Zircaloy Alloys Research Articles

The recycling through melting Zircaloy machining chips, preliminary results

PWR (Pressurized Water Reactor) reactors employ as nuclear fuel UO 2 pellets packed in zirconium alloy tubes, called cladding. In the manufacture of the tubes, machining chips are generated which can not be discarded, since the recycling of this material is strategic in terms of nuclear technology, legislation, economics and the environment. These nuclear alloys are very expensive and are not produced in Brazil and are imported for the manufacture of nuclear fuel. In this work, we will discuss methods not yet studied to recycle Zircaloy chips using electron beam furnaces in order to obtain ingots. In addition, it is intended to carry out the melting of new Zircaloy alloys, from the fusion of zirconium sponge produced in IPEN and imported and Zircaloy bars. The mechanical properties and the present phases of the material should be determined, as well as, the characterization of the microstructures by optical microscopy. This work, therefore, aims at the creation of a new line of research where methods will be approached to recycle the chips and to reduce in 30 times by means of fusion the enormous volume of material stored in the form of machining chips, being able to do others components for nuclear or chemical industry use, as well as conducting basic development research.

High-Temperature Oxidation of a Sponge-Based E110 Cladding Tube Material: Weight Gain and Reaction Layer Kinetics

High-temperature oxidation of WWER fuel cladding under loss-of-coolant accident conditions represents a significant problem and an issue of ongoing study. A study of oxidation kinetics under high-temperature steam oxidation was carried out on a sponge-based E110 cladding tube material in as-received state and exposed to a temperature range between 600°C and 1300°C. Metallographic evaluation of the samples was used to determine the analytical dependencies of the processes of the oxide layer and the oxygen stabilized alpha-zirconium layer formation.The formation of this oxide layer and the layer under the oxide (oxygen stabilized alpha-zirconium) was usually considered to follow a parabolic rate law (e.g., Cathcart-Pawel for Zircaloy, and Solyany for iodide/electrolytic E110), however, the recently performed UJP PRAHA a.s. (UJP) experiments showed different kinetics for oxide layers in the whole temperature range between 600°C and 1300°C. The metallographic evaluation showed changing kinetics of oxide layer formation for several temperatures. Therefore, a new oxide layer formation kinetics, different from the parabolic law, was designed. The parabolic rate law was optimized for the oxygen stabilized alpha-zirconium, which is applicable in the temperature range between 950°C and 1300°C.The UJP experimental database containing more than 1000 data values was compared with the newly developed UJP correlations and other available correlations for iodide/electrolytic E110 and Zircaloy alloys. The UJP correlation for weight gain kinetics was compared with experimental data from different laboratories. Finally, a statistical analysis of all tested correlations is provided.

Proposal of new oxidation kinetics for sponge base E110 cladding tubes material

Study of high temperature steam oxidation kinetics during the high temperature oxidation was carried out on the sponge base E110 cladding tubes material in the temperature range 600–1300 °C. The oxidation kinetics derived from the weight gain measurements showed a parabolic rate law for temperatures 1100 °C and higher only. For lower temperatures in range 800–1050 °C especially, the parabolic law leads to very conservative prediction. Therefore, the new oxidation kinetics, different from the parabolic law, was designed. The experimental database containing more than 800 data points was compared with the new developed UJP-correlation and available correlations for E110 and Zircaloy alloys. Statistical analysis for all tested correlations was provided.

Air oxidation of Zircaloy, Part 1 – Review of correlations

The accident in Fukushima brought up new issues in the area of safety of nuclear reactors. Among others, Spent Fuel Pool accidents gained new focus. The computer codes applicable for safety analyses of Nuclear Power Plants have limited verification and validation in this area and their applicability remains still to be proven. An important phenomenon occurring during loss of water in SFP is air oxidation of Zircaloy cladding material. Mathematical modeling of this phenomenon in computer codes has been under development during the last years. This document presents a review of models for air oxidation of Zircaloy alloys, including up to date models available in open literature, as well as models available in computer codes ASTEC and MELCOR. The models were tested by performing simulations of a number air oxidation experiments from ANL, KIT, and IRSN.It was found out that the ASTEC model gives somewhat better agreement with data than the MELCOR model for the ANL and KIT experiments. The reaction rates from the MELCOR model differ from experiments by factor of 2 and 3 respectively. In case of IRSN tests, the MELCOR model gives better agreement with data. The reaction rates from the ASTEC model differ from experiment by roughly a factor of 3.The difference between the reaction rates predicted by the models and obtained from experiments was found to be up to about factor of three. The ASTEC model gives generally qualitatively better representation of reality. A disadvantage is the use of Boase and Vandergraaf correlation outside its validity range (T>1000K) where it is underestimating the reaction rate. In the MELCOR model the following qualitative behavior is incorrect: breakaway occurs in the entire temperature range, while experiments show that it occurs only for temperatures below about 1050°C; post-breakaway reaction is parabolic, while experiments show linear reaction rate.It is concluded that the air oxidation models exist in the available codes but the accuracy of the models could be improved. Improvement of mathematical modeling of air oxidation is currently ongoing in various institutes, for example the PSI model in MELCOR and SCDAPSim code (Birchley and Fernandez-Moguel, 2012; Fernandez-Moguel, 2015), IRSN and GRS model, including modeling of the nitride itself in the ATHLET code (Hollands, 2007). An attempt to improve accuracy by providing a consistent set of correlations for a full temperatures range and pre- and post-breakaway regimes is a subject of the Part 2 of this paper.

The use of a laser-assisted Local Electrode Atom Probe and TEM to examine the microstructure of Zircaloy and precipitate structure following low dose neutron irradiation at nominally 358 °C

Previous research has shown that the irradiation of Zircaloy at very low fluences (⩽7×1022n/m2 (E>1MeV)) results in measurable irradiation hardening, but the defects or solute clusters responsible for such hardening are generally below the resolution limit of electron microscopy. Efforts to analyze Zr-based alloys using conventional atom probe have been largely unsuccessful due to the poor electrical conductivity of Zirconium at cryogenic temperatures. Laser-assisted atom probe methods were introduced in the 1980s as a means for enabling low electrical conductivity specimens to be analyzed. Some recent examples have been reported in the literature where a Local Electrode Atom Probe (LEAP)® was used to successfully analyze non-irradiated Zirconium based alloys, such as Zircaloy-4 and ZIRLO. In this work, the use of a LEAP to examine the microstructure of non-irradiated Zircaloy alloys is evaluated. Optimum conditions for applying the LEAP to non-irradiated Zircaloy-2 and Zircaloy-4 are investigated. Changes in the local composition of precipitates and grain boundaries following neutron irradiation at nominally 358°C to a fluence of 2.9×1025n/m2 in the High Flux Isotope Reactor (HFIR) are determined using Transmission Electron Microscopy. These results are provided as an example of the local changes in microstructure that can influence the evolution of the defect structure.

Failure characteristics of cladding tubes under RIA conditions

In the frame of actions for improving the safety of its nuclear power plants, Electricité de France needs to build the mechanical criteria ensuring the clad integrity for several operating conditions. This paper presents analytical mechanical models used to derive failure criteria for reactivity insertion accidents (RIA) and interpretation of the CABRI REP-Na experimental tests. Building analytical criteria requires an experimental database. Mechanical tests performed on non-irradiated and irradiated cladding tubes have been provided from French and international programmes (PROMETRA, EPRI, …). These tests consist of tube burst and axial tension, and ring tension. Several strain biaxiality ratios are thus available: pure circumferential tension (from ring tension), pure axial tension (from tube axial tension), and plane strain conditions (from tube burst tests). Several strain rates, temperatures, irradiation conditions are also available. The major feature of our study has been to make it possible that these several thermomechanical conditions be representative of “standard” RIA loading conditions. To this aim, we have derived some biaxiality and strain rate corrections to be applied to the results of experimental tests, in such a way that they could be representative of RIA biaxiality conditions (which are assumed to be strain equibiaxiality), and also representative of RIA strain rate conditions (which are assumed to be 5 s −1). The corrections that we derive are based on the fracture properties of hydrided zirconium alloys (especially in terms of anisotropy), and also on an assumed form of the material constitutive equations. Each test of the “homogenized” database has thus been used to calculate a strain energy density, representative of its fracture (the strain energy density is defined as the integral of the stress times strain rate states, over the duration of the mechanical test). The SED values are plotted against the sample's oxide thickness, and a lower bound limit can be established, with respect to oxide thickness. In order to address the problem of representativeness of the laboratory database, an experimental set-up has been developed that aims at characterizing the failure behavior of cladding tubes under RIA conditions. The developed experimental set-up is based on electromagnetic forming. The development of the test and in particular of the die is delicate but leads to repeatable results with a controlled strain biaxiality ratio higher than those obtained through conventional tests such as ring tests or PSU ring tests. The use of electromagnetic forming process allows testing the specimen with very high strain rates. For the next test series other zircaloy alloys at the reception state and in hydrided conditions will be tested in order to look at hydrides influence on fracture strains. A first finite element simulation of the test was engaged. The simulation and the experimental results are in quite good agreement. In future, the consistency of the previously developed analytical mechanical criterion with the electromagnetic forming experimental results will be verified.

High temperature oxidation of zirconium and zircaloy-4 under applied load: Nuclear microprobe study of the growth of the oxide

The understanding of the mechanisms of high temperature growth of zirconium oxide on zircaloy alloys is of considerable importance owing to the world-wide use of zirconium in nuclear plants. Spallations and growth recovery are known to originate in structural changes of the oxide, but the induced modifications in the oxygen transport have still to be investigated. The zirconia is known to grow mainly by anionic transport, but, due to the very high solubility of oxygen in zirconium, the oxygen penetration profile may be rather complex and extend beyond the oxide/metal interface. The nuclear microprobe offers an attractive way to investigate the oxygen penetration, both from RBS and NRA. In the case of parallel mechanical loading and high temperature oxidation, the oxygen diffusivity is very sensitive to the generated defects in the oxide scales. The use of alternate 16O/ 18O atmosphere and subsequent local microanalysis by means of 18O(p,α) 15N nuclear reaction allows to investigate the effect of the mechanical loading on the oxygen penetration.

Determination of iron in Zircaloy-2 and Zircaloy-4 by differential-pulse polarography

The alloy sample was dissolved in hydrofluoric acid and a portion of the resulting stock solution was heated with sulphuric acid to remove fluoride ion. Tin(II) chloride solution was then added to ensure complete reduction of Fe3+ to Fe2+. Solid sodium citrate was added to form a citrate complex with Fe2+. The polarographic peak of the Fe2+-citrate complex was then used as a measure of the percentage of iron in Zircaloy alloys.