Scientific basis of the nuclear fuel cycle III

Resume : Due to the increasing fuel burnup and changing of its chemical composition, research on the radiation effects on TBP and its solutions in a hydrocarbon diluents is actual for avoiding technological accidents. To irradiate samples the linear accelerator UELV-10-10-C70 was used. Accelerated electron beam was scanned along the vertical axis of the 200 ml cylindrical glass cell (5.5 cm in diameter), equipped with a hydraulic lock and mounted on a rotating carousel table. 4 mole/l HNO3 was used as the hardening liquid. Radiolysis products were defined by IR spectroscopy with Fourier transform and gas chromatography. Distribution coefficient of plutonium is slightly dependent on the irradiation dose, falling in the investigated range equally under either electronic or alpha-irradiation, mainly due to the radiolytic decomposition of HNO3. Compensation of its losses on the next solvent saturation with acid returns distribution coefficient to its previous level. Radiation decomposition of TBP at doses up to 500 kGy occurs only by 8-10%, and has therefore miserable effect on the solvent capacity. The work was performed under financial support from Ministry of Science and Education of the Russian Federation under Agreement between the Ministry and A.N. Frumkin Institute of Physical chemistry and Electrochemistry RAS No. 14.604.21.0153 dated 28/11/2014.

Resume : The safe management of nuclear waste is a serious issue. A possible solution is a long-term preservation of the waste in stable matrices. Currently, several ceramic materials are considered for the immobilization of radionuclides. It requires good understanding of the chemical and physical properties of these materials. By using the state-of-the-art computational techniques and world-class supercomputing resources we contribute to this research with atomistic simulations by providing an atomic-sale insight into the mechanisms of radionuclide incorporation in these materials. Density Functional Theory (DFT) is often the only method of choice for ab initio simulation of chemically complex compounds. However, it often dramatically fails in describing materials containing strongly correlated f-electrons, leaving no other option but to use more resource-demanding and usually unfeasible computational methods (eg. hybrid functionals, CCSD(T), MP2). On the other hand, we found that by using a computationally efficient modification of the DFT method, DFT+U with the Hubbard U parameter derived ab initio with cLDA or cRPA methods, a correct and reliable prediction of various important structural and thermodynamical parameters of f-elements-bearing compounds becomes possible. In particular, we will show the results of computer-aided simulation of monazite and pyrochlore type compounds and solid solutions.

Resume : The thermal stability of extractants mixtures with the UN has been studied very superficially, and the beginning of uncontrolled exothermic reactions therein has not been studied at all. There is absolutely no information concerning effects of the extractant radiolysis and thermolysis products on the thermal stability of these compounds. Meanwhile, the thermal stability of extractant mixtures with nitric acid was found to be affected negatively with their radiolysis. By the transition to new types of fuel with large burn up, the radiation dose to the extractant encreases, which may lead to a considerable both vapor-air and condensed mixtures explosibility growth. The same trend can take place also for mixtures of extractant with uranyl nitrate (UN). TBP*UN extract heating in a closed vessel threw temperature interval of 150-160 °C hasn’t been found to be accompanied by gas and heat release, the appearance of the test specimens having been unchanged. Intensive exothermic reaction begins at temperatures from 165 to 170 °C. It is accompanied by a rapid temperature jump by tens of degrees with the release of large amounts of gaseous products. Increase in the external heating temperature to 230 °C lowers the onset thermal explosion temperature to 147 °C. In the open version the exothermic processes start at 165 °C, but they are very stretched in time and low intensive, nevertheless, UN conversion being sufficiently deep.

Resume : Fusion devices based on thermonuclear reaction are intensely studied nowadays, the most important one being realized in Cadarche, France. A problem in study is the material composition of the first wall of the reaction chamber, erosion, deposition and fuel retention of the mixed layers produced during device operation. One of the ways of simulating such high energy fluxes (10-100MW/m2) is laser irradiation using intense and focussed laser beams. In this paper we are investigated the fusion interest material's (Carbon, Tungsten and Beryllium based) behaviour under single or multiple laser beam pulses. Variable laser pulse wavelenghts were used in these experiments as well as variable laser pulse duration. The irradiated materials were investigated from morphological (using Scanning Electron Microscopy) and structural point of view. Raman investigations gave us information about carbon transformations of the superficial layer after the irradiation process while X-ray Photo Spectroscopy about some superficial and 'in-depth' structural modifications, including oxidation during and after irradiation. Some theoretical considerations about involved processes, based on the photon distribution into the beam and absorbtion processes also included.

Resume : The formation of secondary phases like corrosion product of spent nuclear fuel has been described by different authors [1-3]. Several mineral compounds have been identified as a secondary phases precipitated on the spent fuel under different repository conditions performed in leaching experiment [3]. The oxidizing conditions are likely to occur in the close surface of the spent fuel in the final disposal and as function of the boundary conditions they may lead to the formation of secondary phases. Therefore, the aim of this work is study the alpha-radiolysis effects on the spent fuel behavior and how can be altered by the presence of several parameters under repository conditions. The relevance of these studies are due to find an answer to experimental models and deepens the understanding of the mechanisms of irradiated matrix fuel alteration in order to extrapolate correctly corrosion rates of fuel from laboratory scale to storage scale. Corrosion experiments at room temperature under high concentrated H2O2 solutions of UO2 and SIMFUEL pellets under laboratory conditions were performed to simulate the conditions that the spent fuel will keep in final storage. The experiments were monitored by Raman spectroscopy which allows the in situ characterization of the alterations of the pellet surface, i.e. the identification of the solid secondary phases formed on the surface of these materials. We also performed the pre- and post- leaching characterization of UO2 and SIMFUEL surface by means of Raman spectroscopy, XRD, MO, EDX and SEM. The results of this work showed how after leaching experiments under oxidizing conditions matrix oxidation products as well as secondary phases precipitated during the tests as uranyl peroxides (studtite and schoepite) has been identified. References [1] D.J. Wronkiewicz, J.K. Bates, T.J. Gerding, E. Veleckis, B.S. Tani, Uranium release and secondary phase formation during unsaturated testing of UO2 at 90°C, Journal of Nuclear Materials, 190 (1992) 107-127. [2] C.N. Wilson, Summary of reults from the series 2 and series 3 NWSI bare fuel dissolution test, in: M.J. Apted, R.E. Westerman (Eds.) Scientific Basic for Nuclear Waste Management XI, Material Research Society, Boston, Massachusetts, 1987, pp. 473-483. [3] B. Hanson, B. McNamara, E. Buck, J. Friese, E. Jenson, K. Krupka, B. Arey, Corrosion of comercial spent nuclear fuel. 1. Formation of studtite and metastudtite, Radiochimica Acta, 93 (2005) 159-168.

Resume : For advance nuclear fuel cycles, different approaches based on extraction processes are being developed for minor actinides recycling contained in high level radioactive waste. To predict the behaviour along the extraction process is one of the most important challenges in defining the process development and safety. Solvents used for the partitioning will be in contact with highly radioactive aqueous solutions containing elevated nitric acid concentrations, therefore, it must be identify any unexpected behaviour as well as maximize the regeneration of the spent solvent. Malonamides and diglycolamides are considered as the extractants with higher possibilities to be used at large-scale in Europe. This work describes an overview of the physico-chemical effects of gamma radiation on solvent formulation for the DIAMEX, i-SANEX and GANEX processes based on diglycolamides. Besides, it has been studied the effects of degradation when those solvent formulations are part of an actual extraction process, taking into account the other phases, such as the extractant for selective stripping of trivalent actinides, a water soluble bis-traizinyl-pyridine. The behaviour of the extraction systems under radiation, by using a homogeneous gamma flux of external 60Co sources, has been explored depending on the structural stability of ligand, the nature of the diluents or the presence of different ions, etc. A systematic characterization of degraded samples (composition and extraction properties) and the study of their interferences with other phases are being carried out. The research leading to these results is being performed in the SACSESS project funded by the European Atomic Energy Community Seventh Framework Programme under grant agreement n° 323282.

Resume : The development of new reactor types of the fourth generation demands new types of fuel with high thermal stability. One possibility for new types of fuel is to embed the fissile material into an inert matrix, such as Molybdenum metal. Besides Uranium and Plutonium as fissile material, the Minor Actinide Am is planned to be embedded in the matrix and burned during the operation. The European ASGARD-Project focuses on the reprocessing of the new reactor fuels. During the reprocessing steps fissile material has to be separated from the matrix, most likely in liquid-liquid extraction steps, in order to recycle it. As Molybdenum shows poor solubility as well as slow dissolution kinetics in nitric acid, strategies are needed to enhance the dissolution process of the matrix. Here the addition of Fe(III) leads to an increased solubility of the Molybdenum matrix and enhances the dissolution kinetics. In order to understand the effect of Fe(III) on the process, the solution species Mo forms in the presence of Iron are investigated by means of nano-electrospray ionization mass spectrometry. By this method the ionic species present in solution can be characterized and quantified by transferring them into the gas phase under soft conditions and determining their mass-to-charge ratio. The dissolution of Mo both in presence and absence of Fe(III) in nitric acid are compared. The results hint on an increased solubility due to the formation of mixed Mo-Fe species.

Resume : Minor actinides like americium are produced during the irradiation of the most commonly used nuclear fuels, namely uranium dioxide or mixed uranium-plutonium oxide (MOX). In open nuclear fuel cycles, characterized by a period of surface storage followed by disposal of spent fuel in a geological repository facility, plutonium and the minor actinides will be responsible for the long term radiotoxicity of spent nuclear fuel, and for the medium term heat loading of the fuel. These transuranic elements are mainly alpha-emitters and will generate increasing amounts of alpha-damage and helium during spent fuel storage and in the repository. In closed fuel cycles, in order to minimize long-term radiotoxicity of the waste, it is envisaged to incorporate the plutonium and the minor actinides separated during spent fuel reprocessing in new fuels for transmutation in fast neutron reactors as MABB (Minor Actinide Bearing Blanket). Alternatively, the decay heat produced by some of the short-lived actinides could be advantageous e.g. for applications in radio-isotopic thermal generators (RTG) to power space probes. Alpha-decay generates defects mostly through elastic energy losses from the recoil of daughter nuclei. In addition, when the alpha-particle comes to rest it becomes a helium atom that can alter the microstructure of the material, e.g. by forming microscopic bubbles. All actinide dioxides have the fluorite structure, which is known as being a radiation damage resistant crystal configuration. However, high alpha-dose can have detrimental effects on the long term stability of materials envisaged as fuels, MABB or within RTGs. In this paper we report on experimental observations of alpha-damage effects in 238PuO2, 241AmO2, (U, 241Am)O2, (U, 238Pu)O2 by transmission electron microscopy (TEM), X-ray diffraction (XRD), thermal helium desorption spectrometry (THDS). The results are discussed to explain the differences observed in the various materials.

Resume : Actinides mixed dioxides are considered as reference fuel materials in several nuclear reactors concepts or as matrices for the recycling of minor actinides, either directly in the core or in fertile blankets. This study was thus focused on the dissolution of (MIV,LnIII)O2 and (MIV,MIV)O2 samples (with MIV = Th, U, Ce ; LnIII = La-Yb), as model compounds for future mixed oxide fuels, and highlight the role of often described conventional parameters (temperature, acidity or nature of dissolution medium, …). Moreover, a particular attention was also paid to the less studied structural parameters (role of oxygen vacancies associated to the incorporation of aliovalent elements, crystal structure, crystal defects, crystallite size, …) and microstructural ones (homogeneity, crystallization state, density, pore size and distribution, density and cohesion of the grain boundaries, …). Nevertheless, these latter parameters must be taken into account carefully when studying the evolution of the solid/solution interface. Indeed, ESEM observations performed in operando during the dissolution process allowed imaging the preferential alteration zones for several solids which can be located either at the grain boundaries, triple junctions or within the grains leading to the formation of intragranular corrosion pits. Also, it allowed pointing out the role of surface heterogeneities on the dissolution kinetics as well as the strong evolution of the reactive surface area during the dissolution of the ceramics. This information appears of main importance when working with normalized dissolution rates and led us to consider limit cases for using such variables.

Resume : Actinides mixed dioxides are currently employed in PWR reactors and stand as reference fuels for several Gen-IV concepts such as SFR. As new preparation routes are investigated to optimize the fabrication and the sintering of such ceramics, the preparation of dense (An,Ln)O2 mixed dioxides pellets, used as model compounds, was achieved from the initial precipitation of highly reactive precursors. A wet chemistry route, based on the mixture of cations in solution with large excess of NH4OH, was set up along with a drying step under vacuum aiming to avoid aggregation. Further characterization of the samples by XRD revealed the formation of hydrated MO2.nH2O that probably resulted from the ageing of hydroxide compounds. Also, TEMobservations evidenced the nanosized character of the powder, associated with specific surface areas in the 100-150 m2.g-1 range. The sintering of MO2.nH2O was investigated in a second step. Dilatometry indicated a low temperature of densification compared to other ways of preparation reported in the literature. Also, the pellets sinteredbetween 1350°C and 1550°C showed that a wide range of microstructures was achievable. Particularly, bulk materials with densities of 90-95 % of the calculated value could be prepared with average grain size ranging from around 100 nm to more than 5 µm. This simple process of elaboration of dense materials from highly reactive hydrated oxide precursor thus appears as a very interesting way to prepare oxide materials.

Resume : Many proposed sustainable nuclear fuel cycles include coextraction of minor actinides together with lanthanides from the high level liquid waste. The extraction of minor actinides is of particular importance due to their responsibility for the long-term radiotoxicity of the waste. For the trivalent actinides and lanthanides ions extraction, the diglycolamide family of organic extractants is being used, among others. Since it is well known that the degradation of the extractants in the highly acidic and active environment leads to undesirable effects, it is necessary to demonstrate their hydrolytic and radiolytic stability. For theoretical estimation of the general as well as the local chemical stability of the particular extractant's structure, the ab-initio simulations of TODGA (N,N,N´,N´-tetraoctyldiglycolamide) and related organic extractants were performed. The electronic structure calculations were carried out by the Gaussian and DMol3 computational codes with the use of the Kohn-Sham density functional theory. For the assessment of the chemical stability, various theoretical general and local stability indicators were calculated, such as energies and spatial distribution of the frontier orbitals, electrostatic properties, charge distribution. In the second step, some particular model degradation reaction pathways were investigated using the transition state theory. The results of different computational methods were compared to the published experimental results. Such theoretical chemical stability predictions can provide a valuable support to experimental scientists in development of novel, more stable organic extractants and extraction methods.

Resume : Over the long time periods of deep geologic storage, self-radiation damage of nuclear waste forms is primarily due to the alpha decay of the actinides. Radiation damage and helium accumulation from alpha decay can lead to phase changes and the formation of helium bubbles that may cause swelling and significant changes in mechanical properties. While the conditions for radiation-induced phase changes are reasonably understood, the conditions for formation of helium bubbles are not well known, and existing data suggest that phase changes and the onset of helium bubble formation could occur during interim storage in high actinide ceramic waste forms. On the other hand, glass waste forms appear to be more resistant to helium bubble formation than crystalline ceramic waste forms. We have recently investigated the accumulation of helium and radiation damage in two model waste forms, Gd2Ti2O7 and Gd2Zr2O7, which rapidly undergo radiation-induced crystalline-amorphous and order-disorder transformations, respectively, during interim or early storage time periods. Consequently, our studies of the long-term effects of continued accumulation of radiation damage and helium has focused on the response of amorphous Gd2Ti2O7 and disordered, defect-fluorite Gd2Zr2O7 to helium implantation and high radiation doses representative of 50,000 to 1 million years storage. The results of these studies will be discussed.

Resume : In France, high level radioactive wastes are vitrified in borosilicate glass matrix and are intended for eventual disposal in a deep geological repository. With time, groundwater will corrode the steel overpack and alter the glass. This alteration will lead to the formation of an alteration layer: a gel that is more or less passivating and secondary phases. The presence of chemical elements (Fe, Mg, Ca) in the repository environment can affect the glass alteration and the properties of this alteration layer which will impact the long term behavior of the glass. This study aims to determine the cumulative or competitive nature of the effects of Fe, Mg and Ca on the alteration of a reference glass: the International Simplified Glass (SiO2, B2O3, Na2O, Al2O3, ZrO2). This glass has been altered at 50°C in solutions containing one or more of these elements. In all experiments, the total concentration of elements was kept constant. The alteration kinetics were determined by leachate analysis (ICP-AES) and alteration films were characterized by ESEM, TEM-EDS and ToF-SIMS. Initial results indicate that Ca reduces alteration: it does not form secondary phases and makes the gel more protective. Mg and Fe enhance glass alteration by forming silicate secondary phases with different structures. The multi-element experiments suggest that the effects of these elements are cumulative: each element contributes independently to the altered layer formation and the alteration kinetics.

Resume : Coffinite (USiO4) and associated solid solutions are expected to play an important role in the field of direct storage of spent nuclear fuels in underground repository since they could control the concentration of actinides in the groundwater. However, thermodynamic properties associated with coffinite, especially the solubility product, remain poorly defined and require pure single-phase USiO4. The precipitation of coffinite from a mixture of U(IV)-containing acidic solution and sodium metasilicate appeared as the most promising method to provide USiO4 samples. In this context, a thorough multiparametric study of the formation of synthetic coffinite was achieved by varying various parameters such as pH, heating time, U/Si mole ratio and temperature in order to define optimal operating conditions for the preparation of coffinite. The optimized protocol allowed the preparation of polyphased samples that contained mainly USiO4 associated with oxide side products (amorphous SiO2 and UO2 nanoparticles). A purification process was developed with success to prepare pure synthetic coffinite samples suitable for solubility experiments. The ion activity product in solution equilibrated with USiO4 was determined through dissolution experiments conducted in 0.1 mol L-1 HCl under Ar atmosphere at room temperature. The dissolution was congruent and a constant composition of the aqueous solution was reached after 50 day. The solubility product of coffinite was then calculated (log*KS,USiO4 (298 K) = −6.14 ± 0.08). From these data, coffinite is less stable than the mixture of binary oxides at low temperatures, which is consistent with qualitative evidence from petrographic studies of uranium ore deposits. Finally a tentative mechanism was proposed to explain the formation of USiO4 providing new insights concerning the formation of coffinite in environmental conditions.

Resume : Iron oxidation state and coordination in the series of glasses (mol.%): 40 Na2O, (20-x) Al2O3, x Fe2O3, 40 P2O5 considered as matrices for immobilization of high Fe/Al legacy high level waste were determined by Mössbauer study. Iron in all the glasses was present as octahedrally coordinated Fe(III) and Fe(II). Fe(III) content in the glasses increased with the increase of the x value and reached ~74% of total Fe in the glass with x=15 while reduced to ~67% of total Fe at x=20. At the same time the lowest values of cumulative release from glasses into a deionized water for all the elements were found at x=5. Thus no correlation between the Fe(III) / Fe(II) ratio and cumulative elemental release values was found and iron speciation seems to be not a key factor determining their chemical durability.

Resume : Legacy liquid high level wastes (HLW) from defense programs performance which are under storage in stainless steel tanks at PA Mayak have high Fe/Al contents. They will be vitrified in a new EP-500 Joule-heated ceramic melter, which is planned to be commissioned in 2015-2016. Like previous melters of the same type the new melter will produce aluminophosphate based glass. Due to high content of iron oxides in the HLW the glass obtained will have base sodium-aluminum-iron phosphate composition. Complex sodium-aluminum-iron phosphate glassy materials with various Al2O3 to Fe2O3 ratio containing high level waste (HLW) surrogate were designed, produced, and characterized by X-ray diffraction and scanning electron microscopy and studied in details by Fourier transform infrared (FTIR) spectroscopy. The samples with high Al2O3 concentrations and not containing Fe2O3 were predominantly amorphous but subjected to devitrification under annealing. Addition of B2O3 and partial Fe2O3 substitution for Al2O3 in the materials increases resistance to devitrification whereas further substitution and NiO incorporation increase significantly tendency to devitrification. FTIR spectra demonstrate changes in the structure of glassy materials caused by both structural variations in the anionic motif and occurrence of crystalline phases in the materials. All the glasses had low leachability satisfying to Russian standard R 52126-2003 (similar to MCC-1 at 25 C) but the glasses at Al2O3:Fe2O3 ratio close to 1 were found to be the highest chemically durable.

Resume : Monazite ceramics are considered as potential nuclear waste forms for the conditioning of minor actinides. A combined understanding of their structural and microstructural properties is of great importance with regard to key parameters guiding the long-term stability of ceramic materials. Microstructure impacts mechanical properties and corrosion resistance. A certain porosity in accordance with the degree of waste loading is needed to avoid crack formation upon helium build-up (alpha-decay). The present work focuses on studies of La(1-x)EuxPO4 (x = 0, 0.5, 1) compositions where europium serves as an inactive surrogate for the minor actinides. Lanthanum-europium phosphate powders with different particle morphologies and microstructure were synthesised by using various synthesis methods. The crystallisation process of (La,Eu)PO4 powders was investigated through the combination of TG DSC, XRD, in-situ HT-XRD and HT-Raman spectroscopy. Dynamic aspects of sintering of La 0.5Eu0.5PO4 powders were studied in-situ by high temperature scanning environmental microscopy at 1340oC. The significant influence of the synthesis methods on the microstructure and crystal structure of powders (precursor materials) and pellets has been systematically investigated. As a result, these studies allow for optimisation of sintering parameters and to provide new insights on the microstructure development during heat treatment, including sintering kinetics and grain growth rate.