E.V. van Osch

Research and development on vanadium alloys for fusion applications

The current status of research and development on unirradiated and irradiated V-Cr-Ti alloys intended for fusion reactor structural applications is reviewed, with particular emphasis on the flow and fracture behavior of neutron-irradiated vanadium alloys. Recent progress on fabrication, joining, oxidation behavior, and the development of insulator coatings is also summarized. Fabrication of large (>500 kg) heats of V-4Cr-4Ti with properties similar to previous small laboratory heats has now been demonstrated. Impressive advances in the joining of thick sections of vanadium alloys using GTA and electron beam welds have been achieved in the past two years, although further improvements are still needed.

Simulation of disruptions on coatings and bulk materials

An important issue for the next step thermonuclear fusion devices is the development of a reliable engineering solution for the first wall and in particular for the divertor. Besides severe mechanical loads, these plasma facing components (PFC) will be subjected to energetic pulses during plasma disruptions. To evaluate the performance of plasma facing materials and thus to predict the lifetime of the PFCs, simulation experiments have been performed in the JAERI electron beam irradiation stand (JEBIS). Here different candidate plasma facing materials were evaluated in short beam pulses (1.2–10 ms) with an energy deposition of 2–9 MJ m −2 . The response of the individual materials to single and multiple shot beam pulses was investigated, and in these analyses special attention was given to the material erosion (melting, sublimation, particle emission). The quantification of these effects was done by weight loss measurements and by optical profilometry on the ablation craters. In addition, microstructural and morphological changes in the loaded surface were investigated.

Irradiation hardening of V-4Cr-4Ti

Abstract In the framework of the European Long Term Fusion Technology Program, Advanced Materials Field, ECN has been working on the assessment of low temperature irradiation hardening and embrittlement of vanadium alloys, as being developed for fusion application. Tensile, miniaturized Charpy impact (KLST) and Compact Tension specimens have been irradiated in the High Flux Reactor (HFR) in Petten up to approximately 6 dpa at 600 K. Three alloys were included; V–4Cr–4Ti from the 500 kg IEA reference heat provided by Argonne National Laboratory, and minor amounts of V–3Cr–3Ti and V–6Cr–6Ti, provided by Oak Ridge National Laboratory. The paper presents the results of tensile tests after irradiation. These tensile tests show strong hardening and reduction of ductility.

Plasma-sprayed boron carbide coatings for first-wall protection

Abstract Plasma-sprayed boron carbide coatings have been manufactured by different suppliers onto substrates of type 316L stainless steel. The coating thickness ranges from 0.3 to 2.0 mm. The larger thicknesses could only be achieved by application of an adaptive or gradient bond-layer between substrate and the boron carbide top coating. Measurements of the thermal diffusivity of coating materials are reported. Several high heat flux facilities have been used to study the thermal shock and erosion behaviour of the coated samples. A supporting numerical analysis of the thermal behaviour of the coating under normal and off-normal heat loads is presented, focussing on the differences between electron beam and laser beam tests due to volumetric energy deposition. Some aspects of the applicability of plasma sprayed B 4 C coatings for first-wall protection in a next step device are discussed.

Irradiation testing of 316l(N)-IG austenitic stainless steel for ITER

Abstract In the frame work of the European Fusion Technology Programme and the International Thermonuclear Experimental Reactor (ITER), ECN is investigating the irradiation behaviour of the structural materials for ITER. The main structural material for ITER is austenitic stainless steel Type 316L(N)-IG. The operating temperatures of (parts of) the components are envisaged to range between 350 and 700 K. A significant part of the dose–temperature domain of irradiation conditions relevant for ITER has already been explored, there is, however, very little data at about 600 K. Available data tend to indicate a maximum in the degradation of the mechanical properties after irradiation at this temperature, e.g. a minimum in ductility and a maximum of hardening. Therefore an irradiation program for plate material 316L(N)-IG, its Electron Beam (EB) weld and Tungsten Inert Gas (TIG) weld metal, and also including Hot Isostatically Pressed (HIP) 316L(N) powder and solid–solid joints, was set up in 1995. Irradiations have been carried out in the High Flux Reactor (HFR) in Petten at a temperature of 600 K, at dose levels from 1 to 10 dpa. The paper presents the currently available post-irradiation test results. Next to tensile and fracture toughness data on plate, EB and TIG welds, first results of powder HIP material are included.

Post-irradiation mechanical tests on F82H EB and TIG welds

Abstract The irradiation behaviour of electron beam (EB) and tungsten inert gas (TIG) welded joints of the reduced-activation martensitic steel IEA heat F82H-mod. was investigated by neutron irradiation experiments in the high flux reactor (HFR) in Petten. Mechanical test specimens, such as tensile specimens and KLST-type Charpy impact specimens, were neutron irradiated up to a dose level of 2–3 dpa at a temperature of 300°C in the HFR reactor in Petten. The tensile results for TIG and EB welds are as expected with practically no strain hardening capacity left. Considering impact properties, there is a large variation in impact properties for the TIG weld. The irradiation tends to shift the DBTT of particularly the EB welds to very high values, some cases even above +250°C. PWHT of EB-welded material gives a significant improvement of the DBTT and USE compared to the as-welded condition.

Low temperature irradiation experiments and material testing in Petten

Copyright (c) 1996 Elsevier Science B.V. All rights reserved. JRC, Petten Establishment, and ECN combine the full range of facilities required for the investigation of irradiation damage of fusion reactor structural materials. The high flux reactor (HFRr Petten, owned by the Commission of the European Communities, is one of the few large materials test reactors still in operation in Western Europe. The HFR is particularly suited for research on materials for future thermonuclear fusion reactors. The ECN hot cell laboratory has dedicated facilities for post-irradiation mechanical testing (fracture mechanics, tensile, fatigue, impact testing, creepr of materials. In addition, thermophysical properties can be measured, the weldability of irradiated materials is investigated with laser and TIG welding equipment. For post-testing analysis, various techniques are used, including optical microscopy and scanning and transmission electron microscopy. Materials being investigated for current fusion programmes include ITER-grade (IGr austenitic steel type 316LN-IG (plate, electron beam- and TIG-weldsr, vanadium alloys (e.g. V−4Cr−4Tir and ferritic-martensitic steels (e.g. MANET, F82H, JLF; plate, weldsr. Experimental details on irradiation and post-irradiation testing are presented in the paper.

Material erosion and surface temperature response to plasma-disruption heat load simulations

Abstract Carbon base candidate plasma facing materials have been subjected in the enhanced high power ECN-laser facility to short pulse high heat loads as anticipated to occur during plasma disruptions in next step devices like the International Thermonuclear Experimental Reactor or the Next European Torus. Recently, the first results of measurements on transient surface temperature response up to 3000 K, below erosion threshold, have been presented. After additional modification of the pyrometer, temperatures up to 4000 K can now be measured, permitting analysis of surface temperature response in relation to erosion. Qualitatively, the measured temperature responses are very similar to numerical calculations, and show clearly if erosion is taking place. Quantitatively the interpretation requires some caution, but reasonable fits are obtained. Some results of multiple shot experiments on carbon base materials are also presented. The incremental erosion is found to decrease only slighly after the first shot.

Residual tensile strength of neutron irradiated Inconel 718 bolts

In the frame of the ITER task BL 14.1, the residual strength of precipitation hardened Inconel 718 bolts, preloaded to stresses up to 90% of the yield strength, after neutron irradiation has been investigated. In the ITER design, the Inconel 718 bolts are used to attach the backplate to the radial supports and may experience large shocks from off-normal events during operation. Three irradiation rigs, containing 9 bolts each, were irradiated at the HFR at dose levels of 0.2, 0.7 and 1.1 dpa, respectively, at a temperature of 573 K. The bolts were designed to resemble a scaled-down version of the actual design and the irradiation conditions were chosen to resemble the actual operation conditions. Our results show that two competitive processes occur during neutron irradiation. At low dose, dissolution of precipitates is dominant, which causes a softening of the material. At higher doses, irradiation hardening takes place. The combined effect of the two processes is that the irradiation hardening is relatively small compared to other materials.

Structural materials by powder HIP for fusion reactors

Abstract Tokamak blankets have complex shapes and geometries with double curvature and embedded cooling channels. Usual manufacturing techniques such as forging, bending and welding generate very complex fabrication routes. Hot Isostatic Pressing (HIP) is a versatile and flexible fabrication technique that has a broad range of commercial applications. Powder HIP appears to be one of the most suitable techniques for the manufacturing of such complex shape components as fusion reactor modules. During the HIP cycle, consolidation of the powder is made and porosity in the material disappears. This involves a variation of 30% in volume of the component. These deformations are not isotropic due to temperature gradients in the part and the stiffness of the canister. This paper discusses the following points: (i) Availability of manufacturing process by powder HIP of 316LN stainless steel (ITER modules) and F82H martensitic steel (ITER Test Module and DEMO blanket) with properties equivalent to the forged one. (ii) Availability of powerful modelling techniques to simulate the densification of powder during the HIP cycle, and to control the deformation of components during consolidation by improving the canister design. (iii) Material data base needed for simulation of the HIP process, and the optimisation of canister geometry. (iv) Irradiation behaviour on powder HIP materials from preliminary results.