Development, characterization and dissolution behavior of calcium-aluminoborate glass wasteforms to immobilize rare-earth oxides

Calcium-aluminoborate (CAB) glasses were developed to sequester new waste compositions made of several rare-earth oxides generated from the pyrochemical reprocessing of spent nuclear fuel. Several important wasteform properties such as waste loading, processability and chemical durability were evaluated. The maximum waste loading of the CAB compositions was determined to be ~56.8 wt%. Viscosity and the electrical conductivity of the CAB melt at 1300 °C were 7.817 Pa·s and 0.4603 S/cm, respectively, which satisfies the conditions for commercial cold-crucible induction melting (CCIM) process. Addition of rare-earth oxides to CAB glasses resulted in dramatic decreases in the elemental releases of B and Ca in aqueous dissolution experiments. Normalized elemental releases from product consistency standard chemical durability test were <3.62·10-5 g·m-2for Nd, 0.009 g·m-2for Al, 0.067 g·m-2for B and 0.073 g·m-2for Ca (at 90, after 7 days, for SA/V = 2000m-1); all meet European and US regulation limits. After 20 d of dissolution, a hydrated alteration layer of ~ 200-nm-thick, Ca-depleted and Nd-rich, was formed at the surface of CAB glasses with 20 mol% Nd2O3whereas boehmite [AlO(OH)] secondary crystalline phases were formed in pure CAB glass that contained no Nd2O3

Reactive spark plasma synthesis of CaZrTi2O7 zirconolite ceramics for plutonium disposition

Near single phase zirconolite ceramics, prototypically CaZrTi 2 O 7 , were fabricated by reactive spark plasma sintering (RSPS), from commercially available CaTiO 3 , ZrO 2 and TiO 2 reagents, after processing at 1200 °C for only 1 h. Ceramics were of theoretical density and formed with a controlled mean grain size of 1.9 ± 0.6 μm. The reducing conditions of RSPS afforded the presence of paramagnetic Ti 3+ , as demonstrated by EPR spectroscopy. Overall, this study demonstrates the potential for RSPS to be a disruptive technology for disposition of surplus separated plutonium stockpiles in ceramic wasteforms, given its inherent advantage of near net shape products and rapid throughput

The formation of pitted features on the international simple glass during dynamic experiments at alkaline pH

The forward rate of dissolution of the International Simple Glass (ISG) was determined under alkaline conditions at 40 °C using the Single Pass Flow Through (SPFT) method. Forward rates were consistent with those obtained in the literature for this glass composition. The formation of altered gel layers and surface pits was observed on the surface of glass particles, especially at the very highest pH values, despite the application of high flow rates to prevent the build-up of solubility limiting phases. These features could be attributed to preferential localized dissolution at sites with a higher alkali concentration or from a separate, less durable, vitreous phase. These results may indicate that surface pit and altered gel formation occurs under the forward rate of dissolution as imposed by the SPFT method, particularly for simplified borosilicate glass materials

The HADES facility for high activity decommissioning engineering & science: part of the UK national nuclear user facility

Research and innovation is key to delivering UK Government's civil nuclear energy policy, in particular to accelerate reduction in the hazard, timescale and cost of legacy decommissioning and geological disposal of radioactive wastes. To address this challenge, a national centre of excellence, the HADES Facility, has been established to support research and innovation in High Activity Decommissioning Engineering & Science, as part of the wider network of UK National Nuclear User Facilities. Herein, we describe the development of this user facility, the current status of its capability, and functional equipment specifications. The unique capabilities of the HADES Facility, in the UK academic landscape, are emphasised, including: handling of weighable quantities of 99Tc and transuranics; quantitative electron probe microanalysis of radioactive materials; hot isostatic pressing of radioactive materials; and laboratory-based X-ray absorption and emission spectroscopy. An example case study of the application of the HADES capability is described, involving thermal treatment of a real radioactive ion exchange resin waste to produce a conceptual vitrified waste form

A review of zirconolite solid solution regimes for plutonium and candidate neutron absorbing additives

Should the decision be made to immobilise the UK Pu inventory through a campaign of Hot Isostatic Pressing (HIP) in a zirconolite matrix, prior to placement in a geological disposal facility (GDF), a suite of disposability criteria must be satisfied. A GDF safety case should be able to demonstrate that post-closure criticality is not a significant concern by demonstrating that such an event would have a low likelihood of occurring and low consequence if it were to occur. In the case of ceramic wasteforms, an effective means of criticality control may be the co-incorporation of a requisite quantity of a suitable neutron absorbing additive, either through co-immobilisation within the host structure or the encapsulation of discrete particles within the grain structure. Following an initial screening of a range of potential neutron absorbing additives, a literature-based assessment of the solid solution limits of a number of potential additives (Gd, Hf, Sm, In, Cd, B) in the candidate zirconolite (CaZrTi2O7) wasteform is presented. Key areas of research that are in need of development to further support the safety case for nuclearised HIP for Pu inventories are discussed

Characterisation of a high pH cement backfill for the geological disposal of nuclear waste: The Nirex Reference Vault Backfill

In a conceptual UK geological disposal facility for nuclear waste within a high-strength, crystalline geology, a cement-based backfill material, known as Nirex Reference Vault Backfill (NRVB), will be used to provide a chemical barrier to radionuclide release. The NRVB is required to have specific properties to fulfil the operational requirements of the geological disposal facility (GDF); these are dependent on the chemical and physical properties of the cement constituent materials and also on the water content. With the passage of time, the raw materials eventually used to synthesise the backfill may not be the same as those used to formulate it. As such, there is a requirement to understand how NRVB performance may be affected by a change in raw material supply. In this paper, we present a review of the current knowledge of NRVB and results from a detailed characterisation of this material, comparing the differences in performance of the final product when different raw materials are used. Results showed that minor differences in the particle size, surface area and chemical composition of the raw material had an effect on the workability, compressive strength, the rate of hydration and the porosity, which may influence some of the design functions of NRVB. This study outlines the requirement to fully characterise cement backfill raw materials prior to use in a geological disposal facility and supports ongoing assessment of long-term post-closure safety

Synthesis of simulant ‘lava-like’ fuel containing materials (LFCM) from the Chernobyl reactor Unit 4 meltdown

A preliminary investigation of the synthesis and characterization of simulant ‘lava-like’ fuel containing materials (LFCM), as low activity analogues of LFCM produced by the melt down of Chernobyl Unit 4. Simulant materials were synthesized by melting batched reagents in a tube furnace at 1500 °C, under reducing atmosphere with controlled cooling to room temperature, to simulate conditions of lava formation. Characterization using XRD and SEM-EDX identified several crystalline phases including ZrO2, UOx and solid solutions with spherical metal particles encapsulated by a glassy matrix. The UOX and ZrO2 phase morphology was very diverse comprising of fused crystals to dendritic crystallites from the crystallization of uranium initially dissolved in the glass phase. This project aims to develop simulant LFCM to assess the durability of Chernobyl lavas and to determine the rate of dissolution, behavior and evolution of these materials under shelter conditions

Advanced gas-cooled reactor SIMFuel fabricated by Hot Isostatic Pressing: a feasibility investigation

The manufacture of a simulant UK Advanced Gas Cooled Reactor (AGR) spent nuclear fuel (SIMFuel) was achieved by Hot Isostatic Pressing (HIP). Characterisation of HIP AGR SIMFuels, tailored to burn ups of 25 GWd/t U and 43 GWd/t U (after 100 years cooling) demonstrated fission product partitioning, phase assemblage, microstructure and porosity in good agreement with spent nuclear fuels and SIMFuels, and AGR fuels in particular. A pivotal advantage of the application of the HIP manufacturing method is the retention of volatile fission products within the resultant SIMFuel as the result of using a hermetically-sealed container. This new approach to SIMFuel manufacture should enable the production of more accurate spent nuclear fuel surrogates to support research on spent fuel management, recycle, and disposal, and the thermal treatment of fuel residues and debris

Process development of zirconolite ceramics for Pu disposition: use of a CuO sintering aid

Zirconolite-structured ceramics are candidate wasteform materials for the immobilisation of separated Pu. Due to the refractory properties of zirconolite and other titanates, removing residual porosity remains challenging in the final wasteform product when utilising a conventional solid state sintering route. Herein, we demonstrate that the addition of CuO as a sintering aid increases densification and promotes grain growth. Moreover, zirconolite phase formation was enhanced at lower process temperatures than typically required (≥1350 °C). CuO addition allowed an equivalent density to be reached using process temperatures of 250 °C lower than the undoped composition. At 150 °C lower than the undoped zirconolite, the addition of CuO resulted in a favourable microstructure and phase assemblage, as confirmed via X-ray diffraction and scanning electron microscopy. Secondary phases of CaTiO3 and Ca0.25Cu0.75TiO3 were observed at some processing temperatures, which may prove deleterious to wasteform performance. The use of a CuO sintering aid provides an avenue for the further development of the thermal processing of ceramic wasteform materials

Fenton and Fenton-like wet oxidation for degradation and destruction of organic radioactive wastes

Fenton or Fenton-like oxidation for treatment of organic radioactive wastes is a promising technology with applications to a range of organic wastes. This review details this process; exploring potential challenges, pitfalls and opportunities for industrial usage with radioactive wastes. The application of this process to real radioactive wastes within pilot-plant settings has been documented, with key findings critically assessed in the context of future waste production. Although this oxidation process has not found mainstream success in treatment of radioactive wastes, a lower temperature oxidation system bring certain benefits, specifically for higher volume or problematic organic wastestreams