• Nuclear Engineering and Technology
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• v.50 no.1
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• pp.54-67
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• 2018

There have been recent efforts to establish methods for high-fidelity and multi-physics simulation with coupled thermal-hydraulic (T/H) and neutronics codes for the entire core of a light water reactor under accident conditions. Considering the computing power necessary for a pin-by-pin analysis of the entire core, subchannel-scale T/H analysis is considered appropriate to achieve acceptable accuracy in an optimal computational time. In the present study, the applicability of in-house code CUPID of the Korea Atomic Energy Research Institute was extended to the subchannel-scale T/H analysis. CUPID is a component-scale T/H analysis code, which uses three-dimensional two-fluid models with various closure models and incorporates a highly parallelized numerical solver. In this study, key models required for a subchannel-scale T/H analysis were implemented in CUPID. Afterward, the code was validated against four subchannel experiments under unheated and heated single-phase incompressible flow conditions. Thereafter, a subchannel-scale T/H analysis of the entire core for an Advanced Power Reactor 1400 reactor core was carried out. For the high-fidelity simulation, detailed geometrical features and individual rod power distributions were considered in this demonstration. In this study, CUPID shows its capability of reproducing key phenomena in a subchannel and dealing with the subchannel-scale whole core T/H analysis.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.27 no.3
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• pp.191-198
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• 2014

In this paper, structural integrity evaluation of reactor pressure vessel bottom head without penetration nozzles in core melting accident has been performed. Considering the analysis results of thermal load, weight of molten core debris and internal pressure, thermal load is the most significant factor in reactor vessel bottom head. The failure probability was evaluated according to the established failure criteria and the evaluation showed that the equivalent plastic strain results are lower than critical strain failure criteria. Thermal-structural coupled analyses show that the existence of elastic zone with a lower stress than yield strength is in the middle of bottom head thickness. As a result of analysis, the elastic zone became narrow and moved to the internal wall as the internal pressure increases, and it is evaluated that the structural integrity of reactor vessel is maintained under core melting accident.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.439-445
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• 2021

In view of the complex geometric structure of the rod bundle channel and the limitation of the current CHF visualization experiment technology, it is very difficult to obtain the rod bundle CHF mechanism directly through the phenomenon of the rod bundle CHF visualization experiment. In order to obtain the applicable CHF mechanism assumption for rod bundle channel, firstly, five most representative DNB type round tube CHF mechanistic models are obtained with evaluation and screening. Then these original round tube CHF mechanistic models based on inlet conditions are converted to local conditions and coupled with subchannel analysis code ATHAS. Based on 5 × 5 full-length rod bundle CHF experimental data independently developed by Nuclear Power Institute of China (NPIC), the applicability research of each model for CHF prediction performance in rod bundle channel is carried out, and the commonness and difference of each model are comparatively studied. The CHF mechanism assumption of superheated liquid layer depletion that is most likely to be applicable for the rod bundle channel is selected and two directions that need to be improved are given. This study provides a reference for the development of CHF mechanistic model in rod bundle channel.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2789-2802
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• 2020

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

• Bulletin of the Korean Chemical Society
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• v.33 no.8
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• pp.2642-2646
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• 2012

Pretreatment of spent FCC catalyst and its application in remove trace olefins in aromatics were investigated in this research. The most effective pretreatment route of spent FCC catalyst was calcining at $700^{\circ}C$ for 1 h, washing with 5% oxalic acid solution in ultrasonic reactor and dried. Treated spent FCC catalyst was modified with metal halides, then to prepare catalyst to remove trace olefins in aromatics. X-ray diffraction, Pyridine-FTIR, $N_2$ adsorption-desorption and inductively coupled plasma optical emission spectrometer (ICP-OES) were used to investigate the pretreatment process. The result showed that the performance of the treated spent FCC catalyst was much greater than that of the spent FCC catalyst, which indicted the possibility and improvement of this research.

• Environmental Engineering Research
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• v.19 no.1
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• pp.47-57
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• 2014

The aim of this paper was to study the effect of five different waste combinations (C/N 15, C/N 20, C/N 25, C/N 30, and control) of sewage sludge coupled with sawdust and cattle manure in a pilot scale rotary drum reactor, during 20 days of the composting process. Our results showed that C/N 30 possesses a higher temperature regime with higher % reduction in moisture content, total organic carbon, soluble biochemical oxygen demand and chemical oxygen demand; and higher % gain in total nitrogen and phosphorus at the end of the composting period implying the total amount of biodegradable organic material is stabilized. In addition, $CO_2$ evolution and oxygen uptake rate decreased during the process, reflecting the stable behavior of the final compost. A Solvita maturity index of 8 indicated that the compost was stable and ready for usage as a soil conditioner. The results indicated that composting can be an alternate technology for the management of sewage sludge disposal.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.448-455
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• 2020

The H.B. Robinson Unit 2 (HBR-2) pressure vessel dosimetry benchmark is an in- and ex-Reactor Pressure Vessel (RPV) neutron dosimetry benchmark based on experimental data from the HBR-2 reactor, a 2300-MW PWR designed by Westinghouse and put in operation in March 1971, openly available through the SINBAD Database at OECD/NEA data Bank. The goals of the present work were to carry out three-dimensional (3D) fixed source transport calculations in both Cartesian (X,Y,Z) and cylindrical (R,θ,Z) geometries by using the TORT-3.2 discrete ordinates code on very detailed 3D HBR-2 geometrical models and to test the latest broad-group coupled (47 neutron groups + 20 photon groups) working cross section libraries in FIDO-ANISN format with same structure as BUGLE-96, such as BUGJEFF311.BOLIB, BUGENDF70.BOLIB and BUGLE-B7. The results obtained with all the cited libraries were satisfactory and are here reported and compared.

• Proceedings of the Korean Institute of IIIuminating and Electrical Installation Engineers Conference
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• 2000.11a
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• pp.209-213
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• 2000

In this paper, the removal rate of NO in a reactor is measured first using nonthermal plasma generated from a wire-cylinder type reactor, then the spatial density distribution of NO particles is investigated using ICCD(Intensified Charged Coupled Device) camera. This research uses nonthermal plasma from electrical discharge to analyze the NO characteristics, and the measurements of NO discharge image and Distribution are performed using the ICCD camera to examine the NO characteristics more closely. Furthermore, the method of Laser Induced Fluorescence (LIF) is used to analyze the particular behavior of NO particles more specifically, to suggest a method of reducing exhaust gas, a serious environmental problem.

• Journal of Mechanical Science and Technology
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• v.18 no.12
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• pp.2125-2136
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• 2004

In this paper, a numerical application algorithm for applying the CFAM (Consistent Fluid Added Mass) matrix for a core seismic analysis is developed and applied to the 7-ducts core system to investigate the fluid effects on the dynamic characteristics and the seismic time history responses. To this end, three cases such as the in-air condition, the in-water condition without the fluid coupling terms, and the in-water condition with the fluid coupling terms are considered in this paper. From modal analysis, the core duct assemblies revealed strongly coupled out-of-phase vibration modes unlike the other cases with the fluid coupling terms considered. From the results of the seismic time history analysis, it was also verified that the fluid coupling terms in the CFAM matrix can significantly affect the impact responses and the seismic displacement responses of the ducts.

• 한국연소학회:학술대회논문집
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• 2012.04a
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• pp.213-215
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• 2012

As one of the primary fuel sources, oxy-fuel combustion of coal is actively being investigated because of the climate changing problem such like the emission of green house gases. In this paper research about the pulverized coal technology, which is widely used in both power-generating and iron-making processes was studied to invesgate the ignition behaviour of pulverized coal particles during coal combustion as changing the ambient oxygen concentration of the particle. The ignition phenomenon of the coal particles fed into a laminar flow reactor was imaged with a Integrated charged-coupled device (ICCD) camera. The ignition points were determined throught the analysis of the images, and then the ignition delay times were able to be calculated. The experiment results show that a lower oxygen concentration increases the ignition delay time.