• Nuclear Engineering and Technology
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• v.55 no.11
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• pp.3983-3995
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• 2023

In fast reactors, restructuring of the fuel micro-structure driven by high temperature and high temperature gradient can cause the formation of columnar grains. The non-spheroidal shape and the non-uniform temperature field in such columnar grains implies that standard models for fission gas diffusion can not be applied. To tackle this issue, we present a reduced order model for the fission gas diffusion process which is applicable in different geometries and with non-uniform temperature fields, maintaining a computational requirement in line with its application in fuel performance codes. This innovative application of reduced order models as meso-scale tools within fuel performance codes represents a first-of-a-kind achievement that can be extended beyond fission gas behaviour.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.408-420
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• 2023

Beryllium oxide (BeO) is being re-emphasized and utilized in Micro Modular Reactors (MMR) because of its prominent nuclear and high temperature properties in recent years. The implications of the research about effects of neutron irradiation on the microstructure and properties of BeO are significant. This article comprehensively reviews the effects of neutron irradiation on BeO and proposes the maximum permissible neutron doses at different temperatures for BeO without cracks in appearance according to the data in the previous literature. This maximum permissible neutron dose value has important reference significance for the experimental study of BeO. The effects of neutron irradiation on the thermal conductivity and flexural strength of BeO are also discussed. In addition, microstructure evolution of irradiated BeO during post-irradiation annealing is summarized. This review article has important implications for the application of BeO in MMR.

• Smart Structures and Systems
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• v.24 no.6
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• pp.759-768
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• 2019

In nuclear fusion reactors, the key structural component (i.e., the plasma-facing component) undergoes high heat flux cyclic loading. To ensure the safety of fusion reactors, an experimental study on the temperature-induced creep of stainless steel under heat flux cyclic loading was performed in the present work. The strains were measured using a stereo digital image correlation technique (3D-DIC). The influence of the heat haze was eliminated, owing to the use of a vacuum environment. The specimen underwent heat flux cycles ($500^{\circ}C-1000^{\circ}C$) with different mechanical preloads (0 kN, 10 kN, 30 kN, and 50 kN). The results revealed that, for a relatively large preload (for example, 50 kN), a single temperature cycle can induce a residual strain of up to $15000{\mu}{\varepsilon}$.

• Applied Chemistry for Engineering
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• v.24 no.4
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• pp.337-343
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• 2013

Membrane reactors have been showing a promising future and attracted increasing attention in the scientific community as they possess advantages in terms of enhanced catalytic activity and selectivity, combination of processes (reaction and separation), simplicity in process design, and safety in operation. In particular, solid electrolyte membrane reactor principles are realized in fuel cells, electrolyzers and reactors for hydrogenation of carbon dioxide and other economically viable reactions. In this review, as a young generation of ion conducting materials, high temperature proton conductors are discussed in terms of the current status of material development and their various applications.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.25-78
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• 2005

As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming. After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows: (1) Derivation of the multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (now almost three decades old) diffusion nodal methods. (3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues. (4) Description of the analytic function expansion nodal (AFEN) method. (5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods. (6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis. (7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.

• 한국신재생에너지학회:학술대회논문집
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• 2011.11a
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• pp.175.1-175.1
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• 2011

A rotating activated bacillus contactor (RABC) process with a series of aerobic reactors was tested in pilot scale to treat digested liquid from an anaerobic digester treating swine wastewater and sewage sludge. The influent (digested liquid) for the RABC process showed C/N ratios less than 2 as a typical feature of effluent from anaerobic digesters. The pilot process, which consists of three 3 RABC reactors, four aerobic tanks and a sedimentation tank, was operated for 210 days with a hydraulic retention time of 20 days without pH and temperature control. Since the Bacillus-enriched aerobic reactors shows high efficiencies of nitrogen removal at low DO levels less than 1.0 mg/L, they were operated at reduced aeration intensities. With relatively low concentrations of organics in comparison with nitrogen concentrations, the RABC process tested in this study showed stable and high nitrogen and organics removal efficiencies over 80%. The nitrogen removal process tested in this study was proven to be an effective and operation-cost saving (lower aeration) method to remove nitrogen without adding external carbon sources to meet the optimum C/N ratio.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.663-667
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• 2017

Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below $300^{\circ}C$. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor ($K_{IH}$) to initiate DHC as a function of temperature in Zry-4 for temperatures between $227^{\circ}C$ and $315^{\circ}C$. The experimental technique used in this study was the pin-loading testing technique. To determine the $K_{IH}$, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around $300^{\circ}C$, there was a sharp increase in $K_{IH}$ indicating the upper temperature limit for DHC. The value for $K_{IH}$ at $227^{\circ}C$ was determined to be $2.6{\pm}0.3MPa$ ${\surd}$m.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.2228-2231
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• 2007

Water gas shift reactor in fuel processing is an important part that converts carbon monoxide into hydrogen. Fuel processing system for PEMFC usually has two stages of WGS reactors, which are high temperature and low temperature shifter. In this study we prepared noble metal catalysts and compared their performances with that of a commercial iron chromium oxide catalyst. Noble metal catalysts and the commercial catalyst showed quite different temperature dependence of carbon monoxide conversion. The conversion of carbon monoxide at the commercial catalyst was very low at medium temperature(${\sim}300^{\circ}C$) and increased rapidly as temperature increased while the conversion at noble metal catalysts was high in the medium temperature range and decreased as temperature increased, which is thermodynamically expected. Their characteristics agreed well with the literature published, and we are accomplishing further study for improvement of the noble metal catalysts.

• Korean Journal of Materials Research
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• v.16 no.5
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• pp.318-322
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• 2006

Short-term high temperature degradation test was conducted on Hastelloy X, a candidate tube material for high temperature gas-cooled reactors (HTGR), to evaluate the variation of microstructure and mechanical property in air at $1050^{\circ}C$ during 2000 h. The dominant oxide layer was Cr-oxide and a very shallow Cr-depleted region was observed below the oxide layer. At the beginning of degradation, the island shape $M_6C$ precipitate (M=Mo-rich, Fe, Ni, Cr) was observed in matrix region. After 2000 h degradation, precipitate shape was changed to the chain shape and increased amount of precipitate. These results influenced mechanical property of the specimen which exposed in high temperature. Yield strength was decreased from 115MPa to 89 MPa after 24 h and 2000 h exposure, respectively.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2989-2998
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• 2022

Line contact structures, such as the contact between graphite brick and graphite tenon, widely exist in high-temperature gas-cooled reactors. Due to the stress concentration effect, the line contact area is one of the dangerous positions prone to failure in the nuclear reactor core. In this paper, the failure mechanism of line contact structures composed of IG11 nuclear graphite column and brick were investigated by means of experiment and finite element simulation. It was found that the failure process mainly includes three stages: firstly, the damage accumulation in nuclear graphite material led to the characteristic yielding of the line contact structure, but no macroscopic failure can be observed at this stage; secondly, the stresses near the contact area met Mohr failure criterion, and a crack initiated and propagated laterally in the contact zone, that is, local macroscopic failure occurred at this stage; finally, a second crack initiated in the contact area and developed in to a Y-shape, resulting in the final failure of the structure. This study lays a foundation for the structural design and safety assessment of high-temperature gas-cooled reactors.