Key results and future directions of the JET fusion research programme

Transcription

1 Key results and future directions of the JET fusion research programme Karl Krieger on behalf of the JET contributors This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme under grant agreement No The views and opinions expressed herein do not necessarily reflect those of the European Commission.

2 Acknowledgements K. Krieger 1,2, S. Brezinsek 3, T. Donné 4,5, L. Horton 6,7, X. Litaudon 1,8, I. Nunes 1,7 and JET contributors* 1 EUROfusion Consortium, JET, Culham Science Centre, Abingdon, OX14 3DB, UK. 2 Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, Garching, Germany. 3 Forschungszentrum Jülich, Institut für Energie- und Klimaforschung - Plasmaphysik, Jülich, Germany. 4 EUROfusion Programme Management Unit, Culham Science Centre, Abingdon, OX14 3DB, UK. 5 FOM Institute DIFFER P.O. Box 1207 NL-3430 BE Nieuwegein, The Netherlands. 6 JET Exploitation Unit, Culham Science Centre, OX14 3DB Abingdon, UK. 7 European Commission, B-1049 Brussels, Belgium. 8 CEA, IRFM, Saint Paul Lez Durance, France. 9 Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, Portugal. * See the Appendix of F. Romanelli et al., Proceedings of the 25th IAEA Fusion Energy Conference 2014, Saint Petersburg, Russia. K. Krieger NENE 2015 Portorož Page 2

3 Outline Introduction - JET in the context of nuclear fusion research. Key results of the JET ITER-like Wall (ILW) experiments Plasma-wall interactions T inventory management / 1st wall life time. Integrated high performance plasma operation scenario Confinement properties / heat exhaust / control of transients. Future JET research programme Key issues defining programme of D/T campaign. D/H/He experiments in preparation of D/T campaign. K. Krieger NENE 2015 Portorož Page 3

4 The European fusion road map Scientific foundations to build device with burning plasma Demonstrate technical feasibility Demonstrate industrial feasibility Commercial exploitation ITER DEMO JT-60SA THE AGE OF NUCLEAR FUSION K. Krieger NENE 2015 Portorož Page 4

5 Background - tokamak Tokamak: a device to confine a toroidal plasma column using combined magnetic fields from external coils and an induced plasma current. Recipe for success: divert outer field lines guiding plasma to dedicated target plates*. Cleaner plasmas by keeping surface interactions far away. Heating: injection of neutral fuel atoms (NBI), coupling EM waves to electrons or ions, α-particles. Power exhaust: - Radiation to entire 1st wall - Heat transfer to divertor plates Divertor heat load concentrated to very narrow region 1 cm! [*ASDEX, IPP Garching (1981)] K. Krieger NENE 2015 Portorož Page 5

6 Background - H-mode H-mode*: a plasma mode with a transport barrier at the plasma edge leading to improved ( 2) confinement. Emerges from a phase transition at a certain input power threshold. 1 Plasma Temperature L mode 0 1 Normalised radius r/a CPS b [*F. Wagner et al., ASDEX, IPP Garching (1982)] K. Krieger NENE 2015 Portorož Page 6

7 Background - ELMs Edge Localised Mode: a quasi-periodic instability driven by excessive free energy available in H-mode's steep edge pressure gradient. Plasma bursts are expulsed into edge and travel to target plates. ELMs can cause serious material damage! JET #84778 Plasma pressure before ELM Plasma energy Particle flux at target after ELM K. Krieger NENE 2015 Portorož Page 7

8 ITER mission objectives Physics: ITER is designed to produce a plasma dominated by a-particle heating. Demonstrate significant fusion power amplification factor Q = P fus / P in 10 in long-pulse operation ( s). Demonstrate steady-state operation at Q = 5 for 300s. Optional: controlled ignition at Q 30? 50 MW 500 MW Technology : Test components required for a power plant. Test concepts for a tritium breeding blanket. Demonstrate integrated operation of key systems for a power plant. [D. Campbell, The ITER research plan] K. Krieger NENE 2015 Portorož Page 8

9 Where can JET contribute to ITER? The challenge: sustaining high gain burning plasma by integrating core performance with exhaust constraints. Current ITER top operational risks: Disruption loads and means for disruption mitigation. Uncertainty in H-mode power threshold scaling. Access to H-mode during non-active phase. H-mode periodic edge mode (ELM) mitigation. Vertical stability control. Maintaining in-vessel T inventory within safety margin. Divertor performance with tungsten PFCs. K. Krieger NENE 2015 Portorož Page 9

10 JET ILW principal objectives in support of ITER Exploit JET s unique capability to handle tritium and beryllium Emulate ITER first wall configuration by replacing JET C with Be/W (ILW): Demonstrate low fuel retention plus high fuel recovery fraction in a Be/W environment (benefit of replacing C). Demonstrate compatibility of high confinement plasmas with Be/W wall (more demanding without C). K. Krieger NENE 2015 Portorož Page 10

11 Outline Introduction - JET in the context of nuclear fusion research. Key results of the JET ITER-like Wall (ILW) experiments Plasma-wall interactions T inventory management / 1st wall life time. Integrated high performance plasma operation scenario Confinement properties / heat exhaust / control of transients. Future JET research programme Key issues defining programme of D/T campaign. D/H/He experiments in preparation of D/T campaign. K. Krieger NENE 2015 Portorož Page 11

12 Nuclear safety in-vessel retention of tritium Tritium is retained in vessel due to Implantation ( limited). Wall material (C, Be) migration and co-deposition (unlimited). CX atoms CX atoms ITER prediction safety limit [S. Brezinsek et al., Nucl. Fus. (2013)] K. Krieger NENE 2015 Portorož Page 12

13 Confirmation of low D(T) retention for Be-W Fuel retention mainly via co-deposits at inner baffle. 10 D retention smaller and Be migration 4-8 smaller as that of C in JET-C. Tile analysis lower rates / higher reduction due to long-term outgassing [K. Heinola et al., PFMC (2015)]. [S. Brezinsek et al., Nucl. Fus. (2013), M. Mayer et al., PFMC (2015)] K. Krieger NENE 2015 Portorož Page 13

14 Validate predictive codes modelling T retention Be-migration and resulting D- implantation as well as D-implantation were modelled using the WALLDYN code. Both local distribution and integral value of T-retention can be reproduced. JET ILW data validate code predictions of acceptable T-retention rate in ITER. [K. Schmid et al., IAEA FEC (2014)] K. Krieger NENE 2015 Portorož Page 14

15 Effects of power transients on W-divertor What happens if tungsten divertor is exposed to heat loads exceeding power handling limits? Quasi-stationary case: bulk melting, to be avoided by design. Transient case by plasma instabilities: singular (VDE, disruption) repetitive (ELMs). flash melting of thin surface layer. K. Krieger NENE 2015 Portorož Page 15

16 Effects of W flash melting in the JET divertor Lamellae set with leading edge exposed to q was installed at bulk-w target. Mock-up in ITER avoided by design! K. Krieger NENE 2015 Portorož Page 16

17 Effects of power transients on W-divertor ELMs in H-mode ΔW ELM = 300 kj q II =0.5-1GWm -2 Base temperature 2800 C ΔT by ELMs K Controlled flash melting by ELM induced temperature excursions. [B. Bazylev et al., 21st PSI 2014, J. Coenen et al., 21st PSI 2014] K. Krieger NENE 2015 Portorož Page 17

18 Effects of power transients on W-divertor [J. Coenen et al., 21st PSI 2014] Melt damage progression observed after each of 8 subsequent discharges by In-Vessel Inspection System. K. Krieger NENE 2015 Portorož Page 18

19 Effects of power transients on W-divertor MEMOS code computes thermal response and melt layer motion due to plasma pressure, T, j B forces,... Measurements well reproduced. JET ILW data provide base for improved extrapolation to ITER. [B. Bazylev et al., 21st PSI 2014] K. Krieger NENE 2015 Portorož Page 19

20 Outline Introduction - JET in the context of nuclear fusion research. Key results of the JET ITER-like Wall (ILW) experiments Plasma-wall interactions T inventory management / 1st wall life time. Integrated high performance plasma operation scenario Confinement properties / heat exhaust / control of transients. Future JET research programme Key issues defining programme of D/T campaign. D/H/He experiments in preparation of D/T campaign. K. Krieger NENE 2015 Portorož Page 20

21 Energy confinement properties in JET ILW Sufficient reaction cross-section high T ( 10 kev). Also: sufficient reaction rate high n ( m -3 ). ( ) 3 2 i + e = τ E heat W = n T T dv P Scaling law from many devices "IBP98(y,2)" ITER design τ E JET-C JET-ILW heat Ip n P Be/W 1st wall sub par confinement at high I p What happened? [I. Nunes et al., IAEA FEC, (2014)] K. Krieger NENE 2015 Portorož Page 21

22 Reasons why JET ILW energy confinement suffers Lower "pedestal" T e extends to core ( ) due to "profile stiffness". Reduced fuel recycling Reduced pumping* Increased D 2 fuelling rate* Central radiation losses (W) further decrease of core T e W divertor source W core accumulation * consequences of controlling divertor heat load and W source. [M. Beurskens et al., PPCF (2013)] K. Krieger NENE 2015 Portorož Page 22

23 Plasma behaviour in JET-ILW tungsten W transport dominated by neoclassical transport W accumulation in core plasma Strong cooling of core plasma Plasma radiative collapse What can we do to prevent this? Inhibit W accumulation Increase W flushing (f ELM ) Reduce W source (T e,div, W ELM ) [C. Angioni et al., NF (2014) & PoP (2015), F. Cassonet al., PPCF (2015), M. Valisa et al., IAEA FEC (2014)] K. Krieger NENE 2015 Portorož Page 23

24 Plasma behaviour in JET-ILW tungsten Problem: neutral beam heating leads to temperature profile promoting central W accumulation. Solution: additional RF heating of plasma centre. ASDEX Upgrade [R. Neu et al., J. of Nucl. Mat. (2003)] In JET-ILW central W is successfully controlled by ICRF heating [C. Angioni et al., NF (2014) & PoP (2015), F. Casson et al., PPCF (2015), C. Giroud et al., PPCF (2015))] K. Krieger NENE 2015 Portorož Page 24

25 Tailoring ELMs for W flushing & ELM mitigation ELM plasma bursts help to remove tungsten high / low fuelling Problem: how to increase their frequency? Solution: inject more fuel (D 2 ). ~1MJ But: degrades confinement (by 15%). Other means: kicks, pellets. Best: perturbation coils but not available in JET (yet). [I. Chapman et al., EPS (2015)] K. Krieger NENE 2015 Portorož Page 25

26 Plasma behaviour in JET-ILW beryllium I (CIII at 97.7nm) / n e [arb. units] GB divertor JET with CFC HD divertor deuterium plasma x20 Helium HD JET pulse number JET with W HD divertor and Be first wall deuterium plasma C28-C No chemical erosion of Be. Be getters O. Lower impurity levels in JET-ILW. Increase of fusion performance. No more intrinsic carbon edge radiation cooling. Hotter divertor plasma alternative means of cooling reqd C31 [S. Brezinsek et al., Nucl. Fus. 55 (2015) ] K. Krieger NENE 2015 Portorož Page 26

27 Reducing divertor heat loads Seeded N 2 replaces C Reduction of divertor heat load by power spreading Additional spreading by sweeping of strike-points. Caveat: adversely affects pumping. [L. Aho-Mantila et al., IAEA FEC (2014) A. Järvinen et al., IAEA FEC (2014) M. Wischmeier et al., IAEA FEC (2014)] JET-ILW achieved maximum f RAD = P rad / P in 75% with N 2. However: DEMO will need 90%! How to spread from X-point to radiating mantle? Simultaneous optimisation of f RAD and τ E modelling is key! K. Krieger NENE 2015 Portorož Page 27

28 Outline Introduction - JET in the context of nuclear fusion research. Key results of the JET ITER-like Wall (ILW) experiments Plasma-wall interactions T inventory management / 1st wall life time. Integrated high performance plasma operation scenario Confinement properties / heat exhaust / control of transients. Future JET research programme Key issues defining programme of D/T campaign. D/H/He experiments in preparation of D/T campaign. K. Krieger NENE 2015 Portorož Page 28

29 Present schedule for future JET programme Before the planned D/T campaign several campaigns in pure H, D, T and He are envisaged to optimise high performance scenarios and to study isotope effects. After the D/T campaign JET will be brought into a safe state before decommissioning. [T. Donné, JET Science Meeting, Culham, ] K. Krieger NENE 2015 Portorož Page 29

30 JET key deliverables of DT campaign Demonstrate ITER base line operation scenarios (inductive & hybrid) for the ITER wall configuration. Demonstrate dust and T-recovery from the ITER-like Wall. Understand the influence of isotope mass on H-mode access, confinement, ELMs,... Understand α-particle confinement and instabilities, validate respective code models. Demonstrate 2 nd T heating scheme with ICRF. Test DT relevant diagnostics and neutron rate calibration techniques. [ The scientific case for DT operation in JET, Final Report, (2011)] K. Krieger NENE 2015 Portorož Page 30

31 JET preparation of the DT-campaign Development of ILW-compatible ITER scenario (inductive H-Mode and Hybrid). Tool set: 34 MW NBI & 12 MW ICRH (incl. ITER-Like Antenna). DT-2 challenge: MW target P fus (D/T). Building blocks: 1) High performance scenario with H > 0.9, W th MJ for > 5 s. 2) ITER relevant divertor power handling P sep / R 10MW/m. Heat flux control: Learn to replace N 2 by Ne/Ar seeding. Improve ELM pacing by pellets. Transport / stability control: Improve core-w control actuators. Learn to maintain MHD stability. 3) ITER relevant disruption mitigation efficiency Predictors and mitigation: Diagnostics improvements: Optimise DMV operation. T accounting ("AGHS T-rehearsal"). Neutron detectors calibration. Fast particle detection. K. Krieger NENE 2015 Portorož Page 31

32 Fusion performance - bridging the τ E gap Baseline Hybrid Objective [I. Nunes et al., IAEA FEC, (2014)] [C. Challis et al., Nucl. Fusion 55 (2015)] Good news: recent experiments show JET ILW confinement scales better with P in than IPB98(y,2). Opens path to recover H 1 at higher plasma currents (β N 2). Challenge: maintain acceptable target plate power / energy load. K. Krieger NENE 2015 Portorož Page 32

33 Summary JET ILW experiments have confirmed acceptable tritium retention for the ITER wall materials configuration. W flash melting studies validated MEMOS code now used for optimisation of ITER bulk-w divertor. Origin of sub-optimal energy confinement at low beta operation not yet fully understood. Recovery expected for next campaign because of higher available heating power. JET pre-dt campaigns aim at providing integrated H-mode scenarios combining required confinement and exhaust. JET -DT campaign will for the first time demonstrate max fusion performance under controlled PWI in a true fusion plasma as the final step precluding ITER operation. K. Krieger NENE 2015 Portorož Page 33