Nuclear fuel reprocessing and waste treatment facilities generate significant quantities of off-gas, which contain volatile radioactive and hazardous elements and compounds that must be captured and safely disposed of. To do that, an efficient and integrated off-gas treatment system is required to meet stringent regulatory requirements for operation, monitoring, and emissions control. The specific design and configuration of this system vary depending on the industry and process. However, a common theme is the utilization of solid sorbent materials to efficiently remove contaminants from various gas streams. There are a large number of sorbents at various stages of development that are being investigated and studied to capture mercury and iodine. The downside is that most of them were not tested under relevant process conditions. This presentation will review available sorbents for iodine and mercury against criteria for deployment in off-gas systems, addressing their performance in different environments and possible disposition pathways. Also included will be a discussion of examples of off-gas system designs and flow sheets from nuclear reprocessing facilities and the Hanford Vitrification Plant.

The deep geological repository concept for radioactive wastes is based on the confinement of the radioactivity over long periods of time by a multiple barrier system. Cementitious materials are used as part of the barriers in most of the repository concepts developed internationally (e.g., as backfill, tunnel lining, or in shaft seals and plugs). Although the behaviour of safety-relevant radionuclides in cementitious environments has been investigated extensively in the last decades, the impact of organic degradation products, originating from organic waste components or from superplasticisers in cementitious materials, on the migration of radionuclides under highly-alkaline, cementitious conditions is not yet fully understood. Therefore, the objective of this work, carried out within the framework of EURAD WP CORI (Cement-Organic-Radionuclide Interaction), was to fill knowledge gaps in the understanding of the impacts of the presence of phthalate (C₈H₄O₄²⁻; degradation product from plasticisers in PVC) and tri-methyl-amine (TMA; degradation product of ion exchange resins) on the migration behaviour of ²⁴¹Am and ¹⁵²Eu in cementitious barriers.

In this context, hardened cement pastes (HCP) were prepared with a water/cement ratio of 0.40 from a composite cement (CEM V/A 42.5N; Calcia, Rombas). The uptake and diffusion of ²⁴¹Am and ¹⁵²Eu in HCP was studied under anoxic conditions in the presence and absence of organics. In the absence of organics, a strong retention of both radionuclides on HPC was observed (R_d values between 10⁵ and 10⁶ dm³ kg^-1). In contrast, at phthalate concentrations exceeding ~10^-3 M, a reduction in the uptake of ²⁴¹Am and ¹⁵²Eu on HCP by several orders of magnitude was observed. This reduction in sorption could be the consequence of the decalcification of calcium silicate hydrates (C-S-H), the main sorbing phase in cementitious materials, due to the increasing formation of Ca-phthalate complexes in solution. These results indicate an increase in the mobility and diffusion of ²⁴¹Am and ¹⁵²Eu in cementitious barriers with increasing phthalate concentrations.

Acknowledgements

The EURAD-CORI project leading to this application has received funding from the European Union’s Horizon 2020 research and innovation programme under grant agreement No 847593.

To dispose of radioactive waste in the radioactive waste repository, the radioactive waste should satisfy the waste acceptance criteria of the repository which differ according to the site of the repository. Among the waste acceptance criteria, a leaching rate of the radionuclides in the waste is one of the main criteria which is directly related to the isolation of the radionuclides from the biosphere. However, the leaching rate of the radionuclides varies followed by the test conditions of the leaching test.

According to the chemical environment of the leachant, the chemical form of the radionuclides varies from precipitate to ion. For instance, cobalt exists as a cobalt ion in the H₂O system with a pH lower than about 9 while cobalt exists as cobalt hydroxide when the pH of the leachant is higher than 9. In addition, the adsorption of the nuclides differs followed by the nuclide concentration which affects the leaching rate. However, several studies prepared waste specimens with high concentrations compared to low-level waste to induce the measurable concentration of the leached nuclides. Therefore, the leaching behavior of the nuclides according to the test condition should be compared to avoid both over- and underestimation of the leaching rate.

In this study, the leaching behavior of Cs, Sr, and Co under several leachant types and concentrations is estimated. The cement-solidified specimens containing single Cs, Sr, and Co were manufactured. The leaching test following ANS 16.1 was performed by applying deionized water and cement-saturated groundwater. As a result, a leachability index difference according to the leachant type and nuclide concentration was discussed. The result of this study is expected to be background data that helps understand the actual leaching behavior of the Cs, Sr, and Co in the low- and intermediate level waste repository.

The management of nuclear waste is a major concern for the nuclear industry and society as a whole. Liquid nuclear waste requires special attention due to its potential for environmental contamination and the long half-life of the most common nuclides, such as Cs-137, Sr-90, and Eu-152. Several studies have addressed the use of Alkali Activated Materials (AAMs) for the immobilization of radioactive waste containing the before-mentioned nuclides. These studies have shown that AAMs can effectively immobilize these elements by forming stable phases that incorporate them into the structure of the material. The incorporation of these elements into the AAMs reduces their release and enhances their long-term stability, making them suitable for long-term storage. However, their integration into AAMs can also affect the properties of the encapsulation matrix. It is essential to understand the effect of these radionuclides on the properties of AAMs to ensure that the resulting material meets the necessary criteria for long-term storage. Furthermore, when compared to conventional water technologies, nanomaterials show great promise in removing heavy metals and radioactive ions from water because of their capacity to integrate different properties, creating multifunctional systems. In particular, CeO₂ nanoparticles have proven to be effective free-radical scavengers, providing defense against chemical, biological, and radiological abuse. In this study, inorganic polymers (IP) of different structural compositions were synthesized, and doped with different combinations of CsNO₃, Sr(NO₃)₂, Eu(NO₃)₃, and CeO₂ nanoparticles. The IP samples were developed from copper slag and a sodium silicate solution. Samples were tested on their microstructural (Scanning Electron Microscopy, Energy-Dispersive X-ray Spectroscopy) as well as their physicochemical (X-ray Fluorescence, Calorimetry, Iron oxidation state) properties in order to assess the influence the dopants have on the alkali-activated structures. Furthermore, the ability of IPs to retain the contaminants was tested with an up-flow percolation test.

Calcium silicate hydrate (C-S-H) is formed as a secondary mineral under the condition saturated with groundwater around radioactive waste disposal sites. The C-S-H is also a main component of cementitious materials and significantly adsorbs cationic radionuclides. However, it is considered that if the structure of C-S-H is altered, for example, by containing Mg from groundwater or host rock, the sorption characteristics for radionuclides may also be changed. Thus, in this study, the sorption behavior of Cs, including Cs-134 and C-137 in high-level radioactive wastes, to Mg-containing C-S-H is experimentally evaluated as a fundamental study.

The Mg-containing C-S-H was synthesized with a (Ca+Mg)/Si molar ratio of 0.4 – 1.6 and Mg content of 0 – 20% to Ca amount, by mixing CaO, SiO₂, Mg(NO₃)₂, NaOH to adjust pH, and ultra-pure water in given amounts. The sorption experiment was carried out by simultaneously adding CsCl solution to be 1.0 mM to synthesize the Mg-containing C-S-H without any drying process. The liquid/solid weight ratio was 20 mL/g, and the total volume of the solution was 30 mL. The curing was 7 and 42 days at 298 K with shaking at 120 strokes/min. As the results, the sorption ratio slightly decreased with increasing the (Ca+Mg)/Si molar ratios. Furthermore, the Raman spectra suggested that the incorporation of Mg into the C-S-H structure decreases the sorption site by facilitating the polymerization of the silicate chain. However, high sorption distribution coefficients of K_d= 5.5 – 10.2 mL/g were estimated in (Ca+Mg)/Si=0.8 with Mg-content up to 20%, as an example of secondary mineral. Moreover, the K_d for all samples exceeded the K_d of 0.04 – 0.4 mL/g for the plutonic rocks. This suggests that C-S-H contributes to the immobilization of Cs without decreasing its sorption performance, even if C-S-H incorporates Mg into its structure.

Bentonite has strongly desired properties for its use in an engineered barrier system – swelling and sealing capabilities, high sorption capacity, for containment and sorption of migrating radionuclides, resulting in low permeability. A geological transformation of smectite (the main component of bentonite) to illite, also known as illitization, has been widely studied and occurs with high temperatures, pressures, and an external K⁺ source. However, this is detrimental to the barrier system as the barrier loses its most critical properties and can easily transport radionuclides to the environments in the far field. We have performed extensive research at Sandia National Laboratories to determine what physical and chemical conditions result in the formation of mixed-layer illite/smectite and complete transformation to illite. Our Parr Vessel reaction studies entail the use of different cation exchanged smectite clays (Na⁺, Cs⁺, K⁺), high temperature (200 °C), various reactor solutions, times ranging from 7 to 112 days and liquid to solid ratios of 100, 500 and 1000. The solid reaction products were analyzed with XRD – air dried and ethylene glycolated mounts – and SEM-EDS. The clay recovered from the reactors was also Na-exchanged to determine any K⁺ fixation within the clay interlayer. Analysis from ICP-OES was also used to characterize the liquid chemistry from the hydrothermal reactions. The data recovered reveal illitization of smectite occurring in as little as 28 days, regardless of Na⁺ or K⁺ cations initially in the interlayer. Preliminary results indicate Cs⁺ prolongs the transformation of smectite to illite, forming mixed-layer illite/smectite after 28 days. Understanding the mechanism of illitization will further help to inform the performance assessment in the design of an engineered barrier system.

SNL is managed and operated by NTESS under DOE NNSA contract DE-NA0003525. SAND2023-03806A

In a zero power reactor (ZPR) loaded with salt fuel, the thermal energy released during operation is so small that the fuel remains solid at room temperature with very low burnup and heat rate. Nuclear power plant (NPP) spent fuel is currently reprocessed and recycled at end-of-life to preserve resources and for the reduction of future burden from wastes. Actinides maybe recycled using advanced processes of separation through various routes; PUREX or pyro-processing. The considered zero power salt reactor is in the form of salt fast reactor with high energy neutron flux during operation. Density is one of the critical thermo-physical properties of any reactor. Determining the density of the salt system (NaCl-UCl₄) is very important to evaluate salt fuel-reactivity and behaviour of the core. Actinides and fission products inventories at the end-of-life of reactor are then significant. Presenting the analytical methods of measuring the densities of salt components using multi-scale approaches of X-ray Diffraction (XRD) for nm features, amorphization ratio or any defects, and Scanning Electron Microscopy (SEM) for µm pores in the salt fuel. The emphasis is on a salt mixture with composition of (NaCl-47mol%+UCl₄-53mol%). Densities were measured by changing compositions along with the identification of the complex phase; Na₂UCl₆. Results obtained were in good agreement with the ideal mixed phase (heterogeneous) density model, thereby establishing that XRD and SEM are important techniques to measure the densities of salt fuels. High density fuel in a reactor enhances the reactivity as well as the average neutronic flux. This work provides the salt density measurement which can be used to correct the reactivity of the fuel at end-of-life and for other utilisations. Actinides (Pu, MA) and fission products inventories at the end-of-life are then insignificant in fraction. In these conditions fuel material may be seen as a zero nuclear waste.