• Nuclear Engineering and Technology
• /
• v.10 no.2
• /
• pp.65-72
• /
• 1978

In this paper, it is attempted to investigate the phenomena of iodine stress corrosion cracking of Zircaloy-4 cladding failures in reactor through the results of similar out-of-pile test in iodine vapour. The main result of this experiment is a finding of the relation between the threshold stress which can lead to iodine stress corrosion cracking of Zircaloy-4 tube and the iodine concentration. The values of critical stress and the critical iodine concentration are also obtained. A model which relates failure time of Zircaley-4 tube to failure stress and iodine concentration is suggested as follows: log t$_{F}$ =5.5-(3/2)log$_{c}$-4log $\sigma$ where t$_{F}$ : failure time, minutes c: iodne concentration, mg/㎤ $\sigma$: stress, 10$^4$psi.

• Bulletin of the Korean Chemical Society
• /
• v.30 no.6
• /
• pp.1349-1352
• /
• 2009

We report the results of the surface chemistry of deuterized ethanol exposed Zircaloy-4 (Zry-4) surfaces with various amount of $C_2D_5$OD exposures at 190 K. This system was examined with Auger electron spectroscopy (AES) and temperature programmed desorption (TPD) techniques. In TPD study, $D_2$ was evolved at two different desorption temperature regions accompanying with broad desorption background. The lower temperature feature at around 520 K showed first-order desorption kinetics. The high temperature desorption peak at around 650 K shifted to lower desorption temperature as the exposure of $C_2D_5$OD increased. The Zr(MNV) Auger peak shifted about 2.5 eV from 147 eV to lower electron energy followed by 300 L of $C_2D_5$OD dosing. This implies metallic zirconium was oxidized by deuterized ethanol adsorption. After stepwise annealing of the oxidized Zry-4 sample up to 843 K, the shifted Zr(MNV) peak was gradually shifted back to metallic zirconium peak position. After the sample was heated to 843 K, the oxygen content near the Zry-4 surface was recovered to clean surface level. The concentration of carbon, however, was not recovered by annealing the sample.

• Korean Journal of Materials Research
• /
• v.9 no.6
• /
• pp.589-594
• /
• 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
• /
• v.22 no.1
• /
• pp.75-81
• /
• 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.

• Nuclear Engineering and Technology
• /
• v.53 no.7
• /
• pp.2312-2322
• /
• 2021

In recent studies, the creep rate of Zircaloy-4, one of the basic property parameters of the nuclear fuel code, has been commonly used with the axial creep model proposed by Rosinger et al. However, in order to calculate the circumferential deformation of the fuel cladding, there is a limitation that a difference occurs depending on the anisotropic coefficients used in deriving the circumferential creep equation by using the axial creep equation. Therefore, in this study, the existing axial creep law and the derived circumferential creep results were analyzed through a circumferential creep test by the internal pressurization method in the isothermal conditions. The circumferential creep deformation was measured through the optical image analysis method, and the results of the experiment were investigated through constructed IDECA (In-situ DEformation Calculation Algorithm based on creep) code. First, preliminary tests were performed in the isotropic β-phase. Subsequently in the anisotropic α-phase, the correlations obtained from a series of circumferential creep tests were compared with the axial creep equation, and optimized anisotropic coefficients were proposed based on the performed circumferential creep results. Finally, the IDECA prediction results using optimized anisotropic coefficients based on creep tests were validated through tube burst tests in transient conditions.

• Korean Journal of Materials Research
• /
• v.8 no.3
• /
• pp.268-273
• /
• 1998

Standard Zircaloy-4 sheets, charged with 230-250ppm hydrogen by the gas-charging method and homogenized at $400^{\circ}C$ for 72hrs in a vacuum, were corroded in pure water and aqueous LiOH solutions using static autoclaves at $350^{\circ}C$. Their corrosion behaviors were characterized by measuring their weight gains with the corrosion time and observing their microstructures using an optical microscope and a scanning electron microscope. The elemental depth profiles for hydrogen and lithium were measured using a secondary ion mass spectrometry(S1MS) to confirm their distributions at the oxidelmetal interface. The normal Zircaloy-4 specimens corroded abruptly and heavily at the concentration of Li ions more than 30ppm in the aqueous solution. This is due to accelerations by the rapid oxidation of many Zr- hydrides formed by the large amount of absorbed hydrogen, resulting from the increased substitution of $Li^{+}$ ions with $Zr^{4+}$-sites in the oxide as the Li ion concentration increased. The specimens that had been charged with amounts of hydrogen greater than its solubility corroded early with a more rapid acceleration than normal specimens, regardless of the corrosion solutions. At longer corrosion times. however, normal specimens showed a rather accelerated corrosion rate compared to the hydrogen-charged specimens. These slower corrosion rates of the hydrogen-charged specimens at the longer corrosion times would be due to the pre-existent Zr-hydride in the matrix, which causes the hydrogen pick- up into the specimen to be depressed, when the oxide with an appropriate thickness formed.

• Journal of the Korea Academia-Industrial cooperation Society
• /
• v.7 no.2
• /
• pp.113-119
• /
• 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
• /
• v.17 no.3
• /
• pp.211-215
• /
• 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.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2001.05a
• /
• pp.312.2-312
• /
• 2001

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.10a
• /
• pp.243-243
• /
• 1998