• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1748-1751
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• 2017

Reduced activation ferritic/martensitic (RAFM) steels are widely applied as structural materials in the nuclear industry. To investigate hydrogen's effect on RAFM steels' elastic properties and the mechanism of that effect, a procedure of first principles simulation combined with experiment was designed. Density functional theory models were established to simulate RAFM steels' elastic status before and after hydrogen's insertion. Also, experiment was designed to measure the Young's modulus of RAFM steel samples with and without hydrogen charging. Both simulation and experiment showed that the solubility of hydrogen in RAFM steels would decrease the Young's modulus. The effect of hydrogen on RAFM steels' Young's modulus was more significant in water-quenched steels than it was in tempering steels. This indicated that defects inside martensite, considered to be hydrogen traps, could decrease the cohesive energy of the matrix and lead to a decrease of the Young's modulus after hydrogen insertion.

• Journal of Magnetics
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• v.15 no.4
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• pp.204-208
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• 2010

Local wall thinning in pipelines affects the structural integrity of industries like nuclear power plants (NPPs). In the present study, a pulsed eddy current (PEC) differential probe with two excitation coils and two Hall-sensors was fabricated to measure the wall thinning in insulated pipelines. A stainless steel test sample was prepared with a thickness that varied from 1 mm to 5 mm and was laminated by plastic insulation to simulate the pipelines in NPPs. The excitation coils in the probe were driven by a rectangular current pulse, the difference of signals from two Hall-sensors was measured as the resultant PEC signal. The peak value of the detected signal is used to describe the wall thinning. The peak value increased as the thickness of the test sample increased. The results were measured at different insulation thicknesses on the sample. Results show that the differential PEC probe has the potential to detect wall thinning in an insulated NPP pipelines.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.14 no.5
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• pp.397-404
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• 2001

This study describes the effect of γ-ray irradiation on the properties of insulation materials for low voltage cables in a nuclear power plant. The radiation effects were characterized by measuring OIT, FTIR, electrical properties of the irradiated specimens. As a result, they showed the decrease of OIT and the increase of chemical structural defects with the increase of γ-ray amount. Also, the electrical properties such as dielectrical constant, tan $\delta$ and current were changed by aging. These changes may come from the increase of chemical structural defects by $\delta$-ray irradiation.

• Journal of Powder Materials
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• v.17 no.4
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• pp.270-275
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• 2010

Ethylene glycol-based Cu nanofluids were prepared by pulsed wire evaporation (PWE) method. The structural properties of Cu nanoparticles were studied by X-ray diffraction (XRD) and high resolution transmission electron microscopy (HRTEM). The average diameter and Brunauer Emmett Teller (BET) surface area of Cu nanoparticles were about 100 nm and $7.46\;m^2/g$, respectively. The thermal conductivity and viscosity of copper nanofluid were measured as functions of Cu concentration and temperature. As the volume fraction of Cu nanoparticles increased, both the enhanced ratios of thermal conductivity and viscosity of Cu nanofluids increased. As the temperature increased, the enhanced ratio of thermal conductivity increased, but that ratio of viscosity decreased.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.218-227
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• 2016

Background: Steam explosions may occur in nuclear power plants by molten fuel-coolant interactions when the external reactor vessel cooling strategy fails. Since this phenomenon can threaten structural barriers as well as major components, extensive integrity assessment research is necessary to ensure their safety. Method: In this study, the influence of yield criteria was investigated to predict the failure of a reactor cavity under a typical postulated condition through detailed parametric finite element analyses. Further analyses using a geometrically simplified equivalent model with homogeneous concrete properties were also performed to examine its effectiveness as an alternative to the detailed reinforcement concrete model. Results: By comparing finite element analysis results such as cracking, crushing, stresses, and displacements, the Willam-Warnke model was derived for practical use, and failure criteria applicable to the reactor cavity under the severe accident condition were discussed. Conclusion: It was proved that the reactor cavity sustained its intended function as a barrier to avoid release of radioactive materials, irrespective of the different yield criteria that were adopted. In addition, from a conservative viewpoint, it seems possible to employ the simplified equivalent model to determine the damage extent and weakest points during the preliminary evaluation stage.

• Corrosion Science and Technology
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• v.9 no.4
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• pp.164-170
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• 2010

Thermodynamic consideration was performed for Alloy 617 exposed to an impure helium ($H_2$ 20pa, $H_2O$ 0.5pa, $CH_4$ 2pa and CO 5pa) at $950^{\circ}C$. Oxidation power was decreased in the order Al > Ti > Si > Cr > Mn. Decarburization and carburization reactions were available leading to carbon activity decrease and increase, respectively, depending on carbon and Cr activities. The thermodynamic prediction was compared with the experimental results obtained in similar conditions (($H_2$ 20pa, $H_2O$ 0.05pa, $CH_4$ 5pa and CO 2pa) and $950^{\circ}C$) by others for Alloy 617. The driving force for oxidation of Al, Ti and Si is very large to be oxidized at a given impure helium and the environment is actually carburizing towards the structural alloy, which is consistent with this work.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1871-1880
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• 2020

Round robin analyses for vessel failure probabilities due to PTS events are proposed for plant-specific analyses of all types of reactors developed in Korea. Four organizations, that are responsible for regulation, operation, research and design of the nuclear power plant in Korea, participated in the round robin analysis. The vessel failure probabilities from the probabilistic fracture mechanics analyses are calculated to assure the structural integrity of the reactor pressure vessel during transients that are expected to initiate PTS events. The failure probabilities due to various parameters are compared with each other. All results are obtained based on several assumptions about material properties, flaw distribution data, and transient data such as pressure, temperature, and heat transfer coefficient. The realistic input data can be used to obtain more realistic failure probabilities. The various results presented in this study will be helpful not only for benchmark calculations, result comparisons, and verification of PFM codes developed but also as a contribution to knowledge management for the future generation.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1885-1891
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• 2023

The present work aims to fabricate Na_1+xZr₂Si_xP_3-xO₁₂ compound at various calcination temperatures based on the zircon mineral. The fabricated compound was calcinated at 250, 500, and 1000℃. The effect of calcination temperature on the structure, crystal phase, and radiation shielding properties was studied for the fabricated compound. The X-ray diffraction diffractometer demonstrates that, the monoclinic crystal phase appeared at a calcination temperature of 250℃ and 500℃ is totally transformed to a high-symmetry hexagonal crystal phase under a calcination temperature of 1000℃. The radiation shielding capacity was also qualified for the fabricated compounds using the Monte Carlo N-Particle transport code in the g-photons energy interval between 15keV and 122keV. The impacts of calcination temperature on the g-ray shielding behavior were clarified in the present study, where the linear attenuation coefficient was enhanced by 218% at energy of 122keV, when the calcination temperature increased from 250 to 1000℃, respectively.

• Korean Journal of Materials Research
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• v.18 no.12
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• pp.645-649
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• 2008

Graphite for the nuclear reactor is used to the moderator, reflector and supporter in which fuel rod inside of nuclear reactor. Recently, there are many researches has been performed on the various characteristics of nuclear graphite, however most of them are restricted to the structural and the mechanical properties. Therefore we focused on the thermal property of nuclear graphite. This study investigated the thermal emissivity following the oxidation degree of nuclear graphite with IG-11 used as a sample. IG-11 was oxidized to 6% and 11% in air at 5 l/min at $600^{\circ}C$. The porosity and thermal emissivity of the sample were measured using a mercury porosimeter and by an IR method, respectively. The thermal emissivity of an oxidized sample was measured at $100^{\circ}C$, $200^{\circ}C$, $300^{\circ}C$, $400^{\circ}C$ and $500^{\circ}C$. The porosity of the oxidized samples was found to increase as the oxidation degree increased. The thermal emissivity increased as the oxidation degree increased, and the thermal emissivity decreased as the measured temperature increased. It was confirmed that the thermal emissivity of oxidized IG-11 is correlated with the porosity of the sample.

• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.