• Journal of Korean Society of Water and Wastewater
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• v.8 no.2
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• pp.25-34
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• 1994

In this research, investigations were made on the effect of type and load of organic substrate on phosphorus release. Reactors of three different sizes were operated, being fed on five kinds of organic substrates. The quantitative analyses were made on phosphorus release and substrate utilization under anaerobic condition. The molar ratios of the uptaken organic substrate to the released phosphorus were 0.5 with acetate, 0.6 with glucose, 0.8 with glucose/acetate, and 1.2 with glucose/acids, respectively. The phosphorus release was inhibited at the higher organic load than the normal at stead state. Both acetate and acids/glucose enhanced phosphorus release- as well as uptake-rate, however, the complete phosphorus removal was achieved after the microbial adaptation to the new environment. In case with acetate, operation was hampered by the poor sludge settleability and phosphorus uptake was not enough although the phosphorus release was active. But with milk/starch, the phosphorus release and uptake was well developed even though phosphorus release was not comparatively high. From this study, it was concluded that organic substrates, such as glucose seemed to be converted fatty acids after fast bio-sorption, followed by concurrent uptake of these acids by excess phosphorus removing bacteria.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.647-661
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• 2015

The supercritical $CO_2$ (S-$CO_2$) Brayton cycle has recently been gaining a lot of attention for application to next generation nuclear reactors. The advantages of the S-$CO_2$ cycle are high efficiency in the mild turbine inlet temperature region and a small physical footprint with a simple layout, compact turbomachinery, and heat exchangers. Several heat sources including nuclear, fossil fuel, waste heat, and renewable heat sources such as solar thermal or fuel cells are potential application areas of the S-$CO_2$ cycle. In this paper, the current development progress of the S-$CO_2$ cycle is introduced. Moreover, a quick comparison of various S-$CO_2$ layouts is presented in terms of cycle performance.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.189-199
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• 2016

Fast neutrons with a broad energy spectrum, with which it is possible to evaluate nuclear data for various research fields such as medical applications and the development of fusion reactors, can be generated by irradiating proton beams on target materials such as beryllium. To generate short-pulse proton beam, we adopted a deflector and slit system. In a simple deflector with slit system, most of the proton beam is blocked by the slit, especially when the beam pulse width is short. Therefore, the available beam current is very low, which results in low neutron flux. In this study, we proposed beam modulation using a buncher cavity to increase the available beam current. The ideal field pattern for the buncher cavity is sawtooth. To make the field pattern similar to a sawtooth waveform, a multiharmonic buncher was adopted. The design process for the multiharmonic buncher includes a beam dynamics calculation and three-dimensional electromagnetic simulation. In addition to the system design for pulsed proton generation, a test bench with a microwave ion source is under preparation to test the performance of the system. The design study results concerning the pulsed proton beam generation and the test bench preparation with some preliminary test results are presented in this paper.

• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.915-922
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• 2018

Metallic fuel has been considered for sodium-cooled fast reactors because it can maximize the uranium resources. It generates rare earth elements as fission products, where it is reported by aggravating the fuel-cladding chemical interaction at the operating temperature. Rare earth elements form a multicomponent alloy (Ce-Nd-Pr-La-Sm-etc.) during reactor operation, where it shows a higher reaction thickness than a single element. Experiments have been carried out by simplifying multicomponent alloys for mono or binary systems because complex alloys have difficulty in the analysis. In previous experiments, xCe-yNd was fabricated with two elements, Ce and Nd, which have a major effect on the fuel-cladding chemical interaction, and the thickness of the reaction layer reached maximum when the rare earth elements ratio was 1:1. The objective of this study is to evaluate the effect and relationship of rare earth elements on such synergistic behavior. Single and binary rare earth model alloys were prepared by selecting five rare earth elements (Ce, Nd, Pr, La, and Sm). In the single system, Nd and Pr behaviors were close to diffusion, and Ce showed a eutectic reaction. In the binary system, Ce and Sm further increased the reaction layer, and La showed a non-synergy effect.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1069-1074
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• 2019

A novel fabrication method of oxide dispersion strengthened (ODS) steel cladding tubes for advanced fast reactors has been investigated using the cold spray powder-based materials deposition process. Cold spraying has the potential advantage for rapidly fabricating ODS cladding tubes in comparison with the conventional multi-step extrusion process. A gas atomized spherical 14YWT (Fe-14%Cr, 3%W, 0.4%Ti, 0.2% Y, 0.01%O) powder was sprayed on a rotating cylindrical 6061-T6 aluminum mandrel using nitrogen as the propellant gas. The powder lacked the oxygen content needed to precipitate the nanoclusters in ODS steel, therefore this work was intended to serve as a proof-of-concept study to demonstrate that free-standing steel cladding tubes with prototypical ODS composition could be manufactured using the cold spray process. The spray process produced an approximately 1-mm thick, dense 14YWT deposit on the aluminum-alloy tube. After surface polishing of the 14YWT deposit to obtain desired cladding thickness and surface roughness, the aluminum-alloy mandrel was dissolved in an alkaline medium to leave behind a free-standing ODS tube. The as-fabricated cladding tube was annealed at $1000^{\circ}C$ for 1 h in an argon atmosphere to improve the overall mechanical properties of the cladding.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.249-256
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• 2019

The 9Cr-1Mo ferritic-martensitic ODS steel is a promising structural material for the next generation nuclear power plants including fast reactors for application in reactor vessels and nuclear fuel. The ODS steel was cooled down by furnace cooling, air cooling, oil quenching and water quenching, respectively, after normalizing it at $1150^{\circ}C$ for 1 h and then tempering at $780^{\circ}C$ for 1 h. It is found that grain size, a relative portion of ferrite and martensite, martensitic lath configuration, behaviors of carbide precipitates, and hardness of the ODS steel are strongly dependent on a cooling rate. The grain size and martensitic lath width become smaller with the increase in a cooling rate. The carbides were precipitated at the grain boundaries formed between the ferrite and martensite phases and at the martensitic lath interfaces. In addition, the carbide precipitates become smaller and more widely dispersed with the increase in a cooling rate, resulting in that the faster cooling rate generated the higher hardness of the ODS steel.

• Nuclear Engineering and Technology
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• v.54 no.4
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• pp.1482-1494
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• 2022

Power conversion cycles (Subcritical Steam, Supercritical Steam, Open Air Brayton, Recuperated Air Brayton, Combined Cycle, Closed Brayton Supercritical CO₂ (sCO₂), and Stirling) are evaluated for land-based nuclear microreactors based on technical maturity, system efficiency, size, cost and maintainability, safety implications, and siting considerations. Based upon these criteria, Air Brayton systems were selected for further evaluation. A brief history of the development and applications of Brayton power systems is given, followed by a description of how these thermal-to-electrical energy conversion systems might be integrated with a nuclear microreactor. Modeling is performed for optimized cycles operating at 3 MW(e) with turbine inlet temperatures of 500 ℃, 650 ℃ and 850 ℃, corresponding to: a) sodium fast, b) molten salt or heat pipe, and c) helium or sodium thermal reactors, coupled with three types of Brayton power conversion units (PCUs): 1) simple open-cycle gas turbine, 2) recuperated open-cycle gas turbine, and 3) recuperated and intercooled open-cycle gas turbine. Aeroderivative turboshaft engines employing the simple Brayton cycle and two industrial gas turbine engines employing recuperated air Brayton cycles are also analyzed. These engines offer mature technology that can facilitate near-term deployment with a modest improvement in efficiency.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2373-2379
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• 2023

In sodium systems, leakage is one of the safety concerns; it can cause chemical reactions, which may result in fires. There are contact and non-contact types of leak detectors, and the conventional method of non-contact type detection is by gas sampling. Because of the complexity of this method, there has always been a need for a simple gas sensor, and the resistive-type nanostructure ZnO sensor is a promising option with various advantages. In this study, a ZnO sensor was fabricated, and the concept was tested as a leak detector using a dedicated experiment facility. The experiment results showed distinctive changes in resistance with the presence of sodium aerosol under various conditions. Replacing the conventional gas sampling with the ZnO sensors is expected to enable identification of the leakage location if used as a point-wise instrumentation and to greatly reduce the total cost, making the system simple, light, and effective. For further study, more tests will be performed to evaluate the sensitivity of key parameters under various conditions.

• 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.