• Journal of the Korean Society for Marine Environment & Energy
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• v.9 no.2
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• pp.72-78
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• 2006

Effect of organic loading rate on digester performance was evaluated under the conditions of same surface area/reactor volume ratio and different reactor diameter. At the low loading rate of $0.4\;kg\;COD/m^3{\cdot}d$, high rate of organic removal could be obtained regardless of reactor diameter. It can be estimated that reactor configuration can not affect reactor performance at the low loading rate. However, different performance depending on reactor diameter was observed at the organic loading rate of $6\;kg\;COD/m^3{\cdot}d$. That is, volatile acid accumulation and low COD removal efficiency was observed in reactor having 6.4 cm diameter, while volatile acid was not accumulated at all and high COD removal efficiency was observed in reactor having 3 cm diameter. Such a difference of reactor performance depending on reactor diameter can be explained that sludge bed can be fluidized by evolved gas bubble in narrow reactor while sludge bed can not be fluidized by evolved gas bubble only in wide reactor. At a high organic loading rate of $20\;kg\;COD/m^3{\cdot}d$, it can be judged that there is no relation between reactor configuration and reactor performance because all reactors showed very low COD removal efficiencies regardless of reactor diameter. Sludge bed fluidization is one of the most important factors in achieving efficient start-up of anaerobic digester. Narrow and tall type reactor is favorable condition for making sludge bed fluidization at a constant surface area/reactor volume ratio. Thus, it can be judged that reactor configuration and sludge bed fluidization have great influence to reactor performance.

• Journal of environmental and Sanitary engineering
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• v.11 no.1
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• pp.83-91
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• 1996

It is not too much to say that the territorial inhabitants' concerns are wholly c concentrated on the environmental preservation-problem and development-problem in Korea given effect to the local self-government system. At a time like this I was studied the effect on nutrients removal through lighting period in aerobic high rate pond and we know that waste stabilization pond method is the most economical and energy saving wastewater treatment technology than others. At the results which was studied through operating the reactor-l artifically main-tained at a temperature, $25^{\circ}C$, a light intensity, 3000lux, and a lighting period, 24hrs and the reactor-2 artifically maintained at a tern야rature, $25^{\circ}C$ and a light intensity 3000lux, and a lighting period period, 12hrs, It has appeared for 24hrs.-lighting period -reactor-1 to be prior to the reactor-2. The attained results are that 1. reactor-1 is prior to reactor-2 on oxygen-generation 2. reactor-1 is prior to reactor-2 on algal production 3. COD removal efficiency, 90.76%, T-N removal efficiency, 80%, T-P removal e efficiency, 74.47 % in reactor-2, in reactor-1 COD removal efficiency, 94.85 %, T-N removal efficiency, 98.07%, T-P removal efficiency, 72.13% are, so the treatment efficiency of reactor-1 is more excellent than things of reactor-2 4. it appeared that the detention time is 8, 9days.

• 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.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1017-1023
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• 2019

The increase of the alternate current frequency results in increased rotational speed of the electrical motors and connected pumps. The consequence for the reactor coolant pumps is increased flow in primary coolant system. Increase of the current frequency can be initiated by the subsynchronous resonance phenomenon (SSR). This paper analyses the implications of the SSR and consequential increase of the frequency on the nuclear power plant safety. The Simulink $MATLAB^{(R)}$ model of the steam turbine and governor system and RELAP5 computer code of the pressurized water reactor are used in the analysis. The SSR results in fast increase of reactor coolant pumps speed and flow in the primary coolant system. The turbine trip value is reached in short time following SSR. The increase of flow of reactor coolant pumps results in increase of heat removal from reactor core. This results in positive reactivity insertion with reactor power increase of 0.5% before reactor trip is initiated by the turbine trip. The main parameters of the plant did not exceed the values of reactor trip set points. The pressure drop over reactor core is small discarding the possibility of core barrel lift.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.149-154
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• 1998

A subcritical reactor driven by a linear proton accelerator has been considered as a nuclear waste incinerator at Korea Atomic Energy Research Institute(KAERI). Since the multiplication factor of a subcritical reactor is less than unity, to compensate exponentially decreasing fission neutrons from spallation reactions are essentially required for operating the reactor in its steady state. furthermore, the profile of accelerator beam currents is very important in controlling a subcritical reactor, because the reactor power varies in accordance of the profile of external neutrons. We have developed a code system to find numerical solutions of reactor kinetics equations, which are the simplest dynamic model for controlling reactors. In a due course of our previous numerical study of point kinetics equations for critical reactors, however, we learned that the same code system can be used in studying dynamic behavior of the subcritical reactor. Our major motivation of this paper is to investigate responses of subcritical reactors for small changes in thermal hydraulic parameters. Building a thermal hydraulic model for the subcritical reactor dynamics, we performed numerical simulations for dynamic responses of the reactor based on point kinetics equations with a source term. Linearizing a set of coupled differential equations for reactor responses, we focus our research interest on dynamic responses of the reactor to variations of the thermal hydraulic parameters in transient phases.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• KSBB Journal
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• v.20 no.2 s.91
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• pp.100-105
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• 2005

Removal of nitrogen compound from waste water is essential and often accomplished by biological process. Denitrification bacterium, Paracoccus denitrificans (KCTC 2350) is employed to estimate the denitrification ability and the characteristics. In the immobilized biological reactor system, the measurement of absolute amount of active strain in the reactor is comparatively difficult or impossible. In this. study, a reactor was designed with the unwoven texture wrapped peep holed plastic tube to calculate the absolute amount of active strain by comparing the activity of the permeabilized and or immobilized reactor and the free cell reactor The reactor system was continuous stirred tank reactor and the reaction rate of substrate consumption was assumed to satisfy the Michaelis-Menten equation. The effluent concentration of nitrate and nitrite was measured to estimate the apparent parameter of Michaelis-Menten equation. As a result, we found that the amount of immobilized active strain was figured out to be half of the total active strain in the reactor and the time required to be reached in the equilibrium state in the permeabilized and or immobilized reactor system was figured out to be shorter than that of the free cell reactor system.

• Nuclear Engineering and Technology
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• v.39 no.3
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• pp.193-206
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• 2007

The Korea Atomic Energy Research Institute has developed an advanced fast reactor concept, KALIMER-600, which satisfies the Generation IV reactor design goals of sustainability, economics, safety, and proliferation resistance. The concept enables an efficient utilization of uranium resources and a reduction of the radioactive waste. The core design has been developed with a strong emphasis on proliferation resistance by adopting a single enrichment fuel without blanket assemblies. In addition, a passive residual heat removal system, shortened intermediate heat-transport system piping and seismic isolation have been realized in the reactor system design as enhancements to its safety and economics. The inherent safety characteristics of the KALIMER-600 design have been confirmed by a safety analysis of its bounding events. Research on important thermal-hydraulic phenomena and sensing technologies were performed to support the design study. The integrity of the reactor head against creep fatigue was confirmed using a CFD method, and a model for density-wave instability in a helical-coiled steam generator was developed. Gas entrainment on an agitating pool surface was investigated and an experimental correlation on a critical entrainment condition was obtained. An experimental study on sodium-water reactions was also performed to validate the developed SELPSTA code, which predicts the data accurately. An acoustic leak detection method utilizing a neural network and signal processing units were developed and applied successfully for the detection of a signal up to a noise level of -20 dB. Waveguide sensor visualization technology is being developed to inspect the reactor internals and fuel subassemblies. These research and developmental efforts contribute significantly to enhance the safety, economics, and efficiency of the KALIMER-600 design concept.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• v.38 no.5
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• pp.417-422
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• 2006

The present paper describes the 45 MW High Flux Reactor (HFR) which is located in Petten, The Netherlands. This paper focuses on selected technical aspects of this reactor and on nuclear fuel irradiation experiments. These fuel experiments are mainly experiments on Light Water Reactor (LWR) and Very/High Temperature Reactor (V/HTR) fuels, but also on Fast Reactor (FR) fuels, transmutation fuels and Material Test Reactor (MTR) fuels.