• Proceedings of the KSME Conference
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• 2001.06a
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• pp.489-494
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• 2001

The fuel of light water reactor used far several years at high temperature and pressure, so it needs to clad with high corrosion resistance material. The cladding materials need low absorption of a neutron and high corrosion resistance. Cladding materials used Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor and Zirlo has good for long term corrosion. If fracture of cladding tube occured during operation, it caused disaster. So it is needed to estimate of integrity fur cladding materials. In this paper, tension characteristics of cladding materials are investigate which is basic research far fracture characteristic. Also analysis of residual stress effect between tube type(original type) specimen and flattened type specimen.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.931-932
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• 2004

• 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.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.1377-1380
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• 2008

An irradiation test of a nuclear fuel rod having external and internal tube structure was planned for a performance. To establish fabrication process satisfying the requirements of irradiation test, micro-TIG welding system for fuel rods was developed, and preliminary welding experiments for optimizing process conditions of fuel rod was performed. Fuel rods with 15.9mm diameter and 0.57mm wall thickness of cladding tubes and end caps have been used and optimum conditions of endcap welding have been selected. In this experiment, the qualification test was performed by tensile tests, helium leak inspections, and metallography examinations to qualify the endcap welding procedure. The soundness of the welds quality of a dual cooled fuel rods has been confirmed by mechanical tests and microstructural examinations.

• 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.

• 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.

• Proceedings of the Acoustical Society of Korea Conference
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• 1998.06e
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• pp.247-250
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• 1998

The scattering of plane ultrasonic waves by the nuclear fuel pin of liquid metal reactor in sodium is studied. According to the internal composition in the cladding tube, the fuel pin has three cross sections, i.e. helium gas plenum, sodium-filled section, and fuel insertion section. The scattering spectra for each section of the fuel pin are different. The circumnavigating ultrasonic waves of each section are analyzed by the resonance scattering method. The whispering gallery wave modes are generated in the sodium-filled plenum section and the fuel rod insertion section with a sodium-gap. The circumferential wave modes are propagated in the cladding tube of the helium gas plenum section. The annular gap between the cladding tube and metal uranium pellet rod affects the scattering spectra. The different propagation characteristics can be utilized for the nondestructive method of detecting the unbonded area and measuring the level of the sodium-filled section of the fuel pin.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.256-261
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• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test($400^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test($400^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone.

• Proceedings of the KWS Conference
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• 2001.10a
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• pp.65-68
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• 2001

Corrosion and tensile properties of zircaloy-4 cladding tube having a laser welding part in elevated temperature are studied to present the criterion of quality evaluation in nuclear reactor and to found the scientific basis of SCC, with laser welding method using by coupling up cladding tube to end cap. In the result of tensile test(40$0^{\circ}C$), the fracture is not happened in the welding part but base metal and the result of corrosion test(40$0^{\circ}C$ 1500psi steam), corrosion rate of the molten zone and PMZ is a little higher than the other zone

• 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.