• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.287-295
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• 2020

Group method of data handling (GMDH) is considered one of the earliest deep learning methods. Deep learning gained additional interest in today's applications due to its capability to handle complex and high dimensional problems. In this study, multi-layer GMDH networks are used to perform uncertainty quantification (UQ) and sensitivity analysis (SA) of nuclear reactor simulations. GMDH is utilized as a surrogate/metamodel to replace high fidelity computer models with cheap-to-evaluate surrogate models, which facilitate UQ and SA tasks (e.g. variance decomposition, uncertainty propagation, etc.). GMDH performance is validated through two UQ applications in reactor simulations: (1) low dimensional input space (two-phase flow in a reactor channel), and (2) high dimensional space (8-group homogenized cross-sections). In both applications, GMDH networks show very good performance with small mean absolute and squared errors as well as high accuracy in capturing the target variance. GMDH is utilized afterward to perform UQ tasks such as variance decomposition through Sobol indices, and GMDH-based uncertainty propagation with large number of samples. GMDH performance is also compared to other surrogates including Gaussian processes and polynomial chaos expansions. The comparison shows that GMDH has competitive performance with the other methods for the low dimensional problem, and reliable performance for the high dimensional problem.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.4
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• pp.443-449
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• 2014

To check the characteristics of nuclear fuels during an irradiation test, the nuclear fuel rod needs to be disassembled from the test rig located in the pool of the research reactor. Then, the disassembled fuel rod is delivered to the hot cell for intermediate examination. A fuel rod that passes the intermediate examination is delivered to the reactor pool to be reassembled into the test rig. The irradiation test is resumed with the reassembled test rig. Because nuclear fuel rods irradiated by neutrons are highly radioactive, all the disassembly and reassembly processes should be carried out in the pool of the research reactor to prevent operators being exposed to radiation. In particular, because a test rig is 5.4-m long and the reactor pool of HANARO is 6-m deep, special tools need to be developed for performing the disassembly and reassembly processes. In this study, a new assembly design of nuclear fuel rods for intermediate examination is introduced. Furthermore, tools for treating the irradiated fuel rod assembly are introduced, and their performance is verified by an out pile test.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.30 no.9 s.252
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• pp.850-857
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• 2006

Hydrogen production using current fueling facilities is essential for near-term applications of fuel cells. A preliminary process for developing a natural gas autothermal reforming (ATR) reactor for fuel cells is presented in this paper. A experimental reactor for methane ATR was constructed and used for characterization of Jin reactor. Temperature profiles of the reactor were observed, and reformed gas compositions were analyzed to evaluate efficiency, conversion and reaction heat with varying amounts of $O_2/CH_4$ at selected furnace temperature and $H_2O/CH_4$. The amount of $O_2/CH_4$ showed strong offsets on reactor temperature, efficiency and conversion indicating that $O_2/CH_4$ is a crucial operation condition. Operation conditions which result in thermal neutrality of ATR reactor system were determined for two cases of an ATR system based on the estimation of enthalpy difference between reactants of assumed inlet temperatures and the products from experimental results. The determined conditions for thermally neutral operations could be used for guidelines to design reformers and for determining the operation parameters of a self sustaining ATR reactor.

• Transactions of the Korean Society of Automotive Engineers
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• v.10 no.2
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• pp.48-54
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• 2002

Since photocatalysts are activated by lights of UV wavelengths, plasma is alternatively used as a light source for a photocatalytic reactor. Light intensity generated by plasma is proportional to the surface area of catalytic material, and this, in many practical applications, is prescribed by the geometry of a plasma generator. Thus, it is crucial to increase the surface area far sufficient light intensity for photocatalytic reaction. For example, in a pack-bed type reactor, multitudes of beads are used as a substrate in order to increase the surface area. Honeycomb monolith type substrate, which has very good surface area to volume ratio, has been difficult to apply plasma as a light source due to the fact that light penetration depth through the honeycomb monolith was too short to cover sufficient area, thus resulting in poor intensity for photocatalytic reaction. In this study, nonthermal plasma generation through a photocatalytic reactor of honeycomb monolith substrate is investigated to lengthen this short penetration depth. The ceramic honeycomb monolith substrate used in this study has the same length as a three way catalyst used fur automotive applications, and it is shown that sufficient light intensity for photocatalytic reaction can also be obtained with honeycomb monolith type reactor.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.1
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• pp.117-123
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• 2015

To remove heat generated during a burn-up test of nuclear fuels, the heat generation rate of nuclear fuels should be calculated accurately, and a coolant should be circulated in the test loop at an adequate flow rate. HANARO is an open pool-type reactor with an independent test loop for the burn-up test of nuclear fuels. A test rig is installed in the test loop, and a coolant is circulated through the test loop to maintain the temperature of the nuclear fuel rods within a desired temperature during an irradiation test. The components and sensors in the test rig can be broken or malfunction owing to the flow-induced vibration. In this study, a coolant flow simulation system was developed to verify and confirm the soundness of components and sensors assembled in the test rig with a high flow rate of the coolant.

• Proceedings of the KIPE Conference
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• 1999.07a
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• pp.65-70
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• 1999

In this paper, the cost-effective design of output reactor which is used to suppress the over-voltage at the motor terminal in the Adjustable Speed Drives(ASD) application is proposed. In the elevator drive system, the power cable length is relatively shorter than other ASD applications and then the over-voltage at the motor terminal depends on the frequency characteristics of the output reactor at the over-voltage operating frequency. The over-voltage suppression mechanism of output reactor in ASD application is analyzed and the dominant parameters of output reactor for the over-voltage suppression are extracted. Using these parameters as the design values and considering the high frequency characteristics of iron core in the reactor, a new cost-effective structure of output reactor is proposed. Experimental results of the conventional reactor and the proposed reactor with a 15kW induction motor are given to verify the proposed scheme.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.12
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• pp.1573-1579
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• 2013

To test the performance of nuclear fuels, coolant needs to be circulated through the test rig installed in the test loop. Because the pressure and temperature of the coolant is 15.5 MPa and $300^{\circ}C$ respectively, coolant sealing is one of the most important processes in fabricating a nuclear fuel test rig. In particular, 15 instrumentation cables installed in a test rig pass through the pressure boundary, and brazing is generally applied as a sealing method. In this study, an induction brazing system has been developed using a high frequency induction heater including a vacuum chamber. For application in the nuclear field, BNi2 should be used as a paste, and optimal process variables for Ni brazing have been found by several case studies. The performance and soundness of the brazed components has been verified by a tensile test, cross section test, and sealing performance test.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.443-448
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• 2006

The OPAL (Open Pool Australian Light-water) reactor is currently being constructed to replace HIFAR (HI-Flux Australian Reactor, commissioned in 1958) in mid-2006. HIFAR will be shutdown for decommissioning after several months of simultaneous operation with OPAL for smooth transition of operating systems and business. OPAL is a 20 MW multipurpose research reactor for radioisotope production, irradiation services and neutron beam research. The OPAL reactor uses low enriched uranium fuel in a compact core, cooled by light water and moderated by heavy water, yielding maximum thermal flux not less than $4{\times}10^{14}ncm^{-2}s^{-1}$. The reactor containment building is constructed of reinforced concrete and has been designed to protect the reactor from all external events such as seismic occurrences and impact from a hypothetical light aircraft crash. This paper describes the main elements of the reactor design and its applications.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.968-976
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• 2019

This paper presents a number of verification case studies for a recently developed sensitivity/uncertainty code package. The code package, ROMUSE (Reduced Order Modeling based Uncertainty/Sensitivity Estimator) is an effort to provide an analysis tool to be used in conjunction with reactor core simulators, in particular the Virtual Environment for Reactor Applications (VERA) core simulator. ROMUSE has been written in C++ and is currently capable of performing various types of parameter perturbations and associated sensitivity analysis, uncertainty quantification, surrogate model construction and subspace analysis. The current version 2.0 has the capability to interface with the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) code, which gives ROMUSE access to the various algorithms implemented within DAKOTA, most importantly model calibration. The verification study is performed via two basic problems and two reactor physics models. The first problem is used to verify the ROMUSE single physics gradient-based range finding algorithm capability using an abstract quadratic model. The second problem is the Brusselator problem, which is a coupled problem representative of multi-physics problems. This problem is used to test the capability of constructing surrogates via ROMUSE-DAKOTA. Finally, light water reactor pin cell and sodium-cooled fast reactor fuel assembly problems are simulated via SCALE 6.1 to test ROMUSE capability for uncertainty quantification and sensitivity analysis purposes.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.72-86
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• 2016

A steam explosion may occur during a severe accident, when the molten core comes into contact with water. The pressurized water reactor and boiling water reactor ex-vessel steam explosion study, which was carried out with the multicomponent three-dimensional Eulerian fuel-coolant interaction code under the conditions of the Organisation for Economic Co-operation and Development (OECD) Steam Explosion Resolution for Nuclear Applications project reactor exercise, is presented and discussed. In reactor calculations, the largest uncertainties in the prediction of the steam explosion strength are expected to be caused by the large uncertainties related to the jet breakup. To obtain some insight into these uncertainties, premixing simulations were performed with both available jet breakup models, i.e., the global and the local models. The simulations revealed that weaker explosions are predicted by the local model, compared to the global model, due to the predicted smaller melt droplet size, resulting in increased melt solidification and increased void buildup, both reducing the explosion strength. Despite the lower active melt mass predicted for the pressurized water reactor case, pressure loads at the cavity walls are typically higher than that for the boiling water reactor case. This is because of the significantly larger boiling water reactor cavity, where the explosion pressure wave originating from the premixture in the center of the cavity has already been significantly weakened on reaching the distant cavity wall.