• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.19-24
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• 2011

These experimental studies are carried out to build a database for analyzing fuel performance in nuclear power plants. In particular, this study focuses on the mechanical and irradiation properties of three kinds of zirconium alloy (Alloy A, Alloy B and Alloy C) irradiated in the HANARO (High-flux Advanced Neutron Application Reactor), one of the leading multipurpose research reactors in the world. Yield strength and ultimate tensile strength were measured to determine the mechanical properties before and after irradiation, while irradiation growth was measured for the irradiation properties. The samples for irradiation testing are classified by texture. For the irradiation condition, all samples were wrapped into the capsule (07M-13N) and irradiated in the HANARO for about 100 days (E > 1.0 MeV, $1.1{\times}10^{21}\;n/cm^2$). These tests and results indicate that the mechanical properties of zirconium alloys are similar whether unirradiated or irradiated. Alloy B has shown the highest yield strength and tensile strength properties compared to other alloys in irradiated condition. Even though each of the zirconium alloys has a different alloying content, this content does not seem to affect the mechanical properties under an unirradiated condition and low fluence. And all the alloys have shown the tendency to increase in yield strength and ultimate tensile strength. Transverse specimens of each of the zirconium alloys have a slightly lower irradiation growth tendency than longitudinal specimens. However, for clear analysis of texture effects, further testing under higher irradiation conditions is needed.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.915-916
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.437-438
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.419-420
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.338-339
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• 2005

• Nuclear Engineering and Technology
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• v.49 no.7
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• pp.1472-1482
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• 2017

Zirconium alloy cladding tube specimens were irradiated at $380^{\circ}C$ up to a fast neutron fluence of $7.5{\times}10^{24}n/m^2$ in a research reactor to investigate the effect of neutron irradiation on hydride reorientation and mechanical property degradation. Cool-down tests from $400^{\circ}C$ to $200^{\circ}C$ under 150 MPa tensile hoop stress were performed. These tests indicate that the irradiated specimens generated a smaller radial hydride fraction than did the unirradiated specimens and that higher hydrogen content generated a smaller radial hydride fraction. The irradiated specimens of 500 ppm-H showed smaller ultimate tensile strength and plastic strain than those characteristics of the 250 ppm-H specimens. This mechanical property degradation caused by neutron irradiation can be explained by tensile hoop stress-induced microcrack formation on the hydrides in the irradiation-damaged matrix and subsequent microcrack propagation along the hydrides and/or through the matrix.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.2
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• pp.161-170
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• 2022

The effect of oxidation time on the characteristics and mechanical properties of spent nuclear fuel cladding was investigated using Raman spectroscopy, tube rupture test, and tensile test. As oxidation time increased, the Raman peak associated with the tetragonal zirconium oxide phase diminished and merged with the Raman peak associated with the monoclinic zirconium oxide phase near 333 cm^-1. Additionally, the other tetragonal zirconium oxide phase peak at 380 cm^-1 decreased after 100 d of oxidation, whereas the zirconium monoclinic oxide peak became the dominant peak. The oxidation time had no effect on the tube rupture pressure of the oxidized zirconium alloy tube. However, the yield and tensile stresses of the oxidized nuclear fuel cladding tube decreased after 100 d of oxidation. The results of the scanning electron microscopy and transmission electron microscopy were represented with the in-situ Raman analysis result for the oxide characteristics generated on the cladding of spent nuclear fuel.

• Archives of Metallurgy and Materials
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• v.65 no.3
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• pp.1063-1067
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• 2020

Dynamic charge carrier transport behavior in the zirconium (Zr) oxide was investigated based on the frequency-dependent capacitance-voltage (C-V) and temperature-dependent current-voltage (I-V) measurements. The Zr oxide was formed on the ZIRLO and newly developed zirconium-based alloy (NDZ) by corrosion in the PWR-simulated loop at 360℃. The corrosion test for 90 days showed that the NDZ exhibits better corrosion resistance than ZIRLO alloy. Based on the C-V measurement, dielectric constant values for the Zr oxide was estimated to be 11.28 and 11.52 for the ZIRLO and NDZ. The capacitance difference between low and high frequency was larger in the ZIRLO than in the NDZ, which was attributed to more mobile electrical charge carriers in the oxide layer on the ZIRLO alloy. The current through the oxide layers on the ZIRLO increased more drastically with increasing temperature than on the NDZ, which indicating that more charge trap sites exist in the ZIRLO than in NDZ. Based on the dynamic charge carrier transport behavior, it was concluded that the electrical charge carrier transport within the oxide layers was closely related with the corrosion behavior of the Zr alloys.

• The Journal of Korean Academy of Prosthodontics
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• v.42 no.6
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• pp.619-627
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• 2004

Statement of problem. Mechanisms of tissue-implant interaction and the effect of the implant surface on the behavior of cells has not yet been clarified. Purpose. This study was performed to investigate the tissue reaction to the titanium alloy submerged into rat peritoneum in vivo. Materials and methods. Titanium alloys (titanium-13Zirconium-6Niobium) were inserted inside the peritoneal cavity of Sprague Dawley rats. After 3 months, the tissue formed around the inserted titanium alloys were examined with a light-microscope. Tissue reaction around the material was analyzed by confocal microscopy to evaluate their biocompatibility in a living body. Results. In in vivo study, foreign body type multinucleated giant cells were found in the fibrous tissue formed as a reaction to the foreign material (4 in 20 cases), but the inflammatory reaction was very weak. After experiment, the contaminants of biomaterials was removed from living tissue. In confocal microscopy, we observed that the staining of vinculin and actin showed mixed appearance. In a few cases, we found that the staining of vinculin and beta-catenin showed the prominent appearance. Conclusion. We found that titanium-13Zirconium-6Niobium alloy was an excellent biomaterial.

• Journal of the Korean Society for Heat Treatment
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• v.35 no.1
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• pp.8-19
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• 2022

The metallic implant materials are widely used in biomedical industries due to their specific mechanical strenth, corrosion registance, and superior biocompatability. These metallic materials, however, suffer from the stress-shielding effect and the generation of artifacts in the magnetic resonance imaging exam. In the present study, we develope a Zr-based alloys for the biomedical implant materials with low elastic modulus and low magnetic susceptibility. The Zr-7Si-xSn alloys were fabricated by an arc melting process. The elastic modulus was 24~31 GPa of the zirconium-based alloy. The average magnetic susceptibility value of the Zr-7Si-xSn alloy was 1.25 × 10^-8cm³g^-1. The average I_corr value of the Zr-7Si-xSn alloy was 0.2 ㎂/cm². The Sn added zirconium alloy, Zr-7Si-xSn, is very interested and attractive as a biomaterial that reduces the stress-shielding effect caused by the difference of elastic modulus between human bone and metallic implant.