• Nuclear Engineering and Technology
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• 제49권4호
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• pp.675-691
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• 2017

Jets in cross flow are of fundamental industrial importance and play an important role in validating turbulence models. Two jet configurations related to thermal fatigue phenomena are investigated: ${\bullet}$ T-junction of circular tubes where a heated jet discharges into a cold main flow and ${\bullet}$ Rectangular jet marked by a scalar discharging into a main flow in a rectangular channel. The T-junction configuration is a classical test case for thermal fatigue phenomena. The Vattenfall T-junction experiment was already subject of an OECD/NEA benchmark. A LES modelling and calculation strategy is developed and validated on this data. The rectangular-jet configuration is important for basic physical understanding and modelling and has been analyzed experimentally at CEA. The experimental work was focused on turbulent mixing between a slightly heated rectangular jet which is injected perpendicularly into the cold main flow of a rectangular channel. These experiments are analyzed for the first time with LES. The overall results show a good agreement between the experimental data and the CFD calculation. Mean values of velocity and temperature are well captured by both RANS calculation and LES. The range of critical frequencies and their amplitudes, however, are only captured by LES.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3754-3767
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• 2023

Turbulent jet discharges in subcooled water pools are essential for safety systems in nuclear power plants, specifically in the pressure suppression pool of boiling water reactors and In-containment Refueling Water Storage Tank of advanced pressurized water reactors. The gas and liquid flow in these systems is investigated using multiphase flow analysis. This field has been extensively examined using a combination of experiments, theoretical models, and Computational Fluid Dynamics (CFD) simulations. ANSYS CFX offers two approaches to model multiphase flow behavior. The non-homogeneous Eulerian-Eulerian Model has been used in this work; it computes global information and is more convenient to study interpenetrated fluids. This study utilized the Large Eddy Simulation Model as the turbulence model, as it is better suited for non-stationary and buoyant flows. The CFD results of this study were validated with experimental data and theoretical results previously obtained. The figures of merit dimensionless penetration length and the dimensionless buoyancy length show good agreement with the experimental measurements. Correlations for these variables were obtained as a function of dimensionless numbers to give generality using only initial boundary conditions. CFD numerical model developed in this research has the capability to simulate the behavior of non-condensable gases discharged in water.

• Nuclear Engineering and Technology
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• 제52권5호
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• pp.897-907
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• 2020

The flow characteristics of reactor fuel assembly always intrigue the designers and the experimentalists among the myriad phenomena that occur simultaneously in a nuclear core. In this work, the visual experimental method has been developed on the basis of refraction index matching (RIM) and particle image velocimetry (PIV) techniques to investigate the detailed flow characteristics in China fast reactor fuel subassembly. A 7-rod bundle of simulated fuel subassembly was fabricated for fine examination of flow characteristics in different subchannels. The experiments were performed at condition of Re=6500 (axial bulk velocity 1.6 m/s) and the fluid medium was maintained at 30℃ and 1.0 bar during operation. As for results, axial and lateral flow features were observed. It is shown that the spiral wire has an inhibitory effect on axial flow and significant intensity of lateral flow mixing effect is induced by the wire. The root mean square (RMS) of lateral velocity fluctuation was acquired after data processing, which indicates the strong turbulence characteristics in different flow subchannels.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.299-312
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• 2007

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.719-726
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• 2012

The object of this work is the validation and assessment of the TRACE v5.0 code using the scaled test ATLAS1 facility in the context of a DVI2 line break. In particular, the experiment selected models the 50%, 6-inch break of a DVI line. The same experiment was also adopted as a reference test in the ISP-503. The ISP-50 was proposed to, and accepted by, the OECD/NEA/CSNI due to its technical importance in the development of a best-estimate of safety analysis methodology for DVI line break accidents. In particular, the behavior of the two-phase flow in the upper annulus downcomer was expected to be complicated. What resulted was the need for relevant models to be implemented into safety analysis codes, in order to predict these thermal hydraulic phenomena correctly.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.1020-1028
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• 2021

The nuclear reactor coolant pump transferring heat energy inherently brings with it the unsteady flow and inevitably threatens to the safe operation of the pump unit, especially with the pressure pulsation induced by the rotor-stator interaction. In this paper, the characteristics of pressure pulsation of the diffuser in a nuclear reactor coolant pump were investigated by the numerical simulation with experimental validation. Pressure pulsation signals measured synchronously from sensors mounted on the radial diffuser of a model pump were analyzed via Welch's method. Frequency components induced by the rotor-stator interaction can be revealed by the diameter mode analysis method. The pressure pulsation of the diffuser is dominated by the blade passing frequency and its harmonics, which are free from the effect of flow rate and rotational speed while the corresponding amplitudes are easily affected by different operational conditions and measuring positions. The non-uniformity is much more affected by the rotational speed than the flow rate. This research is helpful for further work to reduce the pressure pulsation for the reactor coolant pump.

• 상하수도학회지
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• 제13권3호
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• pp.73-82
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• 1999

In this study, the optimal analysis for pipe network is performed for the combined ideal pipe network system(CASE 1, CASE 2 and CASE 3) which is composed of 25 nodes, 41 elements, and 1 fixed nodal head with evaluating pressure variation distribution of main and branch in grid composed drainage pipe network. The linear analysis technique used as the analysis method in this study, the KYPIPE being used extensively as the linear technique to design and analysis of pipe network is applied. Firstly, in the analysis of pipe network, the CASE 2 and CASE 3 supply same thing(value) in the result of considering the total flow provided each pipeline, but in the general intension in the case of CASE 2, relative width of supply is more large than CASE 1 and CASE 3. Secondly, in the analysis technique of pipe network, CASE 3 is analysed largest as a result of comparing with same heads, and in the order of their size CASE 2 and CASE 1 were determined but the difference doesn't appear to be obvious. Thirdly, as the result of determining main factor, pressure in the design and analysis of net work. CASE 3 is from Node 3 to 25 than CASE 1 and CASE 2 and it is determined in the order of their size, CASE 2 and CASE 1. Finally, in this study, discharge flow distribution is evaluated in the same condition with 3-type CASE in the case of branch position for designing optimal composed drainage pipe network. As the result of that, branch pipe perform. Therefore, it is thought that the efficient and reasonable management of water supply and sewerage design will be possible if it give all our energies to study at the pipe system design in and out of country in the future.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4296-4309
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• 2022

Mesh and turbulence model sensitivity analyses have been performed on computational fluid dynamics simulations executed with Hydra and ANSYS Fluent for a single CANadian Deuterium Uranium (CANDU) 37M nuclear fuel bundle placed within a standard pressure tube. The goal of this work was to perform a methodical analysis to objectively determine an appropriate mesh and to gauge the sensitivity of different turbulence models for CANDU subchannel flow under isothermal conditions. The boundary conditions and material properties are representative of normal operating conditions in a high-powered channel of the Darlington Nuclear Generating Station. Four meshes were generated with ANSYS Workbench Meshing, ranging from 22 to 84 million cells, and analyzed here to determine an appropriate level of mesh resolution and quality. Five turbulence models were compared in the turbulence model sensitivity analysis: standard k - ε, RNG k - ε, realizable k - ε, SST k - ω, and the Reynolds Stress Model. The intent of this work was to gain confidence in mesh generation and turbulence model selection of a single bundle to inform the decision making of subsequent investigations of an entire fuel channel containing a string of twelve bundles.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1998-2008
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• 2020

Solenoid operated valves (SOV) play important roles in industrial process to control the flow of fluids. Solenoid valves can be found in so many industries as well as the nuclear plant. The ability to be able to detect the presence of faults and predicting the remaining useful life (RUL) of the SOV is important in maintenance planning and also prevent unexpected interruptions in the flow of process fluids. This paper proposes a fault diagnosis method for the alternating current (AC) powered SOV. Previous research work have been focused on direct current (DC) powered SOV where the current waveform or vibrations are monitored. There are many features hidden in the AC waveform that require further signal analysis. The analysis of the AC powered SOV waveform was done in the time and frequency domain. A total of sixteen features were obtained and these were used to classify the different operating modes of the SOV by applying a machine learning technique for classification. Also, a deep neural network (DNN) was developed for the prediction of RUL based on the failure modes of the SOV. The results of this paper can be used to improve on the condition based monitoring of the SOV.

• Journal of Advanced Marine Engineering and Technology
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• 제10권2호
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• pp.136-145
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• 1986

The perfect and accurate methods to control the wear are not made clear so far. For this phenomenon only mating surface has been studied. In order to control the wear the essence of it has to be made clear. It is reported that adhesive wear might occure as a result of plastic deformation, the fracture and removal or transfer asperities on close contacting surfaces. On this view point the plastic flow was attempted to compare with fluid or electromagnetic flow. The partial differential equations of equilibrium for the plane strain deformation will make use of the method of characteristics. The characteristic curves or characteristics of the hyperbolic equation coincide with the slip lines by R. Hill's papers. By Hencky's stress equation, it is evident that if P and .phi. are prescribed for a boundary condition then it may be possible to proceed along constant .alpha. and .betha. lines to determine the value of the hydrostatic pressure everywhere in the slip line field net work. A wedge formation mechanism has been considered for an explanation of this matters. The analysis shows that there is a critical value, which depends on the hardness ratio and the shear stress on the interface, for the top angle of asperity is less than this critical value, the asperity can yield plastically despite of being harder than the mating surface.