• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.218-222
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• 2018

An oxide-dispersion-strengthened (ODS) layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide ($Y_2O_3$)-coated Zircaloy-4 tube to induce the penetration of $Y_2O_3$ particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at $380^{\circ}C$, and from 385 to 470 MPa at $500^{\circ}C$. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to $830^{\circ}C$ at a heating rate of $5^{\circ}C/s$ and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties.

• Korean Journal of Materials Research
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• v.29 no.4
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• pp.271-276
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• 2019

Oxide-dispersion-strengthened (ODS) alloy has been developed to increase the mechanical strength of metallic materials; such an improvement can be realized by distributing fine oxide particles within the material matrix. In this study, the ODS layer was formed in the surface region of Zr-based alloy tubes by laser beam treatment. Two kinds of Zr-based alloys with different alloying elements and microstructures were used: KNF-M (recrystallized) and HANA-6 (partial recrystallized). To form the ODS layer, $Y_2O_3$-coated tubes were scanned by a laser beam, which induced penetration of $Y_2O_3$ particles into the substrates. The thickness of the ODS layer varied from 20 to $55{\mu}m$ depending on the laser beam conditions. A heat affected zone developed below the ODS layer; its thickness was larger in the KNF-M alloy than in the HANA-6 alloy. The ring tensile strengths of the KNF-M and HANA-6 alloy samples increased more than two times and 20-50%, respectively. This procedure was effective to increase the strength while maintaining the ductility in the case of the HANA-6 alloy samples; however, an abrupt brittle facture was observed in the KNF-M alloy samples. It is considered that the initial microstructure of the materials affects the formation of ODS and the mechanical behavior.

• Tribology and Lubricants
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• v.22 no.3
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• pp.144-148
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• 2006

The experimental investigation was performed to find the associated changes in characteristics of fretting wear with various water temperatures. The fretting wear tests were carried out using the zirconium alloy tubes and the grids with increasing the water temperature. The tube materials in water of $20^{\circ}C,\;50^{\circ}C\;and\;80^{\circ}C$ were tested with the applied load of 20 N and the relative amplitude of $200{\mu}m$. The worn surfaces were observed by SEM, EDX analysis and 2D surface profiler. As the water temperature increased, the wear volume was decreased, but oxide layer was increased on the worn surface. The abrasive wear mechanism was observed at water temperature of $20^{\circ}C$ and adhesive wear mechanism occurred at water temperature of $50^{\circ}C,\;80^{\circ}C$. As the water temperature increased, surface micro-hardness was decreased, but wear depth and wear width were decreased due to increasing stick phenomenon. Stick regime occurred due to the formation of oxide layer on the worn surface with increasing water temperatures

• Nuclear Engineering and Technology
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• v.52 no.10
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• pp.2299-2305
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• 2020

Zr alloy specimens were coated with Cr-Al alloy to enhance their resistance to oxidation. The coated samples were oxidized at 1200 ℃ in a steam environment for 300 s and showed extremely low oxidation when compared to uncoated Zr alloy specimens. The microstructure and elemental distribution of the oxides formed on the surface of Cr-Al alloys have been investigated by transmission electron microscopy (TEM) and X-ray photoelectron spectroscopy (XPS). A very thin protective layer of Cr₂O₃ formed on the outer surface of the Cr-Al alloy, and a thin Al₂O₃ layer was also observed in the Cr-Al alloy matrix, near the surface. Our results suggest that these two oxide layers near the surface confers excellent oxidation resistance to the Cr-Al alloy. Even after exposure to a high temperature of 1200 ℃, inter-diffusion between the Cr-Al alloy and the Zr alloy occurred in very few regions near the interface. Analysis of the inter-diffusion layer by high-resolution transmission electron microscopy (HRTEM) and energy dispersive X-ray spectroscopy (EDS) measurement confirmed its identity as Cr₂Zr.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.416-424
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• 2018

Microstructure of oxide formed on Zr-Nb-Sn tube sample was intensively examined by scanning transmission electron microscopy after exposure to simulated primary water chemistry conditions of various concentrations of Zn (0 or 30 ppb) and dissolved hydrogen ($H_2$) (30 or 50 cc/kg) for various durations without applying desirable heat flux. Microstructural analysis indicated that there was no noticeable change in the microstructure of the oxide corresponding to water chemistry changes within the test duration of 100 days (pretransition stage) and no significant difference in the overall thickness of the oxide layer. Equiaxed grains with nano-size pores along the grain boundaries and microcracks were dominant near the water/oxide interface, regardless of water chemistry conditions. As the metal/oxide interface was approached, the number of pores tended to decrease. However, there was no significant effect of $H_2$ concentration between 30 cc/kg and 50 cc/kg on the corrosion of the oxide after free immersion in water at $360^{\circ}C$. The adsorption of Zn on the cladding surface was observed by X-ray photoelectron spectroscopy and detected as ZnO on the outer oxide surface. From the perspective of $OH^-$ ion diffusion and porosity formation, the absence of noticeable effects was discussed further.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.273-280
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• 2012

The electrochemical behaviors of 10 g-scale fresh and oxidized Zircaloy-4 cladding hulls were examined in $500^{\circ}C$ LiCl-KCl molten salts to confirm the feasibility of the electrorefining process for the treatment of hull wastes. In the results of measuring the potential-current response using a stainless steel basket filled with oxidized Zircaloy-4 hull specimens, the oxidation peak of Zr appears to be at -0.7 to -0.8 V vs. Ag/AgCl, which is similar to that of fresh Zircaloy-4 hulls, while the oxidation current is found to be much smaller than that of fresh Zircaloy-4 hulls. These results are congruent with the outcome of current-time curves at -0.78 V and of measuring the change in the average weight and thickness after the electrochemical dissolution process. Although the oxide layer on the surface affects the uniformity and rate of dissolution by decreasing the conductivity of Zircaloy-4 hulls, electrochemical dissolution is considered to occur owing to the defect of the surface and phase properties of the Zr oxide layer.

• Nuclear Engineering and Technology
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• v.19 no.2
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• pp.122-129
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• 1987

Experiments and numerical analysis have been performed to investigate the effect of preoxidation by oxidizing Zircaloy-4 specimens at a higher temperature after a period of exposure at a lower temperature. The oxidation experiments were performed between $700^{\circ}C$ and 85$0^{\circ}C$ after Preoxidation at $650^{\circ}C$ in a steam and water mixture for 600 seconds and 1,800 seconds. As the thickness of preoxidized layer increased, the oxidation rate of preoxidized specimens at higher temperature became lower than that of as-received claddings. A transition region of oxidation rate exist in the preoxidized specimens, and the region disappeared rapidly as the oxidation temperature increased. This effect appeared more clearly at lower temperatures. According to the results of numerical analysis performed in this study, the growth rate of oxide layer thickness and weight gains were similarly affected by the thickness of preoxidized layer.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.576-581
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• 2010

KAERI (Korea Atomic Energy Research Institute) fabricated MOX (Mixed Oxide) fuel pellets as a cooperation project with PSI (Paul Scherrer Institut) for an irradiation test in the Halden reactor. The MOX pellets were fitted into fuel rods that included instrumentation for measurement in IFE (Institutt for Energiteknikk). The fuel rods were assembled into the test rig and irradiated in the Halden reactor up to 50 MWd/kgHM. The irradiated fuel rods were transported to the IFE, where ceramography was carried out. The fuel rods were cut transversely at the relatively higher burn-up locations and then the radial cross sections were observed. Micrographs were analyzed using an image analysis program and grain sizes along the radial direction were measured by the linear intercept method. Radial cracks in the irradiated MOX were observed that were generally circumferentially closed at the pellet periphery and open in the hot central region. A circumferential crack was formed along the boundary between the dark central and the outer regions. The inner surface of the cladding was covered with an oxide layer. Pu-rich spots were observed in the outer region of the fuel pellets. The spots were surrounded by many small pores and contained some big pores inside. Metallic fission product precipitates were observed mainly in the central region and in the inside of the Pu spots. The average areal fractions of the metallic precipitates at the radial cross section were 0.41% for rod 6 and 0.32% for rod 3. In the periphery, pore density smaller than 2 ${\mu}m$ was higher than that of the other regions. The grain growth occurred from 10 ${\mu}m$ to 12 ${\mu}m$ in the central region of rod 6 during irradiation.

• Nuclear Engineering and Technology
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• v.21 no.2
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• pp.111-122
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• 1989

With the Zircaloy-4 tube oxidized in high temperature (1323 K) steam for 5, 10, 30 and 60 minutes, the expanding copper mandrel test was carried out over a temperature range of 673-l173k at $\varepsilon\;=\;3.0\times10^5S\;^1$. The oxidation parameters $(K_i)$ in the present study were linearly proportional to square root of time $(Ki= \delta_{kit})$ and their rate constants ($\delta_{ki}$) are 0.281, 2.82, and 2.313 for weight gain and thickness of Zr02 and $\alpha$(0) layer, respectively. Activation energy for high temperature (873-1073k) plastic deformation of Zircaloy-4 increases from 251 KJ/mol to 323 KJ/mol with increase in oxidation time from 5 minutes to 60 minutes due to the high strengthened Zr02. With the oxide layer thickness [K ; expressed in "Equivalent Cladding Reacted" (ECR,%)] and the yield stress obtained from the mandrel test, an empirical relation was derived as ($\sigma/C)^n=K^mexp$ (Q/RT) with n=6.9, m=5.7, C=0.155, 0.138, 0.051, and 0.046 MPa for Q=251, 258, 316, 323 KJ/mol, respectively.

• Journal of the Korean Ceramic Society
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• v.48 no.1
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• pp.69-73
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• 2011

A $ZrO_2$ thin film as a corrosion protective layer was deposited on Zircaloy-4 (Z-4) clad material using $N_2O$ as a reactive gas by RF reactive magnetron sputtering at room temperature. The Z-4 substrate was located in plasma or out of plasma during the $ZrO_2$ deposition process to investigate mechanical and corrosive properties for the plasma immersion. Tetragonal and monoclinic phases were existed in $ZrO_2$ thin film immersed in plasma. We observed that a grain size of the $ZrO_2$ thin film immersed in plasma state is larger than that of the $ZrO_2$ thin film out of plasma state. In addition, the corrosive property of the $ZrO_2$ thin films in the plasma was characterized using the weight gains of Z-4 after the corrosion test. Compared with the $ZrO_2$ thin film immersed out of plasma, the weight gains of $ZrO_2$ thin film immersed in plasma were larger. These results indicate that the $ZrO_2$ film with the tetragonal phase in the $ZrO_2$ can protect the Z-4 from corrosive phenomena.