TY - JOUR
T1 - Prediction of long-term tritium retention in the divertor of ITER
T2 - 12th International Workshop on Plasma-Facing Materials and Components for Fusion Applications, PFMC-12
AU - Kirschner, A.
AU - Ohya, K.
AU - Borodin, D.
AU - Ding, R.
AU - Matveev, D.
AU - Philipps, V.
AU - Samm, U.
PY - 2009
Y1 - 2009
N2 - 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.
AB - 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.
UR - https://www.scopus.com/pages/publications/77953887156
U2 - 10.1088/0031-8949/2009/T138/014011
DO - 10.1088/0031-8949/2009/T138/014011
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AN - SCOPUS:77953887156
SN - 0281-1847
VL - T138
JO - Physica Scripta Topical Issues
JF - Physica Scripta Topical Issues
M1 - 14011
Y2 - 11 May 2009 through 14 May 2009
ER -