Microstructural Effects on Irradiation Creep of Reactor Core Materials

The processes that control irradiation creep are dependent on the temperature and the rate of production of freely migrating point defects, affecting both the microstructure and the mechanisms of mass transport. Because of the experimental difficulties in studying irradiation creep, many different hypothetical models have been developed that either favour a dislocation slip or a mass transport mechanism. Irradiation creep mechanisms and models that are dependent on the microstructure, which are either fully or partially mechanistic in nature, are described and discussed in terms of their ability to account for the in-reactor creep behaviour of various nuclear reactor core materials. A rate theory model for creep of Zr-2.5Nb pressure tubing in CANDU reactors incorporating the as-fabricated microstructure has been developed that gives good agreement with measurements for tubes manufactured by different fabrication routes having very different microstructures. One can therefore conclude that for Zr-alloys at temperatures < 300 °C and stresses < 150 MPa, diffusional mass transport is the dominant creep mechanism. The most important microstructural parameter controlling irradiation creep for these conditions is the grain structure. Austenitic alloys follow similar microstructural dependencies as Zr-alloys, but up to higher temperature and stress ranges. The exception is that dislocation slip is dominant in austenitic alloys at temperatures < 100 °C because there are few barriers to dislocation slip at these low temperatures, which is linked to the enhanced recombination of irradiation-induced point defects.

Process Optimization and Tailored Mechanical Properties of a Nuclear Zr-4 Alloy Fabricated via Laser Powder Bed Fusion

A nuclear Zr-4 alloy with a near full density was fabricated via laser powder bed fusion (LPBF). The influences of process parameters on the printability, surface roughness, and mechanical properties of the LPBF-printed Zr-4 alloy were investigated. The results showed that the relative density of the Zr-4 alloy samples was greater than 99.3% with the laser power range of 120-160 W and the scanning speed range of 600-1000 mm/s. Under a moderate laser power in the range of 120-140 W, the printed Zr-4 alloy possessed excellent surface molding quality with a surface roughness less than 10 µm. The microstructure of the printed Zr-4 alloy was an acicular α phase with an average grain size of about 1 µm. The Zr-4 alloy printed with a laser power of 130 W and a scanning speed of 400 mm/s exhibited the highest compression strength of 1980 MPa and the highest compression strain of 28%. The findings demonstrate the potential in the fabrication of complex Zr-4 alloy parts by LPBF for industrial applications.

Novel α + β Zr Alloys with Enhanced Strength

Low-alloyed zirconium alloys are widely used in nuclear applications due to their low neutron absorption cross-section. These alloys, however, suffer from limited strength. Well-established guidelines for the development of Ti alloys were applied to design new two-phase ternary Zr alloys with improved mechanical properties. Zr-4Sn-4Nb and Zr-8Sn-4Nb alloys have been manufactured by vacuum arc melting, thermo-mechanically processed by annealing, forging, and aging to various microstructural conditions and thoroughly characterized. Detailed Scanning electron microscopy (SEM) analysis showed that the microstructural response of the alloys is rather similar to alpha + beta Ti alloys. Duplex microstructure containing primary alpha phase particles surrounded by lamellar alpha + beta microstructure can be achieved by thermal processing. Mechanical properties strongly depend on the previous treatment. Ultimate tensile strength exceeding 700 MPa was achieved exceeding the strength of commercial Zr alloys for nuclear applications by more than 50%. Such an improvement in strength more than compensates for the increased neutron absorption cross-section. This study aims to exploit the potential of alpha + beta Zr alloys for nuclear applications.

Site-specific specimen preparation for SIMS analysis of radioactive samples

Secondary Ion Mass Spectrometry is an important technique for the study of the composition of a wide range of materials because of the exceptionally high sensitivity that allows the study of trace elements and the ability to distinguish isotopes that can be used as markers for reactions and transport processes. However, when studying nuclear materials, it is often necessary to analyse highly radioactive samples, and only rather few SIMS facilities are available in active environments. In this paper, we present a methodology using focussed ion beam milling to prepare samples from radioactive specimens that are sufficiently large to undertake SIMS mapping experiments over microstructurally significant regions, but with overall activities small enough to be readily transported and analysed by a SIMS instrument in a normal laboratory environment. Radioactive samples prepared using this methodology can also be used for correlative SIMS analysis with other analytical microscopies. SIMS results showing the distributions of deuterium in oxides on in-reactor corroded zirconium alloys are presented to demonstrate the potential of this sample preparation technique.

Hydrogen Embrittlement and Oxide Layer Effect in the Cathodically Charged Zircaloy-2

The present paper is aimed at determining the less investigated effects of hydrogen uptake on the microstructure and the mechanical behavior of the oxidized Zircaloy-2 alloy. The specimens were oxidized and charged with hydrogen. The different oxidation temperatures and cathodic current densities were applied. The scanning electron microscopy, X-ray electron diffraction spectroscopy, hydrogen absorption assessment, tensile, and nanoindentation tests were performed. At low oxidation temperatures, an appearance of numerous hydrides and cracks, and a slight change of mechanical properties were noticed. At high-temperature oxidation, the oxide layer prevented the hydrogen deterioration of the alloy. For nonoxidized samples, charged at different current density, nanoindentation tests showed that both hardness and Young’s modulus revealed the minims at specific current value and the stepwise decrease in hardness during hydrogen desorption. The obtained results are explained by the barrier effect of the oxide layer against hydrogen uptake, softening due to the interaction of hydrogen and dislocations nucleated by indentation test, and hardening caused by the decomposition of hydrides. The last phenomena may appear together and result in hydrogen embrittlement in forms of simultaneous hydrogen-enhanced localized plasticity and delayed hydride cracking.

The Application of Chemical Polishing in TEM Sample Preparation of Zirconium Alloys

Hydride artefacts are commonly induced by the TEM sample preparation process in Zirconium alloys as hydrogen-sensitive metals, including electron polishing and focused ion beam (FIB) technology. In the research, we present the application of chemical polishing with a solution of 10HF:45HNO₃:45H₂O to prepare the disk samples for TEM observation in zirconium alloys. The thinning efficiency of chemical polishing is 25 μm per minute. XRD patterns indicate that the chemical polishing actually eliminates the macro- and micro-stress induced by mechanical grinding. TEM observation demonstrates that chemical polishing reduces the amount of hydride artefacts, especially hydrides with large size. It is proposed that induced stress provides driving force for hydride artefact formation. Compared with traditional mechanical grinding, the advantages of chemical polishing are high efficiency, free of induced stress, less induced hydride artefacts and bend contours.