Nuclear materials

Resume : China has been committed to peaceful use of nuclear energy for decades. Currently China has the third largest nuclear power industry in the world, with 45 units in operation and 11 units under construction. Here provides the overview of China’s nuclear power policies, authorities, major enterprises and their nuclear power plants, R&D programs and their needs for new materials.

Resume : Materials employed in nuclear environments should be able to exhibit the highest radiation tolerance. Mainly the materials issues such as embrittlement and swelling in fuel cladding and structural components are responsible for the lifetimes of current and even more of new-build and future reactors. Radiation tolerance-related requirements are further boosted by the higher performance expected for Generation IV fission and fusion reactors, which will expose materials to much higher numbers of atomic displacements then current and near-feature reactors. During nuclear energy production, both fuel components and structural materials are subject to substantial radiation damage. They initially appears in the form of local intrinsic point defects within the material (i.e. vacancies and interstitials). Then, these point defects agglomerate, interact with the underlying microstructure and lead to undesirable effects such as blistering and radiation-induced embrittlement, which render the materials. Another important factor is represented by helium embrittlement. He originates from the transmutation of reactor elements which can release alpha particles that acquire electrons to become helium atoms. He is insoluble and mobile in most metals and migrates to grain boundaries and interfaces where it forms bubbles leading to embrittlement. Many of the materials being developed for deployment in these environments are based on incremental improvements to existing materials such as Oxide Dispersion Strengthened steel. Recently, as a new material to be used in nuclear structures, we prepared Zr-Nb metallic multilayer composites using magnetron sputtering technique and showed both theoretically and experimentally that they are mechanically durable under the harsh environments we mentioned above. Here, we present the density functional simulation results of our further investigation, vacancy-interface-He interaction and migration of He in nanoscale Zr-Nb metallic multilayer composite. Our results enlighten where He atoms would be positioned within the material and why as well as how they would move within the material.

Resume : Nuclear fuel cladding materials that possess higher performance attributes than the currently used cladding materials are always desired in order to improve nuclear fuel performance in reactors. In order to improve the properties of the existing light water fuel cladding material, laser shock processing (LSP) has been applied to commercial zirconium alloy. After LSP a thin oxide layer formed on the surface of the specimen. Meanwhile, the surface residual stress and microstructure of the specimen was altered. After being exposed to high temperature steam for a long time, the corrosion weight gains of samples processed by the pulsed laser were less than those that were untreated, and the transition time of corrosion kinetics has been delayed. We also studied the mechanism of zirconium alloy corrosion process, especially the effect of macroscopic compressive stress on corrosion of zirconium alloy. From the above experiments we concluded that LPS could enhance the corrosion resistance of zirconium alloy.

Resume : As the Loss of coolant accident in Fukushima showed, it is vital for public safety and fuel integrity to understand and continuously improve the performance of fuel cladding materials for light water reactor applications, both during in-core use and in long-term post-irradiation storage. One of the major causes of embrittlement and failure in reactor structural materials including fuel cladding materials is the formation and evolution of precipitates by the formation of hydrides or radiation-induced phase separation. In this presentation, we will summarize our recent work to characterize hydride formation in zircaloy 4 fuel cladding and the formation and thermodynamic stability of radiation-induced precipitates in FeCrAl fuel cladding candidate alloys by small-angle neutron scattering. Small-angle neutron scattering (SANS) is a powerful technique for measuring bulk averaged nanometer length scale structures in a variety of materials nondestructively, providing complementary information to direct-geometry small-volume or surface techniques like electron microscopy and atom probe tomography. This technique is well-suited for use in studying the properties of alloys due to the widely varying contrasts of different transition metal isotopes when viewed with neutrons, the sensitivity of neutrons to magnetic structure, and the high penetrating power of neutrons in many materials regardless of atomic number. The large and opposite neutron scattering cross sections of hydrogen and deuterium make SANS and other neutron techniques extremely well-suited to the characterization of hydrides in alloys. These properties, together with the high flux available at modern neutron scattering user facilities and the existing infrastructure for working with radioactive materials, make SANS a uniquely powerful technique for studying reactor structural alloys. The first work that we will describe is a nondestructive neutron scattering method to precisely measure the uptake of hydrogen and the distribution of hydride precipitates in light water reactor (LWR) fuel cladding was developed. Small angle neutron incoherent scattering was performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Our study demonstrates that the hydrogen in commercial Zircaloy-4 cladding can be measured very accurately in minutes by this nondestructive method over a wide range of hydrogen concentrations from a very small amount (~20 ppm) to over 1000 ppm. In the second part of the talk we will describe how SANS is used to characterize the growth of irradiation induced precipitates in Fe-Cr-Al model alloys ranging in composition from 10-18 wt.% Cr and 3-5 wt.% Al that have been irradiated in the Oak Ridge National Laboratory High Flux Isotope Reactor at nominal damage doses up to 13.8 dpa as a function of dose and to probe the in-situ dissolution of the radiation damage precipitates in a single alloy as a function of temperature and time. We also describe the procedures used to perform measurements using on highly radioactive samples shielded to minimize personnel radiation dose and the risk of contamination. One challenge faced when studying alloy systems with SANS or when using thick metallic shielding for irradiated samples is the possibility of contamination of the SANS signal by multiple Bragg diffraction, particularly when working with shorter neutron wavelengths. In this talk, we will show examples of how this can appear and describe a strategy for reducing it impact. Methods such as this will be particularly powerful and valuable as the focus in SANS instrumentation shifts from the traditional, monochromatic steady-state instruments typical of reactor scattering facilities to polychromatic time-of-flight sources like those at pulsed sources like most spallation sources and now being introduced as an option at steady-state sources such as the ILL. The High Flux Isotope Reactor and beamline CG2 of ORNL was sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, U.S. Department of Energy. Beamline CG3 is supported by the Office of Biological and Environmental Research of the U.S. Department of Energy through the ORNL Center for Structural Molecular Biology. A portion of this research was sponsored by the Laboratory Directed Research and Development (LOIS-6502) Program of Oak Ridge National Laboratory, managed by UT-Battelle, LLC, for the U.S. Department of Energy under Contract No. DE-AC05-00OR22725), and EBSD through a user project supported by ORNL’s Center for Nanophase Materials Sciences (CNMS), which is sponsored by the Scientific User Facilities Division, Office of Basic Energy Sciences, U.S. Department of Energy.

Resume : MAX phases have been considered for various applications, such as contact and structural materials in nuclear applications among others. Particularly, Zr- and V-containing MAX phases, exhibit a remarkable resistance towards oxidation, irradiation and thermal solicitation. To this end, we investigated the effect of the substrate on the growth of V2AlC MAX phases on both steel and Zr-4 substrates. The coatings were obtained using cathodic arc deposition using elemental targets. The deposition parameters were determined to deposit films of up to 3 μm in thickness. The films were systematically analyzed to track their growth properties, in terms of orientation and microstructure with respect to the substrate.

Resume : As candidate accident tolerant fuel (ATF) cladding materials for light water reactors Advanced FeCrAl + Ni alloys were developed with nominal compositions of Fe-(13−16)Cr-(5−7)Al-(0−19)Ni. High temperature corrosion behavior of advanced FeCrAl + Ni alloys was examined by high pressure steam thermo-gravimetric analyzer (HP-TGA) at 1100−1300 °C and up to 10 MPa. Effects of pressure on oxides formation at elevated temperature were studied as compared to commercial FeCrAl alloy. Overall, FeCrAl + Ni alloys formed protective alumina under accident-like conditions while intermetallic NiAl phase seemed to have effects on the surface oxide morphology while protective alumina layer was formed regardless of the compositions.

Resume : As candidate accident tolerant fuel (ATF) cladding materials for light water reactors, advanced FeCrAl+Ni alloys were previously developed with nominal compositions of Fe-(13−16)Cr-(5−7)Al-(0−19)Ni. They were aged at 425 °C up to 3000 h and the changes in microstructure and mechanical properties were examined where the degree of embrittlement was compared to commercial ferritic Fe-21Cr-5Al alloy. It was found that intermetallic NiAl phase of advanced FeCrAl+Ni alloys seemed to play an important role in achieving high strength in virgin condition while the low Cr content suppressed α−α′ phase separation in the ferrite phase.

Resume : Precipitation hardened ferritic-martensitic chromium steels of 12% Cr type are considered as a promising fuel cladding materials for the new generation fast neutron reactors with the liquid metal coolant, in particular, lead or lead-bismuth cooled. Problems of interaction between the liquid metal and the fuel cladding materials, as well as the development of methods to improve the corrosion resistance of steels are important tasks to be solved. During last decades different methods were shown to be effective in fabrication of additional surface alloyed layer with elements that form thin and dense protective oxide layers on the surface of steel (such as aluminum and chromium) thus preventing steel from corrosion. Although surface alloying shows promising effect on 12% Cr steel corrosion resistance, there is still an open question of radiation stability of protective oxide scale during neutron irradiation in a reactor. The aim of present research is to reveal the effect of ion irradiation on the evolution of microstructure and radiation stability of oxide scale and oxide/steel interface. Samples of EP823 ferritic-martensitic steel were pre-oxidized in flowing liquid lead loop with controlled oxygen content and irradiated by 4 MeV Au ions at room temperature and at 650 C. Results of transmission electron microscopy characterization of the microstructure of oxide layers and the oxide/steel interface of the ion irradiated steel samples will be presented.

Resume : The development of safer and more efficient nuclear reactors calls for new materials with better radiation tolerance. In recent years, we have made a renewed exploration on the controlling mechanisms of radiation tolerance of alloy matrix by deliberately taking out the effects of complicated microstructure features (or the pre-existing defect sinks) and focusing only on the response of alloy matrices of various composition and crystal structure. Through a systematic study of a large group of nickel-based fcc single-phase concentrated solid solution alloys with ion beam irradiation and TEM analysis, we have found that the resistance to void swelling generally increases with increased number of alloying elements, but similar enhancement in swelling resistance can also be achieved in selected binary alloys by increasing the concentration or changing the species of the alloying element. The result can be explained by the change in the sluggishness of interstitial cluster motion that may affect the recombination rate of the Frenkel defect pairs. The results from the fcc alloys are also compared with that obtained from several bcc alloys, such as FeCrAl and Mo, under the similar irradiation conditions. The effect of alloying composition on radiation hardening studied with nano-indentation will also be presented and discussed in terms of the characteristics of dislocation loops.

Resume : Nanostructured oxide dispersion strengthened (ODS) ferritic steels with two different nominal compositions of Fe-14Cr-2W-0.3Ti-0.3Y2O3 (Si-free ODS steel) and Fe-14Cr-2W-0.3Ti-0.3Y2O3-1Si (Si-containing ODS steel), (wt. %) were synthesized by mechanical alloying for 50 h and consolidated by spark plasma sintering at 900 °C for 5 min at 60 MPa pressure. The effect of Si on the microstructure, mechanical properties as well as on the oxidation behavior was investigated. Addition of silicon caused the reduction in milling time and resulted in finer (5 ± 1.2 µm) and uniform powder particles. No major changes in sinterability have been observed in both the alloys. XRD analysis confirmed the austenite and martensitic formation from the ferrite matrix in Si-free ODS steel. However, phase transformation (α-Fe to -Fe) was hindered in the Si-containing ODS steel. The hardness values of the Si-free ODS steel was higher than the Si-containing ODS steel owing to the presence of martensite in the former steel. Presence of nanoparticles (Cr2TiO4, SiO2, and Y2Ti2O7) caused approximately two times finer grain size in case of Si-containing ODS steel. Synergistic effect of Cr and Si addition and finer microstructure helped to form thin, dense and protective oxide layers, which prevented outward cations and inward anion transport, as a result, excellent oxidation resistance of Si-containing ODS steel. Keywords: Nanostructured ODS ferritic steel; Mechanical alloying; Spark plasma sintering Rietveld refinement; morphology, oxidation resistance

Resume : The advanced oxide dispersion strengthened (ODS) 14YWT ferritic alloy was developed for resisting microstructural changes due to exposure to harsh conditions in future reactor concepts, requiring resistance to high dose neutron irradiations at high temperatures. Resistance to microstructural degradation of ODS alloys to ion irradiation has been demonstrated so far but, unfortunately, there are scarce studies focusing on the microstructural response of 14YWT to neutron irradiation. The investigation in this work presents nanoscale microstructural characterization on neutron irradiated specimens of (ODS) 14YWT ferritic alloy (SM13 heat), with nominal composition Fe-14Cr-3W-0.4Ti-0.3Y2O3 (wt.%) irradiated at the BOR-60 reactor in Russia, at 16.6 dpa and at two different temperatures: 386 C and 412 C. The stability of the microstructure after irradiation is stated by the presence of Y-rich nanoparticles and no evidence of dislocation loops or cavities formation. The main irradiation effect observed on these samples is the alpha prime formation at both temperatures, which was not observed in previous ion irradiation experiments. A detailed study of these events has been performed in terms of atom probe tomography (APT) and transmission electron microscopy (TEM) techniques. These results provide essential data as a baseline to the aim of establishing a correlation between ion and neutron irradiation effects in structural materials.

Resume : The forthcoming Generation IV nuclear reactors will operate at much higher temperatures than the existing nuclear power plants. This feature makes the selection of structural materials to be used in fuel assemblies employed in the reactor core much more challenging. One of the most promising candidates are oxide dispersion strengthened (ODS) steels that are well known for their advanced mechanical properties and radiation resistance. However, increased He/dpa production ratio expected in Generation IV reactors bears potential risks of long-term degradation of steel properties as a result of gas accumulation and interaction with strengthening oxide nanoparticles. The report presents the results of a systematic Transmission Electron Microscopy study of microstructural evolution of ODS-EUROFER steel caused by helium ion implantation up to high doses. The oxide nanoparticles are found to be excellent nucleation sites for helium bubbles, yet their contribution to swelling is found to remain relatively minor as compared to other microstructural defects, especially grain boundaries. However, the bubble growth on oxide particles is shown to be potentially risky in terms of the loss of oxide particle efficiency as dislocation pinning centers and of triggering the bubble-to-void transition, both effects resulting in the severe degradation of steel mechanical properties.

Resume : Steels have important applications in current and advanced nuclear reactors, however, their irradiation tolerance and mechanical properties need to be improved. Bulk ultrafine-grained metals possess drastically higher strength than their conventional coarse-grained counterparts, and may have significantly enhanced irradiation tolerance. In this study, ultrafine-grained austenitic and ferritic-martenstic steels were manufactured by equal-channel angular pressing (ECAP) and high-pressure torsion (HPT). The microstructure and mechanical behavior of the steels manufactured by ECAP and HPT were carefully studied. Advanced microstructural characterization techniques were utilized to investigate the microstructures and chemistry of the steels before irradiation. The thermal stability of the ultrafine-grained steels was also investigated. Neutron irradiation was designed and is being performed to study irradiation behavior of the steels. Limited ion irradiation was also conducted to compared with the neutron irradiation. Results indicated that the ultrafine-grained steels manufactured by HPT and ECAP possess significantly improved hardness/strength compared to their conventionally manufactured coarse-grained counterparts. Grain size of HPT samples is smaller (~100nm) than ECAP samples (~400nm). All the ECAP and HPT steels are thermally stable at least up to 500 °C, and many of them are stable up to 600 °C.

Resume : The expected degradation of the mechanical properties of structural materials under irradiation (i.e. "in operation") in nuclear components represents a significant challenge for the design to account the impact of long-term irradiation effects. Therefore, the development of scientific and engineering expertise to understand and possibly control the severe impact on materials of harsh neutron irradiation is one of the tasks in the current material’s research agenda. Radiation-induced embrittlement is caused by nano-scale defects that obstruct plasticity mediated by dislocations. The present contribution highlights recent research and development efforts addressed towards the assessment of the interrelation between nano-structural features and hardening induced by neutron irradiation in structural high-Cr ferritic/martensitic steels. In this work, we summarize and discuss recent results obtained using several fine-scale experimental techniques (namely: transmission and scanning electron microscopy, small angle neutron scattering, internal friction and magnetic after effect measurements, atom probe tomography) applied to the Fe-Cr model alloys from a single dedicated irradiation campaign. Recent experiments were performed to investigate the impact of such important alloying elements as Ni, Si and P on the microstructural evolution. We provide new microscopy information on the pattern of irradiation damage in the Fe-Cr alloys doped with the above noted impurities. To rationalize and bridge experimental information, we apply material's modelling framework, involving several up-scaling methods from atomistic simulations to the full-scale Monte Carlo model to predict nano-scale irradiation defect evolution. As a result, we distinguish several types of microstructural features created by neutrons which are dislocation loops, voids and solute rich clusters.

Resume : Ferritic-martensitic (F/M) steels have been proposed as the candidate materials for in-core and out-core applications in the Canadian Generation-IV Super Critical Water-cooled (SCW) nuclear reactors. Extensive studies have been conducted to investigate the corrosion behavior of F/M steels in SCW. However, the investigation of the alloying elements on the corrosion behavior of F/M steels exposed to SCW is quite limited. This study examined the corrosion behavior of a series of F/M steels containing various amounts of Si and Mn exposed to SCW at 500 °C and 25 MPa. Gravimetric, Scanning Electron Microscope, Transmission Electron Microscopy (TEM), Energy-dispersive X-ray Spectroscopy (EDX) and X-ray Diffraction (XRD) analyses were conducted to characterize the corrosion behavior of the F/M steels in SCW. The results showed that the concentration of the alloying elements (Si, Mn) impacted both the oxidation kinetics and the oxide scales formed in a synergistic manner. Nevertheless, the increase of Mn content induced in the formation of a non-uniform oxide scale on the steel surface. The corrosion mechanism affected by the alloying elements (Si, Mn) was discussed in this study.

Resume : We conducted reactor performance calculations to assess the potential design basis accident performance of HTGR fuel designs. Three Fully Ceramic Microencapsulated (FCM) fueled HTGR designs were developed in a previous work. The maximum fuel temperature in the cores fueled by these three FCM fuels was predicted to be higher than that in the reference 350-MWt mHTGR core in both normal operating conditions and during representative design basis accidents. To better understand the potential safety margins in mHTGR design basis accidents, we performed thermal-hydraulics sensitivity studies to investigate how maximum fuel temperature varies considering various parameters, e.g. thermal properties, within the ranges corresponding to the differences between the FCM-fueled prismatic mHTGR cores and the reference core. We found that the difference in the steady-state axial power distribution contributed the most to the difference in the maximum fuel temperature, in both normal operation and design basis accidents. Experimental data suggested that the annealing process of irradiation defects in SiC would be rapid at mHTGR relevant fuel temperatures. The bounding potential impact of the SiC annealing on the maximum fuel temperature was analyzed considering both the thermal conductivity recovery and the Wigner energy release due to the annealing of SiC. We found that the defect annealing process in SiC would at most increase the peak maximum fuel temperature of an FCM-fueled core by 40 K in loss of forced cooling accidents and by 10 K in a control rod withdrawal accident. Additional experiments on the SiC defect annealing kinetics and Wigner energy release in more relevant conditions are needed.

Resume : Dense tungsten-iron-borocarbide (W-B-FeCr) materials were synthesized using laboratory scale industrial processes and following on from studies of boron additions in cemented tungsten carbides (cWCs) with non-activating FeCr binder1. Previous studies2 have identified W-B-FeCr materials and cWCs as potential candidates for radiation shielding in compact nuclear reactors. Cemented tungsten carbide is widely used in tooling due to its excellent mechanical properties, combining the hardness from the WC skeleton with the ductility from the metallic binder. The combination of high-Z W and low-Z C is advantageous for shielding against gamma and neutron radiation. However, cWCs has received little attention, due in part because Co and Ni are the most common binder metals and are both activation hazards under neutron irradiation. The recent discovery of FeCr as a ductile, non-activating cWC metallic binder3 enables the use of WC as a candidate radiation shielding material since unlike Co or Ni-based alloys it does not activate when irradiated. Such a cWC-based shield combined with boride-based components may yield radiation shields with unrivalled attenuation properties ? outperforming even metallic tungsten4 This work investigates the phase abundance and microstructure of these materials and the implications with respect to manufacture and mechanical properties. Quantitative determination of phase abundance using XRD, XRF and microscopy aims to provide data on the microstructure of these materials and its implications for incorporation alongside potential cWC-based shields in nuclear reactors and other irradiating environments. 1Humphry-Baker et al. Scripta. Mater. (2018) 155 DOI: 10.1016/j.scriptamat.2018.06.027 2Windsor et al. Nucl.Fus. (2018) 58(7) DOI: 10.1088/1741-4326/aabdb0 3 Humphry-Baker et al. 19th Plansee Seminar. Plansee. Austria (2017) http://wrap.warwick.ac.uk/94047 4 Windsor and Morgan. Nucl.Fus. (2018) 57(8) DOI: org/10.1088/1741-4326/aa7e3e

Resume : Single-phase concentrated solid solution alloys (SP-CSAs), including high entropy alloys (HEAs) are a novel family of materials for studying defect dynamics without preexisting defect sinks. In contrast to conventional alloys, SP-CSAs are composed of two to five principal elements in equal or near-equal molar ratios that form random solid solutions in either a simple face-centered cubic (fcc) or simple body-centered cubic (bcc) crystal lattice structure. Significant suppression of void formation at elevated temperatures has been achieved with increasing compositional complexity in Ni-containing SP-CSAs. In our research, we demonstrated the modification of alloy complexity by increasing the number, the type and the concentration of alloying elements in SP-CSAs. A group of SP-CSAs (Ni, NiCo, NiFe, NiCoFe, NiCoFeCr, NiCoFeCrMn) irradiated by Ni ions at 773 K has been studied by cross-sectional transmission electron microscope (TEM). This study demonstrates the enhancement of radiation tolerance by showing two orders of magnitude of decrease on void swelling with increasing number of alloying elements. The controlling mechanism of defect movements was determined through detailed TEM characterization of defect clusters distributions and Molecular dynamics (MD) simulations. The enhanced swelling resistance is attributed to the tailored interstitial defect cluster motion in the alloys, from a long-range one-dimensional (1-D) mode to a short-range three-dimensional (3-D) mode, which leads to enhanced point defect recombination. The effect of alloying elements on radiation-induced microstructural evolution has been studied in Ni and Ni-20X (X=Fe, Cr, Mn and Pd) binary alloys. The 3-D migration mode is identified to be the dominating migration mechanism for interstitial clusters in these binary alloys, contrary to the 1-D mode dominated in dilute alloys. It is found that the solute atomic volume size factor plays a key role in the migration and interaction of defect clusters. The total void swelling generally decreases as the atomic volume factor increases, accompanying with a significantly sluggish interstitial migration and smaller dislocation loop size. The effects of elemental concentration on radiation tolerance in Ni-Fe alloys have been studied. Void swelling and dislocation loop evolution are both suppressed or delayed with increasing iron concentration. Furthermore, the dominating migration behavior of interstitial clusters shifted from 1-D to 3-D mode with increasing iron concentration. It has been demonstrated that the transition between 1-D and 3-D is a continuous process, and can be quantitatively characterized by the mean free path of the interstitial defect clusters. This talk demonstrates the enhancement of radiation tolerance in SP-CSAs, and more importantly, reveals its controlling mechanism through a detailed analysis of microstructure characterizations and atomistic computer simulations.

Resume : The unit event of radiation-induced defect interaction in structural alloys for nuclear energy applications has a significant impact on the microstructural evolutions and property degradation. In this study, we carry out comparative simulations of defect interaction in bcc Fe and W using atomistic simulations and the self-evolving atomistic kinetic Monte Carlo methods and determine the capture efficiencies and sink strengths in these materials. Since dumbbell is the dominate interstitial defect in bcc iron and crowdion is the primary self-interstitial atom (SIA) in tungsten, the effects of diffusion mechanism of SIA (three dimensional vs. one dimensional) on the defect interaction are examined. The similarities and differences of defect interaction in the two systems will be discussed. This study aims to provide fundamental insights into defect interaction so that strategies might be developed to control the corresponding interaction processes and subsequent microstructural changes. This project is supported by the U.S. Department of Energy, Office of Science, Basic Energy Science Program.

Resume : The advent of High Temperature Superconducting (HTS) magnetic tapes has accelerated the development of compact nuclear fusion reactors. The Rare-Earth Barium Copper Oxides (REBCOs) offer high field strengths that can be accessed at high temperatures (>70 K). During reactor operation high energy neutrons ejected from the plasma will damage the superconducting tapes which may impact their superconducting properties. Experimental observation of the damage process at cryogenic temperatures is difficult without highly specialised facilities that are not currently available. Therefore, here we use molecular dynamics simulations to understand irradiation damage. As a first step to simulating the cascades, we present a new empirical pair potential for YBa2Cu3O7 that combines the Buckingham form with an embedded atom approach. The potential is fitted using thermal expansion coefficients derived from Density Functional Theory (DFT). In this initial study we present cascades in the energy range from 5 keV to 50 keV and consider a number of different directions to account for the anisotropic crystal structure. Cascades are modelled at operational temperatures as well as the temperatures experience in previous experiments to enable a detailed comparison. By comparing the number, type and distribution of defects generated during the cascade we discuss the relevance of the available experimental data to operational conditions. Finally, we examine the intrinsic defect process using DFT.

Resume : It is essential to ensure the integrity of components, structures, and systems in nuclear power plants (NPPs) regarding the protection and the safety of plant-operating personnel and the general public as well as the environment. Especially, reactor containment buildings (RCBs) in NPPs play a role as the final barrier to radioactive material release and provide structural support, and therefore their structural integrity should be maintained over the design lives. Many studies have shown that rebar corrosion is the main degradation mechanism threatening the integrity of RCBs. Also, the diffusion of chloride ions through concrete has been known as a kind of rate-determining step in the rebar corrosion process. For this reason, many efforts have been made to estimate the chloride ion diffusion and to predict the service life reduction of RCBs by the degradation. Recently, NPPs in the MENA (Middle Eastern and North African) countries have been constructed or planned. In this study, considering the atmospheric environmental conditions in the MENA region, chloride diffusion tests using concrete cylinder samples have been performed not only with several constant temperature conditions, but also with temperature gradient along the depth from the surface to simulate the outside temperature conditions. The test results have been incorporated into an analytical model for the RCB life prediction. In this study the experimental and simulation results are presented.

Resume : The Nuclear Science User Facilities (NSUF) is a US Department of Energy Office of Nuclear Energy program that provides the nuclear energy research community access, at no cost to the user, to specialized and often unique capabilities. The NSUF operates as a consortium with Idaho National Laboratory (INL) as the lead and primary laboratory with an additional 20 Partner Facilities and 1 international Affiliate Facility. Current capabilities include a variety of nuclear research and test reactors, ion beam accelerators, hot cell post-irradiation-examination (PIE) equipment, advanced radiologically qualified materials science PIE instrumentation in low activity laboratories, X-ray synchrotron, neutron and positron beam lines, and high-performance computing. The capabilities of the NSUF are directed towards studying irradiation effects in nuclear fuels and materials. Proposals for access to capabilities through the NSUF are accepted through two types of solicitations. The NSUF strives to offer the most complete and state-of-the-art capabilities to the nuclear energy research user community and therefore accepts applications from institutions to become partner or affiliate facilities and is reviewed based on the uniqueness of the capability with respect to the needs of the nuclear energy research community or the capability’s high demand. An overview of some NSUF research will be presented. Further information on the NSUF can be found through the NSUF website at nsuf.inl.gov.

Resume : Understanding the mechanisms of different chemical reactions with actinides at the atomic level is a key step towards safe disposal of nuclear wastes and towards the identification of physical-chemical processes of radionuclides in the environment. This contribution will provide an overview of the recently performed studies on Uranium, Thorium, Plutonium and Cerium contained materials at the Rossendorf Beamline (ROBL) of the European Synchrotron (ESRF) in Grenoble (France). This innovative, recently upgraded, world-wide unique experimental station, funded and operated by HZDR in Dresden (Germany) was used to study actinide systems by several experimental methods: X-ray absorption spectroscopy in high energy resolution fluorescence detection (HERFD) mode, Resonant inelastic X-ray scattering (RIXS) at the An L3 and M4,5 edge and X-ray diffraction (XRD). We will show how the detail information about local and electronic structure of actinide materials can be obtained, including information on the electron-electron interactions, hybridization between molecular orbitals, the occupation and the degree of the f-electron localization. The experimental spectral features have been analysed using a number of theoretical methods based on density functional theory and atomic multiplet theory. It might be of interest for fundamental research in chemistry and physics of actinide systems as well as for the applied science.

Resume : Plutonium is an element of a high importance due to its application as a nuclear fuel. The investigation of the fundamental properties of Pu is crucial to understand the behaviour of Pu compounds in the nuclear fuel cycle especially on the last stages related to the reprocessing and the waste disposal. It was previously shown that plutonium migrates in colloidal form in the subsurface environment with the distance of several kilometers from previously contaminated sites. However, the certain structure and stoichiometry of these colloids, as well as Pu oxidation states present is still debated. This contribution will show the results of plutonium oxide nanoparticles studies at the large-scale facility – The European Synchrotron (ESRF) by complementary methods: X-ray diffraction (XRD), atomic pair distribution function analysis (PDF), several types of spectroscopies: high energy resolution fluorescence detection (HERFD) at L3 and M5-edges, X-ray emission spectroscopy (XES) and extended X-ray absorption fine structure (EXAFS) spectroscopy. The applying multifold synchrotron methods benefits to discover features, which may be unclear or even indistinguishable, these approach is also crucial to confirm results, obtained with individual methods.

Resume : X-ray Absorption Spectroscopy (XAS) is an invaluable tool in nuclear material research, allowing to probe the oxidation state and the local coordination of a selected atomic species. The first part of the spectrum, the X-ray Near Edge Structure (XANES), is less exploited than the Extended X-ray Absorption Fine Structure (EXAFS). However XANES conceals a wealth of information on the electronic structure and the local geometry around the absorber. Nowadays important progresses in the interpretation of XANES have been made thanks to i) the development of dedicated ab-initio codes using powerful computational resources and ii) the use of high resolution XANES, which is especially advantageous for actinides. We present here an example of how to extracted valuable information from XANES. We performed a systematic study of U L3 edge XANES for U5+ and U6+ in different local coordination. It is well known that the presence uranyl bonds gives a characteristic feature in the post-edge of U L3 XANES and it has been observed experimentally that this feature shifts to lower energy when going from the uranyl to the uranate coordination. We found that in U6+ and U5+ mixed systems the U5+ in octahedral coordination gives a characteristic shoulder just before the uranyl feature due to the splitting of the 6d DOS from the octahedral crystal field. We think that the shift of the uranyl feature going to uranate configuration indeed points to the presence of U5+.

Resume : X-ray absorption spectroscopy (XAS) provides a unique and sensitive probe of element speciation and local environment in materials relevant to the nuclear fuel cycle and security. Hitherto, this technique has primarily required access to a synchrotron radiation facility as a broadband X-ray source of high brilliance, which has limited application for routine and high throughput studies. This of particular challenge for the field of nuclear materials where scientific opportunity may be constrained by the absence of an accessible synchrotron source or sample containment requirements and inventory limits. Here, we report our exploitation of a newly available commercial XAS spectrometer, utilising spherically bent crystal analysers to acquire XAS data in the range 5 – 18 keV resolution from actinide, nuclear and radiological materials. We show that XAS data may be acquired in a few hours, or less in favourable circumstances, from moderately dilute to concentrated absorbers to address routine questions of element speciation and co-ordination of particular relevance to radioactive waste immobilisation. These data and analyses are compared with counterpart synchrotron studies, on an identical sample suite, to highlight both the potential opportunities and limitations of laboratory XAS, and feasibility of routine application.

Resume : Molten-salt reactors have exceptional promise as future sources of nuclear energy. Halide salts have superior chemical stability, allowing them to operate at low pressures while achieving high operational temperatures. Actinides can be added to and withdrawn from the fuel during operation, and fission products can be removed during operation, improving stability, neutron economy, and load-following capability. However, the advantage of chemical stability of molten salts presents a great challenge for material selection. Most metallic alloys are unsuitable for use in contact with fluoride salt because they will oxidize to fluorides themselves. A family of high-nickel alloys were developed that were far less susceptible to oxidative attack from fluorides, but improvements will be necessary for long-term operation. The intense radiation environment inside the reactor precludes the consideration of almost all materials except graphite, which also faces development challenges. The chemical processing technologies needed for MSRs also challenge materials, since techniques like fluorination can be quite aggressive. This presentation will review the materials challenge of MSRs, both in the design of the reactor, heat exchangers, and pumps, as well as in the chemical processing equipment. It will describe research improvements presently under investigation as well as system design approaches that can enhance the probability of successful deployment of this promising technology.

Resume : Kinetic rate theory is a mature method that has long been used to model fission gas behaviors in nuclear fuels. However, uncertainties remaining in the key parameters of the kinetic rate theory models often lead to doubts in the accuracy of this method. Well-designed separate effect experiments are believed to be able to assist in reducing the uncertainties of such parameters. In this work, the results of an in situ Xe ion implantation experiment at the IVEM facility were interpreted with a kinetic rate theory model. The complexity of the mathematical model is significantly reduced according to some key experimental information. By fitting the calculation results to the experimentally measured Xe bubble size distributions, a set of parameters including Xe diffusivity, gas bubble resolution factor and gas bubble nucleation factor was obtained. A parametric study was performed to gauge the sensitivity of the calculation results to the values of these parameters. It was shown that the calculated bubble size distribution is highly sensitive to the Xe diffusivity, the gas bubble resolution factor and the gas bubble nucleation factor. Molecular dynamics simulations were also performed to provide estimates of the irradiation enhanced diffusivity of Xe and the gas bubble resolution factor associated with the experimental setup. It was demonstrated that the MD calculated parameters are close to the parameters derived by fitting rate theory results to experimental data.

Resume : YSZ, a potential material for inert matrix fuels, with different grain sizes (tens of nanometers to few microns) was irradiated under different conditions (single beam irradiation with high energy (Se>>Sn) ions at 300 K and 1000 K & simultaneous dual beam irradiation with high and low energy (Sn>>Se) ions at 300 K) to investigate the effect of grain size, irradiation temperature and electronic excitation (Se)/ballistic processes (Sn) on the radiation damage. The low and high energy ions were chosen to simulate the damage produced by alpha recoils and fission fragments respectively. The irradiations at 1000 K and the dual beam irradiations helped to better simulate typical nuclear reactor environment. In case of the single beam irradiations, (i) the nanocrystalline samples were more damaged compared to the micro-crystalline sample irrespective of the irradiation temperature and (ii) the damage for all grain sizes was found to be reduced at 1000 K compared to that at 300 K, which are different from results obtained previously with low energy irradiations. Interestingly, this damage reduction was significantly more for the nanocrystalline samples as compared to the microcrystalline one. For the simultaneous dual beam irradiations, the nanocrystalline sample was less damaged than its micro-crystalline counterpart. The interplay between Se/Sn, grain size and temperature in governing the overall irradiation induced damage will be discussed to explain the observed results. References S. Dey et al., Scientific Reports, 5, 7746 (2015) A. Debelle et al., Journal of Applied Physics, 115, 183504 (2014) P. Kalita et al., Journal of Applied Physics, 122, 025902 (2017)

Resume : Sodium Fast Reactor (SFR) is one of the reference reactor among generation IV nuclear systems. The investigation of severe accidents is essential for the design of SFR reactors. In fact, during a severe accident, the MOX fuel, steel cladding and B4C neutronic absorbers could melt forming a complex mixture of gas, liquid and solid phases called corium. A thermodynamic study of the (U-Pu-O)-(Fe-Cr-Ni)-(B-C) system, representative of a prototypic corium composition, is proposed using the CALPHAD method. The aim is to develop a thermodynamic database as a computational tool to perform thermodynamic calculations on corium for the SFR reactor. It is first necessary to study the most relevant sub-systems. This work is the first effort to fill the lack of thermodynamic data on SFR reactor corium. This research targets on the one hand the investigation of chemical interactions between MOX fuel and steel cladding in case of a steep temperature rise. To mimic fuel/cladding contacts of a reactor, a mix of steel and UO2 powder is heated in a sealed crucible then analysed by SEM and electron microprobe to identify the phases that form. In a second part, the interaction between steel cladding and B4C absorber is studied. In fact, inconsistencies between available models and experimental data exist for the B-Cr-Fe ternary system. A check of the phase transitions will allow a better thermodynamic description of this key system.

Resume : Dissolution or leaching of the spent nuclear fuels (SNF) is a key step either in the field of their reprocessing or their long-term storage in underground repository. Moreover, SNF contain a wide variety of fission products including platinoid elements (PGM’s) either incorporated in the UO2 matrix, or present in various separated phases. Their specific impact on the overall dissolution kinetics was never undoubtly discriminated. In order to answer this question, several samples doped with 0.6 to 3 mol.% of PGM’s (Ru, Rh, Pd) were prepared from hydroxide precursors [1,2]. After conversion then sintering, the prepared pellets were submitted to multiparametric dissolution tests in various nitric acid solutions (0.1 - 4 M HNO3) and temperatures (25°C - 60°C). The macroscopic description of the dissolution showed that the normalized dissolution rates were significantly increased for UO2 doped with PGM’s compared to pure UO2 used as reference compound. This effect was strengthened in less acid media (as instance, a factor of 4500 was observed after 175 days of leaching in 0.1M HNO3). Simultaneously, the dissolution of the pellets was followed operando by ESEM. The combination of macroscopic and microscopic approaches confirmed the modification of the preponderant mechanism occurring at the solid/liquid interface from redox-controlled dissolution in strong nitric acid media to surface-controlled dissolution for less acidic media. [1] Martinez, J. et al., J. Nucl. Mater., 462, 173-181, 2015 [2] Martinez, J. et al., J. Europ. Ceram. Soc., 35, 4535-4546, 2015

Resume : Models of thermodynamic properties and phase transitions of some materials are necessary for analysis and numerical simulations of processes in nuclear reactors under normal and abnormal mode conditions. In the present work, an equation-of-state model for lithium fluoride is presented with taking into account melting and evaporation effects. Thermodynamic characteristics of the material are calculated, and obtained results are compared with available experimental data and theoretical predictions at high temperatures and high pressures. Proposed equation of state can be used efficiently in simulations of different high-temperature, high-pressure processes in molten-salt reactors.

Resume : The objective of this work is to characterize the radial evolution of fission products and microstructure in plutonium bearing mixed oxide (MOX) fuels. Ruthenium, rhodium, technetium, molybdenum, and palladium aggregate to form an insoluble fission product phase in oxide nuclear fuels. To understand the influence these fission product phases have on the thermophysical properties of the fuel, the behavior of these species must be analyzed at high burnups. Palladium has a significantly higher fission yield in Pu-239 compared to U-235, causing the precipitation of a secondary palladium rich phase that is scarcely reported in literature. This work will examine the palladium rich precipitates as well as the other solid fission product phases to evaluate the effects of the thermal gradient on fuel evolution. Scanning electron microscopy, transmission electron microscopy, and electron probe microanalysis were used to characterize the radial profile of metallic fission products observed in irradiated fast reactor MOX fuel pellets. The palladium rich phase forms in the cooler regions on the fuel pellet periphery, commonly nucleating on other metallic fission products. Palladium forms an alloy with tellurium in the secondary metallic phase, potentially mitigating the formation of tellurium containing intermetallics in the steel cladding.

Resume : Sodium-cooled Fast Reactor (SFR) is one of the six technologies that has been chosen as next generation nuclear energy system within the Generation IV International Forum. Uranium and plutonium mixed oxide (MOX) fuels are considered the reference fuel for SFR, thanks to their good stability under irradiation. Operating conditions in SFR are characterized by high linear heat generation rates (>30 kW/m), resulting in fuel centerline temperatures reaching 2000°C. At those temperatures, in addition to swelling and fuel restructuring, significant migration of fission products and fuel components occurs, resulting in fuel-cladding chemical interaction (FCCI). In particular, when the burnup exceeds 7-8% fission of initial heavy atom (FIMA), formation of a mixture of fission products known as “joint oxyde-gaine” (JOG) has been reported. As the FCCI region characteristics influence temperature profiles, knowledge of its composition and microstructure is fundamental to conduct reliable design and performance analysis. Limited amount of data is available regarding the FCCI and its composition is not well known. In this work, we present novel Scanning and Transmission Electron Microscopy results of the FCCI and JOG characteristics in annular MOX fuels irradiated in the Fast Flux Test Facility (FFTF) at burnups between 7 and 20% FIMA.

Resume : In order to enhance the dissolution of oxides hard to dissolve, different options can be considered: either the use of more aggressive dissolution media and/or the increase of surface reactivity of the solids to dissolve. In the latter case, mechanical activation through high-energy milling with the aim of changing microstructure and physicochemical proprieties by mechanical actions such as impacts, frictions or collisions was studied. Ceria was chosen as starting material of this work not only as surrogate of PuO2 but also due to its isotropic cubic structure (fluorite) ensuring a random distribution of the defects generated by milling. Dissolution of several ceria samples in HNO3 8.5M at 95°C were carried out showing an increase of dissolution rate by a factor 400 after 6h that cannot be only explained by an increase of the specific surface area of the oxide powder. Scanning transmission electron microscopy observations on milled samples enable establishing a double fragmentation mechanism. First a debonding of the crystallites constituting the elementary powder particles, then a break of these crystallites by a cleavage effect leading to the generation of dislocations on bigger fragments and the apparition of two particle populations consisting of nanoparticles and the initial crystallites with structural defects. Both generated nanoparticle and structural defects could constitute preferential dissolution sites explaining dissolution enhancement.

Resume : A study has been made of the extraction behavior of N,N-di(2-ethylhexyl) diglycolamic acid (HDEHDGA) for actinide ions, An(IV, VI), and Ln(III) from HNO3 solutions. Different mechanisms are confirmed for the extracted complexes from nitric acid solutions of varying concentration. The N,N-dialkyl-diglycolamic acid has similar tridentate coordination functional group as the well-studied N,N,N’N’-tetraalkyl-diglycolamide, TRDGAs, but it is a carboxylic acid with one of the two amide groups of TRDGA replaced by carboxy group. The results showed that the ligand acts as normal carboxylic acid in the system of low HNO3 concentration and the extraction of actinides and lanthanides is governed by cation-exchange mechanism, while with high nitric acid concentration the ligand forms a cationic complex with actinides and lanthanides, and the extraction goes through ion-pair mechanism.

Resume : Unsymmetrical N,N’-dimethyl-N,N’-dioctyldiglycolamide (DMDODGA) has much better extraction ability for actinide ions than its symmetrical analogues, such as N,N,N’N’-tetraoctyldiglycolamide (TODGA), especially for actinyl ions due to the less steric hindrance from the smallest methyl groups in the extracted complexes. The complexation of U(VI) with DMDODGA and its small molecule analogue tetramethyl-diglycolamide (TMDGA) was studied with spectrophotometry, X-ray crystallography, and distribution ratio slope analysis. Two successive complex species, UO2(L)2+ and UO2(L)22+ (L = TMDGA or DMDODGA) were identified and their stability constants were calculated at 25℃ in 1.0 mol/L NaNO3 and in 1-octonol respectively. UO2(L)2(ClO4)2 (L=TMDGA) is crystallized in a monoclinic space group, P2(1)/n. The two TMDGA ligands are coplanar and each coordinates to U(VI) with three oxygen atoms in the equatorial plane. The molecule is centrosymmetric and the uranium atom is at the inversion center. The structure clearly indicates that alkly groups bigger than methyl will increase the steric hindrance then weaken the bonding between actinyl ions and the ligands due to the limited space in the equatorial plane of the ions. The molar absorption spectra of UO2(L)22+ (L = TMDGA) in the aqueous solution and that of UO2(L)22+ (L = DMDODGA) in 1-octanol deconvoluted from spectrophotometric titrations using Hyperquad program were compared to the spectra of the extracted samples of U(VI) with DMDODGA in 1-octanol. The similarity in the spectra suggests that U(VI) in the extracted samples is also coordinated by two DMDODGA molecule, which is consistent with the results from the slope analysis of the log(D)-log(CL) plot of solvent extraction experiments.

Resume : Spent nuclear fuel direct disposal is not the reference scenario in France, however it is studied as an optional scenario. In this context the main concern is to determine the quantity of radionuclides that would be released from the spent fuel in case of disposal canister breaching1. This source term depends on the aged spent fuel microstructure and on the radionuclides localization. Alpha decay of some of them produces helium. After long time disposal it precipitates in the form of bubbles2 which could affect the fuel microstructure and so the source term of labile radioactivity. Helium precipitation study in UO2 is thus of primary importance and constitutes the main objective of this work. Separated effects studies, coupling ion irradiations/implantations and fine characterizations, have been established to address this objective. TEM observations highlighted the presence of bubbles in the material but also platelets for the first time in UO2. A complete study about helium platelets in this material has been done to identify the conditions for their appearance. In the same time, it has been possible to determine their characteristics (size and habit plane) as a function of different parameters such as implantation conditions or grain orientation. 1 C. Ferry et al., J. Nucl. Mater., 2010 2 Z. Talip et al., J. Nucl. Mater., 2014

Resume : Nuclear fuel can be assessed non-destructively with axial gamma scanning devices and gamma emission tomography (GET), where gamma-ray spectrometry technique is used. In this technique, gamma rays emitting from unstable isotopes produced with the fission of nuclear fuel are utilized. However, self-attenuation of gamma rays in nuclear fuel is inevitable and require correction procedures in quantitative measurements of nuclide concentrations. In a recent study, the energy dependent mass attenuation of gamma rays was evaluated with burnup of nuclear fuel using Serpent 2 reactor physics code and NIST database and found to be non-negligible. This paved the way for a model for predicting mass attenuation with burnup and initial plutonium content in conventional UOX and MOX fuels. Also, a complete model for the evolution of the linear attenuation of gamma rays in nuclear fuel with burnup was assessed by taking into consideration a theoretical change in the density of nuclear fuels due to swelling. In this study, we investigate the degradation of the gamma ray attenuation in Fast Reactor (FR) fuel. Considering that compared to the UOX fuel of Light Water Reactors (LWR), FR fuels may be 1) reaching higher burnups, 2) have a higher actinide concentration and 3) have higher degree of irradiation swelling, we consider it likely that the effect will be enhanced. Therefore, the gamma ray attenuation of high-burnup fuel has been examined for FR fuels. Then, an attenuation model is also constructed for FR fuels in order to be used by the spectroscopy practitioners doing research in Gen-IV reactors. In conclusion, a more accurate data will be obtained on the material properties of the proposed nuclear fuels after irradiation with corrected gamma measurements.

Resume : In-pile irradiation testing is routinely performed on nuclear fuels and materials. For qualification and licensing of Accident Tolerant Fuels or Gen IV reactor fuels, extensive irradiation testing is foreseen in order to fill the gaps of existing validation data, both in normal operational conditions and in order to identify operational limits. Gamma Emission Tomography (GET) has been demonstrated as a viable technique for studies of the behavior of irradiated nuclear fuel, e.g. measurement of fission gas release and inspection of fuel behavior under LOCA conditions. In this work, the aim is to improve the technique of GET for irradiated nuclear fuel by developing a detector concept for an improved tomography system that allows for a higher spatial resolution and/or faster interrogation. This paper covers a review of the literature for gamma-emission tomography devices for nuclear fuel characterization. The detector concepts have been described and quantitatively compared using relevant figures of merit. In addition, we present the working principles of a novel concept for gamma emission tomography using a segmented HPGe detector. The performance of this concept was investigated using the Monte Carlo particle transport codes MCNP and SERPENT. We concluded that the segmented HPGe detector has an advantageous performance based on the evaluated figures of merit, and that its application in tomography of nuclear fuel may facilitate assessment of irradiated nuclear fuel.

Resume : At ISOLDE-CERN and other facilities around the globe, porous uranium carbide is the most used material to produce through nuclear reactions, radioactive isotopes by irradiating it with high-energy particles. These targets are maintained at high temperatures (>2000 °C) during irradiation, to promote the diffusion of isotopes produced in the bulk of the target material. These isotopes diffuse out of the target material to an ion source, where they are ionized and delivered for experiments on nuclear, astro, medical, bio, solid-state and atomic physics. This process is online, which means that while the target is continuously irradiated, the isotopes are extracted and delivered, allowing ISOLDE to deliver very short lived isotopes (down to few ms of half-life). Micrometric sized porous uranium carbide is produced in-house at ISOLDE from uranium oxide mixed with excess graphite, following the reaction: UO2 + 6C -> UC(2-x) + (2+x)C + 2CO (g) up to 2000 °C in vacuum. In order to reduce the diffusion distances of the isotopes in the grains of uranium carbide, a nanometric uranium carbide was developed, which does not sinter at ISOLDE operation temperatures. The standard UO2 was milled to about 160 nm and mixed with multiwall carbon nanotubes (MWCNT) instead of graphite as a carbon source and heated to form uranium carbide. The resulting material was >70% porous and the UCx particles are suspended in a backbone created by the MWCNT halting the sintering. This target was tested successfully at ISOLDE and overall 10-fold was gained in isotope beam yields.

Resume : Purification of rare earth elements is challenging due to their chemical similarities. The interaction and behavior of alloys involving stannous chloride (SnCl2) and lanthanide metals are discussed in the present paper. We investigate the quantitative relationship between the deposition potential of lanthanides with SnCl2 and atomic radius by employing electrochemical techniques, involving cyclic voltammetry (CV), square wave voltammetry (SWV) and open-circuit chronopotentiometry (OCP). Our electrochemical study on the formation intermetallic compounds is based on Sn in LiCl-KCl melts on molybdenum electrodes at 873 K. With the same experimental conditions, different deposits (e.g., Sn-La, Sn-Pr, Sn-Gd, Sn-Dy and Sn-Er) were obtained by using identical substrates. We establish the relationship between the deposition potential and the atomic radii of lanthanides by deriving a mathematical equation from the sorting out and summarizing of the data. The predictions for the existence and the deposition potentials of unknown intermediate phases (e.g., Sn-Ce, Sn-Nd, Sn-Yb and Sn-Lu) were made. From our results, open-circuit chronopotentiometry is potentially a valuable methodology to formally verify the correctness of the forecast. X-ray diffraction pattern (XRD), scanning electron micrograph (SEM) and energy-dispersive spectrometry (EDS) data further verify the reliability of the linear equation.(This paper is funded by the International Exchange Program of Harbin Engineering University for Innovation-oriented Talents Cultivation)

Resume : An innovative molten salt reactor concept, the MSFR (Molten Salt Fast Reactor) is developed by CNRS (France) since 2004. Based on the particularity of using a liquid fuel, this concept is derived from the American molten salt reactors (included the demonstrator MSRE) developed in the 1960s. In MSFR, the ORNL (Oak Ridge National Laboratory) MSBR concept has been revisited by removing graphite and BeF2. The neutron spectrum is fast and the reprocessing rate strongly reduced down to 40 liters per day (compared to 4000l/day in the MSBR concept) to get a positive breeding gain. The reactor is started with 233U or with a Pu and minor actinides (MA) mixture from PWR spent fuel. The MA consumption with burn-up demonstrates the burner capability of MSFR. The structural materials retained for MSR container are Ni-based alloys with a low concentration of Cr. The composition of Hastelloy N (Ni-Mo-Cr system) optimized by ORNL researchers is already a good candidate for temperature up to 750°C. Nevertheless, the chemical reactivity of the material increases with time due to the increase of the redox potential of the salt with the fission reaction. This paper addresses the issue of structural materials considering its chemical compatibility with the fuel salt. Based on thermodynamic calculations and experimental results, the influence of the redox potential on the corrosion rate is clearly shown. Methodology is proposed in order to measure the redox potential and to control it.

Resume : U- 10 wt.% Zr alloy including rare-earth(RE) elements was introduced as a surrogate of transuranium alloys. REs in the metallic fuel derive negative effects such as strong reactivity during casting, FCCI, and distribution of other contents in the fuel. Graphite was chosen as a crucible material according to the cheap and conductive characteristics. Enhanced Y2O3 coating layer was introduced as reaction prevention layer to reduce the interaction during casting. Improved degassing processes were introduced to reduce interaction as well. Vacuum pressure of the chamber was decreased from 0.03 to 5 x 10^(-5) Torr through a diffusion pump. The degassing time was increased from 5 to 60 min to improve the degassing effect. The degassing temperature was decreased from 500 to 200°C to decrease the oxidation of the rare-earth elements before the degassing procedure. The improved degassing procedure reduced humidity and oxide levels of the crucible and chamber, which are easily trapped inside of the crucible and chamber during storage. The coating layer and melt residue were investigated using scanning electron microscopy (SEM). The compositions of the specimens were characterized using energy dispersive spectroscopy (EDS). The improved degassing process and coating layer reduced the interaction between the crucible and melt, and the coating layer remained successfully on the crucible after casting.

Resume : A potential long-term solution for disposing of nuclear waste is to incorporate the material within borosilicate glass and store this within a multi-barrier repository deep underground. For the regulatory safety case, there needs to be confidence that the initially contained radionuclides will not be released into the environment in any significant quantity. Consequently, accurate models are needed to predict glass dissolution behaviour in the presence of groundwater. Such models require robust kinetic parameter selection to minimise the discrepancy between simulated and experimental data. Currently, this is challenging given the many different glass component species and wide range of potential experimental conditions to consider. Such complexity can lead to multiple competing objective functions, and this issue has not been widely addressed within the current glass dissolution literature. Here, results found after applying multi-objective optimisation to borosilicate glass dissolution are presented. Several analytical and numerical models have been used, with both French and UK simulant glasses considered under static and dynamic leaching conditions. Coupled MATLAB-PHREEQC evolutionary optimisation algorithms are applied in conjunction with level diagrams. The results demonstrate a method of appropriately choosing kinetic model parameters in glass dissolution, applicable to nuclear waste disposal, that has the potential to utilise data accumulated over many campaigns.

Resume : Commercial reactors spent nuclear fuel (SNF) reprocessing produces high-level radioactive waste with high residual amount of americium, mainly 241Am which exceeds the amount of 239Pu by more than 10,000 times [1]. As a result, in the USA refused SNF reprocessing is not applied but in France the methods of americium as well as curium separation as a DIAMEX or EXAM process are under development now [2]. The heat release of 241Am and 244Cm are 0.111 and 2.78 W/g, respectively, results in strong decomposition of extractants yielding low efficiency of Am separation and from 1 to 5% of total Am will remain in HLW. In this case the residual HLW (after Am and Cm) separation must be incorporated in glass-ceramics rather than glass as at present [4]. Taking into account Russian experience when HLW is incorporated in sodium-alumino-phosphate glass the sodium aluminophosphate glass-ceramics may be considered as the suitable host phase for the Am-bearing residues. This glass-ceramics is composed of the Am-bearing monazite structure (Ln,Al)PO4 and sodium-iron orthophosphate – Na3Fe2(PO4)3 with minor Al crystalline phases distributed in the interstitial glass. Elemental leach rates of the major elements (Na, Al, Fe, P) determined by Russian standard were found to be 10-5-10-7 g/(cm2d), for REEs – lower 10-5 г/(см2сут) [5]. An alternative matrix is murataite based ceramics or glass-ceramics composed of the zoned structure murataite grains with maximum actinide concentrations in the core of the grains that strongly reduces leaching of the actinide elements. Incorporation of up to 10 wt.% actinide oxides in the murataite structure at excellent chemical durability has been proven [7]. 1. Babaev N.S. et al. At. Energy. 98 (2005) 115. 2. C. Poinssot et al. Procedia Chemistry. 21 (2016) 524-529. 3. V. Vanel et al. Procedia Chemistry. 21 (2016) 190-197. 4. R. Didierlaurent et al. WM2016 Conference, March 6 – 10, 2016, Phoenix, Arizona, USA. 16376. 5. S.V. Stefanovsky et al. J. Nucl. Mater. 500 (2018) 153-165. A.A. Lizin et al. J. Radioanalyt. Nucl. Chem. 318 (2018) 2363-2372.

Resume : Lliquid outflows contaminated with radionuclides, either coming from spent reprocessing fuel plants or from accidental cases such as Fukushima Daiichi, have to be treated. Radionuclides soprtion on solid substrate is very attractive due to the immobilization of radiative compounds on a solid phase that can be added in a second step in conditioning matrices such as glass, cement or bitumen. However, since a couple of years alternative processes avoiding the second step are under investigation, especially for volatile radionuclides such as Cesium or Iodine. One of these alternative processes is to consider the substrate used for the decontamination step as a conditioning matrice through quite simple treatment. These last years, the ICSM has developed several kinds of hierarchical porous silica, functionalized in order to selectively uptake radionuclides with for example the selective sorption of Cs towards Na. The benefits of confining radionuclides in pores comes from the fact that these substrates can be densified quite easily (moderate heating, pressure or irradiation). This talk will give you an overview of the synthesis routes developed to design these materials (post functionnalization of existing porous glass beads, emulsion templated monoliths, nanoparticles…) and the first studies undertaken to densify these porous materials and make it promising solutions for « all-in-one » separation/conditioning process.

Resume : The original interest in plutonium and minor actinides (MA) at the commencement of the nuclear era was colossal [1]. But 70 years later, the knowledge on the chemistry of these elements is still incomplete and sometimes contains diverging information. The limitations concerning nuclear proliferation and radiation protection restrict their handling in a few sites all over the world. Nonetheless, recent challenges related with, e.g., partitioning and transmutation, fuelling fast reactors, nuclear waste management or specific use as thermoelectric generators for deep space missions, all require an in-depth knowledge on the chemistry of Pu and MA. We present here our recent findings on the synthesis, characterisation and stability of AnPO4 and AnAlO3 (An= Pu and Am). These compounds are prepared by a solid state reaction method and their crystal structure at room temperature is solved by powder X-ray diffraction combined with Rietveld refinement [2,3]. We clarify the discrepancy about the crystal structure of PuAlO3 at room temperature, showing that both existing reports based on direct determination [4] and DFT modelling [5] are inaccurate. The thermal stability of AnPO4 and AnAlO3 is studied under oxidising and inert atmospheres, indicating that (at least in these specific cases) the Am-bearing compounds are more stable than the Pu-ones [3,6]. PuPO4 forms solid solutions with LaPO4, with direct benefit for the conditioning of highly irradiated plutonium in monazite-like ceramic waste forms [7]. AmPO4 and AmAlO3 are studied by our group as alternative forms to AmO2 for space applications. Their behaviour under alpha self-irradiation is monitored through XRD, MAS-NMR and Raman [3]. Both compounds show a fast, significant swelling, followed by total amorphisation. The advantages and drawbacks of AmPO4 and AmAlO3 with respect of their thermal and radiation stability will be discussed relative to the americium oxides. References [1] Morss, L.R.; Edelstein N.; Fuger, J.; Katz, J.J. (eds.), The Chemistry of the Actinide and Transactinide Elements, Springer, 2011. [2] Popa, K.; Raison, P.E.; Martel, L.; Martin, P.M.; Prieur, D.; Solari, P.L.; Bouëxière, D.; Konings, R.J.M.; Somers, J. J. Solid State Chem. 2015, 230, 169. [3] Popa, K; Vigier, J.-F. et al., unpublished results. [4] Russell, L.E.; Harrison, J.D.L.; Brett, N.H. J. Nucl. Mater. 1960, 2, 310. [5] Fullarton, M.L.; Qin, M.J.; Robinson, M.; Marks, N.A.; King, D.J.M.; Kuo, E.Y.; Lumpkin, G.R.; Middleburgh, S.C. J.Mater.Chem. A 2013, 1, 14633. [6] Jardin, R.; Pavel, C.C.; Raison, R.E.; Bouëxière, D.; Santa-Cruz, H.; Konings, R.J.M.; Popa, K. J. Nucl. Mater. 2008, 387, 167. [7] Arinicheva, Y.; Popa, K.; Scheinost, A.C.; Rossberg, A.; Dieste-Blanco, O.; Raison, P.; Cambriani, A.; Somers, J.; Bosbach, D.; Neumeier, S. J. Nucl. Mater. 2017, 493, 404.

Resume : Monazite-type ceramic wasteforms are promising candidates for the specific conditioning of trivalent and tetravalent actinides. In this frame, thorium incorporation, resulting in monazite-cheralite solid solution Nd1-2xThxCaxPO4 (x = 0-0.1) was successfully achieved through the precipitation of the single phase hydrated rhabdophane (Nd1-2xThxCaxPO4·nH2O). The synthesis procedure was optimized thanks to a multiparametric study (starting stoichiometry, temperature, heating time). The excess of calcium appeared to be a prevailing factor with a suggested initial Ca:Th mole ratio of 10:1. Similarly, the recommended heating time exceeded 4 days while the optimal temperature of synthesis was 110 °C. The thermal behavior of rhabdophane-type precursors was investigated through the monitoring of in situ HT-PXRD patterns and variation of unit cell parameters. The complete dehydration step took place around 200°C while the phase transition from rhabdophane to cheralite occurred at higher temperatures (650-850°C). The resulting product after heating at 1100°C was single phase. Results of TGA and dilatometry confirmed the increase of the temperatures of dehydration and phase transition with the thorium content. Finally, the direct sintering was performed on Th-rhabdophane to prepare fully densified monazite-cheralite ceramics (i.e. > 95% TD). First sintering maps were established to optimize the sintering procedure.

Resume : Actinides, and mainly plutonium, are the main contributors to the long-term radiotoxicity of spent nuclear fuel. In such conditions, representative for geological conditions, the interactions between these radioelements and silicate species could influence the mobility of the actinides in the environment and thus could affect the safety of the storage facilities. Specifically, the formation of actinide silicate, AnSiO4, has to be taken into account, since thorium and uranium silicate are ubiquitous in environmental silicate rich media and under reductive conditions. However, even if PuSiO4 was synthesized once [1] and if the formation of silicate based compounds was suspected for Pu-containing precipitates, the conditions which allowed the formation of this phase always remain largely unknown. In order to provide more information on the plutonium-silicate system, comprehensive studies have been developed on three surrogate systems, CeSiO4 [2], ThSiO4 [3,4] and USiO4 [5] which crystallize with the same zircon-type structure (space group I41/amd). This study allowed to propose an efficient way of synthesizing PuSiO4, provided key information on the formation of silicate based phases and enabled to identify the differences in terms of reactivity between plutonium and surrogate elements. [1] C. Keller, Nukleonik, 5, 41, 1963 [2] P. Estevenon et al., Inorg. Chem., to be published [3] P. Estevenon et al., Inorg. Chem., 57, 9393, 2018 [4] P. Estevenon et al., Inorg. Chem., 57, 12398, 2018 [5] A. Mesbah et al., Inorg. Chem., 54, 6687, 2015

Resume : Naturally-occurred murataite is a rare cubic structure phase with space group F͞43m, lattice parameter a = 14.9 Å, Z=4, and formula with 22 cations and 43 anions [8]R6[6]М112[5]М24[4]ТХ43, where R = Y, Na, Са, Мn; М1 = Ti, Nb, Na; М2 = Zn, Fe, Ti, Na; T = Zn; X = O, F, OH [1]. In synthetic samples the varieties with three- (like in natural murataite – M3 or 3C; C means cubic) five- (M5 or 5C), seven- (M7 or 7C), and eight-fold (M8 or 8C) elementary fluorite unit cell have been also found. Pyrochlore R2M12X7 and murataite are suggested to form a polysomatic series [2], i.e. the structure of 5C, 7C and 8C phases consists of pyrochlore (two-fold elementary fluorite unit cell – 2C) and murataite (3C) blocks (modules). Later it was confirmed by X-ray structural analysis of single crystals of murataite varieties [3-5]. The four-, five-, and six-coordinated sites are capable to incorporate small Ti, Al, and Fe and other corrosion product cations, the larger sites with CN = 7 and 8 are occupied with Ca, Mn, Zr, REE and An cations. During melt crystallization the firstly segregated polytype is the phase with the highest content of the pyrochlore modules. Thus, the content of Ans, REEs, and Zr reduces in the row: M7 – M5 – M8 – M3 while the Ti, Fe, and Al concentrations increase. The highest concentrations of the An, REE, and Zr located in the core of the grains yield their lowest release from ceramics [6]. 1. T.S. Ercit, F.C. Hawthorne. Canad. Miner. 33 (1995) 1223-1229. 2. V.S. Urusov et al. Crystallogr. Repts. 52 (2007) 37–46. 3. S.V. Krivovichev et al. Minerals as Advanced Materials II. S. Krivovichev (Ed.). Berlin, Heidelberg: Springer-Verlag, 2012, pp. 293–304. 4. A.S. Pakhomova et al. Zeitschrift für Kristal. 228 (2013) 151–156. 5. A.S. Pakhomova et al. Eur. J. Mineral. 28 (2016) 205–214. 6. S.V. Stefanovsky et al. J. Alloys Compds. 444-445 (2007) 618–620.

Resume : Actinides (An) are essential materials for nuclear power plants, but their long-term storage safety has always been a serious concern. Even a simple issue such as oxidation becomes a problem in the storage of radioactive materials. A thorough understanding and control of the reactions and the phases of actinide compounds is therefore very important. It is crucial to establish the spectroscopic signatures of different oxidation states of An in various crystal symmetries. The conversion of U(VI) released into the environment may cause a creation of the U(IV) species in the form of nanoparticles (NPs). For example, the abiotic reduction of U(VI) by green rust or the microbial activity can produce the UO2 nanocrystals. The created U(IV) nano-species might be relatively mobile in the environment. Furthermore, the possibility of formation of the UO2 nanostructure in the fuel during irradiation at high-burn-up has been also identified. The knowledge about UO2 NPs and their reactivity is scarce compared to NPs of semiconductors and metals and therefore requires further research. X-ray spectroscopy turns to be an efficient tool for that. The x-ray absorption spectra (XAS) at An shallow edges can be obtained using electron-energy-loss spectroscopy and can be used for quantitative analysis of An species. This facilitates in-house experiments and helps to avoid additional safety problems and restrictions in case of measurements at synchrotron radiation facilities. However, the interpretation of the data, such as spectra at the actinide 5d edges, is hampered by so-called giant resonances governed by autoionization processes, thus requiring the model calculations. We applied an advanced technique such as HERFD-XAS at An L3, M4,5 edges [1-3] as well as An 5d XAS to study the electronic structure in Th, U, Am, and Cm compounds. The HERFD-XAS measurements were performed at ID26, ESRF and 5d XAS were recorded at beamline I511, MAX-lab and beamline 7.0, ALS. The experiments were supported the Anderson impurity model (AIM) and DFT+U calculations. The following obtained results will be discussed: - signatures of U(V) and U(VI) in different environments/crystal symmetries were established for the U M4 HERFD-XAS and enforced by model calculations; - while the U L3 and M4 HERFD-XAS of the UO2 NPs revealed the presence of U(V) as the surface contribution, the spectra of the ThO2 NPs showed a clear change in the local symmetry of Th(IV) on the surface which was analyzed by AIM and DFT+U calculations; - while the 5d XAS of early An (example of ThO2) was found to be sensitive to the crystal-field interaction and An 5f – ligand 2p hybridization effects, the 5d XAS of Am and Cm only depends on the nominal oxidation state. [1] SM Butorin et al., Chem. Commun. 53, 115 (2017); SM Butorin et al., PNAS 113, 8093 (2016). [2] SM Butorin et al., Chem. Eur. J. 22, 9693 (2016). [3] KO Kvashnina, YO Kvashnin, JR Vegelius, A Bosak, P Martin, SM Butorin, Anal. Chem. 87, 8772 (2015).

Resume : Austenitic stainless steels are widely used in the nuclear industry as a piping material and as a canister material for interim storage of spent nuclear fuels, due to favourable material properties such as corrosion resistance. However, when austenitic stainless steels are exposed to chloride environments and a tensile stress, they may fail via stress corrosion cracking. Here, the influence that the amount of salt deposited on the surface has on the extent of stress corrosion cracking in SS304L is studied. Samples were strained to 5% uniaxially and an upper surface tensile stress of 60 MPa was applied. MgCl2 salt was deposited on the sample surfaces and samples were exposed to conditions of 90oC and 70% relative humidity for a duration of 20 days. Samples were analysed via optical microscopy to study crack densities and corroded area. It was found that there was a linear correlation between salt loading and crack density for loadings within the range of 8 x 10-4 g.cm-2 and 2.6 x 10-2 g.cm-2. Above 2.6 x 10-2 g.cm-2 the crack density reduced from the peak value and at around 4.2 x 10-2 g.cm-2 few cracks were observed. Cross-sectional analysis showed that average crack velocities were relatively consistent (between 1 and 2 μm.hr-1) despite salt loading differing by 2 orders of magnitude suggesting that loading has little effect on crack propagation rate. A second set of samples taken from a SS316L Holtec canister were tested. These included samples of the canister base material and samples that spanned across an axial canister weld. As expected, the SS316L samples showed greater corrosion resistance and after the 20 day test period, base metal samples showed no evidence of cracking. The samples containing the weld were more heavily corroded at the heat affected zone, however, even still very few cracks were observed.

Resume : The Isotope Separator On-Line DEvice ISOLDE is a facility dedicated to the production of radioactive ion beams at CERN. With over 50 years of experience, the ISOLDE facility is able to deliver more than 1000 different isotopes of 74 chemical elements for various applications in nuclear and atomic physics, material science and nuclear medicine. Radionuclides are produced by irradiating thick targets with a 1.4 GeV proton beam. In more than 65% of the beam time, the targets are made of refractory materials such as highly porous depleted uranium carbide with excess graphite (UCx). To increase isotope release efficiency, two new uranium based materials were engineered. The target materials had stable microstructures, however the materials were extremely pyrophoric due to their large surface area and required extreme care in all handling procedures. While UCx cannot be kept in this form for long-term storage, a safe process for the conversion into oxide is investigated. In this contribution, the oxidation kinetics of the next generation target materials which depends highly on the starting microstructure will be discussed. Systematic investigations are underway in order to develop a safe and controlled stabilization process. The work developed could be transferred to other facilities worldwide, as a new waste disposal channel. This project has received funding from the European Union's Horizon 2020 research and innovation programme under grant agreement No. 642889.

Resume : 1. INTRODUCTION The molten salt reactor has been focused possibly due to following reasons: 1) the amount of Pu production is little, if thorium fuel cycle is introduced using ORNL original design MSR, and it is beneficial for non-proliferation. 2) Thorium cycle enables to open new energy source which may solve crisis of population explosion in the world. 3) The liquid fuel of MSR is compatible with pyroprocessing of spent nuclear fuel (SNF). 4) MSRs can burn waste effectively and contribute to the reduction of SNF. The last point has significant impact so that we investigate this potential more closely, introducing different types of MSR. Development of MSR could be consisted of six steps: Materials and salt selection => Salt and materials corrosion tests => Reactor concept development => License framework development => Test reactor design, construction and operation => Commercial plant design and development Almost all venture companies, except SINAP(Shanghai, China) only develop the former 4 steps. 2,Types of Molten Salt Reactors I am going to three types of molten salt reactors. 1) fluoride-fuel, thermal neutron reactor This can extinguish Pu but accumulates Am or Cm. For example; Seaborg technologies now develop Waste Burner reactor. 2) fluoride-fuel, fast neutron reactor In Russia, some type of fluoride fast neutron reactor has been investigated. One is the MOSART reactor. Solubility of the fuel material was investigated in the fluoride molten salt; In the components of 15Li-58NaF-27BeF2, the solubility of Pu is 1.33mol% at 550℃, 1.94mol% at 600℃. (COEMAT, Ignatiev P&T 2003) Neutron physics calculations show fluoride fuel fast neutron molten salt reactor can come into effect for the burning process of Pu and MA materials. 3) chloride-fuel fast neutron reactor There are some arguments of corrosion of materials. In cursory first look at the Molten Chloride Fast Reactor as an alternative to the conventional BATR Concept, only Hastelloy N and molybdenum could be used in material for reactor components. Experiments of material embrittlement in fast neutron environment have to be done. 3,Conclusion The most significant technical barrier in the development of molten salt reactor could be material corrosion and embrittlement. Recently most of the corrosion experiments for MSR are in the static-state molten salt. Further experiments is necessary with molten salt loop system, of the scale of real commercial reactor. It is planned to construct a molten salt test loop system and investigate usability of sus316 in the system of molten salt reactors.

Resume : India has modest natural uranium but vast thorium resources. Accordingly , a 3 stage nuclear power program is underway , with CANDU-PHWRs & LWRs in stage I , Fast Breeder Reactors (FBRs) in stage II with mixed Pu239-U238 fuel core and U238 and Th232 blanket to breed Pu239 and U233 respectively and finally in stage III , a self sustaining thermal breeder based on Th232-U233 fuel cycle. PHWR is the backbone of the nuclear power program in India. Of the 22 reactors in operation, with total installed power of some 6,700 MWe, 18 are small ( < 300 MWe) and medium ( < 700 MWe) PHWRs, 2 are generation I BWRs of 160 MWe each , in operation since 1969, and the latest two reactors are Russian VVER 1000 MWe (PWRs). Eleven reactors are in different stages of construction. All water cooled reactors are large ( ≥ 700 MWe), of which 6 are PHWR 700 MWe and 4 are VVER 1000 . A Prototype Fast Breeder Reactor of 500 MWe (PFBR 500) is in the final stageofcommissioning.Recently , the government of India has approved construction of 10 additional PHWR 700 for which the sites have been allocated. PHWR and related uranium fuel cycle, heavy water and zirconium technologies have attained industrial maturity in the country. Since 2009 , India has been importing natural uranium ore concentrate from Russia , Kazakhstan and Canada for meeting more than 50 % of the fuel requirement for operating PHWR 220 MWe units The government of India has been negotiating with reputed overseas vendors like Westinghouse and General Electric of USA and CEA , France for collaborative construction of Gen III+ LWR parks with life time guarantee of fuel supply from the vendors . As a first step to the FBR program, a Fast Breeder Test Reactor (FBTR 40 MWt) is in operation since 1985 with hitherto untried plutonium rich mixed uranium plutonium monocarbide fuel. The driver fuel for PFBR 500 is mixed uranium plutonium oxide (MOX) containing up to 25 % PuO2. The MOX fuel for PFBR has already been manufactured and awaiting loading in the core. The plutonium and depleted uranium for FBRs are obtained from the three spent fuel reprocessing plant that operate on the PUREX process. For the third stage , design and development are underway for construction of an Advanced Heavy Water Reactor (AHWR) of 300 MWe to be fuelled initially with (Th,Pu) O2 and later with ( Th, U233)O2 containing some 4 % fissile material. The present paper gives an overview of the nuclear power and related fuel cycle program in India, highlighting the authors experience in manufacturing and development of conventional and advanced fuels for thermal and fast breeder reactors in India.