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This article investigates the triggering of ELMs on JET by injection of frozen pellets of isotopes of Hydrogen. A method is established to determine the probability that a specific pellet triggers a particular ELM. This method allows clear distinction between pellet-ELM pairs that are very likely to represent triggering events and pairs that are very unlikely to represent such an event. Based on this, the pellet parameters that are most likely to affect the ability of pellets to trigger ELMs have been investigated. It has been found that the injection location is very important, with injection from the vertical high field side showing a much higher triggering efficiency than low field side (LFS) injection. The dependence on parameters such as pellet speed and size and the time since the last ELM is also seen to be much stronger for LFS injection. Finally, the paper illustrates how improvements to the pellet injection system by streamlining the pellet flight lines and slightly increasing the pellet size has resulted in a significantly improved ability to deliver pellets to the plasma and trigger ELMs.

Joint European Torus (JET) plasma initiations that form a significant quantity of runaway electrons have been studied. It is shown that there is no direct relationship between the prefill pressure and breakdown electric field and signs of runaway electrons during the plasma initiation. Runaway electron generation is determined by the electric field and density development at and after burn-through. A clear criterion of density and electric field at one given point in time, which would ensure the avoidance of runaway electron generation, cannot be determined, because the timescales for the formation of runaway electrons and for the dynamics of the density differ significantly. Moreover, the formation process can be reversed, reducing the influence of runaway electrons on the discharge. Ensuring a high enough density will reduce the likelihood that runaway electron discharges are formed. It is also found that at JET the electric field often exceeds the critical electric field during the early stages of the current ramp-up phase, even when no signs of runaway electrons are present. Expected runaway current dynamics have been analysed using the discharge circuit equation. The comparison of the expected runaway electron current dynamics shows it to be significantly slower compared to theoretical expectations in the presence of a hot and dense thermal background plasma. This could be explained by an enhanced critical electric field and/or a reduced confinement of runaway electrons. The latter is shown to be affected by bursts of magnetohydrodynamic activity that are characteristic during the current ramp-up. The development of discharges in which the current is fully carried by runaway electrons happens on a slow timescale of several seconds, limited by the available flux. Such timescales are sufficient for improved active control of these events, avoiding runaway currents at plasma initiation exceeding values at which they could damage in-vessel components. The results provide insight into the improvement and interpretation of self-consistent modelling of runaway electron generation during the start-up of International Thermonuclear Experimental Reactor discharges.

We report results of benchmarking of core particle transport simulations by a collection of codes widely used in transport modelling of tokamak plasmas. Our analysis includes formulation of transport equations, difference between electron and ion solvers, comparison of modules of the pellet and edge gas fuelling on the ITER baseline scenario. During the first phase of benchmarking we address the particle transport effects in the stationary phase. Firstly, simulations are performed with identical sources, sinks, transport coefficients, and boundary conditions prescribed in the flattop H-mode phase. The transformation of ion particle transport equations is introduced so to directly compare their results to electron transport solvers. Secondly, the pellet fuelling models are benchmarked in various conditions to evaluate the dependency of the pellet deposition on the pellet volume, injection side, pedestal, and separatrix parameters. Thirdly, edge gas fuelling is benchmarked to assess sensitivities of source profile predictions to uncertainties in plasma conditions and detailed model assumptions. At the second phase, we address particle transport effects in the time- evolving plasma including the current ramp-up to the ramp-down phase. The ion and the electron solvers are benchmarked together. Differences between the simulation results of the solvers are investigated in terms of equilibrium, grid resolution, radial coordinate, radial grid distribution, and plasma volume evolution term. We found that the selection of the radial coordinate can yield prominent differences between the solvers mainly due to differences in the edge grid distribution. The simulations reveal that electron and ion solvers predict noticeably different density peaking for the same diffusion and pinch velocity while with the peaked profile of helium, expected in fusion reactors. The fuelling benchmarking shows that gas puffing is not efficient for core fuelling in H-modes and density control should be done by the high field side pellet injection in contrast to present machines.

Although gas breakdown phenomena have been intensively studied over 100 years, the breakdown mechanism in a strongly magnetized system, such as tokamak, has been still obscured due to complex electromagnetic topologies. There has been a widespread misconception that the conventional breakdown model of the unmagnetized system can be directly applied to the strongly magnetized system. However, we found clear evidence that existing theories cannot explain the experimental results. Here, we demonstrate the underlying mechanism of gas breakdown in tokamaks, a turbulent ExB mixing avalanche, which systematically considers multi-dimensional plasma dynamics in the complex electromagnetic topology. This mechanism clearly elucidates the experiments by identifying crucial roles of self-electric fields produced by space-charge that decrease the plasma density growth rate and cause a dominant transport via ExB drifts. A comprehensive understanding of plasma dynamics in complex electromagnetic topology provides general design strategy for robust breakdown scenarios in a tokamak fusion reactor., Gas breakdown mechanism in plasma under the influence of complex electromagnetic field topology is still debatable. Here the authors present the evidence of the E×B mixing avalanche for gas breakdown in magnetized plasmas in fusion devices as tokamak.

The International Tokamak Physics Activity topical group on integrated operational scenarios has compiled a database of stationary H-mode discharges at q(95) similar to 3 from AUG, C-Mod, DIII-D, J ...

The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H = 1 at ßN ~ 1.8 and n/nGW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 under grant agreement No 633053. Peer Reviewed Article signat per 1.173 autors/es: X. Litaudon35, S. Abduallev39, M. Abhangi46, P. Abreu53, M. Afzal7, K.M. Aggarwal29, T. Ahlgren101, J.H. Ahn8, L. Aho-Mantila112, N. Aiba69, M. Airila112, R. Albanese105, V. Aldred7, D. Alegre93, E. Alessi45, P. Aleynikov55, A. Alfier12, A. Alkseev72, M. Allinson7, B. Alper7, E. Alves53, G. Ambrosino105, R. Ambrosino106, L. Amicucci90, V. Amosov88, E. Andersson Sundén22, M. Angelone90, M. Anghel85, C. Angioni62, L. Appel7, C. Appelbee7, P. Arena30, M. Ariola106, H. Arnichand8, S. Arshad41, A. Ash7, N. Ashikawa68, V. Aslanyan64, O. Asunta1, F. Auriemma12, Y. Austin7, L. Avotina103, M.D. Axton7, C. Ayres7, M. Bacharis24, A. Baciero57, D. Baião53, S. Bailey7, A. Baker7, I. Balboa7, M. Balden62, N. Balshaw7, R. Bament7, J.W. Banks7, Y.F. Baranov7, M.A. Barnard7, D. Barnes7, M. Barnes27, R. Barnsley55, A. Baron Wiechec7, L. Barrera Orte34, M. Baruzzo12, V. Basiuk8, M. Bassan55, R. Bastow7, A. Batista53, P. Batistoni90, R. Baughan7, B. Bauvir55, L. Baylor73, B. Bazylev56, J. Beal110, P.S. Beaumont7, M. Beckers39, B. Beckett7, A. Becoulet8, N. Bekris35, M. Beldishevski7, K. Bell7, F. Belli90, M. Bellinger7, É. Belonohy62, N. Ben Ayed7, N.A. Benterman7, H. Bergsåker42, J. Bernardo53, M. Bernert62, M. Berry7, L. Bertalot55, C. Besliu7, M. Beurskens63, B. Bieg61, J. Bielecki47, T. Biewer73, M. Bigi12, P. Bílková50, F. Binda22, A. Bisoffi31, J.P.S. Bizarro53, C. Björkas101, J. Blackburn7, K. Blackman7, T.R. Blackman7, P. Blanchard33, P. Blatchford7, V. Bobkov62, A. Boboc7, G. Bodnár113, O. Bogar18, I. Bolshakova60, T. Bolzonella12, N. Bonanomi97, F. Bonelli56, J. Boom62, J. Booth7, D. Borba35,53, D. Borodin39, I. Borodkina39, A. Botrugno90, C. Bottereau8, P. Boulting7, C. Bourdelle8, M. Bowden7, C. Bower7, C. Bowman110, T. Boyce7, C. Boyd7, H.J. Boyer7, J.M.A. Bradshaw7, V. Braic87, R. Bravanec40, B. Breizman107, S. Bremond8, P.D. Brennan7, S. Breton8, A. Brett7, S. Brezinsek39, M.D.J. Bright7, M. Brix7, W. Broeckx78, M. Brombin12, A. Brosławski65, D.P.D. Brown7, M. Brown7, E. Bruno55, J. Bucalossi8, J. Buch46, J. Buchanan7, M.A. Buckley7, R. Budny76, H. Bufferand8, M. Bulman7, N. Bulmer7, P. Bunting7, P. Buratti90, A. Burckhart62, A. Buscarino30, A. Busse7, N.K. Butler7, I. Bykov42, J. Byrne7, P. Cahyna50, G. Calabrò90, I. Calvo57, Y. Camenen4, P. Camp7, D.C. Campling7, J. Cane7, B. Cannas17, A.J. Capel7, P.J. Card7, A. Cardinali90, P. Carman7, M. Carr7, D. Carralero62, L. Carraro12, B.B. Carvalho53, I. Carvalho53, P. Carvalho53, F.J. Casson7, C. Castaldo90, N. Catarino53, J. Caumont7, F. Causa90, R. Cavazzana12, K. Cave-Ayland7, M. Cavinato12, M. Cecconello22, S. Ceccuzzi90, E. Cecil76, A. Cenedese12, R. Cesario90, C.D. Challis7, M. Chandler7, D. Chandra46, C.S. Chang76, A. Chankin62, I.T. Chapman7, S.C. Chapman28, M. Chernyshova49, G. Chitarin12, G. Ciraolo8, D. Ciric7, J. Citrin38, F. Clairet8, E. Clark7, M. Clark7, R. Clarkson7, D. Clatworthy7, C. Clements7, M. Cleverly7, J.P. Coad7, P.A. Coates7, A. Cobalt7, V. Coccorese105, V. Cocilovo90, S. Coda33, R. Coelho53, J.W. Coenen39, I. Coffey29, L. Colas8, S. Collins7, D. Conka103, S. Conroy22, N. Conway7, D. Coombs7, D. Cooper7, S.R. Cooper7, C. Corradino30, Y. Corre8, G. Corrigan7, S. Cortes53, D. Coster62, A.S. Couchman7, M.P. Cox7, T. Craciunescu86, S. Cramp7, R. Craven7, F. Crisanti90, G. Croci97, D. Croft7, K. Crombé15, R. Crowe7, N. Cruz53, G. Cseh113, A. Cufar81, A. Cullen7, M. Curuia85, A. Czarnecka49, H. Dabirikhah7, P. Dalgliesh7, S. Dalley7, J. Dankowski47, D. Darrow76, O. Davies7, W. Davis55,76, C. Day56, I.E. Day7, M. De Bock55, A. de Castro57, E. de la Cal57, E. de la Luna57, G. De Masi12, J. L. de Pablos57, G. De Temmerman55, G. De Tommasi105, P. de Vries55, K. Deakin7, J. Deane7, F. Degli Agostini12, R. Dejarnac50, E. Delabie73, N. den Harder38, R.O. Dendy7, J. Denis8, P. Denner39, S. Devaux62,104, P. Devynck8, F. Di Maio55, A. Di Siena62, C. Di Troia90, P. Dinca86, R. D’Inca62, B. Ding51, T. Dittmar39, H. Doerk62, R.P. Doerner9, T. Donné34, S.E. Dorling7, S. Dormido-Canto93, S. Doswon7, D. Douai8, P.T. Doyle7, A. Drenik62,81, P. Drewelow63, P. Drews39, Ph. Duckworth55, R. Dumont8, P. Dumortier58, D. Dunai113, M. Dunne62, I. Ďuran50, F. Durodié58, P. Dutta46, B. P. Duval33, R. Dux62, K. Dylst78, N. Dzysiuk22, P.V. Edappala46, J. Edmond7, A.M. Edwards7, J. Edwards7, Th. Eich62, A. Ekedahl8, R. El-Jorf7, C.G. Elsmore7, M. Enachescu84, G. Ericsson22, F. Eriksson16, J. Eriksson22, L.G. Eriksson36, B. Esposito90, S. Esquembri94, H.G. Esser39, D. Esteve8, B. Evans7, G.E. Evans7, G. Evison7, G.D. Ewart7, D. Fagan7, M. Faitsch62, D. Falie86, A. Fanni17, A. Fasoli33, J. M. Faustin33, N. Fawlk7, L. Fazendeiro53, N. Fedorczak8, R.C. Felton7, K. Fenton7, A. Fernades53, H. Fernandes53, J. Ferreira53, J.A. Fessey7, O. Février8, O. Ficker50, A. Field7, S. Fietz62, A. Figueiredo53, J. Figueiredo53,35, A. Fil8, P. Finburg7, M. Firdaouss8, U. Fischer56, L. Fittill7, M. Fitzgerald7, D. Flammini90, J. Flanagan7, C. Fleming7, K. Flinders7, N. Fonnesu90, J. M. Fontdecaba57, A. Formisano79, L. Forsythe7, L. Fortuna30, E. Fortuna-Zalesna19, M. Fortune7, S. Foster7, T. Franke34, T. Franklin7, M. Frasca30, L. Frassinetti42, M. Freisinger39, R. Fresa98, D. Frigione90, V. Fuchs50, D. Fuller35, S. Futatani6, J. Fyvie7, K. Gál34,62, D. Galassi2, K. Gałązka49, J. Galdon-Quiroga92, J. Gallagher7, D. Gallart6, R. Galvão10, X. Gao51, Y. Gao39, J. Garcia8, A. Garcia-Carrasco42, M. García-Muñoz92, J.-L. Gardarein3, L. Garzotti7, P. Gaudio95, E. Gauthier8, D.F. Gear7, S.J. Gee7, B. Geiger62, M. Gelfusa95, S. Gerasimov7, G. Gervasini45, M. Gethins7, Z. Ghani7, M. Ghate46, M. Gherendi86, J.C. Giacalone8, L. Giacomelli45, C.S. Gibson7, T. Giegerich56, C. Gil8, L. Gil53, S. Gilligan7, D. Gin54, E. Giovannozzi90, J.B. Girardo8, C. Giroud7, G. Giruzzi8, S. Glöggler62, J. Godwin7, J. Goff7, P. Gohil43, V. Goloborod’ko102, R. Gomes53, B. Gonçalves53, M. Goniche8, M. Goodliffe7, A. Goodyear7, G. Gorini97, M. Gosk65, R. Goulding76, A. Goussarov78, R. Gowland7, B. Graham7, M.E. Graham7, J. P. Graves33, N. Grazier7, P. Grazier7, N.R. Green7, H. Greuner62, B. Grierson76, F.S. Griph7, C. Grisolia8, D. Grist7, M. Groth1, R. Grove73, C.N. Grundy7, J. Grzonka19, D. Guard7, C. Guérard34, C. Guillemaut8,53, R. Guirlet8, C. Gurl7, H.H. Utoh69, L.J. Hackett7, S. Hacquin8,35, A. Hagar7, R. Hager76, A. Hakola112, M. Halitovs103, S.J. Hall7, S.P. Hallworth Cook7, C. Hamlyn-Harris7, K. 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Jachmich58,35, A.S. Jacobsen83, P. Jacquet7, J. Jansons103, A. Jardin8, A. Järvinen1, F. Jaulmes38, S. Jednoróg49, I. Jenkins7, C. Jeong20, I. Jepu86, E. Joffrin8, R. Johnson7, T. Johnson42, Jane Johnston7, L. Joita7, G. Jones7, T.T.C. Jones7, K.K. Hoshino69, A. Kallenbach62, K. Kamiya69, J. Kaniewski7, A. Kantor7, A. Kappatou62, J. Karhunen1, D. Karkinsky7, I. Karnowska7, M. Kaufman73, G. Kaveney7, Y. Kazakov58, V. Kazantzidis70, D.L. Keeling7, T. Keenan7, J. Keep7, M. Kempenaars7, C. Kennedy7, D. Kenny7, J. Kent7, O.N. Kent7, E. Khilkevich54, H.T. Kim35, H.S. Kim80, A. Kinch7, C. king7, D. King7, R.F. King7, D.J. Kinna7, V. Kiptily7, A. Kirk7, K. Kirov7, A. Kirschner39, G. Kizane103, C. Klepper73, A. Klix56, P. Knight7, S.J. Knipe7, S. Knott96, T. Kobuchi69, F. Köchl111, G. Kocsis113, I. Kodeli81, L. Kogan7, D. Kogut8, S. Koivuranta112, Y. Kominis70, M. Köppen39, B. Kos81, T. Koskela1, H.R. Koslowski39, M. Koubiti4, M. Kovari7, E. Kowalska-Strzęciwilk49, A. Krasilnikov88, V. Krasilnikov88, N. Krawczyk49, M. Kresina8, K. Krieger62, A. Krivska58, U. Kruezi7, I. Książek48, A. Kukushkin72, A. Kundu46, T. Kurki-Suonio1, S. Kwak20, R. Kwiatkowski65, O.J. Kwon13, L. Laguardia45, A. Lahtinen101, A. Laing7, N. Lam7, H.T. Lambertz39, C. Lane7, P.T. Lang62, S. Lanthaler33, J. Lapins103, A. Lasa101, J.R. Last7, E. Łaszyńska49, R. Lawless7, A. Lawson7, K.D. Lawson7, A. Lazaros70, E. Lazzaro45, J. Leddy110, S. Lee66, X. Lefebvre7, H.J. Leggate32, J. Lehmann7, M. Lehnen55, D. Leichtle41, P. Leichuer7, F. Leipold55,83, I. Lengar81, M. Lennholm36, E. Lerche58, A. Lescinskis103, S. Lesnoj7, E. Letellier7, M. Leyland110, W. Leysen78, L. Li39, Y. Liang39, J. Likonen112, J. Linke39, Ch. Linsmeier39, B. Lipschultz110, G. Liu55, Y. Liu51, V.P. Lo Schiavo105, T. Loarer8, A. Loarte55, R.C. Lobel7, B. Lomanowski1, P.J. Lomas7, J. Lönnroth1,35, J. M. López94, J. López-Razola57, R. Lorenzini12, U. Losada57, J.J. Lovell7, A.B. Loving7, C. Lowry36, T. Luce43, R.M.A. Lucock7, A. 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Meigh7, A.G. Meigs7, G. Meisl62, S. Meitner73, L. Meneses53, S. Menmuir7,42, K. Mergia71, I.R. Merrigan7, Ph. Mertens39, S. Meshchaninov88, A. Messiaen58, H. Meyer7, S. Mianowski65, R. Michling55, D. Middleton-Gear7, J. Miettunen1, F. Militello7, E. Militello-Asp7, G. Miloshevsky77, F. Mink62, S. Minucci105, Y. Miyoshi69, J. Mlynář50, D. Molina8, I. Monakhov7, M. Moneti109, R. Mooney7, S. Moradi37, S. Mordijck43, L. Moreira7, R. Moreno57, F. Moro90, A.W. Morris7, J. Morris7, L. Moser26, S. Mosher73, D. Moulton7,1, A. Murari12,35, A. Muraro45, S. Murphy7, N.N. Asakura69, Y.S. Na80, F. Nabais53, R. Naish7, T. Nakano69, E. Nardon8, V. Naulin83, M.F.F. Nave53, I. Nedzelski53, G. Nemtsev88, F. Nespoli33, A. Neto41, R. Neu62, V.S. Neverov72, M. Newman7, K.J. Nicholls7, T. Nicolas33, A.H. Nielsen83, P. Nielsen12, E. Nilsson8, D. Nishijima99, C. Noble7, M. Nocente97, D. Nodwell7, K. Nordlund101, H. Nordman16, R. Nouailletas8, I. Nunes53, M. Oberkofler62, T. Odupitan7, M.T. Ogawa69, T. O’Gorman7, M. Okabayashi76, R. Olney7, O. Omolayo7, M. O’Mullane21, J. Ongena58, F. Orsitto11, J. Orszagh18, B.I. Oswuigwe7, R. Otin7, A. Owen7, R. Paccagnella12, N. Pace7, D. Pacella90, L.W. Packer7, A. Page7, E. Pajuste103, S. Palazzo30, S. Pamela7, S. Panja46, P. Papp18, R. Paprok50, V. Parail7, M. Park66, F. Parra Diaz27, M. Parsons73, R. Pasqualotto12, A. Patel7, S. Pathak46, D. Paton7, H. Patten33, A. Pau17, E. Pawelec48, C. Paz Soldan43, A. Peackoc36, I.J. Pearson7, S.-P. Pehkonen112, E. Peluso95, C. Penot55, A. Pereira57, R. Pereira53, P.P. Pereira Puglia7, C. Perez von Thun35,39, S. Peruzzo12, S. Peschanyi56, M. Peterka50, P. Petersson42, G. Petravich113, A. Petre84, N. Petrella7, V. Petržilka50, Y. Peysson8, D. Pfefferlé33, V. Philipps39, M. Pillon90, G. Pintsuk39, P. Piovesan12, A. Pires dos Reis52, L. Piron7, A. Pironti105, F. Pisano17, R. Pitts55, F. Pizzo79, V. Plyusnin53, N. Pomaro12, O.G. Pompilian86, P.J. Pool7, S. Popovichev7, M.T. Porfiri90, C. Porosnicu86, M. Porton7, G. Possnert22, S. Potzel62, T. Powell7, J. Pozzi7, V. Prajapati46, R. Prakash46, G. Prestopino95, D. Price7, M. Price7, R. Price7, P. Prior7, R. Proudfoot7, G. Pucella90, P. Puglia52, M.E. Puiatti12, D. Pulley7, K. Purahoo7, Th. Pütterich62, E. Rachlew25, M. Rack39, R. Ragona58, M.S.J. Rainford7, A. Rakha6, G. Ramogida90, S. Ranjan46, C.J. Rapson62, J.J. Rasmussen83, K. Rathod46, G. Rattá57, S. Ratynskaia82, G. Ravera90, C. Rayner7, M. Rebai97, D. Reece7, A. Reed7, D. Réfy113, B. Regan7, J. Regaña34, M. Reich62, N. Reid7, F. Reimold39, M. Reinhart34, M. Reinke110,73, D. Reiser39, D. Rendell7, C. Reux8, S.D.A. Reyes Cortes53, S. Reynolds7, V. Riccardo7, N. Richardson7, K. Riddle7, D. Rigamonti97, F.G. Rimini7, J. Risner73, M. Riva90, C. Roach7, R.J. Robins7, S.A. Robinson7, T. Robinson7, D.W. Robson7, R. Roccella55, R. Rodionov88, P. Rodrigues53, J. Rodriguez7, V. Rohde62, F. Romanelli90, M. Romanelli7, S. Romanelli7, J. Romazanov39, S. Rowe7, M. Rubel42, G. Rubinacci105, G. Rubino12, L. Ruchko52, M. Ruiz94, C. Ruset86, J. Rzadkiewicz65, S. Saarelma7, R. Sabot8, E. Safi101, P. Sagar7, G. Saibene41, F. Saint-Laurent8, M. Salewski83, A. Salmi112, R. Salmon7, F. Salzedas53, D. Samaddar7, U. Samm39, D. Sandiford7, P. Santa46, M.I.K. Santala1, B. Santos53, A. Santucci90, F. Sartori41, R. Sartori41, O. Sauter33, R. Scannell7, T. Schlummer39, K. Schmid62, V. Schmidt12, S. Schmuck7, M. Schneider8, K. Schöpf102, D. Schwörer32, S.D. Scott76, G. Sergienko39, M. Sertoli62, A. Shabbir15, S.E. Sharapov7, A. Shaw7, R. Shaw7, H. Sheikh7, A. Shepherd7, A. Shevelev54, A. Shumack38, G. Sias17, M. Sibbald7, B. Sieglin62, S. Silburn7, A. Silva53, C. Silva53, P.A. Simmons7, J. Simpson7, J. Simpson-Hutchinson7, A. Sinha46, S.K. Sipilä1, A.C.C. Sips36, P. Sirén112, A. Sirinelli55, H. Sjöstrand22, M. Skiba22, R. Skilton7, K. Slabkowska49, B. Slade7, N. Smith7, P.G. Smith7, R. Smith7, T.J. Smith7, M. Smithies110, L. Snoj81, S. Soare85, E. R. Solano35,57, A. Somers32, C. Sommariva8, P. Sonato12, A. Sopplesa12, J. Sousa53, C. Sozzi45, S. Spagnolo12, T. Spelzini7, F. Spineanu86, G. Stables7, I. Stamatelatos71, M.F. Stamp7, P. Staniec7, G. Stankūnas59, C. Stan-Sion84, M.J. Stead7, E. Stefanikova42, I. Stepanov58, A.V. Stephen7, M. Stephen46, A. Stevens7, B.D. Stevens7, J. Strachan76, P. Strand16, H.R. Strauss44, P. Ström42, G. Stubbs7, W. Studholme7, F. Subba75, H.P. Summers21, J. Svensson63, Ł. Świderski65, T. Szabolics113, M. Szawlowski49, G. Szepesi7, T.T. Suzuki69, B. Tál113, T. Tala112, A.R. Talbot7, S. Talebzadeh95, C. Taliercio12, P. Tamain8, C. Tame7, W. Tang76, M. Tardocchi45, L. Taroni12, D. Taylor7, K.A. Taylor7, D. Tegnered16, G. Telesca15, N. Teplova54, D. Terranova12, D. Testa33, E. Tholerus42, J. Thomas7, J.D. Thomas7, P. Thomas55, A. Thompson7, C.-A. Thompson7, V.K. Thompson7, L. Thorne7, A. Thornton7, A.S. Thrysøe83, P.A. Tigwell7, N. Tipton7, I. Tiseanu86, H. Tojo69, M. Tokitani67, P. Tolias82, M. Tomeš50, P. Tonner7, M. Towndrow7, P. Trimble7, M. Tripsky58, M. Tsalas38, P. Tsavalas71, D. Tskhakaya jun102, I. Turner7, M.M. Turner32, M. Turnyanskiy34, G. Tvalashvili7, S.G.J. Tyrrell7, A. Uccello45, Z. Ul-Abidin7, J. Uljanovs1, D. Ulyatt7, H. Urano69, I. Uytdenhouwen78, A.P. Vadgama7, D. Valcarcel7, M. Valentinuzzi8, M. Valisa12, P. Vallejos Olivares42, M. Valovic7, M. Van De Mortel7, D. Van Eester58, W. Van Renterghem78, G.J. van Rooij38, J. Varje1, S. Varoutis56, S. Vartanian8, K. Vasava46, T. Vasilopoulou71, J. Vega57, G. Verdoolaege58, R. Verhoeven7, C. Verona95, G. Verona Rinati95, E. Veshchev55, N. Vianello45, J. Vicente53, E. Viezzer62,92, S. Villari90, F. Villone100, P. Vincenzi12, I. Vinyar74, B. Viola90, A. Vitins103, Z. Vizvary7, M. Vlad86, I. Voitsekhovitch34, P. Vondráček50, N. Vora7, T. Vu8, W.W. Pires de Sa52, B. Wakeling7, C.W.F. Waldon7, N. Walkden7, M. Walker7, R. Walker7, M. Walsh55, E. Wang39, N. Wang39, S. Warder7, R.J. Warren7, J. Waterhouse7, N.W. Watkins28, C. Watts55, T. Wauters58, A. Weckmann42, J. Weiland23, H. Weisen33, M. Weiszflog22, C. Wellstood7, A.T. West7, M.R. Wheatley7, S. Whetham7, A.M. Whitehead7, B.D. Whitehead7, A.M. Widdowson7, S. Wiesen39, J. Wilkinson7, J. Williams7, M. Williams7, A.R. Wilson7, D.J. Wilson7, H.R. Wilson110, J. Wilson7, M. Wischmeier62, G. Withenshaw7, A. Withycombe7, D.M. Witts7, D. Wood7, R. Wood7, C. Woodley7, S. Wray7, J. Wright7, J.C. Wright64, J. Wu89, S. Wukitch64, A. Wynn110, T. Xu7, D. Yadikin16, W. Yanling39, L. Yao89, V. Yavorskij102, M.G. Yoo80, C. Young7, D. Young7, I.D. Young7, R. Young7, J. Zacks7, R. Zagorski49, F.S. Zaitsev18, R. Zanino75, A. Zarins103, K.D. Zastrow7, M. Zerbini90, W. Zhang62, Y. Zhou42, E. Zilli12, V. Zoita86, S. Zoletnik113, I. Zychor65 and JET Contributorsa // EUROfusion Consortium JET, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom / 1 Aalto University, PO Box 14100, FIN-00076 Aalto, Finland / 2 Aix Marseille Université, CNRS, Centrale Marseille, M2P2 UMR 7340, 13451, Marseille, France / 3 Aix-Marseille Université, CNRS, IUSTI UMR 7343, 13013 Marseille, France / 4 Aix-Marseille Université, CNRS, PIIM, UMR 7345, 13013 Marseille, France / 5 Arizona State University, Tempe, AZ, United States of America / 6 Barcelona Supercomputing Center, Barcelona, Spain / 7 CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB, United Kingdom / 8 CEA, IRFM, F-13108 Saint Paul Lez Durance, France / 9 Center for Energy Research, University of California at San Diego, La Jolla, CA 92093, United States of America / 10 Centro Brasileiro de Pesquisas Fisicas, Rua Xavier Sigaud, 160, Rio de Janeiro CEP 22290-180, Brazil / 11 Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 12 Consorzio RFX, corso Stati Uniti 4, 35127 Padova, Italy / 13 Daegu University, Jillyang, Gyeongsan, Gyeongbuk 712-174, Republic of Korea / 14 Departamento de Física, Universidad Carlos III de Madrid, 28911 Leganés, Madrid, Spain / 15 Department of Applied Physics UG (Ghent University) St-Pietersnieuwstraat 41 B-9000 Ghent, Belgium / 16 Department of Earth and Space Sciences, Chalmers University of Technology, SE-41296 Gothenburg, Sweden / 17 Department of Electrical and Electronic Engineering, University of Cagliari, Piazza d’Armi 09123, Cagliari, Italy / 18 Department of Experimental Physics, Faculty of Mathematics, Physics and Informatics Comenius University Mlynska dolina F2, 84248 Bratislava, Slovakia / 19 Department of Materials Science, Warsaw University of Technology, PL-01-152 Warsaw, Poland / 20 Department of Nuclear and Quantum Engineering, KAIST, Daejeon 34141, Korea / 21 Department of Physics and Applied Physics, University of Strathclyde, Glasgow, G4 ONG, United Kingdom / 22 Department of Physics and Astronomy, Uppsala University, SE-75120 Uppsala, Sweden / 23 Department of Physics, Chalmers University of Technology, SE-41296 Gothenburg, Sweden / 24 Department of Physics, Imperial College London, London, SW7 2AZ, United Kingdom / 25 Department of Physics, SCI, KTH, SE-10691 Stockholm, Sweden / 26 Department of Physics, University of Basel, Basel, Switzerland / 27 Department of Physics, University of Oxford, Oxford, OX1 2JD, United Kingdom / 28 Department of Physics, University of Warwick, Coventry, CV4 7AL, United Kingdom / 29 Department of Pure and Applied Physics, Queens University, Belfast, BT7 1NN, United Kingdom / 30 Dipartimento di Ingegneria Elettrica Elettronica e Informatica, Università degli Studi di Catania, 95125 Catania, Italy / 31 Dipartimento di Ingegneria Industriale, University of Trento, Trento, Italy / 32 Dublin City University (DCU), Dublin, Ireland / 33 Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne, Switzerland / 34 EUROfusion Programme Management Unit, Boltzmannstr. 2, 85748 Garching, Germany / 35 EUROfusion Programme Management Unit, Culham Science Centre, Culham, OX14 3DB, United Kingdom / 36 European Commission, B-1049 Brussels, Belgium / 37 Fluid and Plasma Dynamics, ULB—Campus Plaine—CP 231 Boulevard du Triomphe, 1050 Bruxelles, Belgium / 38 FOM Institute DIFFER, Eindhoven, Netherlands / 39 Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung—Plasmaphysik, 52425 Jülich, Germany / 40 Fourth State Research, 503 Lockhart Dr, Austin, TX, United States of America / 41 Fusion for Energy Joint Undertaking, Josep Pl. 2, Torres Diagonal Litoral B3, 08019, Barcelona, Spain / 42 Fusion Plasma Physics, EES, KTH, SE-10044 Stockholm, Sweden / 43 General Atomics, PO Box 85608, San Diego, CA 92186-5608, United States of America / 44 HRS Fusion, West Orange, NJ, United States of America / 45 IFP-CNR, via R. Cozzi 53, 20125 Milano, Italy / 46 Institute for Plasma Research, Bhat, Gandhinagar-382 428, Gujarat State, India / 47 Institute of Nuclear Physics, Radzikowskiego 152, 31-342 Kraków, Poland / 48 Institute of Physics, Opole University, Oleska 48, 45-052 Opole, Poland / 49 Institute of Plasma Physics and Laser Microfusion, Hery 23, 01-497 Warsaw, Poland / 50 Institute of Plasma Physics AS CR, Za Slovankou 1782/3, 182 00 Praha 8, Czechia / 51 Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, People’s Republic of China / 52 Instituto de Física, Universidade de São Paulo, Rua do Matão Travessa R Nr.187 CEP 05508-090 Cidade Universitária, São Paulo, Brasil / 53 Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, Lisboa, Portugal / 54 Ioffe Physico-Technical Institute, 26 Politekhnicheskaya, St Petersburg 194021, Russian Federation / 55 ITER Organization, Route de Vinon, CS 90 046, 13067 Saint Paul Lez Durance, France / 56 Karlsruhe Institute of Technology, PO Box 3640, D-76021 Karlsruhe, Germany / 57 Laboratorio Nacional de Fusión, CIEMAT, Madrid, Spain / 58 Laboratory for Plasma Physics Koninklijke Militaire School—Ecole Royale Militaire, Renaissancelaan 30 Avenue de la Renaissance B-1000, Brussels, Belgium / 59 Lithuanian energy institute, Breslaujos g. 3, LT-44403, Kaunas, Lithuania / 60 Magnetic Sensor Laboratory, Lviv Polytechnic National University, Lviv, Ukraine / 61 Maritime University of Szczecin, Waly Chrobrego 1-2, 70-500 Szczecin, Poland / 62 Max-Planck-Institut für Plasmaphysik, D-85748 Garching, Germany / 63 Max-Planck-Institut für Plasmaphysik, Teilinsitut Greifswald, D-17491 Greifswald, Germany / 64 MIT Plasma Science and Fusion Centre, Cambridge, MA 02139, United States of America / 65 National Centre for Nuclear Research (NCBJ), 05-400 Otwock-Świerk, Poland / 66 National Fusion Research Institute (NFRI), 169-148 Gwahak-ro, Yuseong-gu, Daejeon 305-806, Republic of Korea / 67 National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan / 68 National Institute for Fusion Science, Toki, 509-5292, Japan / 69 National Institutes for Quantum and Radiological Science and Technology, Naka, Ibaraki 311-0193, Japan / 70 National Technical University of Athens, Iroon Politechniou 9, 157 73 Zografou, Athens, Greece / 71 NCSR ‘Demokritos’, 153 10, Agia Paraskevi Attikis, Greece / 72 NRC Kurchatov Institute, 1 Kurchatov Square, Moscow 123182, Russian Federation / 73 Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169, United States of America / 74 PELIN LLC, 27a, Gzhatskaya Ulitsa, Saint Petersburg, 195220, Russian Federation / 75 Politecnico di Torino, Corso Duca degli Abruzzi 24, I-10129 Torino, Italy / 76 Princeton Plasma Physics Laboratory, James Forrestal Campus, Princeton, NJ 08543, United States of America / 77 Purdue University, 610 Purdue Mall, West Lafayette, IN 47907, United States of America / 78 SCK-CEN, Nuclear Research Centre, 2400 Mol, Belgium / 79 Second University of Napoli, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 80 Seoul National University, Shilim-Dong, Gwanak-Gu, Republic of Korea / 81 Slovenian Fusion Association (SFA), Jozef Stefan Institute, Jamova 39, SI-1000 Ljubljana, Slovenia / 82 Space and Plasma Physics, EES, KTH SE-100 44 Stockholm, Sweden / 83 Technical University of Denmark, Department of Physics, Bldg 309, DK-2800 Kgs Lyngby, Denmark / 84 The ‘Horia Hulubei’ National Institute for Physics and Nuclear Engineering, Magurele-Bucharest, Romania / 85 The National Institute for Cryogenics and Isotopic Technology, Ramnicu Valcea, Romania / 86 The National Institute for Laser, Plasma and Radiation Physics, Magurele-Bucharest, Romania / 87 The National Institute for Optoelectronics, Magurele-Bucharest, Romania / 88 Troitsk Insitute of Innovating and Thermonuclear Research (TRINITI), Troitsk 142190, Moscow Region, Russian Federation / 89 University of Electronic Science and Technology of China, Chengdu, People’s Republic of China / 90 Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma), Italy / 91 Universidad Complutense de Madrid, Madrid, Spain / 92 Universidad de Sevilla, Sevilla, Spain / 93 Universidad Nacional de Educación a Distancia, Madrid, Spain / 94 Universidad Politécnica de Madrid, Grupo I2A2, Madrid, Spain / 95 Università di Roma Tor Vergata, Via del Politecnico 1, Roma, Italy / 96 University College Cork (UCC), Ireland / 97 University Milano-Bicocca, piazza della Scienza 3, 20126 Milano, Italy / 98 University of Basilicata, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 99 University of California, 1111 Franklin St., Oakland, CA 94607, United States of America / 100 University of Cassino, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 101 University of Helsinki, PO Box 43, FI-00014 University of Helsinki, Finland / 102 University of Innsbruck, Fusion@Österreichische Akademie der Wissenschaften (ÖAW), Innsbruck, Austria / 103 University of Latvia, 19 Raina Blvd., Riga, LV 1586, Latvia / 104 University of Lorraine, CNRS, UMR7198, YIJL, Nancy, France / 105 University of Napoli ‘Federico II’, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 106 University of Napoli Parthenope, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 107 University of Texas at Austin, Institute for Fusion Studies, Austin, TX 78712, United States of America / 108 University of Toyama, Toyama, 930-8555, Japan / 109 University of Tuscia, DEIM, Via del Paradiso 47, 01100 Viterbo, Italy / 110 University of York, Heslington, York YO10 5DD, United Kingdom / 111 Vienna University of Technology, Fusion@Österreichische Akademie der Wissenschaften (ÖAW), Austria / 112 VTT Technical Research Centre of Finland, PO Box 1000, FIN-02044 VTT, Finland / 113 Wigner Research Centre for Physics, PO Box 49, H-1525 Budapest, Hungary

The final phase of the JET Programme in Support of ITER plans to operate with 100% tritium (TT) followed by deuterium-tritium (DT) operation, to help minimise risks and delays in the execution of the ITER Research Plan and the achievement of a fusion gain Q∼10. In preparation for tritium operation, an 8-week technical rehearsal of the procedures and systems is being executed in 2015-16. The aim is to commission (in deuterium or hydrogen) and characterise the NBI performance with a tritium-like gas feed configuration, test the prototype tritium introduction module, commission the torus hall depression and oxygen depletion system, carry out emergency exercises, train personnel and to test and optimise operational procedures. The rehearsal demonstrated that JET is still capable of operating with tritium in a safe and reliable manner, and most importantly, provided operational experience and lead to recommendations for the preparation of these future tritium campaigns on JET.

Extensive preparations are now underway for an experiment in the Joint European Torus (JET) using tritium and deuterium–tritium mixtures. The goals of this experiment are described as well as the progress that has been made in developing plasma operational scenarios and physics reference pulses for use in deuterium–tritium and full tritium plasmas. At present, the high performance plasmas to be tested with tritium are based on either a conventional ELMy H-mode at high plasma current and magnetic field (operation at up to 4 MA and 4 T is being prepared) or the so-called improved H-mode or hybrid regime of operation in which high normalised plasma pressure at somewhat reduced plasma current results in enhanced energy confinement. Both of these regimes are being re-developed in conjunction with JET's ITER-like Wall (ILW) of beryllium and tungsten. The influence of the ILW on plasma operation and performance has been substantial. Considerable progress has been made on optimising performance with the all-metal wall. Indeed, operation at the (normalised) ITER reference confinement and pressure has been re-established in JET albeit not yet at high current. In parallel with the physics development, extensive technical preparations are being made to operate JET with tritium. The state and scope of these preparations is reviewed, including the work being done on the safety case for DT operation and on upgrading machine infrastructure and diagnostics. A specific example of the latter is the planned calibration at 14 MeV of JET neutron diagnostics.

A review of the development of advanced tokamak scenarios at the Joint European Torus (JET) is presented. It has been established that the current profile plays an important role in these regimes, and the presentation of the experimental achievement has been organized with this in mind. The main achievements are discussed: from high beta plasmas starting with a fully diffused plasma current; from hybrid scenarios with a flat current profile and central q around unity; from pellet-enhanced modes where the role of reversed magnetic shear (transiently) was first established; from optimized shear configurations with weakly reversed shear allowing the establishment of internal transport barriers in D-T plasmas for the first time, including the production of 8.2 MW of fusion power; and from strongly reversed shear and steady-state scenarios. The required development of the control techniques for these advanced scenarios is also described. The results obtained have significantly contributed to the development of advanced scenarios for ITER operation. The prospects for further development of hybrid and steady-state scenarios for ITER are indicated in view of the ongoing upgrades to additional heating systems in JET.

Since 1998 ASDEX Upgrade has developed stationary H-modes that routinely obtain confinement enhancement factors H98(y,2) > 1 and normalized beta, βN = 2–3. These discharges are characterized by a q-profile with low magnetic shear in the centre and q(0) ~ 1. New results presented here concentrate on extending the operational range of these improved H-modes at ASDEX Upgrade and extrapolating the results to ITER. Discharges are obtained at high density, over a wide range of plasma collisionality and with a first wall predominantly covered by tungsten coated carbon tiles. The performance is optimized for q95 ranging from 3 to 5. At q95 ~ 3 real time control of βN is used and in some cases ECCD to suppress NTM activity at low βN ~ 2. For the extrapolation to ITER, the fusion power is calculated using the same thermal beta (βN,th) and kinetic profile shapes as obtained in ASDEX Upgrade and setting ne/nGW = 0.85. The fusion gain that could be obtained is evaluated using different confinement scaling expressions. The results indicate that improved H-modes are a candidate for an ITER hybrid scenario or could extend ITER operation beyond what is currently foreseen using standard H-modes.

Safe operation of ASDEX Upgrade is guaranteed by a conventional hierarchy of simple and robust hard-wired systems for personnel and machine protection featuring standardized switch-off procedures. Machine protection and handling of off-normal events is further enhanced and peak and lifetime stress minimized through the plasma control system. Based on a real-time process model supporting safety critical applications with data quality tagging, process self-monitoring, watchdog monitoring and alarm propagation, processes detect complex and critical failures and reliably perform case-sensitive counter measures. Intelligent real-time failure handling is done with hardware or software redundancy and performance degradation, or modification of reference values to continue or terminate discharges with reduced machine stress. Examples implemented so far on ASDEX Upgrade are given, such as recovery from measurement failures, switch-over of redundant actuators, handling of actuator limitations, detection of plasma instabilities, plasma state dependent soft landing, or handling of failed switch-off procedures through breakers disconnecting the machine from grid.

Integrated simulations are performed to establish a physics basis, in conjunction with present tokamak experiments, for the operating modes in the International Thermonuclear Experimental Reactor (ITER). Simulations of the hybrid mode are done using both fixed and free-boundary 1.5D transport evolution codes including CRONOS, ONETWO, TSC/TRANSP, TOPICS and ASTRA. The hybrid operating mode is simulated using the GLF23 and CDBM05 energy transport models. The injected powers are limited to the negative ion neutral beam, ion cyclotron and electron cyclotron heating systems. Several plasma parameters and source parameters are specified for the hybrid cases to provide a comparison of 1.5D core transport modelling assumptions, source physics modelling assumptions, as well as numerous peripheral physics modelling. Initial results indicate that very strict guidelines will need to be imposed on the application of GLF23, for example, to make useful comparisons. Some of the variations among the simulations are due to source models which vary widely among the codes used. In addition, there are a number of peripheral physics models that should be examined, some of which include fusion power production, bootstrap current, treatment of fast particles and treatment of impurities. The hybrid simulations project to fusion gains of 5.6–8.3, βN values of 2.1–2.6 and fusion powers ranging from 350 to 500 MW, under the assumptions outlined in section 3. Simulations of the steady state operating mode are done with the same 1.5D transport evolution codes cited above, except the ASTRA code. In these cases the energy transport model is more difficult to prescribe, so that energy confinement models will range from theory based to empirically based. The injected powers include the same sources as used for the hybrid with the possible addition of lower hybrid. The simulations of the steady state mode project to fusion gains of 3.5–7, βN values of 2.3–3.0 and fusion powers of 290 to 415 MW, under the assumptions described in section 4. These simulations will be presented and compared with particular focus on the resulting temperature profiles, source profiles and peripheral physics profiles. The steady state simulations are at an early stage and are focused on developing a range of safety factor profiles with 100% non-inductive current.

Significant progress has been made in the area of advanced modes of operation that are candidates for achieving steady state conditions in a fusion reactor. The corresponding parameters, domain of operation, scenarios and integration issues of advanced scenarios are discussed in this chapter. A review of the presently developed scenarios, including discussions on operational space, is given. Significant progress has been made in the domain of heating and current drive in recent years, especially in the domain of off-axis current drive, which is essential for the achievement of the required current profile. The actuators for heating and current drive that are necessary to produce and control the advanced tokamak discharges are discussed, including modelling and predictions for ITER. The specific control issues for steady state operation are discussed, including the already existing experimental results as well as the various strategies and needs (qψ profile control and temperature gradients). Achievable parameters for the ITER steady state and hybrid scenarios with foreseen heating and current drive systems are discussed using modelling including actuators, allowing an assessment of achievable current profiles. Finally, a summary is given in the last section including outstanding issues and recommendations for further research and development.

Since the ITER-like wall in JET (JET-ILW) came into operation, dedicated ITER-like plasma current (I p) ramp-up (RU) and ramp-down (RD) experiments have been performed and matched to similar discharges with the carbon wall (JET-C). The experiments show that access to H-mode early in the I p RU phase and maintaining H-mode in the I p RD as long as possible are instrumental to achieve low internal plasma inductance (l i) and to minimize flux consumption. In JET-ILW, at a given current rise rate similar variations in l i (0.7–0.9) are obtained as in JET-C. In most discharges no strong W accumulation is observed. However, in some low density cases during the early phase of the I p strong core radiation due to W influx led to hollow electron temperature (T e) profiles. In JET-ILW Z eff is significantly lower than in JET-C. W significantly disturbs the discharge evolution when the W concentration approaches 10−4; this threshold is confirmed by predictive transport modelling using the CRONOS code. I p RD experiments in JET-ILW confirm the result of JET-C that sustained H-mode and elongation reduction are both instrumental in controlling l i.

Control of the shape of the current density profile is essential to improve the confinement and the stability in the plasma, particularly for advanced tokamak scenarios with internal transport barriers. For real-time control of the current density profile, it is necessary to identify a model that describes the time evolution of the current density profile when additional current is driven by external current drive tools. This paper focuses on the identification of such models in ASDEX Upgrade. Neutral beam injection is planned as a tool to control the current density profile in ASDEX Upgrade. The possibility of modifying the current density profile using neutral beam injection is investigated by the ASTRA code simulations using the Weiland transport model. It is difficult to derive a physics-based model for the current profile modification with neutral beam injection because it is nonlinear and multivariable. Therefore, a numerical model, a state-space model suited for systems with many input and output signals, is employed for the modeling. The matrices of the state-space model are estimated using a database by a standard prediction error method that minimizes the difference between the model output and the reference output. The database consists of a set of perturbed input signals and simulated output signals. The input signals are the variations of neutral beam power from different beam sources, and the output signals are the variations of the total plasma pressure and the current density profile. The ASTRA code with the Weiland transport model is used for the simulations to create the database since experimental data are currently not available at ASDEX Upgrade. A test of identified models is carried out using another database, also produced by ASTRA, applying a step response pattern to the input signals. It is found that the models obtained predict the output of this database with high accuracies. It is possible to apply the approach developed here to other actuators in a similar way for the current profile control in existing and future experiments.

Empirical estimates indicate that the transient divertor power load induced by type-I ELMs in the standard H-mode ITER scenario might result in intolerably short target life times. Significant experimental effort has been put on the exploration of recipes to avoid or at least mitigate the ELM size by external intervention. Experimental ELM mitigation studies (and related modeling) are of interest from different points of views. First, the active intervention gives access to new, externally controlled ELMy discharge scenarios and ELM diagnostic techniques. Second, the ELM response on the type and amplitude of the applied perturbation provides additional hints on the ELM stability. Third, mitigated ELMs, possibly different from spontaneous ones, and their scaling are of interest 'per se', in case that mitigation is to be finally applied in ITER. In this contribution we report on ELM mitigation results in ASDEX Upgrade, where we further restrict the scope to those methods involving 3D, local non-axisymmetric perturbations. Perturbations applied were cryogenic pellet injection, localized supersonic pulsed injection of D gas and laser blow off from targets carrying C or Al micro pellets. The corresponding perturbations differ strongly both in the space and time domain and provide an instrument to probe the local stability. Varying the parameters of the external perturbations by e.g. changing the amount of injected material or the injection speed we mapped out the ELM trigger threshold with high spatial and temporal resolutions.

The 'improved H-mode' regime, realized in ASDEX Upgrade in 1998 and confirmed by other devices, demonstrates the integration of advanced performance beyond the standard H-mode for confinement (confinement enhancement factor H98(y,2) ? 1.6), stability (normalized beta ?N ~ 3?3.5) and, at densities close to the Greenwald density, exhaust in stationary discharges longer than 40 confinement times or up to six resistive diffusion times. The q-profile is characterized by low central magnetic shear and axis safety factor q0 > 1 that is obtained by particular discharge and heating ramp-up scenarios and maintained via fishbones or benign higher (m, n) instabilities without using elaborate current control. Core transport is still governed by drift-wave turbulence with stiff temperature profiles, but density profiles are more strongly peaked and contribute to the increase in global confinement. A further contribution manifests itself by enhanced pressures at the edge barrier pedestal top and at the ? = 0.9 surface both increasing with the input power. (3, 2) NTMs remain small, enabling routine operation up to ?N ~ 3 (limited by (2, 1) NTMs) at ITER relevant collisionalities, for normalized Larmor radii down to four times the ITER value and for a broad range of q95 = 3?5. Tailored heat deposition including central wave heating allows for a compromise in density peaking for enhanced confinement and tolerable high-Z impurity concentrations even with tungsten coated structures.As far as the ITER relevance of this regime is concerned, its compatibility with significant central electron heating, low collisionality and even densities close to the Greenwald density combined with type-II ELMs and ?N ~ 3.5 is of importance. The GLF23 turbulence model still predicts peaked density profiles (R/Ln ~ 3) and sufficient transport to avoid impurity accumulation. At low q95 ~ 3 the fusion performance in terms of is more than doubled compared with the ITER baseline scenario (performance factor ~ 0.2) extrapolating to long Q 10 pulses on ITER. At medium q's bootstrap current fractions up to 50% and performance factors close to 0.2 can be achieved resulting in long inductive pulse lengths of ~1?h allowing ITER 'hybrid' operation at Q ? 9.

The advanced tokamak programme on DIII-D is to develop the scientific basis for steady state, high performance operation in future devices. We report on experiments attempting to demonstrate sustainment of 100% noninductive current for several seconds at high beta, using up to 2.5 MW of off-axis electron cyclotron current drive and up to 15 MW of neutral beam injection with q95 ≈ 5. A 100% noninductive condition was achieved with βT = 3.6%, βN = 3.5, H89 = 2.4 and improved current drive alignment. However, the duration of this phase was limited by the pressure profile evolution leading to magnetohydrodynamic instabilities after about 0.7 s. In a separate discharge, a nearly (~90%) noninductive, stationary condition was maintained for one current relaxation time (1.8 s) only limited by the duration of the hardware system. These experiments have achieved normalized fusion performance with bootstrap current fraction fBS ≈ 60%, consistent with requirements for the ITER Q = 5 steady-state scenarios. The modelling tools that were successfully employed to devise experiments in DIII-D are applied to ITER, indicating that full noninductive operation is plausible for an ITER steady-state scenario with Q ≈ 5.

Experiments in the DIII-D tokamak have demonstrated the ability to sustain ELMing H-mode discharges with high beta and good confinement quality under stationary conditions. These experiments have shown the ability to sustain normalized fusion performance (in terms of {beta}{sub N}H{sub 89P} /q{sub 95}{sup 2}) at or above that projected for Q{sub fus} = 10 operation in the International Thermonuclear Experimental Reactor (ITER) design over a wide range in operating parameters. In the best cases, operation is maintained at the free boundary, n = 1 stability limit. Confinement is found to be better than standard H-mode confinement scalings over a wide range in operation space, and experimentally measured transport is consistent with predictions from the GLF23 transport code. Projections using the standard ITER H-mode scaling laws based on these discharges indicate that Q{sub fus} = 5 can be maintained for >5400 s in ITER at q{sub 95} = 4.5 while Q{sub fus} = 40 can be obtained for {approx}2400 s at q{sub 95} = 3.2. These projected performance levels further validate the ITER design and suggest that long-pulse, high neutron fluence operation as well as very high fusion gain operation may be possible in next-generation tokamaks.

Significant progress has been made on ASDEX Upgrade during the last two years in the basic understanding of transport, in the extension of the improved H-mode in parameter space and towards an integrated operating scenario and in the development of control methods for major performance limiting instabilities. The important features were the understanding of particle transport and the control of impurity accumulation based on it, the satisfactory operation with predominantly tungsten-clad walls, the improved H-mode operation over density ranges and for temperature ratios covering (non-simultaneously) the ITER requirements on ν*, n/nGW and Te/Ti, the ELM frequency control by pellet injection and the optimization of NTM suppression by DC-ECCD through variation of the launching angle. From these experiments an integrated scenario has emerged which extrapolates to a 50% improvement in n T τ or a 30% reduction of the required current when compared with the ITER base-line assumptions, with moderately peaked electron and controllable high-Z density profiles.

Current diffusion, heat transport modelling, and linear gyrokinetic stability analysis have been carried out on a set of seven hybrid discharges from AUG, DIII-D, JET and JT-60U, in order to gain better understanding of the physics underlying this promising candidate scenario for ITER. Within this dataset, the GLF23 model has a higher accuracy than the Weiland model in predicting the temperature profiles in the region 0.3 < ρ < 0.8. The core heat transport appears to be similar between hybrid discharges and standard H-modes, and also among hybrid discharges with very different H factors. Projections to ITER show that Q = 10 can be obtained with the hybrid scenario using an alternative scaling without β degradation. However, additional off-axis current drive and current profile control might be needed for the ITER hybrid scenario, in order to achieve its full potential for high βN on extended duration.

A variety of techniques are being tested on ASDEX Upgrade for avoiding or reducing the divertor energy load caused by ELMs. Quiescent H-mode operation has been improved by operation with recently boronized vessel walls. Zeff values in QH-modes are now ~2.5, comparable with those found in ELMy H-modes of similar density. Type II ELM operation has been generated in discharges with ICRF additional heating alone, thus demonstrating its compatibility with reactor-relevant low core momentum and particle input. Furthermore, Type II ELM operation at low (

Magnetic triggering of edge localized modes (ELMs) was reported first from TCV in ohmic plasmas showing type-III ELMs. This method, showing successful locking of the ELM frequency to an imposed vertical plasma oscillation, has now also been demonstrated in the ITER-relevant type-I ELM regime in ASDEX Upgrade. Our experiments showed the ELM frequency becoming identical to the driving frequency in steady state for an applied motion of only about twice the value caused by an intrinsic ELM event. Triggered ELMs still showing clear type-I features were found when the plasma down-shift velocity reached its maximum, corresponding to the lowest edge current value. This is the opposite of the behaviour expected from the peeling–ballooning nature attributed to the ELM boundary and to TCV observations. The reason for this behaviour is not yet clear.

In a recent campaign with counter-current neutral beam injection (NBI), the quiescent high confinement mode (QH-mode) is studied in ASDEX Upgrade. The QH-mode regime is obtained with all accessible NBI angles, with and without additional ion cyclotron heating, at various values of edge safety factor q95 and at values of Zeff down to 2.5 after fresh boronization. In contrast to the ELMy H-mode, the pedestal pressure depends significantly on radial heat flux in the QH-mode, giving rise to a reduced heating power degradation of H-mode confinement. The 'edge harmonic oscillation', a characteristic MHD feature of the QH-mode that replaces ELMs, occurs at rational surfaces in the H-mode barrier region, and has so far been found only with a toroidal mode number n = 1. These conditions seem to define windows in q95 for favourable QH-mode access.

The ability to recognize the transition from the L-mode to the H-mode or from the H-mode to the improved H-mode reliably from a conveniently small number of measurements in real time is of increasing importance for machine control.Discriminant analysis has been applied to regime identification of plasma discharges in the ASDEX Upgrade tokamak. An observation consists of a set of plasma parameters averaged over a time slice in a discharge. The data set consists of all observations over different discharges and time slices. Discriminant analysis yields coefficients allowing the classification of a new observation. The results of a frequentist and a formal Bayesian approach to discriminant analysis are compared. With five plasma variables, a failure rate of 1.3% for predicting the L-mode and the H-mode confinement regime was achieved. With five plasma variables, a failure rate of 5.3% for predicting the H-mode and the improved H-mode confinement regime was achieved.The coefficients derived by discriminant analysis have been applied subsequently to discharges to illustrate the operation of regime identification in a real time control system.

As the ITER project moves forward in the procurement phase of major components, the baseline design is subject to a series of modifications due to the finalization of the design and interaction with industry. Some modifications have a potential impact on the performance and operation of the machine. A substantial activity is being carried out in EU in support to the design, aimed at assessing such impact. The effect of the 2010 design of the central solenoid on the operational space was assessed. A detailed analysis of the plasma breakdown and start-up phases was performed to assess the impact of withdrawing the booster converters from the baseline design and of the modifications to the power supply voltage ratings. In order to reach a clear conclusion, a review of the start-up strategy was also carried out starting from an analysis of the avalanche and breakdown conditions in ITER. The controllability of the null quality and position was verified in presence of uncertainties and measurement errors. The advantages of higher voltage rating were also quantified. General aim of the paper is to present an overview of the analyses performed and of the issues identified providing an assessment of the impact of some of the proposed changes on ITER operation. © 2011 European Joint Undertaking for ITER and the Development of Fusion for Energy. Published by Elsevier B.V. All rights reserved.

Recent progress in experimental and theoretical studies of edge localized mode (ELM) physics is reviewed for the reactor relevant plasma regimes, namely the high confinement regimes, that is, H-modes and advanced scenarios.Theoretical approaches to ELM physics, from a linear ideal magnetohydrodynamic (MHD) stability analysis to non-linear transport models with ELMs are discussed with respect to experimental observations, in particular the fast collapse of pedestal pressure profiles, magnetic measurements and scrape-off layer transport during ELMs.High confinement regimes with different types of ELMs are addressed in this paper in the context of development of operational scenarios for ITER. The key parameters that have been identified at present to reduce the energy losses in Type I ELMs are operation at high density, high edge magnetic shear and high triangularity. However, according to the present experimental scaling for the energy losses in Type I ELMs, the extrapolation of such regimes for ITER leads to unacceptably large heat loads on the divertor target plates exceeding the material limits. High confinement H-mode scenarios at high triangularity and high density with small ELMs (Type II), mixed regimes (Type II and Type I) and combined advanced regimes at high βp are discussed for present-day tokamaks. The optimum combination of high confinement and small MHD activity at the edge in Type II ELM scenarios is of interest to ITER. However, to date, these regimes have been achieved in a rather narrow operational window and far from ITER parameters in terms of collisionality, edge safety factor and βp.The compatibility of the alternative internal transport barrier (ITB) scenario with edge pedestal formation and ELMs is also addressed. Edge physics issues related to the possible combination of small benign ELMs (Type III, Type II ELMs, quiescent double barrier) and high performance ITBs are discussed for present-day experiments (JET, JT-60U, DIII-D) in terms of their relevance for ITER. Successful plasma edge control, at high triangularity (~0.5) and high density (~0.7nGR), in ITB scenarios in JET is reported.Active control of ELMs by edge current, pellet injection, impurities and external magnetic perturbations creating an ergodic zone localized at the separatrix are discussed for present-day experiments and from the perspective of future reactors.

ASDEX Upgrade has a large variety of technical systems, heating and fuelling as well as digital control systems needed for control and optimisation of high performance plasmas. The neutral beam injection (NBI), electron cyclotron RF (ECRF) and ion cyclotron resonance frequency (ICRF) heating systems are able to operate in heating as well as in current drive (CD) mode. One of two NBI systems has recently been redirected to allow more tangential injection to provide larger CD efficiency. With improved control procedures significant success is achieved in the optimisation of the conventional H-mode (ITER reference scenario) as well as of ‘advanced tokamak’ scenarios. The established operational boundaries encounter or exceed those of the reference scenario. H-modes showing type II ELMs with clearly reduced peak power loads on the target plates are produced at densities close to the empirical Greenwald density limit. Discharges with nearly full non-inductively driven plasma current at Greenwald densities are performed. MHD instabilities like sawteeth and neoclassical tearing modes (NTM) can be tailored and/or completely be stabilised via electron cyclotron frequency CD at relatively low injection power levels. In addition, several non-standard feedback control circuits for plasma performance improvements and machine protection circuits are devised.

ELM-free H-mode plasmas with stationary plasma density and radiation level are obtained in the ASDEX Upgrade tokamak with large clearance between the last closed flux surface and the wall, and neutral beam injection in a toroidal direction opposite to that of the plasma current. This behaviour is accompanied by a characteristic narrow-band magnetohydrodynamic (MHD) oscillation with clear harmonics up to n = 11 visible. This mode is localized in the plasma edge region. Conditions and properties of the stationary ELM-free phases and the edge MHD oscillation closely resemble that of the `quiescent H-mode' and the `edge harmonic oscillation' found in the DIII-D tokamak (Burrell K H et al 2002 Plasma Phys. Control. Fusion 44 A253). In addition, high-frequency MHD oscillations are found with an amplitude correlated with fluctuations of the divertor Dα intensity, suggesting a possible relevance of these modes for particle transport.

An advanced high confinement mode scenario with normalized beta (βN) up to 3.5 in stationary conditions (up to 20 energy confinement times) and βN = 3.8 transiently, has been obtained in ASDEX Upgrade. For transport analysis of high βN discharges, two types of simulations have been performed with the automated system for transport analysis code. One is the simulation of the current density profile with experimental data assuming neoclassical electrical conductivity. The other is the simulation of the energy transport and the current density profile using the Weiland transport model (ion temperature gradient and trapped electron mode limit the temperature gradient length). The comparison of the simulations of the q-profiles and temperature profiles with the experimental profiles shows good agreement and demonstrates that for this plasmas regime the temperature profiles are stiff (no internal transport barrier) as in conventional H-mode discharges. The contribution of the Ohmic, bootstrap and neutral beam driven current to the total current density profiles is calculated. The sum of the non-inductively driven plasma current is in the range 50–57% of the total plasma current. Simulations with the Weiland model are performed to investigate the influence of neoclassical tearing modes, which limit confinement and achievable βN in this scenario, by comparing with the measured kinetic data. In addition, the high βN discharges are compared to the improved H-mode discharges in ASDEX Upgrade with respect to the profile stiffness, MHD activity and experimental conditions.

For the first time, data from eight different tokamaks have been combined in an international database for internal transport barriers (ITBs). An analysis of the data for the formation of an ITB with dominant ion heating shows a clear dependence of the threshold power on the minor radius and line-averaged electron density for the formation of ion ITBs. The dependence of ITB formation on the toroidal magnetic field is weak. For the formation of ITBs with dominant electron heating, the database is smaller, but for the threshold power a strong increase with plasma size and a weak toroidal field dependence could also be identified. Based on these results, an expression for the power required to form an ITB is given using global variables only. These results give a basis for the analysis of the database using local values (like magnetic shear) and a detailed comparison with theory-based models.

The former all-carbon wall on JET has been replaced with beryllium in the main torus and tungsten in the divertor to mimic the surface materials envisaged for ITER. Comparisons are presented between Type I H-mode characteristics in each design by examining respective scans over deuterium fuelling and impurity seeding, required to ameliorate exhaust loads both in JET at full capability and in ITER., 55 pages, 15 figures

After the first high power D-T experiment in JET in 1997 (DTE1), when JET was equipped with Carbon PFC's, a proposed second high power (up to similar to 40MW) D-T campaign (DTE2) in the current Be/W vessel will address essential operational, technical, diagnostics and scientific issues in support of ITER. These experiments are proposed to minimize the risks to ITER by testing strategies for the management of the in-vessel tritium content, by providing the basis for transferring operational scenarios from non-active operation to D-T mixtures and by addressing the issue of the neutron measurement accuracy. Dedicated campaigns with operation in Deuterium, Hydrogen and Tritium before the D-T campaign proper will allow the investigation of isotope scaling of the H-mode transition, pedestal physics, heat, particle, momentum and impurity transport in much greater detail than was possible in DTE1. The D-T campaign proper will include validations of the baseline ELMy H-Mode scenario, of the hybrid H-mode and advanced tokamak scenarios, as well as the investigation of alpha particle physics and the qualification of ICRH scenarios suitable for D-T operation. This paper reviews the scientific goals of DTE2 together with a summary of the results of DTE1.

The eXtreme Shape Controller (XSC) has been originally designed to control the plasma shape at JET during the flat-top phase, when the plasma current has a constant value. During the JET 2012 experimental campaigns, the XSC has been used to improve the shape control during the transient phases of plasma current ramp-up and ramp-down. In order to avoid the saturation of the actuators with these transient phases, a current limit avoidance system has been designed and implemented. This paper presents the experimental results achieved at JET during the 2012 campaigns using the XSC. © 2013 Euratom: EURATOM-ENEA-CREATE/ Universita' di Napoli Federico II; European Atomic Energy Community.

Experiments in JET have concentrated on steady state discharges with internal transport barriers (ITBs). The ITBs are formed during the current rise phase of the discharge with low magnetic shear ( = r/q(dq/dr)) in the centre and with high additional heating power. In order to achieve stability against disruptions at high pressure peaking, which is typical for ITB discharges, the pressure profile can be broadened with an H mode transport barrier at the edge of the plasma. However, the strong increase in edge pressure during an ELM free H mode weakens the ITB owing to a reduction of the rotational shear and pressure gradient at the ITB location. In addition, type I ELM activity during the H mode phase leads to a collapse of the ITB with the input powers available in JET (up to 28 MW). The best ITB discharges are obtained with input power control to reduce the core pressure, and with the edge pressure of the plasma controlled by argon gas dosing. These discharges achieve steady conditions for several energy confinement times (τE) with H97 confinement enhancement factors (τE/τE,ITER97 scaling) of 1.2-1.6 at line averaged densities of around 30-40% of the Greenwald density. Increasing the density by using additional deuterium gas dosing or shallow pellet fuelling leads to a weakening of the ITB. In order to sustain ITBs at higher densities, type III ELMs should be maintained at the plasma edge, giving scope for future experiments in JET.

A reduction in turbulent density fluctuations is observed in internal transport barrier (ITB) discharges in the ASDEX Upgrade tokamak with an L-mode edge and reversed magnetic shear. The turbulence reduction is localized spatially to around the ITB gradient and the ITB foot-point region, coincident with strong shearing in the toroidal rotation velocity. The radial extent and degree of reduction is related to the strength of the ITB. Strong (narrow and steep) ITBs are well localized and display greater turbulence reduction. Weaker ITBs are less localized with less reduction. The formation of a subsequent H-mode edge barrier moves the velocity shearing region to the edge and weakens the ITB to the point of extinction. During the ITB the turbulence is reduced across a broad frequency range, however a residual level of low-frequency fluctuations (f

Optimized shear (OS) experiments in JET can generate an internal transport barrier (ITB) during a high power heating phase early in the plasma discharge. A strong link is generally observed between the formation of the barrier and the location of an integer q magnetic surface within a low magnetic shear (s = r/q(dq/dr)) region of the plasma. However, if the q-profile for such experiments is modified by applying lower hybrid heating and current drive (LHCD) before the main heating pulse, to provide a region of reduced or negative magnetic shear in the plasma core, ITBs can be formed the location of which does not exhibit any apparent association with a particular internal magnetic surface. Initial results suggest that q-profile modification using an LHCD prelude can also be used to reduce the heating power level required for ITB generation.

High performance discharges are routinely obtained on JET with low or reversed magnetic shear (s = (r/q)dq/dr), and the potential for steady state operation of such discharges is under investigation. With the use of the proper heating and fuelling, these `optimized shear' (OS) discharges exhibit an internal transport barrier (ITB), resulting in a strong peaking of the pressure profile, and thus in high fusion performance. These regimes have been extensively studied during the last (DD and DT) JET campaigns in order to promote this type of scenario as the basis for `advanced tokamak' operation. A review is given of the highest performance achieved on JET OS discharges during the last experimental campaigns, in both DD (up to 5.6 × 1016neutrons/s) and DT operation (fusion power up to 8.2 MW, ni0Ti0τE up to 1021 m-3 keV s). The role of the plasma edge is pointed out, as the power required to trigger an ITB is often higher than the H mode power threshold, leading to double barrier regimes. The presence of an H mode pedestal both modifies the ITB and induces edge bootstrap and ELM activity, which should be controlled to prolong such discharges. The operational procedure of optimization is then discussed, addressing the problems of ITB formation (power threshold, timing of the main heating phase, i.e. optimization of the target q profile, influence of the heating scheme, electron versus ion ITBs), ITB evolution (expansion of the ITB footpoint, H mode formation) and ITB termination (disruptive and/or `soft' MHD limits). Finally, the crucial problem of the route to steady state for such OS discharges is addressed, both in terms of ITB sustainment and control within the stability domain and in terms of edge pedestal control by means of impurity injection. The impurity behaviour is found, and examples of high performance discharges sustained for several energy confinement times are given (βN = 1.95, H89 = 2.3, Pfusioneq~10 MW, QDTeq~0.4 sustained for ~3 s). Extrapolation towards fully non-inductive current drive is discussed.

The main limitation to the performance of JET optimized shear (OS) discharges is due to MHD instabilities, mostly in the form of a disruptive limit. The structure of the MHD mode observed as a precursor to the disruption, as measured from SXR and ECE diagnostics, shows a global ideal MHD mode. The measured mode structure is in good agreement with the calculated mode structure of the pressure driven kink mode. The disruptions occur at relatively low normalized beta (1 < βN < 2), in good agreement with calculated ideal MHD stability limits for the n = 1 pressure driven kink mode. These low limits are mainly due to the extreme peaking factor of the pressure profiles. Other MHD instabilities observed in JET OS discharges include, usually benign, chirping modes. These modes, which occur in bursts during which the frequency changes, have the same mode structure as the disruption precursor but are driven unstable by fast particles.

High fusion performance has been achieved in optimized shear discharges in JET with the early appearance of internal transport barriers (ITBs) in the high power phase in both DD and DT plasmas. An ITB has been produced in a large variety of conditions, including LHCD only, ICRH only, NBI only and NBI + ICRH with LHCD preheat. An ITB may coexist with the edge barrier. The high pressure gradient produced by ICRH combined with high power NBI heating starting in a low density plasma with q(0) ≈ 1.5-2 is the main prerequisite for an efficient ITB in high performance discharges. Strong coupling between q profiles, MHD activity, energy confinement and fusion yield has been observed. The transition from L to H mode can be triggered by MHD events. The transition is accompanied by significant momentary changes in the density fluctuation spectrum in the peripheral plasma. The ITB may persist for some time in both ELM-free and ELMy H mode.

The unique roles played by ICRH in the preparation, formation and sustainment of internal transport barriers (ITBs) in high fusion performance JET optimized shear experiments using the Mark II poloidal divertor are discussed. Together with LHCD, low power ICRH is applied during the early ramp-up phase of the plasma current, `freezing in' a hollow or flat current density profile with q(0)>1. In combination with up to ~20 MW of NBI, the ICRH power is stepped up to ~6 MW during the main low confinement (L mode) heating phase. An ITB forms promptly after the power step, revealed by a region of reduced central energy transport and peaked profiles, with the ion thermal diffusivity falling to values close to the standard neoclassical level near the centre of both DD and DT plasmas. At the critical time of ITB formation, the plasma contains an energetic ICRF supported hydrogen minority ion population, contributing ~50% to the total plasma pressure and heating mainly electrons. As both the NBI population and the thermal ion pressure develop, a substantial part of the ICRF power is damped resonantly on core ions (ω = 2 ωcD = 3ωcT), contributing to the ion heating. In NBI step-down experiments, high performance has been sustained by maintaining central ICRH; analysis shows the efficiency of central ICRH ion heating to be comparable to that of NBI. The highest DD fusion neutron rates (RNT = 5.6 × 1016 s-1) yet achieved in JET plasmas have been produced by combining a low magnetic shear core with a high confinement (H mode) edge. In DT, a fusion triple product niTiτE = (1.2 ± 0.2) × 1021 m-3 keV s was achieved with 7.2 MW of fusion power obtained in the L mode and with up to 8.2 MW of fusion power in the H mode phase.

The scaling of the energy confinement in H mode plasmas with different hydrogenic isotopes (hydrogen, deuterium, DT and tritium) is investigated in JET. For ELM-free H modes the thermal energy confinement time τth is found to decrease weakly with the isotope mass (τth ~M-0.25±0.22), whilst in ELMy H modes the energy confinement time shows practically no mass dependence (τth ~M0.03±0.1). Detailed local transport analysis of the ELMy H mode plasmas reveals that the confinement in the edge region increases strongly with the isotope mass, whereas the confinement in the core region decreases with mass (τthcore ∝ M-0.16), in approximate agreement with theoretical models of the gyro-Bohm type (τgB ~M-0.2).

A number of experiments with heating of DT plasmas using ICRF waves have been carried out at JET. The results of these experiments have been analysed by comparing experimentally measured quantities with the results of numerical simulations. In particular, four scenarios have been examined: (a) heating of minority (~5-20%) deuterons at the fundamental ion cyclotron frequency, ω = ωcD; (b) second harmonic heating of tritium, ω = 2ωcT; (c) fundamental minority heating of 3He with a few per cent of 3He; (d) second harmonic heating of deuterium, ω = 2ωcD. An important aim of the analysis was to assess whether the present understanding of the ICRF physics is adequate for predicting the performance of ICRF in DT plasmas. In general, good agreement between experimental results and simulations was found which increases the confidence in predictions of the impact of ICRF heating in future reactors. However, when a relatively high deuterium concentration was used in the ω = ωcD scenario, discrepancies were observed. In order to increase confidence in the simulations, the sensitivity of the simulation results to various plasma parameters has been studied.

Results are presented from a series of dedicated experiments carried out on JET in tritium, DT, deuterium and hydrogen plasmas to determine the dependence of the H mode power threshold on the plasma isotopic mass. The Pthr ∝ Aeff-1 scaling is established over the whole isotopic range. This result makes it possible for a fusion reactor with a 50:50 DT mixture to access the H mode regime with about 20% less power than that needed in a DD mixture. Results on the first systematic measurements of the power necessary for the transition of the plasma to the type I ELM regime, which occurs after the transition to H mode, are also in agreement with the Aeff-1 scaling. For a subset of discharges, measurements of Te and Ti at the top of the profile pedestal have been obtained, indicating a weak influence of the isotopic mass on the critical edge temperature thought to be necessary for the H mode transition.

Steady state high performance with improved core confinement and sustainable plasma edge conditions has been approached on JET in a double barrier (DB) mode. The DB mode combines an internal transport barrier of the optimized shear regime with an edge transport barrier of an ELMy H?mode regime. Improved confinement with an H?factor HITER-89 ? 2 has been maintained for four energy confinement times. Ion and electron temperature profiles remain peaked in the DB mode, while the density profile is broad and similar in shape to the conventional ELMy H?mode profile. The energy confinement improves across the whole plasma cross-section, and the ion heat conductivity falls to the neoclassical level in the core. Particle transport studies show that impurity accumulation can be avoided in the DB mode. In DT discharges the DB mode has attained a fusion gain of Q ? 0.4, producing 6.8?MW of fusion power, compared with Q ? 0.2 in the conventional sawtoothing ELMy H?mode. The ELMs are more benign, with an amplitude an order of magnitude smaller in the DB mode. The DB mode has a high potential to improve performance in reactor relevant conditions.

Reactor relevant ICRH scenarios have been assessed during DT experiments on the JET tokamak using H mode divertor discharges with ITER-like shapes and safety factors. Deuterium minority heating in tritium plasmas was demonstrated for the first time. For 9% deuterium, an ICRH power of 6 MW gave 1.66 MW of fusion power from reactions between suprathermal deuterons and thermal tritons. The Q value of the steady state discharge reached 0.22 for the length of the RF flat-top (2.7 s), corresponding to three plasma energy replacement times. The Doppler broadened neutron spectrum showed a deuteron energy of 125 keV, which was optimum for fusion and close to the critical energy. Thus, strong bulk ion heating was obtained at the same time as high fusion efficiency. Deuterium fractions around 20% produced the strongest ion heating together with a strong reduction of the suprathermal deuteron tail. The ELMs had low amplitude and high frequency and each ELM transported less plasma energy content than the 1% required by ITER. The energy confinement time, on the ITERH97-P scale, was 0.90, which is sufficient for ignition in ITER. 3He minority heating, in approximately 50:50 D:T plasmas with up to 10% 3He, also demonstrated strong bulk ion heating. Central ion temperatures up to 13 keV were achieved, together with central electron temperatures up to 12 keV. The normalized H mode confinement time was 0.95. Second harmonic tritium heating produced energetic tritons above the critical energy. This scheme heats the electrons in JET, unlike in ITER where the lower power density will allow mainly ion heating. The inverted scenario of tritium minority ICRH in a deuterium plasma was demonstrated as a successful heating method producing both suprathermal neutrons and bulk ion heating. Theoretical calculations of the DT reactivity mostly give excellent agreement with the measured reaction rates.

On the basis of experiments carried out in JET in D:T mixtures varying from 100:0 to 5:95 and those carried out in hydrogen plasmas, the isotopic mass dependence of ELM parameters and the edge pedestal pressure in NBI and ICRF heated H mode plasmas is presented. The ELM frequency is found to decrease with the atomic mass number in both ICRH and NBI discharges. However, the frequency in the case of ICRH is about 8-10 times higher than that in the case of NBI. Assuming that ELMs occur at a critical edge pressure gradient, limited by the ballooning instability, the scaling of the maximum edge pressure is most consistent with the assumption that the width of the transport barrier scales as the ion poloidal Larmor radius governed by the average energy of fast ions at the edge. The critical edge pressure in NBI heated discharges increases with the isotopic mass, which is consistent with the higher deduced width of the edge transport barrier in tritium than those in deuterium and hydrogen. The critical edge pressure in ICRH discharges is smaller, presumably, due to the smaller fast ion contribution to the edge region. As a consequence of the edge pressure scaling with isotopic mass, the edge operational space in an ne-Te diagram increases with operation in tritium. If the evidence that the edge pedestal width is governed by the average energy of fast ions in the edge prevails, the pedestal in ITER would be controlled by the slowing down energy spectrum of alpha particles in the edge.