Advanced divertor magnetic configurations for tokamaks: concepts, status, future. Vlad Soukhanovskii Edge Coordinating Committee Fall 2014 Technical Meeting.

Slides:



Advertisements
Similar presentations
The PPPL Perspective on Ten Year Planning S. Prager Princeton Plasma Physics Laboratory.
Advertisements

Introduction to Plasma-Surface Interactions Lecture 6 Divertors.
ASIPP HT-7 belt limiter Houyang Guo, Sizhen Zhu and Jiangang Li Investigation of EAST Divertor Asymmetry in Plasma Detachment & Target Power Loading Using.
ASIPP Characteristics of edge localized modes in the superconducting tokamak EAST M. Jiang Institute of Plasma Physics Chinese Academy of Sciences The.
6th Japan Korea workshop July 2011, NIFS, Toki-city Japan Edge impurity transport study in stochastic layer of LHD and scrape-off layer of HL-2A.
First Wall Heat Loads Mike Ulrickson November 15, 2014.
PhD studies report: "FUSION energy: basic principles, equipment and materials" Birutė Bobrovaitė; Supervisor dr. Liudas Pranevičius.
Introduction to Spherical Tokamak
Implications of Plasma-Material Interactions Dennis Whyte, MIT PSFC & PSI Science Center With contributions from Jeff Brooks (Purdue), Russ Doerner (UCSD),
Physics of fusion power Lecture 14: Anomalous transport / ITER.
Steps Toward a Compact Stellarator Reactor Hutch Neilson Princeton Plasma Physics Laboratory ARIES Team Meeting October 3, 2002.
V. A. SOUKHANOVSKII, POSTER EX/P4-22, 22 ND IAEA FEC, OCTOBER 2008, GENEVA, SWITZERLAND 1 of 22 Divertor Heat Flux Mitigation in High- Performance.
Physics of fusion power
Integrated Effects of Disruptions and ELMs on Divertor and Nearby Components Valeryi Sizyuk Ahmed Hassanein School of Nuclear Engineering Center for Materials.
Physics of fusion power Lecture 8 : The tokamak continued.
Power Extraction Research Using a Full Fusion Nuclear Environment G. L. Yoder, Jr. Y. K. M. Peng Oak Ridge National Laboratory Oak Ridge, TN Presentation.
A. HerrmannITPA - Toronto /19 Filaments in the SOL and their impact to the first wall EURATOM - IPP Association, Garching, Germany A. Herrmann,
Y. Sakamoto JAEA Japan-US Workshop on Fusion Power Plants and Related Technologies with participations from China and Korea February 26-28, 2013 at Kyoto.
Recent JET Experiments and Science Issues Jim Strachan PPPL Students seminar Feb. 14, 2005 JET is presently world’s largest tokamak, being ½ linear dimension.
Nils P. Basse Plasma Science and Fusion Center Massachusetts Institute of Technology Cambridge, MA USA ABB seminar November 7th, 2005 Measurements.
Initial wave-field measurements in the Material Diagnostic Facility (MDF) Introduction : The Plasma Research Laboratory at the Australian National University.
Advanced Tokamak Plasmas and the Fusion Ignition Research Experiment Charles Kessel Princeton Plasma Physics Laboratory Spring APS, Philadelphia, 4/5/2003.
pkm- NCSX CDR, 5/21-23/ Power and Particle Handling in NCSX Peter Mioduszewski 1 for the NCSX Boundary Group: for the NCSX Boundary Group: M. Fenstermacher.
ASIPP EAST Overview Of The EAST In Vessel Components Upgraded Presented by Damao Yao.
Divertor/SOL contribution IEA/ITPA meeting Naka Nov. 23, 2003 Status and proposals of IEA-LT/ITPA collaboration Multi-machine Experiments Presented by.
Simulation Study on behaviors of a detachment front in a divertor plasma: roles of the cross-field transport Makoto Nakamura Prof. Y. Ogawa, S. Togo, M.
V. A. Soukhanovskii 1 Acknowledgements: M. G. Bell 2, R. Kaita 2, H. W. Kugel 2, R. Raman 3, A. L. Roquemore 2 1 Lawrence Livermore National Laboratory,
1 Modeling of EAST Divertor S. Zhu Institute of Plasma Physics, Chinese Academy of Sciences.
Divertor heat flux mitigation with impurity-seeded standard and snowflake divertors in NSTX. V. A. Soukhanovskii, R. E. Bell, A. Diallo, S. Gerhardt, R.
Taming the Plasma-Material Interface with the Snowflake Divertor in NSTX V. A. Soukhanovskii Lawrence Livermore National Laboratory JI nd Annual.
Managed by UT-Battelle for the Department of Energy Stan Milora, ORNL Director Virtual Laboratory for Technology 20 th ANS Topical Meeting on the Technology.
O. Sauter Effects of plasma shaping on MHD and electron heat conductivity; impact on alpha electron heating O. Sauter for the TCV team Ecole Polytechnique.
V. A. Soukhanovskii NSTX Team XP Review 31 January 2006 Princeton, NJ Supported by Office of Science Divertor heat flux reduction and detachment in lower.
Physics of fusion power Lecture 10: tokamak – continued.
V. A. Soukhanovskii 1 Acknowledgement s: R. Maingi 2, D. A. Gates 3, J. Menard 3, R. Raman 4, R. E. Bell 3, C. E. Bush 2, R. Kaita 3, H. W. Kugel 3, B.
The snowflake divertor: a game-changer for magnetic fusion devices ? V. A. Soukhanovskii Lawrence Livermore National Laboratory Workshop on Innovation.
第16回 若手科学者によるプラズマ研究会 JAEA
NSTX-U NSTX-U PAC-31 Response to Questions – Day 1 Summary of Answers Q: Maximum pulse length at 1MA, 0.75T, 1 st year parameters? –A1: Full 5 seconds.
Divertor Design Considerations for CFETR
High  p experiments in JET and access to Type II/grassy ELMs G Saibene and JET TF S1 and TF S2 contributors Special thanks to to Drs Y Kamada and N Oyama.
Radiative divertor with impurity seeding in NSTX V. A. Soukhanovskii (LLNL) Acknowledgements: NSTX Team NSTX Results Review Princeton, NJ Wednesday, 1.
PF1A upgrade physics review Presented by D. A. Gates With input from J.E. Menard and C.E. Kessel 10/27/04.
Physics of fusion power Lecture 9 : The tokamak continued.
NSTX-U cryo/particle control discussion #8 December 19, 2012 What have we done, or plan to do to be responsive to PAC questions, and in prep for 5 year.
Fusion Fire Powers the Sun Can we make Fusion Fire on earth? National FIRE Collaboration AES, ANL, Boeing, Columbia U., CTD, GA, GIT, LLNL, INEEL, MIT,
Compact Stellarator Approach to DEMO J.F. Lyon for the US stellarator community FESAC Subcommittee Aug. 7, 2007.
DIVERTOR INVESTIGATIONS ON NSTX-U LEADING TO FNSF Mike Kotschenreuther Brent Covele Swadesh Mahajan Prashant Valanju Jonathan Roeltgen Zhong-Ping Chen.
PERSISTENT SURVEILLANCE FOR PIPELINE PROTECTION AND THREAT INTERDICTION International Plan for ELM Control Studies Presented by M.R. Wade (for A. Leonard)
EXTENSIONS OF NEOCLASSICAL ROTATION THEORY & COMPARISON WITH EXPERIMENT W.M. Stacey 1 & C. Bae, Georgia Tech Wayne Solomon, Princeton TTF2013, Santa Rosa,
Improved performance in long-pulse ELMy H-mode plasmas with internal transport barrier in JT-60U N. Oyama, A. Isayama, T. Suzuki, Y. Koide, H. Takenaga,
045-05/rs PERSISTENT SURVEILLANCE FOR PIPELINE PROTECTION AND THREAT INTERDICTION Technical Readiness Level For Control of Plasma Power Flux Distribution.
V. A. Soukhanovskii Lawrence Livermore National Laboratory, Livermore, California, USA for the NSTX-U and DIII-D Research Teams Snowflake Divertor Studies.
V. A. Soukhanovskii Lawrence Livermore National Laboratory, Livermore, California, USA H. Reimerdes Ecole Polytechnique Fédérale de Lausanne, Centre de.
1 EAST Recent Progress on Long Pulse Divertor Operation in EAST H.Y. Guo, J. Li, G.-N. Luo Z.W. Wu, X. Gao, S. Zhu and the EAST Team 19 th PSI Conference.
Approach for a High Performance Fusion Power Source Pathway Dale Meade Fusion Innovation Research and Energy ARIES Team Meeting March 3-4, 2008 UCSD, San.
Comments on Fusion Development Strategy for the US S. Prager Princeton Plasma Physics Laboratory FPA Symposium.
SMK – APS ‘06 1 NSTX Addresses Transport & Turbulence Issues Critical to Both Basic Toroidal Confinement and Future Devices NSTX offers a novel view into.
18th International Spherical Torus Workshop, Princeton, November 2015 Magnetic Configurations  Three comparative configurations:  Standard Divertor (+QF)
Fuel Cycle Research Thrust Using A Full Fusion Nuclear Environment
ZHENG Guo-yao, FENG Kai-ming, SHENG Guang-zhao 1) Southwestern Institute of Physics, Chengdu Simulation of plasma parameters for HCSB-DEMO by 1.5D plasma.
Page 1 Alberto Loarte- NSTX Research Forum st - 3 rd December 2009  ELM control by RMP is foreseen in ITER to suppress or reduce size of ELM energy.
MAST Upgrade: Status and plans
1 V.A. Soukhanovskii/IAEA-FEC/Oct Developing Physics Basis for the Radiative Snowflake Divertor at DIII-D by V.A. Soukhanovskii 1, with S.L. Allen.
Snowflake Divertor as Plasma-Material Interface for Future High Power Density Fusion Devices V. A. Soukhanovskii, R. E. Bell, S. P. Gerhardt, S. Kaye,
Stellarator Divertor Design and Optimization with NCSX Examples
Features of Divertor Plasmas in W7-AS
Major aims of IPP-NIFS collaboration on divertor physics
L-H power threshold and ELM control techniques: experiments on MAST and JET Carlos Hidalgo EURATOM-CIEMAT Acknowledgments to: A. Kirk (MAST) European.
Stellarator Program Update: Status of NCSX & QPS
Presentation transcript:

Advanced divertor magnetic configurations for tokamaks: concepts, status, future. Vlad Soukhanovskii Edge Coordinating Committee Fall 2014 Technical Meeting 29 October 2014 New Orleans, Louisiana

Acknowledgements Dmitri Ryutov, LLNL Dale Meade, Princeton, NJ

This talk to discuss poloidal magnetic divertor configurations for tokamaks Outline Introduction Poloidal magnetic divertor configuration Perspectives on advanced magnetic divertor configurations: physics, engineering, history Status of experiments Research plans

Other Plasma-Material Interface areas are as important, however, not discussed in this talk Core and pedestal integration Plasma facing components Divertor target geometry Continuously moving divertor plates Liquid metal divertors Pebble divertors Solid divertor targets with active cooling Operating scenarios Particle control with cryo Radiative regimes Ergodic divertors 3D fields Numerical plasma and material models

Critical divertor tasks Poloidal X-point divertor enabled progress in tokamak physics studies in the last 30 years Critical divertor tasks Power exhaust D/T and He pumping Impurity source reduction Impurity screening Advanced magnetic configurations: potential to perform the divertor tasks better than the standard X-point divertor National Spherical Torus Experiment, Princeton Plasma Physics Lab C. S. Pitcher and P. Stangeby, PPCF 39, 779 (1997)

Significant gaps exist between present divertor solutions and future device requirements Pheat/R Present experiments: ≤ 14 MW/m ITER: ≤ 20 MW/m DEMO: 80-100 MW/m Proposed solution: radiate up to 80-90% Steady-state heat flux Technological limit qpeak ≤ 5-15 MW/m2 ITER: qpeak ≤ 10 MW/m2 (Mitigated) DEMO: qpeak ≤ 150 MW/m2 (Unmitigated) ELM energy, target peak temperature Melting limit 0.1-0.5 MJ/m2 DEMO: Unmitigated, ≥ 10 MJ/m2 Impurity erosion (divertor target plasma temperature) Greenwald report, Toroidal Alternate Panel Report, ReNeW IAEA DEMO Workshops

Advanced magnetic divertor configurations can improve standard divertor properties and performance Radiated power loss Increase via Vdiv, LII Poloidal Target Inclination, etc N Increase plasma-wetted area Increase divertor area at large RSP Increase lq via increased radial transport, LII Number of Divertors / legs Divertor physics is inherently 2D or even 3D Parallel / cross-field transport and turbulence Radiation front (detachment) stability Neutral pressure / density distribution Increase plasma-wetted area via increasing fexp

Engineering and technology aspects define divertor configuration options Plasma equilibria, shaping and control Magnetic coils – inside or outside TF magnet Neutron shielding Cooling Electromagnetic forces Maintenance and remote handling

Advanced divertor magnetic configurations (classified by appearance) Multiple divertors, each with one X-point Higher order (2nd, 3rd) null divertors Divertor with multiple X-points Long-legged divertors with multiple X-points Note on early concepts Envisioned before H-mode discovery (1982) Some concepts envisioned divertor for particle and impurity control only

1. Multiple divertors: triple-null, quadruple-null to share heat and particle fluxes Poloidal Divertor Experiment tokamak at PPPL (1979 – 1983) Significant contribution to divertor physics with double-null configuration D-shaped plasma, high triangularity Increased local shear Enhanced kink stbility K. Bol et.al, Nucl. Fusion 25, 1149 (1985) J. Kesner, Nucl. Fusion 30, 548 (1990)

Snowflake, 2nd order null 2. Higher order null divertors : larger region of very low Bp to affect geometry and transport Snowflake, 2nd order null Bp ~ 0, grad Bp ~ 0 (Cf. first-order null: Bp ~ 0) Bp(r)~r2 (Cf. first-order null: Bp ~ r ) Four divertor legs Cloverleaf, 3rd order null Bp(r)~r3 Six divertor legs Strong plasma convection D. D. Ryutov, Phys. Plasmas 14 (2007), 064502 D. D. Ryutov et. al, Phys. Plasmas 20 (2013), 092509

3. Divertor with multiple X-points: expand SOL in the divertor region N. Ohyabu et. al, Nucl. Fusion 5 519 (1981) N. Ohyabu, J. Plasma Fus. Res. 5 525 (1991) Doublet III Expanded Boundary Poloidal Bundle + Expanded Boundary Analysis of radiation, H-mode compatibility, neutron shielding, coil currents

3. Divertor with multiple X-points: expand SOL in the divertor region H. Takase, J. Phys. Soc. J. 70, 609, 2001 M. Kotschenreuther et. al, IAEA FEC 2004; Phys. Plasmas 14, 072502 (2007) Cusp-like divertor configuration Coil currents acceptable for ITER- like parameters X-Divertor Small dipole coils under each divertor leg, inside the TF Potential to stabilize rad. front

4. Long-legged divertors with multiple X-points: increase connection length, expand SOL F. H. Tenney et.al, J. Nucl. Mater. 53 43 (1974) A.V. Georgievsky et.al, 6th Symp Eng Prob of Fus Energy, p 583, 1975, IEEE 75CH1097-5-NPS, Copyright 1976 F. H. Tenney, PPPL Report 1284, 1976 Conceptual divertor design for Princeton Reference Design Reactor. Long-legged high flux expansion poloidal divertor Long-legged double null poloidal divertor

Long-legged divertors with multiple X-points R. W. Conn et.al, U. Wisc. Report UWFDM-114, 1974 B. Badger et.al, U. Wisc. Report UWFDM-150, 1975 G. L. Kulcinski et.al, U. Wisc. Report UWFDM-173, 1976 UWMAK-II UWMAK-III Tokamak Enginering Test Reactor UW Fusion Technology Institute conceptual reactor systems studies

4. Long-legged divertors with multiple X-points T. F. Yang et.al, Westinghouse Corp. WFPS-TME-055 1977 TNS reactor (w/ ORNL) Divertor heat flux 1-3 MW/m2 Flowing liquid lithium targets for heat and particle removal D. Meade, Private Communication PDX Modification proposal

4. Long-legged divertors remove interaction zone away from plasma Conceptual design Engineering Test Facility R=5.6 m a=1.3 m Bt=5.5 T Ip=6.1 MA PNBI=60 MW Poloidal divertor W target plates Cryo-panels for particle control P. H. Sager et. al, J. Vac. Sci. Tech. 18, 1081 (1981)

4. Long-legged divertors with multiple X-points P. M. Valanju et.al, Phys. Plasmas 16, 056110 (2009) B. LaBombard et.al, APS 2013, IAEA 2014 M. Peng, Steady-state Spherical tokamak TST, Workshop on Edge Plasma for BPX and ITER, 1991 X-point target divertor ADX tokamak proposal Super-X divertor

Advanced magnetic divertor configurations: status of experiments Dedicated experimental devices (1979-1986) PDX Doublet III Expanded boundary Snowflake divertor configuration (existing devices with existing coils, 2008-present) TCV, 2008 NSTX, 2009 DIII-D, 2012 EAST, 2014 Long leg divertor physics (2010-present)

1. Doublet III Expanded Boundary was the first advanced magnetic divertor configuration experiment Stability of divertor configuration with outside PF coils Integrated power and particle exhaust Reduction of intrinsic impurities in the core Radiative divertor plasma cooling with seeded argon Argon screening from main plasma Compatibility with core confinement N. Ohyabu et. al, Nucl. Fusion 5 519 (1981) A. Mahdavi et. al, J. Nuc. Mater. 111 (1982) 355

2. Snowflake configuration has been realized in several devices using existing PF coils TCV Ip≤1 MA Bt≤1.5 T 16 PF coils, Pre- programmed currents NSTX Ip=0.8-1.0 MA Bt≤0.45 T 3 divertor coils, pre- programmed currents DIII-D Ip=0.8-1.0 MA Bt=2 T 3 divertor coils w/control

Divertor heat flux significantly reduced due to snowflake divertor geometry effects Standard Snowflake Standard Snowflake SOL snowflake NSTX DIII-D

Snowflake divertor enables power and particle sharing over multiple strike points Standard Snowflake GSP3 qSP3 qSP1 V. A. Soukhanovskii et.al, IAEA FEC 2014 W. Vijvers et.al, IAEA FEC 2012

Divertor heat transport affected by snowflake geometry Parallel heat flux (MW/m2) lq = 2.40 mm lq = 3.20 mm EMC3-IRENE modeling under-predicts power in additional strike points Suggests additional transport channel in the null region Increased lq may imply increased transport Increased radial spreading due to L|| SOL transport affected by null-region mixing H. Reimerdes et al. PPCF 2013, T. Lunt et al. PPCF 2014 V. A. Soukhanovskii et.al, IAEA FEC 2014

Snowflake configuration favorably affects radiative divertor and detachment Standard Snowflake PSOL = 3-4 MW Standard Snowflake Broader radiated power distribution, nearly complete power detachment in DIII-D Natural partial detachment in NSTX snowflake otherwise inaccessible with standard divertor V. A. Soukhanovskii et.al, NF 2011; POP 2011 V. A. Soukhanovskii et.al, IAEA FEC 2014

Longer connection length vs shorter connection length 3. Recent DIII-D experiments demonstrated benefits of the increased connection length Longer connection length vs shorter connection length Divertor peak heat flux reduced Divertor Te reduced Divertor radiation increased T. Petrie et.al, APS 2013, PSI2014, IAEA FEC 2014, APS 2014

Advanced magnetic divertor configuration development Near-term plans (5 years) Clarify effects of 2nd order null, extra X-points and long legs on Pedestal stability Steady-state and transient transport (heat, ion, impurity) Impurity radiation limits Tokamaks Upgraded: TCV, NSTX-U, HL-2M, MAST-U Existing: DIII-D, EAST Long-term plans (5-15 years) ?

MAST Upgrade to test advanced divertor configurations Super-X Snowflake Bt=0.8 T Ip ≤ 2 MA PNBI ≤ 7.5 MW 8 divertor PF coils Extensive diagnostic set Radial and parallel transport, stability of detachment Pedestal formation, structure, 3D fields G. Fishpool et.al, J. Nucl. Mater. 2013

Snowflake divertor is a leading heat flux mitigation candidate for NSTX-Upgrade New center-stack 2nd neutral beam BT Ip 1 T PNBI pulse 12 MW 2 MA 5 s NSTX-U Mission elements: Advance ST as candidate for Fusion Nuclear Science Facility Develop solutions for the plasma-material interface challenge Explore unique ST parameter regimes to advance predictive capability for ITER Develop ST as fusion energy system J. E. Menard et. al, Nucl. Fusion 52 (2012) 083015

HL-2M tokamak enables a number of advanced magnetic configurations jj Snowflake Snowflake-plus Snowflake-minus Tripod Major radius R = 1.78 m Minor radius a = 0.65 m Toroidal field Bt = 2.2 T Plasma current IP = 2.5 MA Pin=32 MW (Design) G.Y. Zheng et.al, Fus. Eng. Design 89 (2014) 2621

Many pre-DEMO and DEMO designs include snowflake and Super-X configurations Super-X for Aries Slim CS Snowflake for DEMO M. Kotschenreuther et. al, ARIES Workshop 2010 R Albanese et. al, PPCF, 56 (2014) 035008 China Fusion Experimental Reactor Y. Wan, SOFE 2013 Z. Luo et.al, IEEE TRANSACTIONS ON PLASMA SCIENCE, V42, 2014, 1021 Super-X and Snowflake for DEMO N. Asakura et.al, Trans. Fus. Sci. Tech. 63, 2013, 70.

Everything New Is Actually Well-Forgotten Old (Новое – это хорошо забытое старое, the Russian saying). A number of potentially attractive magnetic divertor configurations exist Much research remains to be done to qualify them as Advanced magnetic divertor configurations Divertor candidates for a fusion reactor

Backup

Abstract The present vision for controlling the plasma–material interface of a tokamak is an axisymmetric poloidal magnetic X-point divertor. The divertor must enable access to high core and pedestal plasma performance metrics while keeping target plate heat loads and erosion within the operating limits of plasma-facing component cooling technology and target plate materials. The proposed ITER divertor is based on standard X-point geometry designs tested in large tokamak experiments and uses tilted vertical targets to generate partial radiative detachment of the strike points. However, the standard divertor approach is likely to be insufficient for next step advanced tokamak and spherical tokamak devices such as the proposed fusion nuclear science facilities and for the DEMO reactor. Novel magnetic divertor configuration development and optimization has always been an active area in fusion plasma research. In this talk, advanced poloidal divertor concepts and experimental performance will be reviewed. Advanced divertors have the capability to modify steady-state and transient power exhaust via modifications to parallel and perpendicular transport and dissipative loss channels. The basic physics principles of these concepts will be summarized, from the first divertor for impurity control proposed by L. Spitzer for the stellarator, to long legged divertors, expanded boundaries, multiple X-point divertors, and multipole divertors for tokamaks. Many of the these divertor configurations face practical limitations on magnetic coil layout and construction. In recent years, two advanced divertor concepts have been pursued experimentally: snowflake (2nd order null) divertors, implemented in the TCV, NSTX, DIII-D and EAST tokamaks with existing magnetic coils, and the (long-legged) Super-X divertor, which is presently being implemented in MAST Upgrade using specially designed additional coils. The status and plans for research in these areas will be summarized. Several outstanding physics and engineering problems need to be addressed in order to qualify an advanced divertor concept for a next step tokamak reactor. This talk will discuss the general issues and motivations, including coil design and placement, equilibria design and plasma real-time control, plasma- facing component design, compatibility with highly radiative scenarios, and integration with high-performance core and pedestal plasma. This work is supported by the US Department of Energy under DE-AC5207NA27344.

Conventional divertor history… Poloidal bundle divertor First proposal of a magnetic divertor L. Spitzer, Phys. Fluids 1 (1958) 253 (for impurity control in a stellarator) First use of a poloidal divertor JFT-2a (DIVA), Japan, 1975 (approx.) First demonstration of impurity control using a divertor JFT-2a, Japan, 1975 (approx.) First demonstration of H-mode with a poloidal divertor ASDEX, Germany 1982 A. D. Sanderson et.al, J. Nucl. Mater. 76 530 (1978)