We analyzed beryllium first wall sputtering erosion, sputtered material transport, and T/Be codeposition for a typical next-generation tokamak design - the fusion ignition research experiment (FIRE). The results should be broadly applicable to any future tokamak with a beryllium first wall. Starting with a fluid code scrapeoff layer attached plasma solution, plasma D0 neutral fluxes to the wall and divertor are obtained from the DEGAS2 neutral transport code. The D+ ion flux to the wall is computed using both a diffusive term and a simple convective transport model. Sputtering coefficients for the beryllium wall are given by the VFTRIM-3D binary-collision code. Transport of beryllium to the divertor, plasma, and back to the wall is calculated with the WBC+ code, which tracks sputtered atom ionization and subsequent ion transport along the SOL magnetic field lines. Then, using results from a study of Be/W mixing/sputtering on the divertor, and using REDEP/WBC impurity transport code results, we estimate the divertor surface response. Finally, we compute tritium codeposition rates in Be growth regions on the wall and divertor for D-T plasma shots using surface temperature dependent D-T/Be rates and with different assumed oxygen contents. Key results are: (1) peak wall net erosion rates vary from about 0.3 nm s-1 for diffusion-only transport to 3 nm s-1 for diffusion plus convection, (2) T/Be codeposition rates vary from about 0.1 to 10.0 mg T s-1 depending on the model, and (3) core plasma contamination from wall-sputtered beryllium is low in all cases (< 0.02%). Thus, based on the erosion and codeposition results, the performance of a beryllium first wall is very dependent on the plasma response, and varies from acceptable to unacceptable.
- Tritium codeposition
ASJC Scopus subject areas
- Civil and Structural Engineering
- Nuclear Energy and Engineering
- Materials Science(all)
- Mechanical Engineering