• Journal of the Korean Magnetics Society
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• v.12 no.5
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• pp.189-194
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• 2002

Microstructure and magnetic properties of Sm-Co sintered magnet were investigated with the variation of Zr content and their solution treatment and aging temperatures. The fraction of eutectic structure and the size of eutectic area decreased with increasing x value of cast Sm(C $O_{.688-x}$F $e_{.242}$C $u_{.07}$Z $r_{x}$)$_{7.404}$ alloys. On the other hand, x=0.022 ingot had finer dendritic structure compared to the other alloys. The sintered magnet of Sm(C $O_{.688-x}$F $e_{.242}$C $u_{.07}$Z $r_{x}$)$_{7.404}$ had well defined cell structure which is composed of cell boundary Sm $Co_{5}$ and cell interior S $m_2$Co/ssub 17/ phase. Cell boundary Sm $Co_{5}$ phase has 20nm thickness and its relative angle was 120$^{\circ}$ in x=0.018 and 0.022 alloys. Cell size was decreased with increasing Zr contents. But, x=0.026 alloy has diffuse cell boundary and irregular shape compared to x=0.022 and 0.018 alloys. Maximum value of coercive force and maximum energy Product were obtained from x=0.022 alloys. Optimum solution treatment temperature of Sm(C $O_{.688-x}$F $e_{.242}$C $u_{.07}$Z $r_{x}$)$_{7.404}$ alloy was 1170 $^{\circ}C$ and 1st aging temperature of two step aging process for higher coercivity was 850 $^{\circ}C$.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.734-743
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• 2017

In order to investigate the swelling behavior and fuel-cladding interaction mechanism of U-10Zr alloy metallic fuel at very low burnup, an irradiation experiment was concisely designed and conducted on the China Mianyang Research Reactor. Two types of irradiation samples were designed for studying free swelling without restraint and the fuel-cladding interaction mechanism. A new bonding material, namely, pure aluminum powder, was used to fill the gap between the fuel slug and sample shell for reducing thermal resistance and allowing the expansion of the fuel slug. In this paper, the concise irradiation rig design is introduced, and the neutronic and thermal-hydraulic analyses, which were carried out mainly using MCNP (Monte Carlo N-Particle) and FLUENT codes, are presented. Out-of-pile tests were conducted prior to irradiation to verify the manufacturing quality and hydraulic performance of the rig. Nondestructive postirradiation examinations using cold neutron radiography technology were conducted to check fuel cladding integrity and swelling behavior. The results of the preliminary examinations confirmed the safety and effectiveness of the design.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2016.10a
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• pp.89-90
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• 2016

• Korean Journal of Materials Research
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• v.21 no.3
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• pp.161-167
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• 2011

The stress-strain responses and oxidation properties of cold-rolled Zr-1.5Nb-O and Zr-1.5Nb-O-S alloys were studied. The U.T.S. (ultimate tensile strength) of cold-rolled Zr-1.5Nb-O-S alloy with 160 ppm sulfur (765 MPa) were greater than that of Zr-1Nb-1Sn-0.1Fe alloy (750 MPa), achieving an excellent mechanical strength even after the elimination of Sn, an effective solution strengthening element. The addition of sulfur increased the strength at the expense of ductility. However, the ductile fracture behavior was observed both in Zr-Nb-O and Zr-Nb-O-S alloys. The beneficial effect of sulphur on the strengthening was observed in the cold rolled Zr-1.5Nb-O-S alloys. The activation volume of cold-rolled Zr-1.5Nb decreased with sulfur content in the temperature region of dynamic strain aging associated with oxygen atoms. Insensitivity of the activation volume to the dislocation density and the decrease of the activation volume at a higher temperature where the dynamic strain aging occurs support the suggestion linking the activation volume with the activated bulge of dislocations limited by segregation of oxygen and sulfur atoms. The addition of sulfur was also found to improve the oxidation resistance of Zr-Nb-O alloys.

• Corrosion Science and Technology
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• v.8 no.4
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• pp.162-169
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• 2009

Pure titanium and its alloys are drastically used in implant materials due to their excellent mechanical properties, high corrosion resistance and good biocompatibility. However, the widely used Ti-6Al-4V is found to release toxic ions (Al and V) into the body, leading to undesirable long-term effects. Ti-6Al-4V has much higher elastic modulus than cortical bone. Therefore, titanium alloys with low elastic modulus have been developed as biomaterials to minimize stress shielding. For this reason, Ti-30Ta-xZr alloy systems have been studied in this study. The Ti-30Ta containing Zr(5, 10 and 15 wt%) were 10 times melted to improve chemical homogeneity by using a vacuum furnace and then homogenized for 24 hrs at $1000^{\circ}C$. The specimens were cut and polished for corrosion test and Ti coating and then coated with TiN, respectively, by using DC magnetron sputtering method. The analyses of coated surface were carried out by field emission scanning electron microscope(FE-SEM). The electrochemical characteristics were examined using potentiodynamic (- 1500 mV~+ 2000 mV) and AC impedance spectroscopy(100 kHz~10 mHz) in 0.9% NaCl solution at $36.5{\pm}1^{\circ}C$. The equiaxed structure was changed to needle-like structure with increasing Zr content. The surface defects and structures were covered with TiN/Ti coated layer. From the polarization behavior in 0.9% NaCl solution, The corrosion current density of Ti-30Ta-xZr alloys decreased as Zr content increased, whereas, the corrosion potential of Ti-30Ta-xZr alloys increased as Zr content increased. The corrosion resistance of TiN/Ti-coated Ti-30Ta-xZr alloys were higher than that of the TiN-coated Ti-30Ta-xZr alloys. From the AC impedance in 0.9% NaCl solution, polarization resistance($R_p$) value of TiN/Ti coated Ti-30Ta-xZr alloys showed higher than that of TiN-coated Ti-30Ta-xZr alloys.

• Journal of the Korean Ceramic Society
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• v.25 no.3
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• pp.225-230
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• 1988

The strength and thermal shock resistance of $Al_2O_3-ZrO_2$ composites have been studied. The tetragonal $ZrO_2$ powder containing 1 mol.% $Y_2O_3$ and monoclinic $ZrO_2$ powder were prepared by coprecipitation method and subsequently mixed with $Al_2O_3$ powder and granulated by sieving. Duplex composites were prepared by dry mixing matrix agglomerate with 15 to 30 vol.% of dispersion agglomerate, followed by pressing and sintering at 1$600^{\circ}C$ for1 hr. These $Al_2O_3-ZrO_2$ 2 composites having heterogeneous structure showed improved thermal shock behaviors because of the microcracking and pores in dispersed granules, and compressive stresses around dispersed granules resulting from $ZrO_2$ transformation.

• Proceedings of the IEEK Conference
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• 2001.10a
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• pp.179-182
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• 2001

The chloride volatilization method for the recovery of zirconium and removal of uranium from zirconium containing metallic wastes formed in spent fuel reprocessing was studied using the simulated alloy waste, i.e. the mixture of Zr foil and UO$_2$/U$_3$O$_{8}$ powder. When the simulated waste was heated to react with chlorine gas at 350- l00$0^{\circ}C$, the zirconium metal changed to volatile ZrCl$_4$showing high volatility ratio (Vzr) of 99%. The amount of volatilized uranium increases at higher temperatures causing lowering of decontamination factor (DF) of uranium. This is thought to be caused by the chlorination of UO$_2$ with ZrCl$_4$vapor. The highest DF value of 12.5 was obtained when the reaction temperature was 35$0^{\circ}C$. Addition of 10 vol.% oxygen gas into chlorine gas was effective for suppressing the volatilization of uranium, while the volatilization ratio of zirconium was decreased to 68% with the addition of 20 vol.% oxygen. In the case of the mixture of Zr foil and U$_3$O$_{8}$, the V value of uranium showed minimum (44%) at 40$0^{\circ}C$ with chlorine gas giving the highest DF value 24.3. When the 10 vol.% oxygen was added to chlorine gas, the V value of zirconium decreased to 82% at $600^{\circ}C$, but almost all the uranium volatilized (Vu=99%), which may be caused by the formation of volatile uranium chlorides under oxidative atmosphere.ere.

• The Journal of Advanced Prosthodontics
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• v.11 no.2
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• pp.105-111
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• 2019

PURPOSE. Limited data is available regarding the differences for possible microleakage problems and fitting accuracy of zirconia versus titanium abutments with various connection designs. The purpose of this in vitro study was to investigate the effect of connection design and abutment material on the sealing capability and fitting accuracy of abutments. MATERIALS AND METHODS. A total of 42 abutments with different connection designs [internal conical (IC), internal tri-channel (IT), and external hexagonal (EH)] and abutment materials [titanium (Ti) and zirconia (Zr)] were evaluated. The inner parts of implants were inoculated with $0.7{\mu}L$ of polymicrobial culture (P. gingivalis, T. forsythia, T. denticola and F. nucleatum) and connected with their respective abutments under sterile conditions. The penetration of bacteria into the surrounding media was assessed by the visual evaluation of turbidity at each time point and the number of colony forming units (CFUs) was counted. The marginal gap at the implant- abutment interface (IAI) was measured by scanning electron microscope. The data sets were statistically analyzed using Kruskal-Wallis followed by Mann-Whitney U tests with the Bonferroni-Holm correction (${\alpha}=.05$). RESULTS. Statistically significant difference was found among the groups based on the results of leaked colonies (P<.05). The EH-Ti group characterized by an external hexagonal connection were less resistant to bacterial leakage than the groups EH-Zr, IT-Zr, IT-Ti, IC-Zr, and IC-Ti (P<.05). The marginal misfit (in ${\mu}m$) of the groups were in the range of 2.7-4.0 (IC-Zr), 1.8-5.3 (IC-Ti), 6.5-17.1 (IT-Zr), 5.4-12.0 (IT-Ti), 16.8-22.7 (EH-Zr), and 10.3-15.4 (EH-Ti). CONCLUSION. The sealing capability and marginal fit of abutments were affected by the type of abutment material and connection design.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2556-2566
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• 2023

A first-principle approach within the framework of density functional theory was employed to study the effect of vacancy defects and fission products (FPs) doping on the mechanical, electronic, and thermodynamic properties of uranium monocarbide (UC). Firstly, the calculated vacancy formation energies confirm that the C vacancy is more stable than the U vacancy. The solution energies indicate that FPs prefer to occupying in U site rather than in C site. Zr, Mo, Th, and Pu atoms tend to directly replace U atom and dissolve into the UC lattice. Besides, the results of the mechanical properties show that U vacancy reduces the compressive and deformation resistance of UC while C vacancy has little effect. The doping of all FPs except He has a repairing effect on the mechanical properties of U_1-xC. In addition, significant modifications are observed in the phonon dispersion curves and partial phonon density of states (PhDOS) of UC_1-x, Zr_xU_1-xC, Mo_xU_1-xC, and Rh_xU_1-xC, including narrow frequency gaps and overlapping phonon modes, which increase the phonon scattering and lead to deterioration of thermal expansion coefficient (α_V) and heat capacity (C_p) of UC predicted by the quasi harmonic approximation (QHA) method.

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1547-1554
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• 2017

VESTA (Verification of Ex-vessel corium STAbilization) and VESTA-S (-small) test facilities were constructed at the Korea Atomic Energy Research Institute in 2010 to perform various corium melt experiments. Since then, several tests have been performed for the verification of an ex-vessel core catcher design for the EU-APR1400. Ablation tests of an impinging $ZrO_2$ melt jet on a sacrificial material were performed to investigate the ablation characteristics. $ZrO_2$ melt in an amount of 65-70 kg was discharged onto a sacrificial material through a well-designed nozzle, after which the ablation depths were measured. Interaction tests between the metallic melt and sacrificial material were performed to investigate the interaction kinetics of the sacrificial material. Two types of melt were used: one is a metallic corium melt with Fe 46%, U 31%, Zr 16%, and Cr 7% (maximum possible content of U and Zr for C-40), and the other is a stainless steel (SUS304) melt. Metallic melt in an amount of 1.5-2.0 kg was delivered onto the sacrificial material, and the ablation depths were measured. Penetration tube failure tests were performed for an APR1400 equipped with 61 in-core instrumentation penetration nozzles and extended tubes at the reactor lower vessel. $ZrO_2$ melt was generated in a melting crucible and delivered down into an interaction crucible where the test specimen is installed. To evaluate the tube ejection mechanism, temperature distributions of the reactor bottom head and in-core instrumentation penetration were measured by a series of thermocouples embedded along the specimen. In addition, lower vessel failure tests for the Fukushima Daiichi nuclear power plant are being performed. As a first step, the configuration of the molten core in the plant was investigated by a melting and solidification experiment. Approximately 5 kg of a mixture, whose composition in terms of weight is $UO_2$ 60%, Zr 10%, $ZrO_2$ 15%, SUS304 14%, and $B_4C$ 1%, was melted in a cold crucible using an induction heating technique.