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Engineering Design Evolution of the JT-60SA Project
EngineeringEngineering DesignDesign EvolutionEvolution ofof thethe JTJT --60SA60SA ProjectProject P. Barabaschi, Y. Kamada, S. Ishida for the JT-60SA Integrated Project Team EU: F4E-CEA-ENEA-CNR/RFX-KIT-CRPP-CIEMAT-SCKCEN JA: JAEA 1 JTJT --60SA60SA ObjectivesObjectives •A combined project of the ITER Satellite Tokamak Program of JA-EU (Broader Approach) and National Centralized Tokamak Program in Japan. •Contribute to the early realization of fusion energy by its exploitation to support the exploitation of ITER and research towards DEMO. ITER DEMO Complement Support ITER ITER towards DEMO JT-60SA 2 TheThe NewNew LoadLoad AssemblyAssembly • JT60-U: Copper Coils (1600 T), Ip=4MA, Vp=80m 3 • JT60-SA: SC Coils (400 T), Ip=5.5MA, Vp=135m 3 JTJTJT-JT ---60U60U JTJTJT-JT ---60SA60SA JT-60SA(A ≥2.5,Ip=5.5 MA) ITER (A=3.1,15 MA) 3.0m 6.2m ~2.5m ~4m 1.8m KSTAR (A=3.6, 2 MA) 1.7m EAST (A=4.25,1 MA) 1.1m 3 SST-1 (A=5.5, 0.22 MA) 3 HighHigh BetaBeta andand LongLong PulsePulse • JT-60SA is a fully superconducting tokamak capable of confining break- max even equivalent class high-temperature deuterium plasmas ( Ip =5.5 MA ) lasting for a duration ( typically 100s ) longer than the timescales characterizing the key plasma processes, such as current diffusion and particle recycling. • JT-60SA should pursue full non-inductive steady-state operations with high βN (> no-wall ideal MHD stability limits). 4 4 PlasmaPlasma ShapingShaping • JT-60SA will explore the plasma configuration optimization for ITER and DEMO with a wide range of the plasma shape including the shape of ITER , with the capability to produce both single and double null configurations. -
Overview of Versatile Experiment Spherical Torus (VEST)
Overview of Versatile Experiment Spherical Torus (VEST) Y.S. Hwang and VEST team March 29, 2017 CARFRE and CATS Dept. of Nuclear Engineering Seoul National University 23rd IAEA Technical MeetingExperimental on Researchresults and plans of Using VEST Small Fusion Devices, Santiago, Chille Outline Versatile Experiment Spherical Torus (VEST) . Device and discharge status Start-up experiments . Low loop voltage start-up using trapped particle configuration . EC/EBW heating for pre-ionization . DC helicity injection Studies for Advanced Tokamak . Research directions for high-beta and high-bootstrap STs . Preparation of long-pulse ohmic discharges . Preparation for heating and current drive systems . Preparation of profile diagnostics Long-term Research Plans Summary 1/36 VEST device and Machine status VEST device and Machine status 2/36 VEST (Versatile Experiment Spherical Torus) Objectives . Basic research on a compact, high- ST (Spherical Torus) with elongated chamber in partial solenoid configuration . Study on innovative start-up, non-inductive H&CD, high and innovative divertor concept, etc Specifications Present Future Toroidal B Field [T] 0.1 0.3 Major Radius [m] 0.45 0.4 Minor Radius [m] 0.33 0.3 Aspect Ratio >1.36 >1.33 Plasma Current [kA] ~100 kA 300 Elongation ~1.6 ~2 Safety factor, qa ~6 ~5 3/36 History of VEST Discharges • #2946: First plasma (Jan. 2013) • #10508: Hydrogen glow discharge cleaning (Nov. 2014) Ip of ~70 kA with duration of ~10 ms • #14945: Boronization with He GDC (Mar. 2016) Maximum Ip of ~100 kA • # ??: -
Past, Present and Future of the US-KSTAR Collaboration Hyeon K
Past, Present and Future of the US-KSTAR Collaboration Hyeon K. Park POSTECH at 2009 US-KSTAR Workshop GA.SanDiego, CA On April 15-16, 2009 Past of the international fusion effort • Three large tokamak era: non-steady state device based on Cu coils (pulse length is limited by the cooling system < ~ 20 sec.) – Tokamak Fusion Test Reactor (USA) 1982-1997, Princeton Plasma Physics Laboratory, USA ¾ Fusion power yield: Q ~ 0.3 from D-T experiment – Joint European Tokamak (EU):1983 – present, Culham, Oxfordshore, UK ¾ Fusion power yield: Q ~ 0.7 from D-T experiment – JT-60U (Japan):1985 - present, Japan Atomic Energy Agency (JAEA), Japan ¾ Q~1.25 extrapolated from D-D experiment Internal view of Internal view of TFTR Internal view of JT60-U JET/plasma discharge Future (ITER) • The goal is "to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes". – Demonstration of fusion power yield; Q (output power/input power) ~10 – International consortium (Europe, Japan, USA, Russia, Korea, China, and India) – Total cost ~ $10 B for ~10 years Physics basis is empirical energy confinement scaling KSTAR-US collaboration (past) • US-KSTAR workshop at GA on May 2004 • Areas of interest (first priority of KSTAR)~$1.62M – Steady-state Technology ($14.3M) ($0.37M) – Control, Stability and AT modes($1.48M) ($0.35M) – Conventional Diagnostics ($ 6.5M) ($0.45M) – Advanced Diagnostics ($2.8M) ($0.2M) – Collaboratory ($1.25M)($0.25M) • US-KSTAR workshop 2005 at Daejeon (active work) • US-KSTAR workshop 2006 at Princeton – FY06 International Collaboration to prepare for KSTAR operation (Finals 04/06; $1352K total) – Total allocation for Institution: PPPL (~500k), GA (~400k), ORNL (~200k), LLNL (~20k), MIT(~40k) Columbia (~100k), Escalated KSTAR progress • US-KSTAR workshop (Sept, 2007), Dajeon, Korea – Korean National Assembly passed the Fusion Energy Developm ent Promotion Act on November 30, 2006. -
Alfvén-Character Oscillations in Ohmic Plasmas Observed on the COMPASS Tokamak
44th EPS Conference on Plasma Physics P5.140 Alfvén-character oscillations in ohmic plasmas observed on the COMPASS tokamak T. Markovicˇ1;2, A. Melnikov3;4, J. Seidl1, L. Eliseev3, J. Havlicek1, A. Havránek1;5, M. Hron1, M. Imríšek1;2, K. Kovaríkˇ 1;2, K. Mitošinková1;2, J. Mlynar1, R. Pánek1, J. Stockel1, J. Varju1, V. Weinzettl1, the COMPASS Team1 1 Institute of Plasma Physics of the CAS, Prague, Czech Republic 2 Faculty of Mathematics and Physics, Charles Uni. in Prague, Prague, Czech Republic 3 National Research Centre ’Kurchatov Institute’, Moscow, Russian Federation 4 National Research Nuclear University MEPhi, Moscow, Russian Federation 5 Faculty of Electrical Engineering, Czech Tech. Uni. in Prague, Prague, Czech Republic Energetic Particle (EP) driven plasma modes resulting from interaction of shear Alfvén waves with fast (i.e. comparable to plasma Alfvén velocity VA) ions generated by fu- sion reactions, NBI or ICRH heating [1] are capable of causing fast particle losses, hence negatively affecting the plasma per- formance or having detrimental effects on plasma-facing components or vacuum vessel [2]. A number of comprehensive reviews has already been published, describing properties of EP modes, their appearance, excitation and damping mechanisms, etc. (e.g. [1, 2, 3]). The modes are not trivially expected to appear in ohmic plasmas, however, plasma oscillations bearing their typical signatures have been re- ported in plasmas without an apparent source of EP on a number of devices (such as TFTR [5], ASDEX-U [6], MAST [7], TUMAN-3M Figure 1: High-frequency magnetic fluctuations. [8]). While impact of these specific phenom- fAE curve from eq. -
TH/P8-21 Transport Simulations of KSTAR Advanced Tokamak
1 TH/P8-21 Transport Simulations of KSTAR Advanced Tokamak Scenarios Yong-Su Na 1), 2), C. E. Kessel 3), J. M. Park 4), J. Y. Kim 1) 1) National Fusion Research Institute, Daejeon, Korea 2) Department of Nuclear Engineering, Seoul National University, Seoul, Korea 3) Princeton Plasma Physics Laboratory, Princeton, NJ, USA 4) Oak Ridge National Laboratory, Oak Ridge, TN, USA e-mail contact of main author: [email protected] Abstract. Predictive modeling of KSTAR operation scenarios are performed with the aim of developing high performance steady state operation scenarios. Various transport codes are employed for this study. Firstly, steady state operation capabilities are investigated with time dependent simulations using a free-boundary transport code. Secondly, reproducibility of high performance steady state operation scenario from an existing tokamak to KSTAR is investigated using the experimental data from other tokamak device. Finally, capability of DEMO-relevant advanced tokamak operation is investigated in KSTAR. From those simulations, it is found that KSTAR is able to establish high performance steady state operation scenarios. The selection of the transport model and the current ramp up scenario is also discussed which have strong influence on target profiles. 1. Introduction As the fusion era is rapidly approaching, the necessity of development of steady state operation scenarios becomes more and more important, particularly for fusion reactor models based on the tokamak concept. In addition to the steady state operation, fusion performance of the tokamak needs to be improved compared with conventional H-modes for developing economically viable fusion power plants. In this context, the, so-called, advanced tokamak (AT) scenarios are being developed aiming at satisfying these two reactor requirements simultaneously. -
Losses of Runaway Electrons in MHD-Active Plasmas of the COMPASS Tokamak
Losses of runaway electrons in MHD-active plasmas of the COMPASS tokamak O Ficker1;2, J Mlynar1, M Vlainic1;2;3, J Cerovsky1;2, J Urban1, P Vondracek1;4, V Weinzettl1, E Macusova1, J Decker5,M Gospodarczyk5, P Martin7, E Nardon8, G Papp9,VV Plyusnin10, C Reux8, F Saint-Laurent8, C Sommariva8,J Cavalier1;11, J Havlicek1, A Havranek1;12, O Hronova1,M Imrisek1, T Markovic1;4, J Varju1, R Paprok1;4, R Panek1,M Hron1 and the COMPASS team 1 Institute of Plasma Physics of the CAS, CZ-18200 Praha 8, Czech Republic 2 FNSPE, Czech Technical University in Prague, CZ-11519 Praha 1, Czech Republic 3 Department of Applied Physics, Ghent University, 9000 Ghent, Belgium 4 FMP, Charles University, Ke Karlovu 3, CZ-12116 Praha 2, Czech Republic 5 Swiss Plasma Centre, EPFL, CH-1015 Lausanne, Switzerland 6 Universita’ di Roma Tor Vergata, 00133 Roma, Italy 7 Consorzio RFX, Corso Stati Uniti 4, 35127 Padova, Italy 8 CEA, IRFM, F-13108 Saint-Paul-lez-Durance, France 9 Max Planck Institute for Plasma Physics, Garching D-85748, Germany 10 Centro de Fusao Nuclear, IST, Lisbon, Portugal 11 Institut Jean Lamour IJL, Universite de Lorraine, 54000 Nancy, France 12 FEE, Czech Technical University in Prague, CZ-12000 Praha 2, Czech Republic E-mail: [email protected] Abstract. Significant role of magnetic perturbations in mitigation and losses of Runaway Electrons (REs) was documented in dedicated experimental studies of RE at the COMPASS tokamak. RE in COMPASS are produced both in low density quiescent discharges and in disruptions triggered by massive gas injection (MGI). -
Modeling and Control of Plasma Rotation for NSTX Using
PAPER Related content - Central safety factor and N control on Modeling and control of plasma rotation for NSTX NSTX-U via beam power and plasma boundary shape modification, using TRANSP for closed loop simulations using neoclassical toroidal viscosity and neutral M.D. Boyer, R. Andre, D.A. Gates et al. beam injection - Topical Review M S Chu and M Okabayashi To cite this article: I.R. Goumiri et al 2016 Nucl. Fusion 56 036023 - Rotation and momentum transport in tokamaks and helical systems K. Ida and J.E. Rice View the article online for updates and enhancements. Recent citations - Design and simulation of the snowflake divertor control for NSTX–U P J Vail et al - Real-time capable modeling of neutral beam injection on NSTX-U using neural networks M.D. Boyer et al - Resistive wall mode physics and control challenges in JT-60SA high scenarios L. Pigatto et al This content was downloaded from IP address 128.112.165.144 on 09/09/2019 at 19:26 IOP Nuclear Fusion International Atomic Energy Agency Nuclear Fusion Nucl. Fusion Nucl. Fusion 56 (2016) 036023 (14pp) doi:10.1088/0029-5515/56/3/036023 56 Modeling and control of plasma rotation for 2016 NSTX using neoclassical toroidal viscosity © 2016 IAEA, Vienna and neutral beam injection NUFUAU I.R. Goumiri1, C.W. Rowley1, S.A. Sabbagh2, D.A. Gates3, S.P. Gerhardt3, M.D. Boyer3, R. Andre3, E. Kolemen3 and K. Taira4 036023 1 Department of Mechanical and Aerospace Engineering, Princeton University, Princeton, NJ 08544, USA I.R. Goumiri et al 2 Department of Applied Physics and Applied Mathematics, -
Lyra' Divertor
ENERGY AND PARTICLE CONTROL CHARACTERISTICS OF THE ASDEX UPGRADE `LYRA' DIVERTOR M. Kaufmann, H-S. Bosch, A. Herrmann, A. Kallenbach, K. Borrass, A. Carlson, D. Coster, J.C. Fuchs, J. Gafert, K. Lackner, J. Neuhauser, R. Schneider, J. Schweinzer, W. Suttrop, W. Ullrich, U. Wenzel, and ASDEX Upgrade team Max-Planck-Institut fÈurPlasmaphysik, EURATOM-IPP Association, Garching und Berlin, Germany Abstract In 1997 the new `LYRA' divertor went into operation at ASDEX Upgrade and the neutral beam heating power was increased to 20 MW by installation of a second injector. This leads to the relatively high value of P/R of 12 MW/m. It has been shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. Mea- surements presented in this paper reveal a reduction of the maximum heat ¯ux in the LYRA divertor by more than a factor of two compared to the open Divertor I. This reduction is caused by radiative losses inside the divertor region. Carbon radiation cools the divertor plasma down to a few eV where hydrogen radiation losses become signi®cant. They are increased due to an effective re¯ection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental ®ndings and re®nes the understanding of loss processes in the divertor region. 1. INTRODUCTION The width of the scrape-off layer (SOL) does not necessarily increase in proportion to the size of the device. This poses severe problems for the power exhaust in a fusion reactor. If we take ITER as described in the ®nal design report (FDR) [1], a power ¯ow across the separatrix in the order of 100 to 150 MW might be needed to stay in the H-mode [2]. -
Advanced Tokamak Experiments in Full-W ASDEX Upgrade
1 PDP-4 Advanced tokamak experiments in full-W ASDEX Upgrade J. Stober, A. Bock, E. Fable, R. Fischer, C. Angioni, V. Bobkov, A. Burckhart, H. Doerk, A. Herrmann, J. Hobirk, Ch. Hopf, A. Kallenbach, A. Mlynek, R. Neu, Th. Putterich,¨ M. Reich, D. Rittich, M. Schubert, D. Wagner, H. Zohm and the ASDEX Upgrade Team Max-Planck-Institut fur¨ Plasmaphysik, Boltzmannstr. 2, 85748 Garching, Germany email: [email protected] Abstract. Full W-coating of the plasma facing components of ASDEX Upgrade (AUG) in 2007 ini- tially prevented H-mode operation with little or no gas puff, due to collapses driven by W accumulation, also due to a lack of adequate central wave heating. Additionally, the limits for power loads to the di- vertor were reduced. Under these conditions low density, high beta plasmas with significant external current drive were virtually impossible to operate, i.e. the operational window for advanced tokamak (AT) research was effectively closed. To recover this operational window several major achievements were necessary: the understanding of the changes in pumping and recycling, the change to a full W divertor capable to handle higher power loads, a substantial increase of the ECRH/CD power and pulse length to avoid W accumulation and to adjust the current profile and, finally, new or upgraded diagnostics for the current profile, since the original MSE system turned out to be heavily disturbed by polarizing re- flections on W-coated surfaces. With these efforts it was possible to recover improved H-mode operation as well as to develop a new scenario for fully non-inductive operation at q95 = 5:3, bN of 2.7 and 50% of bootstrap current. -
Snowflake Divertor Studies in DIII-D and NSTX Aimed at the Power
40th EPS Conference on Plasma Physics (EPS 2013) Europhysics Conference Abstracts Vol. 37D Espoo, Finland 1 - 5 July 2013 Part 1 of 2 ISBN: 978-1-63266-310-8 Printed from e-media with permission by: Curran Associates, Inc. 57 Morehouse Lane Red Hook, NY 12571 Some format issues inherent in the e-media version may also appear in this print version. Copyright© (2013) by the European Physical Society (EPS) All rights reserved. Printed by Curran Associates, Inc. (2014) For permission requests, please contact the European Physical Society (EPS) at the address below. European Physical Society (EPS) 6 Rue des Freres Lumoere F-68060 Mulhouse Cedex France Phone: 33 389 32 94 40 Fax: 33 389 32 94 49 [email protected] Additional copies of this publication are available from: Curran Associates, Inc. 57 Morehouse Lane Red Hook, NY 12571 USA Phone: 845-758-0400 Fax: 845-758-2634 Email: [email protected] Web: www.proceedings.com 40th EPS Conference on Plasma Physics 1 - 5 July 2013 Espoo, Finland Snowflake Divertor Studies in O2.101 Soukhanovskii, V.A. DIII-D and NSTX Aimed at the Power Exhaust Solution for the Tokamak Lang, P.T., Bernert, M., Burckhart, A., Casali, L., Fischer, R., Pellet as tool for high density O2.102 Kardaun, O., Kocsis, G., Maraschek, M., Mlynek, A., Ploeckl, B., operation and ELM control in Reich, M., Francois, R., Schweinzer, J., Sieglin, B., Suttrop, W., ASDEX Upgrade Szepesi, T., Tardini, G., Wolfrum, E., Zohm, H., Team, A. Modelling of the O2.103 Panayotis, S. erosion/deposition pattern on the Tore Supra Toroidal Pumped Limiter Non-inductive Plasma Current Start-up in NSTX Raman, R., Jarboe, T.R., Jardin, S.C., Kessel, C.E., Mueller, D., using Transient CHI and O2.104 Nelson, B.A., Poli, F., Gerhardt, S., Kaye, S.M., Menard, J.E., Ono, subsequent Non-inductive M., Soukhanovskii, V. -
June 2018 Fusion in Europe
FUSION IN EUROPE NEWS & VIEWS ON THE PROGRESS OF FUSION RESEARCH “Let us face it: TO DTT OR NOT TO DTT there is no VACUUM – HOW NOTHING REALLY planet B!” MATTERS NOW IS THE TIME TO BE AT JET 2 2018 Fusion Writers … wanted! … and Artists This could be you Fusion in Europe is call- ing for aspiring writers and gifted artists! Introduce yourself to an international audience! Catch one of our topics and turn it into your own! • The future powered by fusion energy • Fusion science and industrial reality • Fusion - a melting pot of different sciences • Fusion drives innovation Find the entire list here: tinyurl.com/ybd4omz7 Your application should include a short CV and a motivational letter. Please apply here: tinyurl.com/ybd4omz7 EUROfusion | Communications Team | Anne Purschwitz (”Fusion in Europe“ Editor) Boltzmannstr. 2 | 85748Application Garching | +49 89 deadline: 3299 4128 | [email protected] 25 June 2018 | Editorial | EUROfusion | “It is the best time to be at JET right now” says a passionate Eva Belo nohy. The member of JET’s Exploitation Unit has recently organised a very suc- cessful workshop. It was aimed to ‘refresh’ the knowledge of European fu- sion researchers regarding the Joint European Torus (JET) but that was a classic understatement. The meeting was a fully-fledged overview on JET’s capabilities which have tremendously changed in the past. The tokamak has gone through major upgrades since it saw its first deuterium tritium campaign in 1997. Those include not only an ITER-like wall but also an Fusion Writers… wanted! … increase of the heating power by 50 percent. -
Disruption Mitigation Studies at ASDEX Upgrade in Support of ITER
Disruption mitigation studies at ASDEX Upgrade in support of ITER G. Pautasso1, M. Bernert1, M. Dibon1, B. Duval2, R. Dux1, E. Fable1, C. Fuchs1, G.D. Conway1, L. Giannone1, A. Gude1, A. Herrmann1, M. Hoelzl1, P.J. McCarthy3, A. Mlynek1, M. Maraschek1, E. Nardon4, G. Papp1, S. Potzel1, C. Rapson1, B. Sieglin1, W. Suttrop1, W. Treutterer1, the ASDEX Upgrade1 and the EUROfusion MST15 teams 1 Max-Plank-Institute fuer Plasma Physik, D-85748, Garching, Germany 2 Swiss Plasma Centre, EPFL, CH-1016 Lausanne, Switzerland 3 Department of Physics, University College Cork, Ireland 4 CEA, IRFM, F-13108 Saint Paul lez Durance, France 5 http://www.euro-fusionscipub.org/mst1 IEA Workshop on Disruptions, July 20-22 2016, PPPL, USA 1 Content Experimental scenarios and rational behind interpretation of pre-thermal quench force mitigation in MGI(*) induced plasma termination; focus on small gas quantities thermal load mitigation runaway electron generation and losses; focus on MGI suppression (*) MGI = massive gas injection IEA Workshop on Disruptions, July 20-22 2016, PPPL, USA 2 Background 22 3 2008-2013: MGI exp.s in AUG aimed at reaching ne ~ nc ~ O (10 ) / m during or just after TQ for RE suppression poor impurity assimilation at large Ninj → attempts to reach nc abandoned ITER DMS consists now of several injectors for TQ & force mitigation + RE suppression TQ: Minimum impurity amount for force & thermal load mitigation? CQ: Is control and/or suppression of REs possible? IEA Workshop on Disruptions, July 20-22 2016, PPPL, USA 3 AUG: Mitigation valves,