Symposium FJ
Materials Challenges for Future Nuclear Fission and Fusion Technologies

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

FJ-1:IL02  Advanced Steels for Fission and Fusion Reactors
A. MOESLANG, KIT, IAM, Karlsruhe, Germany

As the performance goals for fusion and advanced fission systems include higher application temperatures and longer burn-ups, the material requirements become very demanding. With respect to materials development, synergies between fusion and fission becomes obvious, as the majority of fusion and future fission reactor design concepts rely on similar coolants, temperature windows and neutron wall loading parameters. An overview is given of the major materials of construction used (i) for the primary and secondary circuits of the predominant fast breeder and light water power reactors, as well as (ii) for the in-vessel structures (blanket and divertor) of present international fusion DEMOnstration reactor designs. Secondly, the major materials degradation modes in highly neutron loaded energy systems will be summarized. The major part is dedicated to the most challenging operating environments and R&D needs for innovative Steels which are also in next step fission and fusion power plants the most important structural material classes, thereby emphasizing specifically recent achievements in austenitic, ferritic and ferritic martensitic steel R&D, and highlighting the enormous potential of their nanoparticle strengthened variants for challenging radiation tolerant applications.


FJ-1:IL03  Adding Toughness to Nano-structured Ferritic Alloys
THAK SANG BYUN, D.T. HOELZER, Oak Ridge National Laboratory, Oak Ridge, TN, USA; JI HYUN YOON, Korea Atomic Energy Research Institute, Daejeon, S. Korea; JEOUNG HAN KIM, Korea Institute of Materials Science, Changwon, S. Korea; G.R. ODETTE, University of California - Santa Barbara, Santa Barbara, CA, USA; S.A. MALOY, Los Alamos National laboratory, Los Alamos, NM, USA

Nanostructured ferritic alloys (NFAs) have been developed for future applications to fission and fusion reactor core components primarily because of their excellent high temperature creep and irradiation resistance. Such high strength materials, however, often demonstrate low fracture toughness and ductility, which may cause difficulties in manufacturing, as well as in maintaining proper integrity of reactor components. Significant recent efforts have been devoted to developing NFAs with improved fracture toughness. First, it was proposed that strengthening grain boundaries and thus improving fracture toughness could be achieved through carefully designed thermomechanical treatments (TMTs). A new 9Cr NFA (9YWTV) has been produced to study the influence of various treatments on fracture toughness. Second recent research, guided by past experience, also showed that significantly improved fracture toughness of 14YWT can be achieved by integrating of best practice processing, including cleaner mechanical milling to reduce contamination and optimum consolidation and TMT to strengthen grain boundaries. Third, cryogenic milling has been explored to pursue a property improvement through grain refinement.


FJ-1:L05  Study on the Fabrication and Joining Technologies of Ferritic/Martensitic ODS Steel Structural Components for Future Fast Reactor Applications
SUK HOON KANG, SANGHOON NOH, YOUNG-BUM CHUN, JINSUNG JANG, TAE KYU KIM, Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon, Republic of Korea

Oxide dispersion strengthened (ODS) steels are being considered as a prospective candidate material of in-core structural components such as cladding tubes, wire and ducts in the future sodium-cooled fast reactors. Especially, ferritic/martensitic ODS steels have an excellent irradiation resistance to a void swelling as well as a superior creep strength at elevated temperatures. Applications of these FM ODS steels to the fast reactor act to grow faster in nuclear engineering society; however, not so many studies have been made for improving fabrication and joining technologies of ODS steel structural components. In ODS steels, it is well known that uniform nano-oxide dispersoids act as pinning points to obstruct dislocation and grain boundary sliding. However, these advantages will be disappeared due to the segregation of oxide particles if the ODS materials are exposure to extremely high temperatures over than their melting temperatures during the fabrication and joining procedures. In this study, fabrication and joining processes of ODS steels are optimized, homogeneous grain structures and the uniform oxides dispersion are preserved in ODS steels. The response of material for different process variables have been discussed in terms of plastic deformation amount and heat input.


FJ-1:L06  Crystallographic Relationship of Y2Ti2O7 Particles with Matrix in Austenitic ODS Steels
JINSUNG JANG, XIAODONG MAO, CHANG HEE HAN, TAE KYU KIM, Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon, Korea 

Crystallographic relationship of oxide particles with the matrix of ODS steels may give insights into the formation and evolution mechanism of oxide particles, and consequently the resistance to irradiation swelling and strengthening. SUS 316L-based austenitic ODS steels were prepared by mechanical alloying, hot isostatic pressing and hot rolling processes. The coherency of nano-sized Y2Ti2O7 particles with the austenitic matrix was investigated using diffraction contrast techniques and HRTEM. It was revealed that over 90% of the Y2Ti2O7 particles (3-10 nm in diameter) were semi-coherent with the austenitic matrix, and perfectly matched planes that produce no-contrast lines under a two-beam condition in TEM were shown across the oxide/matrix interfaces. Single no-contrast line or multiple no-contrast lines under different active g-vectors were observed and discussed. The lattice distortion in the matrix around the Y2Ti2O7 particles was also estimated by using the diffraction contrast method under TEM.


FJ-1:L07  Behavior at High Temperature of Metallic Liners (Ta, Nb) Used in the Sandwich Cladding Material of the GFR
L. CHARPENTIER, M. BALAT-PICHELIN, PROMES-CNRS, Odeillo, France

The Gas-cooled Fast-Reactor (GFR) is one of the systems developed in the frame of the 4th generation nuclear plants. The helium coolant may contain some residual oxidizing impurities. A sandwich material with a metallic layer (Ta or Nb) inserted between two sheets of SiC/SiC composite is a promising cladding of the nuclear fuel to support mechanical strengths and to retain fission products without blocking the neutrons. Nevertheless SiC may undergo active oxidation with production of gaseous SiO and CO and sublimation in accidental conditions for temperatures above 2000 K, so the aim of this study is to firstly investigate the oxidation resistance of the metallic liner in extreme conditions, alone or covered by the SiC/SiC composite.
The High Pressure and Temperature Solar Reactor (REHPTS) was implemented at the focus of the 5kW Odeillo solar furnace in order to reproduce in helium atmosphere the sudden and huge temperature increase that can occur in the case of a nuclear accident. XRD, SEM, and 3D optical roughness measurements enabled to observe that the high temperature exposure favored the (211) preferential orientation of tantalum and that the amorphization of niobium occurs above 1400 K. Such changes may impact the properties of the cladding elements.


FJ-1:L08  Investigation of Mechanical and Corrosion Properties of Ni-based Alloy for VHTR at 950 °C with Alloying Element and Heat Treatment
DONG-JIN KIM, SU JIN JUNG, BYUNG HAK MOON, SUNG WOO KIM, YUN SOO LIM, HONG PYO KIM, Nuclear Materials Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon, Korea

A very high temperature reactor (VHTR) has lots of advantages, which are to generate highly efficient electricity at near 50% and produce massive amounts of hydrogen while very high temperature is a big challenge to material. In particular, a structural material for an intermediate heat exchanger (IHX) among numerous components is exposed to high temperature of up to 950°C. In this harsh environment, even nickel-based superalloy is degraded unavoidably in spite of its excellent creep resistance at high temperature leading to a great concern with long-term integrity. In this work, the mechanical property and microstructure for nickel-based alloys fabricated at the laboratory were evaluated as functions of alloying elements and heat treatment condition. High-temperature corrosion behavior was also discussed. The Ni-based alloys were melted by VIM (vacuum induction melting), followed by homogenization at 1200°C for 20 hrs and hot rolling in a temperature range of 1050 to 1150°C. A solution annealing, followed by additional heat treatment in the range of 1020 to 1140°C, was conducted at 1175°C. Additional heat treatment was performed to strengthen the grain boundaries through carbide development along the grain boundary. Cooling after all heat treatments was conducted by water quenching. A tensile test was carried out in air at 950°C with a straining rate of 9.3x10^-4 s^-1. The ductility increased from 15% to 48% with the heat treatment temperature and then drastically decreased at 1140°C almost to the ductility of the solution annealed specimen. From the experimental results, it was revealed that the grain boundary is effectively strengthened up to 1110°C. Mo was beneficial to high-temperature ductility while Cr was detrimental to high-temperature ductility. It was found that Co modified the carbide composition providing the synergic effect of Mo and Co on the mechanical property at 950°C. Aluminum improved the high-temperature corrosion resistance.

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

FJ-2:IL01  Indian Test Blanket Module in ITER - Development of RAFM Steel and Fabrication Technology
T. JAYAKUMAR, Metallurgy & Materials Group Indira Gandhi Centre for Atomic Research, Kalpakkam, India; E. RAJENDRA KUMAR, TBM Division, Institute of Plasma Research, Gandhi Nagar, India

A detailed and comprehensive project is in progress in India to fabricate Indian Test Blanket Module (TBM) to be tested in ITER. The project comprises of developments of India-specific RAFM steel and fabrication technologies for TBM. The material development has been realized through the melting and characterization of several heats of 9Cr-RAFM steel of varying tungsten and tantalum contents for optimum strength and toughness. Four heats of RAFM steel with W and Ta contents in the range 1-2 wt. % and 0.06-0.14 wt.% respectively, have been melted through VIM and VAR routes with strict control over the radioactive tramp elements (Mo, Nb, B, Cu, Ni, Al, Co, Ti) and on the elements that promote embrittlement (S, P, As, Sb, Sn, Zr, O). Extensive mechanical testing and metallurgical characterisation of these steels have been carried out. Ductile to brittle transition temperature of the steel increases with both tungsten and tantalum contents. Tensile strength of the steel is found not to influence significantly with the increase in tungsten content, however decreased marginally with the increase in tantalum content with the consequent increase in ductility. Creep rupture strength of the steel increases significantly with increase in tungsten content whereas it decreases with the increase in tantalum content. The increase in tungsten and tantalum contents increase the low cycle fatigue strength of the steel, however extensive cyclic softening is observed for the steel having tungsten content more than 1.4 wt.%. The RAFM steel having 1.4 wt.% tungsten with 0.06 wt.% tantalum possesses better combination of strength and toughness and is considered as India-specific RAFM steel.
The joining technologies adopted for the fabrication of TBM are Hot Isostatic Pressing (HIP) and Tungsten Inert Gas (TIG), Narrow Gap TIG (NG-TIG), Electron Beam (EB), Laser and Laser Hybrid welding processes. HIP is considered for fabrication of first wall and initial trials carried out have demonstrated that plates with internal channels, as required for the first wall of TBM can be produced by joining plates with pre machined grooves by this process. Channel collapsing during HIPing is overcome by inserting leachable ceramic core. EB and Laser Welding are being considered for fabrication of breeder cassettes. Procedures for welding for RAFM steel using these processes have been developed. Properties of the weld metal have been found to be comparable to that of the base metal. TIG, NG-TIG and Laser-Hybrid welding processes are being considered for integration of the various components like first wall, back plate, bottom plate, breeder assembly, flow dividers etc into TBM. RAFM steel welding consumables required for using along with these joining processes have been developed and qualified. Procedure for laser hybrid welding has also been developed independently.
Technologies for inspection and quality assurance of the TBM fabricated using these processes are also being developed in parallel. Use of Ultrasonic C-Scan imaging to examine the bond integrity of the HIP joint has been demonstrated. Phased Array technique that would enable inspection of weld by longitudinal movement of the probe from an optimised distance of the weld has also been developed. This procedure requires access from only one side of the weld and hence ideal for inspection of the welds in a box like structure like TBM.
The presents discusses challenges in developing the India-specific RAFM steel and the fabrication technologies to be adopted in fabrication of Indian TBM.


FJ-2:IL02  Recent Developments on RAFM Steels
E. WAKAI, M. ANDO, N. OKUBO, K. WATANABE, H. NAKAMURA, H. TANIGAWA, Japan Atomic Energy Agency, Naka-gun, Japan; D. BERNARDI, ENEA, Italy

The development of reduced-activation ferritic/martensitic (RAFM) steels for the fusion DEMO reactor has been started from around the 1980s. RAFM steels are the first candidate materials for the first wall and blanket structure of fusion DEMO reactors, the target back-plate and the target assembly of Irradiation Fusion Materials Irradiation facility (IFMIF). 　
In this study, it will be shown in details that recent some research activities for the valuation of RAFM steels such as F82H have been performed in the analysis of microstructures by TEM with EDS and SIMS, and mechanical properties such as tensile, Charpy impact, and fracture behavior. The effect of heat treatment on the microstructures and mechanical properties of RAFM steels is very important for the properties, and some results of F82H steel, F82H+60ppmB steel and F82H+60ppmB+200ppmN steel will be shown in the symposium, and the irradiation response will be also shown as some experimental results of ion irradiations such as TIARA facility of JAEA and neutron irradiations in JMTR and HFIR. In application of RAFM steel to the engineering component, the design and fabrication of F82H steel for the target assembly under IFMIF/EVEDA project has been performed, and the content will be also shown in this symposium.


FJ-2:IL03  Effects of Neutron Irradiation on MAX-phase Ceramics
Y. KATOH, CHUNGHAO SHIH, A. PEREZ-BERGQUIST, K.J. LEONARD, S.J. ZINKLE, Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, USA; T. Toyama, T. Shikama, Institute of Materials Research, Tohoku University, Japan

The use of MAX-phase ternary ceramics is proposed for various nuclear energy applications as the joining materials for SiC-based ceramics and composites, the coating materials for metallic fuel clads and core components, and the matrix materials for ceramic matrix composites. While the oxidation resistance and the ability to deform in a pseudo-ductile manner at elevated temperatures for certain MAX-phase ceramics are attractive, there are concerns for the radiation stability of these materials mainly because of their hexagonal crystallographic structures and the potential instability of the atomically layered structures that may be disturbed by energetic radiation. In the present work, a few representative MAX-phase ceramics in forms of bulk and bonding materials were evaluated following neutron irradiation in High Flux Isotope Reactor to relatively low fluence levels at elevated temperatures. The early results on physical, mechanical, and thermal properties are discussed in relation with the microstructural evolutions.


FJ-2:IL04  Study on the Creep and Fatigue Properties of CLAM Steel
YANYUN ZHAO^{1, 2}, SHAOJUN LIU², QUNYING HUANG², BOYU ZHONG², GANG XU¹, FDS Team ¹Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences, Hefei, Anhui, China; ²University of Science and Technology of China, Hefei, Anhui, China

With good irradiation swelling resistance, thermo-physical and thermo-mechanical properties, the RAFMs (Reduced Activation Ferritic/Martensitic steels) have been considered as the primary candidate structural materials for blankets of the fusion DEMO reactor and the first fusion power plant. As one of the several types of RAFMs, China Low Activation Martensitic (CLAM) steel is being developed in INEST (Institute of Nuclear Energy Safety Technology, Chinese Academy of Sciences) under wide collaboration with many other institutes and universities domestic and overseas.
Creep-rupture behavior at high temperature is one of the key issues for the application of RAFMs in a fusion reactor. And the most common type of mechanical failure is caused by fatigue. In this paper, the creep and fatigue properties of CLAM steel were investigated.
The creep tests of CLAM steel were carried out at 773K, 823K and 873K respectively with the stress ranges from 150MPa to 230MPa. The relationship between minimum creep rate (εmin) and the applied stress (σ) was described by Norton power law. The stress exponent n was 23, 19 and 15 at 773K, 823K and 873K respectively, which means that the creep behavior was controlled by dislocation climb. The Monkman-Grant equation was used to discribethe relationship


FJ-2:L05  Effect of Phase Transformation on Impact Properties In the Weld Heat-affected Zone of a Reduced Activation Ferritic/Martensitic Steel
JOONOH MOON, CHANG-HOON LEE, TAE-HO LEE, Korea Institute of Materials Science (KIMS), Changwon, South Korea

The effect of phase transformation on impact properties in the weld heat affected zone (HAZ) of a reduced activation ferritic/martensitic (RAFM) steel was investigated. The HAZs were simulated using Gleeble simulator, with different heat inputs and peak temperatures. The base steel consisted of tempered martensite, while the HAZs consisted of marteniste, δ-ferrite and small volume of autotempered martensite. The impact properties using Charpy V-notch impact test revealed that the HAZs showed poor impact properties due to the formation of martensite and δ-ferrite as compared with the base steel. In addition, the impact properties of the HAZs further deteriorated with the increase in δ-ferrite fraction by increasing peak temperature. Also, Increasing heat input more deteriorated the impact properties of the HAZs.The impact properties of the HAZs could be improved through the formation of tempered martensite after post weld heat treatment (PWHT), but they were still lower than that of base steel because the δ-ferrite remained in the tempered HAZs.


FJ-2:L06  Effect of Cooling Rate of Normalizing on Microstructures and Mechanical Properties of Low Activation Ferritic Martensitic Steels
CHANG-HOON LEE, JOONOH MOON, TAE-HO LEE, Korea Institute of Materials Science, Changwon, South Korea

The influence of cooling rate of normalizing heat treatment on microstructures and mechanical properties in reduced activation ferritic martensitic steels was investigated. The fractions of ferrite and martensite were controlled by cooling rate after normalizing heat treatment. As cooling rate was slower, the fraction of ferrite phase increased and that of martensite decreased. With increasing the fraction of ferrite, impact toughness decreased drastically in as-cooled specimens after normalizing. On the other hand, tempered specimens did not show a sharp drop in impact toughness with increasing the fraction of ferrite. In this study, the change of mechanical properties, especially impact toughness in this RAFM steel where the fractions of martensite and ferrite were controlled is discussed based on relationship between micro-mechanics of fracture and microstructural analysis.

FJ-2:L06b  Effects of Zr Addition on Mechanical Behavior of Reduced-activation Ferritic-martensitic Steel
YOUNG-BUM CHUN¹, S.H. KANG¹, S. NOH¹, T.K. KIM¹, D.W. LEE², S. CHO³, Y.H. JEONG¹, ¹Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon, Korea; ²Nuclear Fusion Engineering Development Division, Korea Atomic Energy Research Institute, Daejeon, South Korea; ³National Fusion Research Institute, Daejeon, Korea

Reduced activation ferritic-martensitic (RAFM) steel is considered a primary candidate for the structural material in a fusion reactor. The mechanical properties of RAFM steel are affected significantly by the dislocation density and the characteristic of the precipitates. Such microstructural factors are determined by heat treatments such as normalizing and tempering. Accordingly, optimization of the heat treatment variables is important to improve mechanical properties. The present study investigates effects of heat treatments and addition of Zr on tensile behavior such as strain-rate sensitivity, work-hardening rate, and dynamic strain aging (DSA). It was found that the addition of Zr promotes the softening kinetics of a dislocation substructure, which can be related to the reduced yield strength and enhanced impact resistance. Zr addition retards the formation of precipitates, which in turn leads to improved short-term creep resistance but induces DSA at temperatures between 250 and 400°C.

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

FJ-3:IL01  The Comprehensive First Mirror Test in the JET Tokamak for ITER
M. RUBEL¹, D. IVANOVA¹, P. PETERSSON¹, A. GARCIA-CARRASCO¹, J. LIKONEN², A. WIDDOWSON³ and JET-EFDA Contributors*, *JET-EFDA, Culham Science Centre, Abingdon, UK; ¹Alfvén Laboratory, Royal Institute of Technology, Association Euratom - VR, Stockholm, Sweden; ²Association EURATOM-TEKES, VTT, Espoo, Finland; ³CCFE/EURATOM Fusion Association, Culham Science Centre, Abingdon, UK *See the Appendix of F. Romanelli et al, Proceedings of the 24th IAEA Fusion Energy Conference 2012, San Diego, USA

Metallic mirrors, so-called first mirrors, are essential components of all optical systems for plasma diagnosis in a reactor-class device. The First Mirror Test (FMT) was carried out at the JET tokamak. The program has comprised two phases: (a) the test in JET with carbon walls (JET-C) completed in 2005-2009 and (b) exposures done 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, microscopy and a number of ion beam methods including nuclear reaction analysis, heavy ion elastic recoil detection analysis and other techniques.
This presentation shows material modification occurring on Mo mirrors exposed to plasma discharges in JET. The aim is to provide an overview of results obtained in JET-ILW and to compare to the situation in JET-C. The essential result for mirrors exposed in the presence of metal walls is that the deposition of plasma impurity species on mirror surfaces has been reduced by a factor of 10-20 in comparison to JET-C, i.e. from over 20 microns to less than 1 micron for specimens in the divertor region. The implications of these results for the first mirror maintenance in a reactor-class device will also be discussed.


FJ-3:IL02  New Materials for Nuclear Fusion Plasma-facing Components
C. LINSMEIER, Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung - Plasmaphysik, Jülich, Germany

The extreme loads of particle and power fluxes towards the first wall components in present-day and in particular future fusion devices make the selection of materials for these components a critical issue. High surface temperatures establish thermal and stress gradients in the components towards the cooling medium. The use of different materials as plasma-facing armor, bonding layer, heat sink and structural material increase the stresses and potential failure zones in these components. Hydrogen isotope diffusion and permeation, neutron irradiation, and off-normal power loads have to be considered in the design of first-wall components. In addition, the design of a fusion reactor requires a large degree of inherent safety in case of a potential loss-of-coolant event with air ingress. Tungsten or tungsten alloys are currently the most promising armor materials. However, their inherent brittleness even increases under operational heat and neutron loads. This presentation reviews new composite material concepts to solve the brittleness problem of tungsten materials, addresses solutions to avoid tritium permeation into structural materials and cooling media, and demonstrates new self-passivating tungsten alloys to avoid the release of radioactive isotopes in a LOCA scenario.


FJ-3:IL03  Modeling Irradiation Damage and Dislocation Plasticity in Tungsten for Fusion Applications
J. MARIAN, Lawrence Livermore National Laboratory, Livermore, CA, USA

In this talk, we touch on two important aspects of W performance in fusion environments. First, we present results of irradiation damage accumulation in pure W under fast neutron irradiation conditions: those found in the fast reactor JOYO and those in ITER. We calculate the evolution of different defect species with dose and discuss the effect of transmuted He on the overall accumulation of damage. Second, we describe an ongoing effort to develop a model for plasticity in W-Re alloys. The model attempts to explain the ductilization effect of Re atoms in W using stochastic simulations of thermally-activated dislocation motion. The approach is parameterized using dedicated DFT calculations. Our results map the temperature-stress space where the ductile behavior of given W-Re alloys is optimized.


FJ-3:IL04  Plasma Materials Interactions
J. LINKE, J. DU, TH. LOEWENHOFF, G. PINTSUK, T. WEBER, M. WIRTZ, Forschungszentrum Juelich, Euratom Association, Juelich, Germany

The so-called first wall and the divertor in next step thermonuclear fusion devices are exposed to intense fluxes of charged or neutral particles and to radiation in a wide spectral range. These processes, in general referred to as 'plasma wall interaction' will have strong influence on the plasma performance and on the lifetime of the plasma facing components.
For next step machines, various design options for the first-wall and for divertor based on different plasma facing materials (beryllium, fiber-reinforced carbon and tungsten) have been manufactured and successfully tested. The load limits for different geometries using different joining technologies have been determined in electron-beam high heat flux experiments. Beside thermal fatigue effects induced by cyclic heat loads also transients such as plasma disruptions and Edge Localized Modes (type I ELMs) have been evaluated for a wide range of materials, in particular for tungsten grades which have been modified with respect to microstructure, heat treatment, or alloying elements. Another important issue is the materials degradation under hydrogen or helium bombardment with strong impact on the hydrogen retention or the formation of dust particles induced by the growth of nano-scale tungsten fuzz.

 
Session FJ-4 - Functional Materials

FJ-4:IL01  Utilization Research and Development of Hydride Materials in Fast Reactors
K. KONASHI, Tohoku University, Ibaraki, Japan

Metal hydrides have high hydrogen atom density, which is equivalent to that of liquid water. Fast neutrons are efficiently moderated by hydrogen in metal hydrides. Metal hydrides have been studied for their potential application as nuclear materials in the Fast Breeder Reactor (FBR). Two types of the utilizations of metal hydride in FBRs are discussed in this paper. One is the application of Hf-hydride as neutron absorber material for the FBR core. The Hf hydride control rod was designed to replace the boron carbide (B4C) control rod used currently in the FBRs. Fast neutrons generated in the driver fuel region are moderated and efficiently captured in the Hf-hydride region of a control rod assembly. The Hf hydride control rod is superior to the B4C control rod in performance of long life use, which leads to reduce the cost of fabrication and disposal.
The other one is the utilization for transmutation target of long-lived nuclear wastes. Hydride fuel containing 237Np, 241Am and 243Am has been studied as a candidate transmutation target to reduce the radioactivity of long-lived nuclides included in reprocessed nuclear wastes. The hydride target was proposed based on the technologies developed in the above Hf hydride project to enhance the transmutation rate in the FBR.


FJ-4:IL03  In-reactor and In-situ Studies of Dynamic Radiation Effects in Materials for Burning Nuclear Fusion Systems
T. SHIKAMA, Institute for Materials Research, Tohoku University, Katahira, Sendai, Japan

Functional materials, such as electrical insulators and optical materials should play a crucial role in nuclear fusion systems. They are vulnerable to radiation damages and their precedent applications in high-flux irradiation environments are very limited in nuclear fission systems. Especially, their functional properties are susceptible to the dynamic irradiation effects, namely, causing dynamic changes of properties during the irradiation. In general, the in-situ type experiments need some intrigued experimental setups with bulk specimens, as the dynamic irradiation effects are in general correlated with transportation-properties along a long distance and will depend strongly on experimental parameters, such as temperatures, chemical environments, and irradiation parameters. A large experimental volume and good accessibility are mandatory for the concerned irradiation tests. A high flux fusion neutron source will be the best for the in-situ experiments. However, it will need some more time to realize. And, even when it is realized, its utilizations for the in-situ type experiments will be very limited. The in-situ experiments in high flux fission reactors will be the second best or complementing, with strong backups by the experiments with charged particles and gamma-rays. In the present talk, the examples of the in-situ type experiments in the high flux fission reactors, the JMTR, the HFIR, and the BR-2 will be reviewed and the present status of the knowledge-baseline of the dynamic irradiation effects will be summarized in the functional materials for the in-vessel application in the burning nuclear fusion systems. Also, some results of in-situ type experiments with accelerators, which will be also complemental to understand the dynamic irradiation of functional materials, will be reported, in conjunction with the results with fission reactors.


FJ-2:L04  Depleted Uranium as Hydrogen Storage Material
M. YAMAWAKI, Y. ARITA, T. YAMAMOTO, F. NAKAMORI, University of Fukui, Tsuruga-shi, Fukui, Japan; K. OHSAWA, Kyushu University, Japan

Large amounts of depleted uranium is kept as uranium fluoride or solid forms after enrichment of natural uranium to supply enriched uranium fuel to nuclear power stations. Various kinds of uranium alloys have been surveyed on their hydrogen absorption properties to find promising hydrogen storage material. UNiAl intermetallic compound has been found to be attractive as potential hydrogen storage material since it doesn't suffer breaking of its crystal structure on absorbing hydrogen. The kinetics of hydrogen absorption by this compound has been examined experimentally and also its mechanism has been estimated using calculational method. The change of lattice constant of UNiAlHx with the hydrogen content has been calculated, which has been found to simulate well the experimental result. Feasibility of practical use of such uranium-intermetallic compounds is to be discussed in the presentation.

 
Session FJ-5 - Nuclear Fuel Materials

FJ-5:IL01  Ceramic-based Materials for Accident Tolerant Nuclear Fuels
T.M. BESMANN, Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN, USA

A number of materials systems for accident tolerant nuclear fuels are being proposed. These range from replacement of zirconium based cladding with ceramics such as SiC composites with metal alloy barrier layers to UN kernels in TRistructual ISOtropic (TRISO) particles dispersed in a ceramic (SiC) matrix. This paper will review the state of development of these systems and results of fabrication and testing. These include compatibility studies of the hybrid SiC composite-metal barrier layer cladding, where environmental interactions are assessed and interface reactions between the structural materials are explored. For the TRISO fuel concept, wok on preparing UN kernels by carbothermic reduction and nitridation will be described. Results of fuel particle fabrication and analysis of potential in-reactor behavior will be presented and potential stable particle configurations discussed.


FJ-5:IL02  Microstructural Features of the High Burnup Structure
T. WISS, V.V. RONDINELLA, R.J.M. KONINGS, D. STAICU, D. PAPAIOANNOU, S. BREMIER, O. BENES, J.-Y. COLLE, P. PÖML, P. VAN UFFELEN, F. CAPPIA, European Commission, Joint Research Centre, Institute for Transuranium Elements, Karlsruhe, Germany

After a long residence time of the nuclear fuel in a reactor the High Burnup Structure (HBS) forms in the outer colder part of standard Light Water Reactor (LWR) uranium dioxide fuel pellets, where the local burnup is 2-3 times higher than the average pellet burnup. It is characterized by the sub-division of the initial grains into sub-micron sized grains and to a substantial increase of the porosity retaining the fission gases. Also commercial mixed oxide fuel (MOX)exhibits this high burnup structure in the Pu-rich phase.
The HBS needs to be well characterized to ensure that higher burnup fuel can be operated safely. This paper presents a synthesis of the main findings from extensive studies performed at JRC-ITU during the last 25 years to determine properties and behaviour of the HBS. In particular, thermal transport, fission gas behaviour, microstructural features and thermo-mechanical properties of the HBS will be discussed. The main conclusion of the experimental campaigns is that the HBS does not affect the safety of nuclear fuel during normal operation.
There are still investigations ongoing for a more accurate characterisation in order to understand the mechanisms governing this restructuring as well as the behaviour of ceramic fuels under harsh irradiation condition.


FJ-5:L03  Characterization of SiC Ceramic Tube Prepared by CVI/CVD Process
JI YEON PARK, DAEJONG KIM, WEON-JU KIM, Korea Atomic Energy Research Institute, Daejeon, Korea

For much slower degradation in a severe accident scenario: no meltdown, low corrosion rate and less/no hydrogen generation in the light water cooled reactor (LWR), triplex silicon carbide ceramic tube is strongly candidate as replaceable cladding from Zr alloys. In a triplex concept, a SiC tube consists of three SiC layers including a monolithic CVD SiC inner layer, a SiCf/SiC composite middle layer and a CVD SiC environmental barrier coating layer. To fabricate a SiC coated SiCf/SiC composite tube, the combined chemical vapor infiltration (CVI) and deposition (CVD) process was applied. In this study, a rotation speed and the source gas flow path and amounts were controlled as process parameters during fabrication of the SiC coated SiCf/SiC composites tube. A SiC preform was prepared using a filament winding method with a Tyranno-SA SiC fiber. Methyltrichlorosilane (MTS, CH3SiCl3) was used as a source precursor, and purified H2 was used as both a carrier and a diluent gas. The matrix filling behaviors were evaluated with process parameters. The microstructures of cross sections along the length and the longitudinal direction were observed by SEM and micro-CT and then, homogeneity and the porosity of tubes were also analysed. Additionally, the hoop strength was measured with compressive loading using elastomeric inserts.


FJ-5:IL04  Status of the Low Enriched Uranium Fuel Development for High Performance Research Reactors
L. SANNEN, S. VAN DEN BERGHE, SCK.CEN, Mol, Belgium

Historically, uranium enriched to >90%235U has been used for many peaceful applications requiring high fission densities such as driver fuels for research reactors. However, the use of high-enriched uranium or HEU (all enrichments >20% 235U are considered HEU) for civil applications, is considered a proliferation concern. Since the 1970's, efforts have been devoted to the conversion of research reactors operating on HEU to alternative fuels using uranium with enrichment below 20% or LEU. These efforts imply the development of high-density LEU fuels to replace the low volume-density (mostly) UAlx based HEU fuels. The paper updates the presents status of these developments focusing on the UMo dispersion fuel. It aims to provide an overview of the knowledge generated and the lessons learned in roughly 15 years of UMo dispersion fuel R&D in Europe through irradiation experiments and post-irradiation examinations (PIE).


FJ-5:IL05  Fabrication of Nuclear Fuel Beads by a Microfluidic Sol-gel Process
ZHEN-QI CHANG, X. LI, Y.T. YANG, M. ZHANG, Q.Y. HUANG, L.S. SHENG, SNST and INEST of USTC, HeFei, AnHui, P.R. China; C. SERRA, ICPEES - UMR 7515 CNRS and ECPM of UdS, Strasbourg, France

Fuel beads with precise controlled size, narrow size distribution and good sphericity are required for nuclear fuel applications. Here we present a sol-gel microfluidic technology for the preparation of closely size-controlled nuclear fuel beads with extremely narrow size distribution and excellent sphericity. Highly monodisperse CeO2 beads (CV＜5%) as a surrogate for PuO2 beads in the size range of tens to a thousand of microns with the frequency range of tens to few thousands of Hz were prepared precisely by manipulating continuous and dispersed phase flow rate or viscosity. Our works reveal that the relations between the bead size and fluid flow rate, fluid viscosity and capillary size can be described with an empirical equation ddrop/dcap = k(μcνc/μdνd)^-0.22. The porosity of CeO2 beads with the density range of 25% to 93% T.D. can be controlled by the addition of the porogen to the feed solution. The uniform uranium oxide beads with a various size were also prepared in as-built microsystem.


FJ-5:L07  Low-temperature Deposition of SiC Coatings in TRISO Particle Fuel and Evaluation of Properties
WEON-JU KIM, D. KIM, J.Y. PARK, Y.-K. KIM, M.S. CHO, Korea Atomic Energy Research Institute, Daejeon, Korea

TRISO-coated particles for fuels of high-temperature gas-cooled reactors (HTGRs) consist of UO2 microspheres coated with layers of porous pyrolytic carbon (porous PyC), inner dense PyC (IPyC), silicon carbide (SiC), and outer dense PyC (OPyC). Among the TRISO coating layers the SiC layer is particularly important because it acts as a diffusion barrier to gaseous and metallic fission products and as a miniature pressure vessel for the particle. In order to insure the integrity of the SiC layer after fabrication and in use, the microstructure, mechanical properties, and chemical composition of the SiC layer should be controlled properly. It has been known that the release rate of metallic fission products through the SiC layer is higher in a large columnar structure than with small grain sizes. The SiC layer in TRISO-coated particles is normally deposited at temperatures between 1500° and 1650°C. In this study, we investigated various microstructural and chemical features of SiC layers deposited at lower temperatures between 1300° and 1400°C to obtain a finer grain size while fixing the other deposition parameters. The effect of high-temperature annealing on the properties of the SiC coatings was also investigated.


Session FJ-6 - Radiation Effects

FJ-6:IL01  Modeling of Neutron Irradiation Embrittlement of Highly Irradiated Reactor Pressure Vessel Steels in Japan
N. SONEDA, K. NAKASHIMA, K. NISHIDA, A. NOMOTO, K. DOHI, CRIEPI, Yokosuka, Kanagawa, Japan

Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels of nuclear power plants is a phenomenon where the ductile-to-brittle transition temperature shifts towards higher temperatures and the upper-shelf toughness at high temperatures decreases due to neutron irradiation. Appropriate estimation of embrittlement is important for the safe operation of nuclear power plants, and this is achieved by the combinational use of the surveillance tests and embrittlement correlation method. A mechanism-guided embrittlement correlation method was issued in Japan as a part of the JEAC4201- 2007. After this, some new surveillance tests were carried out, and some materials irradiated to high fluences showed larger amount of embrittlement than the predictions. Detailed characterization of the radiation damage in such surveillance steels were performed using atom probe tomography (APT) and transmission electron microscopy (TEM) techniques to fully understand the mechanism of embrittlement, and then the revision of the embrittlement correlation method was performed. In this study, the details of the new surveillance data will be demonstrated, and then the revision of the embrittlement correlation method is described.


FJ-6:IL02  Irradiation Effects on RAFM Steel
E. GAGANIDZE, C. DETHLOFF, J. AKTAA, Karlsruhe Institute of Technology, Institute for Applied Materials, Eggenstein-Leopoldshafen, Germany

Structural materials for in-vessel components of future energy generating Fusion Reactors (FR) will be exposed to high neutron and thermo-mechanical loads. The degradation of the microstructure due to accumulation of an appreciable amount of displacement damage in combination with generated helium will strongly influence the performance of the structural materials. The understanding of the irradiation effects on the mechanical properties of the Reduced Activation Ferritic/Martensitic (RAFM) steels requires detailed investigation of the neutron irradiation induced microstructural defects. The selected mechanical properties will be reviewed for up to 80 dpa irradiated EUROFER97 and other RAFM steels. The results of the quantitative TEM investigation of the major radiation induced microstructural defects e.g. dislocation loops, voids/bubbles and radiation driven precipitation will be discussed. The neutron irradiation induced changes in the microstructure will be correlated to the changes in the tensile properties by applying the appropriate hardening models. Helium embrittlement will be assessed by reviewing the results obtained with selected helium simulating techniques. Recommendations on the operating temperature range for the First Wall and Breeding Blanket will be given.


FJ-6:L04  Effects of Solute Elements on Hardening of Thermally-aged RPV Model Alloys
K. NISHIDA, K. DOHI, A. NOMOTO, N. SONEDA, CRIEPI, Komae, Tokyo, Japan; L. LIU, N. SEKIMURA, Bunkyo-ku, Tokyo, Japan; Z. Li, Beijing, China

Neutron irradiation embrittlement of reactor pressure vessel (RPV) steels is caused by the formation of solute-atom clusters. Characterization of such clusters by atom probe tomography (APT) allows us to obtain information on the size and chemical composition of clusters, and it is now well known that such clusters typically consist of copper, nickel, manganese and silicon atoms, and the volume fraction of the solute atom clusters has good correlation with the corresponding amount of embrittlement. However, the role of each of the four elements on the formation of solute atom clusters is not necessarily well understood. In this study, we have performed the thermal ageing experiment of Fe-Cu based RPV model alloys with different additive chemical elements. Characterization of solute atom clusters were performed by means of APT, and Vickers hardness tests were performed to measure the corresponding mechanical property changes. nickel and manganese addition clearly accelerate the formation of solute atom clusters in terms of the number density and the cluster volume fraction during the initial stage of cluster formation. Detailed analysis on the role of solute elements will be discussed as well as the relationship between the cluster formation and the hardness change.


FJ-6:IL06  Neutron Induced Degradation of Plasma Facing Materials and Components
G. PINTSUK, J. LINKE, M. RÖDIG, Forschungszentrum Jülich, Euratom Association, Jülich, Germany

Besides high thermal fluxes and combined hydrogen/helium particle loads, neutron irradiation of plasma facing materials and components is one of the major issues in the field of nuclear fusion research. The correlated temperature dependent material degradation affects both, the thermo-mechanical and thermo-physical properties of the irradiated materials, and hence influences the maximum applicable heat flux and shortens the lifetime of a component.
In dedicated high heat flux tests in the electron beam facility JUDITH 1 at Forschungszentrum Juelich, reference and neutron irradiated tungsten, CFC, and beryllium materials and components were investigated with respect to their thermal fatigue and thermal shock performance. In both cases the decreasing operational performance is caused by radiation effects, e.g. reduction of thermal conductivity, especially for carbon fibre composites, an embrittlement of tungsten based materials, and the agglomeration of gaseous transmutation products in beryllium. Thereby, in particular in view of material erosion and subsequent plasma contamination, the behaviour under short transients has become an important issue.


FJ-6:L07  Experiments for Helium Bubble Formation in the Microstructure of EUROFER97 with Dual-beam Injection
O. TRÖBER, C. DETHLOFF, E. GAGANIDZE, J. AKTAA, Karlsruhe Institute for Technology, Institute for Applied Materials, Karlsruhe, Baden-Württemberg, Germany; D. BRIMBAL, P. TROCELLIER, L. BECK, CEA, DEN, Service de Recherches de Métallurgie Physique, Laboratoire JANNUS, Gif-sur-Yvette, France

Due to the neutron spectrum of a future fusion reactor the accumulation of displacement damage in the structural steels of in-vessel components will be accompanied by production of large amounts of transmutation helium. The impact of helium on the irradiation behaviour of the structural steels is not fully understood due to absence of a proper neutron irradiation facility. We have planned to simulate fusion relevant helium and dpa production (10 He appm/dpa) in the ferritic martensitic steel EUROFER97 with dual beam irradiation experiments at the JANNuS facility. The calculation of the necessary ion energies and penetration depths by TRIM software yielded that the targeted helium and dpa production can be achieved by using 3 MeV Fe2+ ions and 1.2 MeV He+ ions. The specimens will be irradiated at different temperatures from 330 to 550 °C to doses of 3.2x10e16 ions/cm²(Fe2+) and 1.1x10e16 ions/cm²(He+). The irradiated samples will be prepared by FIB and front-side electropolishing and examined by TEM. We will present the experimental setup, the sample preparation and first results for such irradiation experiments. The obtained results will be used for validation and improvement of a phenomenological rate theory model describing the helium microstructure evolution under irradiation.


FJ-6:L08  Evolution of Microstructure and Nanohardness in Hastelloy N Alloy under Xe26+ Ion Irradiation 
HEFEI HUANG, D.H. LI, J.J. LI, R.D. LIU, G.H. LEI, S.X. HE, Q. HUANG, L. YAN, Division of Nuclear Materials Science and Engineering, Shanghai Institute of Applied Physics, Chinese Academy of Science, Jiading district, Shanghai, China

Hastelloy N alloy, a nickel-base alloy that was invented at Oak Ridge National Laboratory (ORNL), is a good candidate for structural material of Molten Salt Reactor (MSR), because of its good oxidation and corrosion resistance to hot fluoride salts at high temperature (704 to 871°C).
In this paper, the ion irradiation effects of Hastelloy N alloy are reported. The prepared Hastelloy N alloy samples were irradiated by Xe26+ ions at room temperature with irradiation doses from 0.5 to 10 dpa. Nano-scale black spot damage appeared at a lower dose of 0.5 dpa. High number density of network dislocations were observed since the dose of 5 dpa, and grew significantly with the increasing ion dose. In addition, a certain of 3-8nm-diameter, roughly spherical precipitates existed in the samples become amporphous under ion irradiation.
The nanoindentation results show that the nanohardness of irradiated samples increases with the increasing ion dose. The link between irradiation induced microstructural evolution and hardening of Hastelloy N alloy was discussed.

 
Session FJ-7 - Materials Modelling and Database

FJ-7:IL01  Ab Initio Modelling of Dislocations in bcc Metals
F. WILLAIME, L. VENTELON, L. DÉZERALD, CEA Saclay, SRMP, Gif-sur-Yvette, France; D. RODNEY, Université Claude Bernard Lyon 1, ILM, France

The plastic deformation of crystalline materials is governed by the behaviour of dislocations. Quantitative modelling of their properties requires describing interatomic bonding at electronic structure level. We present the DFT study of dislocations in bcc transition metals (V, Nb, Ta, Cr, Mo, W and Fe). These metals form the basis of an important class of structural materials, going from ferritic steels to refractory alloys. In these metals, the electronic structure plays a predominant role, owing to the presence of a marked pseudogap in the electronic density of states. The objective is to provide in bcc transition metals a quantitative description from first principles of the energy landscape seen by dislocations, including in the material under stress. Properties such as core energy, Peierls potential, glide plane, Peierls stress and Schmid law deviation are determined. Their systematic variation as a function of row in the periodic table and column, i.e. d-band filling, is analyzed. Fe exhibits a specific and surprising behavior, with a low relative hard core energy close to that of the saddle configuration between easy cores. A line tension model based solely on the line tension and Peierls barrier values is used to determine the kink-pair formation enthalpy in Fe from DFT.


FJ-7:L03  Characterization and Modeling of the Ratcheting Behavior of the Ferritic-martensitic Steel P91
KUO ZHANG, J. AKTAA, Karlsruhe Institute of Technology (KIT), Institute for Applied Materials, Eggenstein Leopoldshafen, Germany

The ratcheting behavior of the 9%-Cr-1%-Mo ferritic-martensitic steel P91 is investigated by uniaxial cyclic loading tests at room temperature and 550°C. Ratcheting rates with different maximum tensile stresses are tested in order to build the database of P91 for the further application in generation IV fission reactors. The unconventional asymmetry of stress under symmetric strain-controlled tests yield the ratcheting with zero mean stress and this is also approved by the following symmetric stress-controlled tests. A viscoplastic deformation model taking into account the complex non-saturating cyclic softening of RAFM steels is further modified to adapt the ratcheting behavior of P91. It is revealed that the current model for RAFM steel fits cyclic softening behavior in strain-controlled LCF tests well, however it strongly overestimates the uniaxial ratcheting behavior in stress-controlled tests. A new constitutive model is developed based on further analysis of back stresses, which ideally fits the ratcheting behavior of P91 with different types and stress ranges of cyclic loading.


FJ-7:IL04  Modeling of Radiation Effects in SiC
I. SZLUFARSKA, D. MORGAN, M.J. ZHENG, C. JIANG, Department of Materials Science and Engineering, University of Wisconsin, Madison, WI, USA

SiC is a promising structural material for next generation nuclear reactors. In this talk we will discuss the current state of knowledge on radiation effects in SiC, with a focus on modelling studies. For example, our ab initio calculations have shown that there are significant energy barriers to defect recombination in SiC and these reactions are likely rate-controlling during radiation-induced amorphization. If there are other sinks of defects present in SiC (e.g., grain boundaries), the finite barriers to defect recombination can lead to a so-called interstitial starvation. In this phenomenon interstitials, which are mobile, migrate and are annihilated at grain boundaries, leaving behind unrecombined vacancies, which in turn drive amorphization of the material. We will also discuss how radiation resistance of SiC can be further improved by grain boundary engineering. For instance, we have shown both experimentally and in simulations that grain refinement to the nanometer regime combined with processing that results in a large number of low-energy grain boundaries and a high density of stacking faults, leads to significant improvements in radiation resistance of this material. Fundamental insights into this resistance brought by atomistic modelling will also be discussed.


FJ-7:IL05  Impact of Materials Modelling on Fusion Reactor Design
S.L. DUDAREV, EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxfordshire, UK

Neutrons and charged particles produced by nuclear reactions in the fuel assembly of a fission power plant or in the deuterium-tritium plasma of a fusion tokamak device induce significant changes in the physical and mechanical properties of materials. These changes result from processes occurring at atomic scale. For example, fast neutrons initiate collision cascades, in which radiation defects are formed. In contrast to collision cascades, transmutation reactions modify the chemical composition of irradiated materials. Understanding the effect of irradiation on materials requires developing multiscale models for microstructural evolution, which describe how the defects evolve and interact, and how the resulting microstructure responds to external mechanical stress or temperature gradients. The key parameters defining the response of irradiated materials to macroscopic engineering variables are the operating temperature, irradiation dose and dose rate. In this presentation I shall review information about temperature-dependent properties of candidate fusion materials, in the context of fusion reactor design effort and recent developments in the theoretical understanding of defect and dislocation microstructure derived from ab initio based atomistic and mesoscopic simulations.


FJ-5:IL06  Ab Initio Modelling of H and He Effects in Ferritic Alloys
CHU-CHUN FU, E. HAYWARD, LEI ZHANG, CEA, DEN, Service de Recherches de Métallurgie Physique, Gif sur Yvette, France

In future fusion reactors, high energy neutron irradiation is expected to generate large concentrations of hydrogen and helium via transmutation reactions, in addition to the self-defects. It is now well known that these gas elements cause damage to structural materials through bubble formation, embrittlement, swelling, and hardening. Taking α-iron as the basis material, we perform atomistic studies employing a combination of first principles, classical molecular dynamics, and Monte Carlo methods.
In this talk, we present the results on the dissolution, diffusion and clustering of helium and hydrogen in bcc iron, with particular focus on the dependence on local chemical and structural environments. In particular, the impact of impurities (carbon) and alloying elements (chromium) and of grain boundaries will be discussed. Finally, we provide some hints to explain the synergistic effects of helium and hydrogen as suggested by experiments

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

FJ-8:IL01  Steels for Fission and Fusion Applications
J.-L. BOUTARD, Cabinet du Haut-Commissaire à l'Energie Atomique, Gif sur Yvette Cedex, France

Steels used in fission reactors or envisaged for future thermonuclear devices will be presented with their operating conditions in terms of temperature, dose, mechanical stresses and chemical environments. Similarities and specificities of the various neutron spectra will be given. Then main radiation effects and their dependence on crystalline structure, chemical composition and microstructure will be discussed.
Solution-annealed 304 and cold-worked 316 type austenitic, and, low alloy steels are commonly used in Light Water Reactor (LWR) for the internal structures and reactor pressure vessel (RPV) respectively.
Internals have no safety role, but are undergoing rather high dose up to ~80 dpa at high temperature ~350°C, in-service conditions that will induce hardening, loss of ductility and fracture toughness. Possible swelling occurrence will be discussed, inducing dimensional changes and further loss of ductility. Radiation induced segregation is a general phenomenon of alloys under irradiation. In the case of internals austenitic steels, it results in grain-boundary Cr depletion that might promote irradiation assisted corrosion cracking (IASCC).
RPV is the second safety barrier of LWR and is a non-replaceable component. Ageing of RPV steels will determine the reactors life time. Present understanding of microstructure evolution is based on radiation induced segregation on point defects clusters formed in displacements cascades. Resulting hardening and fracture toughness degradation and their multi-scale modelling will be presented and discussed. This topic has certainly strong commonality with fusion 9-12% ferritic-martensitic steels envisaged for fusion tritium breeding blankets.
Improving resistance to radiation effects has been the driving force for developing fast reactor fuel cladding materials with improved resistance to swelling. The various improvements of austenitic steels in terms of chemical composition and cold-working will be briefly described. The good behaviour of 9-12%Cr ferritic-martensitic steels under fission neutron spectrum will be presented. Simulation of He production under fusion neutron spectrum and consequences on mechanical properties will be discussed. Creep resistance of 9-12% Cr ferritic-martensitic steels limited their use below 550°C. A way to improve heat and radiation resistance, both under fission and fusion spectrum, is to increase point defect sinks density. Following this route, recent developments of advanced 9-12%Cr steel and ODS steels will be given.
Finally main challenges in developing advanced steels to resist high temperature and dose will be given.


FJ-8:IL02  Low Activation Structural Materials for Fission and Fusion Power Reactors
V.M. CHERNOV, M.V.LEONTIEVA-SMIRNOVA, M.M. POTAPENKO, V.A. DROBYSHEV, D.A. BLOKHIN, JSC "A.A.Bochvar High-tecnology Research Institute of Inorganic Materials", Moscow, Russia; A.N.TYUMENTSEV, Tomsk State University, Tomsk, Russia; A.I. BLOKHIN, N.I. LOGINOV, A.I. LEYPUNSKY, Institute of Physics and Power Engineering, Obninsk, Russia

Results of the RF R&Ds of low activation materials (LAMAs) for cores of advanced nuclear fusion (neutron sources, DEMO) and fast fission (BN-1200, BREST, others) nuclear power reactors with liq uid metal coolants are presented. Vanadium alloys (V-4Ti-4Cr, V-Cr-W-Zr-C) and RAFMS RUSFER-EK-181 (Fe-12Cr-2W-V-Ta-B-C) are considered.
Problems of selection, modification, manufacturing and processing of the LAMAs are resolved for the most part. Further investigations are related with optimizations of chemical compositions and regimes of thermal-mechanical-chemical treatments of goods to produce the bulk highly homogeneous nanostructured heterophase states ensuring significant enhancement of functional properties and a minimization of impurities concentrations.
The complex ACDAM (codes and libraries) was revised and completed to calculate the nuclear characteristics (neutron absorption, primary radiation damage, activation, transmutation, post reactor cooling) of the LAMAs and coolants irradiated for a long time in neutron spectra of the fusion reactor DEMO-RF (15,3 dpa-Fe/y) and the RF fast power reactor BN-600 (60 dpa-Fe/y) are presented.
The operating temperatu re windows for the LAMA applications 300 C - 800(850) С (vanadium alloys) and 300 C - 670(700) С (RUSFER-EK-181) are recomended. Additional high-dose and high-temperature radiation properties of the LAMAs are necessary and their tests in the BN-600 (2013 - 2018, 60 - 160 dpa-Fe, irradiation temperatures 360 C- 715 C) are scheduled.


FJ-8:IL03  Development of Silicide Coatings on Vanadium Alloys 
S. MATHIEU, N. CHAIA, M. VILASI, IJL-Université de Lorraine, Vandoeuvre lès Nancy, France; F. ROUILLARD, J.L. COUROUAU, LECNA, CEA Saclay, France

The vanadium alloy V-4Cr-4Ti is an attractive structural material for fuel cladding in sodium-cooled fast reactors (SFR). However, the high affinity of vanadium alloys for oxygen, even when the environment contains a relatively low oxygen content such as in liquid sodium at moderate temperature (500-600°C), negatively affects the mechanical properties of alloys, causing embrittlement. Further developments of these alloys either for fission or fusion applications involve the use of protective coatings that i) have the ability to isolate vanadium from oxygen, ii) exhibit a low growing oxide scale at low oxygen pressure, iii) are non-reactive with liquid sodium and (iv) are thermodynamically stable with the vanadium alloy at the operating temperature. These criteria lead us to consider the development of silicide coatings. These coated materials exhibit a very low oxidation rate at 650 °C, both in air and at a low oxygen pressure (He, 5 ppm O2). This silicide developed a protective layer of silica at 650°C in air and was not susceptible to pesting. The coatings were largely unreactive to liquid sodium (<10 ppm O2) during 1000 h of compatibility test at 550 °C. Further details of the oxidation performances and thermal stability at high temperature (1000-1100°C) will be given.
 
 
Poster Presentations

FJ:P04  Microstructural Investigation of Tungsten Exposed to ITER Relevant ELM Loads of Deuterium Plasma Emitted by Powerful Plasma Focus Device
L. CIUPINSKI, Warsaw University of Technology, Warsaw, Poland

Extreme conditions in the present designed and the future fusion reactor chambers require use of modern materials characterized by resistance to high temperatures, thermal shocks, unique resistance to erosion, high thermal conductivity and optimally compatible with plasma. Tungsten (W) will be used in the high-flux region of the divertor in ITER and is a candidate material for plasma facing components in future fusion devices. The main purpose of this study was to investigate the structure and surface morphology of bulk tungsten exposed to fusion plasma generated by PF-1000 plasma focus device as a proxy of the Edge Localized Mode events envisaged in ITER. Scanning Electron Microscopy and Focused Ion Beam have been used to study the surface and cross sections of the exposed polycrystalline tungsten samples. Surface melting and evaporation as well as re-solidification/re-crystallization of near surface volume have been observed. The exposure in the PF device resulted also in development of cracks that propagated into the bulk material to some 500 microns in depth.