Divertor heat load in ITER-like advanced tokamak scenarios on JET

1. Divertor heat load in ITER-like advanced tokamak scenarios on JET G.Arnoux1,(3), P.Andrew1, M.Beurskens1, S.Brezinsek2, C.D.Challis1, P.DeVries1, W.Fundamenski1, E.Gauthier3, C.Giroud1, A.Huber2, S.Jachmich4, X.Litaudon3, R.A.Pitts5, F.Rimini3 and JET-EFDA collaborators* 1EURATOM/UKAEA Fusion Association, Culham Science Center, Abingdon, Oxon, OX14 3DB, UK 2Institüt für Energieforschung – Plasmaphysik, Forschungzentrum Jülich, Trilateral Euregio Cluster, EURATOM-Assoziation, D-52425 Jülich, Germany 3Association EURATOM-CEA, INRFM, CEA/Cadarache, F-13108 St Paul-Lez-Durance, France 4Association “EURATOM-Belgian State” Laboratory for Plasma Physics Koninklijke Militaire scholl – Ecole Royale Militaire Renaissancelaan 20 Avenue de la renaissance, B-1000 Brussels Belgium 5Association EURATOM-Confédération Suisse, Ecole Polytechnique Fédérale de Lausanne (EPFL), CRPP, CH-1015 Lausanne, Switzerland *See Appendix in M.L.Watkins et al., 21st IAEA Fusion Energy Conference, 2006, Chengdu, China

2. Introduction Pin Prad Pcond Inner target Outer target High d configuration • Power balance • Pin=Pcond+Prad • Pin=Pdiv+gPrad • Pdiv=Pcond+(1-g)Prad • Heat load on divertor targets • Pdiv=Pin-gPrad = Pin(1-gfrad) • Prad∝ne2Zeff • Density: ne • Impurities: Zeff

3. Heat load and the ITER-like wall • Steady-state heat loads onto the divertor must be controlled • Wetted area (target geometry and scrape-off layer transport) • Radiative fraction, frad: highdensity + extrinsic impurity seeding • In ITER: steady-state Pdiv,pk≤10MW/m2 • Optimised divertor geometry (wetted area~4m2) • Partially detached plasma (high ne) • frad=75% probably using argon • Pdiv must be assessed for the JET ITER-like wall (ILW) • Divertor tiles in CFC coated with tungsten: Tsurf<1600oC • Tile 5 in tungsten: q⊥,pk<7.5MW/m2 for 10s (higher q⊥,pk for shorter pulse) • Main chamber walls: beryllium (main radiator) • Scenarios in high triangularity configuration compatible with ILW • Baseline scenario (BL): inductive current • Advanced tokamak scenario (AT): steady-state scenario • Up to now, edge of plasmas in AT scenarios poorly documented

4. Scenarios in high d configurations Density Te at target Current Non-inductive maximised 20<Te<40eV 4<ne,av<5∙1019 m-3 0.6<ne,av/ne,gr<0.8 Te≃10eV ne,av≃10∙1019 m-3 ne,av/ne,gr≃1 Inductive Pin>16MW AT BL

5. Divertor heat load measurement SOL SOL PFR PFR SOL PFR SOL • Space and time resolution • Ds=16mm (Dr≃4mm) • Dt=20ms (60ms exposure) • Time averaged profiles • Pin≃Cste • frad≃Cste (inter-ELM) • Strike point position • THEODOR => heat load • Correction factor from thermocouple • Inner target • 1.3<EIR/ETC<2.0 • Outer target • 0.5<EIR/ETC<1.0

6. Heat load reduction on targets 18MW≤Pin≤24MW Neon seeding

7. Radiation

8. Peak heat load reduction => Tsurf=1600oC after… 50s On inner target: 3MW/m2 => shorter pulses than 10s (Pin=24MW) Outer target: 10MW/m2 No clear dependence on injection location. Better cooling with N2? At frad=50% same heat load for AT (seeded) and BL (higher density), but…

9. Plasma contamination? ne,av fgw Zeff • Low density and high Prad => Higher Zeff (∝Prad/ne2) • Higher W sputtering in ILW => Impurity accumulation in the core (ITB)? • Increasing density (fgw>0.8?) => same heat load reduction but lower Zeff. (Zagorski et al.)

10. Summary and conclusion • First measurements of divertor heat load in AT scenario on JET (no ITB) • AT scenario will be an issue in ILW for long pulse operation at Pin=45MW (outer target) • Heat load reduced at the level of baseline scenario with impurity seeding (same Prad) but high Zeff • Will tungsten sputtering be a problem for core accumulation (ITB)? • What is the net gain for divertor heat loads when increasing density and Pin?