• Transactions of the Korean Society of Mechanical Engineers B
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• v.38 no.3
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• pp.271-277
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• 2014

Internal structures, especially those located in the upstream of a reactor core, may have a significant influence on the core inlet flow rate distribution depending on both their shapes and the relative distance between the internal structures and the core inlet. In this study, to examine the effect of the reactor internal structure geometry treatment method on the prediction accuracy for the scale-down APR+ flow distribution, simulations with real geometry modeling were conducted using ANSYS CFX R.14, a commercial computational fluid dynamics software, and the predicted results were compared with those of the porous medium assumption. It was concluded that the core inlet flow distribution could be predicted more accurately by considering the real geometry of the internal structures located in the upstream of the core inlet. Therefore, if sufficient computational resources are available, an exact representation of these internal structures, for example, lower support structure bottom plate and ICI nozzle support plate, is needed for the accurate simulation of the reactor internal flow.

• Nuclear Engineering and Technology
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• v.27 no.4
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• pp.562-570
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• 1995

The recent ABB/CE(Asea Brown Boveri Combustion Engineering) type pressurized oater reactor-s have the on-line monitoring system, i.e., the COLSS(core operating limit supervisory system), to prevent the specified acceptable fuel design limits from being violated during normal operation and anticipated operational occurrences. One of the main functions of COLSS is the on-line monitoring of the DNB(departure from nucleate boiling) overpower margin by calculating the MDNBR(mini-mum DNB ratio) for the measured operating condition at every second. The CETOP-D model, used in the MDNBR calculation of COLSS, is benchmarked conservatively against the TORC mod-el using an inlet flow factor of hot assembly in CETOP-D as an adjustment factor for TORC. In this study, a technique to optimize the CETOP-D inlet flow factor has been developed by elim-inating the excessive conservatism in the ABB/CE's. A correlation is introduced to account for the actual variation of the CETOP-D inlet flow factor within the core operating limits. This technique was applied to the core operating range of the YongGwang Units 3&4 Cycle 1, which results in the increase of 2％ in the DNB overpower margin at the normal operating condition, compared with that from the ABB/CE method.

• Nuclear Engineering and Technology
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• v.23 no.3
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• pp.340-351
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• 1991

Experimental studies were conducted on a l/5.03 scale reactor flow model of the Yong-gwang Nuclear Units 3 and 4. The purpose of the flow model test was to estimate the hydraulic effect in the reactor vessel due to the relative size difference between the ABB-CE's System 80 and the YGN 3&4 reactors. The flow model was designed according to the principle of similarity. Obtained from the test were the core inlet flow distribution, the core exit pressure deviations, and the segmental and overall pressure losses across the flow path from the reactor vessel inlet to outlet nozzle. These data will be used to provide input data for the core thermal margin analysis and to verify the analytical hydraulic design method.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.18 no.10
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• pp.819-825
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• 2006

In a SOFC/GT (solid oxide fuel cell/gas turbine) hybrid power generation system, the recuperator is an indispensible component to enhance system performance. Since the expansion ratio to the recuperator core is very large, generally, the effective header design to distribute the flow uniformly before entering the core is crucial to guarantee the required performance. In the present study, we focus on the design of a diffuser type recuperator header with a 90 degree turn inlet port. To reduce the flow separation and recirculation flows, multiple horizontal vanes are used. The number of horizontal vanes is varied from 0 to 24. The air flow velocity is measured at 40 points just behind the core outlet by using a hot wire anemometer. Then, the flow non-uniformity is evaluated from the measured flow velocity. The experimental results showed that inlet air velocity did not effect on relative flow non-uniformity. According to increasing the number of horizontal vanes, flow non-uniformity reduced about $40{\sim}50%$ than without using horizontal vanes.

• Nuclear Engineering and Technology
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• v.54 no.9
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• pp.3567-3579
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• 2022

This paper provides an assessment of fluid transport and mixing processes inside the primary circuit of the test facility ROCOM through the numerical simulation of Test 2.1 with the system code ATHLET. The experiment represents an asymmetric injection of cold and non-borated water into the reactor coolant system (RCS) of a pressurized water reactor (PWR) to restore core cooling, an emergency procedure which may subsequently trigger a core re-criticality. The injection takes place at low velocity under single-phase subcooled conditions and presents a major challenge for the simulation in lumped parameter codes, due to multidimensional effects in horizontal piping and vessel arising from density gradients and gravity forces. Aiming at further validating ATHLET 3-D capabilities against horizontal geometries, the experiment conditions are applied to a ROCOM model, which includes a newly developed horizontal pipe object to enhance code prediction inside coolant loops. The obtained results show code strong simulation capabilities to represent multidimensional flows. Enhanced prediction is observed at the vessel inlet compared to traditional 1-D approach, whereas mixing overprediction from the descending denser plume is observed at the upper-half downcomer region, which leads to eventual deviations at the core inlet.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• The KSFM Journal of Fluid Machinery
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• v.13 no.1
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• pp.35-41
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• 2010

Experimental study on the flow field inside the nozzle for radial turbine was performed. At design point, the pressure is high and the Mach number is low at the pressure side of the nozzle inlet semi-vaneless space as the flow turns through the nozzle vanes. As the flow accelerates through the nozzle passage to the throat the pressure level at the pressure and suction sides becomes similar. The flow continued accelerating from the throat to the inlet of turbine wheel and the pressure field became uniform in the circumferential direction in the vaneless space. In high expansion ratio condition, strong favorable pressure gradient band region occurred just after the throat in the semi-vaneless space in the circumferential direction and the pressure became uniform in the circumferential direction after this band. In low expansion ratio condition, core flow acceleration is dominant after the throat and this non-uniform pressure field reached to the inlet of turbine wheel.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.617-626
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• 2022

To effectively monitor the variety of distributions of neutron flux, fuel power or temperatures in the reactor core, usually the ex-core and in-core neutron detectors are employed. The thermocouples for temperature measurement are installed in the coolant inlet or outlet of the respective fuel assemblies. It is necessary to reconstruct the measurement information of the whole reactor position. However, the reading of different types of detector in the core reflects different aspects of the 3D power distribution. The feasibility of reconstruction the core three-dimension power distribution by using different combinations of in-core, ex-core and thermocouples detectors is analyzed in this paper to synthesize the useful information of various detectors. A comparison of multilayer perceptron (MLP) network and radial basis function (RBF) network is performed. RBF results are more extreme precision but also more sensitivity to detector failure and uncertainty, compare to MLP networks. This is because that localized neural network could offer conservative regression in RBF. Adding random disturbance in training dataset is helpful to reduce the influence of detector failure and uncertainty. Some convolution neural networks seem to be helpful to get more accurate results by use more spatial layout information, though relative researches are still under way.

• Nuclear Engineering and Technology
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• v.38 no.2
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• pp.109-118
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• 2006

Various indirect power cycle options for a helium cooled gas cooled fast reactor (GFR) with particular focus on a supercritical $CO_2(SCO_2)$ indirect cycle are investigated as an alternative to a helium cooled direct cycle GFR. The balance of plant (BOP) options include helium-nitrogen Brayton cycle, supercritical water Rankine cycle, and $SCO_2$ recompression Brayton power cycle in three versions: (1) basic design with turbine inlet temperature of $550^{\circ}C$, (2) advanced design with turbine inlet temperature of $650^{\circ}C$ and (3) advanced design with the same turbine inlet temperature and reduced compressor inlet temperature. The indirect $SCO_2$ recompression cycle is found attractive since in addition to easier BOP maintenance it allows significant reduction of core outlet temperature, making design of the primary system easier while achieving very attractive efficiencies comparable to or slightly lower than, the efficiency of the reference GFR direct cycle design. In addition, the indirect cycle arrangement allows significant reduction of the GFR &proximate-containment& and the BOP for the $SCO_2$ cycle is very compact. Both these factors will lead to reduced capital cost.

• Nuclear Engineering and Technology
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• v.27 no.5
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• pp.710-720
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• 1995

The propagation of pump-induced pressure pulsation in a reactor is important because of the potential for vibration and resultant damage of reactor internals. A hydrodynamic model has been developed to obtain the pressure fluctuation due to the operation of pumps in the annulus(between the core support barrel and reactor vessel of a pressurized water reactor) including the coolant inlet pipe. The mathematical analysis is formulated in accordance with the linearized Navier-Stokes equation by assuming a compressible, inviscid flow. Two regions are considered separately and by coupling the solutions of the inlet pipe and the annulus, the inlet nozzle pressure(pressure at pipe and annulus interface) is to be calculated without assumptions. The geometric parameter effect on the pump-induced pressure pulsation is evaluated. Comparison of predicted and measured inlet nozzle pressure values for each forcing frequency shows good order of magnitude agreement.