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Abstract: Geopolymer cements are highly promising materials for long-term immobilisation of Strontium-90 radioactive waste, offering superior durability and cation binding sites compared to conventional Portland cement matrices. This study investigates the influence of prolonged leaching on the Sr immobilisation mechanism and structural integrity of metakaolin-based geopolymers using the ANSI/ANS 16.1 semi-dynamic leaching test. All geopolymers demonstrated high Sr retention, with Leachability Indices at least 14.7 for all samples, significantly exceeding the industry guideline of 6.0, confirming their effectiveness. Importantly, potassium silicate–activated geopolymers exhibited reduced Sr release and substantially lower leaching rates than sodium silicate–activated geopolymers. Multiscale spectroscopic and diffractometric analysis, including synchrotron X-ray absorption spectroscopy and multinuclear high-field solid-state MAS NMR probing 39K, 23Na, 27Al, and 29Si, revealed that the alkali aluminosilicate gel framework remained structurally stable after leaching for 28 days, with no significant alterations to Si and Al bonding environments. Sr release is primarily controlled by diffusion, and the dominant immobilisation mechanism is the formation of insoluble SrCO3. Atomic-level Sr structural analysis using XANES/EXAFS revealed an increase in the average Sr coordination number in both systems after leaching, with a more pronounced rise in potassium-based geopolymers, consistent with enhanced SrCO3 formation. Overall, these findings demonstrate that geopolymers maintain structural integrity during leaching and show for the first time that using potassium rather than sodium as an alkali activator is definitively more advantageous for maximising the long-term effectiveness of geopolymer wasteforms. This demonstrates their strong suitability as wasteforms for the safe long-term immobilisation of Sr-containing radioactive wastes.

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Abstract: Selenium-79, a radionuclide present in higher-activity radioactive wastes destined for deep geological disposal, is mobile under oxic conditions, where Se(IV) and Se(VI) dominate. Anoxic batch microcosm incubations were constructed containing Wyoming MX80 bentonite (a candidate buffer material in geological disposal) and artificial groundwater with or without steel coupons to represent canister materials. Se(VI)(aq) bioreduced and was removed by 7 days when lactate was added as an electron donor, after which sulfate reduction occurred. With H2 gas as the electron donor, Se(VI) bioreduction slowed, with complete removal at 14 days and minimal sulfate reduction thereafter. 16S rRNA gene sequencing highlighted the dominance of Anaerobacillus spp. (44% at 28 days) during Se(VI)-reduction, and in the lactate-amended systems, there was a subsequent enrichment in sulfate-reducing bacteria affiliated with Desulfosporosinus spp. (60% relative abundance at 84 days). Extended X-ray absorption fine structure (EXAFS) analyses identified monoclinic Se(0) as the bioreduction product after 28 days, but by 84 days this evolved to trigonal Se(0) in the absence of steel coupons or was further reduced to FeSe2 with steel present. The reduction of Se(VI)(aq) to poorly soluble Se(0)/FeSe2 mediated by indigenous bentonite microbial communities highlights their potential importance in promoting Se-79 retention during deep geological disposal.

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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 <a>-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 <c>-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.