• Nuclear Engineering and Technology
• /
• v.41 no.9
• /
• pp.1191-1204
• /
• 2009

In this study, a CFD methodology is employed to address the problem of the prediction of the flow in a T-junction. An Unsteady Reynolds Averaged Navier-Stokes (URANS) approach has been selected for its low computational cost. Moreover, Unsteady Reynolds Navier-Stokes methodologies do not need complex boundary formulations for the inlet and the outlet such as those required when using Large Eddy Simulation (LES) or Direct Numerical Simulation (DNS). The results are compared with experimental data and an LES calculation. In the past, URANS has been tried on T-junctions with mixed results. The biggest limit observed was the underestimation of the oscillatory behavior of the temperature. In the present work, we propose a comprehensive approach able to correctly reproduce the root mean square (RMS) of the temperature directly downstream of the T-junction for cases where buoyancy is not present.

• Korean Journal of Agricultural Science
• /
• v.43 no.1
• /
• pp.116-126
• /
• 2016

The objective of this study is to predict the fuel efficiency of an agricultural tractor. The fuel efficiency of the tractor during rotary tillage was predicted using numerical modeling. A numerical model was developed using Simulation X. Based on tractor power flow, numerical modeling consisted of an engine, transmission, PTO (power take off), and hydraulics. The specifications of major components utilized in the numerical model were the same as those of a 71 kW tractor (field test tractor). The load that was inputted for fuel efficiency prediction into the simulation model was obtained from a field test. Fuel efficiency predictions were conducted by comparing field test results and simulation results. In addition, it was performed by dividing the rotary tillage and steering section. Main results are as follows: first, t-values of engine torque were measured to be 0.31 in the rotary tillage and 0.92 in the steering section. Second, t-values of fuel consumption were measured to be 0.51 and 5.41 in the rotary tillage and the steering section, respectively. Finally, t-values of fuel efficiency were measured to be 1.72 and 40 in the rotary tillage and the steering section, respectively. The results show no significant differences with t-values of less than 5% in the rotary tillage. But, it shows significant differences in the steering section. Therefore, simulation for accurate fuel efficiency prediction requires a suitable algorithm or detailed design of the simulation model in the steering section.

• Geomechanics and Engineering
• /
• v.21 no.5
• /
• pp.401-411
• /
• 2020

The paper describes the influence of pile reinforcement on the stability of the slope behaviour, and the exploitation of the results of in situ measurements will be conducted. In the second part, a 2D numerical modelling will be conducted by using the finite element code PLAXIS2D; in order to validate the proposed modelling approach by comparing the numerical results with the measurements results carried out on the slides studied; to study the effect of positioning of piles as a function of the shear parameters of the supported soil on the behaviour of the soil. For various shear strength of the soil a row of pile position is found, at which the piles offer the maximum contribution to slope stability. The position of piles is found to influence the safety factor in granular soil whereas it shows a slight influence on the safety factor in coherent soil. The results also indicate that the ideal position for such stabilizing piles is in the middle height of the slope. Comparison of results of present study with literature from publication: indicated that to reach the maximum stability of slope, the pile must be installed with L_x/L ratio (0.37 to 0.62) and the inclination must be between 30° to 60°. Even, after a certain length of the pile, the increasing will be useless. The application of the present approach to such a problem is located at the section of PK 210+480 to 210+800 of the Algerian East-West Highway.

• Journal of Energy Engineering
• /
• v.17 no.4
• /
• pp.218-226
• /
• 2008

A component-scale two-phase analysis code has been developed for a realistic simulation of two-phase flow transients in a light water nuclear reactor component. In the code, a two-fluid three-field model is adopted and the governing equations are solved on an unstructured mesh. For the numerical solution scheme, the semi-implicit method used in the RELAP5 code was selected, which has been proved to be very stable and accurate for most of practical applications. However, some modifications were needed for its application to an unstructured non-staggered grid. This paper presents the modified semi-implicit numerical method for unstructured grid and the preliminary results of the calculations.

• Nuclear Engineering and Technology
• /
• v.53 no.5
• /
• pp.1540-1555
• /
• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

• Proceedings of the Korea Water Resources Association Conference
• /
• 2005.05b
• /
• pp.414-418
• /
• 2005

A finite element model is developed for the numerical simulation of bed elevation change in a curved channel. The SU/PG (Streamline-Upwind/Petrov-Galerkin) method is used to solve 2D shallow water equations and the BG (Bubnov-Galerkin) method is used for the Exner equation. For the time derivative terms, the Crank-Nicolson scheme is used. The developed model is a decoupled model in a sense that the bed elevation does not change simultaneously with the flow during the computational time step. The total load formula with is used for the sediment transport model. The slip conditions are described along the lateral boundaries. The effects of gravity force due to geometry change and the secondary flows in a curved channel are considered in the model. For the verification, the model is applied to two laboratory experiments. The first is $140^{\circ}$ bended channel data at Delft Hydraulics Laboratory and the second is $140^{\circ}$ bended channel data at Laboratory of Fluid Mechanics of the Delft University of Technology. The finite element grid is constructed with linear quadrilateral elements. It is found that the computed results are in good agreement with measured data, showing a point bar at the inner bank and a pool at the outer bank.

• Nuclear Engineering and Technology
• /
• v.38 no.1
• /
• pp.11-32
• /
• 2006

System thermal-hydraulic codes have been used in the past decades in the areas of design, operation, licensing and safety of Nuclear Power Plants (NPPs). The development and validation of these codes have reached a high degree of maturity, through the consideration of huge experiments and advanced numerical models. Nowadays, the analyses are based upon realistic approaches rather than the conservative evaluation models. However the applications of these computational tools require preliminary qualification issues. Although huge amounts of financial and human resources have been invested for the development and improvement of codes, the calculation results are still affected by errors. In the sophisticated nuclear technology, design and safety of NPP, these errors must be quantified. An overview of the state of the art of the current thermal-hydraulic system code is developed and the need of uncertainty analysis in code calculations is emphasized. Several sources of uncertainty have been classified and commented, and typical applications of such methods are shown.

• Nuclear Engineering and Technology
• /
• v.41 no.3
• /
• pp.287-294
• /
• 2009

Due to thermal hydraulics phenomena, such as thermal stratification, various events occur to the parts of a nuclear power plant during their lifetimes: e.g. cracked and dislocated pipes and thermally fatigued, bent, and damaged supports. Due to the operational characteristics of the parts of the steam generator feedwater inlet horizontal pipe, thermal stratification takes place particularly frequently. However, the thermal stress due to thermal stratification at the steam generator feedwater inlet horizontal pipe was not reflected in the design stage of old plants(Kori Unit No.1, 2, 3 and 4, Yeonggwang Unit No. 1 and 2, and Uljin Unit No. 1 and 2; referred to as old-style power plants hereinafter). Accordingly, a verification experiment was performed for thermal stratification in the horizontal inlet nozzle steam generator of old-style plants. If thermal stratification occurred in the horizontal pipe of an old-style power plant, numerical analysis of the temperature distribution of the pipes and fluids was conducted. The temperature distributions were compared at the curved part of the pipe and the horizontal pipe before and after the installation of the improved thermal sleeves designed to alleviate thermal stress due to thermal stratification. The thermal stress reduction measure was proven effective at the steam generator inlet horizontal pipe and the curved part of the pipe.

• Nuclear Engineering and Technology
• /
• v.41 no.7
• /
• pp.885-892
• /
• 2009

This manuscript will discuss a numerical method where the six equations of two-phase flow, the solid heat conduction equations, and the two equations that describe neutron diffusion and precursor concentration are solved together in a tightly coupled, nonlinear fashion for a simplified model of a nuclear reactor core. This approach has two important advantages. The first advantage is a higher level of accuracy. Because the equations are solved together in a single nonlinear system, the solution is more accurate than the traditional "operator split" approach where the two-phase flow equations are solved first, the heat conduction is solved second and the neutron diffusion is solved third, limiting the temporal accuracy to $1^{st}$ order because the nonlinear coupling between the physics is handled explicitly. The second advantage of the method described in this manuscript is that the time step control in the fully implicit system can be based on the timescale of the solution rather than a stability-based time step restriction like the material Courant limit required of operator-split methods. In this work, a pilot code was used which employs this tightly coupled, fully implicit method to simulate a reactor core. Results are presented from a simulated control rod movement which show $2^{nd}$ order accuracy in time. Also described in this paper is a simulated rod ejection demonstrating how the fastest timescale of the problem can change between the state variables of neutronics, conduction and two-phase flow during the course of a transient.

• Nuclear Engineering and Technology
• /
• v.42 no.1
• /
• pp.47-54
• /
• 2010

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/$cm^3$. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.