[en] The effect of oxygen on thermal expansion, elastic moduli, and thermal diffusivity of Zircaloy-4 nuclear reactor fuel cladding was determined for the main purpose of providing baseline data for LOCA evaluation codes. Measurements were made over the temperature range 298 to 1473 K and 0.7 to 28 at.% oxygen. Expansion and moduli were measured in the two relevant directions, while diffusivity was measured in the principal heat transfer direction (through the clad). Thermal expansion and elastic moduli both increased with oxygen, while diffusivity decreased. Similar measurements were made on Zircaloy-2, but only to 5 at.% oxygen. Differences between the two alloys were within experimental error of the measurements made. (orig.)

[en] Previous investigators have reported the separate influences of chemical composition and heat treatment on the nodular and uniform corrosion of zirconium-based alloys. However, many of these compositions lie outside the allowable Zircaloy ranges required by most nuclear fuel specifications. The current study shows that the corrosion performance of the Zircaloys can be optimized by adjusting chemical compositions within current ASTM ranges, and examines the influence of composition on response to heat treatment. It is believed that this approach will yield improved corrosion performance and results in a product that can be introduced for immediate nuclear service application

[en] The steady-state creep behaviour of Zircaloy-4 have been studied over the temperature and stress range of 535-620degC and 6-9kg/mmsup(2) respectively. The following results were obtained; 1. The stress exponents, n were 5.40, 5.15, and 4.91 for temperature of 550,580 and 600degC respectively. 2. The apparent activation energies, Q were 67.9, 63.8 and 62.2kcal/mole for stress of 6,7,8, 9kg/mmsup(2) respectively. In other words the apparent activation energies for creep deformation decreased as the stress increased. 3. The activation energies for creep deformation were nearly equal to that of the volume self diffusion of Zr in the Zr-Sn-Fe-Cr system. (Author)

[en] Much work has been done to determine which compounds may be formed by the reaction of the volatile fission products cesium, iodine, and tellurium with each other and with the nuclear fuel (UO₂) and the Zircaloy cladding, since the behavior of these fission products in case of a reactor accident is governed by their chemical form. Recent measurements of the thermodynamic properties of various compounds involved make it possible to calculate more precisely, as a function of the oxygen potential and the temperature, the equilibrium composition of the chemical species present in the fuel-cladding gap. The computer program SOLGASMIX-PV, obtained from T.M. Besmann of ORNL, was used for this purpose

[en] The segmented expanding mandrel test is modified and improved to yield more quantitative information on the local stress and strain concentrations in Zircaloy tubing samples. The experimental apparatus has undergone changes in segment design, end restraint fixture design and the method of sample heating. These changes result in localization of stress and strain concentrations to known locations in the tubing, the introduction of a small degree of biaxiality and a more uniform sample temperature, respectively. A chamfered specimen design has been introduced for the purpose of producing a near plane-strain condition. Finite-element analysis has been added to provide information on the magnitude of localized stresses and strains given the sample geometry, test procedure and measured diametral strains. The resulting test represents a significant increase in the quantitative information produced in the segmented expanding mandrel technique. (orig.)

No abstract available

[en] Argon and iodine stress-rupture tests were performed on five lots of ZIRCALOY-4 tubing with relatively large differences in texture. The addition of iodine relative to argon decreases the failure time. The iodine data exhibited increasing failure times with decreasing stress until a plateau or threshold stress level was reached. The threshold stress was used to evaluate a model developed from fracture mechanics crack propagation data. Modification of this model was necessary in order to account for tubing texture, tubing fracture surface characteristics, test temperature, and the embrittling effect of iodine. The adjusted model predicts that moderate increases in the iodine threshold stress may be obtained with very low tangential texture tubing

[en] The results of the single effect laboratory investigations, UO₂/Zircaloy interactions and dissolving of UO₂ by molten Zircaloy, and the single rod and bundle experiments, which were previously done in the NIELS test rig, are described. Some information is given on the CORA test rig. The experiments show a comprehensive and detailed understanding of the material behaviour of the different bundle and core components at temperatures above 1200⁰C. Apart from oxidation of the can in steam, the chemical reactions between the cans and the fuel in the preliminary field before core melt-out can be described quantitatively. (DG)

[en] The application of statistical distribution of quality characteristics to process control for Zircaloy fuel cans, the control of fuel can manufacture and the determination of the elastic limit in tensile tests on Zircaloy fuel cans. All the important questions can be examined and answered with the method shown here. (DG)

No abstract available