• Nuclear Engineering and Technology
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• v.9 no.3
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• pp.117-123
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• 1977

The strain ageing behaviour of quenched Zircaloy--4 has been studied as a function of ageing time and temperature in the temperature range 523 to 588 K for a short-ageing time of 1 to 52 seconds. At the test conditions, the strain ageing stress increased with ageing time and temperature at a strain rate of 5.55$\times$10$^{-4}$ sec$^{-1}$ . Applying stress on the Quenched Zircaloy-4, the strain ageing effect indicated following two stages: an initial stage having an activation energy of 0.39 ev considered to be due to Snoek type ordering of intersitial oxygen atoms in the stress field of a dislocation and a second stage having an activation energy of 0.60 ev, due to mainly long-range diffusion of oxygen atoms.

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

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.5
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• pp.823-832
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• 2003

To define the causes of cladding degradation which can take place during the operation of nuclear power plants, it is required to develop the new fracture toughness test of spent fuel cladding. The fracture toughness of Zircaloy-4 cladding was estimated using the recently developed KAERI embedded Charpy (KEC) specimen. Axially notched KEC specimens cut directly from unirradiated fuel claddings, were tested in a way similar to the standard toughness test method of a Single Edge Bending (SEB) specimen. The results of KEC fracture toughness test at room temperatures were discussed and compared with those of the previous other studies. In conclusions, even though the KEC fracture toughness test of nuclear fuel claddings was easier and more reliable than those developed earlier, the results from the cladding fracture tests were not the material characteristics but the specific fracture parameters which were deeply related to the specification of claddings. In addition, the phenomenon of a thickness yielding was not observed from the fracture surface. It was closely related to the fact that the plane strain condition of the KEC specimen was changed to the plane stress condition during crack advancing. It was also supported by the fractographic evidence that the formation of ductile dimples at the crack initiation became the similar appearance such as a quasi-cleavage after the sufficient crack advancing.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2565-2571
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• 2020

Background: Understanding the behaviour of nuclear fuel claddings by conducting burst test on single cladding tube under simulated loss-of-coolant accident conditions and developing theoretical cum empirical predictive computer codes have been the focus of several investigations. The developed burst criterion (a) assumes symmetrical deformation of cladding tube in contrast to experimental observation (b) interpolates the properties of Zircaloy-4 cladding in mixed α+β phase (c) does not account for azimuthal temperature variations. In order to overcome all these drawbacks of burst criterion, it is reasoned that artificial intelligence technique may be a better option to predict the burst parameters. Methods: Artificial neural network models based on feedforward backpropagation algorithm with logsig transfer function are developed. Results: Neural network architecture of 2-4-4-3, that is model with two hidden layers having four nodes in each layer is found to be the most suitable. The mean, maximum, and minimum prediction errors for this optimised model are 0.82%, 19.62%, and 0.004%, respectively. Conclusion: The burst stress, burst temperature, and burst strain obtained from burst criterion have average deviation of 19%, 12%, and 53% respectively whereas the developed neural network model predicted these parameters with average deviation of 6%, 2%, and 8%, respectively.

• Nuclear Engineering and Technology
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• v.20 no.3
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• pp.170-178
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• 1988

An automated ultrasonic flaw detection system was developed for thin-walled and short tubes such as Zircaloy-4 tubes used for cladding heavy-water reactor fuel. The system was based on the two channels immersion pulse-echo technique using 14 MHz shear wave and the specially developed helical scanning technique, in which the tube to be tested is only rotated and the small water tank with spherical focus ultrasonic transducers is translated along the tube length. The optimum angle of incidence of ultrasonic beam was 26 degrees, at which the inside and outside surface defects with the same size and direction could be detected with the same sensitivity. The maximum permissible defects in the Zircaloy-4 tubes, i.e., the longitudinal and circumferential v notches with the length of 0.76mm and 0.38mm, respectively and the depth of 0.04 mm on the inside and outside surface, could be easily detected by the system with the inspection speed of about 1 m/min and the very excellent reproducibility. The ratio of signal to noise was greater than 20 dB for the longitudinal defects and 12 dB for the circumferential defects.

• Journal of Welding and Joining
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• v.24 no.5
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• pp.22-28
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• 2006

A spacer grid assembly, which is an interconnected array of slotted grid straps and is welded at the intersections to form an egg crate structure, is one of the main structural components of the nuclear fuel assembly for Pressurized Water Reactors (PWRs). The weld quality of spacer grids in PWRs fuel is extremely important for the fuel assembly performance in the nuclear renter. The spacer grid welds are currently evaluated mainly by the metallographic examination although it reveals only cross-points which are welded by the laser beam. This experiment is also to compare the weldability of Zircaloy-4 spacer grids using by the GTA and LB. The effect of node geometries of spacer grids for the GTAW and LBW has been studied and optimum conditions of spacer grid welding have been found. Microstructures and micro-hardness of the GTA and LB welded zones have been also compared.

• Journal of the Korean institute of surface engineering
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• v.24 no.1
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• pp.31-41
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• 1991

The nodular corrosion behavior of Zircaloy-4 alloy was investigated by autoclave test at 50$0^{\circ}C$ under 1500 psi for the specimens quenched into water from $700^{\circ}C$, 80$0^{\circ}C$, 90$0^{\circ}C$, and 105$0^{\circ}C$. It was observed that the corrosion resistance of Zircalloy-4 specimen increased with increase in annealing temperature, and annealing at $\alpha$-region temperatures resulted in nodular corrosion while annealing at the temperature range of $\alpha$+$\beta$ and $\beta$ did not show nodular corrosion. It was also found that the size of nodule formed on the surface of the specimens increased with increase in exposure time in autoclave, but the total number of nodule remained uncha-nged. The corrosion of furnace-cooled specimens progressed mostly in the interior of grains where Fe and Cr alloying elements were largely depleted during the cooling process. However, the grain boundary seemed to act as a barrier to the nodular corrosion. From combining the present results with other works, it is suggested that the nodules nuc-leate in the local region where some of alloying elements are depleted.

• 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.

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.2
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• pp.231-238
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• 2003

The fuel of light water reactor is used for several years under high temperature and pressure, so it needs to be clad with high corrosion resistance material. The cladding materials must have the characteristics of low absorption of a neutron and high corrosion resistance. Zircaloy-2 in Boiling Water Reactor, Zircaloy-4 in Pressurized Water Reactor have been used as cladding materials and Zirlo has been developed as the material for preventing the corrosion. If the fracture of the cladding tube occurs during operation, it will cause the economic loss to shut down and replace the system. So it is needed to evaluate the integrity of the cladding materials. In this paper, the tensile characteristics of the cladding materials were investigated for the basic research of fracture characteristics. Also the residual stress was analyzed to compare the tube type(original type) specimen and the flattened type specimen.