Symposium FK
Materials Challenges for Sustainable Nuclear Fission and Fusion Technologies


Session FK-1 - Structural Components for Nuclear Fission and Fusion Applications

FK-1:IL01  Novel Materials and Advanced Design Concepts for DEMO Divertor Targets
JEONG-HA YOU, B. BÖSWIRTH, H. GREUNER, M. LI, A.V. MÜLLER, Max Planck Institute for Plasma Physics, Garching, Germany; E. VISCA, F. CRESCENZI, S. ROCCELLA, ENEA, Depart. Fusion & Technology for Nuclear Safety, Frascati, Italy; CH. VORPAHL, EUROfusion, PMU PPPT, Garching, Germany; T. BARRETT, CCFE, Culham Science Centre, Abingdon, UK; F. GALLAY, M. RICHOU, CEA, IRFM, Saint Paul Lez Durance, France; J. REISER, KIT, IAM, Eggenstein-Leopoldshafen, Germany

One of the major R&D focuses in the European fusion power plant (DEMO) program is to establish a physical as well as technological basis for reliable power exhaust during normal and off-normal operational scenarios. In this regard, the most crucial engineering challenge is to develop robust high-heat-flux components (such as divertor targets) being capable of withstanding extreme surface heat flux up to 20MW/m2 with a sufficient heat removal capacity, structural reliability under neutron irradiation and sufficient lifetime against armor erosion. In the framework of EUROfusion DEMO divertor project (WPDIV), integrated R&D efforts have been conducted to develop advanced design concepts and related key technologies for the DEMO divertor targets. Seven different novel design concepts have been devised for water-cooled divertor targets and the technology R&D is currently progressing. One of the ultimate goals of the WPDIV is to deliver the experimental verification of the required high-heat-flux fatigue performance of these design concepts using small scale mock-ups and neutral beam irradiation facility. In this contribution, a brief overview is presented on the engineering issues of DEMO divertor target, the underlying design rationales, R&D activities for mock-up production, non-destructive inspection methodology and the selected results of the latest high-heat-flux test campaign. Finally, structural failure behaviors are discussed predicted by FEM simulations.

FK-1:IL02  High-temperature Fracture Behaviour of High Chromium Ferritic-martensitic and Nanostructured Ferritic Alloys
THAK SANG BYUN, JUNG PYUNG CHOI, Pacific Northwest National Laboratory, Richland, WA, USA; D.T. HOELZER, Oak Ridge National Laboratory, Oak Ridge, TN, USA; S.A. MALOY, Los Alamos National Laboratory, Los Alamos, NM, USA

Ferritic iron-chromium alloys with ultrafine structures have been prime candidates for future fast reactor and fusion reactor components because of their excellent high temperature strength and radiation damage resistance. The quenched and tempered ferritic-martensitic (FM) steels are hardened primarily by ultrafine laths and precipitates while the powder metallurgy nanostructured ferritic alloys (NFA) are hardened by nanograin structure and nanoclusters. The different hardening processes provide a wide variety of fine microstructures and mechanical properties. This presentation aims to elucidate the differences and similarities in the temperature and strength dependences of fracture behavior of these iron-chromium alloys. The fracture toughness (KJC) versus strength curves confirmed that the fracture toughness of an alloy is inversely proportional to the strength of the alloy. However, some NFAs did not follow this trend and showed outstandingly high fracture toughness after the toughening process. The majority of NFAs have generally higher strength and lower fracture toughness when compared to FM steels. However, some NFAs after toughening processing and FM steels after tailored thermomechanical processing showed comparable strengths as well as fracture toughnesses.

FK-1:IL04  Tungsten Powder Injection Molding @ KIT: Achievements and Trends
S. ANTUSCH, Karlsruhe Institute of Technology, Eggenstein-Leopoldshafen, Germany

In the framework of the European material development programme for fusion power plants beyond the international thermonuclear experimental reactor (ITER), tungsten is being considered as a potential plasma-facing material. This is a result of its favourable properties, such as a high melting point of 3422 °C (3695 K), high strength, high thermal conductivity (173 Wm-1K-1 at room temperature), low tritium inventory, low thermal expansion, low activation, low erosion rate and high temperature yield strength. However, the use of tungsten and tungsten alloys is limited by recrystallisation and brittle-to-ductile transition. Also, an interesting point is the time and cost aspect for tungsten manufacturing. As the production of tungsten parts with conventional technologies is extremely time and cost intensive, a promising fabrication method in view of large-scale production is powder injection moulding (PIM). In the framework of the European material development project (EUROfusion), PIM are intensively investigated at Karlsruhe Institute of Technology (KIT). With its high near-net-shape precision, it`s offer the advantage of reduced costs compared to conventional machining. PIM as a special process allows for the mass production of components, the joining of different materials without brazing and the creation of composite and prototype materials, as well as being an ideal tool for scientific investigations. For example, the addition of oxide or carbide particles into the tungsten matrix enhances its mechanical properties, such as ductility and strength, in comparison to pure tungsten. PIM materials are isotropic with equiaxed grain orientation. One of the possibilities for manufacturing tungsten parts has been shown by the latest produced samples for diagnostics for the French tokamak WEST (Tungsten (W) Environment in Steady-state Tokamak). This contribution describes the characterization and analyses of prototype materials produced via PIM. The investigation of pure tungsten and oxide or carbide doped tungsten materials comprises the microstructure examination, element allocation, texture analyses, and mechanical testing. In addition, fabricated near-net-shape Langmuir probes for diagnostics for the French tokamak WEST will be presented.

FK-1:IL05  The Theory of Precipitation Hardening Revisited: The Effect of Crystal Structure on the Obstacle Strength
YOSHITAKA MATSUKAWA, Tohoku University, Sendai, Japan

Establishing a mechanism-based predictive model of material embrittlement (loss of ductility) is a challenge of nuclear materials research community; however, this is also a long-standing challenge in fundamental physical metallurgy. Although the theory of dislocation is well established for quantitatively describing the strength of materials, the dislocation theory is incapable of directly describing the ductility thus far. Hence, the loss of ductility has often been indirectly scaled by the degree of hardening, based on a generally-accepted empirical rule that stronger materials exhibit less ductility. In the spirit that improving the quantitative precision of the modeling of hardening is a contribution to the precision improvement of the material lifetime prediction, we have revisited the theory of precipitation hardening by using advanced material characterization techniques. Some recent updates are as follows. (1) The crystal structure of precipitates is a factor dominating their obstacle strength against gliding dislocations. (2) The obstacle strength of precipitates changes during precipitation. In a very early stage of precipitation, precipitates have the same crystal structure as that of the matrix rather than the final product of precipitation.

FK-1:L06  Improvement of Density and Strength of CVI-processed SiCf/SiC Composites by Applying SiC Nanowires
DAEJONG KIM, HO WOOK LEE, SEUNGHO LEE, HYEON-GEUN LEE, JI YEON PARK, WOEN-JU KIM, Korea Atomic Energy Research Institute, Daejeon, South Korea

The SiCf/SiC composite has been considered as a core material for various reactors because it has good high-temperature strength and radiation resistance. However, chemically vapor infiltrated (CVI)-SiCf/SiC composites, which are known to have the highest radiation resistance among the SiCf/SiC composites produced by other manufacturing routes, have high porosity whish results in the reduction in strength, toughness and thermal conductivity. In order to increase the density of CVI-processed SiCf/SiC composites, a new deposition site must be provided in large space between the fiber layers in which no fibers are present. In this study, the nanowire was grown in fibers using a CVD method to reduce the porosity, and the composite was prepared by CVI. As a result, it was possible to improve the density of the SiC composite and in particular to greatly reduce the pore size and population of pores between the layers. The thermal conductivity and mechanical strength of SiCf/SiC composites with nanowires were measured.

FK-1:L07  Development of the Nanostructured Ferritic Alloy OFRAC (Fe-12Cr-MoTiNb) for Fast Reactor Advanced Fuel Cladding
D.T. HOELZER, C.P. MASSEY, K.A. TERRANI, Oak Ridge National Laboratory, Oak Ridge, TN, USA

Fuel cladding for sodium fast reactors will require advanced alloys that possess high performance, such as creep, in extreme environments consisting of high-temperatures and neutron doses up to 500 dpa. However, the prospective advanced alloy must be able to be fabricated into thin wall fuel cladding. Over the past 15-20 years, numerous nuclear materials programs around the world have been involved with development of nanostructured ferritic alloys (NFA), such as 14YWT, due to the presence of a high concentration of nano-size, oxygen-enriched particles, or nanoclusters, and ultra-fine grain structure that combine for high sink strength that improves their radiation tolerance. The development of a new NFA, referred to as OFRAC, is based on Fe-12Cr with solute additions of Mo, Ti and Nb. The goal of developing OFRAC is to combine all of the important characteristics mentioned above. This presentation will cover the processing of OFRAC, the tensile properties and microstructure stability after annealing at 1050 and 1150ºC for 8 h, the creep performance at temperatures up to 800ºC obtained from strain-rate jump tests and the recent pilger campaign that successfully resulted in a 1.78 m long thin wall tube that has an outside diameter of 8 mm and wall thickness of 0.5 mm.

FK-1:L08  Microstructural Evolution of Oxide Dispersion Strengthened Alloys under Temperature and Stress
JINSUNG JANG, TAE KYU KIM, WOO GON KIM, CHANG HEE HAN, Korea Atomic Energy Research Institute, South Korea; XIAODONG MAO, Institute of Nuclear Energy Safety Technology, CAS, China; MAN WANG, HEUNG NAM HAN, Seoul National University, South Korea

Oxide dispersion strengthened (ODS) alloys have been good candidate materials for the key components of Gen IV nuclear systems owing to their superior radiation resistance and the good high temperature properties. The excellent performances are usually attributed to the nano-sized stable oxide particles dispersed within the matrix. 12Cr Fe-based ODS steel and alloy 617-based ODS alloy samples were prepared through mechanical alloying (MA) with the addition of yttria (Y2O3) oxide particles of around 30 nm in diameter. Hot consolidation processes such as hot isostatic pressing (HIP), powder extrusion or hot rolling were carried out. After normalizing or solution heat treatment two ODS alloy samples were creep rupture tested at 700 degree C under 70 MPa and at 950 degree C under 25 MPa, respectively. Microstructural change of grain morphology, grain size distribution, precipitation and oxide particles as well as the matrix phase after creep rupture are investigated using scanning electron microscope/electron backscattered diffraction (SEM/EBSD), high resolution transmission electron microscope (HR TEM) on the sample gauge and grip section to examine the effects of stress at the temperatures.

Session FK-2 - Low Activation Structural Materials for Nuclear Fusion Systems

FK-2:IL02  Expanding the Operation Window of RAFM Steels by Optimized Chemical Compositions and Heat Treatments
J. HOFFMANN, M. RIETH, M. KLIMENKOV, S. BAUMGÄRTNER, Karlsruhe Institute of Technology, Karlsruhe, Germany

The improvement of 9%-Cr reduced activation steels, especially the extension of the operation limits, is the present scope of the EUROfusion materials project for advanced steels. Within this programme, new alloys are designed and fabricated to overcome some of the limitations of EUROFER97. In the present study, three 9%-Cr alloys with some variations in the chemical compositions are compared to standard EUROFER97. The main focus lies in the extension of the operation window to both higher and lower temperatures. This is achieved by a variation of the amount of the minor alloying elements which form precipitates and secondary phases. Combining the modified compositions with special heat treatments for the specific alloys leads to a refined distribution of the carbide and nitride phases. Variations in the alloy compositions open the possibility for extended temperature windows for heat treatments (higher tempering temperature). The predicted changes in precipitate composition and size were confirmed by TEM studies. Lower carbon contents proved to be effective to reduce the amount of M23C6 carbides.. Microstructural changes caused by the element variation and/or different treatments were characterized by SEM combined with EBSD mapping.

FK-2:IL03  Sputter-erosion of Low-activation Steel
M. OBERKOFLER1, R. ARREDONDO PARRA1, M. BALDEN1, S. ELGETI1, H. GREUNER1, W. JACOB1, M. MAYER1, R. NEU1, T. SCHWARZ-SELINGER1, T.F. SILVA1,2, KAZUYOSHI SUGIYAMA1, U. VON TOUSSAINT1, 1Max Planck Institute for Plasma Physics, Garching, Germany; 2Instituto de Física da Universidade de São Paulo, São Paulo, Brazil

The suitability of low-activation steels for recessed areas of the plasma-facing wall in a fusion reactor is being investigated. These steels contain small amounts of tungsten. By preferential sputtering a W-enriched surface layer is formed that could lower the erosion rate upon exposure to energetic deuterium ions from the plasma to acceptable levels. We report on simulations and experiments aiming at quantifying the achievable reduction of erosion rates. An extensive series of experiments at the IPP high current ion source has been dedicated to the erosion of EUROFER steel by mono-energetic D ions. Various ion beam analysis techniques have been employed to determine the depth profile within the W-enriched surface layer. Dedicated simulations with SDTrimSP fail to reproduce the experimental observations quantitatively. Surface morphology modifications that are induced by the exposures hint at a possible influence of surface roughening on the experimentally observed reduction in sputtering yields. Further experiments that are reported on are the exposure of an actively cooled EUROFER tile to a high-flux hydrogen beam in GLADIS as well as the performance with respect to erosion of steel tiles mounted in ASDEX-Upgrade.

FK-2:L04  Treatment of Constructional Materials of Fuel Rod Cladding and Fuel Assemblies in the Processing of Spent Nuclear Fuel of the Reactor Facility Brest-300
V. KASCHEEV, High-Tech Institute of Inorganic Materials, Moscow, Russia

The possibilities of radiochemical reprocessing of irradiated Constructional Materials (CM) and the disposal of radioactive waste arising at the handling of irradiated fuel element shells and CM of fuel assemblies. As an example, the characteristics of the induced activity of irradiated in reactor facility BREST-OD-300 ferritic-martensitic EP-823 and EK-181 steels are analyzed. It is shown that the type (chemical composition) of the steel considerably affects the activation of steel during irradiation in a reactor unit. In the selection of CM in addition to solving the basic problem - ensuring the tightness of nuclear fuel in the reactor, it is necessary to take into account the subsequent treatment of the radioactive waste generated. Reduced content of Ni, Co, Nb and Mo in the EK-181 steel forms ( in difference from EP-823) isotopic composition induced by irradiation in the reactor facility, which provides a more rapid decline of dose and heat release with time in irradiated EK-181 steel. The gain in time of dose recession is ultimately reflected in the way of dealing with waste generated and permits to reduce the radiation burden on staff. In terms of future radioactive waste management the use of EK-181 steel as fuel rod cladding and CM of fuel assemblies is more preferable than EP-823 steel.

Session FK-3 - Materials for First Wall Components of Nuclear Fusion Systems

FK-3:IL01  Tungsten Materials for Plasma Facing Components - Status and Research Directions

Tungsten has attractive properties and disadvantages as a material used in plasma-facing components (PFCs) for fusion. While thermal, sputtering, and tritium retention properties are favorable, oxidation and mechanical properties are not favorable for its use in the divertor region. Oxidation resistance is relevant for safety and licensing, but poor mechanical properties is the critical property. It is well-known that tungsten will embrittle under irradiation so that fracture toughness of tungsten-based PFCs is a critical parameter. It is apparent that pure tungsten or tungsten-based alloys are unable to achieve the necessary toughness for use. Therefore, research has focused on tungsten-based composites with engineered toughness as a potential solution to this problem. The status of tungsten research and development in the international fusion materials program is reviewed and discussed with regard to tungsten-based composites, including tungsten wire, foil and laminated structures, graded materials, coatings, and ductile phase toughened composite concepts. New oxidation resistant tungsten alloys are discussed along with an overview of tungsten thermophysical and thermomechanical testing. New developments are considered including additive manufacturing and 3D printing methods.

FK-3:IL02  Advanced Tungsten Materials for Plasma-facing Components of Future Fusion Devices
R. NEU, A. FEICHTMAYER, H. GIETL, Max-Planck-Institut für Plasmaphysik, Garching, Germany, and Technische Universität München, Garching, Germany; J. RIESCH, M. BALDEN, S. ELGETI, T. HOESCHEN, M. LI, S. OLBRICH, Max-Planck-Institut für Plasmaphysik, Garching, Germany; J. Almanstötter, OSRAM GmbH, SP PRE PLM DMET, Schwabmünchen, Germany; J.W. COENEN, Y. MAO, L. RAUMANN, Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung - Plasmaphysik, Partner of the Trilateral Euregio Cluster (TEC), Jülich, Germany

Plasma-facing materials of future fusion power plants will face unique challenges in terms of power, particle and neutron loads. Tungsten (W) is the main candidate material for the first wall as it is resilient against erosion, has the highest melting point of any available metal and shows rather benign behaviour under neutron irradiation. However, W has a fairly high DBTT even increasing under neutron irradiation. To overcome this brittleness, a W-fibre enhanced W-composite material (Wf/W) has been developed incorporating extrinsic toughening mechanisms. The fibres used are drawn potassium doped W wires as used in the lighting industry characterized by very high strength (>2500 MPa), ductility already at room temperature and recrystallization only above 1900°C. Industrial textile techniques have been successfully established to prepare the fibre preform. Large samples (~5x5 cm², more than 2000 long fibres) have been produced by layered W chemical vapour deposition and infiltration. The fibres must be coated by a temperature stable material in order to avoid interdiffusion between matrix and fibres and to allow for the intended extrinsic toughening. The latest results on the production and characterization of the compounds and their expected benefits will be presented.

FK-3:IL03  Tungsten Alloys for Reduced Oxidation under Accident Conditions in Fusion
C. GARCIA-ROSALES, A. CALVO, N. ORDÁS, I. ITURRIZA, Ceit-IK4 and Tecnun (University of Navarra), San Sebastian, Spain; K. SCHLÜTER, R. NEU, M. BALDEN, H. GREUNER, Max-Planck-Institut für Plasmaphysik, Garching, Germany; K. SCHLÜTER, R. NEU, Technische Universtät München, Garching, Germany; F. KLEIN, G. PINTSUK, A. LITNOVSKY, T. WEGENER, Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung - Plasmaphysik, Jülich, Germany; E. Tejado, J.Y. Pastor, Polytechnic University of Madrid, Madrid, Spain

Tungsten (W) is the main candidate material for the first wall armor of future fusion reactors. However, a loss-of-coolant accident with simultaneous air ingress could lead to a temperature rise of the in-vessel components exceeding 1000°C, resulting in the formation of volatile and radioactive W oxides. A way to mitigate this safety issue is the addition of oxide-forming elements, which, in presence of oxygen at high temperatures, promote the formation of a passivating scale protecting W from further oxidation. In this work, bulk alloys of the W-Cr-Y system with different concentrations of alloying elements are studied. Microstructural investigations of bulk material and scale developed after oxidation, as well as thermal conductivity and mechanical properties are presented. Results of different tests performed under the operational conditions expected in a fusion power plant are reported, including oxidation tests under isothermal and accident-like conditions; high heat flux tests at up to 2 MW/m², addressing the load expected at the first wall; and thermal shock tests to simulate loads under transient events. W-Cr-Y alloys exhibit oxidation rates 3 to 4 orders of magnitude lower than those of pure W and a thermal shock resistance comparable to pure W reference material.

FK-3:IL04  Conclusions drawn from Plasma Operation with Beryllium and Tungsten Plasma-facing Components in JET and Linear Plasma Devices
S. BREZINSEK1, D. BORODIN1, M. RUBEL2, R. DOERNER3 and PFC and JET contributors, 1Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung -Plasmaphysik, Jülich, Germany; 2Fusion Plasma Physics, Royal Institute of Technology (KTH), Stockholm, Sweden; 3Center for Energy Research, University of California at San Diego, La Jolla, CA, USA

The ITER material selection consists of Beryllium Plasma-Facing Components (PFCs) in the main chamber and Tungsten PFCs in the divertor. Under harsh conditions a variety of plasma-surface interaction (PSI) processes such as material erosion, transport, deposition as well as fuel retention by implantation and co-deposition will take place and potentially recrystallization and melting will occur. The combination of plasma operation in a tokamak with this material mixture, namely JET ITER-like wall, and a set of linear plasma devices (PISCES-B, PSI-2, MAGNUM) mimicking the first wall and divertor conditions are providing an ideal tool box to study these processes. Though they cannot reproduce the integral physics picture in ITER, they can be used to verify models by detailed experiments in order to predict what can be expected in ITER in its different operational phases. We present in details some key observations such as e.g. the fuel retention, the erosion yields of applied Be and W PFCs in JET and linear plasmas devices, etc. usually as function of surface temperature and local plasma conditions. The power handling and impact of recrystallization and melting on it will be discussed. Conclusions for ITER are drawn by direct experimental extrapolation or by modelling predictions.

FK-3:IL05  Overview of a Comprehensive First Mirror Test in the JET Tokamak for ITER
M. RUBEL1, SUNWOO MOON1, P. PETERSSON1, A. WIDDOWSON2 and JET Contributors, 1Royal Institute of Technology (KTH), Stockholm, Sweden; 2CCFE, Culham Science Centre, Abingdon, UK  

In a reactor-class device so-called first mirrors will be the plasma-facing components in all optical systems for plasma diagnosis. The First Mirror Test (FMT) has been carried out at the JET tokamak since year 2005, first in the presence of the carbon walls (JET-C) and then (since 2011) in JET with the ITER-Like Wall (JET-ILW) constituted of metals: beryllium and tungsten. Before and after exposure mirrors underwent detailed surface analysis using spectrophotometers, high resolution microscopy and a number of ion beam methods including nuclear reaction and heavy ion elastic recoil detection analysis. This work is focused on the modification of molybdenum test mirrors exposed in JET-ILW during either a single campaign or all three campaigns in 2011-2016. The main results show that: (i) reflectivity of all mirrors from the divertor was degraded by 50-90%, while much lower loss of performance occurred on mirrors from the main chamber wall; (ii) surfaces of most samples are covered with co-deposits, dust particles and – in some cases – beryllium splashes from molten limiters. The implications of these results for the first mirror maintenance in a reactor-class device will also be discussed.
*See the list in: X. Litaudon, Nucl. Fusion 57 (2017) 102001.

FK-3:IL06  New Materials and Composites for Fusion Reactor First Wall Components
Ch. LINSMEIER, J.W. COENEN, J. RIESCH*, M. BRAM, J. ENGELS, S. HEUER, A. HOUBEN, B. JASPER, F. KLEIN, A. LITNOVSKY, Y. MAO, G. PINTSUK, L. RAUMANN, M. RASINSKI, J. SCHMITZ, X. TAN, T. WEGENER, Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung - Plasmaphysik, Partner of the Trilateral Euregio Cluster, Jülich, Germany; *Max-Planck-Institut für Plasmaphysik, Garching, Germany

The first wall and divertor components of fusion devices are subject to extreme particle and thermal loads during operation. In terms of durability and plasma compatibility, tungsten is the prime candidate as plasma-facing material. Its intrinsic and operation-induced brittleness, as well as other material properties, however, pose challenges for component manufacturing, lifetime prediction, as well as safety issues. In recent years, tungsten composite and functionally graded materials, new tungsten alloys, as well as new hydrogen isotope permeation barriers haven been developed and produced by various production routes. These new materials allow high performance plasma-facing components for the first wall and divertor components which are compatible with well-established power exhaust concepts for both tokamak and stellarator fusion devices. In this presentation the requirements for new materials, as well as the currents state of the materials and their production routes, will be presented.

Session FK-4 - Functional Materials

FK-4:L01  Uranium Intermetallic Compounds with Superb Hydrogen Absorbability
MICHIO YAMAWAKI, T. YAMAMOTO, Y.ARITA, T.ONITSUKA, B.TSUCHIYA, University of Fukui, Tsuruga, Fukui, Japan

Certain uranium intermetallic compounds such as UNiAl, UNiZn and UPtAl absorb hydrogen up to 2 H atoms per U atom without destruction of the structural integrity of the compounds. Those uranium intermetallic compounds may be high performance hydrogen storage material that can be utilized to store excess electricity generated from sustainable energies. The hydrogen absorption processes of those intermetallic compounds were examined in terms of crystallography, thermodynamics and quantum mechanics. With hydrogen being absorbed, the structure of the intermetallic compounds was observed to change from A-type where some portion of Ni atoms are located on the plane formed by U atoms to B type where none of Ni atoms are located on the plane formed by U atoms. The ab-initio calculation was performed on the structural transformation to prove the observed structural change. In order to search uranium intermetallic compounds with higher hydrogen absorbability, further calculations and experiments have been carried out.

Session FK-5 - Nuclear Fuel Materials

FK-5:IL01  Development of MA-Zr Hydride for Transmutation of Nuclear Wastes by Fast Reactor
KENJI KONASHI1, M. HIRAI2, H. MUTA3, K. KUROSAKI3, K. ITOH4, K, IKEDA5, M. YAMAWAKI6, 1Inst. for Materials Research, Tohoku University, Oarai, Ibaraki-ken, Japan; 2Nippon Nuclear Fuel Development Co. Ltd., Oarai, Ibaraki-ken, Japan; 3Div. of Sustainable Energy and Environmental Engineering, Osaka Univ., Suita, Osaka-fu, Japan; 4Nuclear Development Corp., Tokai-mura, Ibaraki-ken, Japan; 5Mitsubishi FBR Systems, Inc., Shibuya, Tokyo, Japan; 6Research Inst. of Nuclear Engineering, Univ. of Fukui, Tsuruga, Fukui, Japan

Transmutation of long-lived radioactive nuclides of minor actinide (MA), i.e. 237Np, 241Am, 243Am, 244Cm and so on, has been studied. The target material including MA-Zr hydride was considered for neutron irradiation in the blanket region of conventional fast reactor. Fast neutrons generated in the core region are moderated in the MA-Zr hydride target assembly and then produce high flux of thermal or epithermal neutrons, which have large nuclear reaction cross section to nuclides. Preliminary evaluation of MA transmutation has been done by using the MA-Zr hydride targets, which are placed in the radial blanket of a 280 MWe (714 MWth) sodium-cooled mixed (U, Pu) oxide fuelled fast reactor. Calculation results show excellent performance for the transmutation of wastes. One of the most important R&D items to realize the innovative technology is the development of MA-Zr hydride target. Demonstration of fabrication of MA-Zr hydride has been done by using surrogate materials (Nd and Sm) for Am. Another R&D item is research of hydrogen gas release from MA-Zr hydride at high temperature. Special attention was paid to oxygen environment on the surface of MA-Zr hydride. It is revealed that oxygen gas on the surface suppresses the hydrogen gas release.

FK-5:L02  Study of Dissolution Mechanisms for Mixed Actinides Oxides

Several Th1-xUxO2 pellets with various composition, homogeneity and densification rate were prepared by sintering from oxalate precursors (Th being used as a redox-free surrogate of actinides) then finally submitted to dissolution tests. The macroscopic and multiparametric study of Th1-xUxO2 dissolution in nitric media showed the strong impact of composition on their chemical durability [1]. For xU < 0.5, dissolution was mainly driven by surface-controlling reactions at the solid/solution interface involving the adsorption of protons on reactive sites. On the contrary, for xU > 0.5, oxidation of uranium (IV) into uranyl became preponderant at the solid/solution interface, leading to the significant decrease of the chemical durability of the ceramics. Such modifications were validated at the microscopic scale through operando monitoring of evolving interface by ESEM. Preferential dissolution zones (triple junctions, grain boundaries, …) observed for xU < 0.5 induced the rapid increase of the reactive surface area even for short dissolution progress. For xU > 0.5, dissolution was more generalized due to the preponderant oxidation of uranium (IV) at the interface which led to moderate increase of the reactive surface area.
[1] L. Claparede et al.,J. Nucl. Mater. 457, 304-316, 2015

FK-5:L03  Impact of PGM Particles during the Dissolution of Uranium Dioxide

Dissolution of spent nuclear fuels (SNF) is a key step during their reprocessing. Moreover, SNF contains a wide variety of fission products including Platinum Group Elements (PGM) either incorporated in the UO2 matrix or present in various separated phases for which the specific impact on the overall dissolution kinetics has not been yet fully discriminated. To answer this question, UO2 samples doped with 3 mol.% of PGM (55% Ru; 9.6% Rh; 35.4% Pd) were prepared by hydroxide precipitation. After conversion then sintering, the pellets were submitted to multiparametric dissolution tests in various media (0.1 to 4 M HNO3) and temperatures (25°C to 60°C). The macroscopic description of the dissolution showed that the normalized dissolution rates were significantly increased compared to pure UO2, especially in the less acid media. Simultaneously, the microscopic operando monitoring of evolving solid/liquid interface by ESEM during dissolution suggested the existence of catalytic effects occurring at the UO2/PGM/solution interface. The combination of both 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.

FK-5:IL04  Response of Commercial MAX-phases to Neutron Irradiation to Intermediate Fluences
YUTAI KATOH1,2, CAEN ANG2, P. EDMONDSON1, TAKAAKI KOYANAGI1, 1Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA; 2University of Tennessee, Knoxville, TN, USA

MAX-phase ternary compounds have been studied as the potential materials for nuclear energy applications as they present interesting properties including the pseudo-ductile fracture behavior in certain temperature range, moderate thermal conductivity, and oxidation resistance at elevated temperatures. However, understandings of the effects of neutron irradiation on the properties and integrity of the MAX-phases are still inadequate. In particular, consequences of the crystallographically anisotropic dimensional evolutions and the ballistic mixing effects on the atomistically layered structures are among the primary concerns. In the present work, results from the neutron irradiation effects study performed at Oak Ridge National Laboratory on commercial Ti-Si-C and Ti-Al-C MAX phase systems are discussed with the main focus on the two critical radiation stability concerns.
This research was supported by the U.S. Department of Energy, Office of Fusion Energy Sciences. This abstract has been authored by UT-Battelle, LLC, under Contract No. DE-AC05-00OR22725 with the U.S. Department of Energy.

FK-5:L06  High Temperature X-Ray Diffraction Studies of Surrogates for Americium Oxides
E.J. WATKINSON, R.M. AMBROSI, J. NAJORKA, Department of Physics and Astronomy, University of Leicester, Leicester, UK; Natural History Museum, London, UK 

Experimental research into americium oxides are limited but highly relevant to the development of minor actinide targets as well as to the development of European Space Agency americium fuelled radioisotope power systems (RPS). Large changes in temperature could impact the integrity of a ceramic oxide pellet e.g. by phase changes in crystal structure. This is not only relevant to development of target/RPS pellets with handling strength but also to their integrity during non-nominal operating conditions. High temperature X-ray diffraction studies of such materials are thus pertinent. Investigations with inactive surrogate oxides can provide an initial insight into the behaviour of americium oxides under such conditions whilst minimising radiation exposure to personnel. In this talk we present a high temperature X-ray diffraction investigation of different americium oxide surrogates under a range of atmospheres e.g. Ce1-xNdxO2-(x/2) as a surrogate for certain AmO2-(x/2) and discuss future avenues of research.

FK-5:L07  Conversion of Surrogate and Uranium Oxide by Solution Combustion Synthesis
J. MONNIER1, X. DESCHANELS1, C. REY1, E. WELCOMME2, 1CEA Marcoule - Institut de Chimie Séparative de Marcoule (ICSM) - LNER, Bagnols sur Cèze, France; 2CEA Marcoule - ATALANTE - DMRC, France

Within the framework of the researches carried out on the treatment and the recycling of actinides for the 4th generation nuclear plants, various conversion routes are investigated. Solution Combustion Synthesis is an interesting process to synthesize oxides because of its energy efficiency, its low cost and the special properties it confers to materials. The conversion of actinides or surrogates nitrates into oxides was studied according to various parameters (fuel-to-nitrate ratio, fuel chemical properties…). In a first step, the precursor obtained after the sol-gel reaction between the nitrate and the fuel below the ignition temperature was studied by NMR, IR spectroscopy and also by swelling measurements. The thermal transformation of precursor into oxide was followed by TG-DTA and environmental SEM. The final oxide was characterized by BET, XRD and SEM. The structural characteristics of the final oxide powder is clearly influenced by the fuel composition and therefore on its complexing power. Another parameter, the fuel-to-nitrate ratio, seems to make it possible to control, to a certain extent, the structural characteristics of the final oxide, in particular in the uranium nitrate-glycine.

FK-5:L09  Dissolution of Uranium Thorium Mixed Oxides: The Role of Nitrous Acid
T. DALGER, S. SZENKNECT, L. CLAPAREDE, N. DACHEUX, Institut de Chimie Séparative de Marcoule, ICSM UMR 5257, CNRS, CEA, Univ. Montpellier, ENSCM, Bagnols sur Cèze cedex, France; P. MOISY, CEA, Nuclear Energy Division, Research Department of Mining and Fuel Recycling Processes, Bagnols-sur-Cèze, France  

Mixed actinide dioxides are currently used as fuels in Pressurized Water Reactors (PWR) (including Gen III, EPR) and stand as potential candidates for several Gen IV concepts including Sodium-cooled Fast Reactor (SFR) or Gas-cooled Fast Reactor (GFR). In this field, the reprocessing of minor actinides coming from spent nuclear fuel as mixed-oxide fuels or in UO2-based blankets surrounding the core is often considered. For both options, better understanding of mechanism controlling dissolution in the field of reprocessing step is required . Although oxidative dissolution in nitric acid is widely used, the mechanism occurring remains poorly known. The role of several nitrogen based species such as NOx and HNO2 has been investigated and it is yet to understand its exact part in the oxidative dissolution of uranium-based mixed oxides. Dissolution tests were performed on sintered pellets of U0.75Th0.25O2 in nitric acid solutions (from 10-1 M to 4 M) under dynamic conditions. Therefore, a multiparametric study of the dissolution kinetics was achieved in order to determine the influence of the species of interest. The obtained results gave evidence of the strong dependency of the dissolution rate with the nitrous acid activity.

FK-5:L10  Zirconium Carbide (ZrC) for High Temperature Nuclear Environments - Probing the Local Structure using NMR
DHAN-SHAM RANA, I. FARNAN, Department of Earth Sciences, University of Cambridge, Cambridge, UK

Previous studies have characterised the physical, thermal properties and irradiation response of ZrC with C/Zr demonstrating a non-linear variation. Computational studies have concluded vacancy ordering contributes to this behaviour – and vacancy avoidance contributes to the elimination of vacancy clustering inducing instabilities in the structure (low C/Zr stabilised by 3rd nearest neighbours). This study aims provide evidence for preferential carbon site ordering with sintering temperature. ZrC1-x pellets were reactively hot-pressed from appropriate mixtures (C/Zr= 0.60, 0.65, 0.70, 0.80, 0.90, 0.95, 1.00) of ZrH2 and graphite powders (3 batches of each composition were sintered at 1773K, 1973K and 2273K). Non-dispersive infrared carbon analysis (NDICA) and X-ray diffraction (XRD) were used to determine the C/Zr and lattice parameters. Solid-state carbon-13 Nuclear Magnetic Resonance (NMR) was used to observe the carbon local environments. The graphite peak in the NMR spectra highlights a discrepancy between the bulk NDICA and the NMR carbon contents. Graphite was not observed by XRD, hence this peak is sp2 carbon which is unreacted or ex-solved from the structure. Systematic evolution of ZrC1-x sub-peaks demonstrates ordering - no simple model is obeyed for vacancy ordering.

Session FK-6 - Radiation Effects

FK-6:IL01  A Real Space Multiscale Model for the Deformation and Swelling of Components under High-energy Neutron Irradiation
S.L. DUDAREV1, 2, D.R. MASON1, E. TARLETON2, P.-W. MA1,3, A.E. SAND4, 1UK Atomic Energy Authority, Oxfordshire, UK; 2Department of Materials, University of Oxford, Oxford, UK; 3Department of Engineering Science, University of Oxford, Oxford, UK; 4Department of Physics, University of Helsinki, Finland

Neutrons produced by nuclear reactions in the D-T plasma of a fusion device interact with materials surrounding the plasma, and this gives rise to changes in physical and mechanical properties of materials. Magnetic properties of iron and iron alloys also change under irradiation. These changes result from processes occurring at the atomic scale. Collision cascades produce fairly stable and relatively well localized distortions of atomic structure; these are radiation defects. In this study, we focus on the development of a real-space multiscale model for stresses, deformation and swelling of materials and structural components exposed to intense high-energy neutron irradiation. This is a fundamental problem, the solution of which has so far eluded engineers exploring designs of advanced power plants. We show that this problem can be addressed using an approach, free from adjustable parameters, which exploits the fact that the characteristic spatial scale of defect structures is small in comparison with the scale of an engineering component in a reactor. The model uses ab initio and atomistic data to compute strains and stresses associated with defect structures, enabling prediction of deformation of materials and components exposed to high-energy neutron irradiation.

FK-6:IL03  Pancake-like Growth and Coalescence of Intergranular Helium Bubbles: In situ Observation and Analytical Modelling
HEFEI HUANG1, JIE GAO1,2, XIANG LIU3, YAN LI1, 1Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai, China; 2School of Physical Sciences, University of Chinese Academy of Sciences, Beijing, China; 3Department of Nuclear, Plasma, and Radiological Engineering, University of Illinois at Urbana-Champaign, Urbana, IL, USA 

Polycrystalline nickel foil with thickness of ~ 950 nm was irradiated with 0.5 – 1.2 MeV helium ions at room temperature. Using transmission electron microscope (TEM) and positron annihilation spectroscopy (PAS), the irradiation induced helium bubbles in sample can be identified. Moreover, the in situ TEM characterization showed the pancake-like growth of a helium bubble along grain boundaries to achieve coalescence with another bubble at 673 K even in the absence of external stresses. The developed analytical model, addressing the stress field induced by the bubble internal pressure, well reproduced the observed shape evolution during the bubble growth. Moreover, we show that the coalescing rate of intergranular bubbles was controlled by the surface diffusion and can be accelerated by the pressure gradient between bubbles. In this study, a new mechanism governing the coarsening of intergranular sub-10 nm helium bubbles. Our findings shed new light on understanding the coarsening mechanism of intergranular helium bubbles and could open new avenues for future experimental estimations of the extreme low diffusivity of polycrystalline materials.

FK-6:IL05  New Conceptual Advances in Diffusion-mediated Modelling of Dislocation-driven Evolution of Radiation Effects in Fission and Fusion Materials
I. ROVELLI1,2, A.P. SUTTON1, S.L. DUDAREV2, 1Department of Physics, Imperial College London, London, UK; 2Culham Centre for Fusion Energy, UK Atomic Energy Authority, UK

In a nuclear reactor, high-energy neutrons induce the formation of extended defect clusters in structural components, degrading their properties over time. Such damage can be partially recovered via thermal annealing treatments. Therefore, in the context of design and operation of fusion and advanced fission nuclear energy systems, it is critical to predict the annealing timescales of arbitrary configurations of defect clusters. With this goal in mind, we extended the Green’s function formulation of Gu and Srolovitz for the climb of curved dislocations in order to include in the same framework the evaporation and growth of cavities and the effects of free surfaces. We present the mathematical foundations of the model, which makes use of boundary integral equations to solve the steady-state vacancy diffusion problem. Numerical results in the simplified case of a dilute configuration of prismatic dislocation loops and spherical cavities in a finite-size medium show a good agreement with experimental data on defect annealing in ion-irradiated tungsten. In addition, we propose a coupled mean-field description to capture the effect of unobservable small clusters, in order to deal with incomplete data of defect size distributions below the experimental detection limit of TEM.

FK-6:IL06  In situ Ion Irradiation Induced Detwinning in Naotwinned Cu Films
ENGANG FU1, J.L. DU1, K.Y. YU2, M.M. LI3, M. KIRK31School of Physics, Peking University, Beijing, China; 2Dept. of Materials Science, Chinese University of Petroleum, Beijing, China; 3Argonne National Laboratory, Argonne, IL, USA

The mechanism of radiation induced detwinning is different from that of deformation detwinning as the former is dominated by supersaturated radiation induced defects while the latter is usually triggered by global stress. In-situ Kr ion irradiation was performed to study the detwinning mechanism of nanotwinned Cu films with various twin thicknesses. Two types of incoherent twin boundaries (ITBs), so-called fixed ITBs and free ITBs, are characterized based on their structural features, and the difference in their migration behavior is investigated. It is observed that detwinning during radiation is attributed to the frequent migration of free ITBs, while the migration of fixed ITBs is absent. Statistics shows that the migration distance of free ITBs are thickness- and dose-dependent. Potential migration mechanisms are discussed.

FK-6:L08  Behaviour of Spent Nuclear Fuel during Long-term Storage: Accelerated Radiation Damage with 238Pu-doped UO2
E. DE BONA, M. COLOGNA, T. WISS, R.J.M. KONINGS, JRC-Karlsruhe, Eggenstein-Leopoldshafen, Karlsruhe, Germany; G. BALDINOZZI, CentraleSupélec, Gif-sur-Yvette, Paris, France

The radiological activity of spent nuclear fuel after few centuries will be dominated by α-decay. Accumulation of α-damage and radiogenic He might be detrimental to spent fuel integrity. In order to reproduce the ageing at a laboratory timescale, surrogate samples of UO2 containing 2.5 and 10 wt % of α-emitting 238Pu have been synthesized. A homogeneous distribution of the a-damage in the bulk of the material has been achieved by powder coprecipitation and sintering of dense pellets. The kinetics and nature of damage ingrowth has been periodically investigated on samples stored at liquid nitrogen, at room temperature, and at fuel wet storage (200°C) temperature. The combination of the two compositions and the selected periodicity of the measurements cover an equivalent timeframe that spans 10.000 years. The damage evolution is followed by means of XRD (structure), TEM (microstructure), DSC (defect energy), Helium thermal desorption spectrometry, mechanical testing, and Raman spectroscopy. The temperatures chosen for the samples storage should allow differentiating the nature of the defects formed by the a-decay and their temperature-dependent evolution, in particular during spent fuel storage. The outcome of this work should help predicting the long term behaviour of spent fuel.

FK-6:L09  Are Mesoporous Silica Radiation Tolerant?

Because of their high S/V ratio, mesoporous materials could be tolerant to radiation damage defects. To investigate this topic, sol-gel silica thin films (e~100 nm) deposited on Si substrates were irradiated with gold ions (0.514 cm-2 (0.5dpa). Mesoporosity collapse seems different according to the irradiation regime (nuclear versus electronic). The sol-gel samples exhibit a delayed radiation damage compared to material elaborate by classical route, which indicates a higher radiation tolerance. The presentation aims to discuss these different observations, and clarified the role of the interfacial surface on the healing of the defects created by irradiation. From another point of view, the sensitivity of these mesoporous structures to the radiation damage opens interesting prospects for obtaining self-conditioning materials.

FK-6:IL10  Effect of Irradiation Defects on SiC Dissolution in Hot Water

The applicability of SiC composites to the core components in LWRs is a lively discussion topic. Recently, the acceleration of the hydrothermal corrosion of SiC due to self-ion irradiation was reported by authors. The underlying mechanisms, which would be largely different from that for metallic alloys, are still not clear, unfortunately. Particularly in the case of SiC semiconductor, the modification of the electrochemical dissolution is difficult in the absence of assistance from the polarization or illumination. However, the irradiation-induced defects and their localized states could affect the carrier transportation at the water-SiC interface. This work try to associate the irradiation induced defects with the electrochemical properties of the ion-irradiated single-crystalline 3C-SiC in order to understand the underlying mechanisms and to find the corrosion prevention system. Our results showed the strong correlation between the cyclic voltammograms in aqueous solutions saturated with ambient oxygen and the concentrations of defect levels measured by DLTS. No significant correlation was observed in the voltammograms obtained using deoxygenated solutions, implying the anodic SiC dissolution was accelerated by the introduction of defect levels within a bandgap.

FK-6:L12  Helium Precipitation Study in UO2 by Transmission Electron Microscopy
A. MICHEL, G. CARLOT, C. SABATHIER, CEA / DEN / DEC, Saint Paul Lez Durance, France; M. DUMONT, IM2NP, UMR CNRS 7334, Aix-Marseille Université, Marseille, France; M. CABIE, CP2M, Aix-Marseille Université, Marseille, France  

The direct disposal of spent nuclear fuel 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 will be released from the spent fuel at the breaching of the disposal canister1. This source term depends on the microstructure of the aged spent fuel and on the localization of the radionuclides. 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. The study of helium precipitation 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 show bubbles with a maximum size of 1 nm and a density of about 1022 b.m-3. These characterizations do not allow us to link implantation conditions and features of helium bubbles. However, platelets which could lead to spalling3 were observed in case of 3.1016 He/cm2-600°C implantation conditions.
1 C. Ferry et al., J. Nucl. Mater., 2010 2 Z. Talip et al., J. Nucl. Mater., 2014 3 T. Pingault, Orléans, 2016

FK-6:L13  Degradation of Zr Microstructure under Operation as Part of Fuel Assemblies of VVER-type Reactors

At present, the main material for fuel elements cladding (fuel rods) manufacturing of industrial reactors is the E110 zirconium-based alloy doped with 1% niobium. This is due to the relatively small fast neutrons capture cross section and acceptable radiation resistance of this material. During nuclear reactor operation, fuel rods are exposed to such factors as neutron irradiation, elevated temperatures, mechanical stresses due to nuclear fuel swelling and fission products formation, etc., which leads to the structural-phase state degradation, and, ultimately, to mechanical properties degradation. For the safe operation of fuel assemblies, it is necessary to ensure a given resource of the fuel assembly elements, which requires understanding the mechanisms of structure and properties degradation of the cladding materials. Therefore, the complex studies of fuel rod samples after operation within the VVER-1000 core with the different nuclear fuel burnups were carried out in this work, including: • Microstructural studies • Mechanical tests • Fractographic studies A correlation between the structure evolution of E110 alloy and mechanical properties degradation of the studied fuel rod samples is shown.

FK-6:L14  Electronic Structure Calculations of Structural, Electronic, Thermodynamic and Defect Properties in Mixed Uranium-plutonium Oxides (U,Pu)O2
I.C. NJIFON1, M. FREYSS1, R. HAYN2, M. BERTOLUS1, 1CEA, DEN, DEC, de Cadarache, Saint-Paul-Lez-Durance, France; 2Aix-Marseille Université, IM2NP, Campus Scientifique Saint-Jerôme, Marseille Cedex, France

(U,Pu)O2 (MOX) is currently used as nuclear fuel in pressurized water reactors and is envisaged as reference fuel in Generation IV reactors. Under irradiation, nuclear fuel is submitted to multiple phenomena caused by the creation of fission products in the materials, which lead to the formation of a large number of point defects. The aggregation of these defects leads to the creation of dislocations and cavities that modify the fuel microstructure and influence fission product diffusion. A better description of the fuel behaviour at the atomic scale, and especially of the elementary mechanisms involved in the diffusion of point defects and fission products, is necessary to refine the models used in fuel performance codes. While point defect properties have been widely investigated in UO2¹ ², and to a certain extent in PuO2³, the study of (U,Pu)O2 has only started recently and point defects in this material have not been investigated. We will present a study of the structural, electronic and thermodynamic properties of MOX as a function of the Pu content using electronic structure calculations. Results on the defect stability and transport properties will also be shown.
¹ Dorado et al, J. Phys. Condens. Matter 2013. ² Liu et al, J. Mater. Sci. 2012. ³ Petit et al, Science 2003.

Session FK-7 - Materials Modelling and Database

FK-7:IL02  A Large Scale Database of Cascade Configurations: A New Paradigm in Multi-scale Modelling of Radiation Damage Effects in Nuclear Materials
A.E. SAND, University of Helsinki, Helsinki, Finland; S.L. DUDAREV, CCFE, Culham Science Centre, Abingdon, UK

The multi-scale nature of radiation damage formation poses acute challenges for integrated modelling schemes. Collision cascades from energetic particle impacts take hundreds of CPU hours to simulate with molecular dynamics (MD), hence direct integration into larger scale methods to simulate damage build-up is not feasible. The statistics of sizes and spatial distributions of the defect clusters offer general parameters describing the damage, providing a starting point for damage evolution. However, much information is lost when the widely varying cascade damage is condensed into a few statistical distributions. An alternative approach is to directly convert MD configurations to input for evolution codes, yet the rare events generating large defects, which likely dominate the defect evolution, are seldom captured directly in a limited number of cascades. A currently planned open database will facilitate cascade configurations from different projects to be gathered together, providing better statistics and better access to data for researchers developing models of microstructural evolution. In addition, the database will allow systematic investigations of trends in defect production, and offer better opportunity for comparing predictions using different methods and codes.

FK-7:L03  EUROFER97 Ratcheting Behavior at 450 & 550°C and their Modelling
KUO ZHANG, JARIR AKTAA, Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Eggenstein-Leopoldshafen, Germany

EUROFER97 is considered as one of the candidates for structural materials of future fusion reactors. The components in reactor are usually exposed to cyclic loading caused by frequent startups, shutdowns and load changes, leading to ratcheting behavior. According to previous investigation on the ratcheting of mod.9Cr-1Mo ferritic martensitic (FM) steels, it is believed that EUROFER97 will have comparable ratcheting behavior under asymmetric cyclic loading, since these two types of materials have similar chemical compositions. To investigate the ratcheting behavior of EUROFER97, uniaxial stress-controlled cyclic loading tests are performed at 450 & 550°C. Ratcheting rates under various loading conditions, including various peak stresses, stress ratios and stress rates, are determined to build the database of EUROFER97 for the future application in the blankets of fusion reactor. An already developed constitutive model is further modified to adapt the ratcheting behavior of EUROFER97. The developed model describes both cyclic softening in strain-controlled LCF tests and its impact on the ratcheting behavior of mod.9Cr-1Mo FM steel fairly good, both at room temperature and 550°C. It is to be verified, whether the current model can also describe the ratcheting behavior of EUROFER97.

FK-7:IL04  Modelling the Thermophysical and -mechanical Properties of Tungsten Fibre-reinforced Copper Metal Matrix Composites by means of Mean Field Homogenisation
A. VON MÜLLER, M. LI, R. NEU, J.H. YOU, Max-Planck-Institut für Plasmaphysik, Garching, Germany

The exhaust of power and particles is regarded as major challenge for the design of a nuclear fusion demonstration power plant. In such a reactor, highly loaded plasma facing components (PFCs) have to withstand severe heat loads and considerable neutron irradiation. Existing designs for such PFCs combine monolithic W and Cu material grades. Such an approach, however, bears difficulties as W and Cu are materials with inherently different thermomechanical properties and their optimum operating temperatures do not overlap. In order to mitigate these issues, W fibre-reinforced Cu (Wf-Cu) metal matrix composite (MMC) materials are promising advanced candidates for PFC heat sink applications. The contribution will briefly summarise the current status of the topical Wf-Cu material developments. Then, based on available material property data, the macroscopic thermophysical and -mechanical Wf-Cu MMC properties are estimated by means of appropriate mean field homogenisation (MFH) methods. For these predictions, a realistic arrangement of the reinforcing W fibres due to the manufacturing of the MMC is taken into account. As a main result, an estimation of the achievable parameter space in terms of macroscopic properties as well as their implications for PFC application will be given.

FK-7:IL06  Structural Steels for DEMO and Fusion Power Plants
E. GAGANIDZE, C. DETHLOFF, B. KAISER, M. RIETH, J. AKTAA, Karlsruhe Institute of Technology, Institute for Applied Materials (IAM), Eggenstein-Leopoldshafen, Germany

Reduced Activation Ferritic/Martensitic (RAFM) steels are primary candidate structural materials for the First-Wall (FW) and Breeding Blanket (BB) components of future energy generating Fusion Reactors (FR). Although, advanced RAFM steels exhibit clearly better neutron irradiation resistance than commercial martensitic alloys, the performance of these materials under neutron irradiation will still be suffering due to the degradation of the microstructure as a consequence of evolution of displacement damage and due to generation of gaseous transmutation products e.g. helium. The mechanical properties of RAFM steel EUROFER97 will be reviewed in view of its application in the DEMO FW and BB components. The TEM investigation results on the major neutron irradiation induced microstructural defects responsible for the degradation of the mechanical properties e.g. dislocation loops, voids/bubbles and radiation driven precipitation will be summarized. The alteration of the mechanical properties as a consequence of the neutron irradiation will be discussed. Helium effects and embrittlement will be assessed by reviewing the results obtained with selected helium simulating techniques. Recommendations on the operating temperature range for the FW and BB will be given.

Session FK-8 - Crosscutting Materials Issues for Nuclear Fission and Fusion Systems

FK-8:IL01  Low Activation Structural Materials for Nuclear Fission and Fusion Reactors - the RF R&D
V.M. CHERNOV, M.V. LEONTIEVA-SMIRNOVA, A.A. BOCHVAR, High-technology Research Institute of Inorganic Materials, Moscow, Russia

Creation of the large-scale nuclear fission and fusion power engineering under conditions of the requirements of the expansion of safety and efficiency of reactors and the realization of the full closed nuclear fuel cycle and minimization of the radioactivity wastes sets new requirements to structural materials (SMs). New SMs should not be inferior to the reference SMs (high activated) in terms of their functional properties and should provide significant additional opportunities to improve reactor efficiency and safety. The requirements for new SMs are close to the technological limits of their manufacture and can be implemented in a new class of low activated (reduced activation) SMs (LASMs). To a large extent, the problems of selection, fabrication and further modifications of the LASMs have been resolved. The scientific and technological problems of the RF R&D of the current and being developed the RF LASMs are considered: ferritic-martensitic 12% chromium steel RUSFER-EK181 (Fe-12Cr-2W-V-Ta-B), vanadium alloys V-4Ti-4Cr and V-Cr-W-Zr-C-O. The technological levels of readiness of the LASMs, the directions of further R&D works, reactor tests of materials and goods and industrial production are determined.

FK-8:IL02  Challenges of Simulating Neutron-induced Radiation Damage Using Ion Beams
G.S. WAS, University of Michigan, Ann Arbor, MI, USA

Reactor core materials in both fast reactors and LWRs granted life extension must withstand irradiation to high dose at high temperature. To reach high doses, self-ion irradiation is conducted with simultaneous He injection to emulate the reactor irradiation environment. The goal is to create a radiation damage microstructure that mimics that created in reactor. If this can be achieved, then the resulting mechanical property response can be estimated. The advantages of ion irradiation over reactor irradiation are many, including greatly reduced time and cost, thus enabling a huge acceleration in our understanding of radiation damage evolution in specific alloy systems and in developing new materials for reactor core components. Numerous challenges must be met for this equivalence to be established. Among these are the much higher damage rates of ion irradiation, the behavior of implanted gases at such high rates, the existence of a damage gradient through the irradiated region, the proximity of the surface and the non-irradiated bulk, and the thinness of the irradiated layer. Nevertheless, impressive progress has been made in establishing ion irradiation as a valuable technique for understanding microstructure evolution in reactor core materials.

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