• Korean Journal of Materials Research
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• v.12 no.4
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• pp.259-263
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• 2002

Three different ribbons of amorphous $Til_{1-x}Be_x$ alloys such as $Ti_{0.59}Be_{0.41},\;Ti_{0.61}Be_{0.39}\;and\;Ti_{0.63}Be_{0.37}$ were made by melt-spinning method to be used as brazing filler metals for joining Zircaloy-4 nuclear fuel cladding tubes, and their crystallization behavior as well as microstructure of the brazed zone were examined. The crystallization behavior was investigated in teams of thermal stability, crystallization temperature and activation energy. The crystallization of the $Ti_{1-x}Be_x$ alloys proceeded in two steps by the formation of ${\alpha}$-Ti at a lower temperature and of TiBe at a higher temperature. The crystallization temperature and activation energy of $Ti_{1-x}Be_x$ alloys were higher and larger than those of $Zr_{1-x}Be_x$ alloys and PVD Be. Those resulted thinner joining layer with $Ti_{1-x}Be_x$ alloys, which kept sound thickness of Zircaloy-4 nuclear fuel cladding tubes after brazing. But in the brazed zones made by $Ti_{1-x}Be_x$ filler metals, a little solid-solution layers composed of Zr and Ti were formed toward the Zr cladding tube and Zr was detected in the brazed zones. Microstructure of brazed zone was changed from globular to dentrite with decreasing Be content in the $Ti_{1-x}Be_x$ filler metal.

• Korean Journal of Materials Research
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• v.8 no.11
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• pp.1031-1037
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• 1998

To investigate the effect of cooling rate and annealing temperature on the corrosion of Zircaloy-4 and Zr-2. 5Nb alloys, autoclave corrosion tests were performed at $500^{\circ}C$ for the specimens prepared by various heat treatments. The specimens were heat-treated at $1050^{\circ}C$ for 30 minutes and cooled by ice-brine quenching, water quenching, oil quenching, air cooling, and furnace cooling. To investigate the effect of annealing temperature, the specimens were annealed at $\alpha$, ($\alpha$+$\beta$)-, and $\beta$-temperatures. It was observed from the $500^{\circ}C$ corrosion test that nodular corrosion occurred on the Zircaloy-4 alloy but did not occur on the Zr-2.5Nb alloy. The corrosion resistance of Zircaloy-4 increased with increasing the cooling rate. On the other hand, the corrosion resistance of Zr-2.5Nb decreased with increasing the cooling rate and the annealing temperature. It is suggested that corrosion resistance of Zircaloy-4 would be controlled by the distribution of Fe and Cr element in the matrix and precipitates, while that of Zr-2.5Nb alloy the niobium concentration and $\beta_{-Nb}$ phase.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.246-252
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• 2018

A series of simulated reactivity-initiated accident (RIA) tests on irradiated fully recrystallized boiling water reactor Zircaloy-2 cladding has been performed by means of the expansion-due-to-compression (EDC) test method. The EDC method reproduces fuel pellet-clad mechanical interaction (PCMI) conditions for the cladding during RIA transients with respect to temperature and loading rates by out-of-pile mechanical testing. The tested materials had a large variation in burnup and hydrogen content (up to 907 wppm). The results of the EDC tests showed variation in the PCMI resistance of claddings with similar burnup and hydrogen content, making it difficult to clearly identify ductile-to-brittle transition temperatures. The EDC-tested samples of the present and previous work were investigated by light optical and scanning electron microscopy to study the influence of factors such as azimuthal variation of the Zr-hydrides and the presence of hydride rims and radially oriented hydrides. Two main characteristics were identified in samples with low ductility with respect to hydrogen content and test temperature: hydride rims and radial hydrides at the cladding outer surface. Crack propagation and failure modes were also studied, showing two general modes of crack propagation depending on distribution and amount of radially oriented hydrides. It was concluded that the PCMI resistance of irradiated cladding under normal conditions with homogenously distributed circumferential hydrides is high, with good margin to the RIA failure limits. To further improve safety, focus should be on conditions causing nonfavorable hydride distribution, such as hydride reorientation and formation of hydride blisters at the cladding outer surface.

• 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 ㎛.

• Korean Journal of Materials Research
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• v.9 no.6
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• pp.589-594
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• 1999

The effect of $\beta$-heat treatment on th microstructure, mechanical properties and texture in the nuclear fuel cladding of Zircaloy-4 tubes was chosen at 1000, 1100 and 120$0^{\circ}C$, and the tubes were heat-treated by a high frequency vacuum induction furnace. Morphology of the second phase particles and $\alpha$-grain of as-received tubes were markedly changed by heat treatment. The average sizes of second phase particles of as-received and $\beta$-heat treated tubes were 0.1$\mu\textrm{m}$ and 0.076$\mu\textrm{m}$, respectively. However, the average sizes of second phase particles were not much changed in the $\beta$-heated temperatures. With increasing heat treatment temperatures, the 0.2% yield strength and the hoop strength were decreased because of changes in preferred orientation as will as $\alpha$-plate width. Heat treated Zircaloy-4 tubes exhibited texture changes but the preferred orientation of grains still remained.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.75-81
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• 1990

Experiments to predict the thermal contact conductance between the fuel pellet and cladding have been performed, which is important to determine the temperature distibution within the fuel rod. UO$_2$and Zircaloy-2 are used in these experiments. The measuring apparatus is composed of a presser which controls the contact pressure, a thermometer with 5.5 sheathed thermocouples, a vacuum pump, pellet and cladding rods, and two heating devices, etc. The thermal contact conductances were measured with varying the contact pressure and surface roughnesses of UO$_2$and Zircaloy-2 bars. The results show that an increase in the contact pressure and a decrease of surface roughness resulted in increase of the thermal contact conductance. Finally, a fitting correlation has been established and compared with widely-used correlations.

• Tribology and Lubricants
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• v.17 no.6
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• pp.453-458
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• 2001

The fretting wear of a nuclear fuel rod it a dangerous phenomenon. In this study, TiN coating was used to reduce the fretting wear of Zircaloy-4 tube, a nuclear fuel rod cladding material. TiN coating is probably one of the molt frequently and successfully used PVD coatings for the mitigation of fretting wear. The fretting tester was designed and manufactured for this experiment. The number of cycles, slip amplitude and normal load were selected as main factors of fretting wear. The results of this research showed that wear volume was improved 1.3∼3.2 times with TiN coating. The worn surfaces were observed by SEM. Wear mechanism at lower slip amplitude was the brittle cracks and rupture of TiN coating. However, adhesive and abrasive wear were mainly observed on most surfaces at higher slip amplitude.

• Journal of Welding and Joining
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• v.20 no.3
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• pp.98-104
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• 2002

A study on microstructures of welds for Zircaloy-4 sheath end closure by the resistance upset welding methods was carried out. Two upset welding process variations such as magnetic farce and multi-impulse resistance welding were used. Grain size and microhardness across welds were analysed in terms of welding parameters. Magnetic farce resistance weld with one cycle of unbalanced mode has smaller upset length and $\alpha-grain$ size in heat affected zone than those of multi-impulse resistance weld because of lower heat input and shorter welding time. Heat affected zone formed by two upset resistance welding variations revealed fine Widmanstatten structure or martensitic ${\alpha}'$ structure due to the high heating rate and foster cooling rate. Magnetic force resistance welds showed recrystallized grains before grain growth, whereas multi-impulse resistance welds showed full grain growth.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.423-424
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• 2005

The performance of HANA claddings was evaluated in out-of-pile conditions. All the performance test results revealed that HANA claddings were superior to the reference claddings such as Zircaloy-4 and A-cladding. Corrosion resistance was improved by 60 to 70% compared to the commercial claddings. Creep, burst, tensile, LOCA, wear and microstructural properties were shown to be as good as the commercial claddings.

• Nuclear Engineering and Technology
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• v.35 no.5
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• pp.471-481
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• 2003

In order to evaluate the microstructural characteristics of irradiated fuel claddings (Zircaloy-4), two irradiated specimens having different burnups (18 GWD/MTU and 42 GWD/MTU) were prepared from the G23-M4 fuel rods, which were loaded in Kori Unit 1 for 4 fuel cycles. The oxide thickness, hydride morphology and hardness change were characterized by an optical microscope and a micro-hardness tester after preparing the irradiated specimens in the PIE (Post Irradiation Examination) facilities. The dislocation loops and the amorphous transformation of precipitates induced by the neutron irradiation in the nuclear plant were also examined by a TEM. As the burnup increased from 18 GWD/MTU to 42 GWD/MTU, the oxide thickness increased from $6.0{\mu}m\;to\;25.3{\mu}m$ and the contents of hydrogen pick-up in the Zr matrix also greatly increased. In the comparison of the hardness of the unirradiated fuel cladding, the amount of hardness increase was nearly $9.3\%\;and\;24.2\%$ for the irradiated samples of 18 and 42 GWD/MTU, respectively. Both -type dislocation loops and -type dislocation components were observed simultaneously in the irradiated specimens and the densities of the dislocation was increased by increasing the burnup. The precipitates in both the irradiated specimens were amorphously transformed by the neutron irradiation and the trend of the amorphous transformation of the precipitates was enhanced at a higher burnup.