A. Möslang

Status of R&D activities on materials for fusion power reactors

Current R&D activities on materials for fusion power reactors are mainly focused on plasma facing, structural and tritium breeding materials for plasma facing (first wall, divertor) and breeding blanket components. Most of these activities are being performed in Europe, Japan, the Peoples Republic of China, Russia and the USA. They relate to the development of new high temperature, radiation resistant materials, the development of coatings that will act as erosion, corrosion, permeation and/or electrical/MHD barriers, characterization of candidate materials in terms of mechanical and physical properties, assessment of irradiation effects, compatibility experiments, development of reliable joints, and development and/or validation of design rules. Priorities defined worldwide in the field of materials for fusion power reactors are summarized, as well as the main achievements obtained during the last few years and the near-term perspectives in the different investigation areas.

Innovative materials for energy technology

Abstract 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 Y2O3 or Y2Ti2O7 particles 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.

Influence of helium on impact properties of reduced-activation ferritic/martensitic Cr-steels

Abstract Instrumented Charpy impact tests of the reduced activation type 8Cr2WVTa steel F82H have been performed after homogeneous implantation of 300 appm helium at 250°C. The results are compared with investigations on mixed spectrum neutron irradiated (HFR Petten) specimens. After neutron irradiation at 250°C to the same low damage dose of 0.2 dpa, the ductile–brittle transition temperature shift (ΔDBTT) amounts to 18°C, whereas a much higher ΔDBTT of 42°C has been measured after helium implantation. These results are compared with other neutron irradiated ferritic/martensitic steels having different boron levels and thus different helium contents. A model is proposed which describes the dynamic brittle fracture of martensitic/ferritic steels by a stress-induced propagation of micro-cracks, taking into account radiation induced hardening as well as helium bubble formation.

Multimodal options for materials research to advance the basis for fusion energy in the ITER era

Well-coordinated international fusion materials research on multiple fundamental feasibility issues can serve an important role during the next ten years. Due to differences in national timelines and fusion device concepts, a parallel-track (multimodal) approach is currently being used for developing fusion energy. An overview is given of the current state-of-the-art of major candidate materials systems for next-step fusion reactors, including a summary of existing knowledge regarding operating temperature and neutron irradiation fluence limits due to high-temperature strength and radiation damage considerations, coolant compatibility information, and current industrial manufacturing capabilities. There are two inter-related overarching objectives of fusion materials research to be performed in the next decade: (1) understanding materials science phenomena in the demanding DT fusion energy environment, and (2) application of this knowledge to develop and qualify materials to provide the basis for next-step facility construction authorization by funding agencies and public safety licensing authorities. The critical issues and prospects for development of high-performance fusion materials are discussed along with recent research results and planned activities of the international materials research community.

Overview of the IFMIF test facility

Abstract During the past few years, a reference design has been developed for the International Fusion Materials Irradiation Facility (IFMIF). According to the mission and specification of the general requirements, this reference design includes relevant machine parameters and conceptual designs for the major device subsystems – Test Facilities, Lithium Target Facilities and Accelerator Facilities. Major engineering efforts have been undertaken to establish a test cell design that follows closely the users requirements of the fusion materials community and allows safe and completely remote controlled handling. After a short description of the facility requirements, concepts for the two independent test cells, various test assemblies, remote handling equipment and hot cell facilities are presented.

New method for detection of Li inside He bubbles formed in B10-alloyed steel after neutron irradiation.

Electron energy loss spectroscopy (EELS) was used to detect and study the spatial distribution on the nanoscale of He and Li in boron-alloyed steel after neutron irradiation. Li and He are the products of the (10)B(n, α)(7)Li nuclear transmutation reaction and knowledge of their distribution is important to understand their influence on mechanical properties. Here, a new method is presented for the direct detection of Li in Fe, which is based on the analysis of the plasmon structure in EELS spectra. Li drops or particles in He bubbles show pronounced Li plasmon line at 10eV which can be extracted from the Fe/Cr plasmon. The Gaussian or linear interpolation of the Fe/Cr plasmon and its subtraction allows for the calculation of Li and He two-dimensional maps and the study their spatial distribution. The analysis of Li plasmon fine structure allows imaging surface effects in the Li drops.

The international fusion materials irradiation facility IFMIF : an overview of user aspects

Abstract For the future fusion materials development program it is necessary to have a dedicated high intensity neutron source available. IFMIF, an accelerator-based intense D–Li neutron source that produces neutrons with a suitable energy spectrum at high intensity and sufficient irradiation volume, is a near term feasible concept which fulfills all essential requirements of the fusion materials development program defined by the users community. IFMIF performs all necessary kinds of material tests with high flexibility in the materials test matrix and the very attractive capability of accelerated irradiations up to about 50 dpa per full power year.

Study of He-bubble growth in α-particle implanted F82H-mod martensitic steel

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 A in size, evolves during high temperature annealing into a bimodal one with a secondary population of bubbles as large as 100 A approximately.

Design concept for the IFMIF test assemblies

Abstract According to specific materials testing needs, the available irradiation volume of the International Fusion Materials Irradiation Facility (IFMIF) has been partitioned into a high-flux region that has an irradiation damage rate for iron of greater than 20 displacements per atom (dpa) per full-power year (fpy), a medium-flux region with 1–20 dpa/fpy, a low-flux region with fluence levels between 0.1 and 1 dpa/fpy, and a very low-flux zone with an annual accumulation of less than 0.1 dpa. A set of test assemblies, which is located immediately downstream from the neutron producing Li-target, vertically supports (1) test modules used for long-term irradiation of specimens in the high- and medium-flux regions, and (2) an array of tubes used for inserting test capsules in the low- and very low-flux regions. Based on this arrangement and on small specimen test technologies, IFMIF is capable of providing sufficient volume and appropriate irradiation environments to meet the requirements defined by the user community.

Microstructural investigation of low-dose neutron irradiation effects in martensitic steels for nuclear application by means of small-angle neutron scattering

The microstructural effect of low-dose neutron irradiation and subsequent high-temperature tempering in the reduced activation ferritic/martensitic steel F82H-mod. (7.73 Cr, 0.09 C, 0.08 Mn, 0.19 V, 2.06 W, 0.02 Ta, wt%, bal. Fe) has been studied using small-angle neutron scattering (SANS). The investigated samples were irradiated with thermal neutrons at 523 K, to dose levels of 2.4 displacements per atom then tempered for 2 h at 1043 K. The SANS measurements were carried out at the D22 instrument of the High Flux Reactor at the Institut Max von Laue–Paul Langevin, Grenoble, France. The differences observed in nuclear and magnetic small-angle neutron scattering cross-sections after subtraction of the reference sample from the irradiated one suggest that the irradiation and the subsequent post-irradiation tempering produce the growth of non-magnetic precipitates; the results are also compared with those obtained on other ferritic/martensitic steels, with different chemical composition, irradiated under the same conditions.