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Abstract: The capture of xenon (Xe) and krypton (Kr) from the off-gas of used nuclear fuel is of great importance to the treatment of radioactive wastes and production of high purity Xe. Solid sorbents, in particular metal–organic frameworks (MOFs), show promise in gas capture. However, the unknown radiation resistance of MOFs has limited their development. Herein, the efficient capture and separation of Xe/Kr by MFM-520, which strikes a remarkable stability toward 1750 kilogray (kGy) γ-irradiation, is reported. Under ambient conditions, dynamic breakthrough experiments confirm the efficient separation performance, yielding a Xe capacity of 66 and 0.2 mg g−1 from a by-product of air separation (Xe/Kr: 20/80; v/v) and off-gas (Xe/Kr: 400/40 ppm balance in air), respectively. In situ synchrotron X-ray single crystal diffraction and solid-state nuclear magnetic resonance (ssNMR) studies reveal that the optimal micropore of MFM-520 underpins specific host-guest interactions to Xe, resulting in selective Xe capture.

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Abstract: Waterside corrosion of zirconium (Zr) alloy nuclear fuel cladding occurs during normal operating conditions in pressurized water reactors. This corrosion yields zirconium oxide (ZrO2) and hydrogen, with a fraction of the hydrogen absorbed into the fuel cladding where it can redistribute through solid-state diffusion via concentration, temperature, and stress gradients. If the local hydrogen solubility in the metal is exceeded, then zirconium hydrides (ZrHx) precipitate in the zirconium matrix. Zirconium hydrides negatively impact the integrity of zirconium-based nuclear fuel cladding during both normal operation and extended dry storage through cladding embrittlement. With the desire to operate reactors with higher fuel enrichment and to higher fuel burnups, which will result in additional corrosion to the zirconium alloys, modeling hydrogen and hydride behavior is fundamental to ensuring safe reactor operations. Thus, this study aims to understand how applied tensile stress and irradiation-induced defects influence hydride precipitation and dissolution in zirconium alloys. In-situ synchrotron X-ray diffraction experiments were performed to study these influences in Zircaloy-4 under different thermo-mechanical conditions and levels of irradiation fluence. Additional analysis of Zircaloy-4 coupons under applied stress was also performed to explore the effect of stress on hydrogen migration and hydride dissolution. The applied tensile stress studies indicate that stress does not influence the dissolution solvus or the temperature hydrides completely dissolved at. Furthermore, the studies suggest that hydride nucleation is not noticeably changed by an applied tensile stress. However, hydride precipitation, defined by the Terminal Solid Solubility for Precipitation (TSSp), appears to slow under applied tensile stresses greater than or equal to 180 MPa. Furthermore, previous isothermal hydride growth experimental results were replicated and expanded on. Previous results were confirmed and indicate that applied tensile stresses slow hydride growth when the applied tensile stress is above the hydride reorientation threshold stress. Isothermal hydride growth did not result in the concentration of hydrogen in solid solution (Css) reaching the Terminal Solid Solubility for Dissolution (TSSd) during an isothermal hold, potentially reaching a quasi-equilibrium. It is hypothesized that the applied tensile stress impacts hydride growth rather than hydride nucleation. A potential explanation is that hydrides precipitating in different orientations due to the applied tensile stress have differences in hydride structure and dislocation networks. Studies on the influence of irradiation-induced defects on hydride precipitation and dissolution in as-irradiated and post-annealed materials were performed. Annealing the samples removed -dislocations per the reduction in Full-Width at Half-Maximums (FWHMs) for alpha-Zr diffraction peaks. However, the FWHM of the alpha-Zr (002) peak did not change during the anneal, indicating that -dislocation loops are still present and may still act as hydrogen traps. The delta (111) hydride diffraction peak integrated area was measured and found to increase from the as-irradiated to post-annealed state. This indicates hydrogen trapped by irradiation-induced defects in the material was released into solid solution and is available to form hydrides. The amount of hydrogen trapped by irradiation-induced defects was estimated at 12--63 wppm (8--47% of the total hydrogen content), which agrees with values reported in literature. TSSp and TSSd equations were fit to the data in both material states. Compared to the post-annealed data and literature, the as-irradiated TSSp and TSSd were lower. This is the opposite trend reported in literature for as-irradiated data. Little hydride growth during isothermal holds was observed in both the as-irradiated and post-annealed states.

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Abstract: Colloids present a challenge for nuclear decommissioning and disposal due to their potential to mobilise radionuclides. Waste retrieval and decommissioning of storage ponds for spent nuclear fuel and silos for radioactive waste at the Sellafield nuclear facility, UK, are high priorities. The particulates characterised here originate from facilities >60 years old and provide a unique opportunity to investigate the long-term fate of radionuclides in an aquatic, engineered storage environment. Radioactive effluents were obtained from a legacy pond and characterised using ultrafiltration, transmission electron microscopy (TEM) and actinide L3 edge X-ray absorption spectroscopy (XAS). TEM analysis showed discrete UO2-like nanoparticles, 5-10 nm in size, often co-associated with Mg-Al- and Fe-(oxyhydr)oxide colloidal phases. Uranium XAS indicated a mix of uranium oxidation states with EXAFS suggesting U(IV)-oxide nanoparticles and sorbed U(VI). Pu XANES identified Pu(IV) as the dominant oxidation state. Both U and Pu associates with large, Mg/Al- and Fe-(oxyhydr)oxide agglomerates highlights the potential for pseudo-colloid formation, explaining the basis of current particle filtration / abatement of technology. This study, which examines novel samples from a complex, highly radioactive facility using advanced techniques, provides a new understanding of radionuclide speciation and mobility in these environments and informs radioactive effluent treatment and disposal.

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Abstract: Geopolymers are promising materials for safe immobilisation and disposal of complex radioactive waste streams. This work investigates the effect of Sr incorporation and alkali-activator chemistry on 1) geopolymer chemistry, phase assemblage and nanostructure, 2) chemical binding mechanism of Sr2+ into the aluminosilicate framework of (N,K)-A-S-H gels in geopolymers, and 3) mass transport of Sr2+ during leaching, using high-field solid-state nuclear magnetic resonance spectroscopy and synchrotron-based X-ray absorption spectroscopy measurements. All geopolymers studied comprise a fully polymerised, X-ray amorphous Al-rich (N,K)-A-S-H type gel. Si exists predominantly in tetrahedral Q4(4Al) and Q4(3Al) sites and Al exists in tetrahedral sites, resulting in a net negative charge that is balanced by Na+ and/or K+ in extra-framework sites. Sr2+ was incorporated into extra-framework sites within (N,K)-A-S-H gels, without altering the local structure of the aluminosilicate framework by directly substituting for both Na+ and K+ in charge-balancing sites to form a (N,K,Sr)-A-S-H gel, at loadings equal to or below Sr/Na = 0.005. Above this limit, SrCO3 is formed, and the geopolymers simultaneously chemically bind Sr within a (N,Sr,K)-A-S-H gel, and physically encapsulate excess Sr as SrCO3. These findings have significant implications for use of geopolymers as materials for encapsulation and/or immobilisation of radioactive waste containing 90Sr.

Open Access

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Abstract: The Sierra Peña Blanca (SPB) region in Chihuahua, Mexico contains a significant uranium deposit representing about 40% of the country’s reserves. Common uranium minerals in this area include uranophane, schoepite, and weeksite/boltwoodite, with several superficial occurrences. Mining activities in the 1980s left unprocessed uranium ore exposed to weathering, with potential transport towards Laguna del Cuervo. This study presents an experimental simulation of uranium transport in SPB sediments using three approaches: (i) a batch experiment to evaluate the ideal adsorption of (UO2)2+ by fine sediment; (ii) a column system fed with 569 mgU L−1 UO2(NO3)2 to simulate adsorption by different sediment particle sizes; (iii) a column system with an upper horizon of uranophane from the area, fed with deionized water, to simulate uranium weathering and transport in particulate material, determined by liquid scintillation counting, revealed that the clay fraction had the highest adsorption capacity for U. X-ray Absorption Fine Structure (XAFS) analysis at the U L3 edge confirmed the U(IV) oxidation state and the fittings of the extended XAFS spectra confirmed the presence of the uranophane group of minerals. X-ray tomography further corroborated the distribution of particulate minerals along the column. The results suggest that the primary transport mechanism in SPB involves the fragmentation of uranium minerals, accompanied by eventual dissolution and subsequent adsorption of U onto sediments.