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2015 Spring

Materials for energy and environment


Scientific basis of the nuclear fuel cycle III

Nuclear fuel cycle materials are studied for under demanding temperature, pressure and irradiation environments. These materials act as barriers and their properties are investigated with emphasis on mechanical performances, durability, plasticity and stability. Symposium includes sessions dealing with energy vector materials or processes ranging from fuels for thermal or fast reactors, their analysis after irradiation, their reprocessing for recycling and the waste forms. Macro-properties such as thermodynamic, thermophysical and mechanical as well as microstructural analysis of these materials are discussed for example comparing properties prior and after irradiation.


Scope :


The nuclear fuel cycle includes materials that enable the nuclear energy vector to be used after conversion from heat to electricity and proceeds through processes that allow fuels to be reused after reprocessing and recycling.

Fuel materials include oxides, nitrides, carbides or metals in homogeneous form or as composites such as cercer, cermet or metmet that can be used either as fuel or as target for transmutation. Currently the main challenge for fuel studies is to deliver the basis for the scientific understanding of fundamental properties, mainly at the atomic and microstructural level, which determine the evolution of macroscopic properties during irradiation. It is needed to develop predictive tools (models, codes) which allow optimizing the effort for the development of safe and innovative fuel concepts. The FUEL PROPERTIES session aim at introducing the research for fuel production optimization prior to irradiation.

Post-Irradiation Examination (PIE) provides most valuable data to assess the safety of nuclear fuel and materials during in-pile irradiation. During its operational life, nuclear fuel must retain its function within adequate safety margins while being subjected to lattice alteration due to neutron and fission damage, compositional change due to the accumulation of fission products, significant temperature gradients and mechanical constraint conditions. Relevant fuel properties affecting thermal transport efficiency as well as chemical and mechanical integrity can be investigated by PIE techniques carried out in shielded facilities. In the past, systematic irradiation campaigns followed by PIE have established the basis for the safe operation of commercial nuclear reactors worldwide.

The FUEL POST IRRADIATION EXAMINATION session aims at providing an overview of advanced PIE tools and procedures available, and of the main achievements obtained or expected by applying such techniques.

For optimal performance of the nuclear fuel cycle reprocessing and recycling is a must prior reutilisation of the fissile vectors. After then success of PUREX process other partitioning techniques have been profiled to also make use of the actinides as potential fertile elements, and to reduce by re-irradiation the radiotoxicity of the final waste material. The partitioning and transmutation approach allows utilization of the actinides. This process is the topic of an important R&D challenge that could yield to a strong reduction of neptunium, americium or curium in dedicated transmutation units or in thermal or fast reactors. The REPROCESSING session will address these key issues opening the strategy from single to multi-recycling based on the partitioning and transmutation strategies.

The disposal of high-level nuclear waste in deep geological formations poses major scientific and social challenges to be met in the next decades. One of the key issues is related to the long term safety of a waste repository system over extended periods of time (up to 106 years). Innovative waste forms in particular regarding actinide elements and some long-lived fission and activation products can help to improve this long-term safety aspect of deep geological disposal. Some of the relevant materials have been studied for several decades and have clearly demonstrated some of their superior properties. Further more, recent publications clearly show exciting and unexpected results. Also, new materials have been discovered in recent years to have very promising properties - such as murataite.

The WASTE FORM session is intended to provide an overview of recent advances and developments in related research areas including new synthesis routes, radiation damage and related structure - property relations as well as performance under repository conditions using both, experimental as well as computational approaches.


Hot topics to be covered by the symposium:


  • New challenges;
  • Fissile: Green Fuel production;
  • Fuels: Accident tolerant fuels;
  • Waste: from wet to dry storage;
  • Cycles: sustainable aspects.


List of invited speakers (tentative):


  • Janne Wallenius (KTH)
  • Phillipe Martin (CEA)
  • Melissa Denecke (Uni Manchester)
  • Thierry Wiss (ITU)
  • Steven Conradson (Soleil)
  • Charles McCombie (MCM)
  • Andrea Salvatores (CEA)


Scientific committee:


  • Gary Dyck (IAEA)
  • Lars Hallstadius (Westinghouse)
  • Manuel Pouchon (NES/PSI)
  • Giuseppe Modolo (FzJ)
  • Vinzenz Brendler (FzR)
  • Eric Simoni (Uni Orsay)


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Fuel Production and Properties : J. Somers
Authors : Melissa Denecke
Affiliations : Uni Manchester, UK

Resume : The peaceful use of nuclear power is celebrating round about its 60th anniversary. This is an excellent opportunity to review application of synchrotron radiaton (SR)-based X-ray spectroscopic and scattering techniques to characterize radioactive materials and elucidate determinant processes relevant to the nuclear fuel cycle. The penetration capability of intense X-ray sources allows in situ investigations, without having to separate the nuclides of interest out of the material to be studied and within radiological containments or sample cell environements. In my presentation, various examples of application of SR-based techniques in studies related to the nuclear fuel cycle (fuel, recycle, waste disposal) will be presented. Brief discussion of future photon sources will be presented at the close as an outlook.

Authors : Claude Degueldre
Affiliations : NES, Paul Scherrer Institute, 5232 Villigen, Switzerland Mail:

Resume : Uranium extraction is the first step of the nuclear fuel cycle. In today practice the only extractions are carried out on solid ores such as uranium rich minerals (% level) or minerals such as phosphates (ppm level). Since some years extraction of uranium from sea water (ppb level) has been the topic of investigations more especially in Japan because of its national interest. Taking into account the huge volume of ocean ware, the amount of uranium during practical extraction from sea water would remain quasi constant. The paper shows that for potential industrial extraction of uranium would be balanced by the input of soluble fractions recharged by the rivers in the oceans as well as by desorption from the sea floor. Recommendations for the extraction with use of gel panels (e.g. using amidoxime polymer gel absorbent) in high tide environments or fast oceanic stream over pelagic area are suggested.

Authors : D. Bocharov^{1,2}, M. Chollet^1, M. Krack^1, A. Kuzmin^2, J. Bertsch^1, D. Grolimund^1, and M. Martin^1
Affiliations : 1) Paul Scherrer Institute, Villigen PSI, CH-5232, Switzerland 2) Institute of Solid State Physics, University of Latvia, 8 Kengaraga Str., Riga LV-1063, Latvia

Resume : The doping of UO2 with small amounts of chromium is technically applied to obtain a larger average grain size after the fuel sintering process. A change in grain structure results in a higher plasticity of the fuel material as well as retention of fission gases inside the fuel matrix is improved due to larger grain size. It is not fully clear yet, how chromium atoms are incorporated into the UO2 structure. However, such information is needed to improve the material. In this work we study pristine and chromium doped uranium dioxide using force-field methods as well as an electronic structure methods based on density functional theory. Detection of energetically favorable Cr doped structures with CP2K [1] based on force-field calculations was performed. X-ray absorption spectra (XAS) of pure and Cr-doped UO2 were used to validate the modelling approach. We have employed two approaches. The first one, based on molecular dynamic simulations [2], was used to estimate the accuracy of the used force-fields for the simulation of pure UO2. In the second approach [3], the XAS at the Cr K-edge were used to check the accuracy of structure relaxations around chromium dopant by first principles calculations. [1] [2] A. Kuzmin and R.A. Evarestov, J. Phys.: Condensed Matter 21, 055401 (2009) [3] E. Blokhin, E. Kotomin, A. Kuzmin, J. Purans, R. Evarestov, and J. Maier, Appl. Phys. Lett. 102, 112913 (2013)

Authors : Peng Xu1, Ed Lahoda1, Lars Hallstadius1, Jason Harp2, Dave Goddard3
Affiliations : 1. Westinghouse Electric Company; 2. Idaho National Laboratory; 3. National Nuclear Laboratory

Resume : The high density uranium inter-metallic compound uranium silicide (U3Si2) has attracted much attention as a potential nuclear fuel due to several advantages it demonstrates in thermophysical properties, chemical and irradiation stability. It has a higher uranium density of 11.3 g/cm3 (9.6 g/cm3 for UO2), high thermal conductivity (15 W/mK at 27°C), reasonable melting point (~1665°C), and low radiation induced swelling rate (~5% at a fission density of 21021/cm3).These favorable physical and thermal properties lead to margin improvementand also make power maneuvers more flexible. More importantly, replacing the UO2 fuel with the U3Si2 fuelimproves fuel economy and may offset some of the additional production costsof adopting accident tolerant cladding such as SiC. In addition, thanks to the high thermal conductivity, the stored energy will be lower than for UO2 fuel, which is beneficial in an accident scenario. However, the nuclear industry needs to overcome several technical obstacles before implementing U3Si2. There is currently no irradiation data for pelletized U3Si2 at LWR operating temperature and burnup regime, though irradiation datais available fromstudies conducted intest reactors. Another open question is coolant compatibility at elevated temperatures. To learn more about irradiation performance and the water corrosion resistance of U3Si2, both theoretical work and experimental work are performedin the Department of Energy funded Accident Tolerant Fuel program led by Westinghouse. The fuel cycle impact of adopting U3Si2 fuel with zirconium cladding and other alternative claddings will be discussed. The one step synthesis of U3Si2 from UF6 was modeled using thermodynamic calculations and these results will also be presented.

Authors : Motoyasu Kinoshita, Fumihito Chiba, Tsuyoshi Iwashita, Masaaki Furukawa
Affiliations : Univ. of Tokyo/TTS Inc.; TTS inc.; ESTEX Co.; TTS Inc.

Resume : Liquid nuclear fuel has vast flexibility and potential for recycling and waste management. After commercial success of solid fuel technology most of fuel engineers have been away from liquid fuels and even negative to consider its reality. However after the nuclear accident of 3.11 Fukushima we experienced short performance of solid fuels at loosing coolant, reaching very high temperature, melting down the reactor core and releasing volatile radioactive fission products. If liquid fuel has potential to broden our horizon in nuclear industry we should look into it more seriously and bravely. In this presentation performance of liquid fuel was investigated by thermo-hydrodynamics calculations. Proposal will be made to realize use of liquid fuel in the present nuclear industry.

Authors : Laurent Claparede, Florent Tocino, Stéphanie Szenknect, Adel Mesbah, Nicolas Clavier, Nicolas Dacheux
Affiliations : ICSM, UMR 5257 Université Montpellier / CEA / CNRS / ENSCM, Site de Marcoule, Bât 426, BP 17171, 30207 Bagnols sur Ceze, France

Resume : Thorium-based dioxide materials are usually considered to be highly refractoryto dissolution / alteration during leaching/dissolution tests.The dissolution of Th1−xUxO2 solid solutions wasstudied by varying independently several parameters coming from the solution (nature and acidity of the solution,temperature, …) or from the solid (chemical composition, firing temperature, densification state, …). The chemical durability of the samples was significantly affected by the uranium mole loading due to the oxidation of tetravalent uranium into uranyl during the dissolution process. The influence of nitric acid concentration and temperature also showed that the behavior of Th1−xUxO2 solid solutions with higher uranium incorporation ratios (x = 0.52, 0.75 and 1) significantly differs from that of Th0.84U0.16O2 showing potential modifications in the dissolution mechanism occurring at the solid/liquid interface. For the higher uranium contents, it could result in the existence of the fast oxidation of U(IV) to U(VI) at the solid/solution interface then of the detachment of activated complexes formed with U(VI). Furthermore, initial normalized dissolution rate slightly depended on the elimination of crystal defects for firing temperatures below 900 °C but was almost independent of the crystallite size (T ⩾ 900 °C). Finally, dissolution tests on sintered Th0.84U0.16O2 samples showed that RL(Th) values decreased by an order of magnitude when the relative density increased from 79% to 89%.

Authors : M. Steppert(1), M. Cheng(1),(2), E. Ebert(3), C. Walther(1)
Affiliations : (1) Leibniz University Hanover, Institute for Radioecology and Radiation Protection, Herrenhäuser Str. 2, D-30419 Hannover, Germany. (2) Karlsruhe Institute of Technology, Institute forNuclearWasteDisposal (INE), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany. (3) Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung (IEK-6), Nukleare Entsorgung und Reaktorsicherheit, 52425 Jülich, Germany.

Resume : Molybdenum metal is one possible candidate as an inert fuel matrix to embed the fissile material for new proposed Gen IV reactor types. 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. Here the separation processes of the fissile material from the matrix in order to recycle it for new reactor fuel is in focus. To this end the solution species of Mo in strongly acidic media have to be characterized and quantified comprehensively. For this purpose electrospray ionization mass spectrometry (ESI MS), which can probe the stoichiometry and relative abundances of solution species, is applied. The method delivers unique insights into the solution speciation of molybdenum in strongly acidic media. The presentation will focus on the investigations on pure Molybdenum solutions as well as on the formation of mixed species Mo forms with analogues for the Actinides Am and Pu. Possible implications of the formation of the solution species on reprocessing steps will be discussed.

Authors : M.Chollet, J. Bertsch, D. Grolimund, M. Martin
Affiliations : Paul Scherrer Institut, Switzerland

Resume : The structure of the UO2nuclear fuel is radically transformed including at the atomic scale under neutron irradiation and fission in a power plant. Since this deep restructuration influences some relevant properties of the fuel like thermal conductivity or mechanical behavior, it is of first relevance for safety issues to characterize this evolution up to high burn-up. X-ray absorption spectroscopy (XAS) has the ability to reveal the element-specific atomic environment of material, but such studies on active material are still scarce up to now.The X05LA-microXAS beamline of the Swiss Light Source (PSI, Switzerland) offers the worthwhile opportunity to analyze radioactive materials by synchrotron techniques. The comprehensive interpretation of XAS spectra of such highly transformed material is not straightforward without a reliable reference because of, among others, the eventual changes in oxygen stoichiometry, cation valences and chemistry. In the first step of the study, we have provided high quality pristine UO2 XAS spectrum up to 14 k. The sample thickness was prepared by focused ion beam (FIB) to optimize the transmission measurements. X-ray diffraction data complete the structural dataset. This has permitted the understanding of the XAS spectra of a 9-cycles UO2 fuel (80 MWd/kg). The resultsprovide a better knowledge of the mechanism of the structural evolution of UO2 material at high burn-up. Moreover, we are confident that this kind of reference spectrum provides also valuable data useful in more theoretical approaches.

Authors : Daniel Shepherd
Affiliations : NNL Preston Laboratory, Springfields Works, Salwick, Preston, Lancashire, PR4 0XJ, UK

Resume : The Fukushima accident in 2011 highlighted some of the shortcomings of current nuclear fuel in LWRs and HWRs. Of particular concern is the potential runaway exothermic oxidation of the zirconium alloy cladding in steam at accident temperatures, which generates very significant heat and potentially explosive hydrogen gas. For this reason, modification or replacement of current zirconium cladding alloys is being sought.The low thermal conductivity of UO2 and MOX fuels is also a factor and hence means to improve the thermal conductivity are being researched as well as the use of alternative higher thermal conductivity fuel compounds. New design geometries using current fuel and clad materials are also being investigated as a means for improving safety margins.Of the plethora of ATF fuel concepts that have been suggested since Fukushima, some would offer greater accident tolerance than others. Economic favourability once deployed would also vary considerably. Furthermore, some concepts have a higher technology readiness and are hence foreseen to have a lower R&D requirement to qualify the fuel for reactor deployment.Quantifying these competing factors is therefore very important in order to allow international R&D to be prioritised to focus on the ATF concepts that are predicted to yield the greatest net benefit. This papersummarises current ATF concepts with discussion on their respective advantages and disadvantages.

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Authors : Z.V. Dzivanova, G.P Thorzhnitsky, D.I. Danilin, E.V. Belova, S.V. Stefanovsky
Affiliations : A.N. Frumkin Institute of Physical chemistry and Electrochemistry RAS

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.

Authors : Yan Li, George Beridze, Ariadna Blanca-Romero, Yaqi Ji, Victor Vinograd, Dirk Bosbach, Piotr M. Kowalski
Affiliations : nstitute of Energy and Climate Research: Nuclear Waste Management and Reactor Safety (IEK-6), Forschungszentrum Juelich GmbH, 52425 Juelich, Germany

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.

Authors : E.V. Belova, I.V. Skvortsov, Z.V. Dzivanova, A.V.Rodin, S.V. Stefanovsky
Affiliations : A.N. Frumkin Institute of Physical chemistry and Electrochemistry RAS

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.

Authors : Aurelian Marcu1, Liga Avotina2, Mihai Lungu3, Corneliu Porosnicu1, Cristiana A Grigorescu4, Cristian P Lungu1, Alexandru Marin5, Gunta Kizane3, Stefan Antohe6
Affiliations : 1National Institute for Laser Plasma and Radiation Physics, Laser Department, Atomistilor 409, P.O.BOX-MG-36, Bucharest-Magurele, Romania, 2Institute of Chemical Physics, University of Latvia, Kronvalda 4, LV 1010 Riga, Romania, 3National Institute for Material Physics, 077125 Bucharest-Magurele, Romania, 4National Institute R&D for Optoelectronics INOE 2000, 077125 Bucharest, Romania, 5Institute of Physical Chemistry “Ilie Murgulescu”, 060021, Bucharest, Romania, 6University of Bucuresti, Faculty of Physics, Bucharest-Magurele, Romania

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.

Authors : L. J. Bonales, S. Royuela, J. Cobos
Affiliations : Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid, Spain.

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.

Authors : Hitos Galán (1), Ana Núñez (1), Amparo González (1), Joaquín Cobos (1), Aritz Durana (2), Javier de Mendoza (2), Richard J. M. Egberink (3), Willem Verboom (3), Denise Munzel (4), Udo Müllich (4), Andreas Geist (4)
Affiliations : (1) Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), Madrid, Spain. (2) Institut Catalá d’Investigació Química (ICIQ), Tarragona, Spain. (3) University of Twente, Laboratory of Molecular Nanofabrication, Enschede, The Netherlands (4) Karlsruhe Institute of Technology (KIT), Germany

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.

Authors : M. Cheng(1),(2), E. Ebert(3), M. Steppert(1), C. Walther(1)
Affiliations : (1) Leibniz University Hanover, Institute for Radioecology and Radiation Protection, Herrenhäuser Str. 2, D-30419 Hannover, Germany. (2) Karlsruhe Institute of Technology, Institute forNuclearWasteDisposal (INE), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany. (3) Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung (IEK-6), Nukleare Entsorgung und Reaktorsicherheit, 52425 Jülich, Germany.

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.

Partitioning and recycling : Ch. Poinssot
Authors : T. Wiss(1), O. Dieste(1), A. Janssen(2), P. Raison(1), J.-Y. Colle(1), O. Benes(1), R.J.M. Konings(1), D. Prieur(1), J.-F. Vigier(1), V.V. Rondinella(1), J. Somers(1), P. Martin(3)
Affiliations : (1) European Commission, Joint research Centre, Institute for Transuranium Elements, P.O. Box 2340, 76125 Karlsruhe, Germany (2) The University of Manchester,Oxford Rd, Manchester, M13 9PL, United Kingdom (3) Commissariat à l’Energie Atomique et aux Energie Alternatives, Centre de Cadarache, 13108 Saint-Paul-Lez-Durance Cedex, France

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.

Authors : Laurent Claparede*, Florent Tocino, Denis Horlait, Johann Ravaux, Adel Mesbah, Nicolas Clavier, Stephanie Szenknect, Nicolas Dacheux
Affiliations : ICSM, UMR 5257 Université Montpellier / CEA / CNRS / ENSCM, Site de Marcoule, Bât 426, BP 17171, 30207 Bagnols sur Ceze, France

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.

Authors : J. Martinez(a),(b), N. Clavier(a), F. Audubert(b), N. Vigier(c), N. Dacheux(a)
Affiliations : (a) ICSM, UMR 5257 CEA/CNRS/UM2/ENSCM, Site de Marcoule Bât. 426, BP 17171, 30207 Bagnols/Cèze cedex, France (b) CEA, DEN, Cadarache, DEC/SPUA/LTEC, Bât 717, 13108 St Paul lez Durance, France (c) AREVA NC / BUR / DIRP / RDP, Boîte à lettre 406B, 1 Place Jean Millier, 92 084 Paris La Défense, France

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.

Authors : Tomas Koubsky, Ladislav Kalvoda
Affiliations : Czech Technical University in Prague, Faculty of Nuclear Sciences and Physical Engineering

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.

Waste form I : D. Bosbach
Authors : William J. Weber, Caitlin A. Taylor, Maulik Patel, Ke Jin, Haizhou Xue, Yanwen Zhang
Affiliations : Department of Materials Science & Technology, University of Tennessee, Knoxville, TN, USA

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.

Authors : Hélène Aréna, Nicole Godon, Diane Rébiscoul, Renaud Podor
Affiliations : CEA, DEN, DTCD, SECM, F-30207 Bagnols-sur-Cèze, France ; CEA, DEN, DTCD, SECM, F-30207 Bagnols-sur-Cèze, France ; ICSM, L2ME, F-30207 Bagnols-sur-Cèze, France ; ICSM, LNER, F-30207 Bagnols-sur-Cèze, France

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.

Authors : S. Szenknect(1), A. Mesbah(1), T. Cordara(1), N. Clavier(1), C., Poinssot(2), and N. Dacheux(1)
Affiliations : (1) ICSM, UMR 5257 CEA/CNRS/UM2/ENSCM, Site de Marcoule – Bât. 426, BP 17171, 30207 Bagnols-sur-Cèze cedex, France (2) CEA, Nuclear Energy Division, DRCP/DIR, CEA Marcoule, Bât. 400, BP 17171, 30207 Bagnols-sur-Cèze cedex, France

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.

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Authors : I.S. Glazkova, S.N. Kalmykov, I.A. Presnyakov, A.V. Sobolev, O.I. Stefanovskaya, S.V. Stefanovsky, S.E. Vinokurov, B.F. Myasoedov
Affiliations : Lomonosov Moscow State University; Institute of Physical Chemistry and Electrochemistry RAS, Vernadsky Institute of Geochemistry and Analytical Chemistry RAS

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.

Authors : S.V. Stefanovsky, O.I. Stefanovsky, M.B. Remizov, P.V. Kozlov, E.A. Belanova, R.A. Makarovsky
Affiliations : Institute of Physical Chemistry and Electrochemistry RAS; FSUE PA Mayak

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.

Authors : Y. Arinicheva 1, S. Neumeier 1, N. Clavier 2, A. Bukaemskiy 1, R. Podor 2, N. Dacheux 2, D. Bosbach 1
Affiliations : 1 Forschungszentrum Juelich GmbH, Institute of Energy and Climate Research - IEK-6: Nuclear Waste Management and Reactor Safety, Juelich, Germany; 2 Institut de Chimie Separative de Marcoule, UMR 5257 CEA/CNRS/UM2/ENSCM Site de Marcoule - Bat. 426, BP 17171, 30207 Bagnols / Ceze cedex, France

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.


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Symposium organizers
Claude DEGUELDRELancaster University

Engineering Department, Lancaster, UK

+ 44 1524 592716
Joseph SomersITU

Karlsruhe Germany

+49 7247 9510

Marcoule France

+33 (0)4 6679 6618
Dirk BOSBACHJulich F Zentrum

05.3 Raum: 288 Julich Germany

+49 2461 61 5299
+49 2461 61 2450