• Nuclear Engineering and Technology
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• v.46 no.2
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• pp.169-182
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• 2014

High-performance research reactors require fuel that operates at high specific power to high fission density, but at relatively low temperatures. Research reactor fuels are designed for efficient heat rejection, and are composed of assemblies of thin-plates clad in aluminum alloy. The development of low-enriched fuels to replace high-enriched fuels for these reactors requires a substantially increased uranium density in the fuel to offset the decrease in enrichment. Very few fuel phases have been identified that have the required combination of very-high uranium density and stable fuel behavior at high burnup. U-Mo alloys represent the best known tradeoff in these properties. Testing of aluminum matrix U-Mo aluminum matrix dispersion fuel revealed a pattern of breakaway swelling behavior at intermediate burnup, related to the formation of a molybdenum stabilized high aluminum intermetallic phase that forms during irradiation. In the case of monolithic fuel, this issue was addressed by eliminating, as much as possible, the interfacial area between U-Mo and aluminum. Based on scoping irradiation test data, a fuel plate system composed of solid U-10Mo fuel meat, a zirconium diffusion barrier, and Al6061 cladding was selected for development. Developmental testing of this fuel system indicates that it meets core criteria for fuel qualification, including stable and predictable swelling behavior, mechanical integrity to high burnup, and geometric stability. In addition, the fuel exhibits robust behavior during power-cooling mismatch events under irradiation at high power.

• Journal of Mechanical Science and Technology
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• v.19 no.10
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• pp.1891-1901
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• 2005

Recently the authors tried to find damage position only using measured frequency response functions. According to their work, it seems that the algorithm is very practical since it needs only measured frequency responses while other methods require exact analytic model. But when applying the method to a real structure, it requires lots of experiment. The authors, in this time, propose a method to reduce its experimental load by detecting damage within a substructure. This method searches damages not within an entire structure but within substructures. In addition, damage severity was treated in this paper since it is worthy to know damage severity. Optimization technique is used to estimate damage level using measured responses and damage model. Two test examples, a plate and a jointed structure, are chosen to verify the suggesting method.

• Korean Journal of Materials Research
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• v.8 no.9
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• pp.783-790
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• 1998

New process development of nuclear fuel fabrication for nuclear power plant was attempted by induction plasma technology with yttria-stabilized-zirconia ($\textrm{ZrO}_{2}$-$\textrm{Y}_{2}\textrm{O}_{3}$)powder, similar to $\textrm{UO}_{2}$, in the respect of melting point and physicochemical characteristics. Extent of powder melting was affected greatly by plasma plate power and particle size. Being optimized such as, sheath gas composition, probe position, particle size and spraying distance, dense deposit of 97.91% T.D. with deposition rate 20mm/min was attained at the condition of 120/20$\ell$/min of Ar/$\textrm{H}_{2}$ flow rate, 80kw of plate power, 8cm of probe position, 200Torr of chamber pressure and 18cm of spraying distance. The pellet of 96.5% of theoretical density was formed with homogeneity and nice exterior view at the best condition of deposition experiments, and the possibility of new nuclear pellet fabrication process was confirmed. The main and interrelated effects on deposit density were assessed by ANOVA(Ana1ysis of Variance).

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.

• Journal of Electrochemical Science and Technology
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• v.10 no.2
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• pp.104-114
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• 2019

Energy is an essential driving force for modern society. In particular, electricity has become the standard source of power for almost every aspect of life. Electric power runs lights, televisions, cell phones, laptops, etc. However, it has become apparent that the current methods of producing this most valuable commodity combustion of fossil fuels are of limited supply and has become detrimental for the Earth's environment. It is also self-evident, given the fact that these resources are non-renewable, that these sources of energy will eventually run out. One of the most promising alternatives to the burning of fossil fuel in the production of electric power is the proton exchange membrane (PEM) fuel cell. The PEM fuel cell is environmentally friendly and achieves much higher efficiencies than a combustion engine. Water management is an important issue of PEM fuel cell operation. Water is the product of the electrochemical reactions inside fuel cell. If liquid water accumulation becomes excessive in a fuel cell, water columns will clog the gas flow channel. This condition is referred to as flooding. A number of researchers have examined the water removal methods in order to improve the performance. In this paper, a new water removal method that investigates the use of vibro-acoustic methods is presented. Piezo-actuators are devices to generate the flexural wave and are attached at end of a cathode bipolar plate. The "flexural wave" is used to impart energy to resting droplets and thus cause movement of the droplets in the direction of the traveling wave.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.39 no.11
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• pp.905-911
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• 2015

In this study, to examine the effect of a hole size change(smaller hole diameter) in the outer region of the lower-support-structure-bottom plate(LSSBP) on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.15. The predicted results were compared with those of the original LSSBP. Through these comparisons, it was concluded that a more uniform distribution of the mass flow rate at the core-inlet plane could be obtained by reducing the hole size in the outer region of the LSSBP. Therefore, from the nuclear regulatory perspective, design change of the hole pattern in the outer region of the LSSBP may be desirable in terms of improving both the mechanical integrity of the fuel assembly and the core thermal margin.

• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.111-120
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• 1992

This paper investigates the dynamic characteristics of spacer grid impact loads and the effects of variations in the amplitude and frequency of the core plate motions on the resultant impact loads. A model of the longest row (15 fuel assemblies) across the core is analyzed using the input motions generated from safe shutdown earthquake. Input excitations consist of time history motions applied to the core support plate, fuel alignment plate and core shroud. The responses are determined for a set of four parameter runs with respect to the amplitude and frequency changes. Spacer grid impact loads and normalized input values for all cases are presented. The results show that changing the natural frequency has negligible effect but changing the amplitude of the input motions has a significant effect on the grid impact loads Therefore, time history analysis is not necessary for a shifted case to get the core responses under the seismic excitation.

• Nuclear Engineering and Technology
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• v.18 no.4
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• pp.267-272
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• 1986

To obtain various necessary data for the manufacturing and the use of the nuclear fuel cladding tube, the effects of deformation and heat treatment on Properties of Zircalof-4 material have been studied. The hardness is increased rapidly at a low degree of cold work and increased rapidly at cold work above 10%. Recrystallization has been completed at 64$0^{\circ}C$, 59$0^{\circ}C$, and 555$^{\circ}C$ in 30%, 60% and 80% cold worked specimen, respectively. The transformation of microstructure with increasing cooling rate after $\beta$-annealing is as follows; coarse Widmanstatten ($\alpha$) longrightarrow fine parallel plate ($\alpha$) longrightarrow martensite ($\alpha$$^{'}$). At the same time, hardness increased with increasing cooling rate. rate.

• Korean Journal of Materials Research
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• v.9 no.6
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• pp.589-594
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• 1999

The effect of $\beta$-heat treatment on th microstructure, mechanical properties and texture in the nuclear fuel cladding of Zircaloy-4 tubes was chosen at 1000, 1100 and 120$0^{\circ}C$, and the tubes were heat-treated by a high frequency vacuum induction furnace. Morphology of the second phase particles and $\alpha$-grain of as-received tubes were markedly changed by heat treatment. The average sizes of second phase particles of as-received and $\beta$-heat treated tubes were 0.1$\mu\textrm{m}$ and 0.076$\mu\textrm{m}$, respectively. However, the average sizes of second phase particles were not much changed in the $\beta$-heated temperatures. With increasing heat treatment temperatures, the 0.2% yield strength and the hoop strength were decreased because of changes in preferred orientation as will as $\alpha$-plate width. Heat treated Zircaloy-4 tubes exhibited texture changes but the preferred orientation of grains still remained.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.3
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• pp.329-345
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• 2023

This paper described a method for analyzing the structural performance of a metal container used for disposing radioactive waste generated during the decommissioning of a nuclear power plant, and numerical analysis results of a method for reinforcing the container. The containers to be analyzed were those that can be used in near-surface and landfill disposal facilities scheduled to be operated at the Gyeongju radioactive waste disposal facility. Structural reinforcement of the container was performed by lattice reinforcement, column reinforcement, and bottom plate reinforcement. Accordingly, a total of 14 reinforcement cases were modeled. The external force causing damage to the container was set equivalent to the impact of a 9-m fall, accounting for the height of the vault at the near-surface disposal facility. The reinforcement methods with a high contribution to the structural performance of the container were concluded to be lattice and column reinforcements.