• Nuclear Engineering and Technology
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• v.45 no.2
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• pp.191-202
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• 2013

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy may be due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

• KEPCO Journal on Electric Power and Energy
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• v.2 no.3
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• pp.433-436
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• 2016

In the area of environmental chemistry of power plant, flow analysis of the reactor with built-in impeller is a very important part from the perspective of the improvement of the efficiency of the entire process. As a wide range of methods are being proposed for the analysis of the flow pattern within the reactor, this study analyzed the flow within the reactor according to the baffle structure (height) installed on the internal wall of the reactor in order to improve the reaction efficiency through the inducing of the up and down stirring with the reactor. As the results of the execution of the flow analysis for each of a diverse range of cases by utilizing the Computational Fluid Dynamics (CFD) method, it was possible to confirm that the flow is markdely improved by inducing the up and down stirring among the reactants within the reactor if the baffle is elevated to the level below the water surface. In particular, as the results of the analysis of the general cases in which the baffle is elevated all 4 steps and the cases in which the baffle is elevated only 2 steps, elevating the baffle only 2 steps achieve the same effect as the elevating of the baffle by 4 steps. Therefore, it was possible to expect to improve the efficiency with out the need to increase the use of electric power substantially if the outcomes of this study is applied to the actual sites of power plants in the future.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.243-249
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• 2001

In this study, the progressive inelastic deformation, so called, thermal ratchet phenomenon which can occur in high temperature liquid metal reactor was simulated with thermal ratchet structural test facility and 316L stainless steel test cylinder. The inelastic deformation of the reactor baffle cylinder can occur due to the moving temperature distribution along the axial direction as the hot free surface moves up and down under the cyclic heat-up and cool-down of reactor operations. The ratchet deformations were measured with the laser displacement sensor and LVDTs after cooling the structural specimen which experiences thermal load up to $550^{\circ}$ and the temperature differences of about $500^{\circ}C$. During structural thermal ratchet test, the temperature distribution of the test cylinder along the axial direction was measured from 28 channels of thermocouples and the temperatures were used for the ratchet analysis. The thermal ratchet deformation analysis was performed with the NONSTA code whose constitutive model is nonlinear combined kinematic and isotropic hardening model and the test results were compared with those of the analysis. Thermal ratchet test was carried out with respect to 9 cycles of thermal loading and the maximum residual displacements were measured to be 1.8mm. It was shown that thermal ratchet load can cause a progressive deformation to the reactor structure. The analysis results with the combined hardening model were in reasonable agreement with those of the tests.

• Structural Engineering and Mechanics
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• v.39 no.1
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• pp.125-145
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• 2011

Pushover analysis has gained significant popularity as an analytical tool for realistic determination of the inelastic behaviour of RC structures. Though significant work has been done to evaluate the demands realistically, the evaluation of capacity and realistic failure modes has taken a back seat. In order to throw light on the inelastic behaviour and capacity evaluation for the RC framed structures, a 3D Reinforced concrete frame structure was tested under monotonically increasing lateral pushover loads, in a parabolic pattern, till failure. The structure consisted of three storeys and had 2 bays along the two orthogonal directions. The structure was gradually pushed in small increments of load and the corresponding displacements were monitored continuously, leading to a pushover curve for the structure as a result of the test along with other relevant information such as strains on reinforcement bars at critical locations, failure modes etc. The major failure modes were observed as flexural failure of beams and columns, torsional failure of transverse beams and joint shear failure. The analysis of the structure was by considering all these failure modes. In order to have a comparison, the analysis was performed as three different cases. In one case, only the flexural hinges were modelled for critical locations in beams and columns; in second the torsional hinges for transverse beams were included in the analysis and in the third case, joint shear hinges were also included in the analysis. It is shown that modelling and capturing all the failure modes is practically possible and such an analysis can provide the realistic insight into the behaviour of the structure.

• Journal of Industrial Technology
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• v.21 no.B
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• pp.141-147
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• 2001

For the preparation of calcium carbonate particles from aqueous $Ca(OH)_2$ slurry, carbonation reaction of aqueous $Ca(OH)_2$ slurry was carried out by batch method the $CO_2$ into reactor filled with aqueous slurry of $Ca(OH)_2$. The concentration of $Ca(OH)_2$ varies from 1.00 to 7.00wt%, reactor temperature at 20 and $40^{\circ}C$, and reactor pressure from atmospheric pressure to $6.0kg_f/cm^2$. Crystal structure of calcium carbonate was of calcite, the particle size were about $0.05{\sim}2.0{\mu}m$, and the particle shape was cubic and spindle. When reactor temperature was higher, particle size of calcium carbonate was bigger and particle shape was varied, but reaction rate was increased. When reactor pressure was higher, particle size of calcium carbonate was smaller, particle shape was cubic, and reaction rate was increased.

• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.408-422
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• 1999

To reduce the fast neutron fluence which deteriorates the RPV integrity, additional shields were assumed to be installed at the outer core structures of the Kori Unit 1 reactor, and its reduction effects were examined. Full scope Monte Carlo simulation with MCNP4A code was made to estimate the fast neutron fluence at the RPV. An optimized design option was found from various choices in geometry and material for shield structure. It was expected that magnitude of fast neutron fluence would be reduced by 39% at the circumferential weld of the RPV, resulting in extension of plant lifetime by 4.6 EFPYs based on the criterion of PTS requirement It was investigated that the nuclear characteristics and thermal hydraulic factors at the internal core were only negligibly influenced by the installation of additional shield structure.

• Proceedings of the SAREK Conference
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• 2003.07a
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• pp.9-9
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• 2003

• Transactions of the KSME C: Technology and Education
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• v.5 no.2
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• pp.127-134
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• 2017

The reactor internals shall be secured in safety and structural integrity under various vibrational loading conditions. Thus, U.S. NRC, Regulatory Guide 1.20 requires the evaluation for the reactor internals due to acoustic induced vibration including the response to the reactor coolant pump pressure pulsation. This paper suggests a methodology to develop an analytical model of the core support barrel accounting for the fluid around the structure and to analyze the responses to the pump pulsation loads using acoustic structure interaction analysis. The analysis results were compared with those of US Palo Verde 1 CVAP and showed a good agreement. Thus, it is expected that the suggested methodology could be an efficient way to evaluate the response of the core support barrel to the pump pulsation loads.

• Journal of Industrial Technology
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• v.22 no.A
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• pp.255-260
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• 2002

We prepared the nano-sized $TiO_2$ particles by the diffusion flame reactor and investigated the effects of several process variables on the generation and transport properties of $TiO_2$ particle. As the length from the tip of diffusion flame reactor increases, the size of $TiO_2$ particle increases by the coagulation between particles. The structure of $TiO_2$ particles prepared is almost found to be anatase. It was found that the $TiO_2$ particle size depends more largely on the change of reactor temperature than on the change of inlet $TiCl_4$ concentration.

• 제어로봇시스템학회:학술대회논문집
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• 2001.10a
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• pp.133.6-133
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• 2001

Reactor vessel is one of the most important equipment of Nuclear Power Plant (NPP) with regard to the nuclear safety. Thus reactor vessel must be examined periodically by certified experts. Currently, ultra-sonic(UT) non-destructive inspection is executed on reactor vessel. Two different techniques are used in this inspection. One is using the movable manipulator fixed with the support-guide placed on the vessel, and the other is using mobile robot moving in the vessel. Movable manipulator machine is very heavy, hard to handle, and very expensive. Mobile robot equipment is small and convenient but has a weak point on positional precision. To solve these problems we developed a reactor inspection system based on laser-driven mobile robot. This paper describes the main concept and structure of integrated inspection units and the feature of implemented units.