• Corrosion Science and Technology
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• v.2 no.6
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• pp.253-259
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• 2003

Zirconium based alloys have been extensively used as a cladding material for fuel rods in nuclear reactors, due to their low thermal neutron absorption cross-section, excellent corrosion resistance and good mechanical properties at high temperatures. However, a cladding material for fuel rods in nuclear reactors was contact water during long time at high-temperature, so it is necessary to improve the wear and corrosion resistance of the fuel cladding, At ambient environment, there are few data or paper on the characteristic of corrosion in chloride solution and acidic solution. The specimens used in this work are pure Zr and Zircaloy-4. Zircaloy-4 is a specific zirconium-based alloy containing, on a weight percent basis, 1.4% Sn, 0.2% Fe, 0.1% Cr. Pitting corrosion resistance of two alloys by ASTM G48 is higher than that of electrochemical method. Passive film formed on Zircaloy-4 is mainly composed of $ZrO_2$, metallic Sn, and iron species regardless of formation environments. Also, passive film formed on Zr alloys shows n-type semiconductic property on the base of Mott-Schottky plot.

• Transactions of the Korean hydrogen and new energy society
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• v.9 no.3
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• pp.101-110
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• 1998

The change of hydrogen storage characteristics by substituting zirconium for a portion of titanium in Ti-Cr-V alloys has been studied. The zirconium substitution decreased the plateau pressure and hysteresis of the PC isotherm. However, it decreased the hydrogen storage capacity and increased slopping in PC isotherm by forming $Cr_2Zr$ phase. By modifying the composition ratio of titanium to chromium, thereby suppressing the formation of $Cr_2Zr$ phase, we got an alloy having very high hydrogen storage capacity. The heat treatment of the alloys improved the flatness of plateau very much without a decrease in the maximum and the effective hydrogen storage capacities.

• Nuclear Engineering and Technology
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• v.51 no.6
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• pp.1589-1595
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• 2019

In this paper, the CLAM steel strengthened by micro-scale Y-Zr-O was prepared by vacuum induction melting followed by electroslag remelting (VIM-ESR). Yttrium (Y) and zirconium (Zr) were easy to aggregates into massive yttrium-zirconium-rich inclusions in the steel melted by vacuum induction melting (VIM), which would interrupt the continuity of the matrix and reduce the mechanical properties of steel. Micron-sized Y-Zr-O inclusions would be produced with the removal of original blocky Y-Zr-rich inclusions and the submicron-sized inclusions smaller than $0.2{\mu}m$ could be retained in the steel. The small grain size and the better refinement and distribution uniformity of Y-Zr-O inclusions after remelting would be responsible for the better yield strength and toughness. For VIM-ESR alloy, the ultimate tensile strength is 749 MPa and the yield strength is 642 MPa at room temperature, meanwhile they are 391 MPa and 367 MPa at $600^{\circ}C$, respectively. Meanwhile, the ductile-brittle transition temperature (DBTT) reduced from $-43^{\circ}C$ (VIM) to $-76^{\circ}C$ (VIM-ESR).

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1579-1587
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• 2022

The effects of hydride amount (20-850 wppm), orientation (circumferential and radial), and temperature (room temperature, 100 ℃, 200 ℃) on the axial mechanical properties of Zircaloy-4 cladding were comprehensively examined. The fraction of radial hydride fraction in the cladding was quantified using PROPHET, an in-house radial hydride fraction analysis code. Uniaxial tensile tests (UTTs) were conducted at various temperatures to obtain the axial mechanical properties. Hydride orientation has a limited effect on the axial mechanical behavior of hydrided Zircaloy-4 cladding. Ultimate tensile stress (UTS) and associated uniform elongation demonstrated limited sensitivity to hydride content under UTT. Statistical uncertainty of UTS was found small, supporting the deterministic approach for the load-failure analysis of hydrided Zircaloy-4 cladding. These properties notably decrease with increasing temperature in the tested range. The dependence of yield strength on hydrogen content differed from temperature to temperature. The ductility-related parameters, such as total elongation, strain energy density (SED), and offset strain decrease with increasing hydride contents. The abrupt loss of ductility in UTT was found at ~700 wppm. Demonstrating a strong correlation between total elongation and offset strain, SED can be used as a comprehensive measure of ductility of hydrided zirconium alloy.

• Membrane Journal
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• v.34 no.2
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• pp.140-145
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• 2024

In modern times, when a change in the energy paradigm is required, hydrogen is an attractive energy source. Among these hydrogen purification technologies, technology using a membrane is attracted attention as a technology that can purify high purity hydrogen at low cost. However, palladium(Pd), which is mostly used because of its excellent hydrogen separation performance, is very expensive, so a replacement material is needed. In this study, a alloy membrane was manufactured from an alloy of niobium (Nb), which has high hydrogen permeability but is weak to hydrogen embrittlement, and nickel (Ni) and zirconium (Zr), which have low hydrogen permeability but are highly durable. Hydrogen permeation characteristics were confirmed under conditions of 350~450 ℃ at 1 to 4 bar. The maximum hydrogen permeation flux was 0.69 ml/cm²/min for the Ni₄₈Nb₃₂Zr₂₀ alloy membrane without Pd coating, and 13.05 ml/cm²/min for the Pd coated alloy membrane.

• Journal of the Korean Society for Heat Treatment
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• v.17 no.5
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• pp.271-277
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• 2004

Zr-xNb alloys (x = 0.2, 0.8, 1.5 wt.%) were prepared to study the characteristics of the phase transformation in Zr-Nb system. The samples were heat treated at ${\beta}$-temperature ($1020^{\circ}C$) for 20 min and then cooled with different cooling rate. The microstructures of the specimens having the same compositions were changed with cooling rate and Nb content. The Widmanst$\ddot{a}$tten structure was observed on the furnace-cooled sample. The relationship between ${\alpha}$-Widmanst$\ddot{a}$tten and ${\beta}$-phase was the {0001}${\alpha}$//{110}${\beta}$, <11$\bar{2}$0>//<111>. The ${\beta}$-phase in Widmanst$\ddot{a}$tten structure of Zr-Nb alloys containing Nb more than solubility limit was identified as ${\beta}_{Zr}$ phase which was a stable phase at high temperature. In the water quenched samples, two kinds of martensite structures were observed depending on the Nb-concentration. The lath martensite was formed in Zr-0.2, 0.8 wt.% Nb alloys and the plate martensite having twins was formed in Zr-1.5 wt.% Nb alloy.

• Korean Journal of Materials Research
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• v.11 no.10
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• pp.879-888
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• 2001

The effects of final annealing temperature on the microstructure and creep characteristics were investigated for the Zr-lNb-0.2X (X=0, Mo, Cu) and Zr-lNb- 1Sn-0.3Fe-0.1X (X=0, Mo, Cu) alloys. The microstructures were observed by using TEM/EDS, and grain size and distributions of precipitates were analyzed using a image analyzer. The creep test was performed at $400^{\circ}C$ under applied stress of 150 MPa for 10 days. The $\beta$-Zr was observed at annealing temperature above $600^{\circ}C$. In the temperature above$ 600^{\circ}C$, the grain sizes of both alloy systems appeared to be increased with increasing the final annealing temperature. The creep strengths of Zr-1Nb-1Sn-0.3Fe-0.1X alloys were higher than those of Zr-1Nb-0.2X ones due to the effect of solid solution hardening by Sn in Zr-lNb-lSn-0.3Fe-0.1X alloy system. Also, Mo addition showed the strong effect of precipitate hardening in both alloy systems. The creep strength rapidly decreased with increasing the annealing temperature up to $600^{\circ}C$. However, a superior creep resistance was obtained in the sample that annealed to have a second phase of $\beta$-Zr. It was considered that the appearance of $\beta$-Zr would play an important role in the strengthening mechanism of creep deformation.

• The Journal of Korean Academy of Prosthodontics
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• v.44 no.6
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• pp.740-750
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• 2006

Statement of problems. In an attempt to reduce screw loosening, dry lubricant coatings such as pure gold or tefron have been applied to the abutment screw. However, under repeated tightening and loosening procedures, low wear resistance and adhesion strength of coating material produced free particles on the surface of abutment screw and increased frictional resistance resulting in screw tightening problems. Purpose. The aim of this study was to compare friction coefficient, adhesion strength, vickers hardness and evaluate coating surface of titanium alloy specimens coated with TiN(titanium nitride), ZrN(zirconium nitride) and WC(tungsten carbide). Material and method. Titanium alloy(Ti-6Al-4V) discs of 12mm in diameter and 1mm in thickness divided into 4 groups. TiN, ZrN and WC was coated for the specimens of 3 groups respectively, and those of 1 group were not coated. Each group was made up of 4 specimens. In this study, sputtering method was used among the PVD(Physical Vapor Deposition) techniques available for TiN, ZrN and WC coatings. Friction coefficient, adhesion strength, vickers hardness and coating surface of 4 groups were measured. Results. 1. For all three coating conditions, friction coefficient was significantly decreased. Especially, ZrN coated surface showed the lowest value. $TiN(0.39{\pm}0.02)$, $ZrN(0.24{\pm}0.01)$, $WC(0.31{\pm}0.03)$. 2. TiN coating showed the highest adhesion strength, however ZrN coating had the lowest value. $TiN(25.3N{\pm}1.6)$, $ZrN(14.8N{\pm}0.6)$, $ WC(18.4N{\pm}0.7)$. 3. Vickers hardness of all three coatings was remarkably increased as compared with that of none coated specimen. TiN coating had the highest Vickers hardness, however WC coating showed the lowest value. $TiN(1865.2{\pm}33.8)$, $ZrN(1814.4{\pm}18.6)$, $WC(1008.5{\pm}35.9)$. 4. The ZrN or WC coated specimen showed a homogeneous and smooth surface, however the rough surface with defects was observed for TiN coating. Conclusions. When TiN, ZrN and WC coating applied to the abutment screw, frictional resistance would be reduced, as a result, the greater preload and prevention of the screw loosening could be expected.

• Transactions of Materials Processing
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• v.17 no.5
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• pp.364-372
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• 2008

In this study, the circumferential creep behaviors ofpilgered advanced Zirconium alloy tubes such as Zr-Nb-O and Zr-Nb-Sn-Fe were investigated in the temperature range of $400\sim500^{\circ}C$ and in the stress range of 80$\sim$150MPa. The test results indicate that the stress exponent for the steady-state creep rate of the Zr-Nb-Sn-Fe alloy decreases with the increase of stress(from 6$\sim$7 to 4), while that of the Zr-Nb-O alloy is nearly independent of stress(5$\sim$6). The activation energy of creep deformation is found to be nearly the same as the activation energy for Zr self diffusion. This indicates that the creep deformation may be controlled by dislocation climb mechanism in Zr-Nb-O. On the other hand, the transition of stress exponent(from 6-7 to 4) in Zr-Nb Sn-Fe strongly suggests the transition of the rate controlling mechanism at high stresses. The lower stress exponent at high stresses in Zr-Nb-Sn-Fe can be explained by the dynamic deformation aging effect caused by interaction of dislocations with Sn substitutional atoms.

• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.16-25
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• 2016

In severe loss of coolant accidents (LOCA), similar to those experienced at Fukushima Daiichi and Three Mile Island Unit 1, the zirconiumalloy fuel claddingmaterials are rapidlyheateddue to nuclear decay heating and rapid exothermic oxidation of zirconium with steam. This heating causes the cladding to rapidly react with steam, lose strength, burst or collapse, and generate large quantities of hydrogen gas. Although maintaining core cooling remains the highest priority in accident management, an accident tolerant fuel (ATF) design may extend coping and recovery time for operators to restore emergency power, and cooling, and achieve safe shutdown. An ATF is required to possess high resistance to steam oxidation to reduce hydrogen generation and sufficient mechanical strength to maintain fuel rod integrity and core coolability. The initiative undertaken by Electric Power Research Institute (EPRI) is to demonstrate the feasibility of developing an ATF cladding with capability to maintain its integrity in $1,200-1,500^{\circ}C$ steam for at least 24 hours. This ATF cladding utilizes thin-walled Mo-alloys coated with oxidation-resistant surface layers. The basic design consists of a thin-walled Mo alloy structural tube with a metallurgically bonded, oxidation-resistant outer layer. Two options are being investigated: a commercially available iron, chromium, and aluminum alloy with excellent high temperature oxidation resistance, and a Zr alloy with demonstratedcorrosionresistance.Asthese composite claddings will incorporate either no Zr, or thin Zr outer layers, hydrogen generation under severe LOCA conditions will be greatly reduced. Key technical challenges and uncertainties specific to Moalloy fuel cladding include: economic core design, industrial scale fabricability, radiation embrittlement, and corrosion and oxidation resistance during normal operation, transients, and severe accidents. Progress in each aspect has been made and key results are discussed in this document. In addition to assisting plants in meeting Light Water Reactor (LWR) challenges, accident-tolerant Mo-based cladding technologies are expected to be applicable for use in high-temperature helium and molten salt reactor designs, as well as nonnuclear high temperature applications.