• Corrosion Science and Technology
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• v.8 no.2
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• pp.89-92
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• 2009

Pure Ti as well as Ti-6Al-4V alloy exhibit excellent properties for dental implant applications. However, for a better biocompatibility it seems important to avoid in the composition the presence of V due to the toxic effects of V ion release. Thus Al and V free and composed of non-toxic element such as Nb, Zr alloys as biomaterials have been developed. Especially, Zr contains to same family in periodic table as Ti. The addition of Zr to Ti alloy has an excellent mechanical properties, good corrosion resistance, and biocompatibility. In this study, the electrochemical characteristics of Ti-Zr alloys for biomaterials have been investigated using by electrochemical methods. Methods: Ti-Zr(10, 20, 30 and 40 wt%) alloys were prepared by arc melting and homogenized for 24 hr at $1000^{\circ}C$ in argon atmosphere. Phase constitutions and microstructure of the specimens were characterized by XRD, OM and SEM. The corrosion properties of the specimens were examined through potentiodynamic test (potential range of -1500 ~ 2000 mV), potentiostatic test (const. potential of 300 mV) in artificial saliva solution by potentiostat (EG&G Co, PARSTAT 2273. USA).

• Korean Journal of Materials Research
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• v.8 no.4
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• pp.368-374
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• 1998

The corrosion characteristics of zirconium alloys have been investigated in various aqueous solutions of LiOH. NaOH, KOH, RbOH. and CsOH at 3S$0^{\circ}C$. The concentrations of solutions were set to 4.3 mmol and 32.Smmol with equimolar $M^+$ and OH . The oxide characterization was performed using TEM on the samples corroded in 32. Smmol LiOH, NaOH, and KOH solution. The samples were prepared to have the same oxide thickness for the pretransition and post- transition regimes. Considering the trend of experimental data, the cation would playa major role in the corrosion process of Zr alloys in alkali hydroxide solutions. The microstructures of the oxides formed in various solutions were quite different. In LiOH solution the oxides grown in pre-transition as well as post-transition had the equiaxed structures with many pores and open grain boundaries. The oxides grown in NaOH solution had the protective columnar structures in pre-transition and the equiaxed structures with many open grain boundaries in post- transition. On the other hand. in KOH solution the columnar structure was maintained from pre- transition to post- transition. It was considered that the cation incorporation into zirconium oxide controlled the oxide characteristics and the corrosion acceleration in alkali hydroxide solutions.

• Proceedings of the Korean Institute of Surface Engineering Conference
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• 2018.06a
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• pp.23-23
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• 2018

In this study, new titanium alloys were prepared by adding elements such as tantalum (Ta), zirconium (Zr) and the like to complement the biological, chemical and mechanical properties of titanium alloys. The Ti-40Ta-xZr ternary alloy was formed on the basis of Ti-40Ta alloy with the contents of Zr in the contents of 0, 3, 7 and 15 wt. %. Plasma electrolytic oxidation (PEO), which combines high-voltage sparks and electrochemical oxidation, is a novel method to form ceramic coatings on light metals such as Ti and its alloys. These oxide film produced by the electrochemical surface treatment is a thick and uniform porous form. It is also composed of hydroxyapatite and calcium phosphate-based phases, so it has the characteristics of bone inorganic, non-toxic and very high bioactivity and biocompatibility. Ti-40Ta-xZr alloys were homogenized in an Ar atmosphere at $1050^{\circ}C$ for 1 hour and then quenched in ice water. The electrochemical oxide film was applied by using a power supply of 280 V for 3 minutes in 0.15 M calcium acetate monohydrate ($Ca(CH_3COO)_2{\cdot}H_2O$) and 0.02 M calcium glycerophosphate ($C_3H_7CaO_6P$) electrolyte. A small amount of 0.0075M zinc acetate and magnesium acetate were added to the electrolyte to enhance the bioactivity. The mechanical properties of the coated surface of Ti-40Ta-xZr alloys were evaluated by Vickers hardness, roughness test, and elastic modulus using nano-indentation, and the surface wettability was evaluated by measuring the contact angle of the coated surface. In addition, cell activation and differentiation were examined by cell culture of HEK 293 (Human embryonic kidney 293) cell proliferation. Surface properties of the alloys were analyzed by scanning electron microscopy(FE-SEM), EDS, and X-ray diffraction analysis (XRD).

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1013-1014
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• 2004

• Journal of Welding and Joining
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• v.33 no.1
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• pp.14-23
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• 2015

Recent developing tendency of nuclear power plant are studied by searching of NDSL, KIPRIS, Science Direct and so on. Welding materials such as low alloyed steels, stainless steels, nickel-based alloys, zirconium alloy and welding methods such as narrow gap welding, laser beam welding, friction stir welding, overlay welding are investigated.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.1
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• pp.115-147
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• 2023

The Fukushima accident in 2011 revealed some major flaws in traditional nuclear fuel materials under accidental conditions. Thus, the focus of research has shifted toward "accident tolerant fuel" (ATF). The aim of this approach is to develop fuel material solutions that lead to improved reactor safety. The application of protective coatings on the surface of nuclear fuel cladding has been proposed as a near-term solution within the ATF framework. Many coating materials are being developed and evaluated. In this article, an overview of different zirconium-based alloys currently in use in the nuclear industry is provided, and their performances in normal and accidental conditions are discussed. Coating materials proposed by different institutions and organizations, their performances under different conditions simulating nuclear reactor environments are reviewed. The strengths and weaknesses of these coatings are highlighted, and the challenges addressed by different studies are summarized, providing a basis for future research. Finally, technologies and methods used to synthesize thin-film coatings are outlined.

• Korean Journal of Materials Research
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• v.18 no.6
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• pp.326-333
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• 2008

Since the 1990s, the second generation of Zirconium alloys containing main alloy compositions of Nb, Sn and Fe have been used as a replacement of Zircaloy-4 (Zr-Sn-Fe-Cr), a first-generation Zirconium alloy, to meet severe and rigorous reactor operating conditions characterized by high-burn-up, high-power and high-pH operations. In this study, the mechanical properties and creep behaviors of Zr-Sn-Fe-Cr and Zr-Nb-Sn-Fe alloys were investigated in a temperature range of $450{\sim}500^{\circ}C$ and in a stress range of $80{\sim}150\;MPa$. The mechanical testing results indicate that the yield and tensile strengths of the Zr-Nb-Sn-Fe alloy are slightly higher compared to those of Zr-Sn-Fe-Cr. This can be explained by the second phase strengthening of the $\beta$-Nb precipitates. The creep test results indicate that the stress exponent for the steady-state creep rate decreases with the increase in the applied stress. However, the stress exponent of the Zr-Sn-Fe-Cr alloy is lower than that of the Zr-Nb-Sn-Fe alloy in a relatively high stress range, whereas the creep activation energy of the former is slightly higher than that of the latter. This can be explained by the dynamic deformation aging effect caused by the interaction of dislocations with Sn substitutional atoms. A higher Sn content leads to a lower stress exponent value and higher creep activation energy.

• Korean Journal of Materials Research
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• v.12 no.4
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• pp.317-323
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• 2002

Effect of V and Sb content on characteristics of ${\beta}\;to\;{\alpha}$ phase transformation in Zr-0.84Sn alloy has been studied using optical microscopy and transmission electron microscopy. As V content increased, the ${\beta}{\to}{\beta}+{\alpha}$ transformation temperature was lowered, thus allowing the width of $\alpha$-lath in air-cooled Zr-0.86Sn-0.40V alloy to be fine. The width of ${\alpha}$-lath in air-cooled Zr-0.84Sn-xSb, however, was rarely changed with Sb content. The ${\beta}\;to\;{\alpha}$ transformed microstructures of water-quenched Zr-0.84Sn, Zr-0.84Sn-0.10V and Zr-0.84Sn-0.19V alloys were mainly slipped martensite. On the other hand, those of wafter-quenched Zr-0.86Sn-0.40V and Zr-0.85Sn-0.05Sb alloys were predominantly twinned martensite. In case of water-quenched Zr-0.85Sn-0.12Sb and Zr-0.84Sn-0.17Sb alloys, basketweave structure was observed. The transition of slipped martensite to twinned martensite in Zr-0.84Sn-xV alloys and the transition of twinned martensite to basketweave structure in Zr-0.84Sn-xSb alloys were due to the decrease of $M_s$ temperature.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.570-587
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• 1993

Recently a number of severe fuel degradation events, seemingly due to internal secondary hydriding, have been reported. This paper reviews internal hydriding of defected zircaloy cladding. First, the history of zircaloy cladding development and the environment of the zircaloys in service in the nuclear reactor are introduced. Fundamental aspects of zircaloy hydriding, such as hydrogen permeability in zirconium oxide, terminal solubility and precipitation in zirconium and its alloys, and the deleterious effect of hydrides are reviewed. The mechanism of massive internal hydriding in defected zircaloy fuel rods is qualitatively described based on the observed phenomena. Significant factors affecting the hydriding process are discussed. A quantitative model for the massive hydriding as a part of an effort to mitigate fuel degradation is briefly mentioned and necessary information and recommended future work for improvement of the model are outlined.

• Korean Journal of Materials Research
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• v.29 no.8
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• pp.483-490
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• 2019

Currently, the Korean nuclear industry uses ZIRLO as material for nuclear fuel cladding(zirconium alloy). KEPCO Nuclear Fuel is in the process of developing a HANA alloy to enable domestic production of cladding. Cladding manufacture involves multistage heat treatments and pickling processes, the latter of which is vital for the removal of defects and impurities on the cladding surface. SMUT that forms on the cladding surface during such pickling process is a source of surface defects during heat treatment and post-treatment processes if not removed. This study analyzes ZIRLO, HANA-4, and HANA-6 alloy claddings to extensively study the SEM/EDS, XRD, and particle size characteristics of SMUT, which are second phase particles that are formed on the cladding surface during pickling processes. Using the analysis results, this study observes SMUT formation characteristics according to Nb concentration in Zr alloys during the washing process following the pickling process. In addition, this study observes SMUT removal characteristics on cladding surfaces according to concentrations of nitric acid and hydrofluoric acid in the acid solution.