• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.123-124
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.125-126
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• 2018

• 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.

• Nuclear Engineering and Technology
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• v.43 no.2
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• pp.149-158
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• 2011

Metal waste generated from the pyroprocessing of 10 MtU of spent fuel was classified by comparing the specific activity of a relevant radionuclide with the limit value of the specific activity specified in the Korean acceptance criteria for a lowand intermediate-level waste repository. A Korean Optimized Fuel Assembly design with a 17${\times}$17 array, an initial enrichment of 4.5 weight-percent, discharge burn-up of 55 GWD/MtU, and a 10-year cooling time was considered. Initially, the mass and volume of each structural component of the assembly were calculated in detail, and a source term analysis was subsequently performed using ORIGEN-S for these components. An activation cross-section library generated by the KENO-VI/ORIGEN-S module was utilized for top-end and bottom-end pieces. As a result, an Inconel grid plate, a SUS plenum spring, a SUS guide tube subpart, SUS top-end and bottom-end pieces, and an Inconel top-end leaf spring were determined to be unacceptable for the Gyeongju low- and intermediate-level waste repository, as these waste products exceeded the acceptance criteria. In contrast, a Zircaloy grid plate and guide tube can be placed in the Gyeongju repository. Non-contaminated Zircaloy cladding occupying 76% of the metal waste was found to have a lower level of specific activity than the limit value. However, Zircaloy cladding contaminated by fission products and actinides during the decladding process of pyroprocessing was revealed to have 52 and 2 times higher specific activity levels than the limit values for alpha and $^{90}Sr$, respectively. Finally, it was found that 88.7% of the metal waste from the 17${\times}$17 Korean Optimized Fuel Assembly design should be disposed of in a deep geological repository. Therefore, it can be summarized that separation technology with a higher decontamination factor for transuranics and strontium should be developed for the efficient management of metal waste resulting from pyroprocessing.

• Journal of Mechanical Science and Technology
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• v.15 no.9
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• pp.1274-1280
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• 2001

The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In the study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.

• Nuclear Engineering and Technology
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• v.55 no.7
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• pp.2454-2465
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• 2023

Film boiling may lead to burnout of the heating element. Even though burnout does not occur, the heating element is subject to deformation because it is not sufficiently strong to withstand external loads. In particular, the ballooning and rupture of a tube under film boiling are important phenomena in the field of nuclear reactor safety. If the tube-type cladding of nuclear fuel ruptures owing to high internal pressure and thermal load, radioactive materials inside the cladding are released to the coolant. Therefore, predicting the ballooning and rupture is important. This study presents numerical simulations to predict the ballooning behavior and rupture time of a horizontal tube at high internal pressure under saturated film boiling. To do so, a multi-step coupled simulation of conjugated film boiling heat transfer and ballooning using creep model is adopted. The numerical methods and models are validated against experimental values. Two different nonuniform heat flux distributions and four different internal pressures are considered. The three-step simulation is enough to obtain a convergent result. However, the single-step simulation also successfully predicts the rupture time. This is because the film boiling heat transfer characteristics are slightly affected by the tube geometry related to creep ballooning.

• Proceedings of the Korean Society of Tribologists and Lubrication Engineers Conference
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• 1996.04b
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• pp.25-29
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• 1996

The fretting wear characteristics for Zircaloy-4 tube used as fuel rod in the nuclear power plant have been investigated. The fretting wear tester was designed and manufactured for this experiment. This study was focused on main factors of fretting wear, cycle, slip amplitude and normal load. The worn surfaces were observed by SEM.

• 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.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.7 no.2
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• pp.113-119
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• 2006

Brazing is an important manufacturing process in the fabrication of Heavy Water Reactor fuel rods, in which bearing and spacer pads are joined to Zircaloy-4 cladding tubes. The physical vapor deposition(PVD) technique is currently used to deposit metallic Be on the surfaces of pads as a filler metal. Amorphous Zr-Be binary alloys which are manufactured by rapid solidification process are under developing to substitute the conventional PVD-Be coating. In the present study, brazed joint with PVD and amorphous alloys of $Zr_{1-x}Be_{x}(0.3{\le}x{\le}0.5)$ as filler metals are compared by mechanism, microstructure and hardness. The thickness of brazed joint with amorphous alloys became much smaller than that of PVD-Be. The erosion of base metal did not occur in the brazed joint with amorphous alloys. The brazing mechanism for PVD-Be seems to be Be diffusion into Zr-4 with capillary action resulting from eutectic reaction while that for amorphous alloys are associated with the liquid phase formation in the brazed joint. The brazed joint microstructure with PVD-Be consists of dendrite while that with amorphous alloys is globular. The $Zr_{0.7}Be_{0.3}$ alloy shows the smooth interface with little erosion in the base metal and is recommended a most suitable brazing filler metal for Zircaloy-4.

• Nuclear Engineering and Technology
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• v.17 no.3
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• pp.211-215
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• 1985

The effects of temperature and strain rate on the I-SCC behaviors of Zircaloy-4 were investigated by constant load test at 30$0^{\circ}C$ and constant elongation rate test at 300, 350 and 40$0^{\circ}C$ in 3.34mg $I_2$/㎤. The results showed that I-SCC susceptibility increased as the strain rate decreased or the temperature increased. The empirical relation between the stress and the time to failure at 30$0^{\circ}C$ was given by 1/ $t_{f}$∝exp (0.3$\sigma$/$\sigma$$_{UTS}$-31.5) When the I-SCC susceptibility was expressed by the ratio of fracture energy in iodine atmosphere to that in the inert atmosphere, severe I-SCC susceptibility was found near 7.6$\times$10$^{-6}$ sec at 30$0^{\circ}C$ and the maximum point of I-SCC susceptibility tended to shift to the higher strain rate with increasing the temperature. The quasi-cleavage fracture was observed in I-SCC fracture surface. From these results, it was certain that the film repture step was involved as an important process in the I-SCC mechanism of Zircaloy-4.4.