1. Alberto Loarte 7 th ITPA Divertor Meeting – Toronto 6/9 – 11 – 2006 1 ITER Issue Card FW-3. Modification of Upper Be-blanket modules, material and/or PFC geometry for various Be-blanket modules Alberto Loarte EFDA CSU Garching

2. Alberto Loarte 7 th ITPA Divertor Meeting – Toronto 6/9 – 11 – 2006 2 Outline of the ITER Design/Assumptions Blanket modules in ITER covered by Be panels have straight shape in toroidal and poloidal direction Typical toroidal length = 1.2 – 2 m Typical poloidal length = 0.85 – 1.2 m Poloidal/toroidal gaps = 20 mm Installation tolerance = ± 3 mm (?) Castellation = 20 x 20 mm x 0.7 mm (?) Plasma thermal loads are assumed to be deposited uniformly on module face (radiation)

3. Alberto Loarte 7 th ITPA Divertor Meeting – Toronto 6/9 – 11 – 2006 3 Issue We know now that :  Plasma flux will reach the main wall in ITER (> 10 23 s -1 T ~ 10-30 eV) along B  Power fluxes during ELMs IIB at upper X-point can reach up to 10 GWm -2 (for large ELMs)  Power fluxes during disruptions IIB at upper X-point can reach up to 40 GWm -2 (for large disruptions)  If distance between separatrices is reduced (ref ~5 cm) significant power fluxes can reach upper Be modules Survivability of Be wall to plasma flow is directly correlated to avoidance of edges and to effective use of large wall area for power deposition Given decay of fluxes with distance to separatrix & dominance of outer midplane outfluxes (not for disruptions ?) most exposed modules are upper Be modules and main chamber modules 16 & 17

4. Alberto Loarte 7 th ITPA Divertor Meeting – Toronto 6/9 – 11 – 2006 4 Action  evaluate plasma particle and power fluxes on Be-blanket modules with present design (incl. castellations)/installation tolerances  evaluate consequences for Be erosion/melting of present design  investigate possibility of simple design modifications to account for plasma impact in most exposed modules and/or in all modules  determine “safe” plasma operation range  If “safe” range considered insufficient evaluate possibility of using other PFM in most exposed modules and/or a secondary upper divertor Benefits to ITER : larger device flexibility and resilience to transient power fluxes Implications on cost and schedule : depend on final solution adopted Risks : Low if correct shaping is achieved and sufficient for Be, if W is necessary  n W in plasma Hiding toroidal edges in outer upper X-point modules (± 3 mm) for  80%