Studies were made into the influence of oversized rare-earth atoms on the processes of radiation defect accumulation and annealing in two-component zirconium alloys. Zr and Zr-X alloys (where X = Sc, Dy, Y, Gd and La) have been irradiated with 2 MeV electrons at 82 K. The radiation-induced resistivity has been measured in situ as a function of dose. As compared to unalloyed zirconium, the alloys have exhibited a decrease in the resistivity gain, this decrease being proportional to both the concentration and the size of dopant atoms. A possible explanation for the effect is offered. The difference between the recovery processes in zirconium and in its alloys has been studied. To this end, the irradiated specimens were subjected to isochronal annealing at temperatures between 82 and 350 K. It is shown that Dy, Y, Gd and La atoms trap interstitial atoms at stage I of the recovery. The dissociation of interstitial-impurity complexes takes place at stage II. In zirconium alloys with Dy, Y and Gd, splitting of recovery stage III into two substages has been revealed. The Zr-La alloy has not shown this splitting. Isothermal annealing data were used to determine the activation energies of recovery stages, and also to calculate the activation energy spectra for zirconium and its alloys. The oversized atoms of rare-earth metals are shown to interact effectively with both the interstitials and the vacancies in the zirconium matrix. This effect must be taken into account when developing new radiation-resistant Zr-base alloys or modifying the ones already existing.

The grain boundary has been recognized for one of the major defect structures in determining the material strength. It is increasingly important to understand the individual characteristics of various types of grain boundaries due to the recent advances in material miniaturization technique.

In the present study three types of grain boundaries of coincidence site lattice (CSL), small angle (SA), and random types are considered as the representative example of grain boundaries. The grain boundary energies and atomic configurations of CSL are first evaluated by first-principle density functional theory (DFT) and the embedded atom method (EAM) calculations. SA and random grain boundaries are subsequently constructed by the same EAM and the fundamental characteristics are investigated by the discrete dislocation mechanics models and the Voronoi polyhedral computational geometric method. As the result, it is found that the local structures are well accorded with the previously reported high resolution-transmission electron microscope (HR-TEM) observations, and that stress distributions of CSL and SA grain boundaries are localized around the grain boundary core. The random grain boundary shows extremely heterogeneous core structures including a lot of pentagon-shaped Voronoi polyhedral resulting from the amorphous-like structure.

Efforts have been made to develop long term sustainable materials for use above 900°C in the SO3/SO2 environments utilized in Nuclear Hydrogen Production Systems. In this study, the surface of Hastelloy X was serially modified by evaporative SiC coating, ion beam mixing (IBM), additional coatings, and final ion beam hammering (IBH). Subsequent heating above 900°C results in no peeling-off of the SiC coating layer in spite of the huge difference in their coefficients of thermal expansion (CTE). It was also found that ion beam hammering (irradiation) suppresses the vacuum sublimation of the low density (∼40% of bulk) ceramic film (bulk materials begin to sublime at ∼ 750°C in a vacuum of 1.5×10−5 torr). The sublimation rate is ≤ 30% of the bulk rate after an annealing at 950 for 2 hrs but is decreased to ≤10% of the bulk rate after irradiation with 70 keV ions to a total dose of 1×1017 N ions/cm2. Further irradiation (up to 4×1017 N ions/cm2) does not further decrease the rate. Both an immersion test in 98% sulfuric acid and the potentiodynamic polarization test suggest that the surface modified Hastelloy X has a greatly prolonged life time in the corrosive sulfuric acid atmosphere, suggesting the serial surface modification process is applicable to the thermo-chemical system for the nuclear hydrogen production.

Technetium has a long half life of up to 2.13×105 years. It is separated from liquid waste streams with tetraphenylphosphonium bromide [1], which upon degradation releases Tc as the pertechnetate anion, TcO4−. Pertechnetate is highly mobile in groundwater and it is therefore highly desirable to capture and immobilise this anion within a solid for interim and ultimately long term storage. Layered Double Hydroxide (LDH) materials are known to possess excellent anion sorption capabilities due to their structure which consists of ordered positively charged sheets intercalated with interchangeable hydrated anions. The composition can be tailored to produce suitable precursors for ceramic phases by varying the divalent and trivalent cations and the anions. LDHs with the general formula Ca1-x (Fe1-y, Aly)x (OH)2 (NO3)x . nH2O were produced by a co-precipitation method from a solution of mixed nitrates. Calcination leads to the formation of Brownmillerite Ca2(Al,Fe)2O5 like compounds for temperatures as low as 400°C, this is close to the lowest temperature at which Tc is known to volatilise (310.6 °C Tc2O7). It was shown that after calcining up to 600°C, the LDH structure is recovered in water allowing rapid ion capture to occur. This suggests that these materials have potential for both capture and as a storage medium for Tc.

Crystal and interfacial structures of oxide nanoparticles in 16Cr-4Al-2W-0.3Ti-0.3Y2O3 ODS ferritic steel have been examined using high-resolution transmission electron microscopy (HRTEM) techniques. Oxide nanoparticles with a complex-oxide core and an amorphous shell were frequently observed. The crystal structure of complex-oxide core is identified to be mainly monoclinic Y4Al2O9 (YAM) oxide compound. Orientation relationships between the oxide and matrix are found to be dependent on the particle size. Large particles (> 20 nm) tend to be incoherent and have a spherical shape, whereas small particles (< 10 nm) tend to be coherent or semi-coherent and have a faceted interface. The observations of partially amorphous nanoparticles lead us to propose three-stage mechanisms in order to rationalize the formation of oxide nanoparticles containing core/shell structures in as-fabricated ODS steels.

Pre-irradiated thermodynamic and microstructural properties of nuclear fuels form the necessary set of data against which to gauge fuel performance and irradiation damage evolution. This paper summarizes recent efforts in mixed-oxide and minor actinide-bearing mixed-oxide ceramic fuels fabrication and characterization at Los Alamos National Laboratory. Ceramic fuels (U1-x-y-zPuxAmyNpz)O2 fabricated in the compositional ranges of 0.19≤x≤0.3 Pu, 0≤y≤0.05 Am, and 0≤z≤0.03 Np exhibited a uniform crystalline face-centered cubic phase with an average grain size of 14μm; however, electron microprobe analysis revealed segregation of NpO2 in minor actinide-bearing fuels. Immersion density and porosity analysis demonstrated an average density of 92.4% theoretical for mixed-oxide fuels and an average density of 89.5% theoretical density for minor actinide-bearing mixed-oxide fuels. Examined fuels exhibited mean thermal expansion value of 12.56×10−6/°C-1 for temperature range (100°C<T<1500°C) and ambient temperature Young's modulus and Poisson's ratio of 169 GPa and of 0.327, respectively. Internal dissipation as determined from mechanical resonances of these ceramic fuels has shown promise as a tool to gauge microstructural integrity and to interrogate fundamental properties.

Fine bulk samples of delta-phase Hf hydride with various hydrogen contents (CH) ranging from 1.62 to 1.72 in the atomic ratio (H/Hf) were prepared, and their thermal and mechanical properties were characterized. In the temperature range from room temperature to around 650 K, the heat capacity and thermal diffusivity of the samples were measured and the thermal conductivity was evacuated. The elastic modulus was calculated from the measured sound velocity. The Vickers hardness was measured at room temperature. Effects of CH and/or temperature on the properties of Hf hydrides were discussed. At room temperature, the thermal conductivity values of the Hf hydrides were 23 Wm−1K−1. The Young's and shear moduli and the Vickers hardness of Hf hydride decreased with increasing CH.