changes. At a certain flow velocity, the stability of tube array reaches the first critical state, a dynamic bifurcation occurs. The tube array returns to a stable state with continues to increase the flow velocity. At another certain flow velocity, the stability of tube array reaches the second critical state, another dynamic bifurcation occurs. However, if system damping is big, there is only one critical state with increases the flow velocity. Compared the results of experiments to numerical analysis, it shows a good agreement.