A. Cardella

Assessment and selection of materials for ITER in-vessel components

Abstract During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)–IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti–6Al–4V alloy and two copper alloys, CuCrZr–IG and CuAl25–IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

Armor and heat sink materials joining technologies development for ITER plasma facing components

An extensive program on the development of the joining technologies between armor (beryllium, tungsten and carbon fibre composites) and copper alloys heat sink materials for ITER plasma facing components (PFCs) has been carried out by ITER home teams. A brief review of this R&D program is presented in this paper. The critical problems related to these joints are described. Based on the results of this program and new requirements on the reduction the manufacturing cost of ITER PFC, reference technologies for use in ITER have been selected and recommended for further development.

Design and Material Selection for ITER First Wall/Blanket, Divertor and Vacuum Vessel

Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

ITER plasma facing components, design and development

Abstract The paper summarizes the collaborative effort of the ITER Conceptual Design Activity (CDA) on Plasma Facing Components (PFC) which focused on the following main tasks: (a) the definition of basic design concepts for the First Wall (FW) and Divertor Plates (DP), (b) the analysis of the performance and likely lifetime of these PFC designs including the identification of major critical issues, (c) the start of R&D work giving already first results, and the definition of the required further R&D program to support the contemplated ITER Engineering Design Activity (EDA). From the ITER CDA effort on PFC it is mainly concluded that: (a) the expected PFC operating conditions lead to design solutions at the limit of present technology in particular for the divertor, which may constrain the overall machine performance, (b) the development of convincing PFC designs requires an intensified R&D effort both on PFC technology and plasma physics.

Engineering design of the ITER blanket and relevant research and development results

The design of the ITER blanket is presented together with the related technology which has been developed. The evolution of this component since the beginning of the EDA is explained in relation to the developing understanding of the thermal deformations and of the electromagnetic forces. These loads lead to a system composed of compact modules protecting a continuous support shell called a backplate. The backplate is a stiff double wall construction which conveys the coolant to the modules. The supports of the module are flexible and allow relative thermal expansions. They are connected and disconnected to the backplate by bolts operated through holes in the front face of the module. The coolant connections and the electrical straps located on the back of the modules are reached similarly. The first wall is integral with the module and cooled in series. A research and development program on materials and joining methods defined the construction path which has been tested in prototypes. The main body is built of stainless steel by forging and drilling or powder hot isostatic pressing (HIP), depending on the complexity of the shape. The first wall includes a dispersion strengthened copper heat sink which is hot isostatic pressed onto the steel body. Beryllium is the basic plasma facing material and is attached by HIP to the copper. Prototypes of the module attachment have been built and are under integrated tests.

Design of the ITER EDA plasma facing components

Abstract The design of the plasma facing components (PFC) in ITER has evolved with the detailed design of the reactor. The structures exposed to the plasma have different requirements according to their functions. The primary wall, surrounding most of the plasma along the last closed magnetic surface, is exposed to a moderate heat flux (0.5 MW m−2) but has to withstand the highest neutron load. The baffle wall is exposed to a peak heat flux of 3 MW m−2 and to severe erosion from neutral particles due to their high neutrals pressure in the divertor. The limiter is subjected to the same loads as the primary wall during plasma burn conditions and a higher peak heat flux (depending on its location) during the start-up and shut down phases when the plasma is leaning on its surface. The divertor vertical targets intercept the open magnetic flux surfaces near the separatrix and have to withstand the highest heat flux and erosion in their lower part. The divertor dome is located directly below the null point and works in conditions similar to the baffle. The divertor wings receive similar thermal loads as the dome but can be subjected to high heat shocks and electromagnetic forces during plasma disruption. The paper describes the solutions adopted for the PFC and the results of analyses performed to validate the design. The description is focused on the part of the PFC which is exposed to the plasma.

ITER first wall:shield blanket

The blanket system comprises first wall/shield modules which are supported on a structural shell called a Back Plate containing water channels for cooling the modules. A modular design of the first wall/shield was chosen to allow independent maintenance through the horizontal mid-plane ports using in-vessel remote handling equipment. The modules are mechanically attached to the back plate for assembly and maintenance. The first wall/shield modules can be replaced with breeding modules for the EPP. In the first wall (FW) region, the water is contained in 316 LN stainless steel pipes surrounded by a copper heat sink except in high heat flux regions where copper pipes with a stainless steel liner are used in the limiter and baffles. The plasma facing surface of the FW will be beryllium except for the lower region of the baffles, where tungsten is used. Electromagnetic analysis and structural analysis has been performed on the back plate and FW/shield modules during a plasma disruption and for a VDE with toroidally asymmetric halo currents. A major R and D project is being conducted to provide needed input for design of the blanket system and to confirm the fabrication technology.

FW/Blanket and vacuum vessel for RTO/RC ITER

Abstract The design has progressed on the vacuum vessel and First Wall (FW)/blanket for the Reduced Technical Objective/Reduced Cost (RTO/RC) ITER. The basic functions and structures are the same as for the 1998 ITER design. The design has been improved to achieve, along with the size reduction, ∼50% target reduction of the fabrication cost. The number of blanket modules has been minimized according to smaller dimensions of the machine and a higher payload capacity of the blanket Remote Handling tool. A concept without the back plate has been designed and assessed. The blanket module concept with flat separable FW panels has been developed to reduce the fabrication cost and future radioactive waste.

Design and test of the support elements of the W7-X superconducting magnets

The Wendelstein 7-X stellarator is presently under construction at the Max-Planck-Institute for Plasma Physics in Greifswald with the goal to verify that a stellarator magnetic confinement concept is a viable option for a fusion power plant. The superconducting magnet system has to fulfill demanding requirements regarding magnetic field, loads, manufacturing and assembly. The magnet support system consists of several types of structural components. The main one is the central support structure (CSS) to which the superconducting coils are connected through Central Support Elements (CSE). These are bolted interfaces that allow for flange opening to reduce loads on the components. The non-planar coils (NPC) are toroidially interconnected via lateral support elements (LSE) and narrow support elements (NSE). NSE are contact supports consisting of Al bronze pads that allow for sliding under large compressive loads between the coils. The planar coils (PC) are connected to the NPC through planar support elements (PSE). At the module and half-module separation planes Contact Elements (CTE) connect the neighbouring NPC. An integrated programme of design, FE analysis, experiments and assembly trials has been undertaken. The NSE experimental program provided confidence that the pads can cope with the requirements regarding loads and cycles. Weld trials provided procedures for installing the LSE whilst keeping shrinkage and distortion within tight limits. Tests have been carried out to provide insight on the functioning of the CSE, in particular of the bolts and high performance Superboltreg-nuts during pre-load. This paper gives an overview of the integrated program on the W7-X support elements.

Recent progress in the modelling of helium and tritium behaviour in irradiated beryllium pebbles

Abstract One of the key issues of the European Helium Cooled Pebble Bed blanket is the behaviour under irradiation of beryllium pebbles, which have the function of neutron multiplier. An intense production of helium occurs in-pile, as well as a non negligible generation of tritium. Helium bubbles induce swelling and a high tritium inventory is a safety issue. Extensive studies for a better understanding, characterisation and modelling of the behaviour of helium and tritium in irradiated beryllium pebbles are being carried out, with the final aim to enable a reliable prediction of gas release and swelling in the full range of operating and accidental conditions of a Fusion Power Reactor. The general strategy consists in integrating studies on macroscopic phenomena (gas release) with the characterisation of corresponding microscopic diffusion phenomena (bubble kinetics) and the assessment of some fundamental diffusion parameter for the models (gas atomic diffusion coefficients). The present work gives a summary of the latest achievements in this context. By an inverse analysis of experimental out-of-pile gas release from weakly irradiated pebbles, coupled to the study of the characteristics of bubble population, it has been possible to assess the thermal diffusion coefficients of helium and tritium in and to improve and validate the classical model of gas precipitation into bubbles inside the grain. The improvement of the description of gas atomic diffusion and precipitation is the first step to enable a more reliable prediction of gas release.