• Korean Journal of Materials Research
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• v.4 no.4
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• pp.445-452
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• 1994

Three-axial deformation behavior of the Zircaloy cladding tube under the irradiation condition of the fuel in pressurized water reactor can be analyzed by the anisotropy in the creep and the irra- diation growth, which depends on the texture parameter. A methodology to evaluate the impact of the anisotropic creep and irradiation growth on the strain in each axial direction of the cladding tube has been proposed. Based on the measured strains after irradiation and predicted ones with the help of a fuel performance analysis code, it is found that a tangential strain of the cladding tube is caused mainly by the creep, whereas a axial strain of the cladding is caused mainly by the irradiation growth but with a considerable contribution of the creep at low irradiation.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.10b
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• pp.185-186
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• 2002

Fretting of fuel rod cladding material, Zircaloy-4 tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to be an ideal solution to fretting damage since fretting is closely related to wear. corrosion and fatigue. Therefore. in this study the fretting wear experiment was performed using TiAlN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaloy-4 as on of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAlN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and fretting wear mechanisms were brittle fracture and plastic flow at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher ship amplitude.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2047-2053
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• 2020

Single-tube burst tests on hydrogenated Zircaloy-4 nuclear fuel cladding under simulated loss-of-coolant accident are conducted to evaluate the impact of hydrogen on burst parameters. The heating rate and initial pressure are varied from 5 K/s to 150 K/s and 5 bar-80 bar, respectively. The hydrogen concentration in the cladding is in the range of 0-2000 wppm. Burst stress is lower for hydrogenated cladding in α-phase. A significant loss of ductility is observed in α-phase and lower α + β-phase for hydrogenated cladding. However, the burst strain is higher for hydrogenated cladding in β-phase. There is a sigmoidal dependency of rupture area with initial stress and rupture area is larger for hydrogenated cladding. A novel burst stress correlation for hydrogenated Zircaloy-4 cladding has been proposed.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2002.05a
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• pp.88-95
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• 2002

Fretting of fuel rod cladding material, Zircaloy-4 Tube, in PWR nuclear power plants must be reduced and avoided. Nowadays the introduction of surface treatments or coatings is expected to bean ideal solution to fretting damage since fretting is closely related to wear, corrosion and fatigue. Therefore, in this study the fretting wear experiment was peformed using TiAIN coated Zircaloy-4 tube as the fuel rod cladding and uncoated Zircaioy-4 tube as one of grids, especially concentrating on the sliding component. Fretting wear resistance of TiAIN coated Zircaloy-4 tubes was improved compared with that of TiN coated tubes and uncoated tubes and the fretting wear mechanisms were delamination and plastic flow following by brittle fracture at lower slip amplitude but severe oxidation and spallation of oxidative layer at higher slip amplitude.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2054-2063
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• 2020

The high-temperature oxidation behavior of ZrSi₂ used as a coating material for nuclear fuel cladding was investigated for developing accident-tolerant fuel cladding of light water reactors. Bulk ZrSi₂ samples were prepared by spark plasma sintering. In situ X-ray diffraction was conducted in air at 900, 1000, and 1100 ℃ for 20 h. The microstructures of the samples before and after oxidation were examined by scanning electron microscopy and transmission electron microscopy. The results showed that the oxide layer of zirconium silicide exhibited a layer-by-layer structure of crystalline ZrO₂ and amorphous SiO₂, and the high-temperature oxidation resistance was superior to that of Zircaloy-4 owing to the SiO₂ layer formed. ZrSi₂ was coated on the Zircaloy-4 tube surface using laser 3D printing, and the coated tube was oxidized for 2000 s at 1200 ℃ under a vapor/argon mixture atmosphere. The outer surface of the coated tube was hardly oxidized (10-30 ㎛), while the inner surface of the uncoated tube was significantly oxidized to approximately 300 ㎛.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.218-227
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• 1986

Studies were conducted to determine the extent of oxidation and same of the mechanical property changes of Zircaloy-4 fuel cladding after it was exposed to hot steam environment. The purpose of these tests was to provide some informations on the embrittlement behavior of CANDU type fuel cladding, which could be experienced under the loss-of-coolant accident conditions. The Zircaloy fuel cladding tubes were exposed in a steam environment at the temperature of 90$0^{\circ}C$, 1,00$0^{\circ}C$. The growth of the ZrO$_2$ layer combined with an oxygen rich $\alpha$-phase layer into the Zircaloy tube material was found as a function of time t and temperature of steam exposure, E=1.1√Dt+0.002 where D is a temperature dependent diffusion coefficient. The tensile strength of the specimens exposed for a short period increased but decreased continuously with further exposure. The circumferential elongation was drastically changed with the exposure time while the hoop strength did't decrease greatly. The X-ray measurement of preferred orientation of the Zircaloy tube material indicated that grains in the as received tube were oriented such that the poles of the basal (0001) planes were predominantly radial, while the poles of the basal plane in the tube materials heattreated at 1,00$0^{\circ}C$ were oriented tangentially. It appears that this reoriented texture may contribute to lessening the decrease of the hoop strength of the heat treated Zircaloy tube material.

• Journal of the Korean institute of surface engineering
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• v.35 no.5
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• pp.312-318
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• 2002

A study on the change of mechanical properties and oxidation behavior of Zircaloy-4 fuel cladding after exposing at 90$0^{\circ}C$ and $1000^{\circ}C$ for various periods of exposure time under the steam atmosphere was carried out. The growth of the $ZrO_2$ layer combined with an oxygen-rich-phase layer into the Zircaloy tube material can be described by an expression, E = 1.1√Dt + $2 $\times$ 10^{-4}$ . The tensile strength of Zircaloy tubes increased for a short period of exposure time and decreased rapidly with further exposure while the hoop strength was not decreased greatly. In the meantime, the axial and circumferential elongations of oxidized Zircaloy tubes were decreased drastically with increasing exposure time as a result of embrittlement phenomena.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 2001.11a
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• pp.19-26
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• 2001

The wear characteristic of Zircaloy-4 tube, which is used for a cladding of light water reactor fuel rod, is investigated experimentally. The experiment is conducted with contacting the crossed tube specimens in air as well as in water at room temperature with various combination of contact normal force and sliding distance of reciprocating motion. The contour and the volume of each wear are examined to study the effect of contact condition and environment on wear. As a result, it is found that the wear volume in the water environment is larger than that in the air for all the contact (i.e., force and sliding distance) conditions. However, the wear depth is greater in air than in water if the contact normal force and the sliding distance are larger. These are explained by the ease of detachment of wear particles from the contact surface. On the other hand, workrate model is applied with the contact shear force range measured by our wear tester. Investigated is the correlation between the workrate and the wear volume increase rate of the present experiment. The parabolic curve is found to fit well for the present wear data.

• 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.

• Analytical Science and Technology
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• v.12 no.6
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• pp.490-497
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• 1999

A hydrogen determinator was modified and installed in the glove box to analyse total hydrogen content in irradiated zircaloy tube. The analysis method of hydrogen is Inert Gas Fusion(IGF)-Thermal Conductivity Detection(TCD). The hydrogen recoveries of no tin method using Ti and Zr matrix standards, respectively, were available within $3{\mu}g$ of hydrogen. Also the smaller size of sample showed the better hydrogen recovery. It was found that the hydrogen standard of Ti matrix is avaliable to hydrogen analysis in zircaloy sample. The mean radioactivity of irradiated zircaloy sample was 10 mR/hr and hydrogen concentration was 130 ppm.