The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

Assessment of tungsten for use in the ITER plasma facing components

As the most promising plasma-facing material for actual and future nuclear fusion devices, tungsten has to face and withstand a broad variety of severe operational conditions. These comprise high transient thermal loads, neutron irradiation-induced material degradation and transmutation, hydrogen and helium attack at the plasma-facing surface, and thermal fatigue under steady state heat fluxes as part of a plasma-facing component. The characterization and understanding of the material’s behavior under these conditions is essential for finding a grade of tungsten or tungsten alloy compatible with such an extreme environment.

4.17 – Tungsten as a Plasma-Facing Material

Critical heat flux analysis and R&D for the design of the ITER divertor

The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux tests for R&D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m 2 in the divertor targets and around 0.5 MW/m 2 at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components.

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

Non-destructive testing of divertor components

The main European contribution to the ITER divertor project was the development of the divertor target with severe operating requirements such as peak heat loads of up to 20 MW/m2. This development involving EU laboratories and industry included R&D on armour materials, thermo-hydraulics testing, component manufacture, high heat flux testing, design and manufacture of prototypes for later testing. The 4-year EU R&D effort achieved the demonstration of the feasibility of a robust divertor target design based on carbon and tungsten armour. This EU solution has eventually been adopted for the ITER reference design and could be valid also for other ITER high heat flux components such as limiters or baffles.

European development of the ITER divertor target

Manufacturing and testing of a prototypical divertor vertical target for ITER

High heat flux tests on CuCrZr actively water cooled elements were performed with geometric and thermal hydraulic parameters relevant to ITER (International Thermonuclear Experimental Reactor) divertor conditions. Different types of mock-ups with the same width were tested and compared: double smooth tubes (SM2), swirl tubes (ST2; ST4), annular flow tubes (AF1; AF3) and hypervapotron tubes (HV1; HV3). Analyses of tests were done using the CEA method [1;2] first developed by Sandia Laboratory [ 3 ]. Finite Element calculations were used with a set of correlations in order to express the wall heat flux as a fonction of wall temperature in the convective regime as well as in the subcooled boiling regime (this set is now available in the EUPITER code). Maximum wall heat flux was compared with modified TONG-75 correlation. In terms of ICHF (Incident Critical Heat Flux) and for the same thermal hydraulic conditions, results gave this decreasing order: HV1, HV3, ST2, AF1, ST4, AF3, SM2. Versus lineic pumping power, the previous order was slightly changed: HV1, HV3, ST2, ST4, AF1, SM2, AF3. A typical HV1 result is a 38 MW/m 2 ICHF for 135°C local subcooling, 10 m/s water velocity, 3.5 MPa local pressure, 0.3 MPa/m lineic pressure drop and 380 W/m lineic pumping power.

Comparison between various thermal hydraulic tube concepts for the ITER divertor

This task within the EU R&D for ITER is primarily aimed at the development of solutions for the divertor target which has to be designed for up to 1000 off-normal transients at 20 MW/m2. Representative small scale mock-ups with carbon and tungsten armour have been manufactured using a similar technology. First tests support the analysis which indicates the best high heat flux capability for carbon mono-blocks.

I. Smid

Papers

Non-destructive testing of divertor components

European development of the ITER divertor target

Manufacturing and testing of a prototypical divertor vertical target for ITER

Comparison between various thermal hydraulic tube concepts for the ITER divertor

Manufacture and Testing of Small-Scale Moch-ups for the ITER Divertor