The global expansion of nuclear energy will generate increasing quantities of waste with low levels of plutonium or other nuclear materials (NM) potentially subject to International Atomic Energy Agency (IAEA) safeguards. Reducing requirements on retained wastes has the potential to reduce future demands on already strained IAEA resources. We describe an effort to help the IAEA and Member States better estimate projected waste loads and associated safeguards obligations by developing a reporting tool to estimate types and sizes of future waste-storage and -disposal facilities. States can use such information to plan waste facilities, including size and type. The IAEA can use these data for inclusion in multiple agency reports and products for the benefit of Member States.

Leaching results to compare the dissolution behavior of a new type of fuel with additives (Advanced Doped Pellet Technology, ADOPT) with standard UO2 fuel are presented. Both fuels were irradiated in the same assembly of a commercial boiling water reactor to a local burnup of ∼58 MWd/kgU. Fuel fragments are leached in simplified groundwater in two autoclaves under hydrogen atmosphere, representing conditions in a canister failure scenario resulting in water intrusion for a spent nuclear fuel repository. Preliminary results indicate the uranium concentration decreased to 3-4x10-8 M after 421 days, slightly above the solubility of amorphous UO2. Xe has been detected in the gas phase of both autoclaves. The concentration of Cs and I seems to gradually approach constant values, yet the redox sensitive elements continue to slowly increase with time. The preliminary data obtained supports the hypothesis that there is no major difference in leaching behavior between the two fuels.

In a deep geological repository (DGR) scenario, uranium oxidized in aqueous systems will be stabilized as UO22+ (hexavalent uranium), as a consequence of tetravalent uranium oxidation by radiolytic byproducts. Uranyl cationic species (UO22+) in different speciation forms are expected to be found at the whole pH range conditions. The importance of UO22+ lies in its potential incorporation of trace radioelements onto secondary uranyl phases. In view of the difficulty of U chemistry in natural groundwater, it is necessary to improve speciation assessment techniques so as to understand chemical processes. Raman spectroscopy has been shown as a powerful tool to analyze the speciation of various actinyl (UO22+,NpO2+ and PuO22+) and to determine the distribution of those elements which are more likely to be stable in a near-field groundwater environment. Therefore, the aim of this work is to follow UO22+ changes as a consequence of γ radiation in aqueous media under DGR conditions, and to understand the behavior of UO22+ as a function of aqueous media, which help to understand and predict the potential precipitation of the solid phases formed. In this work, the use of Raman spectroscopy adapted to the empirical analysis of different nuclear applications for initial uranium concentrations 0.04M at ambient atmosphere is shown, i.e. as monitoring tool for UO22+ precipitation as a function of pH, studying UO2(NO3)2·6H2O stability in aqueous solutions representative of groundwater, in particular at ionic strength I = 0.02 – 0.4 M and pH from 7 to 13.2; and to evaluate the effect of γ radiation fields. At 10−4-10-3 M of radiolytically formed H2O2 concentration, the amount of uranium in solution decreased, as a results of secondary phases precipitation. The results obtained will be useful to the performance assessment studies of the Spent Nuclear Fuel (SNF) stored in DGRs. The work performed provides a partial picture of secondary phase formations, as a result of corrosion of SNF in a DGR.

A 1D reactive transport model has been implemented in iCP (interface COMSOL Multiphysics and PhreeqC) to assess the corrosion of Spent Fuel (SF), considered as homogeneous UO2(am,hyd) doped with Pd. The model couples: i) generation of water radiolysis species by alpha and beta radiation considering the complete water radiolysis system with the kinetic reactions involving: H+, OH-, O2, H2O2, H2, HO2-, HO2·, O·, O-, O2-, H·, ·OH and e- ii) processes occurring in the spent fuel surface: oxidative dissolution reactions of UO2(am,hyd) and subsequent reduction of oxidized fuel, considering H2 activation by Pd, and iii) corrosion of Fe(s) in oxic and anoxic conditions. Process i) has been implemented in COMSOL and processes ii) and iii) have been implemented in PHREEQC with their kinetic constants being calibrated with different sets of experimental data published in the open literature. The model yields a UO2(am,hyd) dissolution rates similar to the values selected in safety assessments.

A range of stoichiometric and mixed A-site cation brannerite glass-ceramics have been synthesised and characterised. The formation of UTi2O6 in glass is reliant on ensuring all uranium remains tetravalent by processing in an inert atmosphere. ThTi2O6 forms in glass under both inert and oxidising atmospheres due to the lack of other easily available oxidation states. CeTi2O6 could not be made to form within this glass system. The formation of A0.5B0.5Ti2O6 phases depends strongly on the oxidation states of the A and B cations available in the process atmosphere, with the most successful compositions having an average final oxidation state of (A,B)4+. Mixed cation brannerite compositions that formed in argon include U0.75Th0.25Ti2O6 and U0.71Ce0.29Ti2O6. Those forming in air include U0.23Th0.77Ti2O6, Th0.37Ce0.63Ti2O6, and U0.41Ce0.59Ti2O6.

The oxygen ions in the high temperature superionic state of uranium dioxide (UO2) are known to be in an arrested or jammed state, exhibiting characteristic features of jammed kinetics such as low dimensional string-like ion hopping and dynamical heterogeneity (DH). This thermally-jammed state entails a configurational entropic cost. Using atomistic simulations and the 2PT method, we compute the solid-like (vibrational) and hard sphere-like (configurational) contributions to the total entropy across a temperature range of 1500 K to 2800 K that envelop both the onset of superionic conduction (2000 K) and the second order λ-transition (2610 K). To properly account for the thermally jammed state of the ions, we use an equation of state that is appropriate for the metastable fluid branch. Our simulation results are in excellent agreement with the entropy data extracted from specific heat experiments with a mean error of less than 2%.

Cladding tubes of water-cooled nuclear reactors are usually made of Zircaloy and are an important retaining element for radionuclides present in the fuel both during predisposal activities such as reloading of fuel assemblies from interim storage casks to final disposal casks and during final disposal in the case of canister breaching. However, cladding integrity is affected by various processes during reactor operation and beyond, e.g. fuel cladding chemical interaction and fission product precipitation onto the inner cladding surface. Using experimental and modelling methods, the radionuclide inventory of an irradiated Zircaloy-4 plenum section is analyzed. Quantities of 235/238U, 237Np, 238/239/240/241/242Pu, 241/243Am, 243/244Cm besides 14C, 55Fe, 125Sb, 154Eu, and 134/137Cs were (radio-)chemically determined in digested Zircaloy-4 subsamples. Measured inventories of activation products in the Zr-alloy are in good agreement with calculated values. However, amounts of actinides and fission products exceed the calculated inventory by factor ∼57 (minor actinides and non-volatile fission products) and ∼114 (137Cs). Excess of minor actinides and part of enhanced Cs inventory originate from fuel residues deposited on the inner cladding surface during fuel rod fabrication, whereas vast amount of cesium is volatilized from subjacent fuel pellets and transported to the plenum.

A good understanding of the spent fuel matrix (UO2) behavior under predisposal activities conditions is required for the proper performance assessment of a final repository. Hence, the oxidation evolution of UO2 under dry interim storage conditions, as a main predisposal action within the Spanish strategy, needs to be addressed. For this aim, in this work a detailed in situ Raman spectroscopy study of the surface oxidation of a UO2.00 disk heated in the presence of synthetic air at 573 K is presented. The spectra analysis required two previous studies. In the first one, UO2+x powder samples with controlled degree of non-stoichiometry were identified by thermogravimetric analysis and subsequently characterized by Raman spectroscopy. The equations obtained from this study enable estimating the oxidation degree of any UO2+x sample (for x < 0.20) at atmospheric conditions. The second one was performed in order to use these equations for the in situ experiments (at 573 K), since the shift of the bands due to temperature needs to be taken into account. Thus, the behavior of the Raman spectra as a function of temperature was analyzed and a correction term thereafter introduced in the initial equations.

Nuclear power is a crucial source of clean energy for India. In the near-term, India is focusing on thermal reactors using natural and enriched uranium. In the long-term, India is exploring various options to use its large thorium reserves.

India’s present nuclear installed capacity is 5680 MW, which contributes to about 3.4% of the annual electricity generation. However, nuclear power is an important source of energy in India’s aspirations for energy security and also in achieving its Intended Nationally Determined Contributions (INDC), of 40% fossil free electricity, by 2030. India has limited uranium reserves, but abundant thorium reserves. The Nuclear Suppliers Group (NSG) lifted restrictions on trade with India, in 2008, enabling India to import uranium (natural and enriched) and nuclear reactors. In the near–term (2030), the nuclear capacity could increase to about 42,000 MW. This would be from a combination of domestic Pressurized Heavy Water Reactors (PHWR) and imported Pressurized Water Reactors (PWR). For the long–term (2050), India is exploring various options for utilising its vast thorium reserves. This includes Advanced Heavy Water Reactor and Molten Salt Breeder Reactor. However, generating public acceptance will be crucial to the expansion of the nuclear power program.

Due to their high actinide content MOX residues require immobilization within a robust host matrix. Although it is possible to immobilize actinides in vitreous wasteforms; ceramic phases, such as brannerite (UTi2O6), are attractive due to their high waste loading capacity and relative insolubility. Brannerites Gd0.1U0.9Ti2O6, Ce0.1U0.9Ti2O6 and Gd0.1U0.81Ce0.09Ti2O6 were prepared using an oxide route. Charge compensation of trivalent cations was expected to occur via the oxidation of U (IV) to higher valence states (U (V) or U (VI)). Gd was added to act as a neutron absorber in the final Pu bearing wasteform and Ce was used as a structural surrogate for Pu. X-ray absorption spectroscopy showed that Ce (IV) was reduced to Ce (III) in all cases. X-ray powder diffraction of synthesized specimens found that the final phase assemblage was strongly affected by processing atmosphere (air or argon). Prototypical brannerite was formed in all compositions, secondary phases observed were found to vary according to processing atmosphere and stoichiometry. Microstructural analysis (SEM) of the sintered samples confirmed the results of the X-ray powder diffraction.

Large stores of unstable waste uranic materials such as fluorides or nitrates exist internationally due to legacy civil nuclear enrichment activities. Conversion of these uranic materials to layered metal uranates prior to disposal is possible via aqueous quench - precipitation type reactions. Previous studies1 have shown facile in-situ formation of geologically persistent and labile uranate colloids2 under simulated nuclear waste repository conditions, though the effects of local solution metal-uranium ratios on uranate stoichiometry have yet to be covered. This affects our understanding of how key radionuclides present in repository porewaters such as strontium or caesium may be sequestered in these uranate structures. In this work, we demonstrate a synthesis reaction for calcium monouranate particles via rapid anhydrous curing of a sol-gel. We present some results showing aqueous nucleation of uranate nanoparticles and their phase transformations during thermal curing as well as the effects of solution phase calcium loading on uranate phase purity in the cured particles.

Helium is produced in spent nuclear fuel by α-decays of actinides. After 10,000 years, the concentration of He accumulated in UO2 spent fuel is about 0.23 at.%. For direct disposal of spent nuclear fuel, consequences of helium build-up on the fuel matrix microstructure must be evaluated since it can modify the radionuclide release when water comes into contact with the spent fuel surface, after breaching of the disposal canister. An operational model has been proposed in order to evaluate the effect of helium on the microstructure of spent fuel in a repository. Based on conservative assumptions and different scenarios of bubble population, the calculated helium critical concentration, that could lead to a partial loss of integrity of the spent fuel pellet, is 0.37 at.%. However, observations on He-implanted UO2, α-doped UO2 pellets and natural analogues evidence a macroscopic damage only for He concentrations, which are more than one order of magnitude higher.

Several diverse types of UH3-based hydrides can be prepared by hydrogenation of bcc U-based alloys. Pair Distribution Function (PDF) analysis using high-energy X-rays identified that the (UH3)1-x Mox hydrides are nanocrystalline, with the structure motif based mainly on the β-UH3 structure. α-UH3 represents a minority component. On the other hand, PDF of the (UH3)1-x Zrx hydrides corresponds well to the α-UH3 crystal structure. All the hydrides are ferromagnetic, with the Curie temperature T C reaching up to 203 K.

Raman spectroscopy studies have been performed on one hand to identify different materials related to spent nuclear fuel (SNF), and on the other hand to study the behavior of SNF at different storage conditions. Specifically, the expected oxidation of the SNF matrix under dry storage conditions and the formation of secondary phases (SP), as a result of corrosion of SNF in a deep geological repository, have been studied. In order to perform these experiments, two protocols based on the Raman spectroscopy technique have been developed. The results show U4O9/U3O7 and U3O8 as oxidation products of UO2 powder at high temperatures in air, and the secondary phase formation (rutherfordine, UO2(CO3), soddyite, (UO2)2SiO4•2H2O, uranophane alpha Ca(UO2)2(SiO3OH)2•5H2O and kasolite, PbUO2SiO4•H2O), due to uraninite corrosion under the conditions of Sierra Albarrana (Spain).

In the framework of the Euratom FIRST-Nuclides program, SCK•CEN conducted leaching experiments on a PWR UOX fuel, taken from the Belgian Tihange 1 reactor, with an average rod burnup of 50 MWd/kgHM in order to determine the rapid release of some of the most critical radionuclides. Experiments were conducted with an intact and a declad spent fuel segment in contact with a reference carbonate solution. During one year, samplings were performed to determine the release of the most relevant radionuclides. The samples taken after six months of leaching were analysed by means of Thermal Ionisation Mass Spectrometry (TIMS). These results were compared to the total inventory measured with radiochemical burnup analyses, as well as with the calculated inventory profile along the pellet radius. The comparison gives information about the origin of the dissolved uranium.

Under high burnup UO2 fuel pellets can experience high burnup structure (HBS) at the rim also known as rim effect. The HBS is exceptionally porous with fine grain sizes. HBS increases the swelling further than it would have achieved at a larger grain size. A theoretical swelling model is used in conjunction with a grain subdivision simulation to calculate the swelling of UO2. In UO2 the nucleation sites are at vacancies and the bubbles are concentrated at grain boundaries. Vacancies are created due to irradiation and gas diffusion is dependent on vacancy migration. In addition to intragranular bubbles, there are intergranular bubbles at the grain boundaries. Over time as intragranular bubbles and gas atoms accumulate on the grain boundaries, the intergranular bubbles grow and cover the grain faces. Eventually they grow into voids and interconnect along the grain boundaries, which can lead to fission gas release when the interconnection reaches the surface. This is known as the saturation point. While the swelling model used does not originally incorporate a changing grain size, the simulation allows for more accurate swelling calculations by introducing a fractional HBS based on the temperature and burnup of the pellet. The fractional HBS is introduced with a varying grain size. Our simulations determine the level of swelling and saturation as a function of burnup by combining an independent model and simulation to obtain a more comprehensive model.

Metallographical examination of Uranium alloys can benefit from electron back scatter diffraction (EBSD) technique. Various methods of surface preparation for microstructural characterization are described and compared. The aim of the study was to optimize the preparation technique for surfaces of U alloy splats for EBSD mapping, particularly in the context of U1-x Mox alloys, as properties of γ-U surfaces (e.g. with more Mo) are very different than for mostly α-U type (low-alloyed U).

The combination of density functional theory in local density approximation and dynamical mean field theory (LDA+DMFT) was employed in a preliminary study of the strong electron correlation effects in a promising nuclear fuel—uranium mononitride (UN). For the ferromagnetic phase, the effective impurity problem arising in the LDA+DMFT [1-3] cycle is solved with the spin-polarized T-matrix fluctuation exchange (SPTF) solver, which includes spin–orbit interactions. Concerning the paramagnetic phase, the disordered local moment (DLM) approach was used, based on both standard local density approximation (LDA) and LDA+U. Basic spectral properties and material properties, such as the spin, orbital and total magnetic moments on U atom were calculated for various values of the Hubbard parameter U with a fixed exchange parameter J. Our main focus was to compare the calculated spectral functions (density of states) for different magnetic phases and different methods to the experimental XPS data [4]. On top of that, the total moments of the paramagnetic and ferromagnetic phases are compared with the measured values by neutron spectroscopy [4, 5].

The formation of uranyl peroxide phases was identified as a corrosion product of spent fuel by Hanson et al [1]. The subsequent analysis of this phase showed that metastudtite retained 241Am, 237Np and 239Pu [2]. In this study, the retention of radionuclide Pu4+ and An3+, released from the spent fuel matrix into studtite structure, has been evaluated by the precipitation of studtite from uranyl dissolution with variable concentrations of REE (Th, Nd, Sm and Eu). Three different precipitation conditions parameters were studied: media of synthesis, time of synthesis and REE concentration. Synthesized phases were characterized by XRD and the cell parameter was calculated. The REE incorporation was determined by ICP-MS analysis. The results showed that studtite could incorporate 63% of Th in solution during its precipitation. Changes in the “a” cell parameter were identified. The results suggest that studtite coprecipitated with REE could play a role as a limiting for the REE mobility.

Preparation and characterization of a Simulated Spent Nuclear Fuel (SIMFuel), which replicates the chemical state and microstructure of Spent Nuclear Fuel (SNF) discharged from UK Advanced Gas-cooled Reactor (AGR) after a cooling time of 100 years is described. Thirteen stable elements were added to depleted UO2 and sintered to simulate the composition of fuel pellets after burn-ups of 25 and 43 GWd/tU and, as a reference, pure UO2 pellets were also investigated. The fission product distribution was calculated using the Fispin code provided by NNL. SIMFuel pellets exhibit a microstructure up to 92% TD. During the sintering process in H2 atmosphere Mo-Ru-Rh-Pd metallic precipitates and grey-phase ((Ba, Sr)(Zr, RE)O3 oxide precipitates) formed within the UO2 matrix. These secondary phases are present in real PWR and AGR SNF, although they are smaller in size than those examined in this study. The grain size of the produced SIMFuel is in good agreement with literature references.