Status of measurement requirements for the ITER divertor

Size: px
Start display at page:

Download "Status of measurement requirements for the ITER divertor"

Transcription

1 Status of measurement requirements for the ITER divertor 1 R. A. Pitts, 2 G. Vayakis, 2 A. Costley with thanks for comments to A. Kukushkin, D. Whyte 1 Centre de Recherches en Physique des Plasmas, Association EURATOM-Confédération Suisse, École Polytechnique Fédérale de Lausanne, CH-1015 Switzerland 2 ITER-IT, Naka, Japan

2 Recent history of these measurement requirements 13/10/2000: A. Costley gives review of measurement of ITER divertor plasma and target parameters at SOL and Divertor Physics Expert Group, Garching, - specified (at that time) and estimated measurement performance. 09/07/2001: A. Costley gives update on divertor measurement requirements at the 14th Divertor and SOL Expert Group meeting in Naka on the basis of some feedback following the Garching meeting. Presentation given in the form of 6 discussion points in key areas requiring attention. Action item on R. A. Pitts at the Naka meeting to propose a minimum set of requirements on behalf of the SOL and Divertor group, addressing each of the discussion points. 15/11/2001: G. Vayakis makes a presentation at the first ITPA diagnostics meeting in St. Petersburg using as a base a document transmitted to G. Vayakis and A. Costley containing feedback on the discussion points.

3 This process has led to a set of modified requirements and some actions on the ITPA Diagnostics Group. Aim here is to summarise these measurement specifications for the ITPA Divertor Group following the lines of the previous presentation by G. Vayakis. This is in preparation for the debate to be held in a joint session of the Divertor and Diagnostics ITPA groups on the first morning of the 2nd Diagnostics ITPA meeting: Monday March 4, Apologies in advance for some repetition of previous material - there are some new members in the group.

4 Six key discussion points where minimum measurement requirement uncertain * 1 Is the proposed spatial resolution for divertor total radiation (profiles and inverted data) good enough? 2 What is the minimum required time resolution and precision for target plate heat flux measurements? 3 Given the measurement difficulties, what are the resolution requirements for target plate erosion? 4 The question of T e, n e, ion flux measurement at the target plates. 5 How serious is it if T e, n e measurements have relatively low spatial resolution in the outer divertor leg? How serious is it if there are no measurements in the inner leg? 6 Is visible spectroscopy sufficient in the divertor (given the difficulty of doing UV spectroscopy)? * A. Costley & G. Vayakis, 14th Divertor and SOL EG meeting, Naka, 9-11 July 2001

5 Discussion point #1: radiated Power A spatial resolution of 5 cm in the radiated power is all that is likely to be available. Is it enough? What are the consequences if it is > 5 cm, say 10 cm?

6 Current proposed LOS: main chamber/divertor Upper Port 60 lines of sight X mm A Upper Port 60 lines of sight H F C B G E D Equatorial Port 80 lines of sight Div Bolom/LAMx Divertor Cassette 120 lines of sight Current design has ~340 lines of sight. Chordal resolution: ~5 cm, spatial resolution after inversion > 10 cm. Concern about lifetime of divertor cameras and strong effect of neutrals. Only very coarse resolution possible if too many divertor chords are lost. Performance improvements presently limited by cash.

7 From G. Janeschitz, L. de Kock, A. Kukushkin et al., Diagnostic Requirements for the ITER Divertor in Proc. Int. Workshop on Diagnostics for ITER, Varenna, 1997 Partially attached case div. inner main & SOL div. outer 1%C (#136) 80 total carbon neon helium Carbon + 0.2% Ne seeding 10% He at coreedge interface 2D radiated power profile per unit volume (MW/m 3 ) from all C charge states 20 neutral A.Kukushkin, H.D.Pacher 5/ target X X target length [m] Cumulative poloidal radiation integral in inner, outer divertor and SOL Most of the radiation comes from the divertors, but otherwise well spread out along the divertor legs In general, proposed spatial resolution (say 10x10 cm pixel) in an inverted image should be acceptable except in strike point zones (especially outer target). Within a tight budget, performance improvements should concentrate on improving resolution in the target zones if possible.

8 Discussion point #1: Radiated Power - conclusion Presently envisaged potential resolution offered by the divertor (in combination with main chamber LOS) acceptable. If possible, performance enhancements to be directed toward improving resolution near the targets. Is it worth spending some time investigating the possibility of complementing the system with dead layer AXUV diodes if radiation hardness can be demonstrated?

9 Discussion point #2: time resolution and precision of target plate surface temperature/power flux We have been asked to consider increasing the time resolution in the target plate measurement [of heat flux] to 20 µs (from 2 ms). On this timescale, we expect the accuracy [of the present system] to be poor ( T > 200 C). Is this serious? What is the minimum requirement?

10 Current proposed target IR measurement 1.0 m z 4780 nm P6 P5 P nm 4340 nm P nm P6 P5 P4 P3 P2 P1? P nm Horizontal Viewing Duct 0.4 m Ellipsoidal M irror Grating 200 L/mm Slit1.6 mm x 40 mm 3210 nm 7.0 m 7.4 m R Concept unchanged from ITER-1998 design. Figures here from DDD 5.5.G.06 P1 A system exists for both target plates Spatial/time resolution: 3 mm/2 ms temperature range: C, wavelength range: 3-5 µm, accuracy, 10% Combined with wide-angle viewing from equatorial ports Wide viewing angle makes borescope optics difficult and mulit-element - use wavelength multiplexing collection optics Inverse Rowland circle spectrometer - spatial information encoded into wavelength and recovered with a second spectrometer outside biological shield

11 Requirement for monitoring effects of ELMs Temperature,ûC Transient ELMs CFC (20 mm) (0.8 MJ/m2-200µs, initial heat flux=10 MW/m2) Vaporised thickness Temperature Time, s Transient ELMs W (15 mm) (0.8 MJ/m2, initial Heat Flux=10 MW/m2) Melted layer Vaporised thickness Temperature Time, s Vaporised/Melted Thickness (µm) From G. Federici, 14th Divertor SOL EG meeting, Naka, July 2001 Simulations show that a 200 µs ELM depositing 0.8 MJ/m 2 raises T surf to melting (sublimation) limit of W (CFC) in a time < 100 µs Depending on inter-elm power flux, starting temperature could be as low as ~400 C. ITER FDR instrumental resolution would permit time resolution of the order of several µs for T surf exceeding 1000 C with high accuracy. Any events leading to T surf rise above ~1000 C can be monitored with high accuracy Poorer performance at low T surf not a serious problem - time resolution can be sacrified if no transients. Need more simulation (and understanding) of exactly what to expect for ITER ELMs: Instrument could be in difficulty for low starting T surf and smaller ELMs Sensitivity can also be strongly affected by changes in surface ε.

12 Discussion point #2: Surface temperature - conclusion Quoted spatial resolution (~ 3 mm) adequate. Very high time resolution (µs) and sensitivity for high T surf near divertor target operating limits. Perfectly adequate for ELM monitoring. Some potential problems if low starting T surf and smaller ELMs. Potential difficulties under conditions when surfaces change radically (and quickly) under erosion/ redeposition (common to any system). Need to use any new ITER ELM simulations as they come to re-evaluate instrument performance. More emphasis should be placed on the use of tile thermocouples (but water cooling makes it harder).

13 Discussion point #3: target plate erosion It is suggested that the resolution in the measurement of target plate erosion should be 0.1 µm for a single Type I ELM in real time. It will not be possible to meet this; 100 µm is more likely. How serious is this?

14 Quiescent case: no ELMs no disruptions ( semidetached, most recent case )* Peak net erosion rate: ~6 (< ) nm/s for C (W) Tritium co-deposition rate: 1-5 (~0) mg/s for C(W) If C used in the divertor, ~3 µm per 500s pulse eroded and ~500g T retained after 200, 8 min. pulses. Situation unknown but probably much worse in case of ELMs, disruptions etc. (depending on the ELMs ITER will have). Erosion measurements are difficult and so important to specify as carefully as possible the minimum requirements What should be the functionality of the system(s)? *From G. Federici, 14th Divertor SOL EG meeting, Naka, July 2001

15 Function 1: tread wear: signs of divertor end of life. Function 2: Real time capability: avoid dangerous regimes for the divertor when they are encountered, monitor ELM induced erosion etc. For C targets, erosion measurement can also be an indirect indicator of approximate T-retention. Real time would allow erosion budgets to be set for the operators - could even be one of the control categories. Function 1: Range finding system forseen (periscopic insertion between pulses) covering ~80% of the targets and first wall with spot size ~1mm and depth resolution ± mm. Function 2: Interferometric/range finding techniques might be extendable to 10 s µm and could conceivably be made real time (several sec. time res.) but this seems unlikely (and will be toroidally localised).

16 A full remote handling inspection system is forseen for ITER. This system can, if enough modules are equipped, image almost the entire first wall and divertor surfaces with 1 mm spatial resolution and ± 0.5 mm resolution for metrology (range finding, interferometry etc). It is possible that these systems could go down as low as ± 0.1 mm But not really an inter-pulse option.

17 Feasibility studies etc. required to find limitations of any proposed technique for real time (or at least inter-pulse), divertor localised observations. What minimum realistic measurement requirements can physics impose?

18 Discussion point #4: target plate probe measurements It will probably not be possible to measure reliably the plasma parameters (n e and T e ) at the target but only the ion flux with Langmuir probes. How serious is this? What are the consequences?

19 Proposed system has 3 groups of ~80 CFC single probes with ~1 cm pitch arranged as triple probes (as in JET) Probe is of JET design Potential problems with RIC and RIED Unknown how long these will survive in full performance discharges (or even before then). Survival only realistically possible in partially detached, cold divertors Probes an excellent indicator of detachment (via ion flux), with no problems of interpretation (except changes in A proj ). The usual problems of interpretation of absolute value of T e in high recycling - appears difficult to solve, but relative changes in T e could be usefully employed. System should be more than adequate and should be included. Lifetime issues impossible to judge for ITER using only JET experience.

20 Discussion points #5,6: T e and spectroscopy in the divertor Measurement of T e along the outer divertor leg is very difficult. There will be high resolution measurements in the upper SOL region (main chamber) and there could be some across the X- point. How serious is it if there are no measurements along the divertor leg? If there are, what should be the spatial resolution? It is difficult to make UV measurements in the divertor - will visible be good enough if good bolometry measurements are available. Presently no provision for measurements along the inner leg. How serious is this?

21 12 23 Divertor (imaging) and X-point (LIDAR) Thomson Centre de Recherches en Physique des Plasmas Divertor impurity monitor UPPER-PORT (2 port apart Toridally) EQUATORIAL-PORT (next Troidal port) X-POINT IV IH OV OH DIVERTOR-PORT DIVERTOR-CASSETTE BIO-SHIELD Two systems forseen: outer leg and X-pt. LIDAR system expected to provide 5 cm spatial resolution, target for divertor system is ~10 cm (both with 1 ms time res.) Inner leg viewing practically impossible (3 potential sightlines along inner leg could be provided by divertor reflectometry). Divertor TS cannot be swept poloidally, at least not in real time. Poloidal res. ~ 1 cm Combined viewing from main chamber and divertor give resolution along legs of ~ 4 cm and possibility for crude inversion giving 5-10 cm resolution for main impurities (C, W, Be) and D,T ( nm) Number of sightlines reduced in going from FDR to FEAT (shorter divertor legs and cost).

22 What divertor Thomson needs to do n e (sep.) 20 T e (sep.) T e x-point target 40 x-point target Poloidal distance [m] Poloidal distance [m] From A. Kukushkin, P SOL = 100 MW, n sep = 3.2x10 19 m -3, carbon divertor, no additoinal seeding Would ideally require better spatial resolution (say ~ 2 cm) within 20 cm of the target, more uniform (say 5-10 cm) away from the target. But must be some flexibility to vary the poloidal position of the main sightline once measurements begin (strike pt. sweeps difficult in ITER) This system extremely important as an aid toward interpretation of spectroscopic influxes but there are serious concerns for lifetime of front end components.

23 Regarding the need for UV spectroscopy: If Carbon is in the machine, then can use visible C spectroscopy in combination with bolometry to compute contributions to total radiation both diagnostics must have similar spatial resolution. C spectroscopy can also be used as an ionisation front position detector. If only W targets and no C elsewhere, UV spectroscopy probably essential if details of W transport required. Recommendation is that divertor (and X-pt.) Thomson essential, divertor UV not essential depending on impurities present. Divertor Thomson spatial resolution needs to be < 5 cm near plate, but can be lower further from target. Chordal resolution of impurity monitor ~ 4 cm along legs adequate but should not be lower.

24 Summary of revisions to ITER-FEAT FDR divertor requirements (changes in red)* MEASUREMENT PARAMETER RANGE or COVERAGE 16. Divertor operational parameters 37. Radiation profile 38. Heat loading profile in divertor 41. Divertor electron parameters 41. Divertor ion temperature RESOLUTION Temporal Spatial ACCUR ACY Max. Surface Temp C 2 ms - 10% Erosion rate µm/s 2 s 1 cm 30% Net Erosion 0-3 mm Per pulse 1 cm 12 µm Ionis. front position 0 - TBD m 1 ms 10 cm - X-pt./MARFE TBD region P rad MWm ms a/15 20% Divertor P rad TBD MWm ms 5 cm 30% Surface Temp C C Power load (default) 2 ms 20 µs 3 mm 10% 10% TBD - 25 MWm -2 2 ms 3 mm 10% n e m -3 1 ms 5 cm along leg, 3 mm across leg T e ev 1 ms 5 cm along leg, 3 mm across leg T i ev 1 ms 5 cm along leg, 3 mm across leg 20% 20% 20% * Minutes of 1st ITPA Diagnostics group N CX MI F1

25 Summary Centre de Recherches en Physique des Plasmas Divertor bolometry probably adequate as is. Very fast T surf excursions (eg. ELMs) could be followed with the proposed IR thermography if rise in T surf large enough. More work required to quantify effect of lower starting temp. and lower transient amplitude. Erosion measurement is the hardest. We need to discuss in more detail exactly what to specify here and to justify the need for real time measurements. Proposed target Langmuir probe diagnostic perfectly adequate (spatial resolution). Lifetime is another issue. Some changes required to divertor TS spatial resolution, but this diagnostic is essential and must be retained. Visible spectroscopy resolution should not be decreased below proposed (already reduced) specs. Could live without divertor UV if resources are limited.

Diagnostics for Burning Plasma Physics Studies: A Status Report.

Diagnostics for Burning Plasma Physics Studies: A Status Report. Diagnostics for Burning Plasma Physics Studies: A Status Report. Kenneth M. Young Princeton Plasma Physics Laboratory UFA Workshop on Burning Plasma Science December 11-13 Austin, TX Aspects of Plasma

More information

Power Deposition Measurements in Deuterium and Helium Discharges in JET MKIIGB Divertor by IR-Thermography

Power Deposition Measurements in Deuterium and Helium Discharges in JET MKIIGB Divertor by IR-Thermography EFDA JET CP(02)01/03 T Eich, A Herrmann, P Andrew and A Loarte Power Deposition Measurements in Deuterium and Helium Discharges in JET MKIIGB Divertor by IR-Thermography . Power Deposition Measurements

More information

Steady State and Transient Power Handling in JET

Steady State and Transient Power Handling in JET Steady State and Transient Power Handling in JET G.F.Matthews * on behalf of the JET EFDA Exhaust Physics Task Force and JET EFDA Contributors + + See annex of J. Pamela et al, "Overview of JET Results",

More information

Divertor power deposition and target current asymmetries during type-i ELMs in ASDEX Upgrade and JET

Divertor power deposition and target current asymmetries during type-i ELMs in ASDEX Upgrade and JET Journal of Nuclear Materials 363 365 (2007) 989 993 www.elsevier.com/locate/jnucmat Divertor power deposition and target current asymmetries during type-i ELMs in ASDEX Upgrade and JET T. Eich a, *, A.

More information

Divertor Requirements and Performance in ITER

Divertor Requirements and Performance in ITER Divertor Requirements and Performance in ITER M. Sugihara ITER International Team 1 th International Toki Conference Dec. 11-14, 001 Contents Overview of requirement and prediction for divertor performance

More information

ITER A/M/PMI Data Requirements and Management Strategy

ITER A/M/PMI Data Requirements and Management Strategy ITER A/M/PMI Data Requirements and Management Strategy Steven Lisgo, R. Barnsley, D. Campbell, A. Kukushkin, M. Hosokawa, R. A. Pitts, M. Shimada, J. Snipes, A. Winter ITER Organisation with contributions

More information

ITER. Power and Particle Exhaust in ITER ITER

ITER. Power and Particle Exhaust in ITER ITER Power and Particle Exhaust in ITER ITER G. Janeschitz, C. Ibbott, Y. Igitkhanov, A. Kukushkin, H. Pacher, G. Pacher, R. Tivey, M. Sugihara, JCT and HTs San Diego.5.2 Power and Particle Exhaust in ITER

More information

Divertor Heat Load in ITER-Like Advanced Tokamak Scenarios on JET

Divertor Heat Load in ITER-Like Advanced Tokamak Scenarios on JET EFDA JET CP(8)2/3 G. Arnoux, P. Andrew, M. Beurskens, S. Brezinsek, C.D. Challis, P. De Vries, W. Fundamenski, E. Gauthier, C. Giroud, A. Huber, S. Jachmich, X. Litaudon, R.A. Pitts, F. Rimini and JET

More information

Tokamak Divertor System Concept and the Design for ITER. Chris Stoafer April 14, 2011

Tokamak Divertor System Concept and the Design for ITER. Chris Stoafer April 14, 2011 Tokamak Divertor System Concept and the Design for ITER Chris Stoafer April 14, 2011 Presentation Overview Divertor concept and purpose Divertor physics General design considerations Overview of ITER divertor

More information

Information Session for the ITER CPTS System

Information Session for the ITER CPTS System Information Session for the ITER CPTS System Fusion for Energy Barcelona, 15 April 2015 1 Introduction to the meeting Information provided is preliminary and subject to Agenda change ahead of formal tendering

More information

Critical Gaps between Tokamak Physics and Nuclear Science. Clement P.C. Wong General Atomics

Critical Gaps between Tokamak Physics and Nuclear Science. Clement P.C. Wong General Atomics Critical Gaps between Tokamak Physics and Nuclear Science (Step 1: Identifying critical gaps) (Step 2: Options to fill the critical gaps initiated) (Step 3: Success not yet) Clement P.C. Wong General Atomics

More information

Impurity accumulation in the main plasma and radiation processes in the divetor plasma of JT-60U

Impurity accumulation in the main plasma and radiation processes in the divetor plasma of JT-60U 1 EX/P4-25 Impurity accumulation in the main plasma and radiation processes in the divetor plasma of JT-6U T. Nakano, H. Kubo, N. Asakura, K. Shimizu and S. Higashijima Japan Atomic Energy Agency, Naka,

More information

Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor

Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor 1 FIP/3-4Ra Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor N. Asakura 1, K. Hoshino 1, H. Utoh 1, K. Shinya 2, K. Shimizu 3, S. Tokunaga 1, Y.Someya 1, K. Tobita 1, N. Ohno

More information

Bolometry. H. Kroegler Assciazione Euratom-ENEA sulla Fusione, Frascati (Italy)

Bolometry. H. Kroegler Assciazione Euratom-ENEA sulla Fusione, Frascati (Italy) Bolometry H. Kroegler Assciazione Euratom-ENEA sulla Fusione, Frascati (Italy) Revised May 28, 2002 1. Radiated power Time and space resolved measurements of the total plasma radiation can be done by means

More information

ITER Divertor Plasma Modelling with Consistent Core-Edge Parameters

ITER Divertor Plasma Modelling with Consistent Core-Edge Parameters CT/P-7 ITER Divertor Plasma Modelling with Consistent Core-Edge Parameters A. S. Kukushkin ), H. D. Pacher ), G. W. Pacher 3), G. Janeschitz ), D. Coster 5), A. Loarte 6), D. Reiter 7) ) ITER IT, Boltzmannstr.,

More information

Modelling of JT-60U Detached Divertor Plasma using SONIC code

Modelling of JT-60U Detached Divertor Plasma using SONIC code J. Plasma Fusion Res. SERIES, Vol. 9 (2010) Modelling of JT-60U Detached Divertor Plasma using SONIC code Kazuo HOSHINO, Katsuhiro SHIMIZU, Tomonori TAKIZUKA, Nobuyuki ASAKURA and Tomohide NAKANO Japan

More information

Physics of fusion power. Lecture 14: Anomalous transport / ITER

Physics of fusion power. Lecture 14: Anomalous transport / ITER Physics of fusion power Lecture 14: Anomalous transport / ITER Thursday.. Guest lecturer and international celebrity Dr. D. Gericke will give an overview of inertial confinement fusion.. Instabilities

More information

Characterization of Tungsten Sputtering in the JET divertor

Characterization of Tungsten Sputtering in the JET divertor EX/P5-05 Characterization of Tungsten Sputtering in the JET divertor G.J. van Rooij 1, J.W. Coenen 2, L. Aho-Mantila 3, M. Beurskens 4, S. Brezinsek 2, M. Clever 2, R. Dux 5, C. Giroud 4, M. Groth 6, K.

More information

Estimating the plasma flow in a recombining plasma from

Estimating the plasma flow in a recombining plasma from Paper P3-38 Estimating the plasma flow in a recombining plasma from the H α emission U. Wenzel a, M. Goto b a Max-Planck-Institut für Plasmaphysik (IPP) b National Institute for Fusion Science, Toki 509-5292,

More information

Hydrogen and Helium Edge-Plasmas

Hydrogen and Helium Edge-Plasmas Hydrogen and Helium Edge-Plasmas Comparison of high and low recycling T.D. Rognlien and M.E. Rensink Lawrence Livermore National Lab Presented at the ALPS/APEX Meeting Argonne National Lab May 8-12, 2

More information

STEADY-STATE EXHAUST OF HELIUM ASH IN THE W-SHAPED DIVERTOR OF JT-60U

STEADY-STATE EXHAUST OF HELIUM ASH IN THE W-SHAPED DIVERTOR OF JT-60U Abstract STEADY-STATE EXHAUST OF HELIUM ASH IN THE W-SHAPED DIVERTOR OF JT-6U A. SAKASAI, H. TAKENAGA, N. HOSOGANE, H. KUBO, S. SAKURAI, N. AKINO, T. FUJITA, S. HIGASHIJIMA, H. TAMAI, N. ASAKURA, K. ITAMI,

More information

DEMO Concept Development and Assessment of Relevant Technologies. Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor

DEMO Concept Development and Assessment of Relevant Technologies. Physics and Engineering Studies of the Advanced Divertor for a Fusion Reactor FIP/3-4Rb FIP/3-4Ra DEMO Concept Development and Assessment of Relevant Technologies Y. Sakamoto, K. Tobita, Y. Someya, H. Utoh, N. Asakura, K. Hoshino, M. Nakamura, S. Tokunaga and the DEMO Design Team

More information

Fusion Development Facility (FDF) Divertor Plans and Research Options

Fusion Development Facility (FDF) Divertor Plans and Research Options Fusion Development Facility (FDF) Divertor Plans and Research Options A.M. Garofalo, T. Petrie, J. Smith, M. Wade, V. Chan, R. Stambaugh (General Atomics) J. Canik (Oak Ridge National Laboratory) P. Stangeby

More information

Divertor Detachment on TCV

Divertor Detachment on TCV Divertor Detachment on TCV R. A. Pitts, Association EURATOM-Confédération Suisse,, CH- LAUSANNE, Switzerland thanks to A. Loarte a, B. P. Duval, J.-M. Moret, J. A. Boedo b, L. Chousal b, D. Coster c, G.

More information

Experimental results and modelling of ASDEX Upgrade partial detachment

Experimental results and modelling of ASDEX Upgrade partial detachment Experimental results and modelling of ASDEX Upgrade partial detachment M. Wischmeier 1 With thanks to X. Bonnin 2, P. Börner 3, A. Chankin 1, D. P. Coster 1, M. Groth 4, A. Kallenbach 1, V. Kotov 3, H.

More information

Temperature measurement and real-time validation

Temperature measurement and real-time validation Temperature measurement and real-time validation A. Herrmann, B. Sieglin, M. Faitsch, P. de Marné, ASDEX Upgrade team st IAEA Technical Meeting on Fusion Data Processing, Validation and Analysis ITER-

More information

Electron Bernstein Wave Heating in the TCV Tokamak

Electron Bernstein Wave Heating in the TCV Tokamak Electron Bernstein Wave Heating in the TCV Tokamak A. Mueck 1, Y. Camenen 1, S. Coda 1, L. Curchod 1, T.P. Goodman 1, H.P. Laqua 2, A. Pochelon 1, TCV Team 1 1 Ecole Polytechnique Fédérale de Lausanne

More information

Divertor power and particle fluxes between and during type-i ELMs in ASDEX Upgrade

Divertor power and particle fluxes between and during type-i ELMs in ASDEX Upgrade Divertor power and particle fluxes between and during type-i ELMs in ASDEX Upgrade A. Kallenbach, R. Dux, T. Eich, R. Fischer, L. Giannone, J. Harhausen, A. Herrmann, H.W. Müller, G. Pautasso, M. Wischmeier,

More information

LASER-AIDED PLASMA DIAGNOSTICS FOR ITER. Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein, The Netherlands;

LASER-AIDED PLASMA DIAGNOSTICS FOR ITER. Euregio Cluster, P.O. Box 1207, 3430 BE Nieuwegein, The Netherlands; LASER-AIDED PLASMA DIAGNOSTICS FOR ITER A.J.H. DONNÉ, 1 O. BUZHINSKIJ, 2 A.E. COSTLEY, 3 K. EBISAWA, 3 S. KASAI, 4 A. KOCH, 5 T. KONDOH, 4 I. MOSKALENKO, 6 P. NIELSEN, 7 G. RAZDOBARIN, 8 G. VAYAKIS, 3

More information

1 EX/3-5. Material Erosion and Redeposition during the JET MkIIGB-SRP Divertor Campaign

1 EX/3-5. Material Erosion and Redeposition during the JET MkIIGB-SRP Divertor Campaign 1 Material Erosion and Redeposition during the JET MkIIGB-SRP Divertor Campaign A. Kirschner 1), V. Philipps 1), M. Balden 2), X. Bonnin 3), S. Brezinsek 1), J.P. Coad 4), D. Coster 2), S.K. Erents 4),

More information

THE ADVANCED TOKAMAK DIVERTOR

THE ADVANCED TOKAMAK DIVERTOR I Department of Engineering Physics THE ADVANCED TOKAMAK DIVERTOR S.L. Allen and the team 14th PSI QTYUIOP MA D S O N UCLAUCLA UCLA UNIVERSITY OF WISCONSIN THE ADVANCED TOKAMAK DIVERTOR S.L. Allen and

More information

Physics of the detached radiative divertor regime in DIII-D

Physics of the detached radiative divertor regime in DIII-D Plasma Phys. Control. Fusion 41 (1999) A345 A355. Printed in the UK PII: S741-3335(99)97299-8 Physics of the detached radiative divertor regime in DIII-D M E Fenstermacher, J Boedo, R C Isler, A W Leonard,

More information

Estimation of the contribution of gaps to tritium retention in the divertor of ITER

Estimation of the contribution of gaps to tritium retention in the divertor of ITER Estimation of contribution of gaps to tritium retention in the divertor of ITER 1 Estimation of the contribution of gaps to tritium retention in the divertor of ITER 1. Introduction D. Matveev 1,2, A.

More information

Materials for Future Fusion Reactors under Severe Stationary and Transient Thermal Loads

Materials for Future Fusion Reactors under Severe Stationary and Transient Thermal Loads Mitglied der Helmholtz-Gemeinschaft Materials for Future Fusion Reactors under Severe Stationary and Transient Thermal Loads J. Linke, J. Du, N. Lemahieu, Th. Loewenhoff, G. Pintsuk, B. Spilker, T. Weber,

More information

Scaling of divertor heat flux profile widths in DIII-D

Scaling of divertor heat flux profile widths in DIII-D 1 Scaling of divertor heat flux profile widths in DIII-D C.J. Lasnier 1, M.A. Makowski 1, J.A. Boedo 2, N.H. Brooks 3, D.N. Hill 1, A.W. Leonard 3, and J.G. Watkins 4 e-mail:lasnier@llnl.gov 1 Lawrence

More information

Spatio-temporal investigations on the triggering of pellet induced ELMs

Spatio-temporal investigations on the triggering of pellet induced ELMs Spatio-temporal investigations on the triggering of pellet induced ELMs G. Kocsis, S. Kálvin, P.T. Lang*, M. Maraschek*, J. Neuhauser* W. Schneider*, T. Szepesi and ASDEX Upgrade Team KFKI-RMKI, EURATOM

More information

Introduction to the Diagnosis of Magnetically Confined Thermonuclear Plasma

Introduction to the Diagnosis of Magnetically Confined Thermonuclear Plasma Introduction to the Diagnosis of Magnetically Confined Thermonuclear Plasma EDGE-SOL II: Plasma Wall Interactions J. Arturo Alonso Laboratorio Nacional de Fusión EURATOM-CIEMAT E6 P2.10 arturo.alonso@ciemat.es

More information

ARTICLES PLASMA DETACHMENT IN JET MARK I DIVERTOR EXPERIMENTS

ARTICLES PLASMA DETACHMENT IN JET MARK I DIVERTOR EXPERIMENTS ARTICLES PLASMA DETACHMENT IN JET MARK I DIVERTOR EXPERIMENTS A. LOARTE, R.D. MONK, J.R. MARTÍN-SOLÍSa,D.J.CAMPBELL, A.V. CHANKIN b, S. CLEMENT, S.J. DAVIES, J. EHRENBERG, S.K. ERENTS c,h.y.guo, P.J. HARBOUR,

More information

Introduction to the Diagnosis of Magnetically Confined Thermonuclear Plasma

Introduction to the Diagnosis of Magnetically Confined Thermonuclear Plasma Introduction to the Diagnosis of Magnetically Confined Thermonuclear Plasma Core diagnostics II: Bolometry and Soft X-rays J. Arturo Alonso Laboratorio Nacional de Fusión EURATOM-CIEMAT E6 P2.10 arturo.alonso@ciemat.es

More information

Mission Elements of the FNSP and FNSF

Mission Elements of the FNSP and FNSF Mission Elements of the FNSP and FNSF by R.D. Stambaugh PERSISTENT SURVEILLANCE FOR PIPELINE PROTECTION AND THREAT INTERDICTION Presented at FNST Workshop August 3, 2010 In Addition to What Will Be Learned

More information

Development of an experimental profile database for the scrape-off layer

Development of an experimental profile database for the scrape-off layer Development of an experimental profile database for the scrape-off layer M. Groth, 1 G.D. Porter, 1 W.M. Meyer, 1 A.W. Leonard, 2 T.H. Osborne, 2 D.P. Coster, 3 A. Kallenbach, 3 M. Wischmeier 3 N.H. Brooks,

More information

1 EX/P4-8. Hydrogen Concentration of Co-deposited Carbon Films Produced in the Vicinity of Local Island Divertor in Large Helical Device

1 EX/P4-8. Hydrogen Concentration of Co-deposited Carbon Films Produced in the Vicinity of Local Island Divertor in Large Helical Device 1 EX/P4-8 Hydrogen Concentration of Co-deposited Carbon Films Produced in the Vicinity of Local Island Divertor in Large Helical Device T. Hino 1,2), T. Hirata 1), N. Ashikawa 2), S. Masuzaki 2), Y. Yamauchi

More information

Divertor Power Handling Assessment for Baseline Scenario Operation in JET in Preparation for the ILW

Divertor Power Handling Assessment for Baseline Scenario Operation in JET in Preparation for the ILW EFDA JET CP(9)6/54 I. Nunes, P.J. Lomas, G. Saibene, T. Eich, G. Arnoux, H. Thomsen, E de la Luna and JET EFDA contributors Divertor Power Handling Assessment for Baseline Scenario Operation in JET in

More information

Power Exhaust on JET: An Overview of Dedicated Experiments

Power Exhaust on JET: An Overview of Dedicated Experiments Power Exhaust on JET: An Overview of Dedicated Experiments W.Fundamenski, P.Andrew, T.Eich 1, G.F.Matthews, R.A.Pitts 2, V.Riccardo, W.Sailer 3, S.Sipila 4 and JET EFDA contributors 5 Euratom/UKAEA Fusion

More information

Divertor Heat Flux Reduction and Detachment in NSTX

Divertor Heat Flux Reduction and Detachment in NSTX 1 EX/P4-28 Divertor Heat Flux Reduction and Detachment in NSTX V. A. Soukhanovskii 1), R. Maingi 2), R. Raman 3), R. E. Bell 4), C. Bush 2), R. Kaita 4), H. W. Kugel 4), C. J. Lasnier 1), B. P. LeBlanc

More information

Carbon Deposition and Deuterium Inventory in ASDEX Upgrade

Carbon Deposition and Deuterium Inventory in ASDEX Upgrade 1 IAEA-CN-116 / EX / 5-24 Carbon Deposition and Deuterium Inventory in ASDEX Upgrade M. Mayer 1, V. Rohde 1, J. Likonen 2, E. Vainonen-Ahlgren 2, J. Chen 1, X. Gong 1, K. Krieger 1, ASDEX Upgrade Team

More information

Exhaust scenarios. Alberto Loarte. Plasma Operation Directorate ITER Organization. Route de Vinon sur Verdon, St Paul lez Durance, France

Exhaust scenarios. Alberto Loarte. Plasma Operation Directorate ITER Organization. Route de Vinon sur Verdon, St Paul lez Durance, France Exhaust scenarios Alberto Loarte Plasma Operation Directorate ITER Organization Route de Vinon sur Verdon, 13067 St Paul lez Durance, France Acknowledgements: Members of ITER Organization (especially R.

More information

Deuterium Balmer/Stark spectroscopy and impurity profiles: first results from mirror-link divertor spectroscopy system on the JET ITER-like wall

Deuterium Balmer/Stark spectroscopy and impurity profiles: first results from mirror-link divertor spectroscopy system on the JET ITER-like wall CCFE-PR(13)35 A.G. Meigs, S. Brezinsek, M. Clever, A. Huber, S. Marsen, C. Nicholas, M.Stamp, K-D Zastrow, and JET EFDA Contributors Deuterium Balmer/Stark spectroscopy and impurity profiles: first results

More information

Measurement Requirements and the Diagnostic System on ITER: Modifications Following the Design Review.

Measurement Requirements and the Diagnostic System on ITER: Modifications Following the Design Review. 1 Topic: IT/P6-21 Measurement Requirements and the Diagnostic System on ITER: Modifications Following the Design Review. A E Costley 1), S Allen 2), P Andrew 1), L Bertalot 1), R Barnsley 1), X R Duan

More information

Review of experimental observations of plasma detachment and of the effects of divertor geometry on divertor performance

Review of experimental observations of plasma detachment and of the effects of divertor geometry on divertor performance Review of experimental observations of plasma detachment and of the effects of divertor geometry on divertor performance Alberto Loarte European Fusion Development Agreement Close Support Unit - Garching

More information

Driving Mechanism of SOL Plasma Flow and Effects on the Divertor Performance in JT-60U

Driving Mechanism of SOL Plasma Flow and Effects on the Divertor Performance in JT-60U EX/D-3 Driving Mechanism of SOL Plasma Flow and Effects on the Divertor Performance in JT-6U N. Asakura ), H. Takenaga ), S. Sakurai ), G.D. Porter ), T.D. Rognlien ), M.E. Rensink ), O. Naito ), K. Shimizu

More information

Comparison of tungsten fuzz growth in Alcator C-Mod and linear plasma devices!

Comparison of tungsten fuzz growth in Alcator C-Mod and linear plasma devices! Comparison of tungsten fuzz growth in Alcator C-Mod and linear plasma devices G.M. Wright 1, D. Brunner 1, M.J. Baldwin 2, K. Bystrov 3, R. Doerner 2, B. LaBombard 1, B. Lipschultz 1, G. de Temmerman 3,

More information

ITER DIAGNOSTIC PORT PLUG DESIGN. N H Balshaw, Y Krivchenkov, G Phillips, S Davis, R Pampin-Garcia

ITER DIAGNOSTIC PORT PLUG DESIGN. N H Balshaw, Y Krivchenkov, G Phillips, S Davis, R Pampin-Garcia N H Balshaw, Y Krivchenkov, G Phillips, S Davis, R Pampin-Garcia UKAEA, Culham Science Centre, Abingdon, Oxon,OX14 3DB, UK, nick.balshaw@jet.uk Many of the ITER diagnostic systems will be mounted in the

More information

Radiative type-iii ELMy H-mode in all-tungsten ASDEX Upgrade

Radiative type-iii ELMy H-mode in all-tungsten ASDEX Upgrade Radiative type-iii ELMy H-mode in all-tungsten ASDEX Upgrade J. Rapp 1, A. Kallenbach 2, R. Neu 2, T. Eich 2, R. Fischer 2, A. Herrmann 2, S. Potzel 2, G.J. van Rooij 3, J.J. Zielinski 3 and ASDEX Upgrade

More information

2.6 Plasma Diagnostic System

2.6 Plasma Diagnostic System 2.6 Plasma Diagnostic System 2.6.1 Selected Diagnostic Systems and Startup Set 1 2.6.2 Diagnostic Integration 1 2.6.2.1 In-vessel Installations 3 2.6.2.2 Equatorial and Upper Ports 4 2.6.2.3 Divertor Ports

More information

Analyses of Visible Images of the Plasma Periphery Observed with Tangentially Viewing CCD Cameras in the Large Helical Device

Analyses of Visible Images of the Plasma Periphery Observed with Tangentially Viewing CCD Cameras in the Large Helical Device Analyses of Visible Images of the Plasma Periphery Observed with Tangentially Viewing CCD Cameras in the Large Helical Device M. SHOJI, T. WATANABE, S. MASUZAKI, H. YAMADA, A. KOMORI and LHD Experimental

More information

DIVIMP simulation of W transport in the SOL of JET H-mode plasmas

DIVIMP simulation of W transport in the SOL of JET H-mode plasmas DIVIMP simulation of W transport in the SOL of JET H-mode plasmas A. Järvinen a, C. Giroud b, M. Groth a, K. Krieger c, D. Moulton d, S. Wiesen e, S. Brezinsek e and JET- EFDA contributors¹ JET-EFDA, Culham

More information

Hydrocarbon transport in the MkIIa divertor of JET

Hydrocarbon transport in the MkIIa divertor of JET INSTITUTE OF PHYSICS PUBLISHING Plasma Phys. Control. Fusion 45 (23) 39 319 PLASMA PHYSICS AND CONTROLLED FUSION PII: S741-3335(3)37188-X Hydrocarbon transport in the MkIIa divertor of JET A Kirschner

More information

1. Motivation power exhaust in JT-60SA tokamak. 2. Tool COREDIV code. 3. Operational scenarios of JT-60SA. 4. Results. 5.

1. Motivation power exhaust in JT-60SA tokamak. 2. Tool COREDIV code. 3. Operational scenarios of JT-60SA. 4. Results. 5. 1. Motivation power exhaust in JT-60SA tokamak 2. Tool COREDIV code 3. Operational scenarios of JT-60SA 4. Results 5. Conclusions K. Gałązka Efficient power exhaust in JT-60SA by COREDIV Page 2 The Institute

More information

Dust collected in MAST and in Tore Supra. Nanoparticle growth in laboratory plasmas

Dust collected in MAST and in Tore Supra. Nanoparticle growth in laboratory plasmas FDR-FM Association EURATOM-EA Dust collected in MAST and in Tore Supra. Pardanaud 1,. Martin 1, P. Roubin 1,. Arnas 1 and G. De Temmerman 2 1 Lab. PIIM, NRS-Université de Provence, UMR 6633, 13397 Marseille,

More information

Implementation of a long leg X-point target divertor in the ARC fusion pilot plant

Implementation of a long leg X-point target divertor in the ARC fusion pilot plant Implementation of a long leg X-point target divertor in the ARC fusion pilot plant A.Q. Kuang, N.M. Cao, A.J. Creely, C.A. Dennett, J. Hecla, H. Hoffman, M. Major, J. Ruiz Ruiz, R.A. Tinguely, E.A. Tolman

More information

ADVANCES IN PREDICTIVE THERMO-MECHANICAL MODELLING FOR THE JET DIVERTOR EXPERIMENTAL INTERPRETATION, IMPROVED PROTECTION, AND RELIABLE OPERATION

ADVANCES IN PREDICTIVE THERMO-MECHANICAL MODELLING FOR THE JET DIVERTOR EXPERIMENTAL INTERPRETATION, IMPROVED PROTECTION, AND RELIABLE OPERATION D. IGLESIAS et al. ADVANCES IN PREDICTIVE THERMO-MECHANICAL MODELLING FOR THE JET DIVERTOR EXPERIMENTAL INTERPRETATION, IMPROVED PROTECTION, AND RELIABLE OPERATION D. IGLESIAS CCFE Abingdon, UK Email:

More information

DEMO diagnostics and impact on controllability

DEMO diagnostics and impact on controllability Member of the Helmholtz Association DEMO diagnostics and impact on controllability IAEA DEMO workshop, 17 th -20 th Dec 2013, Vienna (Austria) W. Biel 1, A. Dinklage 2, F. Felici 3, R. König 2, H. Meister

More information

ANALYSIS OF PLASMA FACING MATERIALS IN CONTROLLED FUSION DEVICES. Marek Rubel

ANALYSIS OF PLASMA FACING MATERIALS IN CONTROLLED FUSION DEVICES. Marek Rubel ANALYSIS OF PLASMA FACING MATERIALS IN CONTROLLED FUSION DEVICES Marek Rubel Alfvén Laboratory, Royal Institute of Technology, Association EURATOM VR, Stockholm, Sweden Acknowledgements Paul Coad and Guy

More information

Erosion and Confinement of Tungsten in ASDEX Upgrade

Erosion and Confinement of Tungsten in ASDEX Upgrade ASDEX Upgrade Max-Planck-Institut für Plasmaphysik Erosion and Confinement of Tungsten in ASDEX Upgrade R. Dux, T.Pütterich, A. Janzer, and ASDEX Upgrade Team 3rd IAEA-FEC-Conference, 4.., Daejeon, Rep.

More information

Impurity transport analysis & preparation of W injection experiments on KSTAR

Impurity transport analysis & preparation of W injection experiments on KSTAR Impurity transport analysis & preparation of W injection experiments on KSTAR J. H. Hong, H. Y. Lee, S. H. Lee, S. Jang, J. Jang, T. Jeon, H. Lee, and W. Choe ( ) S. G. Lee, C. R. Seon, J. Kim, ( ) 마스터부제목스타일편집

More information

Power balance of Lower Hybrid Current Drive in the SOL of High Density Plasmas on Alcator C-Mod

Power balance of Lower Hybrid Current Drive in the SOL of High Density Plasmas on Alcator C-Mod Power balance of Lower Hybrid Current Drive in the SOL of High Density Plasmas on Alcator C-Mod I.C. Faust, G.M. Wallace, S.G. Baek, D. Brunner, B. LaBombard, R.R. Parker, Y. Lin, S. Shiraiwa, J.L. Terry,

More information

EU Plasma-Wall Interactions Task Force

EU Plasma-Wall Interactions Task Force Recent results on material migration and fuel retention in JET V. Philipps and JET TFE co-workers* Overview on present results on erosion, deposition and fuel retention in last JET campaign (2001-2004,C5-C15)

More information

Some Notes on the Window Frame Method for Assessing the Magnitude and Nature of Plasma-Wall Contact

Some Notes on the Window Frame Method for Assessing the Magnitude and Nature of Plasma-Wall Contact Some Notes on the Window Frame Method for Assessing the Magnitude and Nature of Plasma-Wall Contact Peter Stangeby 4 September 2003 1. Fig. 1 shows an example of a suitable magnetic configuration for application

More information

Evidence for enhanced main chamber wall plasma loads in JET ITER-like Wall at high radiated fraction

Evidence for enhanced main chamber wall plasma loads in JET ITER-like Wall at high radiated fraction EUROFUSION WPJET1-PR(16) 14698 C Guillemaut et al. Evidence for enhanced main chamber wall plasma loads in JET ITER-like Wall at high radiated fraction Preprint of Paper to be submitted for publication

More information

Exploration of Configurational Space for Quasi-isodynamic Stellarators with Poloidally Closed Contours of the Magnetic Field Strength

Exploration of Configurational Space for Quasi-isodynamic Stellarators with Poloidally Closed Contours of the Magnetic Field Strength Exploration of Configurational Space for Quasi-isodynamic Stellarators with Poloidally Closed Contours of the Magnetic Field Strength V.R. Bovshuk 1, W.A. Cooper 2, M.I. Mikhailov 1, J. Nührenberg 3, V.D.

More information

Long Term Reduction of Divertor Carbon Sources in DIII-D

Long Term Reduction of Divertor Carbon Sources in DIII-D Long Term Reduction of Divertor Carbon Sources in DIII-D D.G Whyte, UCSD R. Doerner, W.P. West, R.L. Lee, N.H. Brooks, R.D. Isler, M.R. Wade, G.D. Porter APS-DPP, Seattle, Nov. 1999 NATIONAL FUSION FACILITY

More information

Plasma Spectroscopy in ISTTOK

Plasma Spectroscopy in ISTTOK Plasma Spectroscopy in ISTTOK J. Figueiredo 1, R. B. Gomes 1, T. Pereira 1, H. Fernandes 1, A. Sharakovski 2 1 Associação EURATOM/IST, Centro de Fusão Nuclear, IST, 1049-001 Lisboa, Portugal 2 Association

More information

Cross-Field Plasma Transport and Main Chamber Recycling in Diverted Plasmas on Alcator C-Mod

Cross-Field Plasma Transport and Main Chamber Recycling in Diverted Plasmas on Alcator C-Mod Cross-Field Plasma Transport and Main Chamber Recycling in Diverted Plasmas on Alcator C-Mod B. LaBombard, M. Umansky, R.L. Boivin, J.A. Goetz, J. Hughes, B. Lipschultz, D. Mossessian, C.S. Pitcher, J.L.Terry,

More information

Fusion Nuclear Science Facility (FNSF) Divertor Plans and Research Options

Fusion Nuclear Science Facility (FNSF) Divertor Plans and Research Options Fusion Nuclear Science Facility (FNSF) Divertor Plans and Research Options A.M. Garofalo, T. Petrie, J. Smith, V. Chan, R. Stambaugh (General Atomics) J. Canik, A. Sontag, M. Cole (Oak Ridge National Laboratory)

More information

ITER operation. Ben Dudson. 14 th March Department of Physics, University of York, Heslington, York YO10 5DD, UK

ITER operation. Ben Dudson. 14 th March Department of Physics, University of York, Heslington, York YO10 5DD, UK ITER operation Ben Dudson Department of Physics, University of York, Heslington, York YO10 5DD, UK 14 th March 2014 Ben Dudson Magnetic Confinement Fusion (1 of 18) ITER Some key statistics for ITER are:

More information

Comparison of tungsten fuzz growth in Alcator C-Mod and linear plasma devices

Comparison of tungsten fuzz growth in Alcator C-Mod and linear plasma devices Comparison of tungsten fuzz growth in Alcator C-Mod and linear plasma devices G.M. Wright 1, D. Brunner 1, M.J. Baldwin 2, K. Bystrov 3, R. Doerner 2, B. LaBombard 1, B. Lipschultz 1, G. de Temmerman 3,

More information

Chamber Development Plan and Chamber Simulation Experiments

Chamber Development Plan and Chamber Simulation Experiments Chamber Development Plan and Chamber Simulation Experiments Farrokh Najmabadi HAPL Meeting November 12-13, 2001 Livermore, CA Electronic copy: http://aries.ucsd.edu/najmabadi/talks UCSD IFE Web Site: http://aries.ucsd.edu/ife

More information

PISCES Laser Transient Systems

PISCES Laser Transient Systems Laser Transient Systems Karl R. Umstadter for Team Center for Energy Research University of California San Diego, USA February 11, 2009 Overview Introduction Use of Laser Heat Pulses PA Short Pulse Update

More information

BOLOMETRY FOR DIVERTOR CHARACTERIZATION AND CONTBaOL IL.; 0

BOLOMETRY FOR DIVERTOR CHARACTERIZATION AND CONTBaOL IL.; 0 GA-A22138 BOLOMETRY FOR DIVERTOR CHARACTERIZATION AND CONTBaOL IL.; 0 by A.W. LEONARD, J. GOETZ, C. FUCHS, M. MARASHEK, F. MAST, and R. RElCHLE OCTOBER 1995 GENERAL ATOMfCS DISCLAIMER This report was prepared

More information

Plasma shielding during ITER disruptions

Plasma shielding during ITER disruptions Plasma shielding during ITER disruptions Sergey Pestchanyi and Richard Pitts 1 Integrated tokamak code TOKES is a workshop with various tools objects Radiation bremsstrahlung recombination s line s cyclotron

More information

First plasma operation of Wendelstein 7-X

First plasma operation of Wendelstein 7-X First plasma operation of Wendelstein 7-X R. C. Wolf on behalf of the W7-X Team *) robert.wolf@ipp.mpg.de *) see author list Bosch et al. Nucl. Fusion 53 (2013) 126001 The optimized stellarator Wendelstein

More information

L-Mode and Inter-ELM Divertor Particle and Heat Flux Width Scaling on MAST

L-Mode and Inter-ELM Divertor Particle and Heat Flux Width Scaling on MAST CCFE-PR(13)33 J. R. Harrison, G. M. Fishpool and A. Kirk L-Mode and Inter-ELM Divertor Particle and Heat Flux Width Scaling on MAST Enquiries about copyright and reproduction should in the first instance

More information

Impact of Neon Injection on Electron Density Peaking in JET Hybrid Plasmas

Impact of Neon Injection on Electron Density Peaking in JET Hybrid Plasmas 1 P/233 Impact of Neon Injection on Electron Density Peaking in JET Hybrid Plasmas D. Frigione 1, M. Romanelli 2, C. Challis 2, J. Citrin 3, L. Frassinetti 4, J. Graves 5, J. Hobirk 6, F. Koechl 2, M.

More information

ARIES-AT Blanket and Divertor Design (The Final Stretch)

ARIES-AT Blanket and Divertor Design (The Final Stretch) ARIES-AT Blanket and Divertor Design (The Final Stretch) The ARIES Team Presented by A. René Raffray and Xueren Wang ARIES Project Meeting University of Wisconsin, Madison June 19-21, 2000 Presentation

More information

Tokamak operation at low q and scaling toward a fusion machine. R. Paccagnella^

Tokamak operation at low q and scaling toward a fusion machine. R. Paccagnella^ Tokamak operation at low q and scaling toward a fusion machine R. Paccagnella^ Consorzio RFX, Associazione Euratom-ENEA sulla Fusione, Padova, Italy ^ and Istituto Gas Ionizzati del Consiglio Nazionale

More information

Erosion and Confinement of Tungsten in ASDEX Upgrade

Erosion and Confinement of Tungsten in ASDEX Upgrade IAEA-CN-8/EXD/6- Erosion and Confinement of Tungsten in ASDEX Upgrade R. Dux, A. Janzer, T. Pütterich, and ASDEX Upgrade Team Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching

More information

Surface temperature measurement and heat load estimation for carbon targets with plasma contact and machine protection

Surface temperature measurement and heat load estimation for carbon targets with plasma contact and machine protection Surface temperature measurement and heat load estimation for carbon targets with plasma contact and machine protection A. Herrmann Max-Planck-Institut für Plasmaphysik, Boltzmannstr. 2, D-85748 Garching,

More information

Heat Flux Management via Advanced Magnetic Divertor Configurations and Divertor Detachment.

Heat Flux Management via Advanced Magnetic Divertor Configurations and Divertor Detachment. Heat Flux Management via Advanced Magnetic Divertor Configurations and Divertor Detachment E. Kolemen a, S.L. Allen b, B.D. Bray c, M.E. Fenstermacher b, D.A. Humphreys c, A.W. Hyatt c, C.J. Lasnier b,

More information

EFDA European Fusion Development Agreement - Close Support Unit - Garching

EFDA European Fusion Development Agreement - Close Support Unit - Garching Multi-machine Modelling of Divertor Geometry Effects Alberto Loarte EFDA CSU -Garching Acknowledgements: K. Borrass, D. Coster, J. Gafert, C. Maggi, R. Monk, L. Horton, R.Schneider (IPP), A.Kukushkin (ITER),

More information

Progress on Design and R&D for ITER Diagnostic Systems in Japan Domestic Agency

Progress on Design and R&D for ITER Diagnostic Systems in Japan Domestic Agency 1 ITR/P5-35 Progress on Design and R&D for ITER Diagnostic Systems in Japan Domestic Agency Y. Kawano 1, Y. Kusama 1, T. Kondoh 1, T. Hatae 1, K. Sato 1, H. Ogawa 1, M. Ishikawa 1, T. Sugie 1, E. Yatsuka

More information

Power Balance and Scaling of the Radiated Power in the Divertor and Main Plasma of Alcator C-Mod

Power Balance and Scaling of the Radiated Power in the Divertor and Main Plasma of Alcator C-Mod PFC/JA-94-15 Power Balance and Scaling of the Radiated Power in the Divertor and Main Plasma of Alcator C-Mod J.A. Goetz, B. Lipschultz, M.A. Graf, C. Kurz, R. Nachtrieb, J.A. Snipes, J.L. Terry Plasma

More information

EXD/P3-13. Dependences of the divertor and midplane heat flux widths in NSTX

EXD/P3-13. Dependences of the divertor and midplane heat flux widths in NSTX 1 Dependences of the ertor and plane heat flux widths in NSTX T.K. Gray1,2), R. Maingi 2), A.G. McLean 2), V.A. Soukhanovskii 3) and J-W. Ahn 2) 1) Oak Ridge Institute for Science and Education (ORISE),

More information

Chemical Erosion and Critical Issues for ITER

Chemical Erosion and Critical Issues for ITER Chemical Erosion and Critical Issues for ITER J. Roth Max-Planck-Institut für Plasmaphysik, Garching Chemical Erosion Studies Erosion yields: Dependence on temperature, energy and flux Emitted hydrocarbons

More information

Atomic physics in fusion development

Atomic physics in fusion development Atomic physics in fusion development The next step in fusion development imposes new requirements on atomic physics research by R.K. Janev In establishing the scientific and technological base of fusion

More information

Investigation of Water Fragments

Investigation of Water Fragments National Nuclear Research University MEPhI Federal State Autonomous Institution for Higher Education 31 Kashirskoe shosse 115409 Moscow, Russia VAT registration number, 7724068140 REG. No 1037739366477

More information

Development of Langmuir Probes on Divertor Cassettes in JT-60SA

Development of Langmuir Probes on Divertor Cassettes in JT-60SA Development of Langmuir Probes on Divertor Cassettes in JT-60SA Masakatsu FUKUMOTO, Shinji SAKURAI, Nobuyuki ASAKURA 1) and Kiyoshi ITAMI Japan Atomic Energy Agency, Naka, Ibaraki 311-0193, Japan 1) Japan

More information

and expectations for the future

and expectations for the future 39 th Annual Meeting of the FPA 2018 First operation of the Wendelstein 7-X stellarator and expectations for the future Hans-Stephan Bosch Max-Planck-Institut für Plasmaphysik Greifswald, Germany on behalf

More information

Thermographic measurements of power loads to plasma facing components at Wendelstein 7-X

Thermographic measurements of power loads to plasma facing components at Wendelstein 7-X Thermographic measurements of power loads to plasma facing components at Wendelstein 7-X M.W. Jakubowski 1, A. Ali 1, P. Drewelow 1, H. Niemann 1, F. Pisano 4, A. Puig Sitjes 1, G. Wurden 3, C. Biedermann

More information

LH Generated Hot Spots on the JET Divertor

LH Generated Hot Spots on the JET Divertor LH Generated Hot Spots on the JET Divertor K.M. Rantamäki 1, V. Petržílka 2, F. Žáček 2, A. Ekedahl 3, K. Erents 4, V. Fuchs 2, M. Goniche 3, G. Granucci 5, S.J. Karttunen 1, J. Mailloux 4, and M.-L.Mayoral

More information