Description
The new ITER baseline considers changing the first wall material from beryllium (Be) to tungsten (W), and the inclusion of a boronization system to mitigate plasma operation risks related to the use of W [97S4QU]. Boronization applied through a glow discharge (GD) is shown to reduce the intrinsic oxygen contamination in full W tokamaks, easing plasma start-up, extending gas prefill ranges and reducing the edge radiation levels.The frequent introduction of boron (B) in ITER has the potential to trap significant amounts of tritium (T) in-vessel. As in the case of Be, eroded B is expected to migrate in the plasma edge and accumulate with hydrogen isotopes, including T, and impurity species into thick co-deposited layers at the divertor and in recessed main chamber wall areas. The T retention in B deposits is capped by the quantity of B introduced, estimated at 321.6g T in DT-1 if no specific action to recover T, within or between campaigns is taken. This emphasizes the importance of tritium removal techniques to maintain a minimal level of retained tritium within the layers. A combination of removal techniques is proposed, similar to those applied to remove T from Be co-deposits in JET following the DTE2 experiment [4]: baking, glow discharge conditioning (GDC), ion cyclotron wall conditioning (ICWC) as well as tokamak plasmas.
Methods to recover T rely on thermal release, isotopic exchange, or a combination of both. Isotopic exchange occurs through plasma exposure of surfaces, and in the specific case of B layers, the depth of effective interaction is thought to depend strongly on the energy of the plasma particles impacting the surface. To assess the removal capabilities as a function of particle flux conditions and boron layer thickness, and to validate the tritium removal strategy of ITER, experimental input is required.
The purpose of this expertise is to provide the ITER Organization with a laboratory assessment of fuel removal capabilities from pre-characterized boron layers, limited to the techniques of GDC and ICWC, combined with detailed post-exposure analysis. The targeted plasma facilities is:
• The toroidal TOMAS device equipped with ICWC and GDC systems, for which the exposure parameters (particle flux and energy distribution) of material surfaces can be characterized and optimized to ITER conditions.
The anticipated results will inform IO on the efficiency of T removal from boron layers by ICWC and GDC, thereby validating their role in ITER's tritium removal strategy.
| Période | 1 févr. 2025 → 31 déc. 2026 |
|---|---|
| Travaille pour | FORSCHUNGSZENTRUM JULICH GMBH, Allemagne |
| Degré de reconnaissance | International |