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A substantial reduction of net compared to gross erosion of molybdenum and tungsten was observed in experiments conducted in the lower divertor of DIII-D using the divertor material evaluation system. Post-exposure net erosion of molybdenum and tungsten films was measured by Rutherford backscattering (RBS) yielding net erosion rates of 0.4–0.7 nm s−1 for Mo and ~0.14 nm s−1 for W. Gross erosion was estimated using RBS on a 1 mm diameter sample, where re-deposition is negligible. Net erosion on a 1 cm diameter sample was reduced compared to gross erosion by factors of ~2 for Mo and ~3 for W. The experiment was modeled with the REDEP/WBC erosion/re-deposition code package coupled to the Ion Transport in Materials and Compounds—DYNamics mixed-material code, with plasma conditions supplied by the Onion skin modeling + Eirene + Divimp for edGE modeling code with input from divertor Langmuir probes. The code-calculated net/gross erosion rate ratios of 0.46 for Mo and 0.33 for W are in agreement with the experiment.

A series of quiescent L-mode discharges have been used with varying degrees of divertor attachment to study carbon erosion in the DIII-D tokamak divertor. Spectra of atomic and molecular carbon plasma emissions are measured across the divertor. Predictions of carbon emission resulting from chemical and physical sputtering at the outer strikepoint are provided by the WBC transport code. For attached ionizing plasmas (Te ~ 20 eV and carbon surface T ~ 350 K) the CD and C2 carbon radical emissions are consistent with a total chemical sputtering yield, Ychem ~ 0.3%. During sweeps of the inner divertor leg, CD and C2 emissions indicate the main-wall tiles has a six times higher Ychem than the divertor tiles despite identical incident plasma conditions. Emission of atomic (CI) and singly ionized carbon are dominated by physical sputtering with a measured yield Yphys ~ 2% as expected from laboratory data, indicating that chemical sputtering plays a minor role as a divertor carbon source. The Doppler broadening of the CI emission agrees with the physical sputtering model, but the predicted Doppler shift is a factor three larger than the experiment. CD and C2 emission are reduced below detection during divertor detachment (Te ~ 1–2 eV), an indication of the suppression of chemical erosion with Ychem < 10-4 based on expected photon intensity from WBC.

Erosion of the divertor and first-wall plasma-facing components, tritium uptake in the re-deposited films, and direct implantation in the armour material surfaces surrounding the plasma, represent crucial physical issues that affect the design of future fusion devices. In this paper we present the derivation, and discuss the results, of current predictions of tritium inventory in ITER-FEAT due to co-deposition and implantation and their attendant uncertainties. The current armour materials proposed for ITER-FEAT are beryllium on the first-wall, carbon-fibre-composites on the divertor plate near the separatrix strike points, to withstand the high thermal loads expected during off-normal events, e.g., disruptions, and tungsten elsewhere in the divertor. Tritium co-deposition with chemically eroded carbon in the divertor, and possibly with some Be eroded from the first-wall, is expected to represent the dominant mechanism of in-vessel tritium retention in ITER-FEAT. This demands efficient in-situ methods of mitigation and retrieval to avoid frequent outages due to the reaching of precautionary operating limits set by safety considerations (e.g., ~350 g of in-vessel co-deposited tritium) and for fuel economy reasons. Priority areas where further R&D work is required to narrow the remaining uncertainties are also briefly discussed.