Jean-Christophe Brachet

Studies of second phase particles in different zirconium alloys using extractive carbon replica and an electrolytic anodic dissolution procedure

Zirconium alloys are widely studied for applications as cladding tubes and structural components of PWR fuel assemblies. Due to their influence on some of the alloys properties (corrosion resistance, irradiation growth, …), the crystallographic structure and the chemical stoichiometry of the second phase particles (SPP) precipitated in these alloys have to be well established. The aim of this paper is to present the results obtained using two methods of SPP extractions. The first one, the extractive carbon replica method, allowed us to determine the chemical composition of SPP in different zirconium alloys: Zr–Sn–Fe–Cr (Zircaloy-4®), Zr–Sn–Fe–Cr–(V,Mo), Zr–Nb and Zr–Nb–Fe alloys. The second one, an anodic dissolution procedure of the matrix, is an interesting way of isolating SPP from the surrounding α-Zr matrix, giving access to a precise determination of the crystallographic structure and lattice parameters of the SPP by X-ray diffraction. This procedure was validated for Zy-4 by comparing the SPP size distribution obtained by extraction with that directly measured on a massive Zy-4 alloy (i.e. the SPP size distributions were the same for both measurements).

Relationship between Crystallographic Texture and Dilatometric Behaviour of a Hexagonal Polycrystalline Material

The main objective of the present paper is to describe the relationship between the texture of polycrystalline samples having hexagonal lattice symmetry with their dilatometric behaviour during continuous heating. For this purpose, a thick Zr base alloy tube (Zr>98%) has been used. The initial texture of the tube was characterised by x-ray diffraction techniques using pole figures and the Orientation Distribution Function (ODF). The experimental results highlight a strong influence of the cutting direction of samples on the dilatometric behaviour. Introducing the single crystal Coefficients of Thermal Expansion (CTE) obtained from the literature, and using the coefficients of the spherical harmonics development of the ODF, it has been possible to calculate the CTE of textured polycrystals in the α phase field along any direction of the sample. The calculated CTE compare quite well with the experimental ones. Moreover, for alloys with a typical α grain size of 15μm in diameter, the experimental dilatometric contraction/dilation associated with the α to β phase transformation upon heating has been correlated to the CTE in the α-phase field and can therefore be correlated to the initial texture.

Mechanical Behavior at High Temperatures of Highly Oxygen- or Hydrogen-Enriched α and Prior-β Phases of Zirconium Alloys

Mechanical behavior at high temperature of highly oxygen-or hydrogen-enriched α and (prior-) β phases of zirconium alloys ABSTRACT: During a hypothetical loss-of-coolant accident (LOCA), zirconium alloy fuel claddings can be loaded by internal pressure and exposed to steam at high temperature (HT, potentially up to 1200°C), then cooled and water quenched. A significant fraction of the oxygen reacting with the cladding during HT oxidation diffuses beneath the oxide through the metallic substrate. This induces a progressive transformation of the metallic βZr phase layer into an intermediate layer of αZr(O) phase containing up to 7 wt.% of oxygen. Furthermore, in some specific conditions, the cladding may rapidly absorb a significant amount of hydrogen during steam exposition at HT. Being a βZr-stabilizer, hydrogen would mainly diffuse and concentrate up to several thousands of wt.ppm into the inner βZr phase layer. Oxygen and hydrogen are known to modify the metallurgical and mechanical properties of zirconium alloys but data are scarce for high contents, especially at HT. However, such data are important basic components to improve the assessment of the oxidized cladding mechanical behavior and integrity during and after LOCA-like thermal-mechanical transients. This study intended to provide new, more comprehensive data on the HT mechanical behavior of the αZr(O) and the (prior-) βZr phases containing high contents of oxygen and hydrogen, respectively. Model samples, produced from M5® 5 and Zircaloy-4 cladding tubes, homogeneously charged in oxygen (≤6 wt.%) and in hydrogen (≤3000 wt.ppm) respectively, were prepared. Their mechanical behavior was determined under vacuum between 800 and 1100°C for the oxygen-enriched αZr phase, and in air between 700 and 20°C, after cooling from the βZr temperature domain, for the hydrogen-enriched (prior-) βZr phase. The αZr phase is substantially strengthened and embrittled by oxygen. Power-law and nearly linear creep regimes are observed and were modelled for stress levels beyond and below 15 MPa, respectively. The model αZr(O) material experiences a ductile-to-brittle transition at 1000-1100°C for oxygen contents between 3.4 and 4.3 wt.%. The viscoplastic behavior of the αZr(O) phase was used to evaluate the contribution of the αZr(O) layer to the HT creep behavior of an oxidized fuel cladding tube subjected to internal pressure. The model (prior-) βZr phase becomes macroscopically brittle at temperatures ≤135°C and ≤350-400°C for average hydrogen contents

Atomic-scale interface structure of a Cr-coated Zircaloy-4 material

Highly adherent, thin Cr coatings on Zr-based nuclear fuel claddings can be potentially used for the development of accident-tolerant fuels in light water reactors. To guarantee the successful implementation of Cr-coated Zr alloys as cladding tubes in nuclear power plants, the adhesive strength of the Cr coatings must be assessed. The interface between Cr and Zr was characterized via high-resolution transmission electron microscopy. We observed the formation of nanometer-thick Zr(Fe, Cr)2 poly-type, structured Laves phases at the interfacial region that display both C14 and C15 lattice symmetries. Although the crystallinity was preserved throughout the interfacial region, different atomic configurations were observed for all the interfaces studied. In most cases, coherent or semicoherent crystallographic relationships were observed, ensuring the adhesive strength of the coating.

Microstructure and Properties of a Three-Layer Nuclear Fuel Cladding Prototype Containing Erbium as a Neutronic Burnable Poison

To increase cycle length and/or fuel burnup, several theoretical and experimental studies have been performed at CEA. Among them, prospective neutronic calculations have shown that the addition of a few weight percents of erbium into the cladding materials could be a promising alternative to the introduction of the neutronic poison directly into the nuclear fuel pellets. Thus, fabrication of homogeneous Zr-Er alloys has been assessed, at least up to 10 wt. % of erbium and, based on the as-received mechanical properties, an optimum erbium concentration ranging from 3 to 6 wt. % has been derived. However, because of the high-oxygen thermodynamic affinity of erbium, thermal treatments have to be controlled during the fabrication route to limit Er2O3 precipitation and coarsening, which may have detrimental effects on the ductility/toughness of Zr-Er alloys. In parallel, to get more fundamental insights into the underlying phase diagrams, thermodynamic studies have been devoted to experimental assessment and modeling of the Zr-Er-(H-O) system. Because of the detrimental influence of erbium on the corrosion resistance, a three-layer sandwich clad prototype has been developed using corrosion-resistant inner/outer Zr-1Nb layers to protect the intermediate Zr-Er layer from direct water exposure. Compared to a reference Zr-1Nb(O) alloy that has been subjected to the same fabrication route, the three-layer clad prototype shows limited decrease in ductility because of pre-hydriding or after high-temperature steam oxidation e.g., in the case of a loss-of-coolant accident). Moreover, the studies performed so far have shown a spectacular hydride trapping capacity of the intermediate Zr-Er layer both for hydrogen coming from nominal outer corrosion or because of massive secondary hydriding in case of the direct access of water to the Zr-Er intermediate layer. Using μ-ERDA (elastic recoil detection analysis) measurements, detailed studies of the hydrogen spatial redistribution upon thermal cycling has been done. A simple model has been successfully used to characterize the cooling rate influence on the through-wall clad thickness partitioning of hydrogen/hydrides between the three layers, after cooling from a temperature corresponding to full dissolution of hydrides