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2019 Spring Meeting



Nuclear materials

Nuclear energy production requires materials that are very resistant under demanding environment: temperature, pressure and irradiation field. These materials act as barriers and their structural properties are investigated with emphasis on mechanical performance, durability, plasticity and stability. The symposium J includes sessions dealing with materials ranging from structural components of fission thermal or fast reactors, fuel materials to waste forms. Production of these materials at the industrial level is discussed on the basis of economical and safety considerations. Macro – properties such as quality, thermodynamics thermophysical and mechanical as well as micro-structural analysis are studied from the atomic scale to the materials within a multi-scale approach.


The component materials for Gen II and III, as well as for Gen IV reactors are tested for their behaviour under irradiation with particles (neutrons, protons, deutons …) of high energy and large flux. These are structural materials such as alloys (steels) or composites (cercer, cermet …) or are the coolant e.g. liquid metals or molten salts. The session concerns the R&D for high temperature gas reactors and molten salt reactors.

The fuels consist of solids (or liquids) with their components (homogeneous/heterogeneous, matrices, fissiles, burnable poisons, fertiles and additives). These fuel materials (oxides, nitrides, carbides, silicides / solid, or fluoride, chloride / liquid  …) are presented in a comprehensive way with emphasis of their intrinsic properties (thermal conductivity, high melting points). The new accident tolerant fuels and the inert matrix fuels shall be discussed in specific sub sessions. This section also includes liquid fuels such as molten fluorides (thermal) or chlorides (fast) for the molten salt reactors. Other liquids may also be investigated if acceptable. Again key properties such as melting points, thermal capacity, … are discussed in this session.

The waste forms must finally be recognized for their stability, durability, low solubility or leachability over geological time scale. The research includes materials such as homogeneous amorphous (glass) crystalline (spent UO2 fuels) or heterogeneous (Synroc, spent MOX fuel).

In all cases irradiation with accelerator guide the investigators in choosing optimal components prior irradiation in reactor. The challenge this century will be to work with much reliable and robust materials that make their use safer in nuclear system. Specific attention shall be given on the cost in Energy (EJ) during their production and their performance in term of retention of contaminants. The symposium shall contribute in enhancing the safety of the nuclear systems.

Hot topics to be covered by the symposium:

  • Green uranium/thorium mining
  • Extraction U from seawater
  • Green fuel production (energy cheaper routes)
  • ‘Stronger’ structural materials
  • Liquid fuel reactor components
  • New nuclear industrial applications
  • More performing waste forms
  • Integrated Computational Materials Engineering
  • Education in Nuclear Materials

List of invited speakers:

  • Sylvie Delpech, substituted for Else Merle (Lancaster Univ.): Materials for Molten Salt Reactor: the Chemical Control
  • Dave Goddard (NNL UK): Manufacturing Challenges for High Density Fuel Materials
  • Kirk Sorensen (Flibe Energy): Materials challenge of the MSR
  • Liu Tong (CCNC): Status of Nuclear Energy in China
  • Eva Uribe (Sandia National Laboratories – Livermore): Th fuel cycle
  • Lumin Wang (University of Michigan): Recent Progress on the Study of Radiation Tolerant Materials with Ion Beams

Tentative list of scientific committee members:

  • Christine Delafoy, Framatom, France
  • Peng Xu, Westinghouse, USA
  • Liu Tong, China General Nuclear, China
  • Lingfeng He, Idaho National Laboratory, USA
  • Daniel Shepherd, National Nuclear Laboratory, UK  
  • Michael Ojovan, IAEA, Vienna, Austria
  • Eric Simoni, Uni Orsay, France
  • Rory Kennedy, INL, USA
  • Christina Trautmann, GSF, Germany
  • Yongsun Yi, El Kahlifa Uni, Abu Dabi, UAE
  • Motoyasu Kinoshita, CRIEPI, Japan
  • Xu Hongjie, SINAP, CAS, Shanghai, China
  • Stephane Bourg, CEA Marcoule, France
  • Thierry Stora, ISOLDE, CERN, Switzerland
  • Rudi Konings, ITU JRC, EU 
  • Sergey Stefanovky, Frumkin Institute, Moscow, Russia
  • Eric Vance, ANSTO, Australia


The best papers shall be published in  J. Mater. Sci.

Start atSubject View AllNum.
09:00 Welcome address by G. Kiriakidis (E-MRS President)    
09:05 Introduction to the Nuclear Materials Symposium by C. Degueldre et al. (Symposium organizers)    
Fuel from mining to production : NN
Authors : D. T. Goddard
Affiliations : National Nuclear Laboratory, Preston Laboratory, Springfields, Preston, Lancs. PR4 0XJ

Resume : Uranium dioxide (UO2) has been the fuel of choice for Light Water Reactors (LWRs) due to its excellent irradiation performance and chemical compatibility with Zr alloy claddings and the water coolant. The manufacture of this fuel is highly automated with good process control and low reject levels. Despite these obvious advantages the drive for improvements in reactor safety and especially economics means that alternative fuel materials are now under consideration. These are referred to as Accident Tolerant Fuels (ATF), although the terminology Advanced Technology Fuels is probably more appropriate, since both economics and safety are important. Uranium nitride (UN) and uranium silicide (U3Si2) both have a uranium density higher than that of UO2 which could result in fuel cycle cost savings. The thermal conductivities of these materials are also significantly higher, leading to improvements in their response to certain reactor fault sequences. The introduction of a new fuel material is not without its challenges however. Not only is the irradiation performance of these materials in LWR conditions largely untested, but the consequences of a fuel pin leak must be addressed, as well as the fabrication of these fuels at a commercial scale. NNL are working with academic and industrial collaborators to develop new routes for the fabrication of high density fuel materials. This includes the construction of a rig to enable testing of reactions between uranium hexafluoride (UF6) and silicon at elevated temperatures which it is hoped may provide a more direct conversion process for U3Si2 production. Pelleting processes for U3Si2 fabrication are also described, which may need to introduce concepts to reduce water reactivity, such as the use of dopants, composite materials or surface coatings.

Authors : Zhenyuan Bai 1, Jun Wang 1,2
Affiliations : 1 College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, P.R. China; 2 Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, 150001, PR China.

Resume : In this study, the zinc oxide nanorod/ramie fiber-amidoxime (ZnO/RF-AO) adsorbent was successfully synthesized by two steps, including amidoxime and solvothermal self-assembly. The use of amidoxime to modify RF effectively improves the selectivity of the adsorbent, and the introduction of ZnO, the adsorbent effectively increases the active sites on the RF-AO surface, thereby improving the adsorption capacity of RF-AO. The SEM was used to characterize the material morphology. A series of batch adsorption experiments were carried out to study the effects of pH value and initial concentration of uranium solution, contact time and temperature on the adsorption of U(Ⅵ). The result show that the optimal adsorption amount of U(VI) of ZnO/RF-AO adsorbent was 266.88 mg g-1 at 298.15 K, pH = 7, C0 =200 mg L-1, and t =120 min, which is consistent with the pseudo-second-order kinetics model and the Langmuir isotherm model. The adsorption is an endothermic and spontaneous process. The adsorbent exhibits excellent ion selectivity and cycle reproducibility. It is worth noting that the removal efficiency of ZnO/RF-AO composites for low-concentration uranium is 89% and for real seawater and simulated seawater contaminated by radioactive waste can also reach more than 45%, demonstrating the potential of the adsorbent in the field of uranium extraction from seawater. Based on FTIR spectroscopy and XPS, a plausible U(VI) adsorption mechanism for ZnO/RF-AO composites was proposed. (This paper is funded by the International Exchange Program of Harbin Engineering University for Innovation-oriented Talents Cultivation.)

Authors : Ju Peihai1,Wang Jun1,2
Affiliations : 1 Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, Harbin 150001,Heilongjiang,China; 2College of Material Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, Heilongjiang

Resume : In this paper, a new Mg2CO3(OH)2/ polyacrylonitrile (PAN) adsorption material was prepared by hydrothermal method for uranium recovery, and the adsorption performance of the material was studied. The structure and morphology of the material were characterized by scanning electron microscopy, X-ray photoelectron spectroscopy and Fourier transform infrared spectroscopy. The adsorption conditions, kinetics and thermodynamics of the materials were studied by adsorption experiments. The adsorption performance of materials in practical applications was also discussed by ion competition experiments, desorption experiments, and low concentration adsorption experiments. The results show that the optimal adsorption conditions for Mg2CO3(OH)2/PAN was pH = 5.0; under laboratory conditions, the maximum adsorption capacity can reach 277.3 mg/g. Mg2CO3(OH)2/PAN has good selectivity and circulation. What’s more, the adsorption efficiency of low concentration uranium reaches more than 90%. (This paper is funded by the International Exchange Program of Harbin Engineering University for Innovation-oriented Talents Cultivation.)

Authors : Huiquan Gu1, Qi Liu1, Jun Wang1,2
Affiliations : (1) College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, P.R. China. (2) Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, 150001, PR China.

Resume : Uranium adsorption from seawater was wildly investigated for decades, as the rapid development of nuclear power. Due to the high adsorption affinity for metal ions and the hydrophilicity of polyethylenimine (PEI) and the industrial productivity of polyacrylonitrile (PAN), the material of PEI modified PAN fibers (PPFs) were designed and synthesized. After prepared, the PPFs were woven into cloths (PPCs). The micrographs of PPCs were analyzed by scanning electron microscope (SEM). Batch experiments of uranium adsorption were used to investigate the properties of PPCs. Pseudo-first-order kinetics model, pseudo-second-order kinetics model, the Langmuir isotherm model and Freundlich isotherm model were chosen for further investigation. The results show that the best adsorption pH is 5 and the maximum adsorption capacity can reach 343.58 mg g-1 in 300 min, followed by the pseudo-second-order kinetics model and the Langmuir isotherm model. Marine tests of PPCs were also utilized in Huanghai, China with the maximum adsorption amount of 14.01 mg kg-1. However, the capacity of PPCs was influenced by marine fouling, barnacle to be specific. For further study of marine adsorbent for uranium, the antifouling property should be considered by functionalizing antifouling groups to the adsorbent, such as cationic polymers, antibiotics, antimicrobial peptides, antimicrobial enzymes, etc. (This paper is funded by the International Exchange Program of Harbin Engineering University for Innovation-oriented Talents Cultivation.)

10:30 Coffee break    
Authors : Jiawu Gao1, Jingyuan Liu1, 2, Jun Wang1, 2
Affiliations : 1 College of Material Science and Chemical Engineering, Harbin Engineering University, 150001, PR China; 2 Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, 150001, PR China.

Resume : The shortage of energy is an urgent problem which can be solved with the development of nuclear. Uranium is the fundamental element for nuclear fuel existing 4.6 billion tons in the oceans. However, the extremely low concentration, other competition metal ions and background high salinity restrict the uranium extraction from seawater. In this study, phosphonate-functionalized luffa cylindrical (PLC) was synthesis by Trans esterification reaction. Based on FT-IR and XPS, the PLC was proved to be synthesis successfully. The adsorbents showed high adsorption capacity which fitted Langmuir model and maximum sorption capacity can be reached 598.8 mg g-1 at 298K. The kinetic data followed the pseudo-second-order model, the adsorption equilibrium time took 15 mins only, indicating the adsorption process is chemisorption dominantly. Meanwhile, the PLC also presents superior recyclability and stability after five cycle times. In the end, the PLC showed high removal rate to natural seawater which demonstrating that PLC adsorbent is a potential material for uranium extraction from seawater. (This paper is funded by the International Exchange Program of Harbin Engineering University for Innovation-oriented Talents Cultivation.)

Authors : Sang Gyu Park, Jeong Yong Park
Affiliations : Korea Atomic Energy Research Institute

Resume : The motivation for innovative fuel development is the development of the advanced ultra-high burnup sodium-cooled fast reactor metallic fuel concepts. The fabrication experiment seeks to investigate advanced fuel designs with the following features: decreased fuel smeared density (SD), venting of the fission gas to the sodium coolant, a uranium-molybdenum (U-Mo) based alloy fuel system, coating or liner on the cladding inner surface, and/or targeted fuel alloy additions to reduce FCCI (Fuel Cladding Chemical Interaction), and an advanced fabrication method that includes consideration of annular fuel and co-extruded fuel and cladding. From the experiment result, annular fuel shows the possibility of the reduction of swelling effect and then prevention of the FCMI (Fuel Cladding Mechanical Interaction). However, the fabrication technology of the annular fuel has not been developed yet. Therefore, KAERI has started to study the annular fuel fabrication method by using hot extrusion method. In this study, the prototype of annular fuel has been fabricated by using Cu billet. The dimension of billet and annular fuel has been determined, and then and design requirements and material for the mold has been simulated by using Deform 3D program. After the mold fabrication, the prototype annular fuel has been fabricated and its texture were examined by using EBSD (Electron Back Scatter Diffraction).

Affiliations : 1- CEA, Centre de Cadarache-DEN⁄DEC, 13108 St Paul Les Durance, France 2- Institut Néel CNRS/UJF, 25 rue des Martyrs, BP 166, F38042 Grenoble Cedex 9, France 3- CEA, Centre de Grenoble-DRF⁄INAC⁄PHELIQS⁄IMAPEC, 38054 Grenoble Cedex 9, France

Resume : Progressing in the thermomechanical modeling of polycrystalline uranium dioxide UO2 under irradiation requires taking into consideration the phenomenon of grain boundaries (GB) decohesion. The description of the thermomechanical behavior of UO2 down to the polycrystalline aggregate scale, especially considering the grain orientations and the crystallographic structure of the GBs, is a key strategy for improving the simulation tools. It involves, on the one hand, identifying the GBs of interest in the polycrystalline UO2 and, on the other hand, experimentally determining their energy, a valuable quantity for the simulation. In this work, our purpose was to study the effects of sintering parameters (temperature and duration) and of the nature of powder on the evolution of the population and the distribution of GBs. Thus, several pellets possessing different microstructures were investigated by using EBSD-SEM technique. Then, the energies of the GBs were assessed using the thermal grooving technique, which consists in performing a thermal etching to reveal the GB grooves and in measuring their angle. Therefore, UO2 pellets were annealed at 1400°C for 4 h under Ar 5%H2 atmosphere. Special GBs of interest were then identified using EBSD-SEM. The dihedral angle (ψ) of their grooves were determined with atomic force microscopy (AFM). The GB energies (Ɣgb) were finally estimated from the Herring equation.In parallel with these experimental investigations, GBs were simulated at the atomic scale and their energy calculated by molecular dynamics using different empirical potentials. Our first results are very promising as the comparison between measured and calculated values shows a good agreement. Such coupled experimental - simulation approach gives confidence in the reliability of the experimental results and, at the same time, strengthens the simulation by allowing the choice of the best-suited potential. This will indeed allow extending numerical simulations to other properties such as cleavage energies and toughness. Keywords: Uranium dioxide, Grain boundary energy, Simulation, Thermal grooving, AFM

Authors : Victor TRILLAUD, Renaud PODOR, Nicolas DACHEUX, Nicolas CLAVIER
Affiliations : Institut de Chimie Séparative de Marcoule, UMR 5257 CEA-CNRS-ENSCM-Univ. Montpellier, Site de Marcoule – 30207 Bagnols/Cèze cedex, France

Resume : As a key-step for the elaboration of nuclear fuels, the sintering of UO2 has been studied for years. If grain growth processes were investigated thanks to both experimental works and calculations, the elaboration of necks during the first step of sintering was generally assessed only through numerical models, frequently based on simple configurations, such as two spherical single crystals in contact. In order to complement such approaches, the elaboration of necks during the sintering of UO2 microspheres, herein used as model compounds, was experimentally observed by High Temperature Environmental Scanning Electron Microscopy (HT-ESEM). In this aim, we first developed an original synthesis protocol yielding size-controlled UO2 spherical particles through the hydrothermal treatment of U(IV) aspartate. The kinetics associated with the evolution of neck radius, contact angles and centers displacement during the sintering of two microspheres were then evaluated, by in situ observations in the 800-1200°C range. Also, as UO2 sintering is well-known to be strongly impacted by atmosphere, different experiments were performed with variable pO2 to determine its impact on the kinetics of neck formation. Finally, complex assemblies made of three spheres or more were also investigated and led to preliminary results aiming to extrapolate the data obtained on model systems to real-life UO2 fuel pellets.

Authors : Jérémie Manaud, Jérôme Maynadié, Daniel Meyer, Adel Mesbah, Nicolas Dacheux, Nicolas Clavier
Affiliations : ICSM - UMR 5257 - CEA, CNRS, ENSCM, Univ Montpellier, Site de Marcoule, BP 17171, 30207 Bagnols/Ceze cedex, France

Resume : French nuclear fuel fabrication process is currently based on the blending of UO2 and PuO2 powders prior to their sintering at high temperature. Within the development of new generations of reactors, wet chemistry routes are studied to improve the cations homogeneity within the final mixed oxide in order to increase the incorporation of plutonium and to optimise the fabrication process. In this framework, this work aims at directly synthesize uranium oxide samples from solutions by combining oxalic precipitation step with hydrothermal conditions in order to avoid the preliminary calcination step. The powders obtained after mixing uranium (IV) hydrochloric solution and oxalic acid, and after the subsequent hydrothermal treatment between RT and 250°C, were characterised by XRD, SEM and TGA-MS. U(C2O4)2, nH2O, was obtained up to 170°C whereas UO2+x was detected above 180°C. Simultaneously, the morphology changed from platelets to nearly-spherical aggregates. Rietveld refinements showed variations of the unit cell volume and crystallite size as a function of the temperature. Moreover, hydrothermal treatment at 250°C led to very low amounts of residual water and carbon, comparable to those obtained after thermal conversion of oxalates at 1000°C. On the basis of these results, a multiparametric study was undertaken to explore the effects of kinetics and pH on oxides properties. Finally, the sintering of final uranium oxides was studied by the means of dilatometry.

Authors : V.A. Borodin(1,2), A.V. Tenishev(1), D.P. Shornikov(1)
Affiliations : (1) National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia; (2) NRC "Kurchatov Institute", Moscow, Russia

Resume : The primary compaction of uranium dioxide powders into pellets is an important stage of the pellet production cycle, which largely determines the quality of the resulting product after subsequent sintering. Within standard procedures, the compaction is typically done at room temperature, leading to the geometric pellet density at the level of roughly 50% of the theoretical UO2 density. In this report we consider the question how the compaction efficiency is affected by increasing the compaction temperature up to high values that remain still below those of the quick sintering onset. The experiments on uranium dioxide powder compaction at relatively low external pressures (20-40 MPa), but at temperatures ranging up to 1200ºC have demonstrated that at temperatures above ~700ºC one observes a drastic increase in the efficiency of pellet densification. The effect becomes more pronounced with temperature increase, implying the launching of a certain temperature sensitive accelerated compaction mechanism. Both the scanning electron microscopy observations and the molecular dynamics modelling indicate that the mechanism behind the powder compaction efficiency increase is related to the activation of plastic strain, most probably the grain boundary sliding.

12:15 Lunch    
Authors : Eva C. Uribe
Affiliations : Systems Research and Analysis Group, Sandia National Laboratories, California

Resume : Nuclear energy from thorium reactors has the potential to be safer, cleaner, and more resistant to the clandestine diversion of materials that could be used for nuclear weapons programs, compared to conventional reactors, which are fueled with uranium and plutonium. Thorium reactors will produce fissile material in the form of U-233. The International Atomic Energy Agency has defined a significant quantity of U-233 as 8 kg, the same as for Pu-239. Nevertheless, thorium reactors are regarded as more proliferation-resistant because they generate less plutonium than conventional light water reactors, and because the handling and transport of any U-233 generated is significantly complicated by the co-production of U-232, which decays through a series of highly radioactive daughters. There are several pathways by which U-232 is produced in thorium reactors, and thus various means by which U-232 formation can be reduced or eliminated altogether, rendering the U-233 produced a serious proliferation concern. Chemical reprocessing of recently discharged spent thorium fuel change the amount of U-232 in the final U-233 product. Protactinium-233 (half-life = 27 days) is an intermediate in the formation of U-233. Spent fuel reprocessing may intentionally or inadvertently isolate Pa-233, which would decay to isotopically pure U-233. This presentation will discuss the role of protactinium in nuclear energy from thorium, the chemistry of protactinium in proposed reprocessing schemes, and the potential for protactinium to diminish the nonproliferation benefits of thorium fuel cycles. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525. SAND2019-0394 A

Authors : Claude Degueldre*, Malcolm Joyce
Affiliations : Nuclear Engineering Group, Engineering Department, Lancaster University LA1 4YW, UK *E-mail:

Resume : The actinides are thought to have been generated by the rapid neutron capture (r)-process in a supernova at the origin of the solar system (4.6109 a ago). Thus, given the generation processes for thorium, Th, and uranium, U, (and their precursors) are similar, the difference of half-life (232Th: 14.05109 a, and 238U: 4.468109 a) would suggest a larger abundance of Th than U be present today. However, the chemistry of U is very different from that of Th which could suggest different pathways to explain the different concentrations of Th and U in various formations and local deposits of UO2. Focussing on the accessible geosphere of the Earth (i.e., the crust and hydrosphere), the evidence for the presence of these actinide elements in the liquid and solid rocks was examined. This study discusses why, in seawater, soluble Th (510-11 M) is more depleted than U (1.410-8 M), and why solid ores of pure UO2 or U3O8 are relatively common while ThO2 deposits are quasi-non-existent. Thorium is rather diluted in rock, relative to uranium, requiring greater mining, milling and treatment to recover it on comparable scales. This paper discusses and compares the abundances of U and Th in the Earth’s accessible geosphere as well as the extent of their known, economically-recoverable resources.

Authors : Michael A Bromley (a),* Colin Boxall (a), Robin Taylor (b), Mark Sarsfield (b)
Affiliations : (a) Lancaster University, Engineering, Gillow Avenue, Lancaster, LA1 4YW, UK (b) National Nuclear Laboratory, Central Laboratory, Sellafield, Cumbria, CA20 1PG, UK

Resume : Nuclear fuel recycling processes are well established and the recovery of U and Pu for the fabrication of new fuel has been successfully operated over several decades. However, the increasing usage of mixed metal oxide (MOx) fuels in LWR’s across Europe and Japan, combined with the need to reduce the volume of waste arising from spent nuclear fuel, continues to drive the need for cleaner, more efficient reprocessing plants and routes to manufacturer for MOx fuel pellets. Within the UK Nuclear Innovation Programme, work has begun on the development of high Pu content MOx fuel fabrication processes which aim to reduce the heavy metal throughput of reprocessing plants while maintaining a level of proliferation resistance with the use of U / Pu co-treatment stages throughout. As a contribution to this, a rapid and clean photochemical process for the reduction of U(VI) and co-reduction of U(VI) / Ce(IV) (as a Pu surrogate) is under development at Lancaster University. By directly utilising the advantageous photochemistry of U, with isopropanol as a reductant, we report the convenient photo-excitation and chemical reduction of U(VI) upon exposure to 407 nm light energy. Using a purpose built laboratory scale photo reactor, 407 nm light is delivered by an ultra-bright 5,600 mW LED array within an integrating sphere for highly efficient energy transfer to the U solution. The effect of illumination is immediate with the successful conversion of 12 g / dm3 U(VI) in 2 mol / dm3 nitric acid demonstrated within 90 seconds. With the addition of hydroxylamine nitrate as a stabilising agent / nitrous acid scavenger, this results in the rapid generation of a stable U(IV) product solution. Through a combination of direct photochemical reduction and the indirect chemical reduction of Pu(IV) by photo-generated U(IV), we believe this method is promising as a viable co-reduction process suitable for use prior to oxalate precipitation and subsequent calcination as a homogeneous mixed metal oxide fuel. Furthermore, with no requirements for non-CHON reagents and the elimination of hazardous substances, such as hydrazine and hydrogen gas as used in more traditional methods, the photochemical reduction process is compatible with existing reprocessing conditions and the need to improve process safety.

Authors : Lu Cai(1), Kathryn Metzger(1), Edward J. Lahoda(1), Frank Boylan(1), Peng Xu(1), Andrew Atwood(1), Robert L. Oelrich(1), Bowen Gong(2), Jie Lian(2)
Affiliations : (1) Westinghouse Electric Company LLC, U.S.A.; (2) Rensselaer Polytechnic Institute

Resume : As a promising candidate for next generation fuel, U3Si2 offers higher thermal conductivity and higher uranium loading than UO2. However, the inadequate water corrosion resistance of U3Si2 is unfavorable compared to UO2 when there is a leaking fuel rod in normal operation. The oxidation reaction between U3Si2 and water vapor at around 400oC can cause rapid pellet fragmentation. The advanced U3Si2 have been developed to improve the water corrosion resistance by grain boundary modification with highly corrosion-resistant additives. An experimental thermal analysis of the advanced U3Si2 is undertaken to quantitatively determine the behaviors of these compounds at elevated temperatures in a water vapor atmosphere. The behavior of these materials is revealed by thermogravimetry (TG) isothermal experiments at the typical light water reactor (LWR) coolant temperatures in a water vapor atmosphere. The results will be discussed and compared to UO2 for better understanding of their behaviors.

Authors : Mattia Cabrioli, Erkka Frankberg, Matteo Vanazzi, Fabio Di Fonzo
Affiliations : Politecnico di Milano Energy Department, Center for Nano Science and Technology Istituto Italiano di Tecnologia; Center for Nano Science and Technology Istituto Italiano di Tecnologia; Center for Nano Science and Technology Istituto Italiano di Tecnologia; Center for Nano Science and Technology Istituto Italiano di Tecnologia

Resume : Accident Tolerant Fuel (ATF) concepts aim at providing radical performance and safety improvements at economically attractive conditions. Coatings as near-term evolutionary option allow licensing to rely on traditional claddings for structural purposes. In such case, they are expected to reduce corrosion rates while retaining structural integrity and adhesion, minimizing hydrogen uptake and improving high temperature oxidation resistance. In this work, nano-ceramic coatings are designed, characterized and tested in the framework of the European project IL TROVATORE. Coatings are grown on relevant substrates as Zirconium alloys and AISI 316L, by the Pulsed Laser Deposition (PLD) technique. PLD is a particularly well-suited method for growing high-quality ceramic films, with strong adhesion and interfacial bonding. Different oxides are investigated as potential solutions. Candidate materials are chosen by matching neutronic requirements and chemical stability in pressurized-water environment. Coating performance is evaluated through the exposure of specimens to PWR-relevant conditions and high-temperature steam environment. Pre- and post-exposure structural features are characterized by FT-IR, SEM, EDX and XRD. Finally, burst tests are performed on coated tubes, to challenge structural integrity and adhesion of the films at extreme deformation rates. Overall, the results show that the selected PLD coatings are a promising option as coated-cladding ATF concept.

Authors : Haiming Wen, Jiaqi Duan, Hans Pommerenke, Adam Bratten
Affiliations : Missouri University of Science and Technology, USA

Resume : The fuel design for high-temperature gas reactors (HTGRs) consists of small tristructural isotropic (TRISO) fuel particles that have a UC or UO2 fuel kernel surrounded by layers of pyrolytic carbon and SiC. The TRISO particles are bonded using graphitic matrix materials into a cylindrical fuel compact or spherical pebble. Unlike nuclear-grade graphite, the matrix material (matrix-grade graphite) consists of partially-graphitized material. Matrix-grade graphite is not nuclear-grade graphite by definition; it is graphitic material composed of multiple types of graphite and carbonized phenolic resin. While HTGRs use pure helium as a reactor coolant, in some accident scenarios significant amounts of moisture or air can be introduced into the helium coolant and reactor core. The effects of oxidants on TRISO fuel integrity are essential considerations that are part of HTGR safety analysis. It is noted that no studies have been performed on oxidation of TRISO fuel matrix material in water vapor. In this study, TRISO fuel matrix materials were subjected to oxidation by moisture with conditions relevant to the moisture ingress accident. The kinetic parameters of oxidation were measured through the oxidation studies at different temperatures with different partial pressure of water vapor. The microstructures of the materials before and after oxidation were carefully characterized. The oxidation mechanisms were ascertained in relation to the microstructures of the materials.

Authors : Martin LEBLANC, Gilles LETURCQ, Eléonore WELCOMME, Xavier DESCHANELS, Thibaud DELAHAYE
Affiliations : CEA, DEN, MAR/DMRC/SFMA/LPCA, F-30207 Bagnols-sur-Cèze Cedex, France and Institut de Chimie Séparative de Marcoule, ICSM UMR5257, Centre de Marcoule, F-30207; CEA, DEN, MAR/DMRC/SFMA/LPCA, F-30207 Bagnols-sur-Cèze Cedex, France; CEA, DEN, MAR/DMRC/SFMA/LPCA, F-30207 Bagnols-sur-Cèze Cedex, France; Bagnols/Cèze, France; Institut de Chimie Séparative de Marcoule, ICSM UMR5257, Centre de Marcoule; CEA, DEN, MAR/DIR, F-30207 Bagnols-sur-Cèze Cedex, France

Resume : Within the U and Pu recycling process from nuclear spent fuels, the conversion of purified U and Pu solution into oxide powder is a key step at the interface between the separation / purification processes and the fabrication of uranium-plutonium oxide fuels called MOx ("Mixed Oxides"). This study deals with the development of a new "direct" conversion route based on advanced thermal denitration to synthetize mixed actinide oxide (U1-xPuxO2±δ). This new synthetize method consists in the gelation of an actinide nitrate solution within a crosslinked polymer, followed by dehydration and calcination under controlled conditions to obtain the targeted mixed actinide oxide. On the basis of 0.5 g batch productions, the feasibility of the synthesis of all the solid solution U1-xPuxO2±δ with x ranging from 0 to 1 was demonstrated without any redox adjustment of the actinide feeding solutions and whatever nitric acidity (up to 7 M) or total actinide concentrations. Moreover, a first scale up was operated on the basis of a 15 g U0.80Pu0.20O2±δ batch, also dedicated to study the pellet fabrication using the powder as synthetized (i.e. without any grinding step). 94% of the theoretical density pellet was obtained with oxygen to metal ratio of 2.00. Therefore, such a conversion route allows fabrication of any kind of MOx fuels (PWR or SFR) offering several advantages: no redox adjustments, no solid/liquid partitioning required and reduction of actinide dissemination risks.

Authors : Prantik Chakraborty, A. Chartier, C. Guéneau
Affiliations : Den-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette, France

Resume : In frame of the Generation-IV forum, the Sodium-cooled Fast Reactor (SFR) is investigated in order to fulfil the energy needs of present time. One of the key aspect of any nuclear reactor system is its fuel technology, with implications for fuel fabrication, safe operation, reusability and overall economics (Crawford, et al., 2007). Mixed oxide of uranium and plutonium (U,Pu)O2 (MOx) is the reference fuel considered for the SFR reactor. During operational in-reactor condition, the MOx fuel pellet undergoes very high temperature at its centre up to 2000°C along with an expected radial thermal gradient of 1000°C/cm. This notably high thermal gradient induces the migration and redistribution of the fuel elements from the centre towards the rim region. This thermal diffusion process results in the formation of a central void in the center of the pellet and further leads to the variation of thermo-mechanical and physico-chemical properties of the fuel material. Therefore, the development of a thermo-kinetic model of the fuel elements in the MOx is of utmost importance to ensure the safe operation of SFRs. A thermal diffusion model of oxygen in MOx fuel has been developed using the DICTRA code by Moore et al (Moore E., 2017). The aim of the present work is to extend this model with the assessment of cation diffusion parameters. However, the data on diffusion phenomenon of the cations in the MOx fuel are not very well known. In fact, the experimental studies are scarce and difficult to carry out, due to the radioactivity level of the samples and to the problem related to the control of the oxygen stoichiometry (and/or oxygen partial pressure) of the samples. Moreover no full thermo-kinetic model exists for the whole (U,Pu)O2±x solid solution.. Thus, in the present work, the cation diffusion, principally the plutonium diffusion data in MOx fuel, is investigated as function of plutonium content, oxygen stoichiometry and temperature. In the inception of the study, the lack of experimental measurements is circumvented by the ‘cBΩ’ model. This model estimates the activation energy of self-diffusion using the bulk properties of the material, so long as diffusion operates by a single type of defect. Prior model relies on the fitting of one single parameter only, whereas the activation energy explained by G=cBΩ, B being the bulk modulus, Ω the mean volume per atom and c the fitting parameter (Varotsos, et al., 1978). In case of MOx, we took the advantage of plethora of experiments and equations that are available to eptomize the evolution of bulk properties with the variation of plutonium proportion & oxygen-to-metal ratio of the fuel and temperature of the system (MATPRO). Hence, plutonium self-diffusion data have been roughly estimated for different plutonium and oxygen compositions and temperature in order to build a database of plutonium diffusion in the whole solid solution (U,Pu)O2±x . In a second step, using experimental data from the literature and the estimations from the ‘cBΩ’ model, a plutonium mobility database has been developed using the DICTRA code to provide a new model for plutonium diffusion in MOx fuel. The starting point is the CALPHAD thermodynamic model developed by Guéneau et al (Guéneau, et al., 2011) coupled with the oxygen mobility database assessed by Moore et al (Moore E., 2017) .The present thermo-kinetic model for (U,Pu)O2±x is used to perform application calculations with DICTRA to simulate diffusion couple tests and heat treatments on MOX samples. _______________________________________________________________________ (Crawford, et al., 2007): Crawford Douglas C., Porter Douglas L. et Hayes Steven L. Fuels for sodium-cooled fast reactors: US perspective [Revue]. - [s.l.] : Journal of Nuclear Materials, 2007. - 1-3 : Vol. 371. (Guéneau, et al., 2011): Guéneau Christine [et al.] Thermodynamic modelling of advanced oxide and carbide nuclear fuels: Description of the U–Pu–O–C systems [Revue] // Journal of Nuclear Materials. - 2011. - 1-3 : Vol. 419. - pp. 145-167. (MATPRO): MATPRO A Library of Materials Properties for Light-Water-Reactor Accident Analysis [Revue]. - [s.l.] : U.S. Nuclear Regulatory Commission. - SCDAP/RELAP5/MOD 3.1 Code Manual. (Moore E., 2017): Moore E. Guéneau C., Crocombette J.P. [Revue] // Journal of Nuclear Materials. - 2017. - Vol. 485. - pp. 216-230. (Varotsos, et al., 1978): Varotsos P., Ludwig W. et Alexopoulos K. Calculation of the formation volume of vacancies in solids [Revue]. - [s.l.] : Phys. Rev. B, 1978. - 6 : Vol. 18.

Authors : Karin Popa, Olaf Walter
Affiliations : European Commission, Joint Research Centre, P.O. Box 2340, D-76125 Karlsruhe, Germany

Resume : In 2016, we have proposed for the first time the hydrothermal decomposition of tetravalent actinides oxalates as a straightforward method to produce reactive actinide oxide nanocrystals [1]. The method could be easily applied at very low temperature (95-250 °C) in order to generate nanocrystalline nano-AnO2 (An= Th, U, Np and Pu) or different solid solutions. With respect to other thermal methods employing organic solvents, the hydrothermal decomposition of oxalates presents the advantage that the material obtained is free of any residual carbon impurities possibly blocking the nanocrystals surface [2]. Reproducible synthesis of a full series of (U,Th)O2 solid solutions and extended mechanical characterization of spark plasma sintered pellets have been also reported [3]. We present here the first results on the production of several nano-sized (U,Pu)O2 and we discuss the oxidation state of the actinides. [1] O. Walter, K. Popa, O. Dieste Blanco, "Hydrothermal decomposition of the actinide(IV) oxalates: a new aqueous route towards reactive actinide oxides nanocrystals", Open Chem. 14 (2016) 170-174 [2] K. Popa, O. Walter, O. Dieste Blanco, A. Guiot, D. Bouëxière, J.-Y. Colle, L. Martel, M. Naji, D. Manara, "A low-temperature synthesis method for AnO2 nanocrystals (An = Th, U, Np, and Pu) and associate solid solutions", CrystEngComm. 20 (2018) 6414-6422 [3] L. Balice, D. Bouëxière, M. Cologna, A. Cambriani, J.-F. Vigier, E. De Bona, D.G. Sorarù, C. Kübel, O. Walter, K. Popa, "Nano and micro U1-xThxO2 solid solutions: from powders to pellets", J. Nucl. Mater. 498 (2018) 307-313

16:00 Coffee break    
Poster J: Nuclear Materials : NN
Authors : Daniele Iadicicco, Boris Paladino, Matteo Vanazzi, Patricia Munoz, Serena Bassini, Marco Utili, Fabio Di Fonzo
Affiliations : Center for NanoScience and Technology (CNST) Istituto Italiano di Tecnologia (IIT); Center for NanoScience and Technology (CNST) Istituto Italiano di Tecnologia (IIT), Department of Energy at Politecnico Milano; Center for NanoScience and Technology (CNST) Istituto Italiano di Tecnologia (IIT); Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT); Agenzia nazionale per le nuove tecnologie, l'energia e lo sviluppo economico sostenibile (ENEA); Agenzia nazionale per le nuove tecnologie, l'energia e lo sviluppo economico sostenibile (ENEA); Center for NanoScience and Technology (CNST) Istituto Italiano di Tecnologia (IIT)

Resume : Materials are proving to be one of the major bottlenecks for future generation nuclear systems. New fission and fusion reactors share similar technological issues: corrosion by heavy liquid metals (HLM) and molten salts, radiation damage and Tritium confinement. We report on multifunctional nanoceramic oxide coatings grown by Pulsed Laser Deposition (PLD). Oxides were chosen because of their chemical inertia and thermodynamic stability. Among these, Alumina and Yttria were identified as promising candidates. Coatings of these materials have been tested as Tritium permeation barriers with Hydrogen at different temperatures (from 350°C to 650°C). They showed a permeation reduction factor (PRF) up to 10^5 at 650°C. These results have been confirmed also in the case of Deuterium permeation, both under 1.8 MeV electron irradiation and for samples that were previously irradiated with ions. In addition, to evaluate the chemical compatibility of the films in HLM, samples have been exposed to static corrosion tests (both Pb-16Li and pure Pb) up to 10000 hours. No corrosive attacks on the substrate were detected. Finally, as for the magnetohydrodynamic drag in fusion applications, the electrical resistivity of the coatings has been measured, showing good insulating behavior. To conclude, PLD-grown oxide coatings present interesting properties as multifunctional protective barriers, proving to be one of the possible solutions for the technological issues of future nuclear systems.

Authors : Jeremiah Emanuel Josey
Affiliations : Thorium Salt Network, MECi Group, Baar, CH-6340, Switzerland

Resume : Molten Salt Reactors (MSRs) are currently receiving increased attention as one of the Generation Ⅳ systems [1]. Among other features, MSRs envisage increased safety, reduced costs and compactness [2]. These, combined with the potential of using thorium as fuel as well as reusing the waste of the existing nuclear power plants, lead to active research in MSR technology during the past two decades. However, the conventional approach to regulating nuclear technologies makes it challenging to share the knowledge and the technical experience accumulated in this field among all involved parties. Establishing a platform which would enable decentralized and highly secure data collaboration as well as allow providing services relevant to the MSR development, could become ground-breaking for the nuclear power technologies and open up a path for making nuclear a publicly accepted solution for producing clean energy worldwide. In this article we explain how the Thorium Salt Network is planning to build a blockchain based platform which would allow MSR experts and supporters to not only collaborate and share knowledge but also provide and receive services in MSR material research, fuel carrier salt purchase, etc. 1. 2. D. LeBlanc, Molten salt reactors: A new beginning for an old idea, Nuclear Engineering and Design, 2010

Authors : S.N. Kalmykov, V.G. Petrov, A.A. Zherebtsov, A.P. Varlakov, A.V. Dmitrieva, V.V. Kapustin
Affiliations : Lomonosov Moscow State University, department of chemistry

Resume : A promising method of reprocessing of advanced nitride spent nuclear fuel (SNF) is the application of pyroelectrochemical technologies where SNF is dissolved in the LiCl-KCl eutectic with subsequent electrolytic deposition of uranium and plutonium. A salt residue containing fission products and minor actinides is generated in this process as a waste. Such material have to be immobilized in a stable waste forms among which crystalline matrices are the most promising materials. We have used an argillaceous material that has high sorption affinity towards alkali metal ions as a base of crystalline matrix. The optimal temperature and annealing time were established to produce mechanically stable matrix with high load of LiCl-KCl eutectic. The compressive strength of the material is higher than 10 MPa. X-ray diffraction revealed the formation of new phases containing lithium after annealing. Morphology and crystallinity of the samples were investigated with scanning electron microscopy and high-resolution transmission electron microscopy. The samples were irradiation by different sources: gamma-quanta and electron beam. Irradiation of the samples up to a dose of 100 MGy did not result in any structural changes or compressive strength decrease regardless of the type of radiation. Leaching test demonstrate high chemical durability of the material.

Authors : Xiaoyan Jing, Wenting Li
Affiliations : (1) College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, P.R. China. (2) Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, 150001, PR China.

Resume : With the development of nuclear power in the world and the gradual industrialization of the country, the demand for uranium resources is also increasing. Capturing U(VI) from wastewater and seawater is highly attractive for the environment and clean energy with the increasing deficiency of land sources. Howbeit, the huge volume of water and the very low concentration of U(VI) pose a substantial challenge to the industrial application. Accordingly, we describe an organic-inorganic hybrid adsorbent with MOF particles grown on electrospun polyacrylonitrile fibers (PAN) covalently modified with amidoxime groups (AOPAN/ZIF). The synergistic reaction between Co-N bond and amidoximation improves the adsorption performance in a wide pH range, which is favorable for U(VI) capture under nuclear wastewater and seawater. As a result, the AOPAN/ZIF fiber presents high adsorption amount of 498.39 mg g-1 in U(VI) contaminated aqueous solution at pH 4. Furthermore, the amount of U(VI) reached 2.0 mg g-1 in natural seawater after contact of 36 days, which implies that the AOPAN/ZIF adsorbent may promote the development of U(VI) recovery from nuclear wastewater and seawater.

Authors : A. I. Popov, E.A. Kotomin, V.N. Kuzovkov
Affiliations : Institute of Solid State Physics, University of Latvia, 8 Kengaraga Str., Riga LV1063, Latvia

Resume : The The binary and complex oxide (MgO, ZnO, BeO, Al2O3, MgAl2O4, Y3Al5O12 etc) insulators as well as some other wide-band gap insulator and semiconductor compounds (AlN, BN, PLZT) are very important for many applications, including those when they are either exposed to radiationbeams or function in radiation environment. This includes a huge number of applications, such as optical materials, phosphors, laser active elements, dosimeter, imaging plates, materials for fusion technology etc. The radiation-induced defects determine and affect functional characteristics of the corresponding devices and/or their specific elements. Thus, it is very important to predict/simulate not only the kinetics of diffusion-controlled defect accumulation under irradiation, but also a longtime defect structure evolution including the thermal annealing of radiation-induced defects. After introducing some basics on the radiation point defects in halides, binary oxides and oxide perovskites [1] as well as the mechanisms of point defect formation under particle irradiation (neutron, ion, proton, electron), the current understanding of the point defect thermal annealing processes will be briefly reviewed. We will present recent results and discuss current understanding of the kinetics of the F-type center annealing in above-mentioned compounds after electron, heavy ions or neutron irradiation, which were treated as the bimolecular process with equal concentrations of the complementary F and Oi defects. The process is controlled by the interstitial oxygen ion mobility, which is much higher than that of the F centers. The appropriate Oi migration energies were obtained from available in the literature experimental annealing kinetics for electron, neutron and ion irradiated MgO, Al2O3, MgAl2O4, Y3Al5O12, BeO, ZnO, YSZ, PLZT etc. The results obtained are used for the evaluation of interstitial oxygen migration parameters and are compared with the available ab initio calculations. References 1. A.I. Popov, E.A. Kotomin, J. Maier, Nucl. Instr. Meth. B 268 (2010) 3084. 2. E.A.Kotomin, V.N. Kuzovkov, A.I. Popov, R. Vila, Nucl. Instr. Meth. B 374 (2016) 107. 3. E.A. Kotomin, V.N. Kuzovkov, A.I. Popov, J. Maier, R. Vila, J. Phys Chem A 122 (2018) 28. 4. V.N. Kuzovkov, E.A. Kotomin, A.I. Popov, J. Nucl. Mat. 502 (2018) 295.

Authors : S McGowan, C. Degueldre
Affiliations : Department of Engineering, University of Lancaster LA1 4YW

Resume : An analytical expression that evaluates the effect of the pH and the redox potential (Eh) was developed for studying the sorption of actinides onto substrates mimicking inorganic organic and bioorganic absorbers in seawater conditions. It includes surface complexation with one type of surface site and its formulation yields to a distribution coefficient (Kd) as function of the pH (hydrolysis) and Eh (redox sensitive species). The formulation considers also the values of the stability and hydrolysis constants for all species in solution and sorbed at the surface, and makes use of semi-empirical correlations between hydrolysis and surface complexation constants, for each surface. The presence of complexing ligands in solution (such as carbonates) was also taken into account. The model was applied to the sorption of uranium onto aluminol, hydrated iron oxide and silanol, as well as carboxylic and phenolic (both mono- and bi-) in the presence and in absence of carbonates in solution. The primary driver on this approach is to simulate the absorption of uranium species in seawater. When carbonates are present in solution the calculated values of the distribution coefficient were lower than those calculated in the absence of carbonates, and no redox effect was detected. The distribution coefficient (Kd) values obtained with the developed model are compared with values reported for the sorption of uranium onto specific absorbants from the literature. It is known that in the water stability region U(IV) and their hydroxides are the primary stable species in surface waters; however, some reduction effects are possible when interacting with the surface, which must be taken into account in the model. All the possible redox reactions of uranium were considered as a consequence. Moreover, this model is applicable to study the sorption of other redox sensitive elements of interest. This model will help to derive the best conditions for absorption of uranium from seawater.

Authors : J. O’Connell1, G. Aralbayeva2, V. Skuratov3,4,5, M. Saifullin3, A. van Vuuren1, A. Akilbekov2, M. Zdorovets6,7, A. Seitbayev2, A. Dauletbekova2
Affiliations : 1CHRTEM, NMU, University Way, Summerstrand, Port Elizabeth, South Africa 2L.N. Gumilyov Eurasian National University, Astana, Kazakhstan, 010008 3FLNR, JINR, Joliot-Curie 6, 141980 Dubna, Russia 4National Research Nuclear University MEPhI, Moscow, Russia 5Dubna State University, Dubna, Russia 6Institute of Nuclear Physics, Astana, Kazakhstan 7Ural Federal University, Yekaterinburg, Russia.

Resume : Nanosized hillock-like surface defects produced by swift heavy ions have been registered in Al2O3 and MgO, having a relatively high threshold of specific ionizing energy loss for structural disorder enhancement. Because these insulators are considered as candidates for inert matrix fuel hosts for fission reactors, SHI irradiation simulating a fission product’s impact is of considerable practical interest due to the large number of fission track recoils in reactor fuel. In this report we give a brief review of recent AFM and XTEM data on hillocks morphology in TiO2 and Al2O3 crystals induced by hundreds MeV Xe and Bi ions. It was found that all hillocks in TiO2 are crystalline and epitaxial with the original crystal surface while the hillocks in Al2O3 are amorphous. Analysis of rutile crystals irradiated in the temperature range 80-1000 K has revealed that average hillock height increased with irradiation temperature although large specimen-to-specimen and hillock-to-hillock size differences were also recorded. These experimental observations were discussed in the framework of a simple model based on the inelastic thermal spike (i-TS) model.

Authors : N. Khiara [1], L. Dupuy [1], F. Onimus [1], J.P. Crocombette [1], T. Pardoen [2], J-P. Raskin [2], Y. Bréchet [3]
Affiliations : [1] CEA Saclay, Gif-sur-Yvette, France ; [2] Université Catholique de Louvain, Louvain-la-Neuve, Belgique ; [3] INPG, Université Grenoble Alpes, 38402 Saint Martin d'Hères - France

Resume : Metals and alloys, such as stainless steels and zirconium alloys, used as structural materials in the nuclear core of pressurized water undergo irradiation creep deformation. At a macroscopic level, the mechanical behavior is well characterized. Yet, the underlying microscopic mechanisms are still unclear. Many theoretical mechanisms have been proposed in the literature, but only few experimental results were conclusive. Recent in situ TEM straining experiments under ion irradiation conducted on Zircaloy-4 have demonstrated that, at high stress levels, dislocations pinned on irradiation induced point defects clusters start to glide once under irradiation. One of the proposed hypotheses was that the observed dislocation glide assisted by irradiation was due to a direct interaction between the displacement cascade and the pinned dislocation. To test this hypothesis a molecular dynamics study in α-zirconium was conducted. The interaction of a screw dislocation with an interstitial loop in the prismatic plane was first studied. The objective was to pin the dislocation on the irradiation defect and form a helical turn, which is the configuration the most difficult to unpin. The atomic positions corresponding to a stress below the unpinning stress were then extracted and displacement cascade calculations were performed with a PKA (Primary Knock-on Atom) energy of 20 keV. It was shown that for a high applied stress and for particular PKA positions the displacement cascade can directly unpin the dislocation in agreement with the experiments. Based on these numerical simulations, a simple analytical probabilistic model was then proposed to explain the irradiation creep deformation under high applied stress.

Authors : A. I. Popov, E.A. Kotomin, V.N. Kuzovkov
Affiliations : Institute of Solid State Physics, University of Latvia, 8 Kengaraga Str., Riga LV1063, Latvia

Resume : The radiation-resistant binary oxides are very important materials for applications in fusion reactors. In this work, we analyzed the kinetics of the F-type and V center annealing after neutron irradiation. F-type center kinetics were treated as the bimolecular process with equal concentrations of the complementary F and Oi defects. Such process is controlled by the interstitial oxygen ion mobility, which is much higher than that of the F centers. The F center annealing begins at temperatures 500-700 K, when both F and F+ centers are practically immobile, due to the recombination with mobile Oi defects. It is demonstrated how the shape of the F-annealing curves is determined by two control parameters: Ea and effective pre-exponential factor and strongly depends on irradiation conditions. Special attention is paid to a detailed comparison of diffusion-controlled F center thermal annealing in neutron-irradiated BeO and ZnO with similar data obtained for for MgO, Al2O3 and MgAl2O4 [1-3]. The results obtained are used for the evaluation of interstitial oxygen migration parameters and are compared with the available ab initio calculations for MgO and Al2O3. Finally, we also treated the kinetics of F-center annealing in thermochemically-reduced crystals. The obtained activation energy allows to evaluate both the intrinsic F-center migration energy and also the conditions for metal colloid formation in BeO and ZnO. References: 1. E.A.Kotomin, V.N. Kuzovkov, A.I. Popov, R. Vila, Nucl. Instr. Meth. B 374 (2016) 107. 2. E.A. Kotomin, V.N. Kuzovkov, A.I. Popov, J. Maier, R. Vila, J. Phys Chem A 122 (2018) 28. 3. V.N. Kuzovkov, E.A. Kotomin, A.I. Popov, J. Nucl. Mat. 502 (2018) 295.

Authors : JunWang, Peipei Yang
Affiliations : (1) College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, P.R. China. (2) Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, 150001, PR China.

Resume : With the rapid growth of industrialization, water pollution caused by heavy metals has become one of the most serious environmental problems, and attracted considerable attention. Heavy metals are non-degradable and can accumulate in living tissues, so they must be removed from wastewater. Extraction of heavy metals from wastewater embraces reverse osmosis, chemical precipitation, electrodialysis, organic−inorganic ion exchange, and adsorption. Among these methods, adsorption stand out from the aforementioned methods due to its simple operation and cost-effectiveness. Thus far, the most efficient adsorbent for adsorption heavy metals have high adsorption capacity and removal rate. However, in recent years, the adsorbents designed for the capture of heavy metal have not been suitable for application at seawater pH. Currently, the design of a suitable wastewater adsorbent shows more promise for the extraction of heavy metals. In this report, we report a facile approach to construct a suitable wastewater pH and large surface area material that polyaniline (PANI) covalent grafted onto the surface of GO nanosheets. In the progress, transmission electron microscope (TEM), X-ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared spectroscopy (FTIR) were used to determine the effectiveness of the synthesis of GO-PANI composites. Meanwhile, the GO-PANI composites were investigated for adsorption of U(VI) from aqueous solution. It is clear that the GO-PANI composites have a high adsorption capacity (qm=589.4 mg g-1, T=298.15 K) at a suitable wastewater pH with a high removal rate (>95%). Based on the FTIR spectroscopy, Zeta potential, and X-ray photoelectron spectroscopy (XPS), a possible adsorption mechanism of U(VI) onto GO-PANI composites is revealed. Finally, the result also exhibited outstanding adsorption efficiency and adsorption capacity under the operating conditions for the adsorption-desorption of U(VI) from aqueous solution, which indicated a promising potential in the application of the absorbent in wastewater. (This paper is funded by the International Exchange Program of Harbin Engineering University for Innovation-oriented Talents Cultivation.)

Authors : Shengli Chen, David Bernard, Jean Tommasi, Cyrille De Saint Jean
Affiliations : CEA, Cadarache

Resume : Atomic displacement is one of the key factors that influence the behaviors of material properties during and after irradiation. For example, the life of a nuclear reactor is determined by the irradiation resistance of the reactor pressure vessel. Many models, including the international standard metric Norgett-Robinson-Torrens model (NRT), have been developed to calculate the Displacement per Atom (DPA) using the energy of Primary Knocked-on Atom (PKA) as a major parameter. However, extensive experiments and simulations indicate that the NRT-DPA model seriously overestimates (about 3 times) the actual DPA. Nordlund recently developed the Athermal Recombination-Corrected DPA (ARC-DPA) model. The ARC-DPA model shows that the Molecular Dynamics (MD) simulations can be directly used to compute DPA by fitting the simulated data for each isotope. We have developed an improved expression for the efficiency function to calculate the DPA without requiring fitting parameters as needed in the ARC-DPA model. Our DPA calculation results utilizing the improved efficiency function are validated against the experimental data for the Fe, Ni, and Cu. The applications in fast breeder nuclear reactors show good agreement with the ARC-DPA metric for 56Fe.

Authors : Michael T. Benson, Yi Xie, Jason M. Harp, Lingfeng He, James A. King, Jinsuo Zhang, Indrajit Charit, Samrat Choudhury
Affiliations : Michael T. Benson; Yi Xie; Jason M. Harp; Lingfeng He; James A. King - Idaho National Laboratory Jinsuo Zhang - Virginia Polytechnic Institute and State University Indrajit Charit; Samrat Choudhury - University of Idaho

Resume : Fuel-cladding chemical interaction (FCCI) occurs when the nuclear fuel or fission products react with the cladding material. A major cause of FCCI in metallic fuels during irradiation is fission product lanthanides, which tend to migrate to the fuel periphery, coming in contact with the cladding. The result of this interaction is degradation of the cladding, and will eventually lead to rupture of the fuel assembly. In order to extend fuel life and safely reach higher burn-up, a method to control the lanthanides, to either decrease or mitigate FCCI, is needed. Fuel additives are one method being investigated for this purpose. The rationale behind fuel additives is to have an element dispersed throughout the fuel matrix that will react with the lanthanides as they are produced. To date, Pd, Sn, Sb, Te, and In have been investigated as fuel additives. A comparison of these additives, in both U-Zr and U-Pu-Zr based fuels, will be presented, along with post-irradiation examination results for the Pd additive in a U-Zr fuel.

Authors : Qi Liu, Xuejie Guo
Affiliations : 1. College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, P.R. China. 2. Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, 150001, PR China.

Resume : Herein, a design of bifunctional hybrid adsorbent made of a magnetic core and a zeolite imidazolate framework (ZIF-8) shell embedded with carbon dots (CDs) is synthesized for U (VI) efficient removal. The carbon dots and magnetic core (Fe3O4) were prepared by one-pot hydrothermal approach using carboxymethyl cellulose (CMC) as precursor and stabilizer. Then the magnetic/luminescent bifunctional adsorbent was prepared successfully via in situ synthesis, and the ZIF-8 shell growth process has been explored at different conditions. The adsorbents retained the strong superparamagnetic behavior of Fe3O4 nanoparticles and favorable biocompatibility. The U(VI) adsorption on Fe3O4-CMC@ZIF-8 and Fe3O4-CMC@ZIF-8@CDs were investigated under various experimental conditions. Interestingly, the adsorbents not only clearly display a typical photoluminescence effect of carbon dots, but also obviously enhance U (VI) adsorption performance via assisting with the carbon dots. The adsorption processes of U(VI) on Fe3O4-CMC@ZIF-8 and Fe3O4-CMC@ZIF-8@CDs well simulated by pseudo-second-order model. The maximum adsorption capacity of U(VI) on Fe3O4-CMC@ZIF-8@CDs was calculated to be 606.06 mg/g at pH = 4.0 and T = 298 K, which was significantly higher than that of U(VI) on Fe3O4-CMC@ZIF-8 (564.97 mg/g). Particularly, BET, FTIR and XPS results show that the higher adsorption capacity of Fe3O4-CMC@ZIF-8@CDs was mainly attributed to higher specific surface area and more abundant surface oxygen-containing functional groups, suggesting that the main interaction mechanisms were surface complexation and electrostatic interactions. (This paper is funded by the International Exchange Program of Harbin Engineering University for Innovation-oriented Talents Cultivation.)

Authors : Elisa Re, Jérôme Maynadie, Daniel Meyer, Xavier Le Goff
Affiliations : Institut de Chimie Séparative de Marcoule, UMR 5257 CEA/CNRS/UM/ENSCM, Site de Marcoule – Bât. 426, BP 1717130207, Bagnols / Cèze cedex, France

Resume : Over the past two decades, nanoscience and nanotechnology have attracted increasing attention of the scientist community and industry. As a consequence of nanoscale, this type of materials exhibit size- and shape-dependent properties. Indeed, the physical and chemical properties show a lot of difference between the bulk and the nanoparticles (electronic, magnetic, optical, catalytic…). In 1993, Murray and al have reported a surfactant-assisted method in organic medium enable to produce highly monodisperse nanocrystals [1]. Since, scientists have devoted a significant effort on the improvement of this synthetic method for the production of size- and shape-controlled nano-objects witch can be envisaged as novel building blocks. These parameters are crucial for the organization of nano-objects in 2D or 3D architectures and for the study of their intrinsic and collective properties. This project aims to elaborate and characterize new nanostructured micrometric objects using two bottom-up approaches. The first one, called “one-pot”, is based on simultaneous growth and self-assembly of nanoparticles using a bridging organic ligand. The second one need two steps: the first step consists in nanoparticles synthesis and the second one involves nanoparticle surface modification by reactive molecules able to promote self-assembling by formation of covalent bridges (Click Chemistry) or by electrostatic interactions. Using the first synthesis method, hybrid nanomaterial was produced as highly organized sheets presenting a lamellar structure. The distance between sheets can be controlled by temperature and the inter-lamellar distance can be modulated by the nature of the organic linker. The self-assembly synthesis was used to obtain 3D nanostructured hybrid materials. The first part of this work has been done on uranium oxide. The next goal for this work is to assemble mixed actinide nanoparticles with platinoid nanoparticles in order to produce model materials. [1] C. B. Murray, D. J. Norris, et M. G. Bawendi, J. Am. Chem. Soc., 115, 19, p. 8706‑8715, (1993).

Authors : Yatendra Kumar Pandey, David Nesaraj, Gautam Biswas*
Affiliations : Nuclear Power Corporation of India Limited, Nabhikiya Urja Bhavan, Anushaktinagar, Mumbai – 400094, India

Resume : The current fleet of operating nuclear power reactors in India consists of 22 number of - small, medium and large size reactors. These include two boiling water reactors (BWRs, 160 MWe), sixteen small size pressurized heavy water reactors (PHWRs, 220 Mwe), two medium size PHWRs of 540 Mwe capacity and 2 VVER-1000 reactors. In addition, 6 reactors of 700 MWe PHWRs and 4 VVER-1000 reactors are under construction and a fleet of 10 reactors of 700 MWe PHWR are being taken up for construction shortly. Over the years, different types of fuel combinations have been used in these reactors. This paper talks about the experiences gained and future outlook with regard to fuel cycle management of Light Water Reactors. The paper focuses on the strategies adopted for the achievement of this goal. The twin BWRs at Tarapur, built by GE in 1969, are the oldest nuclear power reactors in the country. The better design of fuelling patterns and optimization of control rod withdrawal sequences have tremendously improved the fuel performance in these reactors. With the objective of utilization of Plutonium, the indigenously developed Mixed Oxide Fuel (MOX) was loaded in these reactors. The paper discusses experience gained with loading of MOX fuel in these reactors. Two units of VVER-1000 reactors at Kudankulam were constructed with Russian Cooperation. These units viz. Unit-1 & Unit-2 were commissioned in 2013 and 2016 respectively. Currently Unit-1 is operating in 4th cycle while unit-2 is operating in 2nd cycle. The fuel of design type UTVS is supplied by TVEL Corporation of Russian Federation. Commissioning tests performed during first (physical) startups and subsequent operation of both units demonstrated that fuel and core behaviour were in close agreement with theoretical estimations. In unit-1, the first set of thermal surveillance specimens (coupons) have been dismantled after Cycle-3 for material analysis. The paper brings out salient features of fuel performance for both units and future outlook with regard to further improvements.

Authors : Peipei Hu (a,b), Jie LIu (a), Jian Zeng(a), Pengfei Zhai (a), Shengxia Zhang (a),Lijun Xu (a,b), Zongzhen Li (a,b), Liu Li (a,b)
Affiliations : a.Institute of Modern Physics, Chinese Academy of Sciences (CAS), Lanzhou 730000, PR China; b.University of Chinese Academy of Sciences (UCAS), Beijing 100049, PR China;

Resume : This article reports a comprehensive study on damage evolution in InP crystal due to different species of swift heavy ions (SHI, Ar, Fe, Kr, Ta and Bi) irradiation. The damage effects caused by different ions were investigated by Raman spectroscopy. The significant red-shifts of LO mode were observed, which reveals the presence of lattice defects and stress induced during the irradiation process. The new peaks LO’ modes were clearly identified in the samples irradiated by SHIs. A peak point of the intensity ratio of LO’ peak to LO peak (ILO’/ILO) was detected, and the corresponding fluence decreased with the increase in (dE/dx)e. While in the case of Bi ions irradiation with a higher (dE/dx)e, the intensity of TO mode rapid increased with the increase in ion fluences. The growth of TO model is closely related to the defects caused by SHI irradiation. The presence of this mode also provides a new way to identify the nature defects in InP crystalline.

Authors : S. Piskunov, I. Isakovica, A. I. Popov
Affiliations : Institute of Solid State Physics, University of Latvia, 8 Kengaraga Str., Riga LV1063, Latvia

Resume : F centers play an important role in characterization and determination of the radiation resistance of functional materials for nuclear materials science. In this work the influence of the induced F-center to the atomic structure and electronic properties of Mn ion substituted for the host Al at YAlO3 have been studied from the first principles. The 2 x 2 x 2 supercell adopted for the bulk crystal phase of Pbnm symmetry allows us to simulate substitutional point defect with concentration of about 3%. To perform ab initio modeling of Mn-doped YAlO3 we were using approach of hybrid exchange-correlation functional HSE within density functional theory. Calculated atomic coordinates and lattice parameters, as well as bulk modulus and band gap for perfect YAlO3 are in a very good agreement with the most recent experimental observations. The electronic charge redistribution calculated for perfect orthorhombic YAlO3 crystal suggests notable covalency of the Al-O bond, that can be validated from further X-ray powder difraction analysis. This covalency is somewhat decreased in the Mn-O bond after doping. Formation of F0-center in the vicinity of the Mn dopant is accompanied by a well pronounced shift (0.17 A) of the nearest Al and Y atoms towards the vacancy. The F-center traps 0.54 e. The remaining 0.9 e from the missing anion are mainly localized on nearest Mn atoms forming the Mn2+ ion. Created Mn2+ ion induces four defect level in the middle of the band gap of perfect YAlO3, that certainly influences its electronic structure and gives us knowledge necessary to describe spectroscopic data obtained in the future. Funding from Latvian-Ukranian biletaral project is greatly acknowledged.

Authors : Kyuhong Lee, Wonjae So, Jonghwan Kim, Ki Nam Kim, Jung Min Do, Sunghwan Kim, Yong Jin Jeong, Jong Man Park
Affiliations : Korea Atomic Energy Research Institute, 111, Daedeok-daero 989Beon-gil, Yuseong-gu, Daejeon, Republic of Korea

Resume : HANARO (High-flux Advanced Neutron Application ReactOr) fuels were designed by KAERI and ACEL on the basis of the driver fuel for the NRU reactor in Canada. ACEL supplied the driver fuels to HANARO from the beginning of the rector operation. KAERI localized the HANARO fuels in 2005 and fabricated a total of 353 fuel assemblies so far. As a key manufacturing process, electron beam welding is applied to seal the gap between the cladding and the end plug of the fuel rod. KAERI installed the new electron beam welding machine and the welding jig to weld multiple fuel rods at the same time. Up to seven fuel rods can be mounted on the new jig and welded at one time. The number of vacuums compared to existing processes has been reduced to 1/14. Welded fuel rods were checked the welding integrity such as the presence of pores or cracks by X-ray inspection and liquid penetration test. Also, the microstructural analysis of the cross-section was conducted to inspect the penetration depth and the soundness of weldment. X-ray and microstructural analysis results showed that the rods was fully welded and satisfied with the manufacturing specification. KAERI successfully established the new electron beam welding system for HANARO fuel rods. As a result, the fabrication efficiency was remarkably improved.

Authors : Claude Degueldre, Daniel Ulloa Garza, Hans-Holger Wiese
Affiliations : Engineering Department, Lancaster University, LA1 4YW, UK Laboratory of Nuclear Materials, Paul Scherrer Institute, CH-5232Villigen Hot Laboratory, Paul Scherrer Institute, CH-5232Villigen

Resume : Calculations and measurements of the isotopic fraction of the fission gases released during burn-up in light water reactor have been successfully achieved as function of the burn up for the fuel and in the plenum. The burn-up explored are ranging from 0 to 100 MW h kg-1. The model calculation takes into account all isotope inventory considering the neutron absorption, fission, fission product build-up as well as precursor and fission gas isotope decay and their reactivity in the neutron flux of 0.1 x 1014, 0.3 x 1014 and 0.6 x 1014 cm-2 s-1 in both boiling water reactors and pressured water reactors. The experimental data are gained by puncture and mass spectroscopy analysis. Comparison of experimental and modeled data fit reasonably well for both Kr and Xe isotopes.

Authors : Matteo Vanazzi, Boris Paladino, Luca Ceseracciu, Gaelle Gutierrez, Celine Cabet, Jing Hu, Meimei Li, Fabio Di Fonzo
Affiliations : Center for Nano Science and Technology (CNST@PoliMi) - Istituto Italiano di Tecnologia (IIT); Dipartimento di Energia - Politecnico di Milano; Smart Materials, Nanophysics - Istituto Italiano di Tecnologia (IIT); DEN, Service de Recherches de Métallurgie Physique - CEA; DEN, Service de Recherches de Métallurgie Physique - CEA; Nuclear Engineering Division - Argonne National Laboratory (ANL); Nuclear Engineering Division - Argonne National Laboratory (ANL); Center for Nano Science and Technology (CNST@PoliMi) - Istituto Italiano di Tecnologia (IIT)

Resume : In order to qualify innovative materials for next generation nuclear systems, their radiation resistance must be assured. Irradiations studies with neutrons present overwhelming complications related to cost, availability and duration. Ions have been found as a valid alternative to simulate neutrons and gather useful data. Previously, we reported on amorphous Alumina (Al2O3) coatings under heavy ions irradiation up to 450 dpa, showing a general radiation-induced crystallization. In this work, we use different ions to irradiate up to 3 dpa. Here, we concentrate on the low dpa regime, to study the first stages of crystallization and obtain dpa values compatible with neutrons. Irradiations are now performed at different temperatures in order to decouple the thermal contribution from the radiation-induced effects. The evolution appears strictly temperature-dependent, with no structural changes until 500°C. Based on the experimental data, a preliminary kinetic model is proposed. From a mechanical point of view, an evident size-effect is manifested. The growth of nano-crystalline domains increase rapidly the hardness, in accordance with the inverse Hall-Petch trend. A second irradiation campaign is carried out with in-situ TEM tandem apparatus. Tests are repeated on free-standing films, to collect dynamically microstructural changes and phases transformation. To conclude, a consistent and coherent picture of the evolution of amorphous Al2O3 under irradiation is presented.

Authors : John David Simnett, Phil Holden, Claude Degueldre.
Affiliations : John David Simnett; Phil Holden'; Sustainable Energy Microsystems Community Interest Company ("Sustenerg") Claude Degueldre; Lancaster University

Resume : Sustenerg is designing a thorium-fueled MSR (“ThEA”) with 10 MW output weighing 5 tons and transportable. It would operate maintenance-free for 10 to 20 years and choice of materials is thus of paramount importance. One notable problem of molten salts is that they can corrode steel and melt aluminium, so non-traditional materials and manufacturing methods must be used. Attention must therefore be directed to formulating salt mixes (generally fluorides) which have the least corrosive properties and introducing low-corrosion materials from which the various parts of the reactor are constructed. With regard to quantity rather than quality of materials, nuclear reactors, including thorium MSRs, use only a fraction of the materials such as concrete and steel used by other forms of sustainable energy. Some of the applications of ThEA are as follows: SHIPS PROPULSION Maersk, the world’s largest container shipping company has decided to introduce zero-carbon ships from 2022 and achieve a completely carbon-free fleet by 2050. Small reactors such as ThEA could do this. DEVELOPING WORLD ThEA could be trucked or airlifted to remote areas and being air-cooled would need no access to bodies of water. DESALINATION Water for domestic and industrial use is becoming short in supply. Of the 193 nations recognized by the UNO, 145 have access to the sea. Production of fresh from seawater is therefore an option for the majority of the world’s countries and energy for this could be provided by ThEA. SPACE APPLICATIONS Establishment of lunar and Martian colonies is now being considered. ThEA would be suitable for powering these since it is light enough to be taken by rocket. It could also power orbiting space stations, permitting a wider range of zero-gravity industries.

Authors : C. Martin1, K.H. Miller2, D. Craciun3, R. Martin1, G. Dorcioman3, M. D. Dracea4, D. Pantelica4, B. S. Vasile5, and V. Craciun3, 6
Affiliations : 1Ramapo College, New Jersey, USA; 2NASA Goddard Space Flight Center, Greenbelt, MD, USA; 3National Institute for Lasers, Plasma and Radiation Physics, Măgurele, Romania; 4Horia Hulubei National Institute for Physics and Nuclear Engineering, Măgurele, Romania; 5Faculty of Applied Chemistry and Material Science, Polytechnic University of Bucharest, Bucharest, Romania; 6Extreme Light Infrastructure-Nuclear Physics, Magurele, Romania

Resume : We measured broadband optical reflectance (50 meV to 6 eV) of 1 MeV Au-irradiated ZrN thin films. The films were deposited by pulsed laser deposition technique using a KrF laser on Si substrates heated at 500 °C. Grazing and symmetrical incidence X-ray diffraction investigations showed that films were nanocrystalline, with grain size around 10 nm, dense and (111) textured. Rutherford backscattering spectrometry investigations showed that films were slightly Zr rich and their chemical composition was not changed by Au irradiation. X-ray photoelectron spectroscopy investigations found a low (<2 at. %) oxygen contamination in bulk. Results showed that an irradiation fluence of 1×10E15 cm-2 enhances the far and mid infrared reflectance, and shifts slightly to higher frequency the plasma edge associated with free (conduction) carriers. Lorentz-Drude analysis confirms that gold irradiation increases slightly the carrier concentration, hence acts as dopant, but it also enhances significantly the spectral weight of the impurity localization band in mid infrared. Another important effect of irradiation is the improvement of the plasmonic performance of ZrN films, especially in the visible range.

Authors : G. Kauric, A.L. Smith, S.Bordier, S.Gossé, C. Guéneau
Affiliations : DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette cedex, France ; Delft University of Technology, Faculty of Applied Sciences, Radiation Science & Technology Department, Reactor Physics and Nuclear Materials (RPNM), Mekelweg 15, 2629 JB Delft, The Netherlands ; DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette cedex, France ; DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette cedex, France ; DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette cedex, France

Resume : Study of the Cs-Na-Mo-O system for the safety assessment of the Sodium-cooled Fast Reactor: From experimental measurements to Calphad modelling. G. Kauric a, A.L. Smith b, S.Bordier a, S.Gossé a, C. Guéneau a a DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191 Gif-sur-Yvette cedex, France b Delft University of Technology, Faculty of Applied Sciences, Radiation Science & Technology Department, Reactor Physics and Nuclear Materials (RPNM), Mekelweg 15, 2629 JB Delft, The Netherlands The chemistry of irradiated mixed oxide fuel U1-xPuxO2 in a Sodium-cooled Fast Reactor (SFR) is complex due to the accumulation of fission products and progressive increase of the oxygen potential. The high temperature gradient between the centre and the periphery of the fuel pellet induces the migration of the most volatile fission products (Cs, I, etc) towards the pellet rim. This leads to the formation at high burnup of a layer called “Joint Oxide Gain” filling the gap between the irradiated fuel and the cladding. In the potential event of a clad breach, the metallic sodium would interact with the fuel, and hence with this layer, mainly composed by the fission products: cesium, barium, molybdenum, iodine, tellurium, with phases such as Cs2MoO4, CsI, Cs2Te, BaMoOx, etc. For a thorough safety assessment of the SFR, the possible reactions between sodium and the compounds of the irradiated fuel, in particular cesium molybdate, need to be investigated. In fact, cesium and molybdenum are produced with high fission yields and cesium molybdate is one of the most stable compound formed at high burnup in the JOG under operating conditions of the reactor. Therefore, the interaction between sodium, cesium and molybdenum will start at the beginning of the accident. Moreover, in case of a severe accident, the fast increase of the fuel temperature can lead to the fission product release and to the fuel melting. However, these extreme temperature conditions are challenging to reproduce in the laboratory. Therefore, a thermodynamic model is needed to be able to predict the phases formed under these extreme conditions. In this work, the investigation of the system Cs-Na-Mo-O is presented. For such multicomponent systems and large scale of compositions and temperatures, the Calphad method is the most suitable to analyse the chemical interaction between fission products, sodium and the phases formed in solid, liquid and gas states. CALPHAD, which stands for CALculation of PHAse Diagram, is a semi-empirical method that enables to develop a thermodynamic model based on the Gibbs free energy of the gas, liquid and solid phases as a function of the temperature, composition and pressure of the system. It describes the equilibrium state of the system by modelling the total Gibbs energy of the system, expressed as a linear function of the Gibbs energy of the phases in the whole system. To do so, adjustable parameters are optimised to fit the experimental data available (structural and thermodynamic properties or phase diagram data). Recently, Zolotova et al. [1] performed a Differential Scanning Calorimetry (DSC) study of the pseudo-binary phase diagram Cs2MoO4-Na2MoO4 and reported a new quaternary compound Cs3Na(MoO4)2 with a phase transition at low temperature (around 390°C). While studying the thermodynamic properties of the newly determined quaternary compound, Smith et al. [2] did not find a low temperature phase which is not consistent with the work of Zolotova et al. As this system is key for the safety assessment of the SFR, a reinvestigation has been performed in this work in order to develop an accurate model of the pseudo-binary section and then of the quaternary system Cs-Na-Mo-O. To solve the discrepancy, a new DSC study of the pseudo-binary Cs2MoO4-Na2MoO4 system has been performed in this work. The results are consistent with the ones obtained by Zolotova et al. [3], except for the high enriched part in Na2MoO4, where the liquidus determined is closer to the phase diagram reported in the PhD work of F.Tête [4]. Moreover, we did not observe the low temperature phase transition of the quaternary compound in agreement with the study of Smith et al. [2]. In addition, the enthalpy of fusion of the quaternary compound Cs3Na(MoO4)2 is reported for the first time, and the hexavalent state of molybdenum in Cs3Na(MoO4)2 has been confirmed with an X-ray absorption spectroscopy experiment. These data combined with the ones measured by Smith et al. [2] [5] have been used to develop an accurate model of the pseudo-binary section using the Calphad method. In the system Cs-Na-Mo-O, another quaternary compound exists, namely CsNaMo3O10. In this work, an X-Ray diffraction pattern of the compound has been reported for the first time. To refine the structure by the Rietveld method, the compound RbNaMo3O10 (space group Pnma 62) was used as starting model. The isostructurality of the two compounds can be expected from the fact that the ionic radii of rubidium and cesium are very close [6]. Moreover, to determine precisely the Mo-O bond distances, an X-Ray Absorption Spectroscopy experiment has been performed. As the compound was pure enough to measure its thermodynamic properties, a DSC analysis has also been performed. Thanks to this measurement, the melting temperature and enthalpy of fusion have been determined and no phase transition has been observed. To summarize, a DSC study of the pseudo-binary phase diagram Na2MoO4-Cs2MoO4, has been carried out in order to solve the discrepancies raised in previous studies. Rietveld refinements of the two quaternary compounds in the Cs-Na-Mo-O system, namely Cs3Na(MoO4)2 and CsNaMo3O10, have been made combined with X-ray Absorption Spectroscopy measurements to confirm the hexavalent state of molybdenum and determine precisely the Mo-O bond distances. As the compounds were pure enough for thermodynamic analyses, DSC studies on Cs3Na(MoO4)2 and CsNaMo3O10 have been done. The enthalpies of fusion of the two quaternary compounds have been measured and the melting temperature of the CsNaMo3O10 compound has been determined for the first time. Thanks to this study and the data reported in the literature, an accurate model of the Cs-Na-Mo-O phase diagram has been developed for the first time. [1] E.S. Zolotova, Z.A. Solodovnikova, V.N. Yudin, S.F. Solodovnikov, E.G. Khaikina, O.M. Basovich, I.V. Korolkov, I.Y. Filatova, Phase relations in the Na2MoO4 –Cs2MoO4 and Na2MoO4 –Cs2MoO4 –ZnMoO4 systems, crystal structures of Cs3Na(MoO4)2 and Cs3NaZn2(MoO4)4, J. Solid State Chem. 233 (2016) 23–29. [2] A.L. Smith, G. Kauric, L. van Eijck, K. Goubitz, N. Clavier, G. Wallez, R.J.M. Konings, Structural and thermodynamic study of Cs3Na(MoO4)2: Margin to the safe operation of sodium cooled fast reactors, J. Solid State Chem. 269 (2019) 1–8. doi:10.1016/j.jssc.2018.08.033. [3] E.S. Zolotova, Z.A. Solodovnikova, V.N. Yudin, S.F. Solodovnikov, E.G. Khaikina, O.M. Basovich, I. V Korolkov, I.Y. Filatova, Journal of Solid State Chemistry Phase relations in the Na2MoO4 – Cs2MoO4 and, J. Solid State Chem. 233 (2016) 23–29. doi:10.1016/j.jssc.2015.10.008. [4] F. Tête, La réaction Cs2MoO4/ Na : Application à l’intéraction combustible / sodium lors d’une rupture de gaine à fort taux de combustion dans un RNR, Université de Provence, 1999. [5] A.L. Smith, M.-C. Pignié, L. van Eijck, J.-C. Griveau, E. Colineau, R.J.M. Konings, Thermodynamic study of Cs3Na(MoO4)2: Determination of the standard enthalpy of formation and standard entropy at 298.15 K, J. Chem. Thermodyn. 120 (2018) 205–216. doi:10.1016/j.jct.2018.01.016. [6] Shannon-Prewitt Effective Ionic Radius - Part 3 | The Elements Handbook at KnowledgeDoor, KnowledgeDoor. (n.d.). (accessed December 4, 2018).

Authors : A. V. Ochkin, S. V. Stefanovsky
Affiliations : D. Mendeleev University of Chemical Technology; Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences

Resume : The nuclear fuel reprocessing is based on PUREX process. Usually 30% TBP in hydrocarbon diluents is applied. Then the composition of the organic phase can be expressed as H2O-HNO3-UO2(NO3)2-Pu(NO3)4-TBP-diluent. But the plutonium concentration is low and thus the system can be considered as a five-component one. This system has been considered early [1]. Modeling of the system includes few points: 1. The application of mole fractions and volume ones and rational activity coefficients, 2. The calculation of concentration of free (unbounded with solvates) water. 3. The methods of activity coefficient calculation. 4. The methods to estimate errors during the calculation of various coefficients. Gibbs-Duhem equation is used to calculate dependence of component “i” activity on concentration of other components xidlnai +∑ xjdlnaj = 0 (1) xidlnfi +∑ xjdlnfj = 0 (2) The dodecane is commonly used as a diluent. The main system is usually divided into sub-systems and every sub-system is calculated through thermodynamic activities. The model of H2O-HNO3-TBP-dodecane was given in [2]. Activities of nitric acid and uranyl nitrate were calculated in [3-5]. In order to calculate the system H2O-HNO3-UO2(NO3)2-TBP-dodecane it is necessary to determine a formation constant of di-solvate UO2(NO3)2·2TBP and its error. 1. Ochkin, A., Gladilov, D., Nekhaevskiy, S. Procedia Chemistry 7, 315-321 (2012). 2. Ochkin, A., Gladilov, D., Nekhaevskiy, S.Theoretical Foundations of Chemical Engineering, 2015, Vol. 49, No. 5, pp. 649–655. 3. Davis W., Lawson P.S., de Bruin H.J., Mrochek J. J. Physical. Chem. 1965. V. 69. P.1904. 4. Yu Yang-Xin, Zhang Qing-Yin, Gao Guang-Hua. J. Radioanal. Nuclear Chem. 2001. V.245. P. 581. 5. Ochkin A., Merkushkin A., Nekhaevskiy S., Tiupina E. Radiochemistry, 2016, Vol. 58, No. 3, pp. 280–286.

Authors : A.V. Ochkin, S.V. Stefanovsky
Affiliations : D. Mendeleev University of Chemical Technology; Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences

Resume : The reprocessing of spent fuel is usually performed by the PUREX process [1]. The process currently includes the extraction of uranium, plutonium, and neptunium with 30% tri-n-butylphosphate (TBP) in a hydrocarbon solvent. In this process, uranium, plutonium, and neptunium are concentrated in the organic phase, and americium, curium, and fission products are concentrated in the aqueous phase. High_level radioactive wastes (HLRW) are further formed based on the aqueous phase. The concentration of target elements in HLRWs does not exceed 0.01, 0.025, and 0.5% with reference to uranium, plutonium, and neptunium, respectively. HLRW specific activities were determined by 90Sr and 137Cs fission products and americium and curium radionuclides. Now it is necessary to isolate americium and curium radionuclides. The process includes the following stages: (1) the separation of a mixture of rare_earth elements (fission products), americium, and curium from the aqueous phase of the extraction process; (2) the extraction separation of americium and curium from rare_earth elements. Here, there are the two possible variants: 1) the coextraction of americium and curium – DIAMEX – process; 2) the extraction of americium only, whereas curium remains in the aqueous phase EXAm -process. 1. Ochkin, A.V., Some problems in reprocessing of fuel spent by modern power reactors. Theor. Found. Chem. Eng., 2014, vol. 48, pp. 34–38.

Authors : Hengbin Xu, Milin Zhang
Affiliations : Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, 150001, PR China

Resume : The potential consequences of nuclear events and the complexity of nuclear waste management have recently received extensive attention in the nuclear energy field. While extraction of radioactive uranium (U(VI)) from seawater/wastewater further motivate the development of selective sorbents. Herein, a facile yet versatile strategy for fabrication of adsorbents were reported. The fabricated mesoporous magnesium hydroxide (Mgs) material was obtained via the in situ conversion of the natural ore powder (magnesite) precursor, MgCO3. The unique internal pore structure to be well suited as a platform for the deployment of highly efficient sorbents, and combined with carboxymethyl cellulose (CMC) lock water features that they exhibit remarkable performance and production cost advantages. Scanning electron microscopy (SEM), X-ray diffraction (XRD), and Fourier transform infrared spectroscopy (FTIR) were used to determine the effectiveness of the synthesis of CMC/Mgs composites. The adsorption behavior of U (VI) by the porous CMC/Mgs was studied by static adsorption experiments, and also the effects of adsorption time, pH of wastewater/seawater and U (VI) concentration were discussed. The results showed that the CMC/Mgs composites had a high adsorption capacity at pH=5 or pH=8 with a high removal rate of U(VI) at the ppb and ppm levels. (This paper is funded by the International Exchange Program of Harbin Engineering University for Innovation-oriented Talents Cultivation.)

Authors : Neil Hyatt, Amber Mason, Lucy Mottram, Shi Kuan Sun, Martin Stennett and Claire Corkhill.
Affiliations : Department of Materials Science & Engineering, The University of Sheffield, Mappin Street, Sheffield, S1 3JD, UK.

Resume : X-ray absorption spectroscopy was applied to understand the speciation of elements relevant to the immobilization and disposal of radioactive plutonium bearing wastes, utilizing Ce as a Pu surrogate. Cl K-edge XANES studies of zirconolite glass-ceramics, designed for immobilization of Pu residues, demonstrated incorporation within the aluminosilicate glass phase as the chloride anion, likely co-ordinate to Na as a network modifying species, below the Cl solubility limit of 1.0 wt%. Ce L3 XANES characterization of the crystallised glass material produced by cold crucible plasma vitrification, at demonstration, evidenced incorporation as Ce3+ within the glass phase, providing an important validation of laboratory scale studies. U L3 and Ce L3 XANES investigation of brannerite (UTi2O6) glass ceramics, synthesized under oxidizing, neutral and reducing conditions, established the charge compensation mechanism as incorporation of Ce3+ through formation of U5+. In each of these examples, X-ray Absorption Spectroscopy has provided a pivotal understanding of element speciation in relation to the mechanism of incorporation within the host waste form intended for geological disposal.

Authors : Ki Nam Kim, Jong Hwan Kim, Sunghwan Kim, Kyuhong Lee, Yong Jin Jeong, Jong Man Park
Affiliations : Korea Atomic Energy Research Institute

Resume : Mo-99 decays to Tc-99m, which is the most widely used radiopharmaceutical isotope for medical diagnostic purposes. Recently, Mo-99 producers have been attempting to replace conventional highly enriched uranium (HEU) targets with low enriched uranium (LEU) targets by international non-proliferation policies. As a result, it is necessary to develop high-uranium-density targets with LEU to improve the Mo-99 production efficiency of LEU targets. Korea Atomic Energy Research Institute (KAERI) has been developing commercial LEU targets with a uranium density of 2.6 gU/cc and high-density LEU targets using atomized U-Al alloy powder. Atomization is a key technology for achieving high-uranium-density because atomized powder is able to have various U-Al compositions and a high uranium content. We successfully mass-produced uranium alloy powder for high-density targets through centrifugal atomization and fabricated a high density target with a uranium density of 3.2 gU/cc by using U-15Al powder. The development of higher density target fabrication technology (over 4.0 gU/cc; above 50% improvement in uranium density over commercial targets), including forming, rolling, heat treatment, and non-destructive testing, is currently in progress.

Authors : Shouzheng Su 1, Hong xia 1 and Jun Wang 1,2,3
Affiliations : 1 College of Nuclear Science and Technology, Harbin Engineering University, Harbin 150001, China; 2 Key Laboratory of Superlight Material and Surface Technology, Ministry of Education, Harbin Engineering University, Harbin 150001, China; 3 College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, China;

Resume : A novel hierarchical micro/nano composite adsorbent NiCo-LDHs/ZIF-67, in which nanocrystalline ZIF-67 was immobilized on the surface of flower-like NiCo-LDHs by an in situ growth method, was fabricated for the extraction of uranium from seawater. Compared with NiCo-LDHs and ZIF-67, NiCo-LDHs/ZIF-67 exhibited much higher adsorption capacity for uranium at pH 8.0, which was close to natural seawater pH. Electrostatic interactions and possible "ligand separation" mechanism was proposed to explain this phenomenon. The adsorption process of NiCo-LDHs/ZIF-67 fitted well with Langmuir isotherm model and pseudo-second-order model. Moreover, NiCo-LDHs/ZIF-67 exhibited remarkable adsorption performance at low concentration of U(VI) (μg/L) in simulated seawater. This paper provided a new insight for designing the effective adsorbent for U(VI) recovery at seawater pH.

Authors : V.N. Kuzovkov (a), A.I. Popov (a), E.A. Kotomin (a), M. Izerrouken (b), R. Villa (c)
Affiliations : (a) Institute of Solid State Physics, University of Latvia, Riga, Latvia, (b) Nuclear Research Center of Draria, Algiers, (c) CIEMAT, Madrid, Spain

Resume : The radiation-resistant oxide insulators (Al2O3, MgAl2O4, Y3Al5O12 etc) are important materials for applications in fusion reactors, e.g. in optical windows. It is very important to predict/simulate not only the kinetics of diffusion-controlled defect accumulation under neutron irradiation, but also a long-time defect structure evolution including thermal defect annealing after irradiation. Primary radiation defects in ionic solids consist of Frenkel defects—pairs of anion vacancies with trapped electrons (F-type centers) and interstitial ions. Using phenomenological theory of diffusion-controlled recombination of the F-type centers with much more mobile interstitial ions (complementary hole centers) [1], we suggested theory, how to extract from experimental data the migration energy of interstitials and pre-exponential factor of diffusion. The correlation of these two parameters satisfies the so-called Meyer–Neldel rule (MNR) [2] observed more than once earlier in glasses, liquids, and disordered materials, but not in the irradiated materials. Our results [2] allow to establish a direct relation between irradiation fluence and its post-irradiation thermal annealing. We analyzed here the available experimental kinetics of the F-type center annealing in a wide temperature range (300-1000 K) for two different ionic solids: neutron/ion-irradiated Al2O3 (corundum) [1,2] and ion-irradiated Y3Al5O12 (YAG) [3]. We compare results for corundum and YAG -- both wide gap insulating materials but with different crystalline structures. As the result, it is demonstrated for corundum that with the increase of radiation fluence both the migration energy and pre-exponent are decreasing, irrespective of the type of irradiation. This is MNR with normal dose dependence. For YAG we have confirmed MNR, but the dose dependence is inverse. We discuss the cause of this phenomenon. Thus, in this study, we demonstrated that the dependence of defect migration parameters on the radiation fluence plays an important role in the quantitative analysis of the radiation damage of real materials and cannot be neglected. [1] V.N. Kuzovkov, E.A. Kotomin, A.I. Popov, R.Villa, Nucl. Inst. Meth. B 374, 107 (2016). [2] E. Kotomin, V. Kuzovkov, A.I. Popov, J. Maier, R. Vila, J. Phys. Chem. A, 122, 28 (2018). [3] M. Izerrouken, A. Meftah, M. Nekkab, Nucl. Instr. Meth. B 258, 395 (2007).

Authors : R. Rakesh(1), V. P. Sinha(1), and R. Tewari(2)
Affiliations : (1) Metallic Fuels Division, (2) Materials Science Division, Bhabha Atomic Research Centre, Mumbai 400085

Resume : U-Mo alloys are promising alternative as high density fuel material for conversion of Research & Test reactors from HEU to LEU under the GTRI (RERTR) initiative. In U-Mo, presence of metastable -phase (bcc) is required to make the alloys swelling resistant and stable under irradiation environment. In this alloy system, the high temperature -phase undergoes a eutectoid transformation at around 560oC generating a two phase lamellae of -U and U2Mo ( commonly known as ' phase). It has been reported in literature that increasing Mo content in the alloy causes beneficial delay in the decomposition of the -phase during an isothermal ageing heat treatment. Detailed crystallographic analysis carried out in the present study has shown that all the three phase display structural relationship among them. This uniquely defined structural relationship is a consequence of the strain associated with the phase transformation. Such strain further retards the already kinetically slow decomposition of the -phase. This aspect allows tweaking the composition in such a way that the -phase can be stabilized even in dilute off-eutectoid alloys. In the present study, it has been analysed that the maximum in the gamma stabilizing effect in terms of suppressing γ’formation can be realised in the U-8Mo alloy. Systematic study of the decomposition behaviour of three alloys viz. U-8Mo, U-9Mo & U-10Mo were carried out at 500°C & 350°C to understand the progressive decomposition behaviour of these alloys. It is noteworthy to observe the cellular decomposition from metastable γ-phase in the form of discontinuous precipitation reaction (U+ U) rather than direct eutectoid decomposition to the stable α-U and γ’(U2Mo) phases. The in situ transformation of γ interlamellar regions to the γ’-phase was noticed occasionally within the γ + α lamellar colonies with specific orientation relationship with both & phases. Many such transformation related issues will be discussed in the present paper.

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Structural materials : NN
Authors : Tong Liu
Affiliations : China Nuclear Power Technology Research Institute

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.

Authors : Huseyin Sener Sen, Tomas Polcar
Affiliations : ADVAMAT Group, Department of Control Engineering, Czech Technical University in Prague, Technicka 2, 16627 Prague 6, Czech Republic

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.

Authors : Junkai LIU, Wenbo LIU, Di YUN, Guanghai BAI, Meiyi Yao, Zhixin XIA
Affiliations : Xi'an Jiaotong University; Xi'an Jiaotong University; Xi'an Jiaotong University; Suzhou Nuclear Power Research Institute; Shanghai University; Suzhou University;

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.

Authors : Kenneth C. Littrell
Affiliations : Oak Ridge National Laboratory

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.

Authors : Clio Azina, Andrejs Petruhins, Igor Zhirkov, Binbin Xin, Johanna Rosen, Per Eklund
Affiliations : Thin Film Physics Division, Department of Physics, Chemistry and Biology (IFM), Linköping University, SE-58183 Linköping, Sweden

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.

10:00 Coffee Break    
Authors : Chaewon Kim(a), Hyunmyung Kim(a), Lee Sung Yong(b), Changheui Jang(a)*
Affiliations : (a) Korea Advanced Institute of Science and Technology (KAIST), Daejeon 34141, Republic of Korea; (b) KEPCO Nuclear Fuel, Daejeon 34057, Republic of Korea

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.

Authors : Hyunmyung Kim(a), Chaewon Kim(a), Hun Jang(b), and Changheui Jang(a)*
Affiliations : (a) Korea Advanced Institute of Science and Technology (KAIST), Daejeon 34141, Republic of Korea; (b) KEPCO Nuclear Fuel, Daejeon 34057, Republic of Korea

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.

Authors : P. Dzhumaev (1), V. Borodin (1,2), O. Emelianova (1,3), M. Ganchenkova (1), A. Gentils (3) , V. Polsky (1), V. Yakushin (1)
Affiliations : (1) National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Moscow, Russia; (2) NRC "Kurchatov Institute", Moscow, Russia; (3) CSNSM, Université Paris-Sud, CNRS/IN2P3, Université Paris-Saclay, Orsay, France

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.

11:00 Plenary    
12:30 Lunch    
Authors : Lumin Wang
Affiliations : Department of Nuclear Engineering and Radiological Sciences, Department of Materials Science and Engineering, University of Michigan-Ann Arbor, Michigan 48109, USA

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.

Authors : Wahida R Ilaham and Tapas Laha
Affiliations : Department of Metallurgical and Materials Engineering, Indian Institute of Technology Kharagpur, India

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

Authors : Maria A. Auger, David T. Hoelzer, Kevin G. Field, Michael P. Moody
Affiliations : Physics Department, Universidad Carlos III de Madrid. Avenida Universidad 30, 28911 Leganes, Madrid, Spain; Oak Ridge National Laboratory, Oak Ridge, TN 37831, Tennessee, USA; Oak Ridge National Laboratory, Oak Ridge, TN 37831, Tennessee, USA; Department of Materials, University of Oxford. Parks Road OX1 3PH, Oxford, UK

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.

Authors : O. Emelyanova (1,2), A. Gentils (1), M.G. Ganchenkova (2), V.A. Borodin (2,3), P.V. Vladimirov (4), P. Dzhumaev (2) , I. Golovchanskiy (5), A. Moeslang (4)
Affiliations : (1) CSNSM, Université Paris-Sud, CNRS/IN2P3, Université Paris-Saclay, Orsay, France; (2) National Research Nuclear University MEPhI, Moscow, Russia; (3) NRC "Kurchatov Institute", Moscow, Russia; (4) Institute of Applied Materials, Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany; (5) National University of Science and Technology MISIS, Moscow, Russia

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.

Authors : Haiming Wen1, Andrew K. Hoffman1, Rinat K. Islamgaliev2
Affiliations : 1 Missouri University of Science and Technology, USA; 2 Ufa State Aviation Technical University, Russia

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.

Authors : N. Castin, A. Dubinko, D. Terentyev, M. Konstantinovic
Affiliations : Institute of Nuclear Material Science, SCK•CEN, Boeretang 200, Mol, Belgium;

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.

Authors : Ziqiang Dong; Jingli Luo; Zihan Wang, Ming Zhang
Affiliations : Ziqiang Dong, Materials Genome Institute, Shanghai University; Jingli Luo, Chemical and Materials Engineering Department, University of Alberta; Zihan Wang, Materials Genome Institute, Shanghai University; Ming Zhang, Materials Genome Institute, Shanghai University

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.

15:30 Coffee Break    
Authors : Cihang Lu 1, Takaaki Koyanagi 2, Yutai Katoh 2, Kurt A. Terrani 2, Nicholas R. Brown 3
Affiliations : 1 Pennsylvania State University, University Park, PA USA; 2 Oak Ridge National Laboratory, Oak Ridge, TN USA; 3 University of Tennessee-Knoxville, Knoxville, TN USA

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.

Authors : J. M. Marshall, D. Walker and P. A. Thomas
Affiliations : University of Warwick

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) 4 Windsor and Morgan. Nucl.Fus. (2018) 57(8) DOI: org/10.1088/1741-4326/aa7e3e

Authors : Chenyang Lu1,2, Fei Gao2, Yanwen Zhang3, Lumin Wang2
Affiliations : 1. Department of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049, China 2. Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109, United States 3. Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831, USA

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.

Authors : Haixuan Xu
Affiliations : The University of Tennessee, Knoxville

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.

Authors : Rebecca Gray, Michael Rushton, Samuel Murphy
Affiliations : Lancaster University; Bangor University; Lancaster University

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.

Authors : Boohyun An, Yongsun Yi, Pyungyeon Cho, Taeyeon Kim*, Paul Rostron, Akram AlFantazi
Affiliations : Khalifa University, Abu Dhabi, UAE. *presenting author

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.

Fuel analysis and post irradiation management I : NN
Authors : J. Rory Kennedy
Affiliations : Idaho National Laboratory

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

Authors : K. O. Kvashnina
Affiliations : 1 Rossendorf Beamline at ESRF – The European Synchrotron, CS40220, 38043 Grenoble Cedex 9, France. 2 Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden, Germany

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.

Authors : Evgeny Gerber, Anna Romanchuk, Ivan Pidchenko, Christoph Hennig, Alexander Trigub, Stephan Weiss, Andreas Scheinost, Andre Rossberg, Gavin Vaughan, Lucia Amidani, Stepan Kalmykov, Kristina Kvashnina
Affiliations : Rossendorf Beamline at ESRF – The European Synchrotron, CS40220, 38043 Grenoble Cedex 9, France Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden Lomonosov Moscow State University, Department of Chemistry, 119991 Moscow, Russia; Lomonosov Moscow State University, Department of Chemistry, 119991 Moscow, Russia; Rossendorf Beamline at ESRF – The European Synchrotron, CS40220, 38043 Grenoble Cedex 9, France Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden; Rossendorf Beamline at ESRF – The European Synchrotron, CS40220, 38043 Grenoble Cedex 9, France Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden; National Research Centre “Kurchatov Institute”, 123182 Moscow, Russia; Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden; Rossendorf Beamline at ESRF – The European Synchrotron, CS40220, 38043 Grenoble Cedex 9, France Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden; Rossendorf Beamline at ESRF – The European Synchrotron, CS40220, 38043 Grenoble Cedex 9, France Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden; European Synchrotron Radiation Facility (ESRF), Grenoble, 38000, France; Rossendorf Beamline at ESRF – The European Synchrotron, CS40220, 38043 Grenoble Cedex 9, France Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden; Lomonosov Moscow State University, Department of Chemistry, 119991 Moscow, Russia; Rossendorf Beamline at ESRF – The European Synchrotron, CS40220, 38043 Grenoble Cedex 9, France Helmholtz Zentrum Dresden-Rossendorf (HZDR), Institute of Resource Ecology, PO Box 510119, 01314 Dresden;

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.

Authors : Lucia Amidani, Ivan Pidchenko, Kristina Kvashnina
Affiliations : Helmholtz-Zentrum Dresden-Rossendorf, ROBL, Bautzner Landstraße 400, 01328 Dresden

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+.

Authors : Neil Hyatt, Martin Stennett, Sarah O'Sullivan, Lewis Blackburn, Chris Dixon Edwards, Amber Mason, Lucy Mottram and Claire Corkhill.
Affiliations : Department of Materials Science & Engineering, The University of Sheffield, Mappin Street, Sheffield, S1 3JD, UK.

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.

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Fuel analysis and post irradiation management II : NN
Authors : Kirk Sorensen
Affiliations : CEO of Flibe Energy, Inc., Huntsville, Alabama, USA

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.

Authors : Di YUN, Wenhua ZHANG, Xin XIE, Wenbo LIU
Affiliations : Xi'an Jiaotong University; Xi'an Jiaotong University; Xi'an Jiaotong University; Xi'an Jiaotong University;

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.

Authors : Parswajit Kalita, Santanu Ghosh, Gaël Sattonnay, Gaëlle Gutierrez, Udai B. Singh, Pawan K. Kulriya, Vinita Grover, Rakesh Shukla, A.K. Tyagi, Devesh K. Avasthi
Affiliations : Dept. of Physics, Indian Institute of Technology Delhi, New Delhi – 110016, India; Dept. of Physics, Indian Institute of Technology Delhi, New Delhi – 110016, India; LAL, Université Paris-Sud, Bât 200 F-91405, Orsay, France; CEA Saclay, DEN, SRMP, Labo JANNUS, 91191 Gif-sur-Yvette, France; Dept. of Physics, Deen Dayal Upadhyaya Gorakhpur University, Gorakhpur – 273009, India; Materials Science Group, Inter University Accelerator Centre, New Delhi – 40085, India; Chemistry Division, Bhabha Atomic Research Centre, Mumbai – 400085, India; Chemistry Division, Bhabha Atomic Research Centre, Mumbai – 400085, India; Chemistry Division, Bhabha Atomic Research Centre, Mumbai – 400085, India; Amity Institute of Nanotechnology, Amity University, Noida – 201313, India

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)

Authors : M.Garrigue, A. Quaini, C.Guéneau, C.Bonnet, T. Alpettaz
Affiliations : DEN-Service de la Corrosion et du Comportement des Matériaux dans leur Environnement (SCCME), CEA, Université Paris-Saclay, F-91191, Gif-sur-Yvette, France

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.

Authors : N. Dacheux 1, T. Cordara 1, S. Szenknect 1, L. Claparede 1, A. Mesbah 1, C. Lavalette 2, R. Podor 1,
Affiliations : 1- ICSM, CNRS, CEA, ENSCM, Univ Montpellier 2- AREVA NC/BG aval/DO recyclage/RDP, Direction AREVA

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

Authors : K. V. Khishchenko
Affiliations : Joint Institute for High Temperatures RAS, Moscow, Russia

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.

10:00 Coffee Break    
Authors : Riley J. Parrish1, Karen E. Wright2, Alexander J. Winston2, Jason M. Harp2, Assel Aitkaliyeva1
Affiliations : 1. University of Florida, Department of Materials Science and Engineering, Gainesville, FL, USA 2. Idaho National Laboratory, Idaho Falls, ID, USA

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.

Authors : F. Cappia B. Miller D. Murray L. He J. Harp
Affiliations : Idaho National Laboratory

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.

Authors : Julia Hidalgo, P. Roussel, T. Delahaye, G. Leturcq, J.L. Rouviere,
Affiliations : CEA, Nuclear Energy Division, DMRC, SFMA, F-30207, Bagnols sur Cèze, France ; UCCS, USTL-ENSCL-EC Lille, 59655 VIlleneuve d’Ascq, France; CEA, Nuclear Energy Division, DMRC, SFMA, F-30207, Bagnols sur Cèze, France ; CEA, Nuclear Energy Division, DMRC, SFMA, F-30207, Bagnols sur Cèze, France ; CEA-Grenoble, Department de Recherche Fondamentale sur la Matière Condencée/SP2M, 17 rue des Martyres, 38054 Grenoble Cedex 9, France;

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.

Authors : Zhang Yan, Yang Suliang, Liu Qian, YangZhihong, Zhao Yonggang, Tian Guoxin*
Affiliations : China Institute of Atomic Energy

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.

Authors : Yang Suliang, Yang Zhihong, Zhang Yan, Zhou Jin, Liu Qian, Tian Guoxin*
Affiliations : China Institute of Atomic Energy

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.

Authors : A. Michel, G. Carlot, C. Sabathier, C. Onofri, M. Dumont, M. Cabie
Affiliations : CEA / DEN / DEC, Saint Paul Lez Durance, France ; CEA / DEN / DEC, Saint Paul Lez Durance, France ; CEA / DEN / DEC, Saint Paul Lez Durance, France ; CEA / DEN / DEC, Saint Paul Lez Durance, France ; IM2NP – UMR CNRS 7334 – Aix-Marseille Université, Marseille, France ; CP2M – Aix-Marseille Université Marseille, France

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

Authors : Peter Andersson, Haluk Atak, Anastasios Anastasiadis
Affiliations : Department of Physics and Astronomy, Uppsala University; Department of Nuclear Engineering, Hacettepe University; Department of Physics and Astronomy, Uppsala University

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.

Authors : P. Andersson 1, V. Rathore 1, E. Andersson Sundén 1, H. Atak 2, S. Holcombe 3, A. Håkansson 1, P. Jansson 1, J. Nyberg 1
Affiliations : 1. Dep. Of Physics and Astronomy, Uppsala University, 2. Dep. of Nuclear Engineering, Hacettepe University 3. Dep. Of Nuclear Fuels and Materials Experiments, Institute for Energy Technology

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.

Authors : J.P. Ramos, N.-T. Vuong, A. Gottberg, R. Catherall, T. Stora
Affiliations : CERN (Switzerland) and KU Leuven (Belgium), CERN and EPFL (Switzerland), TRIUMF (Canada), CERN (Switzerland), CERN (Switzerland)

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.

Authors : Zhang‒Mi Lin
Affiliations : Key Laboratory of Superlight Materials and Surface Technology, Ministry of Education, College of Materials Science and Chemical Engineering, Harbin Engineering University, Harbin 150001, China

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)

12:45 Lunch    
Waste forms : NN
Authors : S. Delpech, G. Duran-Klie and D. Rodrigues
Affiliations : Institut de Physique Nucléaire - CNRS/IN2P3 Univ. Paris Sud, Université Paris Saclay, 91406 Orsay, France

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.

Authors : Seoung Woo Kuk, SangGyu Park, Kyung Chai Jeong, Seok Jin Oh, Ki Hwan Kim, Jeong-Yong Park

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.

Authors : Mr Joseph Lillington, Professor Ian Farnan
Affiliations : Department of Earth Sciences, University of Cambridge (Both Authors)

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.

Authors : S.V. Stefanovsky, O.I. Stefanovsky, S.V. Yudintsev, A.V. Ochkin
Affiliations : Frumkin Institute of Physical Chemistry and Electrochemistry

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.

Authors : Jeremy Causse, Cyrielle Rey, Sandrine Dourdain, Xavier Deschanels
Affiliations : ICSM, UMR 5257, CEA, CNRS, ENSCM, Univ. Montpellier, Marcoule, 30207 Bagnols sur Cèze, France

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.

15:30 Coffee Break    
Authors : Karin Popa; Jean-François Vigier; Laura Martel; Dario Manara; Daniel Freis; Rudy J.M. Konings
Affiliations : European Commission, Joint Research Centre, P.O. Box 2340, D-76125 Karlsruhe, Germany

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.

Authors : Danwen Qin, Adel Mesbah, Nicolas Clavier, Stéphanie Szenknect, Nicolas Dacheux
Affiliations : ICSM, CNRS, CEA, ENSCM, Univ Montpellier, Site de Marcoule, BP 17171, 30207 Bagnols/Cèze, France

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.

Authors : Nicolas Dacheux 1 , Paul Estevenon 1,2 , Eleonore Welcomme 2 , Stéphanie Szenknect 1 , Adel Mesbah 1 , Philippe Moisy 2 , Christophe Poinssot 2
Affiliations : 1-ICSM, CNRS, CEA, ENSCM, Univ Montpellier, Site de Marcoule, Bat 426, BP 17171, 30207 Bagnols sur Ceze, France 2-CEA, Nuclear Energy Division, DMRC, BP 17171, 30207 Bagnols sur Ceze, France

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

Authors : S.V. Stefanovsky, S.V. Yudintsev, B.S. Nikonov, O.I. Stefanovsky, M.V. Skvortsov
Affiliations : Frumkin Institute of Physical Chemistry and Electrochemistry of the Russian Academy of Sciences; Institute of Geology of Ore Deposits, Mineralogy, Petrography, and Geochemistry of the Russian Academy of Sciences

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.

Authors : Sergei M. Butorin
Affiliations : Department of Physics and Astronomy, Uppsala University, Sweden

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).

Authors : Peilong Dong, Giuseppe Scatigno, Mark Wenman
Affiliations : Imperial College London, EDF, Imperial College London

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 and 2.6 x 10-2 Above 2.6 x 10-2 the crack density reduced from the peak value and at around 4.2 x 10-2 few cracks were observed. Cross-sectional analysis showed that average crack velocities were relatively consistent (between 1 and 2 μ 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.

Authors : N.-T. Vuong 1 2, A. Gottberg 3, Sanjib Chowdhury 4, A. P. Gonçalves 4, K. Sivula 2, T. Stora 1
Affiliations : 1 – CERN, 1211 Genève 23, Switzerland; 2 – École Polytechnique Fédérale de Lausanne, EPFL, 1015 Lausanne, Switzerland; 3 –Present address : TRIUMF, 4004 Wesbrook Mall, Vancouver, B.C. V6T 2A3, Canada; 4 – C2TN Instituto Superior Técnico, Universidade Lisboa, Estrada Nacional 10, Km 139.7, 2695-066 Bobadela LRS, Portugal

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.

Authors : Shotaro Akahori (physician in Japan) Motoyasu Kinoshita
Affiliations : Standard Power Ltd

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.

Authors : Chaitanyamoy Ganguly
Affiliations : Pandit Deendayal Petroleum University (PDPU) Gandhinagar , Gujarat , India

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.

18:00 Poster Awards - Final discussion    
19:00 Graduate Student Award ceremony followed by the social event    

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Symposium organizers
Chaitanyamoy GANGULYUniversity Gandhinagar

Department of Nuclear Science and Technology, Gujarat 382007, India

+91 8008222016
Claude DEGUELDRELancaster University

Engineering Department, Lancaster, UK

+ 44 1524 592716

Locked Bag 2001, Kirrawee DC, NSW 2232, Australia

+61 2 9717 3742
Rodney C. EWINGStanford University

Center for International Security and Cooperation:, University of Stanford, CA, USA

+1 650 725 8641