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99Mo is an essential medical isotope that comprises of at least 70% of radioactive procedures globally. Currently an essential component of 99Mo manufacturing is the uranium target from which 99Mo is produced by fission. As the world moves towards low enriched uranium (LEU) targets due to non-proliferation concerns it is becoming of interest to find methods to increase the efficiency of the LEU targets in order to reduce the ever increasing nuclear waste levels of which a long term solution for disposal or treatment has yet to be satisfactorily found. Advantages and disadvantages of various target designs are investigated and discussed along current disposal and reprocessing methods. The idea of a reusable target is introduced as a way forward in reducing the nuclear waste burden for future generations. • Fission based manufacturing of 99Mo is moving towards Low Enriched Uranium methods. • Low Enriched Uranium leads to increased radioactive waste volumes. • The majority of uranium targets are single use and highly inefficient. • Reusable targets have the potential to lower the nuclear waste burden. [ABSTRACT FROM AUTHOR]

Abstract The simple conversion of zirconium phosphonate sorbent materials, with known affinity for lanthanide elements, to durable ceramic waste forms via thermal treatment has been demonstrated. The use of zirconium phosphonate enables both removal of targeted elements from spent nuclear fuel and immobilisation into leach resistant solid products to be achieved using a single material. Thermal conversion was performed on the zirconium phosphonate both before and after loading with europium, which acted as a surrogate for the chemically similar minor actinides. Without europium loaded, the zirconium phosphonate sorbent formed predominantly KZr 2 (PO 4) 3 upon heating, independent of the processing conditions used. A maximum relative density of 87% was achieved with cold isostatic pressing (200 MPa) and sintering at 1200 °C for 12 h. When the zirconium phosphonate sorbent was loaded with europium, the phase composition formed upon thermal treatment was more complex. Specifically, mixtures of ZrP 2 O 7 , Eu 0.33 Zr 2 (PO 4) 3 , EuPO 4 and Zr 2 O(PO 4) 2 were formed, with phase compositions depending on the temperatures and pressures used. The simplest phase composition for the europium loaded material was achieved via uniaxial pressing (120 MPa) and sintering at 1300 °C for 1 h, although the ceramic pellet produced under these conditions had a relative density of only 53%. The loaded europium deported primarily to a EuPO 4 phase, which is known to be highly stable and leach resistant. As such, these zirconium phosphonate materials have potential utility for treatment of nuclear wastes. [ABSTRACT FROM AUTHOR]

In order to obtain insight into the hardening mechanism of Cr(N,O) thin films with respect to increase in oxygen content, the microstructure of Cr(N,O) grains in thin films was studied. High resolution transmission electron microscopy provided evidence that the grains in Cr(N,O) thin films prepared by pulsed laser deposition, which have an oxygen content of > 27 mol%, contained stacking faults. Phase identification via electron diffraction, indicated that the grains consisted of two phases, which had a NaCl-type and corundum-type structure and these phases had the same orientation as that of the Cr 0.67 O thin film. The estimated stacking fault energy of Cr(N,O) with an oxygen content of 34 mol% was 20 mJ/m 2 . Our results revealed the Cr(N,O) grains had the same nano-lamellar morphology as that of the Cr 0.67 O thin film which supports the possibility that the nano-lamellar morphology was formed by the introduction of extended dislocations. It is also possible that the hardening of Cr(N,O) was caused by dislocation pinning at the boundaries of the nano-lamellar morphology. [ABSTRACT FROM AUTHOR]

CuI is a candidate wasteform for the immobilization of the fission product 129 I. CuI can be made simply by the addition of CuCl to an I − bearing solution such that exchange of Cl − with I − takes place. The CuI material can then be consolidated into a wasteform by sintering at approximately 550 °C in argon or by hot isostatically pressing at 550 °C with 100 MPa of pressure. A waste loading of greater than 60 wt.% is achievable with good water leach resistance, in keeping with the low solubility product of CuI. However, like the well known wasteform candidate AgI, CuI decomposes in water containing metallic Fe. To compensate this deficiency, the sintered CuI wasteform can be further protected by surrounding it by Sn powder and HIPing at the low temperature of 200 °C. [ABSTRACT FROM AUTHOR]

A new class of materials, which rely on the dispersion strengthening of SiC particles in addition to precipitation strengthening by nano-precipitates is being developed for the application in molten salt nuclear reactors. A battery of dispersion and precipitation strengthened (DPS) NiMo-based alloys containing varying amount of SiC (0.5–2.5 wt.%) was prepared via a mechanical alloying (MA) route followed by spark plasma sintering (SPS), rapid cooling, high-temperature annealing and water quenching. Lab X-ray Diffraction (XRD), Electron Back Scattering Diffraction (EBSD), and Transmission Electron Microscopy (TEM) were employed in the microstructural characterization of this new type of alloys. It is shown that the NiMo matrix of these alloys is effectively reinforced by dispersion of SiC from the initial powder mixture and nano-Ni 3 Si precipitates, which precipitated during the sintering/annealing process. Furthermore, the matrix is strengthened by solid-solution of Mo in Ni. As a result, these newly developed NiMo alloys take advantage of dispersion, precipitation and solid solution strengthening, which leads to their superior mechanical properties. [ABSTRACT FROM AUTHOR]

Copper-doped nanosized zeolite A crystals were synthesized by an in situ templating approach using [Cu(EDTA)] 2− -complex. The structural properties of the copper containing zeolite crystals were characterized by a suite of different techniques including SEM–EDX, ESR, mid-IR and Far-IR, Raman, in situ XRD and non-ambient neutron powder diffraction. The SEM investigations on the morphology show spheroidal zeolite A crystals with average size ∼200 nm. The asymmetric ESR spectrum shows that the Cu 2+ ion is in a tetragonal-distorted octahedral crystal field. FT-IR and Raman spectroscopies provide information on coordination environment of the copper ion. The band due to stretching vibration of C–N bond, where N is coordinated to the copper ion (C–N–Cu), was observed at 1109 cm −1 in the mid-infrared region. The Raman band due to the Cu–O bond is present at 630 cm −1 indicating the coordination of the Cu 2+ -cation to COO − -group of the EDTA-ion. The XRD data shows an enlarged d -spacing between the adjacent zeolite lattice planes due to the presence of the [Cu(EDTA)] 2− -complex in comparison to template-free LTA zeolite structure. LeBail fitting approach on temperature-dependent in situ X-ray and neutron diffraction profiles have demonstrated the expansion of the zeolite cell during the thermal treatment followed by subsequent contraction with the decomposition of the organic template. [ABSTRACT FROM AUTHOR]

The β -phase of Au 7 Cu 5 Al 4 undergoes a reversible shape-memory phase transformation for which several conflicting martensite phases have been reported. Here we show the significance of the cooling temperature used to obtain the martensite. If Au 7 Cu 5 Al 4 is cooled from the parent phase condition to cryogenic temperatures, e.g. below 200 K, the martensitic phase is orthorhombic (space group Pcmn , a = 4.4841 Å, b = 5.8996 Å, c = 17.8130 Å); however, when this composition is cooled to only ∼260 K it will in general consist of a mixture of orthorhombic and monoclinic phase (the latter has space group P 2 1 / m , a = 4.4742 Å, b = 5.9265 Å, c = 13.3370 Å, β = 91.425°). In contrast, a sample with decreased Al content (Au 7 Cu 5.7 Al 3.3 ) transforms fully to monoclinic phase if cooled to ∼260 K. [ABSTRACT FROM AUTHOR]

Current fission-based methods of 99Mo production require single use uranium targets which leads to spent uranium waste. This waste could be reduced if a target is developed that does not require dissolution so that it can be reused for multiple production runs. MCNP6.2 was used to model reusable targets of 20% and 1% enrichment for activity produced, target efficiency and burnup. The 1% enriched target was found to be much more efficient but had a lower activity produced compared to the 20% enriched target. The ideal target design for 99Mo production that optimises efficiency and reusability and reduces the self-shielding effect of UO 2 was found to be a target that is made from 1% enriched UO 2 with density as high as allowable for sufficient yields, efficient 99Mo extraction and having an irradiation time of 5 days, with the target able to be re-irradiated and re-processed 2–4 times. • Current fission based manufacturing of 99Mo using uranium targets is based on target destruction. • Target destruction leads to increased radioactive waste volumes. • MCNP6 modelling of uranium targets show that higher density targets have more self-shielding effects and are less efficient. • Low density uranium targets with 5 days irradiation maximise 99Mo output and minimising radioactive waste. [ABSTRACT FROM AUTHOR]