Symposium : E
|Fuel : Ch. Poinssot & K. Backman|
|09:00||Scientific and technical challenges for the development of generation IV fast reactor nuclear fuels|
Authors : N. CHAUVIN, M. PHELIP, P. GUEDENEY
Affiliations : DEN/CAD/DEC/SESC/DIR, CEA / Cadarache, France
Resume : Among the six concepts selected by the GENIV International Forum, two are of particular interest: sodium-cooled fast reactors (SFR) which already benefit from past experience and gas-cooled fast reactors (GFR) that could be deployed in the longer term taking into account their more innovative design. The key issues GENIV systems are faced with are plutonium multirecycling, optimising uranium resources, and, possibly, transmuting minor actinides while meeting safety requirements as stringent as for 3rd generation reactors. In the framework of the ASTRID project (SFR prototype of 1500 MWth), core pre-conceptual design studies are conducted in accordance with GEN IV reactor criteria, with regard to safety improvements particularly. A new design based on tight pin bundles with large pins and an axial fertile area at mid plane has thus been proposed with the particular aim of lowering the sodium void effect. The fuel is a standard (U,Pu)O2 with a 15-15Ti cw type stainless steel cladding. The challenge for the fuel element design is the optimisation of geometry and irradiation conditions in order to enhance safety features under nominal conditions as well as during power transients. This objective is mainly based on the feed back of several thousands of SFR fuel elements, calculations with new generation of fuel element performance codes and specific tests for qualification of fuel sub-assemblies. Heat resistant fuel forms constitute the key issue with respect to the feasibility and performance of GFRs. Requirements estimated at first to insure a safe management of the most severe cooling accidents are: preservation of the fuel element integrity as a barrier for fission products up to 1600 ◦C, and preservation of the fuel element geometry up to 2000 ◦C. Plate or pin shaped carbide fuels with SiC–SiCfiber composite cladding have been the subject of modelling and laboratory-scale R&D since 2001. The main focus was put on plate shaped fuel elements that were selected as reference for the GFR baseline concept in 2007. Current plans include R&D on pin shaped fuel element with a multilayer composite cladding and compliant thermal bridges between fuel pellet and cladding. Carbide fuel (U, Pu)C is taken as reference fuel for its high heavy atom content and good thermal conductivity that provide excellent neutronic properties (critical core size, breeding) and moderate normal operating temperatures (~1100 ◦C at 100MW/m3 average core power density vs a melting temperature above 2350 ◦C). A number of scientific and technical challenges remain to meet Gen IV criteria, in particular for GFRs in the field of nuclear materials. Some of them will be presented here.
|09:40||Ab initio simulations of oxygen adsorption and migration upon uranium nitride surfaces|
Authors : E.A. Kotomin, D. Gryaznov, D. Bocharov, Yu.F. Zhukovskii
Affiliations : Institute of Solid State Physics, University of Latvia, 8 Kengaraga str., Riga, Latvia Faculty of Computing, University of Latvia, Rainis blvd.29, Riga, latvia Faculty of Physics and Mathematics, University of Latvia, Riga
Resume : Although uranium mononitride (UN) is a promising fuel material for the future Generation IV nuclear reactors, which possesses several advantages as compared to the oxide fuels, oxygen impurities, which are always present in nitrides and carbides, lead to its unwanted contamination and further degradation. Thus, it is important to understand the mechanism of the initial stage of UN oxidation, in order to reduce or eliminate this process. A number of experiments were performed so far, aimed at analysis of oxygen effects on UN properties. These studies clearly show that oxygen contact with UN can lead to the transformation of UN surface layer into oxynitrides UOxNy . The results of ab initio DFT-GGA calculations of oxygen adsorption and migration on the UN (001) and (110) surfaces will be presented and compared. It is shown that surface O atoms are highly mobile and easily penetrate into surface nitrogen vacancies without considerable energy barrier. The energy diagram is presented which shows all elementary steps for oxygen penetration into UN surface, including O2 molecule adsorption, dissociation, O migration and interaction with surface vacancy. The temperature effects are incorporated through the oxygen chemical potentials. Results for slabs of different thicknesses and sizes are compared. The atomistic mechanism of UN oxidation via formation of oxynitrides is discussed
|10:20||The rationale of the future fuel cycles in view of sustainable nuclear energy|
Authors : Ch.POINSSOT, Ch.ROSTAING, M.MASSON, S.GRANDJEAN
Affiliations : CEA, Nuclear Energy Division, Radiochemistry & Processes Department. CEA MARCOULE
Resume : The sustainability of the current nuclear fuel cycles is not completely achieved since they do not optimise the consumption of natural resource (only a very small part of uranium is burnt) and they do not ensure a complete and efficient recycling of the potential energetic material like the actinides. Promoting nuclear energy as a future energy source requires proposing new nuclear systems that could meet the criteria of sustainability in terms of durability, bearability and liveability. In particular, it requires shifting towards more efficient fuel cycles, in which natural resources are saved, nuclear waste are minimised, efficiently confined and safely disposed of, in which safety and proliferation-resistance are ensured. Such evolution will require (i) implementing the recycling of the major actinides U and Pu since they can be used as energetic materials in fast neutron spectra, (ii) implementing the recycling of minor actinides which are the main contributors to the long term heat power and radiotoxicity of nuclear waste. Both options will require fast neutrons reactor to ensure an efficient consumption of actinides. In such a context, the back-end of the fuel cycle will be significantly modified: implementation of treatment/recycling processes, minor-actinides recovery and transmutation, production of lighter final waste requiring lower repository space. In view of the 2012 French milestones in the framework of the 2006 Waste Management Act, this paper will depict the current state of development with regards with these perspectives and will enlighten the consequences for the subsequent nuclear waste management.
|10:40||Calcined Resin Microspheres Pelletization (CRMP): a novel process for making metallic oxide ceramic of tailored microstructure.|
Authors : E. Remy, S. Picart, I. Bisel, T. Delahaye, N. Herlet, P. Allegri, O. Dugne, R. Podor, N. Clavier , P. Blanchart, A. Ayral
Affiliations : CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze, France, RadioChemistry and Processes Department, SCPS, LC2A ; CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze, France, RadioChemistry and Processes Department, SCPS, LC2A ; CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze, France, RadioChemistry and Processes Department, SCPS, LC2A ; CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze, France, Fuel Cycle Technology Department, SDTC, LEMA ; CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze, France, Fuel Cycle Technology Department, SDTC, LEMA ; CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze, France, Fuel Technology Department, SGCS, LMAC ; CEA, Nuclear Energy Division, F-30207 Bagnols-sur-Cèze, France, Fuel Technology Department, SGCS, LMAC ; Institut de Chimie Séparative de Marcoule, UMR 5257 CEA-CNRS-UM2-ENSCM, F-30207 Bagnols-sur-Cèze, France ; Institut de Chimie Séparative de Marcoule, UMR 5257 CEA-CNRS-UM2-ENSCM, F-30207 Bagnols-sur-Cèze, France ; Heterogeneous Materials Research Group, Centre Européen de la Céramique, F-87068 Limoges, France ; Institut Européen des Membranes, UMR 5635 CNRS-ENSCM-UM2, Université Montpellier 2, F-34095 Montpellier Cedex 5, France.
Resume : One of the most promising routes studied to reduce the radiotoxicity and the residual heat of nuclear waste is the transmutation of the long-lived radionuclides, such as americium in sodium fast reactor. Our study concerns the elaboration of mixed oxide precursors of uranium and minor actinides dedicated to pellet fabrication of Minor Actinide Bearing Blanket (MABB) for the transmutation of minor actinides , , . Mixed oxide precursors consist in uranium and minor actinides oxide microspheres, with a diameter in the range 100 to 1000 µm, obtained by the Weak Acid Resin process ,,. This conversion process is based on the loading of beads of ion exchanger with uranium and minor actinides and on their mineralization into oxide microspheres. The spherical morphology is preserved during calcination and the homogeneous precursor materials of millimetric size are then compacted and sintered into ceramic pellets (Fig 1). This process provides an interesting option for minor actinides bearing fuels preparation because the microspheres are suitable for remote control operations and significantly reduces the presence of highly contaminating dusts in the fabrication line ,. This present work constitutes a first step in the elaboration of oxide pellets by the CRMP process with the elaboration of cerium (as surrogate of americium) and uranium oxide microspheres precursors. The second step of the process concerns the compaction of those microspheres followed by the sintering of green pellets and has been tested for CeO2 and UO2±x materials: pellets with variable density could be obtained. Especially, two types of UO2±x pellets microstructure could be achieved: a relatively dense pellet and an open pore microstructure. References  B. Boullis, D. Warin In 11th Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, OECD-NEA, San Francisco, USA, 1-4 November 2010.  S. Pillon, D.Warin, Technique de l’ingénieur, BN3645, 2010.  A. Jankowiak, F. Jorion, Nuclear Science and Engineering, 160, 378–384 ,(2008).  P. Haas, (1974) US Patent 3,800,023  G. W. Weber, R. L. Beatty et al, Nucl Technol, 35, 217-226, (1977).  S. Picart, H. Mokhtari et al, (2009) FR Patent, 2936348.  E. Zimmer, C. Ganguly et al, J. Nucl. Mat., 152, 169-177, (1988).  C. Ganguly, U. Linke, and E. Kaiser, Metallography,20,1-14,(1987).
|11:00||An experimental study of oxalate to oxide decomposition mechanisms|
Authors : F. de Bruycker (a), B. Arab-Chapelet (a), N. Clavier (b), F. Abraham (c), S. Grandjean (a)
Affiliations : (a) CEA, Nucler Energy Division, RadioChemistry & Processes Department, BP 17171, 30207 Bagnols/Cèze, France; (b) ICSM, UMR 5257 CNRS/CEA/UM2/ENSCM, site de Marcoule, Bât. 426, BP 17171, 30207 Bagnols/Cèze, France; (c) UCCS, solid Chemistry, UMR CNRS 8181, ENSCL-USTL, BP 108, 59652 Villeneuve d’Ascq, France
Resume : Oxalic coconversion of actinides produces mixed oxides which can be used for fabrication of advanced fuel. Thus, the oxalate synthesis and the entire calcination mechanisms must be controlled to obtain appropriate oxide for each specific need. These mechanisms involve numerous steps from the initial crystallized mixed oxalate. During decomposition, the actinides redox has also a key-role as they create secondary chemical reactions. Thereby, the calcination mechanisms are quite complex. A first approach is based on the study of simplified systems via experiments on Th-lanthanide and Th-Pu systems. As thorium presents an invariant oxidation state, the study of such mixed oxalate decomposition allows quantifying the redox influence of the second metal. Comparison of two oxalate crystalline structures with a solid solution domain permits to study the influence of molecular and crystallographic characteristics on the calcination path. Hexagonal solid solutions have been prepared substituting thorium by either plutonium or lanthanides. Calcinations were performed under air or inert atmosphere. Complete characterisation of the Th-lanthanide oxalates will be presented and influences of the thorium substitution will be stressed through XRD and IR-absorption analyses. Evolution of the sample will be presented via analyses performed on quenched samples and in situ measurements carried out on MEBE and μ-Raman. Preliminary results will also be presented on the thorium plutonium systems.
|11:20||Helium behaviour in UO2 and actinide containing fuels|
Authors : Z. Talip, T. Wiss, A. Janssen, J-Y. Colle, E. Gilabert, D. Staicu, R. Eloirdi, R. Konings
Affiliations : Institute for Transuranium Elements
Resume : Nuclear waste management is one of the most important issues of nuclear technology. Research related to radioactive waste should help to predict the long term behaviour of the spent fuel from nuclear reactors with the aim to protect the humanity and the environment now and in the future. Alpha-decaying actinides produce large quantities of helium in the spent fuel. Although helium behaviour in UO2 has been studied since the 1960s, there is still considerable discrepancy between the published helium diffusion coefficients, justifying additional work to better understand the helium behaviour in UO2. The present study focuses on helium behaviour in different materials starting from fresh uranium dioxide to actinide-containing fuels. UO2 single crystal samples were used to investigate the intrinsic behaviour of helium without the effects of grain boundaries. Helium was introduced into the matrix by infusion under high temperature and high pressure. Helium diffusion coefficients in UO2 were determined following isothermal and isochronal helium release measurements by high resolution thermal desorption mass spectrometry. Many properties at the atomic scale, like defect mobility or self-diffusion, are strongly dependent on the stoichiometry of UO2. The effect of this latter parameter on the helium solubility was studied by using samples of UO2+x single crystals. Characterization of the samples was carried out by thermogravimetry, laser flash method and x-ray diffraction. The presence of irradiation defects in the spent fuel should increase the helium solubility and change its mobility. To better understand helium behaviour in more complex materials, UO2 samples doped with 238PuO2 and irradiated fuel samples were used. The results were interpreted based on thermal desorption measurements, TEM and XRD.
|11:40||Irradiation effects on the lattice parameter of chromia-doped UO2 investigated by micro-beam X-ray diffraction|
Authors : C. Mieszczynski 1, G. Kuri 1, C. Degueldre 1, M. Martin 1, J. Bertsch 1, C. N. Borca 1, D. Grolimund 1, Ch. Delafoy 2, E. Simoni 3
Affiliations : 1 NES & SYN, Paul Scherrer Institute, 5232 PSI-Villigen, Switzerland; 2 AREVA NP, 10 Rue Juliette Récamier 69006 Lyon, France; 3 Institut de Physique Nucléaire, Université Paris-Sud, 91406 Orsay, France
Resume : The performance of conventional UO2 fuel is mostly limited today by the phenomena of pellet-cladding interaction, fuel swelling and fission gas release. The impact of these phenomena increases with increasing burnup or power uprate. Consequently, the concept of UO2 doping has been overtaken to manufacture advanced fuels. Chromia (Cr2O3) has been also successfully used for the fabrication of larger UO2 grain structure. Therefore, it is necessary to understand the microstructural changes occurring in the fuel, due to both addition of dopants and incorporation of fission products upon irradiation. In this work synchrotron based micro-diffraction (micro-XRD) has been used to determine changes of UO2 lattice parameters resulting from irradiation induced defects and/or dissolved fission products in an irradiated chromia doped UO2. Several diffraction line profiles [e.g., (220), (311), (222) and (400) reflections] have been analyzed in details. The observed shifts towards lower or higher scattering angles in irradiated samples compared to those of fresh UO2 indicate a change in the d(hkl) interplanar spacing in irradiated UO2. A strong local burnup dependence of lattice parameters has been also observed. The micro-XRD patterns allow also the evaluation of the crystalline domain size and sub-grain formation at different locations of the fuel pellet. All these experimental results together with theoretical calculations of the UO2 lattice parameters will be presented and discussed.
|13:40||New SIMS data processing for Xenon depth profiling in Uranium dioxide|
Authors : B. Marchand(a,b), N. Moncoffre(a), Y. Pipon(a,c), N. Bérerd(a,c), C. Garnier(b), L. Raimbault(d), P.Sainsot(e), C. Delafoy(b), M. Fraczkiewicz(b), A. Perrat-Mabillon(a).
Affiliations : a- Université de Lyon, CNRS/IN2P3, Université Lyon 1, Institut de Physique Nucléaire de Lyon, 4 rue Enrico Fermi, 69 622 Villeurbanne cedex, France b- AREVA, AREVA NP, 10 rue Juliette Récamier, 69 456 Lyon, France c- Université de Lyon, Université Lyon 1, IUT Lyon 1, 43 bd du 11 novembre 1918, 69 622 Villeurbanne cedex, FRANCE d- Ecole des Mines de Paris, Centre de Géosciences, 35 rue Saint Honoré, F-77305 Fontainebleau cedex, FRANCE e- Université de Lyon, Université Lyon 1, LaMCoS, INSA-Lyon, CNRS UMR5259, F-69621 Villeurbanne cedex, FRANCE
Resume : During PWR reactor operation, around 25% of the created Fission Products (FP) are Xenon and Krypton. They have a low solubility in the nuclear fuel (UO2) and can (i) agglomerate into bubbles which induce mechanical stress in the fuel pellets or (ii) be released from the pellet, increasing the pressure within the cladding and decreasing the thermal conductivity of the gap between pellets and cladding. Fifty years of studies on the nuclear fuel did not succeed to fully understand all mechanisms of fission gas release (FGR). The deep understanding of gas atom behavior within the nuclear fuel is one of the major issues for nuclear fuel modeling with a safety purpose. In the literature, the Xenon diffusion coefficient in UO2 is mostly studied indirectly through release measurements applying the Booth model with some assumptions (for instance: grains are perfectly spherical, diffusion is the only Xenon release mechanism …). This induces some variability on the value of the diffusion coefficient. Direct measurements of Xenon depth concentration profiles using the SIMS technique can allow discriminating Xenon migration mechanisms. This is the goal of this study which proposes a new data processing in order to improve the Xenon depth profile achievement in UO2, taking into account the sputtering velocity of each grain.
|14:00||Phase equilibria of corium prepared from UO2 and Zircalloy-2 at temperatures of 1473 K – 1873 K|
Authors : M. Akashi, S. Hirooka, T. Uchida, K. Morimoto, M. Kato
Affiliations : Japan Atomic Energy Agency
Resume : Chemical and physical properties of corium are important data to evaluate the fuel behavior in severe accident of LWR. Those properties were affected by the phase state of corium. In this study, the phase equilibria of the corium prepared from UO2 and Zircalloy-2 were investigated as function of oxygen chemical potential. A sample which simulated corium was prepared from UO2 and Zircalloy-2 powders, whose molar ratio of U/Zr was 1:1. The sample was pressed and sintered at 1923 K for 2 h in 0.05%-H2/Ar mixed gas atmosphere adding moisture and for 1 h in 5%H2/Ar mixed gas without moisture adding. The sintered sample was melted at 2873 K, and the melted sample consisted chiefly of single fcc phase. The sample was crushed and used in oxygen potential measurement by using thermal gravimeter. The oxygen chemical potential was controlled by adjusting the ratio of PH2/PH2O. The oxygen-to-metal (O/M) ratio of the sample was estimated from the weight change of sample. Experimental result at 1473 K showed that the O/M ratio increased with increasing PO2. The O/M was about 2.1 in the PO2 of over 10-3 MPa. The O/M change at 1673 K and 1873 K were smaller than that at 1473 K. The data measured are contributed to analyze fuel behavior in severe accident and to treat the corium in post-severe accident.
|14:20||Relation between the elaboration route and the phase equilibria in (U0.55Pu0.45)O2-x MOX fuels|
Authors : T. Truphémus, R. C. Belin, S. Vaudez, J. Rogez
Affiliations : CEA Cadarache, CEA Cadarache, CEA Cadarache, CNRS-Université Paul Cézanne d’Aix Marseille III
Resume : (U1-yPuy)O2-x fuels with a high Pu-content are potential candidates for IVth generation nuclear reactors. In the hypo-stoichiometric range, the U-Pu-O system is known to exhibit a large biphasic domain depending on the Pu content. The precipitation of a second phase may significantly affect sintering properties as well as irradiation performance. However, phase equilibria are still to be fully described as various representations are proposed in the literature. In that prospect, a thorough knowledge of the uranium-plutonium-oxygen ternary phase diagram is of utmost importance. In this work, we provide new insights into the phase separation occurring in the U-Pu-O system by coupling high temperature XRD and optical microscopy experiments. We studied 45% Pu samples elaborated both by standard co-grinding and innovative co-precipitation processes. First, we show phase equilibria within this biphasic domain are slightly different at room temperature than that expected from the literature with the presence of a base-centered cubic phase. The comparison of samples with various oxygen to metal ratios provides valuable information regarding tie-lines in the miscibility gap, yet to be described in the literature. We believe such results will contribute to assess the thermodynamic description of IVth generation nuclear fuels. Surprisingly, we also observe a strong influence of the preparation on the phase equilibria at this composition, which will be detailed during the presentation.
|14:40||Electron energy loss spectroscopy investigation through a nano ablated uranium dioxide layer|
Authors : C. Degueldre*, R. Schaueblin**, J. Krbanjevic**
Affiliations : * NES, Paul Scherrer Institut, 5232 Villigen, Switzerland ** CRPP, EPFL, 5232 Villigen, Switzerland  Corresponding author: t. +41 56 310 41 76. email@example.com
Resume : A sample of uranium dioxide was produced (~10 x 10 x 0.05-0.15 mm) by focused ion beam for TEM/EELS examinations. This sample allows investigations of the EELS spectra recorded not only for the UO2 itself but also as a function of the thickness. The M, N and O edges were recorded over the 4000 eV energy range. The M4 and M5 edges were analysed using XAFS code FEFF8.4 for comparison of the fine structure over 0.5 nm from the actinide absorber. This edge analysis allows a better modeling and understanding of the EELS spectrum as well as a better separation of the plasmon part from the core loss part. In addition, the energy lost was studied through the range of thicknesses going from 50 to 150 nm to derive the electron mean free path and cross section for inelastic scattering in the plasmon part of the spectrum. The inverse of mean free path of inelastic electron for uranium dioxide is compared with that of reported earlier for other oxides from Be to Bi and for 200 keV. The derived data could be used for estimating the high temperature component of the thermal conductivity of UO2. This data shall be crucial for the analysis of nuclear fuels at high temperature.
|15:20||RECYLING THE MINOR ACTINIDES FOR DECREASING THE WASTE BURDEN TOWARDS FUTURE GENERATIONS. METHODOLOGY AND MAIN ACHIEVEMENTS BY 2012.|
Authors : Ch.POINSSOT, C.ROSTAING, P.BARON
Affiliations : CEA, Nuclear Energy Division, RadioChemistry & Processes Department, CEA MARCOULE
Resume : Minor actinides are demonstrated to be the key radionuclides for the long-term radiotoxicity of the high level nuclear waste. Furthermore, they bring most of the long-term residual heat power of the waste, and therefore are responsible for the geological repository waste density. Implementing their recycling is therefore of high interest for improving the burden associated to the ultimate waste towards future generations. Since 1991, France strongly supported a wide R&D program in this field, the aim of which is (i) developing partitioning processes able to recover minor actinides with a hight efficiency, and (ii) demonstrating their transmutation in fast reactors. In this framework, specific extracting molecules have been developed and demonstrated to be effective enough to recover either the sole americium, or the americium and curium, or all the minor actinides together with plutornium. This paper will depict both the methodology which has been developed for that purpose and the main achievements while a synthetic report is to be produced towards the French government on the industrial feasibility of minor actinides recycling by the end of 2012.
|15:40||Silicon Carbide and its Composites as Inert Matrix and Cladding Material for Minor Actinide Transmutation in Light Water Reactors|
Authors : 1) Dr. Sergei E. LEMEHOV; 2) Prof. Nikolai N. Ponomarev-Stepnoy
Affiliations : 1) Nuclear Materials Science Institute, Belgian Nuclear Research Centre K•CEN, Boeretang 200, B-2400 Mol, Belgium; 2) Russian Federation Academy of Science
Resume : Previous E-MRS 2011 Conference revealed industry supported interest to silicon carbide (SiC) in context of its use in Light Water Reactors as a clad material. Indeed, SiC exhibits favourable mechanical and chemical properties at high temperatures revealing dimensional stability in the 500-1000 0C temperature range which is the range where standard LWR fuels are used. However, in a neutron radiation environment SiC also exhibits dramatic changes in thermal properties showing decrease in thermal conductivity from very high values for as unirradiated material to somewhat ~ 10 W/m-K for as-irradiated SiC. Proposed paper first analyses existing public database and recommends an update for properties of SiC and its composites that can be used to model in-reactor behaviour of SiC as a clad material for “replacement” of standard zirconium based claddings. Secondly, paper gives an overview of important properties of SiC as being considered as an inert matrix for the use in LWR systems together with MOX and Am/Cm oxide fuel compositions. In particular, thermal properties of porous and dense SiC are reviewed and modelled. Paper proposes a number of model correlations that can be used for evaluation of SiC-based fuels with fuel performance codes to simulate temperature/dose dependences of SiC-based fuel swelling rate, thermal conductivity, and fission gas release. Exercises performed with the full-scale fuel performance code MACROS give clear confirmations that indeed use of SiC-based composites for a cladding and β-SiC for inert matrix in equivalent conditions will show performance that from many points of view is better than that of standard LWR fuels. Most visible advantages are due to mechanical stability of SiC claddings, low fission gas release and better responding to off-normal and accidental events.
|16:00||Analysis of In-reactor behaviour and Post Irradiation Examination data from HELIOS and FUTURIX-FTA tests with minor actinide fuels|
Authors : Sergei E. LEMEHOV
Affiliations : Nuclear Materials Science Institute, Belgian Nuclear Research Centre K•CEN, Boeretang 200, B-2400 Mol, Belgium
Resume : In Europe there exist two distinguishable approaches towards closing the nuclear fuel cycle by means of Participating and Transmutation (P&T), i.e. (a) Transmutation in future Gen IV fast reactor systems and (b) Transmutation in dedicated sub-critical reactor systems, such as Accelerated Driven Systems (ADS). The latter system can be regarded as dedicated “nuclear waste burner”. Initiated in SCK•CEN Myrrha project shall probably result in one of the very first practical examples of ADS burners. Minor actinide recycling is a prerequisite for both Gen IV and ADS strategies. In order to practically implement a sustainable, closed fuel cycle, P&T requires a coherent approach. In particular, numerous studies are to be undertaken to assess the transmutation of minor actinides in (CerCer and CerMet) fuels and targets. The European FAbrication, Irradiation and Reprocessing of FUELS and targets for transmutation (FAIRFUELS) multi-task project intends to address these challenges in a coherent way with considerable support by modelling and fundamental studies of primary effects and phenomenon in fuels/targets and materials. In short, FAIRFUELS is devoted to the experimental validation of different inert-matrix fuels/targets concepts (CerCer, CerMet). FAIRFUELS objectives will be achieved by the research activities, separated in two scientific domains, i.e. “Fabrication & Performance” with primary focus on fabrication technology and “PIE Experimental and Modelling” domain. The latter dThis paper will report recent modifications and model developments carried out to improve predictability and understanding of helium/fission gas behaviour, swelling and thermal behaviour of Gen-IV and ADS targets and fuels. Mo and MgO based inert matrices bearing high content of (U,Am)O2 and (Pu,Am)O2 compositions are considered. The experimental PIE of the dedicated fuel samples irradiated in the European transmutation irradiation programs (HELIOS, MARIOS, FUTURIX-FTA) as well as test conducted in OECD Halden Reactor Project are used for verification calculations, model validation and understanding further needs for new PIE data and model developments. Mo, MgO and other CerCer fuels are heterogeneous fuels which have some advantages and disadvantages compared to homogeneous fuels. Heterogeneity aspects need special attention and dedicated tools in order to have realistic assessments of CerCer fuels and PIE data.
|16:20||Uranyl oxo group functionalisation, rearrangement, and magnetic coupling|
Authors : P. L. Arnold*,(1) A.-F. Pecharman,(1) G. M. Jones,(1) E. Hollis,(1) J. B. Love,(1) N. Magnani,(2) R. Caciuffo,(3) H. G. Schreckenbach.(4)
Affiliations : 1 EaStCHEM School of Chemistry, University of Edinburgh, Edinburgh, UK; 2 Chemical Sciences Division, Lawrence Berkeley National Lab., Berkeley, CA, USA; 3 Institute of Transuranic Elements, JRC, Karlsruhe, Germany; 4. Department of Chemistry, University of Manitoba, Winnipeg, MB, Canada R3T 2N2.
Resume : The most common motif in uranium chemistry is the d0f0 uranyl ion [UO2]2 in which the oxo groups are rigorously linear and inert. This contrasts with Group 6 transition metals whose dioxo complexes are widespread in both enzymatic and industrial oxidation catalysis. The singly-reduced uranyl cation [O=U=O] is unstable in aqueous solution, but a few complexes have now been reported in recent years that are stable in anaerobic conditions, and although retaining the linear, covalent metal dioxo core, exhibits an enhanced oxo basicity. This basicity causes problems in nuclear waste separation, but is key to the removal of uranium from groundwater. Coordination in a wedge-shaped Pacman macrocycle has allowed us to study the chemistry of a chosen single uranyl oxo group of the linear dioxo dication, and avoid the standard equatorial U-ligand substitution chemistry. We will show how this can generate previously unseen uranyl chemistry, such as the thermally facile cleavage of hydrocarbon C-H bonds, the first covalent oxo-group functionalisation of the uranyl ion, and rearrangement of an oxo group to the cis-position to give a remarkably stable bis(uranyl) geometry with the strongest magnetic U(V)-U(V) coupling yet reported. The macrocycle also allows us to study the influence of secondary metal cations in the control of oxo group reactivity. We have isolated some homo and heteropolymetallic adducts that display unusual electronic and magnetic properties. The relevance of these structures to actinide chemistry present in nuclear waste will be discussed. References: 1. P. L. Arnold, D. Patel, C. Wilson, J. B. Love, Nature 2008, 451, 315. 2. P. L. Arnold, A.-F. Pecharman, E. H. Hollis, A. Yahia, L. Maron, S. Parsons and J. B. Love, Nature Chem., 2010, 2, 1056. 3. P. L. Arnold, E. Hollis, F. J. White, N. Magnani, R. Caciuffo and J. B. Love, Angew. Chem. Int. Ed., 2011, 50, 887. 4. P. L. Arnold, A.-F. Pecharman, and J. B. Love, Angew. Chem. Int. Ed., 2011, 50, 9456. 5. P. L. Arnold, G. M. Jones, S. O. Odoh, H. G. Schreckenbach, and J. B. Love, Nature Chem., in press, 2012.
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