• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1340-1346
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• 2020

The aim of this paper is to explore the 8th-order Adams-Bashforth-Moulton (ABM8) method in the solution of the point reactor kinetics equations. The numerical experiment considers feedback reactivity by Doppler effects, and insertions of reactivity. The Doppler effects is approximated with an adiabatic nuclear reactor that is a typical approximation. The numerical results were compared and discussed with several solution methods. The CATS method was used as a benchmark method. According with the numerical experiments results, the ABM8 method can be considered as one of the main solution method for changes reactivity relatively large.

• Korean Chemical Engineering Research
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• v.55 no.6
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• pp.874-880
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• 2017

In this study, we performed numerical simulation of the adsorptive desulfurization reactor for a 100 kW fuel cell. Using experimental results and the adsorption kinetics theory, the adsorption rate of sulfur in diesel was estimated and verified by numerical analysis. By analyzing the performance of desulfurization according to reactor size, the optimal reactor size was determined. By maximizing processed diesel amount, optimal diesel flow rate was determined. Regeneration process was also confirmed for the obtained optimal reactor size. The present work will be utilized to design a diesel desulfurization reactor for a fuel cell used in a ship by further process modeling and economic analysis.

• Journal of Korean Society for Atmospheric Environment
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• v.17 no.4
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• pp.347-354
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• 2001

SCR (Selective Catalytic Reduction) process is presently considered as one of the most effective techniques for removing nitric oxides from exhaust gases. In this study, based on the conceptually designed SCR reactor of 500 MW coal fired power plant. a reduced scale (1/20) SCR reactor model was made to analyze the flow pattern in front of catalyst layer according to the guide vane's design factors such as the number, interval, and angle of vanes. The results of the test were compared to those numerical simulation in order to assure the reliability of two methods. On the basis of our study. the critical Reynolds number (2.0$\times$ 10$^{5}$ ) was proposed for ensuring the similarity between the reduced scale model and the prototype of SCR reactor. Optimum design parameters of guide vanes were determined as follows, 4 vanes, the first vane angle of 93$^{\circ}$, and the vane intervals of 0.85 S/n, 1.05 S/n, 1.1 S/n, 1.0S/n, 1.0S/n (S: the distance of duct, n: the number of guide vanes). The excellent agreement between the results of the numerical simulation and the reduced scale model provides the validation of two methods for prediction of flow through SCR reactor.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.31-44
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• 2019

This study simulated a control rod assembly (CRA), which is a part of reactor shutdown systems, in immersed and fluid flow conditions. The CRA was inserted into the reactor core within a predetermined time limit under normal and abnormal operating conditions, and the CRA (which consists of complex geometric shapes) drop behavior is numerically modeled for simulation. A full-scale prototype CRA drop test is established under room temperature and water-fluid conditions for verification and validation. This paper describes the details of the numerical modeling and analysis results of the several conditions. Results from the developed numerical simulation code are compared with the test results to verify the numerical model and developed computer code. The developed code is in very good agreement with the test results and this numerical analysis model and method may replace the experimental and CFD method to predict the drop behavior of CRA.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.13 no.7
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• pp.637-646
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• 2001

VOC(Volatile Organic Compound) removal characteristics in continuous flow reactors have been numerically investigated. The photocatalytic reaction have been simulated with the binding constant and the reaction rate constant obtained from experimental data for the constant-volume batch reactor, and then VOC abatement in continuous flow reactors with the same conditions as those of batch reactor has been analyzed. The standard 4\kappa-\varepsilon$ model and mass conservation equation have been employed for numerical calculation, and heterogeneous reaction rate has been used in terms of the boundary condition of the conservation equation. in the case of the continuous flow reactor, reaction characteristics have been estimated with various inlet velocities and with different number of baffles. The result shows that the concentration distribution and flow patterns are strongly affected by the inlet velocity, and that with the increased inlet velocity, VOC removal rate is increased, while removal efficiency is decreased. This result may be useful in the design of reactors with improved VOC removal efficiency.

• Journal of the Korean Society of Safety
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• v.32 no.2
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• pp.147-152
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• 2017

The external reactor vessel cooling (ERVC) is well known strategy to mitigate a severe accident at which nuclear fuel inside the reactor vessel is molten. In order to compare the heat removal capacity of ERVC between the nuclear reactor designs quantitatively, numerical method is often used. However, the study for ERVC using computational fluid dynamics (CFD) is still quite scarce. As a validation study on the numerical prediction for ERVC using CFD, the subcooled boiling flow and natural circulation of coolant at the ULPU-V experiment was simulated. The commercially available CFD software ANSYS-CFX was used. Shear stress transport (SST) model and RPI model were used for turbulence closure and wall-boiling, respectively. The averaged flow velocities in the downcomer and the baffle entry under the reactor vessel lower plenum are in good agreement with the available experimental data and recent computational results. Steam generated from the heated wall condenses rapidly and coolant flows maintains single-phase flow until coolant boils again by flashing process due to the decrease of saturation temperature induced by higher elevation. Hence, the flow rate of coolant natural circulation does not vary significantly with the change of heat flux applied at the reactor vessel, which is also consistent with the previous literatures.

• Structural Engineering and Mechanics
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• v.51 no.6
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• pp.1005-1016
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• 2014

Roles of reactor internals are to support nuclear fuel, provide insertion and withdrawal channels of nuclear fuel control rods, and carry out core cooling. In case of functional loss of the reactor internals, it may lead to severe accidents caused by damage of nuclear fuel assembly and deterioration of reactor vessel due to attack of fallen out parts. The present study is to examine fluid flows in reactor internals subjected to hydrodynamic loads. In this context, an integrated model was developed and applied to two kinds of numerical analyses; one is to analyze periodic loading effect caused by pump pulsation and the other is to analyze random loading effect employing different turbulent models. Acoustic pressure distributions and flow velocity as well as pressure and temperature fields were calculated and compared to establish appropriate analysis techniques.

• Nuclear Engineering and Technology
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• v.33 no.4
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• pp.397-408
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• 2001

A marine reactor shows very different thermal-hydraulic characteristics compared to a land- based reactor. Especially, study on the variation of flow field due to ship motions such as inclination, heaving and rolling is essential since the flow variation has great influence on the reactor cooling capability. In this study, the natural circulation characteristics of integral type marine reactor with modular steam generators were analyzed using computational fluid dynamics code, CFX-4, for inclined conditions. The numerical analyses are performed using the results of natural circulation experiments for integral reactor which are already conducted at Seoul National University. From the results, it was found that the flow rate in the ascending steam generator cassettes increases due to buoyancy effect. Due to this flow variation, temperature difference occurs at the outlets of the each steam generator cassettes. which is mitigated through downcomer by thermal mixing. Also, around the upper pressure header the flow from descending hot leg goes up to the ascending steam generator cassettes due to large natural circulation driving force in ascending steam generator cassettes. From this result, the increase of How rate in the ascending steam generator cassettes could be understood qualitatively.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.414-422
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• 2021

The aim of this work considers a second order point reactor kinetics model based on the P1 approximation of transport theory, called in this work as P1 point reactor model. The P1 point reactor model implicitly considers the time derivative of the neutron source which has not been thus considered previously. The inverse method to calculate the reactivity in nuclear reactors -chosen because its high accuracy- Matrix Formulation. The numerical results shown that the Matrix Formulation for the reactivity estimation constitutes a method with insignificant calculation errors.

• Journal of computational fluids engineering
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• v.17 no.4
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• pp.41-48
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• 2012

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 is due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.