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Prediction of long-term tritium retention in the divertor of ITER: influence of modelling assumptions on retention rates

2009; IOP Publishing; Volume: T138; Linguagem: Inglês

10.1088/0031-8949/2009/t138/014011

1402-4896

A. Kirschner, K. Ohya, D. Borodin, Rui Ding, D. Matveev, V. Philipps, U. Samm,

Nuclear Materials and Properties

ERO modelling of long-term tritium (T) retention has been done for the divertor of ITER with graphite target plates assuming a certain beryllium influx into the divertor, eroded from the main chamber.The divertor beryllium (Be) influx relative to the deuterium ion flux has been fixed at 0.1% for the outer divertor and 1.0% for the inner divertor.In addition to the original B2-Eirene plasma background, the influence of variations of temperature and density in the divertor has been studied.Moreover, assumptions for enhanced erosion of redeposited carbon and effective sticking for hydrocarbons have been analysed.With graphite target plates, long-term tritium retention is dominated by T co-deposition in deposits.Within the studied parameter range, the modelling yields 200-500 possible ITER discharges without cleaning before reaching the safety limit of 700 g of in-vessel retained tritium.Surface temperature-dependent tritium amounts in carbon and beryllium deposits have been taken into account.