• 한국기계가공학회지
• /
• 제16권6호
• /
• pp.109-116
• /
• 2017

In this study, the optimization of the overall dimensions of an automobile bumper was investigated through CAE and experiment using the Six Sigma method and design of experiment (DOE) method, respectively. Injection pressure, injection speed, injection time, cooling time, holding time, injection temperature, and holding pressure were selected as the vital parameters affecting the overall width of product through analysis of trivial many using CAE. The optimal values were determined using the DOE method, and we analyzed the improvement by applying the optimal conditions to the production process. As a result, the mean value of the overall width was close to the target value, with a deviation of 0.05mm, and the processability and I-MR control were remarkably improved. Finally, the dimension pass rate of the product improved by 20%.

• Nuclear Engineering and Technology
• /
• 제37권6호
• /
• pp.587-594
• /
• 2005

In an effort to develop a thermal-hydraulic (TH) safety analysis code for Gas-cooled Reactors (GCRs), the MARS code, which was primarily developed for TH analysis of water reactor systems, has been extended here for application to GCRs. The modeling requirements of the system code were derived from a review of major processes and phenomena that are expected to occur during normal and accident conditions of GCRs. Models fur code improvement were then identified through a review of existing MARS code capability. Among these, the following priority models necessary fur the analysis of limiting high and low pressure conduction cooling events were evaluated and incorporated in MARS-GCR/V1 : 1) Helium (He) and Carbon Dioxide ($CO_2$) as main system fluids, 2) gas convection heat transfer, 3) radiation heat transfer, and 4) contact heat transfer models. Each model has been assessed using various conceptual problems for code-to-code benchmarks and it was demonstrated that MARS-GCR/V1 is capable of capturing the relevant phenomena. This paper describes the models implemented in MARS-GCR/V1 and their verification and validation results.

• KIEE International Transactions on Electrophysics and Applications
• /
• 제5C권6호
• /
• pp.233-241
• /
• 2005

The present model of a SF6 arc accounts for the copper vapour contamination from the electrodes inside a Laval nozzle of a circuit breaker. Steady state simulations have been done for the arc with electrode gap of 60mm and DC electric current of 500A, 1000A and 1500A for both cases with and without copper contamination. The effects of electrode polarity are considered for the arc current of 1000A. It was found out that evaporation of copper from the anode results in a cooling of the arc in a region close to the electrodes. The electrical potential across the electrodes is not sensitive to the presence of copper vapour, typically less than $4\%$ difference. Transient analysis has been done in order to obtain the arc properties at current zero. The arc current is increased linearly from -1000 to 0A when the upstream electrode is cathode with constant dI/dt of $27.0A/{\mu}s$ (or decreased linearly from 1000 to 0A when upstream electrode is anode). It has been predicted that the presence of copper vapour reduces the interruption capability of the breaker.

• Nuclear Engineering and Technology
• /
• 제30권6호
• /
• pp.531-540
• /
• 1998

The DYLAM-3 code which overcomes the limitation of event tree/fault tree was applied to LOOP (Loss of Off-site Power) in the mid-loop operation employing HEPs (Human Error Probabilities) supplied by the ASEP (Accident Sequence Evaluation Program) and the SEPLOT (Systematic Evaluation Procedure for Low power/shutdown Operation Task) procedure in this study. Thus the time history of core uncovery frequency during the mid-loop operation was obtained. The sensitivity calculations in the operator's actions to prevent core uncovery under LOOP in the mid-loop operation were carried out. The analysis using the time dependent HEP was performed on the primary feed & bleed which has the most significant effect on core uncovery frequency. As the result, the increment of frequency is shown after 200 minutes duration of simulation conditions. This signifies the possibility of increment in risk after 200 minutes. The primary feed & bleed showed the greatest impact on core uncovery frequency and the recovery of the SCS (Shutdown Cooling System) showed the least impact. Therefore the efforts should be taken on the primary feed & bleed to reduce the core uncovery frequency in the mid-loop operation. And the capability of DYLAM-3 in applying to the time dependent concerns could be demonstrated.

• Nuclear Engineering and Technology
• /
• 제45권4호
• /
• pp.469-476
• /
• 2013

In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the calandria vault water which acts as the ultimate heat sink. However, the questions remain: how long would the vault water be an ultimate heat sink, and what would be the failure mode of the calandria vessel if the heat sink capability of the reactor vault water is lost? In the present study, a numerical analysis is performed to evaluate the thermal loads and the stresses in the calandria vessel following the above accident scenario. The heat transfer from the molten corium pool to the surrounding is assumed to be by a combination of radiation, conduction, and convection from the calandria vessel wall to the vault water. From the temperature distribution in the vessel wall, the transient thermal loads have been evaluated. The strain rate and the vessel failure have been evaluated for the above scenario.

• Smart Structures and Systems
• /
• 제24권1호
• /
• pp.111-125
• /
• 2019

Metallic shape memory alloys present fascinating physical properties such as their super-elastic behavior in austenite phase, which can be exploited for providing a structure with both a self-centering capability and an increased ductility. More or less accurate numerical models have been introduced to model their behavior along the last 25 years. This is the reason for which the literature is rich of suggestions/proposals on how to implement this material in devices for passive and semi-active control. Nevertheless, the thermo-mechanical coupling characterizing the first-order martensite phase transformation process results in several macroscopic features affecting the alloy performance. In particular, the effects of day-night and winter-summer temperature excursions require special attention. This aspect might imply that the deployment of some devices should be restricted to indoor solutions. A further aspect is the dependence of the behavior from the geometry one adopts. Two fundamental lacks of symmetry should also be carefully considered when implementing a SMA-based application: the behavior in tension is different from that in compression, and the heating is easy and fast whereas the cooling is not. This manuscript focuses on the passive devices recently proposed in the literature for civil engineering applications. Based on the challenges above identified, their actual feasibility is investigated in detail and their long term performance is discussed with reference to their fatigue life. A few available semi-active solutions are also considered.

• Nuclear Engineering and Technology
• /
• 제51권5호
• /
• pp.1251-1260
• /
• 2019

The water-cooled WCLL blanket is one of the possible candidates for the blanket of the fusion power reactors. The plasma-facing first wall manufactured from the reduced-activation ferritic-martensitic steel Eurofer97 will be cooled with water at a typical pressurized water reactor (PWR) conditions. According to new estimates, the first wall will be exposed to peak heat fluxes up to $7MW/m^2$ while the maximum operated temperature of Eurofer97 is set to $550^{\circ}C$. The performed analysis shows the capability of the designed flat first wall concept to remove heat flux without exceeding the maximum Eurofer97 operating temperature only up to $0.75MW/m^2$. Several heat transfer enhancement methods (turbulator promoters), structural modifications, and variations of parameters were analysed. The effects of particular modifications on the wall temperature were evaluated using thermo-hydraulic three-dimensional numerical simulation. The analysis shows the negligible effect of the turbulators. By the combination of the proposed modifications, the permitted heat flux was increased up to $1.69MW/m^2$ only. The results indicate the necessity of the re-evaluation of the existing first wall concepts.

• 한국기계가공학회지
• /
• 제17권6호
• /
• pp.98-104
• /
• 2018

Deflected 4-way panels of ceiling air conditioners produced by injection molding process have caused dew condensation at the edge of products. In order to prevent this drawback with reducing weight and deformation, this study proposed renovated process adopting microcellular foaming. According to results from 2-sample t-test and analysis of variance(ANOVA), the critical factors affecting weight were melt temperature and injection speed. In addition, the vital effects on deformation were structure at the edge, mold temperature and cooling time. Optimal conditions of these parameters were derived by regressive analysis with CAE and response surface method(RSM), and then applied to an actual design and process stage to analyze performance. As a results, it clearly showed that new process improved process capability as well as reduced both weight and deformation by 18.8% and 71.9% respectively compared to the conventional method.

• 한국추진공학회지
• /
• 제26권6호
• /
• pp.10-20
• /
• 2022

본 연구에서는 케로신 다단연소사이클 엔진용 예연소기를 설계하기 위해, 고압의 산화제 과잉 조건에서 예연소가스를 계산하고 냉각유로에서 극저온 유체의 복합열전달 및 수력 특성을 해석할 수 있는 설계코드를 개발하였다. 사용자 편의성과 범용성을 가진 오픈 소스 라이브러리 Cantera를 활용하였으며, 실제유체의 열역학/전달 상태량을 정확히 계산하기 위해 관련 소스 코드들을 새로 작성하여 Cantera에 추가하였다. 현재 예비설계 중인 100톤급 부스터 엔진용 예연소기에 적용하였으며, CFD 해석결과와 비교를 통해 설계코드로서의 예측 정확도와 활용성을 확인하였다.

• Nuclear Engineering and Technology
• /
• 제55권5호
• /
• pp.1911-1923
• /
• 2023

The nuclear safety assessment involving large transient simulations is forcing the community to develop methods for coupling thermal-hydraulics and neutronic codes and three-dimensional (3D) Computational Fluid Dynamics (CFD) codes. In this paper a set of dynamic boundary conditions are implemented in OpenFOAM^Ⓡ in order to apply zero-dimensional (0D) approaches coupling with 3D thermal-hydraulic simulation in a single framework. This boundary conditions are applied to model pipelines, tanks, pumps, and heat exchangers. On a first stage, four tests are perform in order to assess the implementations. The results are compared with experimental data, full 3D CFD, and system code simulations, finding a general good agreement. The semi-implicit implementation nature of these boundary conditions has shown robustness and accuracy for large time steps. Finally, an application case, consisting of a simplified open pool with a cooling external circuit is solved to remark the capability of the tool to simulate thermal hydraulic systems commonly found in nuclear installations.