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Research and development projects on materials for applications exposed to high temperatures and radiation include comprehensive mechanical and metallographic examinations as well as analysis methods based on electron microscopy investigations using SEM (scanning electron microscope) and TEM (transmission electron microscope). In addition to 2D EDS and EBSD analyses in the µm range, the imaging 3D analysis method has become more and more important in the past few years. Particularly, the combination of SEM images with simultaneously obtained EDS and EBSD data during 3D analyses is a test method which complements previous material testing and produces clear results since much larger data sets are used in order to determine the composition of a material and the phases present in its microstructure. Since 2014, our institute has been working with a workstation (FIB/SEM) from the company Zeiss, a crossbeam AURIGA 40 with 3D analysis.

The influence of titanium alloying (Ti content of 0, 0.2, 0.3, and 0.4 wt %) on the nanostructure of yttrium oxide (Y2O3) dispersion strengthened steel with a composition Fe-13.5% Cr-2% W-0.3% Y2O3is investigated. The spatial distribution of chemical elements is analyzed in the investigated volumes. The matrix composition and average size and concentration of nanoscale clusters are compared for different samples. It is shown that the average nanocluster size (∼3 nm) is almost unchanged with increasing Ti concentration, while the cluster concentration grows from ∼1 × 1023m−3(for Ti-free steel) to ∼1.5 × 1024m−3(for 0.4 wt % Ti alloy).

SANS was used to investigate the growth of He-bubbles in a martensitic steel developed for fusion reactors ("MANET"). The investigated sample were simultaneously irradiated at high temperature with high-energy protons and alpha particles and then submitted to fatigue. Differences in the SANS curves were observed between the irradiated and non irradiated samples ; they were attributed to radiation induced inhomogeneities and in particular to helium bubbles.

AbstractAs the core region of IFMIF, the test cell (TC) suffers intense neutron and gamma irradiations. Major material challenges of the TC faced during engineering design phase are outlined and the current key material allocations are described. Actively cooled magnetite concrete is selected as the major biological shielding material for the TC, and actively cooled closed liner made of 316L stainless steel is selected to cover the complete TC internal surfaces. Material selections for sealing gaskets and electric insulations inside the TC are preliminarily defined based on dose rate maps at different locations. Metal based sealing gaskets and glass/ceramic electric insulations are applied in the areas with high dose rate, while organic based gaskets and conventional insulation materials can only be arranged behind sufficient biological shielding. Leak tight welding seams between removable interface shielding plugs and the TC liner are located in the region with very low helium generation rate (<0.01 appm/fpy) in steel so that cutting and re-welding during the complete IFMIF life span is guaranteed.

AbstractIn future nuclear fusion reactors, structural materials will undergo a large thermal cycling due to pulsed operation and the occurrence of several maintenance periods. Therefore, the investigation of the combined role of creep and fatigue loading is of major importance. In this study, we focused on Eurofer 3 electron beam welds. Two different post-welding heat treatments were carried out: a two-step heat treatment (30 minutes at 980°C followed by 2 hours at 750°C) and a one-step heat treatment (2 hours at 750°C). Fatigue, creep and creep-fatigue tests were performed. A 550°C test temperature was chosen, corresponding to the upper operation temperature currently foreseen for this material. Creep-fatigue experiments were achieved by interrupting a fatigue test and then applying a creep loading until the fracture of the specimen. Several fatigue pre-stress conditions were studied. The post-weld heat treatment influence was analyzed. The damage contributions of fatigue and creep were studied using electron microscopy. The results were compared to previous results obtained on base material.

Abstract. Ferritic-martensitic steels like Eurofer-97 are candidate structural materials for future fusion reactors. In the tempered state, this steel contains fine particles dispersed in the ferritic matrix. The aim of this work is to investigate abnormal grain growth in Eurofer-97 steel. The microstructural evolution was followed by isothermal annealing between 200 and 800°C (ferritic phase field) after cold rolling to 70, 80, and 90% reductions. Representative samples were characterized by scanning electron microscopy in the backscattered electron mode. Microtexture was evaluated by electron backscattered diffraction. We propose a mechanism based on the size advantage acquired by nuclei with misorientation angles above 45º relative to their nearest neighbors to explain abnormal grain growth. Abnormal grain growth textures have components belonging to the α- and γ-fibers with predominance of {111}, {111}, and {100}.

Four alloys on the basis of 8-10% CrWVTa steel were melted to study the influence of boron on the structure and precipitation behavior. Boron contents in the alloys amounted to 0.0083 and 0.1 wt.%, respectively. One batch was nearly free of boron. The individual alloys were examined metallographically and characterized in the initial state and after aging treatments up to 800 °C. It was found that the grains of boron-containing alloys were coarsened when the alloys were subjected to tempering. At the same time, the precipitate content was reduced. The different phases of precipitates were analyzed as primary carbides with tantalum and tungsten, and due to the higher content of boron BN, B(C, N), and (Fe, Cr)2B. M23C6 was developed in all alloys after the heat treatment.

Reduced activation ferritic-martensitic steels are considered for future applications in fusion power plants. The thermal stability of two steels (ODS-Eurofer and Eurofer-97) was evaluated in cold-rolled samples annealed below 800oC. These grades have similar chemical compositions except for the presence of nanosized Y2O3 particles in ODS-Eurofer steel. This nanosized dispersion is very effective to prevent recrystallization. On the other hand, full recrystallization occurs in Eurofer-97 steel when annealed above 700oC. The low volume fraction of recrystallized grains (&lt; 0.1) in ODS-Eurofer steel annealed at 800oC can be explained by nucleation at prior grain boundaries and around large M23C6 particles. Further growth of these nuclei is impeded by Zener drag and concurrent recovery. Static recovery is the main softening mechanism in the ODS-steel below 800oC.

AbstractOxide-dispersion-strengthened (ODS) ferritic-martensitic steels are candidates for applications in fusion power plants where microstructural long-term stability at temperatures of ˜650°C to 700°C are required. The microstructural stability of 80% cold-rolled reduced-activation ferritic-martensitic 9% Cr ODS-Eurofer steel was investigated within a wide range of temperatures (300°C to 1350°C). Fine oxide dispersion is very effective to prevent recrystallization in the ferritic phase field. The low recrystallized volume fraction (<0.1) found in samples annealed at 800°C is associated with the nuclei found at prior grain boundaries and around coarse M23C6particles. The combination of retarding effects such as Zener drag and concurrent recovery decrease the local stored energy and impede further growth of the recrystallization nuclei. Above 900°C, martensitic transformation takes place with consequent coarsening. Significant changes in crystallographic texture are also reported.

Complementing the efforts towards the realization of ITER, KIT is pursuing, within the overall EURATOM fusion program, a number of important long-term technology developments towards a magnetic confinement fusion power plant (FPP), taking into account the features that will distinguish such facility from ITER.To this end, structural materials on the basis of both low-activation steels and refractory metals, as well as concepts for breeding blankets and divertor designs, are being developed along with suitable manufacturing and joining technologies. In parallel, KIT contributes to the engineering design and validation phase of the International Fusion Materials Irradiation Facility (IFMIF) necessary for qualifying the materials to be used in an FPP. The specific characteristics of an FPP fuel cycle, i.e., substantial tritium quantities within huge mass flows of gases and the related tritium compatible high throughput vacuum and pumping technologies, are being translated into viable engineering approaches. High temperature superconducting magnet solutions are being developed, with a view to overall plant efficiency. In order to increase the wall-plug efficiency of plasma heating, advanced gyrotron tubes with power levels significantly beyond what is envisaged for ITER are being developed along with a frequency tunability option for efficiently counteracting plasma instabilities.

Specimens of the European reference 9Cr-WTaV steel EUROFER 97 alloyed with ~0.1 wt.% 10B and 0.5 wt.% Cu were neutron irradiated up to 16 dpa at 300 and 400 °C to study an interplay between helium generated by nuclear reaction on 10B and copper precipitates on microstructure and mechanical properties. Before irradiation 2 – 3 nm copper clusters often trapped on grain or lath boundaries were observed. It was shown that after irradiation copper segregates either in the bulk material by forming Cu-clusters which are in many cases attached to helium bubbles or it further precipitates at lath and grain boundaries. The size of both copper clusters and helium bubbles increases with radiation temperature.

AbstractOxide dispersion strengthened (ODS) ferritic/martensitic (FM) steels are promising candidates for structural applications in future fusion power reactors. In order to evaluate the thermal stability of 80% cold-rolled ODS-EUROFER, samples were annealed for 1 h at temperatures up to about 0.9 Tm, where Tmis the absolute melting point. The characterization of the annealed samples was performed using transmission electron microscopy and electron backscatter diffraction. Results show that static recovery is the main softening mechanism of this steel when annealed below 800°C. The volume fraction of recrystallized grains is quite small (below 0.10). Above 900°C, martensitic transformation takes place causing pronounced hardening. Large M23C6particles are found at the grain boundaries after tempering at 750°C for 2 h.

High performance materials are indispensable for economical and environmentally acceptable operation of fusion power and innovative fission reactor systems. In order to provide materials design data for short term needs of the ITER Test Blanket Modules and for a Fusion Demonstration Reactor, the international fusion materials community is concentrating their R&D work mainly on the four reduced-activation structural materials 8 – 9 CrWTa ferritic/martensitic steels, V – Ti – Cr alloys, W alloys and SiCf/SiC composites as well as on functional materials such as CVD diamond windows. Besides the development of the European reduced activation reference steel EUROFER, emphasis will be placed on synergies between fusion and fission materials R&D, namely the development of ductile oxide dispersion strengthened ferritic/martensitic steels with nano-scaled Y2O3or Y2Ti2O7particles that increase not only high temperature strength but also substantially the microstructural stability against irradiation embrittlement and high temperature ageing. Finally, the corrosion behaviour of zirconium-1.5 wt.% tin alloy, a well established cladding material for nuclear fuel rods, will be shown in different environments.

AbstractTritium as one of the two fuel components for fusion power plays a special role in any fusion device. Due to its volatile character, radioactivity and easy incorporation as HTO it needs to be controlled with special care and due to its scarcity on earth it has to be produced in-situ in future fusion power plants. The paper discusses the present tritium R&D activities in fusion ongoing in the EU and presents the various processes/techniques envisaged for controlling tritium in future fusion reactors focusing mainly on the issues of breeding blankets and the fuel cycle in DEMO.

The development of first wall, blanket and divertor materials that are capable of withstanding for many years high neutron and heat fluxes, is a critical path to fusion power. Therefore, the timely availability of a sound materials database has become an indispensable element in international fusion road maps. In order to provide materials design data for the short-term needs of ITER test blanket modules (TBMs) and for a DEMOnstration fusion reactor, a wealth of R&D results on the European reduced activation ferritic–martensitic steel EUROFER, and on oxide dispersion strengthened variants are being characterized, mainly in the temperature window 250–650°C. The characterization includes irradiations up to 15 dpa in the mixed spectrum reactor HFR and up to 75 dpa in the fast breeder reactor BOR60. Industrial EUROFER-batches of 3.5 and 7.5 tons have been produced with a variety of semi-finished, quality-assured product forms. To increase the thermal efficiency of blankets, high temperature resistant SiC_f/SiC channel inserts for liquid metal coolant tubes are also developed. Regarding radiation damage resistance, broad based reactor irradiation programmes deal with several ranges from ≤ 5 dpa (ITER TBMs) up to 75 dpa (DEMO). For the European divertor designers, a materials database is currently being set up for pure W and its alloys, and related reactor irradiations are foreseen with temperatures from 650°C to 1000°C.

Ein stochastisches Modell zur Simulation der Schädigungsentwicklung in dem martensitischen Stahl F82H-mod unter Ermüdungsbelastung wurde entwickelt. Empirische Beziehungen für Rissinitiierung, Risswachstum und Zusammenlagerung von Rissen werden aus experimentell beobachteten Rissmustern abgeleitet und in ein stochastisches Simulationsmodell basierend auf einer zufälligen Zellstruktur übertragen. Die Simulationsergebnisse werden mit experimentellen Rissmustern verglichen.

The inherent properties of reduced activation ferritic/martensitic (RAFM) steels include reduced swelling and high recycling potential, which make them likely candidates for application in commercial fusion power plants. The International Energy Agency (IEA) agreement has been an effective framework for international co-operation in developing RAFM steels. The progress and critical issues observed in this co-operation are reported. The production of RAFM steels on an industrial scale has been demonstrated. Various methods of fusion welding and solid hot isostatic pressing (HIP) are feasible for joining the steels. Manufacturing of complex shapes with the powder HIP method works well for RAFM steels. Major critical issues addressed concern the effects of simultaneous introduction of helium and displacement damage. The availability of a 14 MeV neutron source is identified as an essential tool to determine this effect. Finally, the potential of oxide dispersion strengthening to increase the operating temperature of RAFM steels is considered as an issue that has to be resolved to enlarge the application temperature window of RAFM steels.

This paper presents the results of a small-angle neutron scattering (SANS) study of He-bubble growth in martensitic steel F82H-mod. The investigated samples had been homogeneously implanted with 400 appm He at 250°C, then submitted to 2 h annealings up to 975°C, together with a non-implanted sample of the same material (one for each temperature) serving as a reference to isolate the SANS signal arising from the He-bubbles. An as-implanted sample was also investigated together with its thermal reference. The He-bubble volume distributions, obtained from the SANS data and discussed with reference to uncertainties arising from background subtraction, show that the uniform distribution of the as-implanted sample, with bubbles approximately 15 Å in size, evolves during high temperature annealing into a bimodal one with a secondary population of bubbles as large as 100 Å approximately.

μSR experiments were performed on electron-irradiated magnetic samples. In dilute iron alloys, the isolated vacancy (30 MHz line) was observed up toTA (annealing temperature) = 200 K. Above this temperature a new line (30 MHz +δv), assigned to vacancy-impurity pairs, was found in some alloys. The frequency shiftδv and the annealing behaviour of the new line are characteristic for each impurity. In electron-irradiated Nino newμSR line was observed.

The relaxation rateλ* of anomalous muonium in silicon with different dopant concentrations was investigated as a function of temperature. Below 140 K, a close correlation betweenλ* and the concentration of conduction electrons was found. We conclude from this behavior that the relaxation of anomalous muonium in this temperature region is caused by the scattering of conduction electrons. A microscopic model is developed and the value of the exchange interactionJex is derived.

Muon Spin Rotation (γSR) experiments were performed on diluted (400 ppm) iron alloys. The annealing behaviour of electron irradiated samples was studied. We found that below TA=200 K, muons are trapped at isolated vacancies (ν=30 MHz). Around 200 K the isolated vacancies disappear and, in some of the samples, a new signal, which is assigned to a vacancy-impurity pair, shows up. From the disappearance of this signal, the vacancy-impurity binding energy is derived.

In order to simulate fusion neutron displacement damage, irradiations with light ions can be used. For this purpose irradiations of iron with protons and α-particles were experimentally and theoretically investigated. The main emphasis was put on the behavior of displacement damage production caused by high-energetic recoils induced by charged particle nuclear reactions. Evaluation of excitation functions and mean recoil ranges lead to non-elastic PKA-spectra. The comparison between charged particle and fusion neutron induced PKA-spectra points out, that the charged particle-induced non-elastic fraction contributes a considerable part to the PKA-spectra. The closest fit was found for the 20 MeV proton-induced PKA-spectrum. Further comparisons of DPA-data show, that at the highest examined energies the non-elastic/total DPA-ratio is 0.43 for protons and 0.47 for α-particles.

Aluminum nitride has been proposed as electrically insulating coating in connection with blankets made of vanadium alloys. Irradiations were performed with an applied DC electric field of 100 kV/m, using a 104 MeV α-particle beam. The dose achieved were 0.1 dpa at 300°C, 0.05 dpa at 400°C, and 0.06 dpa at 500°C. The out-of-beam conductivity σ0decreased at all temperatures employed during early stage of irradiation. Before 0.02 dpa had been reached, σ0dropped to low values of typically (2–4) × 10−11(Ω m)−1. Although a moderate increase can be observed during further irradiation, σ0remained below the initial value of the unirradiated specimen for all temperatures. That is, there is no significant evidence for radiation induced electrical degradation. The in-beam conductivity σ, measured continuously during irradiation, also decreased by at least one order of magnitude. Relevant damage dose and ionization rate dependencies of σ and σ0will be discussed.

Strain controlled fatigue experiments on the modified ferritic—martensitic 12% Cr-steel Manet have been performed during, after and without irradiation. Continuous push—pull cycling of the hollow fatigue specimens with a cycle ratio R= −1, an axial strain rate of 8 · 10−4s−1and total strain ranges Δεtbetween 0.42% and 1.0% has been applied in vacuum (5 · 10−3Pa) at T= 250°C. For the irradiation, a 104 MeV α-particle beam was used. Compared to unirradiated specimens, in-beam specimens undergo irradiation hardening at an advanced fatigue/implantation stage. Post-irradiation specimens show significantly higher total stress amplitudes and reduced fatigue lives. TEM investigations of in-beam and post-irradiation specimens several helium bubbles with different diameters on internal surfaces and in the matrix.

Tensile specimens of the normalized and tempered fully martensitic steel DIN 1.4914 were helium implanted up to 340 appm within the temperature range 300–720°C using a 104 MeV alpha particle beam. The high He/damage ratio of 1850 appm He/dpa allowed to determine the effect of helium on material properties. After irradiation tensile strength and ductility were investigated mainly at test temperatures equal to the irradiation temperatures. While short time irradiations with helium contents up to 100 appm have shown no visible effect on the tensile behaviour at all temperatures investigated, long time irradiations reduced the material strength at irradiation temperatures above about 440°C. This irradiation induced softening obscured the predicted minor hardening effect of the observed He-bubbles completely and increases with irradiation time and temperature. The fracture mode of helium-implanted and control specimens is found to remain always ductile and transgranular.