Developing physics basis for the snowflake divertor in the DIII-D tokamak

Science.gov (United States)

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; Lasnier, C. J.; Makowski, M. A.; McLean, A. G.; Meyer, W. H.; Ryutov, D. D.; Kolemen, E.; Groebner, R. J.; Hyatt, A. W.; Leonard, A. W.; Osborne, T. H.; Petrie, T. W.; Watkins, J.

2018-03-01

Recent DIII-D results demonstrate that the snowflake (SF) divertor geometry (see standard divertor) enables significant manipulation of divertor heat transport for heat spreading and reduction in attached and radiative divertor regimes, between and during edge localized modes (ELMs), while maintaining good H-mode confinement. Snowflake divertor configurations have been realized in the DIII-D tokamak for several seconds in H-mode discharges with heating power P_NBI ≤slant 4 -5 MW and a range of plasma currents I_p=0.8-1.2 MA. In this work, inter-ELM transport and radiative SF divertor properties are studied. Significant impact of geometric properties on SOL and divertor plasma parameters, including increased poloidal magnetic flux expansion, divertor magnetic field line length and divertor volume, is confirmed. In the SF-minus configuration, heat deposition is affected by the geometry, and peak divertor heat fluxes are significantly reduced. In the SF-plus and near-exact SF configurations, divertor peak heat flux reduction and outer strike point heat flux profile broadening are observed. Inter-ELM sharing of power and particle fluxes between the main and additional snowflake divertor strike points has been demonstrated. The additional strike points typically receive up to 10-15% of total outer divertor power. Measurements of electron pressure and poloidal beta βp support the theoretically proposed churning mode that is driven by toroidal curvature and vertical pressure gradient in the weak poloidal field region. A comparison of the 4-4.5 MW NBI-heated H-mode plasmas with radiative SF divertor and the standard radiative divertor (both induced with additional gas puffing) shows a nearly complete power detachment and broader divertor radiated power distribution in the SF, as compared to a partial detachment and peaked localized radiation in the standard divertor. However, insignificant difference in the detachment onset w.r.t. density between the SF and the standard

D III-D divertor target heat flux measurements during Ohmic and neutral beam heating

International Nuclear Information System (INIS)

Hill, D.N.; Petrie, T.; Mahdavi, M.A.; Lao, L.; Howl, W.

1988-01-01

Time resolved power deposition profiles on the D III-D divertor target plates have been measured for Ohmic and neutral beam injection heated plasmas using fast response infrared thermography (τ ≤ 150 μs). Giant Edge Localized Modes have been observed which punctuate quiescent periods of good H-mode confinement and deposit more than 5% of the stored energy of the core plasma on the divertor armour tiles on millisecond time-scales. The heat pulse associated with these events arrives approximately 0.5 ms earlier on the outer leg of the divertor relative to the inner leg. The measured power deposition profiles are displaced relative to the separatrix intercepts on the target plates, and the peak heat fluxes are a function of core plasma density. (author). Letter-to-the-editor. 11 refs, 7 figs

Studies of power exhaust and divertor design for a 1.5 GW-level fusion power DEMO

Science.gov (United States)

Asakura, N.; Hoshino, K.; Suzuki, S.; Tokunaga, S.; Someya, Y.; Utoh, H.; Kudo, H.; Sakamoto, Y.; Hiwatari, R.; Tobita, K.; Shimizu, K.; Ezato, K.; Seki, Y.; Ohno, N.; Ueda, Y.; Joint Special TeamDEMO Design

2017-12-01

Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of {{P}sep ~ }   =  205-285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of {{P}out}   =  250 MW and the total radiation fraction at the edge, SOL and divertor ({{P}rad}/{{P}out}   =  0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load ({{q}target} ) at the attached region was reduced to ~5 MW m-2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak {{q}target} was less than 10 MW m-2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5-1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak {{q}target} of 10 MWm-2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m-2 was distributed in the major part

Numerical analysis of tungsten erosion and deposition processes under a DEMO divertor plasma

Directory of Open Access Journals (Sweden)

Yuki Homma

2017-08-01

Full Text Available Erosion reduction of tungsten (W divertor target is one of the most important research subjects for the DEMO fusion reactor design, because the divertor target has to sustain large fluence of incident particles, composed mainly of fuel ions and seeded impurities, during year-long operation period. Rate of net erosion and deposition on outer divertor target has been studied by using the integrated SOL/divertor plasma code SONIC and the kinetic full-orbit impurity transport code IMPGYRO. Two background plasmas have been used: one is lower density ni and higher temperature case and the other is higher ni and lower temperature case. Net erosion has been seen in the lower ni case. But in the higher ni case, the net erosion has been almost suppressed due to increased return rate and reduced self-sputtering yield. Following two factors are important to understand the net erosion formation: (i ratio of the 1st ionization length of sputtered W atom to the Larmor gyro radius of W+ ion, (ii balance between the friction force and the thermal force exerted on W ions. DEMO divertor design should take into account these factors to prevent target erosion.

Evaluating Stellarator Divertor Designs with EMC3

Science.gov (United States)

Bader, Aaron; Anderson, D. T.; Feng, Y.; Hegna, C. C.; Talmadge, J. N.

2013-10-01

In this paper various improvements of stellarator divertor design are explored. Next step stellarator devices require innovative divertor solutions to handle heat flux loads and impurity control. One avenue is to enhance magnetic flux expansion near strike points, somewhat akin to the X-Divertor concept in Tokamaks. The effect of judiciously placed external coils on flux deposition is calculated for configurations based on the HSX stellarator. In addition, we attempt to optimize divertor plate location to facilitate the external coil placement. Alternate areas of focus involve altering edge island size to elucidate the driving physics in the edge. The 3-D nature of stellarators complicates design and necessitates analysis of new divertor structures with appropriate simulation tools. We evaluate the various configurations with the coupled codes EMC3-EIRENE, allowing us to benchmark configurations based on target heat flux, impurity behavior, radiated power, and transitions to high recycling and detached regimes. Work supported by DOE-SC0006103.

Preliminary concept design of the divertor remote handling system for DEMO power plant

Energy Technology Data Exchange (ETDEWEB)

Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Di Gironimo, G. [ENEA/CREATE/University of Naples Federico II, 80125 Naples (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2014-11-15

Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor Mover: Hydraulic telescopic boom concept design. An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • Transportation cask conceptual studies and logistic. - Abstract: This paper is based on the remote maintenance system project (WPRM) for the demonstration fusion power reactor (DEMO). Following ITER, DEMO aims to confirm the capability of generating several hundred of MW of net electricity by 2050. The main objective of these activities is to develop an efficient and reliable remote handling (RH) system for replacing the divertor cassettes. This paper presents the preliminary results of the concept design of the divertor RH system. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections of 4 m each, and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative design of the end effector to grip and manipulate the divertor cassette are also presented in this work. Both the concepts are hydraulically actuated, basing on the ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. The main objective of this paper is to illustrate the feasibility of DEMO divertor remote maintenance operations.

Fast-wave power flow along SOL field lines in NSTX and the associated power deposition profile across the SOL in front of the antenna

International Nuclear Information System (INIS)

Perkins, R.J.; Bell, R.E.; Diallo, A.; Gerhardt, S.; Hosea, J.C.; Jaworski, M.A.; LeBlanc, B.P.; Kramer, G.J.; Maingi, R.; Phillips, C.K.; Podestà, M.; Roquemore, L.; Scotti, F.; Ahn, J.-W.; Gray, T.K.; Green, D.L.; McLean, A.; Ryan, P.M.; Jaeger, E.F.; Sabbagh, S.

2013-01-01

Fast-wave heating and current drive efficiencies can be reduced by a number of processes in the vicinity of the antenna and in the scrape-off layer (SOL). On NSTX from around 25% to more than 60% of the high-harmonic fast-wave power can be lost to the SOL regions, and a large part of this lost power flows along SOL magnetic field lines and is deposited in bright spirals on the divertor floor and ceiling. We show that field-line mapping matches the location of heat deposition on the lower divertor, albeit with a portion of the heat outside of the predictions. The field-line mapping can then be used to partially reconstruct the profile of lost fast-wave power at the midplane in front of the antenna, and the losses peak close to the last closed flux surface as well as the antenna. This profile suggests a radial standing-wave pattern formed by fast-wave propagation in the SOL, and this hypothesis will be tested on NSTX-U. RF codes must reproduce these results so that such codes can be used to understand this edge loss and to minimize RF heat deposition and erosion in the divertor region on ITER. (paper)

Analyses of erosion and re-deposition layers on graphite tiles used in the W-shaped divertor region of JT-60U

International Nuclear Information System (INIS)

Gotoh, Y.; Yagyu, J.; Masaki, K.; Kizu, K.; Kaminaga, A.; Kodama, K.; Arai, T.; Tanabe, T.; Miya, N.

2003-01-01

Erosion and re-deposition profiles were studied on graphite tiles used in the W-shaped divertor of JT-60U in June 1997-October 1998 periods, operated with all-carbon walls with boronizations and inner-private flux pumping. Continuous re-deposition layers were found neither on the dome top nor on the outer wing, while re-deposition layers of around 20 μm thickness were found on the inner wing, in the region close to the dome top. On the outer divertor target, erosion was found to be dominant: maximum erosion depth of around 20 μm was measured, while on the inner target, re-deposition was dominant: columnar structure layers of maximum thickness at around 30 μm on the inner zone while laminar/columnar-layered structures of maximum thickness around 60 μm were found on the outer zone. Poloidal distributions of the erosion depth/re-deposition layer thickness were well correlated with the frequency histograms of strike point position, which were weighted with total power of neutral beam injection, on both the outer and inner targets. Through X-ray photoelectron spectroscopy, composition of the re-deposition layers at a mid zone on the inner target were 3-4 at.% B and <0.6 at.% O, Fe, Cr, and Ni with remaining C. Boron atoms are mostly bound to C atoms but some may precipitated as boron

Divertor detachment

Science.gov (United States)

Krasheninnikov, Sergei

2015-11-01

The heat exhaust is one of the main conceptual issues of magnetic fusion reactor. In a standard operational regime the large heat flux onto divertor target reaches unacceptable level in any foreseeable reactor design. However, about two decades ago so-called ``detached divertor'' regimes were found. They are characterized by reduced power and plasma flux on divertor targets and look as a promising solution for heat exhaust in future reactors. In particular, it is envisioned that ITER will operate in a partly detached divertor regime. However, even though divertor detachment was studied extensively for two decades, still there are some issues requiring a new look. Among them is the compatibility of detached divertor regime with a good core confinement. For example, ELMy H-mode exhibits a very good core confinement, but large ELMs can ``burn through'' detached divertor and release large amounts of energy on the targets. In addition, detached divertor regimes can be subject to thermal instabilities resulting in the MARFE formation, which, potentially, can cause disruption of the discharge. Finally, often inner and outer divertors detach at different plasma conditions, which can lead to core confinement degradation. Here we discuss basic physics of divertor detachment including different mechanisms of power and momentum loss (ionization, impurity and hydrogen radiation loss, ion-neutral collisions, recombination, and their synergistic effects) and evaluate the roles of different plasma processes in the reduction of the plasma flux; detachment stability; and an impact of ELMs on detachment. We also evaluate an impact of different magnetic and divertor geometries on detachment onset, stability, in- out- asymmetry, and tolerance to the ELMs. Supported by the U.S. Department of Energy Office of Science, Office of Fusion Energy Sciences under Award Number DE-DE-FG02-04ER54739 at UCSD.

Initial study of divertor particle and heat flux width scaling in lower-single-null configuration on EAST

International Nuclear Information System (INIS)

Wang Liang; Xu Guosheng; Guo Houyang; Gan Kaifu; Gong Xianzu; Hu Liqun

2013-01-01

The dependence of divertor particle and power deposition widths on plasma current (I_p) for lower hybrid current driven (LHCD) L- and H-mode plasmas was initially studied in the Experimental Advanced Superconducting Tokamak (EAST) under a lower single null (LSN) divertor configuration. And the profile widths were obtained from the divertor triple Langmuir probe array and an infra-red (IR) camera. It is shown that the deposition widths of divertor particle and heat flux profiles both display a strong negative dependence on increasing plasma current, in L-mode, ELM-free H-mode and ELMy H-mode scenarios. The experimental results show good agreement with the heuristic SOL width model proposed by Goldston. (author)

Automated magnetic divertor design for optimal power exhaust

Energy Technology Data Exchange (ETDEWEB)

Blommaert, Maarten

2017-07-01

The so-called divertor is the standard particle and power exhaust system of nuclear fusion tokamaks. In essence, the magnetic configuration hereby 'diverts' the plasma to a specific divertor structure. The design of this divertor is still a key issue to be resolved to evolve from experimental fusion tokamaks to commercial power plants. The focus of this dissertation is on one particular design requirement: avoiding excessive heat loads on the divertor structure. The divertor design process is assisted by plasma edge transport codes that simulate the plasma and neutral particle transport in the edge of the reactor. These codes are computationally extremely demanding, not in the least due to the complex collisional processes between plasma and neutrals that lead to strong radiation sinks and macroscopic heat convection near the vessel walls. One way of improving the heat exhaust is by modifying the magnetic confinement that governs the plasma flow. In this dissertation, automated design of the magnetic configuration is pursued using adjoint based optimization methods. A simple and fast perturbation model is used to compute the magnetic field in the vacuum vessel. A stable optimal design method of the nested type is then elaborated that strictly accounts for several nonlinear design constraints and code limitations. Using appropriate cost function definitions, the heat is spread more uniformly over the high-heat load plasma-facing components in a practical design example. Furthermore, practical in-parts adjoint sensitivity calculations are presented that provide a way to an efficient optimization procedure. Results are elaborated for a fictituous JET (Joint European Torus) case. The heat load is strongly reduced by exploiting an expansion of the magnetic flux towards the solid divertor structure. Subsequently, shortcomings of the perturbation model for magnetic field calculations are discussed in comparison to a free boundary equilibrium (FBE) simulation

Automated magnetic divertor design for optimal power exhaust

International Nuclear Information System (INIS)

Blommaert, Maarten

2017-01-01

The so-called divertor is the standard particle and power exhaust system of nuclear fusion tokamaks. In essence, the magnetic configuration hereby 'diverts' the plasma to a specific divertor structure. The design of this divertor is still a key issue to be resolved to evolve from experimental fusion tokamaks to commercial power plants. The focus of this dissertation is on one particular design requirement: avoiding excessive heat loads on the divertor structure. The divertor design process is assisted by plasma edge transport codes that simulate the plasma and neutral particle transport in the edge of the reactor. These codes are computationally extremely demanding, not in the least due to the complex collisional processes between plasma and neutrals that lead to strong radiation sinks and macroscopic heat convection near the vessel walls. One way of improving the heat exhaust is by modifying the magnetic confinement that governs the plasma flow. In this dissertation, automated design of the magnetic configuration is pursued using adjoint based optimization methods. A simple and fast perturbation model is used to compute the magnetic field in the vacuum vessel. A stable optimal design method of the nested type is then elaborated that strictly accounts for several nonlinear design constraints and code limitations. Using appropriate cost function definitions, the heat is spread more uniformly over the high-heat load plasma-facing components in a practical design example. Furthermore, practical in-parts adjoint sensitivity calculations are presented that provide a way to an efficient optimization procedure. Results are elaborated for a fictituous JET (Joint European Torus) case. The heat load is strongly reduced by exploiting an expansion of the magnetic flux towards the solid divertor structure. Subsequently, shortcomings of the perturbation model for magnetic field calculations are discussed in comparison to a free boundary equilibrium (FBE) simulation. These flaws

Power deposition on misaligned edges in COMPASS

Directory of Open Access Journals (Sweden)

R. Dejarnac

2017-08-01

Full Text Available If the decision is made not to apply a toroidal chamfer to tungsten monoblocks at ITER divertor vertical targets, exposed leading edges will arise as a result of assembly tolerances between adjacent plasma-facing components. Then, the advantage of glancing magnetic field angles for spreading plasma heat flux on top surfaces is lost at the misaligned edges with an interaction occurring at near normal incidence, which can drive melting for the expected inter-ELM heat fluxes. A dedicated experiment has been performed on the COMPASS tokamak to thoroughly study power deposition on misaligned edges using inner-wall limited discharges on a special graphite tile presenting gaps and leading edges directly viewed by a high resolution infra-red camera. The parallel power flux deducted from the unperturbed measurement far from the gap is fully consistent with the observed temperature increase at the leading edge, respecting the power balance. All the power flowing into the gap is deposited at the leading edge and no mitigation factor is required to explain the thermal response. Particle-in-cell simulations show that the ion Larmor smoothing effect is weak and that the power deposition on misaligned edges is well described by the optical approximation because of an electron dominated regime associated with non-ambipolar parallel current flow.

Snowflake divertor configuration studies in National Spherical Torus Experiment

Energy Technology Data Exchange (ETDEWEB)

Soukhanovskii, V. A.; McLean, A. G.; Rognlien, T. D.; Ryutov, D. D.; Umansky, M. V. [Lawrence Livermore National Laboratory, Livermore, California 94551 (United States); Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; Menard, J. E.; Paul, S. F.; Podesta, M.; Roquemore, A. L.; Scotti, F.; Battaglia, D.; Bell, M. G.; Gates, D. A.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); and others

2012-08-15

Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4-6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width {lambda}{sub q} was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5-1 MW/m{sup 2} (from 4-7 MW/m{sup 2} in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, P{sub rad} in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.

The contribution of radio-frequency rectification to field-aligned losses of high-harmonic fast wave power to the divertor in the National Spherical Torus eXperiment

Energy Technology Data Exchange (ETDEWEB)

Perkins, R. J., E-mail: rperkins@pppl.gov; Hosea, J. C.; Jaworski, M. A.; Diallo, A.; Bell, R. E.; Bertelli, N.; Gerhardt, S.; Kramer, G. J.; LeBlanc, B. P.; Phillips, C. K.; Podestà, M.; Roquemore, L.; Taylor, G.; Wilson, J. R. [Princeton Plasma Physics Laboratory, Princeton, New Jersey 08543 (United States); Ahn, J.-W.; Gray, T. K. [Oak Ridge National Laboratory, Oak Ridge, Tennessee 37831 (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Sabbagh, S. [Department of Applied Physics and Applied Mathematics, Columbia University, New York, New York 10027 (United States)

2015-04-15

The National Spherical Torus eXperiment (NSTX) can exhibit a major loss of high-harmonic fast wave (HHFW) power along scrape-off layer (SOL) field lines passing in front of the antenna, resulting in bright and hot spirals on both the upper and lower divertor regions. One possible mechanism for this loss is RF sheaths forming at the divertors. Here, we demonstrate that swept-voltage Langmuir probe characteristics for probes under the spiral are shifted relative to those not under the spiral in a manner consistent with RF rectification. We estimate both the magnitude of the RF voltage across the sheath and the sheath heat flux transmission coefficient in the presence of the RF field. Although precise comparison between the computed heat flux and infrared (IR) thermography cannot yet be made, the computed heat deposition compares favorably with the projections from IR camera measurements. The RF sheath losses are significant and contribute substantially to the total SOL losses of HHFW power to the divertor for the cases studied. This work will guide future experimentation on NSTX-U, where a wide-angle IR camera and a dedicated set of coaxial Langmuir probes for measuring the RF sheath voltage directly will quantify the contribution of RF sheath rectification to the heat deposition from the SOL to the divertor.

Toroidal asymmetries in divertor impurity influxes in NSTX

Directory of Open Access Journals (Sweden)

F. Scotti

2017-08-01

Full Text Available Toroidal asymmetries in divertor carbon and lithium influxes were observed in NSTX, due to toroidal differences in surface composition, tile leading edges, externally-applied three-dimensional (3D fields and toroidally-localized edge plasma modifications due to radio frequency heating. Understanding toroidal asymmetries in impurity influxes is critical for the evaluation of total impurity sources, often inferred from measurements with a limited toroidal coverage. The toroidally-asymmetric lithium deposition induced asymmetries in divertor lithium influxes. Enhanced impurity influxes at the leading edge of divertor tiles were the main cause of carbon toroidal asymmetries and were enhanced during edge localized modes. Externally-applied 3D fields led to strike point splitting and helical lobes observed in divertor impurity emission, but marginal changes to the toroidally-averaged impurity influxes. Power coupled to the scrape-off layer SOL plasma during radio frequency (RF heating of H-mode discharges enhanced impurity influxes along the non-axisymmetric divertor footprint of flux tubes connecting to plasma in front of the RF antenna.

Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

International Nuclear Information System (INIS)

Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh

2013-01-01

Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes

Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake

Energy Technology Data Exchange (ETDEWEB)

Kotschenreuther, Mike; Valanju, Prashant; Covele, Brent; Mahajan, Swadesh [Institute for Fusion Studies, The University of Texas at Austin, Austin, Texas 78712 (United States)

2013-10-15

Advanced divertors are magnetic geometries where a second X-point is added in the divertor region to address the serious challenges of burning plasma power exhaust. Invoking physical arguments, numerical work, and detailed model magnetic field analysis, we investigate the magnetic field structure of advanced divertors in the physically relevant region for power exhaust—the scrape-off layer. A primary result of our analysis is the emergence of a physical “metric,” the Divertor Index DI, which quantifies the flux expansion increase as one goes from the main X-point to the strike point. It clearly separates three geometries with distinct consequences for divertor physics—the Standard Divertor (DI = 1), and two advanced geometries—the X-Divertor (XD, DI > 1) and the Snowflake (DI < 1). The XD, therefore, cannot be classified as one variant of the Snowflake. By this measure, recent National Spherical Torus Experiment and DIIID experiments are X-Divertors, not Snowflakes.

Divertor power load studies for attached L-mode single-null plasmas in TCV

NARCIS (Netherlands)

Maurizio, R.; Elmore, S.; Fedorczak, N.; Gallo, A.; Reimerdes, H.; Labit, B.; Theiler, C.; Tsui, C. K.; Vijvers, W. A. J.; TCV team,; MST1 Team,

2018-01-01

This paper investigates the power loads at the inner and outer divertor targets of attached, Ohmic L-mode, deuterium plasmas in the TCV tokamak, in various experimental situations using an Infrared thermography system. The study comprises variations of the outer divertor leg length and target flux

Characterization of deuterium retention and co-deposition of fuel with lithium on the divertor tile of EAST using laser induced breakdown spectroscopy

International Nuclear Information System (INIS)

Li, Cong; Zhao, Dongye; Hu, Zhenhua; Wu, Xingwei; Luo, Guang-Nan; Hu, Jiansheng; Ding, Hongbin

2015-01-01

A laser induced breakdown spectroscopy (LIBS) system has been developed to measure and monitor the composition evolution on plasma facing materials (PFMs) of Experimental Advanced Superconducting Tokamak (EAST). As a necessity and important proof of principle experiment, LIBS analysis has been performed for lithium–deuterium co-deposition layer diagnosis of EAST divertor tiles in lab experiments. The distribution of deuterium retention has been obtained from the depth of 0.5–4 μm in the divertor tiles. The deuterium/hydrogen concentration ratio was estimated as 0.17 ± 0.02 in lithium–deuterium co-deposition layer. Moreover, the depth profile behaviors of lithium and deuterium indicate that the deuterium retention in divertor tile came from lithium–deuterium co-deposition processes during deuterium discharge in EAST. This work would improve the understanding of deuterium retention and lithium–deuterium co-deposition mechanism and give a guidance to optimize the LIBS system which will be a unique and useful diagnostic approach in EAST 2014-campaign

DiMES Studies of Temperature Dependence of Carbon Erosion and Re-Deposition in the DIII-D Divertor

Energy Technology Data Exchange (ETDEWEB)

Rudakov, D L; Jacob, W; Krieger, K; Litnovsky, A; Philipps, V; West, W P; Wong, C C; Allen, S L; Bastasz, R J; Boedo, J A; Brooks, N H; Boivin, R L; De Temmerman, G; Fenstermacher, M E; Groth, M; Hollmann, E M; Lasnier, C J; McLean, A G; Moyer, R A; Stangeby, P C; Wampler, W R; Watkins, J G; Wienhold, P; Whaley, J

2007-03-15

A strong effect of a moderately elevated surface temperature on net carbon deposition and deuterium co-deposition in the DIII-D divertor was observed under detached conditions. A DiMES sample with a gap 2 mm wide and 18 mm deep was exposed to lower-single-null (LSN) L-mode plasmas first at room temperature, and then at 200 C. At the elevated temperature, deuterium co-deposition in the gap was reduced by an order of magnitude. At the plasma-facing surface of the heated sample net carbon erosion was measured at a rate of 3 nm/s, whereas without heating net deposition is normally observed under detachment. In a related experiment three sets of molybdenum mirrors recessed 2 cm below the divertor floor were exposed to identical LSN ELMy H-mode discharges. The first set of mirrors exposed at ambient temperature exhibited net carbon deposition at a rate of up to 3.7 nm/s and suffered a significant drop in reflectivity. In contrast, two other mirror sets exposed at elevated temperatures between 90 C and 175 C exhibited practically no carbon deposition.

Experimental study of heating scheme effect on the inner divertor power footprint widths in EAST lower single null discharges

Science.gov (United States)

Deng, G. Z.; Xu, J. C.; Liu, X.; Liu, X. J.; Liu, J. B.; Zhang, H.; Liu, S. C.; Chen, L.; Yan, N.; Feng, W.; Liu, H.; Xia, T. Y.; Zhang, B.; Shao, L. M.; Ming, T. F.; Xu, G. S.; Guo, H. Y.; Xu, X. Q.; Gao, X.; Wang, L.

2018-04-01

A comprehensive work of the effects of plasma current and heating schemes on divertor power footprint widths is carried out in the experimental advanced superconducting tokamak (EAST). The divertor power footprint widths, i.e., the scrape-off layer heat flux decay length λ q and the heat spreading S, are crucial physical and engineering parameters for fusion reactors. Strong inverse scaling of λ q and S with plasma current have been demonstrated for both neutral beam (NB) and lower hybrid wave (LHW) heated L-mode and H-mode plasmas at the inner divertor target. For plasmas heated by the combination of the two kinds of auxiliary heating schemes (NB and LHW), the divertor power widths tend to be larger in plasmas with higher ratio of LHW power. Comparison between experimental heat flux profiles at outer mid-plane (OMP) and divertor target for NB heated and LHW heated L-mode plasmas reveals that the magnetic topology changes induced by LHW may be the main reason to the wider divertor power widths in LHW heated discharges. The effect of heating schemes on divertor peak heat flux has also been investigated, and it is found that LHW heated discharges tend to have a lower divertor peak heat flux compared with NB heated discharges under similar input power. All these findings seem to suggest that plasmas with LHW auxiliary heating scheme are better heat exhaust scenarios for fusion reactors and should be the priorities for the design of next-step fusion reactors like China Fusion Engineering Test Reactor.

Advanced divertor configurations with large flux expansion

Energy Technology Data Exchange (ETDEWEB)

Soukhanovskii, V.A., E-mail: vlad@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA (United States); Bell, R.E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); McLean, A. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Menard, J.E.; Paul, S.F.; Podesta, M. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Raman, R. [University of Washington, Seattle, WA (United States); Ryutov, D.D. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Scotti, F.; Kaita, R. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Maingi, R. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mueller, D.M.; Roquemore, A.L. [Princeton Plasma Physics Laboratory, Princeton, NJ (United States); Reimerdes, H.; Canal, G.P. [Ecole Polytechnique Fédérale de Lausanne, Centre de Recherches en Physique des Plasmas, Association Euratom Confédération Suisse, Lausanne (Switzerland); and others

2013-07-15

Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3–7 MW/m{sup 2} to 0.5–1 MW/m{sup 2} was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L–H power threshold, enhanced stability of the peeling–ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2–3) Type I ELM frequency and slightly increased (20–30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX

Engineering conceptual design of CFETR divertor

Energy Technology Data Exchange (ETDEWEB)

Peng, Xuebing, E-mail: pengxb@ipp.cas.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Ye, Minyou [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China); Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China); Mao, Xin [Institute of Plasma Physics, Chinese Academy of Sciences, Shushanhu Road 350, 230031 Hefei Anhui (China); Chen, Peiming; Qian, Xinyuan [School of Nuclear Science and Technology, University of Science and Technology of China, Jinzhai Road 96, 230026 Hefei Anhui (China)

2015-10-15

Highlights: • Three divertor structures for two plasma configurations, ITER-like and snowflake. • Property of enlarging wet area for all three divertors is analyzed. • The divertor accommodating with both the plasma configurations is unfeasible. • Divertor cooling system is developed. - Abstract: The China Fusion Engineering Test Reactor (CFETR), which is in conceptual design phase, aims at producing fusion power of 50–200 MW with tritium breeding ratio of ∼1.2 and duty cycle time of 0.3–0.5. Its designed main parameters are major/minor radii of 5.7 m/1.6 m and plasma current of 10 MA. Although the fusion power is lower than the one of ITER, the relative smaller machine dimensions and planed much higher auxiliary heating power of 100–140 MW make that the power exhausting for the CFETR divertor is a very critical issue. To solve this issue, the divertor should be better designed with advanced physical operation mode, advanced configuration/geometry or high efficient cooling structure. In the paper, much effort was put on the divertor configuration and geometry. With designed magnet system, three divertor configurations can be realized, ITER-like, snowflake and super-X. However, considering structural design feasibility and remote handling compatibility, only the first two configurations were selected for the first step of engineering design. Three divertors were designed. They have different first wall geometries to accommodate with different plasma configurations, one for the ITER-like, one for the snowflake and the third one for both the configurations. All three divertors employ the same cassette body as the support and the cooling water manifold for the first wall. This feature simplifies the interface of the divertor to other components in the vacuum vessel. Besides, the cooling structure and the remote maintenance concept are also introduced in the paper.

Advanced divertor experiments on DIII-D

International Nuclear Information System (INIS)

Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

1991-04-01

The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix ''strike'' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. 5 refs., 5 figs

Advanced divertor experiments on DIII-D

International Nuclear Information System (INIS)

Schaffer, M.J.; Mahdavi, M.A.; Osborne, T.; Petrie, T.W.; Stambaugh, R.D.; Buchenauer, D.; Hill, D.N.; Klepper, C.C.

1991-01-01

The poloidal divertor is presently favored for next-step, high-power tokamaks. The DIII-D Advanced Divertor Program (ADP) aims to gain increased control over the divertor plasma and tokamak boundary conditions. This paper reports experiments done in the first phase of the ADP. The DIII-D lower divertor was modified by the addition of a toroidally symmetric, graphite-armoured, water-cooled divertor-biasing ring electrode at the entrance to a gas plenum. (In the past DIII-D operated with an open divertor.) The plenum will eventually contain a He cryogenic loop for active divertor pumping. The separatrix 'strike' position is controlled by the lower poloidal field shaping coils and can be varied smoothly from the ring electrode upper surface to the divertor floor far from the entrance aperture. External power, at up to 550 V and 8 kA separately, has been applied to the electrode to date. (author) 5 refs., 5 figs

Measurement of LHCD edge power deposition through modulation techniques on Alcator C-Mod

Energy Technology Data Exchange (ETDEWEB)

Faust, I. C.; Brunner, D.; LaBombard, B.; Parker, R. R.; Baek, S. G.; Chilenksi, M. A.; Hubbard, A.; Hughes, J. W.; Terry, J. L.; Shiraiwa, S.; Walk, J. R.; Wallace, G. M.; Whyte, D. G. [MIT Plasma Science and Fusion Center, Cambridge, MA USA (United States); Edlund, E. [Princeton Plasma Physics Laboratory, Princeton, NJ USA (United States)

2015-12-10

The efficiency of LHCD on Alcator C-Mod drops exponentially with line average density. At reactor relevant densities (> 1 · 1020 [m{sup −3}]) no measurable current is driven. While a number of causes have been suggested, no specific mechanism has been shown to be responsible for the loss of current drive at high density. Fast modulation of the LH power was used to isolate and quantify the LHCD deposition within the plasma. Measurements from these plasmas provide unique evidence for determining a root cause. Modulation of LH power in steady plasmas exhibited no correlated change in the core temperature. A correlated, prompt response in the edge suggests that the loss in efficiency is related to a edge absorption mechanism. This follows previous results which found the generation of n{sub ||}-independent SOL currents. Multiple Langmuir probe array measurements of the conducted heat conclude that the lost power is deposited near the last closed flux surface. The heat flux induced by LH waves onto the outer divertor is calculated. Changes in the neutral pressure, ionization and hard X-ray emission at high density highlight the importance of the active divertor in the loss of efficiency. Results of this study implicate a mechanism which may occur over multiple passes, leading to power absorption near the LCFS.

The control of divertor carbon erosion/redeposition in the DIII-D tokamak

International Nuclear Information System (INIS)

Whyte, D.G.; West, W.P.; Wong, C.P.C.

2001-01-01

The DIII-D tokamak has demonstrated an operational scenario where the graphite-covered divertor is free of net erosion. Reduction of divertor carbon erosion is accomplished using a low temperature (detached) divertor plasma that eliminates physical sputtering. Likewise, the carbon source rate arising from chemical erosion is found to be very low in the detached divertor. Near strikepoint regions, the rate of carbon deposition is ∼3 cm/burn-year, with a corresponding hydrogenic codeposition rate >1kg/m 2 /burn-year; rates both problematic for steady-state fusion reactors. The carbon net deposition rate in the divertor is consistent with carbon arriving from the core plasma region. Carbon influx from the main wall is measured to be relatively large in the high-density detached regime and is of sufficient magnitude to account for the deposition rate in the divertor. Divertor redeposition is therefore determined by non-divertor erosion and transport. Despite the success in reducing divertor erosion on DIII-D with detachment, no significant reduction is found in the core plasma carbon density, illustrating the importance of non-divertor erosion and the complex coupling between erosion/redeposition and impurity plasma transport. (author)

In situ measurement of erosion/deposition in the DIII-D divertor by colorimetry

International Nuclear Information System (INIS)

Weschenfelder, F.; Wienhold, P.; Winter, J.

1996-01-01

Colorimetry was introduced into the DIII-D tokamak to measure in situ the growth and erosion of transparent wall coatings (a-C:H) on the divertor. The colorimetric measurement system consisting of a halogen light source, a set of three filters and a black/white camera is described together with a first erosion measurement. An insertable graphite sample with a diameter of 4.7 cm was precoated with a 300 nm thick amorphous carbon film and was exposed in the divertor for several discharges with its surface coplanar to the surrounding graphite tiles. For each of the discharges the plasma strike point was moved onto the sample for 1 s to erode the coating. Between the discharges a camera signal with each filter was recorded and the film thickness was evaluated along a radial line across the DIMES sample. Thus it has been possible for the first time to measure erosion and deposition of divertor material in situ and shot-by-shot. The average peak heat flux with the strike point on DIMES was about 110 W cm -2 . The measurement shows a strong decrease in the film thickness almost over the entire sample with an average erosion rate of ∼ 9 nm s -1 . (Author)

NSTX Plasma Response to Lithium Coated Divertor

Energy Technology Data Exchange (ETDEWEB)

H.W. Kugel, M.G. Bell, J.P. Allain, R.E. Bell, S. Ding, S.P. Gerhardt, M.A. Jaworski, R. Kaita, J. Kallman, S.M. Kaye, B.P. LeBlanc, R. Maingi, R. Majeski, R. Maqueda, D.K. Mansfield, D. Mueller, R. Nygren, S.F. Paul, R. Raman, A.L. Roquemore, S.A. Sabbagh, H. Schneider, C.H. Skinner, V.A. Soukhanovskii, C.N. Taylor, J.R. Timberlak, W.R. Wampler, L.E. Zakharov, S.J. Zweben, and the NSTX Research Team

2011-01-21

NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Zeff and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, <0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.

NSTX plasma response to lithium coated divertor

International Nuclear Information System (INIS)

Kugel, H.W.; Bell, M.G.; Allain, J.P.; Bell, R.E.; Ding, S.; Gerhardt, S.P.; Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.M.; LeBlanc, B.P.; Maingi, Rajesh; Majeski, R.; Maqueda, R.J.; Mansfield, D.K.; Mueller, D.; Nygren, R.E.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Sabbagh, S.A.; Schneider, H.; Skinner, C.H.; Soukhanovskii, V.A.; Taylor, C.N.; Timberlake, J.; Wampler, W.R.; Zakharov, L.E.; Zweben, S.J.

2011-01-01

NSTX experiments have explored lithium evaporated on a graphite divertor and other plasma-facing components in both L- and H- mode confinement regimes heated by high-power neutral beams. Improvements in plasma performance have followed these lithium depositions, including a reduction and eventual elimination of the HeGDC time between discharges, reduced edge neutral density, reduced plasma density, particularly in the edge and the SOL, increased pedestal electron and ion temperature, improved energy confinement and the suppression of ELMs in the H-mode. However, with improvements in confinement and suppression of ELMs, there was a significant secular increase in the effective ion charge Z(eff) and the radiated power in H-mode plasmas as a result of increases in the carbon and medium-Z metallic impurities. Lithium itself remained at a very low level in the plasma core, < 0.1%. Initial results are reported from operation with a Liquid Lithium Divertor (LLD) recently installed.

ARIES-III divertor engineering design

International Nuclear Information System (INIS)

Wong, C.P.C.; Schultz, K.R.; Cheng, E.T.; Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S.; Herring, J.S.; Valenti, M.; Steiner, D.

1992-01-01

This paper reports the engineering design of the ARIES-III double- null divertor. The divertor coolant tubes are made from W-3Re alloy and cooled by subcooled flow boiling of organic coolant. A coating of 4 mm thick tungsten is plasma sprayed onto the divertor surface. This W layer can withstand the thermal deposition of a few disruptions. At a maximum surface heat flux of 5.4 MW/m 2 , a conventional divertor design can be used. The divertor surface is contoured to have a constant heat flux of 5.4 MW/m 2 . The net erosion of the W-surface was found to be negligible at about 0.1 mm/year. After 3 years of operation, the W-3Re alloy ARIES-III divertor can be disposed of as Class A waste. In order to control the prompt dose release at site boundary to less than 200 Rem, isotopic tailoring of the W-alloy will be needed

ARIES-III divertor engineering design

Energy Technology Data Exchange (ETDEWEB)

Wong, C.P.C.; Schultz, K.R. [General Atomics, San Diego, CA (United States); Cheng, E.T. [TSI Research, Solana Beach, CA (United States); Grotz, S.; Hasan, M.A.; Najmabadi, F.; Sharafat, S. [California Univ., Los Angeles, CA (United States). Dept. of Mechanical, Aerospace and Nuclear Engineering; Brooks, J.N.; Ehst, D.A.; Sze, D.K. [Argonne National Lab., IL (United States); Herring, J.S. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Valenti, M.; Steiner, D. [Rensselaer Polytechnic Inst., Troy, NY (United States). Plasma Dynamics Lab.

1992-01-01

This paper reports the engineering design of the ARIES-III double- null divertor. The divertor coolant tubes are made from W-3Re alloy and cooled by subcooled flow boiling of organic coolant. A coating of 4 mm thick tungsten is plasma sprayed onto the divertor surface. This W layer can withstand the thermal deposition of a few disruptions. At a maximum surface heat flux of 5.4 MW/m{sup 2}, a conventional divertor design can be used. The divertor surface is contoured to have a constant heat flux of 5.4 MW/m{sup 2}. The net erosion of the W-surface was found to be negligible at about 0.1 mm/year. After 3 years of operation, the W-3Re alloy ARIES-III divertor can be disposed of as Class A waste. In order to control the prompt dose release at site boundary to less than 200 Rem, isotopic tailoring of the W-alloy will be needed.

A review of progress towards radiative divertor

International Nuclear Information System (INIS)

Zagorski, Roman

1997-07-01

A solution of the problem of the power and particle exhaust from the next step tokamaks, will require new techniques which redistribute the power entering the SOL onto much larger surface area than conventional divertor design permits, while maintaining good impurity retention in divertor volume and allowing for efficient helium pumping. Progress made in developing such techniques is discussed. Status of the modelling studies of dynamic gas target divertor and impurity seeded radiating divertors is presented. Recent results of experiments on radiative and gas target divertors are reviewed

Integrated simulations of H-mode operation in ITER including core fuelling, divertor detachment and ELM control

Science.gov (United States)

Polevoi, A. R.; Loarte, A.; Dux, R.; Eich, T.; Fable, E.; Coster, D.; Maruyama, S.; Medvedev, S. Yu.; Köchl, F.; Zhogolev, V. E.

2018-05-01

ELM mitigation to avoid melting of the tungsten (W) divertor is one of the main factors affecting plasma fuelling and detachment control at full current for high Q operation in ITER. Here we derive the ITER operational space, where ELM mitigation to avoid melting of the W divertor monoblocks top surface is not required and appropriate control of W sources and radiation in the main plasma can be ensured through ELM control by pellet pacing. We apply the experimental scaling that relates the maximum ELM energy density deposited at the divertor with the pedestal parameters and this eliminates the uncertainty related with the ELM wetted area for energy deposition at the divertor and enables the definition of the ITER operating space through global plasma parameters. Our evaluation is thus based on this empirical scaling for ELM power loads together with the scaling for the pedestal pressure limit based on predictions from stability codes. In particular, our analysis has revealed that for the pedestal pressure predicted by the EPED1  +  SOLPS scaling, ELM mitigation to avoid melting of the W divertor monoblocks top surface may not be required for 2.65 T H-modes with normalized pedestal densities (to the Greenwald limit) larger than 0.5 to a level of current of 6.5–7.5 MA, which depends on assumptions on the divertor power flux during ELMs and between ELMs that expand the range of experimental uncertainties. The pellet and gas fuelling requirements compatible with control of plasma detachment, core plasma tungsten accumulation and H-mode operation (including post-ELM W transient radiation) have been assessed by 1.5D transport simulations for a range of assumptions regarding W re-deposition at the divertor including the most conservative assumption of zero prompt re-deposition. With such conservative assumptions, the post-ELM W transient radiation imposes a very stringent limit on ELM energy losses and the associated minimum required ELM frequency. Depending on

Divertor radiation in the ASDEX upgrade tokamak

Energy Technology Data Exchange (ETDEWEB)

Sehmer, Till; Bernert, Matthias; Koll, Juergen; Meister, Hans; Wischmeier, Marco; Fantz, Ursel [Max-Planck-Institut fuer Plasmaphysik, Boltzmannstrasse 2, 85748 Garching (Germany); Reimold, Felix [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, 52425 Juelich (Germany); Collaboration: The ASDEX Upgrade Team

2016-07-01

To reduce in ITER the expected power flux density onto the divertor target, the plasma-wall interaction in the divertor needs to be strongly reduced. The fundamental path to achieve this is using radiation from seeded impurities, whereas the localization of this radiation (e.g. inside/outside confined region), which could have an impact onto the power balance, is a key challenge. The absolute radiated power distribution can be measured by foil bolometers. To study at the ASDEX Upgrade tungsten divertor the localization and quantification of radiation, the respective line of sight density of the bolometers has been improved by two additional cameras. The divertor radiation enhanced by nitrogen (N{sub 2}) seeding has been investigated, using variations of (1) the external heating power or (2) the N{sub 2} seeding rate. While in both cases the inner divertor stays fully detached, measurements indicate that the region of dominant radiation moves from the inner divertor through the X-Point into the confined region. In the outer divertor however, the measurements indicate either an immediate upwards shift or a continuous movement of the radiation away from the target, depending on experimental conditions.

Divertor Design and Physics Issues of Huge Power Handling for SlimCS Demo Reactor

Energy Technology Data Exchange (ETDEWEB)

Asakura, N.; Hoshino, K.; Tobita, K.; Someya, Y.; Utoh, H.; Nakamura, M., E-mail: asakura.nobuyuki@jaea.go.jp [Japan Atomic Energy Agency, Rokkasho (Japan); Shimizu, K. [Japan Atomic Energy Agency, Naka (Japan); Takizuka, T. [Osaka University, Osaka (Japan)

2012-09-15

Full text: Power exhaust scenario for a 3 GW class fusion reactor with the ITER-size plasma has been developed with enhancing the radiation loss from seeding impurity. Transport of plasma, impurity and neutrals was simulated self-consistently, for the first time, under the Demo divertor condition using an integrated divertor simulation code SONIC. The total heat load, q{sub target}, was evaluated including radiation power load and neutral load, in addition to the plasma heat load. It was found that heat and particle diffusion coefficients significantly affect the plasma detachment. For the case of increasing the coefficients by the factor of two, peak q{sub target} is reduced from 18 MW/m{sup 2} to below the engineering design level of 10 MW/m{sup 2}, while the characteristic width of the heat flux at the midplane SOL increases slightly from 2.2 to 2.7 mm. It was also found that that enhancement of the local {chi} and D at the outer SOL affects a reduction in the peak q{sub target} near the separatrix. Effects of the divertor geometry such as the divertor leg were investigated. Outer divertor leg length was extended to 2.7 m, while the magnetic flux expansion at the target is reduced to a half compared to the reference case of 1.8 m. Large radiation volume is shifted further upstream from the target due to a reduction in T{sub e}. The peak q{sub target} decreases to 10 MW/m{sup 2} due to reduction in both the plasma heat load and the radiation power load. (author)

Characterizing Low-Z erosion and deposition in the DIII-D divertor using aluminum

Directory of Open Access Journals (Sweden)

C.P. Chrobak

2017-08-01

Full Text Available We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth and realistic (rough surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms.

Dissipative divertor operation in the Alcator C-Mod tokamak

International Nuclear Information System (INIS)

Lipschultz, B.; Goetz, J.; LaBombard, B.; McCracken, G.M.; Terry, J.L.; Graf, M.; Granetz, R.S.; Jablonski, D.; Kurz, C.; Niemczewski, A.; Snipes, J.

1995-01-01

The achievement of large volumetric power losses (dissipation) in the Alcator C-Mod divertor region is demonstrated in two operational modes: radiative divertor and detached divertor. During radiative divertor operation, the fraction of SOL power lost by radiation is P R /P SOL ∼0.8 with single null plasmas, n e 20 m -3 and I p e,div ≤6x10 20 m -3 . As the divertor radiation and density increase, the plasma eventually detaches abruptly from the divertor plates: I SAT drops at the target and the divertor radiation peak moves to the X-point region. Probe measurements at the divertor plate show that the transition occurs when T e ∼5 eV. The critical n e for detachment depends linearly on the input power. This abrupt divertor detachment is preceded by a comparatively long period ( similar 1-200 ms) where a partial detachment is observed to grow at the outer divertor plate. ((orig.))

A Lithium Vapor Box Divertor Similarity Experiment

Science.gov (United States)

Cohen, Robert A.; Emdee, Eric D.; Goldston, Robert J.; Jaworski, Michael A.; Schwartz, Jacob A.

2017-10-01

A lithium vapor box divertor offers an alternate means of managing the extreme power density of divertor plasmas by leveraging gaseous lithium to volumetrically extract power. The vapor box divertor is a baffled slot with liquid lithium coated walls held at temperatures which increase toward the divertor floor. The resulting vapor pressure differential drives gaseous lithium from hotter chambers into cooler ones, where the lithium condenses and returns. A similarity experiment was devised to investigate the advantages offered by a vapor box divertor design. We discuss the design, construction, and early findings of the vapor box divertor experiment including vapor can construction, power transfer calculations, joint integrity tests, and thermocouple data logging. Heat redistribution of an incident plasma-based heat flux from a typical linear plasma device is also presented. This work supported by DOE Contract No. DE-AC02-09CH11466 and The Princeton Environmental Institute.

Modelling of island divertor physics and comparison to W7-AS experimental results

International Nuclear Information System (INIS)

Feng, Y.; Sardei, F.; Grigull, P.; McCormick, K.; Giannone, L.; Kisslinger, J.; Reiter, D.; Igitkhanov, Y.; Wenzel, U.

2003-01-01

Extensive parameter studies have been carried out with the EMC3-EIRENE code. Major code predictions, namely the absence of high recycling prior to detachment, additional momentum losses associated with the specific island divertor geometry and the jump of the radiation at detachment transition have been verified by the W7-AS divertor experiments. Measurements and simulations are compared for high density, high power W7-AS divertor discharges and the physics related to rollover and detachment is discussed in detail. Local comparisons with the W7-AS experiment have been started with a new code version accounting for the real open-island geometry. Specifically, the observed asymmetric power unloading of the target plates at detachment transition could now be reproduced and explained. Agreement with the experiment was also found for the unexpected spatial structure of particle deposition by including classical ExB drifts into the code

Thermal deposition analysis during disruptions on DIII-D using infrared scanners

International Nuclear Information System (INIS)

Lee, R.L.; Hyatt, A.W.; Kellman, A.G.; Taylor, P.L.; Lasnier, C.J.

1995-12-01

The DIII-D tokamak generates plasma discharges with currents up to 3 MA and auxiliary input power up to 20 MW from neutral beams and 4 MW from radio frequency systems. In a disruption, a rapid loss of the plasma current and internal thermal energy occurs and the energy is deposited onto the torus graphite wall. Quantifying the spatial and temporal characteristics of the heat deposition is important for engineering and physics-related issues, particularly for designing future machines such as ITER. Using infrared scanners with a time resolution of 120 micros, measurements of the heat deposition onto the all-graphite walls of DIII-D during two types of disruptions have been made. Each scanner contains a single point detector sensitive to 8--12 microm radiation, allowing surface temperatures from 20 C to 2,000 C to be measured. A zinc selenide window that transmits in the infrared is used as the vacuum window. Views of the upper and lower divertor regions and the centerpost provide good coverage of the first wall for single and double null divertor discharges. During disruptions, the thermal energy is not deposited evenly onto the inner surface of the tokamak, but is deposited primarily in the divertor region when operating diverted discharges. Analysis of the heat deposition during a radiative collapse disruption of a 1.5 MA discharge revealed power densities of 300--350 MW/m 2 in the divertor region. During the thermal quench of the disruption, the energy deposited onto the divertor region was more than 70% of the stored thermal energy in the discharge prior to the disruption. The spatial distribution and temporal behavior of power deposition during high β disruptions will also be presented

ADX: a high field, high power density, advanced divertor and RF tokamak

Science.gov (United States)

LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

2015-05-01

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept

Effects of neutral gas collisions on the power transmission factor at the divertor sheath

International Nuclear Information System (INIS)

Futch, A.H.; Matthews, G.F.; Buchenauer, D.; Hill, D.N.; Jong, R.A.; Porter, G.D.

1992-01-01

We show that charge-exchange and other ion-neutral collision can reduce the power transmission factor of the plasma sheath, thereby lowering the ion impact energy and target plate sputtering. The power transmission factor relates the heat flux reaching the divertor target to the plasma density and temperature just in front of the surface: δ=Q surf /J ew k T e . Experimental data from the DIII-D tokamak suggests that δ could be as low as 2-3 near the region of peak divertor particle flux, instead of the 7-8 expected from usual sheath theory. Several effects combine to allow ion-neutral interactions to be important in the divertor plasma sheath. The shallow angle of incidence of the magnetic field (1-3deg in DIII-D) leads to the spatial extension of the sheath from approximately ρ i ∝1 mm normal to the plate to several centimeters along the field lines. Ionization reduces the sheath potential, and collisions reduce the ion impact energy. (orig.)

Power transport to the poloidal divertor experiment scoop limiter

International Nuclear Information System (INIS)

Kugel, H.W.; Budny, R.; Fonck, R.

1987-01-01

Power transport to the Poloidal Divertor Experiment graphite scoop limiter was measured during both ohmic- and neutral-beam-heated discharges by observing its front face temperatures using an infrared camera. Measurements were made as a function of a plasma density, current, position, fueling mode, and heating power for both co- and counter-neutral beam injection. The measured thermal load on the scoop limiter was 25 to 50%. of the total plasma heating power. The measured peak front face midplane temperature was 1500 0 C, corresponding to a peak surface power density of 3 kW/cm/sup 2/. This power density implies an effective parallel power flow of 54 kW/cm/sup 2/ in agreement with the radial power distribution extrapolated from television Thomson scattering and calorimetry measurements

Tungsten erosion and redeposition in the all-tungsten divertor of ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Mayer, M; Krieger, K; Matern, G; Neu, R; Rasinski, M; Rohde, V; Sugiyama, K; Wiltner, A [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, Boltzmannstrasse 2, 85748 Garching (Germany); Andrzejczuk, M; Fortuna-Zalesna, E; Kurzydlowski, K J; Zielinski, W [Faculty of Materials Science and Engineering, Warsaw University of Technology, Association EURATOM-IPPLM, 02-507 Warsaw (Poland); Hakola, A; Koivuranta, S; Likonen, J [VTT Materials for Power Engineering, EURATOM Association, PO Box 1000, FI-02044 VTT (Finland); Ramos, G [CICATA-Qro, Instituto Politecnico Nacional, Queretaro (Mexico); Dux, R, E-mail: matej.mayer@ipp.mpg.de

2009-12-15

Net erosion and deposition of tungsten (W) in the ASDEX Upgrade divertor were determined after the 2007 campaign by using thin W marker stripes. ASDEX Upgrade had full-W plasma-facing components during this campaign. The inner divertor and the roof baffle were net W deposition areas with a maximum deposition of about 1x10{sup 18} W-atoms cm{sup -2} in the private flux region below the inner strike point. Net erosion of W was observed in the whole outer divertor, with the largest erosion close to the outer strike point. Only a small fraction of the W eroded in the main chamber and in the outer divertor was found in redeposits in the inner divertor, while a large fraction was either redeposited at unidentified places in the main chamber or has formed dust.

Snowflake Divertor Configuration in NSTX

International Nuclear Information System (INIS)

Soukhanovskii, V.A.; Ahn, Joonwook; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.; Maingi, Rajesh; Maqueda, R.J.; McLean, Adam G.; Menard, J.E.; Mueller, D.; Paul, S.F.; Raman, R.; Roquemore, L.; Ryutov, D.D.; Scott, H.A.

2011-01-01

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

'Snowflake' divertor configuration in NSTX

International Nuclear Information System (INIS)

Soukhanovskii, V.A.; Ahn, J.-W.; Bell, R.E.; Gates, D.A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H.W.; LeBlanc, B.P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J.E.; Mueller, D.M.; Paul, S.F.; Raman, R.; Roquemore, A.L.; Ryutov, D.D.; Scott, H.A.

2011-01-01

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel 'snowflake' divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

"Snowflake" divertor configuration in NSTX

Science.gov (United States)

Soukhanovskii, V. A.; Ahn, J.-W.; Bell, R. E.; Gates, D. A.; Gerhardt, S.; Kaita, R.; Kolemen, E.; Kugel, H. W.; Leblanc, B. P.; Maingi, R.; Maqueda, R.; McLean, A.; Menard, J. E.; Mueller, D. M.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Ryutov, D. D.; Scott, H. A.

2011-08-01

Steady-state handling of divertor heat flux is a critical issue for present and future conventional and spherical tokamaks with compact high power density divertors. A novel "snowflake" divertor (SFD) configuration that takes advantage of magnetic properties of a second-order poloidal null has been predicted to have a larger plasma-wetted area and a larger divertor volume, in comparison with a standard first-order poloidal X-point divertor configuration. The SFD was obtained in 0.8 MA, 4-6 MW NBI-heated H-mode discharges in NSTX using two divertor magnetic coils. The SFD led to a partial detachment of the outer strike point even in low-collisionality scrape-off layer plasma obtained with lithium coatings in NSTX. Significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with good core confinement and MHD properties.

Examining Innovative Divertor and Main Chamber Options for a National Divertor Test Tokamak

Science.gov (United States)

Labombard, B.; Umansky, M.; Brunner, D.; Kuang, A. Q.; Marmar, E.; Wallace, G.; Whyte, D.; Wukitch, S.

2016-10-01

The US fusion community has identified a compelling need for a National Divertor Test Tokamak. The 2015 Community Planning Workshop on PMI called for a national working group to develop options. Important elements of a NDTT, adopted from the ADX concept, include the ability to explore long-leg divertor `solutions for power exhaust and particle control' (Priority Research Direction B) and to employ inside-launch RF actuators combined with double-null topologies as `plasma solution for main chamber wall components, including tools for controllable sustained operation' (PRD-C). Here we examine new information on these ideas. The projected performance of super-X and X-point target long-leg divertors is looking very promising; a stable fully-detached divertor condition handling an order-of-magnitude increase in power handling over conventional divertors may be possible. New experiments on Alcator C-Mod are addressing issues of high-field side versus low-field side heat flux sharing in double-null topologies and the screening of impurities that might originate from RF actuators placed in the high-field side - both with favorable results. Supported by USDoE Awards DE-FC02-99ER54512 and DE-AC52-07NA27344.

Thermal effects of runaway electrons in an armoured divertor

International Nuclear Information System (INIS)

Stad, R.C.L. van der.

1993-12-01

This report describes the results of a numerical thermal analysis of the heat deposition of runaway electrons accompanying plasma disruptions in a armoured divertor. The divertor concepts studied are carbon on molybdenum and beryllium on copper. The conclusion is that the runaway electrons can cause melting of the armour as well as melting of the structure and can damage the divertor severely. (orig.)

Modeling detachment physics in the NSTX snowflake divertor

Energy Technology Data Exchange (ETDEWEB)

Meier, E.T., E-mail: emeier@wm.edu [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Bell, R.E.; Diallo, A.; Kaita, R.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States); Podestà, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08540 (United States); Rognlien, T.D.; Scotti, F. [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

2015-08-15

The snowflake divertor is a proposed technique for coping with the tokamak power exhaust problem in next-step experiments and eventually reactors, where extreme power fluxes to material surfaces represent a leading technological and physics challenge. In lithium-conditioned National Spherical Torus Experiment (NSTX) discharges, application of the snowflake divertor typically induced partial outer divertor detachment and severalfold heat flux reduction. UEDGE is used to analyze and compare conventional and snowflake divertor configurations in NSTX. Matching experimental upstream profiles and divertor measurements in the snowflake requires target recycling of 0.97 vs. 0.91 in the conventional case, implying partial saturation of the lithium-based pumping mechanism. Density scans are performed to analyze the mechanisms that facilitate detachment in the snowflake, revealing that increased divertor volume provides most of the parallel heat flux reduction. Also, neutral gas power loss is magnified by the increased wetted area in the snowflake, and plays a key role in generating volumetric recombination.

The effect of feedback-controlled divertor nitrogen seeding on the boundary plasma and power exhaust channel width in Alcator C-Mod

Science.gov (United States)

Labombard, B.; Brunner, D.; Kuang, A. Q.; McCarthy, W.; Terry, J. L.

2017-10-01

The scrape-off layer (SOL) power channel width, λq, is projected to be 0.5 mm in power reactors, based on multi-machine measurements of divertor target heat fluxes in H-mode at low levels of divertor dissipation. An important question is: does λq change with the level of divertor dissipation? We report results in which feedback controlled nitrogen seeding in the divertor was used to systematically vary divertor dissipation in a series of otherwise identical L-mode plasmas at three plasma currents: 0.55, 0.8 and 1.1 MA. Outer midplane profiles were recorded with a scanning Mirror Langmuir Probe; divertor plasma conditions were monitored with `rail' Langmuir probe and surface thermocouple arrays. Despite an order of magnitude reduction in divertor target heat fluxes (q// 400 MW m-2 to 40 MW m-2) and corresponding change in divertor regime from sheath-limited through high-recycling to near-detached, the upstream electron temperature profile is found to remain unchanged or to become slightly steeper in the near SOL and to drop significantly in the far SOL. Thus heat in the SOL appears to take advantage of this impurity radiation `heat sink' in the divertor by preferentially draining via the narrow (and perhaps an increasingly narrow) λq of the near SOL. Supported by USDoE award DE-FC02-99ER54512.

The Design and Use of Tungsten Coated TZM Molybdenum Tile Inserts in the DIII-D Tokamak Divertor

Energy Technology Data Exchange (ETDEWEB)

Murphy, Christopher [General Atomics, San Diego; Nygren, R. E. [Sandia National Laboratories (SNL); Chrobak, C P. [General Atomics, San Diego; Buchenauer, Dean [Sandia National Laboratories (SNL); Holtrop, Kurt [General Atomics, San Diego; Unterberg, Ezekial A. [ORNL; Zach, Mike P. [ORNL

2017-08-01

Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as theleading candidate. However, tokamak experiments with tungsten divertors have seen significantdetrimental effects on plasma performance. The DIII-D tokamak presently has carbon as theplasma facing surface but to study the effect of tungsten on the plasma and its migration aroundthe vessel, two toroidal rows of carbon tiles in the divertor region were modified with high-Zmetal inserts, composed of a molybdenum alloy (TZM) coated with tungsten. A dedicated twoweek experimental campaign was run with the high-Z metal inserts. One row was coated withtungsten containing naturally occurring levels of isotopes. The second row was coated withtungsten where the isotope 182W was enhanced from the natural level of 26% up to greater than90%. The different isotopic concentrations enabled the experiment to differentiate between thetwo different sources of metal migration from the divertor. Various coating methods wereexplored for the deposition of the tungsten coating, including chemical vapor deposition,electroplating, vacuum plasma spray, and electron beam physical vapor deposition. The coatingswere tested to see if they were robust enough to act as a divertor target for the experiment. Testsincluded cyclic thermal heating using a high power laser and high-fluence deuterium plasmabombardment. The issues associate with the design of the inserts (tile installation, thermal stress,arcing, leading edges, surface preparation, etc.), are reviewed. The results of the tests used toselect the coating method and preliminary experimental observations are presented.

Overview of co-deposition and fuel inventory in castellated divertor structures at JET

International Nuclear Information System (INIS)

Rubel, M.J.; Coad, J.P.; Pitts, R.A.

2007-01-01

The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 x 10 15 cm -2 . Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced

VUV Spectroscopy in DIII-D Divertor

International Nuclear Information System (INIS)

Alkesh Punjabi; Nelson Jalufka

2004-01-01

The research carried out on this grant was motivated by the high power emission from the CIV doublet at 155 nm in the DIII-D divertor and to study the characteristics of the radiative divertor. The radiative divertor is designed to reduce the heat load to the target plates of the divertor by reducing the energy in the divertor plasma using upstream scrape-off-layer (SOL) radiation. In some cases, particularly in Partially Detached Divertor (PDD) operations, this emission accounts for more than 50% of the total radiation from the divertor. In PDD operation, produced by neutral gas injection, the particle flow to the target plate and the divertor temperature are significantly reduced. A father motivation was to study the CIV emission distribution in the lower, open divertor and the upper baffled divertor. Two Vacuum Ultra Violet Tangential viewing Television cameras (VUV TTV) were constructed and installed in the upper, baffled and the lower, open divertor. The images recorded by these cameras were then inverted to produce two-dimensional distributions of CIV in the poloidal plane. Results obtained in the project are summarized in this report

Textor bundle divertor

International Nuclear Information System (INIS)

Yang, T.F.; Wan, A.; Gierszewski, P.; Rapperport, E.; Montgomery, D.B.

1982-01-01

This report presents a preliminary bundle divertor conceptual design for installation on the TEXTOR tokamak. An advanced cascade T-shaped coil configuration is used. This divertor design has the following important characteristics: (1) the current density in the conductor is less than 6 kAmp/cm 2 , and the maximum field is less than 6 Tesla; (2) the divertor can be operated at steady-state either for copper or superconducting conductors; (3) the power consumption is about 7 MW for a normal conductor; (4) the divertor can be inserted into the existing geometry of TEXTOR; (5) the ripple on axis is only 0.3% and the mirror ratio is 2 to 4; (6) the stagnation axis is concave toward the plasma, therefore q/sub D/ is smaller, the acceptance angle is larger, and the efficiency may be better than the conventional circular coil design

TEXTOR bundle divertor

International Nuclear Information System (INIS)

Yang, T.F.; Wan, A.; Gierszewski, P.; Rapperport, E.; Montgomery, D.B.

1982-01-01

This report presents a preliminary bundle divertor conceptual design for installation on the TEXTOR tokamak. An advanced cascade T-shaped coil configuration is used. This divertor design has the following important characteristics: (1) the current density in the conductor is less than 6 kAmp/cm 2 , and the maximum field is less than 6 Tesla; (2) the divertor can be operated at steady-state either for copper or superconducting conductors; (3) the power consumption is about 7 MW for a normal conductor; (4) the divertor can be inserted into the existing geometry of TEXTOR; (5) the ripple on axis is only 0.3% and the mirror ratio is 2 to 4; (6) the stagnation axis is concave toward the plasma, therefore q/sub D/ is smaller, the acceptance angle is larger, and the efficiency may be better than the conventional circular coil design

The ITER divertor concept

International Nuclear Information System (INIS)

Janeschitz, G.; Borrass, K.; Federici, G.; Igitkhanov, Y.; Kukushkin, A.; Pacher, H.D.; Pacher, G.W.; Sugihara, M.

1995-01-01

The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions T e in the divertor can be reduced to <5 eV along a flame like ionisation front by impurity radiation and CX losses. Downstream of the front, neutrals undergo more CX or i-n collisions than ionisation events, resulting in significant momentum loss via neutrals to the divertor chamber wall. The pressure reduction by this mechanism depends on the along-field length for neutral-plasma interaction, the parallel power flux, the neutral density, the ratio of neutral-neutral collision length to the plasma-wall distance and on the Mach number of ions and neutrals. A supersonic transition in the main plasma-neutral interaction region, expected to occur near the ionisation front, would be beneficial for momentum removal. The momentum transfer fraction to the side walls is calculated: low Knudsen number is beneficial. The impact of the different physics effects on the chosen geometry and on the ITER divertor design and the lifetime of the various divertor components are discussed. ((orig.))

Engineering design of the Aries-IV gaseous divertor

International Nuclear Information System (INIS)

Hasan, M.Z.; Najmabadi, F.; Sharafat, S.

1994-01-01

ARIES-IV is a conceptual, D-T burning, steady-state tokamak fusion reactor producing 1000 MWe net. It operates in the second plasma stability regime. The structural material is SiC composite and the primary coolant is helium at 10MPa base pressure. ARIES-IV uses double-null divertors for particle control. Total thermal power recovered from the divertors is 425MW, which is 16% of the total reactor thermal power. Among the desirable goals of divertor design were to avoid the use of tungsten and to use the same structural material and primary coolant as in the blanket design. In order to reduce peak heat flux, the innovative gaseous divertor has been used in ARIES-IV. A gaseous divertor reduces peak heat flux by increasing the surface area and by distributing particle and radiation energy more uniformly. Another benefit of gaseous divertor is the reduction of plasma temperature in the divertor chamber, so that material erosion due to sputtering, can be diminished. This makes the use of low-Z material possible in a gaseous divertor

Time and space-resolved energy flux measurements in the divertor of the ASDEX tokamak by computerized infrared thermography

International Nuclear Information System (INIS)

Mueller, E.R.; Steinmetz, K.; Bein, B.K.

1984-06-01

A new, fully computerized and automatic thermographic system has been developed. Its two central components are an AGA THV 780 infrared camera and a PDP-11/34 computer. A combined analytical-numerical method of solving the 1-dimensional heat diffusion equation for a solid of finite thickness bounded by two parallel planes was developed. In high-density (anti nsub(e) = 8 x 10 13 cm -3 ) neutral-beam-heated (L-mode) divertor discharges in ASDEX, the power deposition on the neutralizer plates is reduced to about 10-15% of the total heating power, owing to the inelastic scattering of the divertor plasma from a neutral gas target. Between 30% and 40% of the power is missing in the global balance. The power flow inside the divertor chambers is restricted to an approximately 1-cm-thick plasma scrape-off layer. This width depends only weakly on the density and heating power. During H-phases free of Edge Localized Mode (ELM) activity the energy flow into the divertor is blocked. During H-phases with ELM activity the energy is expelled into the divertor in very short intense pulses (several MW for about one hundred μs). Sawtooth events are able to transport significant amounts of energy from the plasma core to the peripheral zones and the scrape-off layer, and they are frequently correlated with transitions from the L to the H mode. (orig./AH)

Divertor scenario development for NSTX Upgrade

Science.gov (United States)

Soukhanovskii, V. A.; McLean, A. G.; Meier, E. T.; Rognlien, T. D.; Ryutov, D. D.; Bell, R. E.; Diallo, A.; Gerhardt, S. P.; Kaita, R.; Kolemen, E.; Leblanc, B. P.; Menard, J. E.; Podesta, M.; Scotti, F.

2012-10-01

In the NSTX-U tokamak, initial plans for divertor plasma-facing components (PFCs) include lithium and boron coated graphite, with a staged transition to molybdenum. Steady-state peak divertor heat fluxes are projected to reach 20-30 MW/m^2 in 2 MA, 12 MW NBI-heated discharges of up to 5 s duration, thus challenging PFC thermal limits. Based on the recent NSTX divertor experiments and modeling with edge transport code UEDGE, a favorable basis for divertor power handling in NSTX-U is developed. The snowflake divertor geometry and feedback-controlled divertor impurity seeding applied to the lower and upper divertors are presently envisioned. In the NSTX snowflake experiments with lithium-coated graphite PFCs, the peak divertor heat fluxes from Type I ELMs and between ELMs were significantly reduced due to geometry effects, increased volumetric losses and null-point convective redistribution between strike points. H-mode core confinement was maintained at H98(y,2)<=1 albeit the radiative detachment. Additional CD4 seeding demonstrated potential for a further increase of divertor radiation.

Characteristics of the Secondary Divertor on DIII-D

Science.gov (United States)

Watkins, J. G.; Lasnier, C. J.; Leonard, A. W.; Evans, T. E.; Pitts, R.; Stangeby, P. C.; Boedo, J. A.; Moyer, R. A.; Rudakov, D. L.

2009-11-01

In order to address a concern that the ITER secondary divertor strike plates may be insufficiently robust to handle the incident pulses of particles and energy from ELMs, we performed dedicated studies of the secondary divertor plasma and scrape-off layer (SOL). Detailed measurements of the ELM energy and particle deposition footprint on the secondary divertor target plates were made with a fast IR camera and Langmuir probes and SOL profile and transport measurements were made with reciprocating probes. The secondary divertor and SOL conditions depended on changes in the magnetic balance and the core plasma density. Larger density resulted in smaller ELMs and the magnetic balance affected how many ELM particles coupled to the secondary SOL and divertor. Particularly striking are the images from a new fast IR camera that resolve ELM heat pulses and show spiral patterns with multiple peaks during ELMs in the secondary divertor.

ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

Science.gov (United States)

Whyte, Dennis; ADX Team

2015-11-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

Experimental studies of the snowflake divertor in TCV

Directory of Open Access Journals (Sweden)

B. Labit

2017-08-01

Full Text Available To address the risk that, in a fusion reactor, the conventional single-null divertor (SND configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD, are investigated in TCV. The expected benefits of the SFD-minus in terms of power load and peak heat flux are discussed and compared to experimental measurements. In addition, key results obtained during the last years are summarized.

Experimental studies of the snowflake divertor in TCV

NARCIS (Netherlands)

Labit, B.; Canal, G. P.; Christen, N.; Duval, B. P.; Lipschultz, B.; Lunt, T.; Nespoli, F.; Reimerdes, H.; Sheikh, U.; Theiler, C.; Tsui, C. K.; Verhaegh, K.; Vijvers, W. A. J.

2017-01-01

To address the risk that, in a fusion reactor, the conventional single-null divertor (SND) configuration may not be able to handle the power exhaust, alternative divertor configurations, such as the Snowflake divertor (SFD), are investigated in TCV. The expected benefits of the SFD-minus in terms of

Divertor design for the Tokamak Physics Experiment

International Nuclear Information System (INIS)

Hill, D.N.; Braams, B.

1994-05-01

In this paper we discuss the present divertor design for the planned TPX tokamak, which will explore the physics and technology of steady-state (1000s pulses) heat and particle removal in high confinement (2--4x L-mode), high beta (β N ≥ 3) divertor plasmas sustained by non-inductive current drive. The TPX device will operate in the double-null divertor configuration, with actively cooled graphite targets forming a deep (0.5 m) slot at the outer strike point. The peak heat flux on, the highly tilted (74 degrees from normal) re-entrant (to recycle ions back toward the separatrix) will be in the range of 4--6 MW/m 2 with 18 MW of neutral beams and RF heating power. The combination of active pumping and gas puffing (deuterium plus impurities), along with higher heating power (45 MW maximum) will allow testing of radiative divertor concepts at ITER-like power densities

Detached divertor plasmas in JET

Energy Technology Data Exchange (ETDEWEB)

Horton, L D; Borrass, K; Corrigan, G; Gottardi, N; Lingertat, J; Loarte, A; Simonini, R; Stamp, M F; Taroni, A [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking; Stangeby, P C [Toronto Univ., ON (Canada). Inst. for Aerospace Studies

1994-07-01

In simulations with high radiated power fractions, it is possible to produce the drop in ion current to the divertor targets typical of detached plasmas. Despite the fact that these experiments are performed on beryllium target tiles, radiation from deuterium and beryllium cannot account for the measured power losses. The neutral deuterium levels in the SOL in these plasmas are higher than the model predicts. This may be due to leakage from the divertor or to additional wall sources related to the non-steady nature of these plasmas. In contrast, a surprisingly high level of carbon is present in these discharges; higher even than would be predicted are the divertor target tiles pure carbon. This level may well be large enough to produce the measured radiation. (authors). 6 refs., 2 figs., 1 tab.

Retention of Hydrogen Isotopes in Divertor Tiles Used in JT-60U

International Nuclear Information System (INIS)

Hirohata, Y.; Shibahara, T.; Tanabe, T.; Oya, Y.; Arai, T.; Gotoh, Y.; Masaki, K.; Yagyu, J.; Oyaidzu, M.; Okuno, K.; Nishikawa, M.; Miya, N.

2005-01-01

Retention characteristics of deuterium and hydrogen retained in graphite tiles placed in the divertor region of JT-60U were investigated by thermal desorption spectroscopy (TDS). The deuterium retained in the near surface of all graphite tiles was mostly replaced by hydrogen due to exposure to hydrogen plasma at the final stage operations, resulting in main deuterium retention in the deeper region. The dominant species desorbed from the divertor tiles were H 2 , HD, D 2 and CH 4 . The smallest retention of hydrogen isotopes (H+D) was observed in the outer divertor tile which was eroded with maximum of 20 μm depth. The amount of H+D retained in the inner divertor tiles covered by the re-deposited layers increased with the thickness of the re-deposited layers. Hydrogen isotopes concentration ((H+D)/C) in the re-deposited layers was ∼0.02, which was much smaller than those observed in JET and other devices

Mitigation of divertor heat loads by strike point sweeping in high power JET discharges

Science.gov (United States)

Silburn, S. A.; Matthews, G. F.; Challis, C. D.; Frigione, D.; Graves, J. P.; Mantsinen, M. J.; Belonohy, E.; Hobirk, J.; Iglesias, D.; Keeling, D. L.; King, D.; Kirov, K.; Lennholm, M.; Lomas, P. J.; Moradi, S.; Sips, A. C. C.; Tsalas, M.; Contributors, JET

2017-12-01

Deliberate periodic movement (sweeping) of the high heat flux divertor strike lines in tokamak plasmas can be used to manage the heat fluxes experienced by exhaust handling plasma facing components, by spreading the heat loads over a larger surface area. Sweeping has recently been adopted as a routine part of the main high performance plasma configurations used on JET, and has enabled pulses with 30 MW plasma heating power and 10 MW radiation to run for 5 s without overheating the divertor tiles. We present analysis of the effectiveness of sweeping for divertor temperature control on JET, using infrared camera data and comparison with a simple 2D heat diffusion model. Around 50% reduction in tile temperature rise is obtained with 5.4 cm sweeping compared to the un-swept case, and the temperature reduction is found to scale slower than linearly with sweeping amplitude in both experiments and modelling. Compatibility of sweeping with high fusion performance is demonstrated, and effects of sweeping on the edge-localised mode behaviour of the plasma are reported and discussed. The prospects of using sweeping in future JET experiments with up to 40 MW heating power are investigated using a model validated against existing experimental data.

Mitigation of divertor heat loads by strike point sweeping in high power JET discharges

International Nuclear Information System (INIS)

Silburn, S A; Matthews, G F; Challis, C D; Belonohy, E; Iglesias, D; Keeling, D L; King, D; Kirov, K; Lomas, P J; Frigione, D; Graves, J P; Mantsinen, M J; Hobirk, J; Lennholm, M; Moradi, S; Sips, A C C; Tsalas, M

2017-01-01

Deliberate periodic movement (sweeping) of the high heat flux divertor strike lines in tokamak plasmas can be used to manage the heat fluxes experienced by exhaust handling plasma facing components, by spreading the heat loads over a larger surface area. Sweeping has recently been adopted as a routine part of the main high performance plasma configurations used on JET, and has enabled pulses with 30 MW plasma heating power and 10 MW radiation to run for 5 s without overheating the divertor tiles. We present analysis of the effectiveness of sweeping for divertor temperature control on JET, using infrared camera data and comparison with a simple 2D heat diffusion model. Around 50% reduction in tile temperature rise is obtained with 5.4 cm sweeping compared to the un-swept case, and the temperature reduction is found to scale slower than linearly with sweeping amplitude in both experiments and modelling. Compatibility of sweeping with high fusion performance is demonstrated, and effects of sweeping on the edge-localised mode behaviour of the plasma are reported and discussed. The prospects of using sweeping in future JET experiments with up to 40 MW heating power are investigated using a model validated against existing experimental data. (paper)

Advantages and Challenges of Radiative Liquid Lithium Divertor

Science.gov (United States)

Ono, Masayuki

2017-10-01

Steady-state fusion power plant designs present major divertor technology challenges, including high divertor heat flux both in steady-state and during transients. In addition to these concerns, there are the unresolved technology issues of long term dust accumulation and associated tritium inventory and safety issues. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid Li divertor (RLLD) concept and its variant, the active liquid Li divertor concept (ARLLD), taking advantage of the enhanced Li radiation in relatively poorly confined divertor plasmas. It has been suggested that radiation-based liquid lithium (LL) divertor concepts with a modest Li-loop could provide a possible solution for the outstanding fusion reactor technology issues such as divertor heat flux mitigation and real time dust removal, while potentially improving the reactor plasma performance. Laboratory tests are also planned to investigate the Li-T recover efficiency and other relevant research topics of the RLLD. This work supported by DoE Contract No. DE-AC02-09CH11466.

TCV divertor upgrade for alternative magnetic configurations

Directory of Open Access Journals (Sweden)

H. Reimerdes

2017-08-01

Full Text Available The Swiss Plasma Center (SPC is planning a divertor upgrade for the TCV tokamak. The upgrade aims at extending the research of conventional and alternative divertor configurations to operational scenarios and divertor regimes of greater relevance for a fusion reactor. The main elements of the upgrade are the installation of an in-vessel structure to form a divertor chamber of variable closure and enhanced diagnostic capabilities, an increase of the pumping capability of the divertor chamber and the addition of new divertor poloidal field coils. The project follows a staged approach and is carried out in parallel with an upgrade of the TCV heating system. First calculations using the EMC3-Eirene code indicate that realistic baffles together with the planned heating upgrade will allow for a significantly higher compression of neutral particles in the divertor, which is a prerequisite to test the power dissipation potential of various divertor configurations.

Operating conditions of the BPX divertor

International Nuclear Information System (INIS)

Hill, D.N.; Milovich, J.; Rognlien, T.; Braams, B.J.; Brooks, J.N.; Campbell, R.; Haines, J.; Knoll, D.; Prinja, A.; Stotler, D.P.; Ulrickson, M.

1991-01-01

In this paper we discuss the expected operating conditions at the divertor of the BPX tokamak (Burning Plasma Experiment), the next- step US tokamak proposed for the study of self-heated plasmas at Q ≅ 5 to ignition. In this double-null device (κ ≅ 2), the predicted first-wall loading is high because of is compact size (R = 2.6m, α = 0.8m, I p = 10.6 MA, and B T ) and its high projected fusion power output (100--500 MW with up to 20 MW of ICRH). Present designs call for inertially cooled carbon-based target plate material and X-point sweeping to handle the divertor heat flux during the 3--5 s flat-top at full power. The X-point is maintained about 15--20 cm off the target plates (a distance of ∼5m along field lines), which represents a reasonable compromise between lowering the divertor electron temperature (T e,d ) by increasing the connection length, and lowering the peak divertor heat flux (q d ) by increasing the magnetic flux expansion (which is about 15--20 in this case). It is planned for the BPX device to operate with H-mode confinement; ELMs are expected because of the relatively high power flow through the edge plasma (P sep ≅ 0.6 MW/m 2 for P fus = 500 MW). The ELMs will help reduce the impurity concentration in the core plasma (Z eff ≅ 1.7) and keep the density down, but should not add significantly to the divertor heat flux since their measured contribution to the global power balance drops with increasing input power

Fast wave power flow along SOL field lines in NSTX

Science.gov (United States)

Perkins, R. J.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Hosea, J. C.; Jaworski, M. A.; Leblanc, B. P.; Kramer, G. J.; Phillips, C. K.; Roquemore, L.; Taylor, G.; Wilson, J. R.; Ahn, J.-W.; Gray, T. K.; Green, D. L.; McLean, A.; Maingi, R.; Ryan, P. M.; Jaeger, E. F.; Sabbagh, S.

2012-10-01

On NSTX, a major loss of high-harmonic fast wave (HHFW) power can occur along open field lines passing in front of the antenna over the width of the scrape-off layer (SOL). Up to 60% of the RF power can be lost and at least partially deposited in bright spirals on the divertor floor and ceiling [1,2]. The flow of HHFW power from the antenna region to the divertor is mostly aligned along the SOL magnetic field [3], which explains the pattern of heat deposition as measured with infrared (IR) cameras. By tracing field lines from the divertor back to the midplane, the IR data can be used to estimate the profile of HHFW power coupled to SOL field lines. We hypothesize that surface waves are being excited in the SOL, and these results should benchmark advanced simulations of the RF power deposition in the SOL (e.g., [4]). Minimizing this loss is critical optimal high-power long-pulse ICRF heating on ITER while guarding against excessive divertor erosion.[4pt] [1] J.C. Hosea et al., AIP Conf Proceedings 1187 (2009) 105. [0pt] [2] G. Taylor et al., Phys. Plasmas 17 (2010) 056114. [0pt] [3] R.J. Perkins et al., to appear in Phys. Rev. Lett. [0pt] [4] D.L. Green et al., Phys. Rev. Lett. 107 (2011) 145001.

Characterizations of power loads on divertor targets for type-I, compound and small ELMs in the EAST superconducting tokamak

DEFF Research Database (Denmark)

Wang, L.; Xu, G.S.; Guo, H.Y.

2013-01-01

-III ELMy H-modes. The energy loss and divertor power load are systematically characterized for these different ELMy H-modes to provide a physics basis for the next-step high-power long-pulse operations in EAST. Both type-I and compound ELMs exhibit good confinement (H98(y,2) ∼ 1). A significant loss......The Experimental Advanced Superconducting Tokamak (EAST) has recently achieved a variety of H-mode regimes with different edge-localized mode (ELM) dynamics, including type-I ELMs, compound ELMs, which are manifested by the onset of a large spike followed by a sequence of small spikes on Dα......-III ELMs. It is remarkable that the new very small ELMy H-modes exhibit even lower target power deposition than type-III ELMs, with the peak heat flux generally below 1 MW m−2. These very small ELMs are usually accompanied by broadband fluctuations with frequencies ranging from 20 to 50 kHz, which may...

Hydrogen isotope distributions and retentions in the inner divertor tile of JT-60U

Energy Technology Data Exchange (ETDEWEB)

Oya, Y. [Radioisotope Center, University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-0032 (Japan)]. E-mail: yoya@ric.u-tokyo.ac.jp; Hirohata, Y. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Tanabe, T. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Shibahara, T. [Graduate School of Engineering, Nagoya University, Nagoya 464-8603 (Japan); Kimura, H. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Oyaidzu, M. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Arai, T. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Masaki, K. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Gotoh, Y. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Okuno, K. [Faculty of Science, Shizuoka University, Shizuoka 422-8529 (Japan); Miya, N. [Naka Fusion Establishment, Japan Atomic Energy Research Institute, Naka-machi, Naka-gun, Ibaraki 311-0193 (Japan); Hino, T. [Graduate School of Engineering, Hokkaido University, Sapporo 060-8628 (Japan); Tanaka, S. [Graduate School of Engineering, University of Tokyo, Tokyo 113-8656 (Japan)

2005-11-15

Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature.

Hydrogen isotope distributions and retentions in the inner divertor tile of JT-60U

International Nuclear Information System (INIS)

Oya, Y.; Hirohata, Y.; Tanabe, T.; Shibahara, T.; Kimura, H.; Oyaidzu, M.; Arai, T.; Masaki, K.; Gotoh, Y.; Okuno, K.; Miya, N.; Hino, T.; Tanaka, S.

2005-01-01

Retention profiles of hydrogen and deuterium in graphite tiles placed in the inner divertor region of JT-60U were analyzed by secondary ion mass spectroscopy (SIMS) and thermal desorption spectroscopy (TDS). The difference in hydrogen and deuterium retention behaviour is discussed considering the frequency of the strike-point hit and history of NBI heating power. It was found that most of hydrogen/deuterium was retained in the deposited layers, HH deposition layers/DD deposition layers or co-deposited with carbon. Owing to the higher heating power of DD discharges, the deuterium concentration in the DD deposition layers was much lower than that of hydrogen in the HH deposition layers. On the area showing no deposition, very shallow profile of deuterium dominated hydrogen profile. These results indicate that the tritium retention is strongly influenced by the history of discharge and temperatures. Tritium retention on graphite tiles and deposition layers could be significantly reduced with increasing the operation temperature

Low energy neutral particle fluxes in the JET divertor

International Nuclear Information System (INIS)

Reichle, R.; Horton, L.D.; Ingesson, L.C.; Jaeckel, H.J.; McCormick, G.K.; Loarte, A.; Simonini, R.; Stamp, M.F.

1997-01-01

First measurements are presented of the total power loss through neutral particles and their average energy in the JET divertor. The method used distinguishes between the heat flux and the electromagnetic radiation on bolometers. This is done by comparing measurements from inside the divertor either with opposite lines of sight or with a tomographic reconstruction of the radiation. The typical value of the total power loss in the divertor through neutrals is about 1 MW. The average energy of the neutral particles at the inner divertor leg is 1.5-3 eV when detachment is in progress, which agrees with EDGE2D/NIMBUS modelling. (orig.)

FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT

International Nuclear Information System (INIS)

O'NEIL, RC; STAMBAUGH, RD

2002-01-01

OAK A271 FINAL REPORT FOR THE DIII-D RADIATIVE DIVERTOR PROJECT. The Radiative Divertor Project originated in 1993 when the DIII-D Five Year Plan for the period 1994--1998 was prepared. The Project Information Sheet described the objective of the project as ''to demonstrate dispersal of divertor power by a factor of then with sufficient diagnostics and modeling to extend the results to ITER and TPX''. Key divertor components identified were: (1) Carbon-carbon and graphite armor tiles; (2) The divertor structure providing a gas baffle and cooling; and (3) The divertor cryopumps to pump fuel and impurities

Integrated core-edge-divertor modeling studies

International Nuclear Information System (INIS)

Stacey, W.M.

2001-01-01

An integrated calculation model for simulating the interaction of physics phenomena taking place in the plasma core, in the plasma edge and in the SOL and divertor of tokamaks has been developed and applied to study such interactions. The model synthesises a combination of numerical calculations (1) the power and particle balances for the core plasma, using empirical confinement scaling laws and taking into account radiation losses (2), the particle, momentum and power balances in the SOL and divertor, taking into account the effects of radiation and recycling neutrals, (3) the transport of feeling and recycling neutrals, explicitly representing divertor and pumping geometry, and (4) edge pedestal gradient scale lengths and widths, evaluation of theoretical predictions (5) confinement degradation due to thermal instabilities in the edge pedestals, (6) detachment and divertor MARFE onset, (7) core MARFE onsets leading to a H-L transition, and (8) radiative collapse leading to a disruption and evaluation of empirical fits (9) power thresholds for the L-H and H-L transitions and (10) the width of the edge pedestals. The various components of the calculation model are coupled and must be iterated to a self-consistent convergence. The model was developed over several years for the purpose of interpreting various edge phenomena observed in DIII-D experiments and thereby, to some extent, has been benchmarked against experiment. Because the model treats the interactions of various phenomena in the core, edge and divertor, yet is computationally efficient, it lends itself to the investigation of the effects of different choices of various edge plasma operating conditions on overall divertor and core plasma performance. Studies of the effect of feeling location and rate, divertor geometry, plasma shape, pumping and over 'edge parameters' on core plasma properties (line average density, confinement, density limit, etc.) have been performed for DIII-D model problems. A

Role of molecular effects in divertor plasma recombination

Directory of Open Access Journals (Sweden)

A.S. Kukushkin

2017-08-01

Full Text Available Molecule-Activated Recombination (MAR effect is re-considered in view of divertor plasma conditions. A strong isotopic effect is demonstrated. In deuterium plasmas, the reaction chain through D2+ formation, usually considered dominant and included in 2D edge plasma models, is negligible. However, in this case the other branch, through D−, usually neglected in modelling, becomes relatively strong. The overall share of MAR in divertor plasma recycling stays within 20%. The operational parameters of the divertor plasmas, such as the peak power loading on the divertor targets or the pressure limit for partial detachment of the divertor plasma, are insensitive to the presence of MAR, although the latter may be important for correct interpretation of the divertor diagnostics.

Innovative Divertor Development to Solve the Plasma Heat-Flux Problem

International Nuclear Information System (INIS)

Rognlien, T.; Ryutov, D.; Makowski, M.; Soukhanovskii, V.; Umansky, M.; Cohen, R.; Hill, D.; Joseph, I.

2009-01-01

) and generating Resonant Magnetic Perturbations by the SOL currents (3). However, the specific approaches discussed here are part of a wider class of innovative divertor ideas that have come from the community in the last several years, and we certainly advocate the need to consider a range of options. Indeed, the most effective solution to the heat-flux problem may well contain features of various ideas. For example, there are the X-divertor (Kotschenreuther et al. (4)) that expands the magnetic flux surface in the vicinity of the near-X-point divertor plate, and the super X-divertor (Valanju et al. (5)) that guides the near-separatrix SOL flux tubes to a larger major radius to increase the surface area available for power deposition. These approaches have the common feature of manipulation of the edge magnetic geometry. Another approach is the use of liquid divertor surfaces that can increase the heat-flux capability by flowing the heated material to a cooling region and eventually out of the machine, and/or by being able to withstand a higher peak heat flux (6). All of these areas are only emerging concepts that require substantially more analysis and definitive experimental tests, and given the need for a large improvement in this area, we advocate a substantial program to systematically assess the approaches. Because of space limitation here, we present some details of one of the concepts, namely the Snowflake divertor configuration. The Snowflake (SF) divertor (2) exploits a tokamak geometry in which the poloidal magnetic field varies quadratically with distance from the X-point null, Δr. The name stems from the characteristic hexagonal, snowflake-like, shape of the multi-branched separatrix for this exact second-order null. In contrast, the standard X-point configuration has a poloidal field varying linearly with ?r. The different variations mean that a flux expansion is much larger in the vicinity of a null of a snowflake divertor, and one can try to exploit

An Asdex-type divertor for ITER

International Nuclear Information System (INIS)

Fowler, T.K.

1989-01-01

An Asdex-type local divertor is proposed for ITER consisting of a copper poloidal field coil adjacent to the plasma. Estimates indicate that the power consumption is acceptable. Advantages would be a much reduced heat load not very sensitive to magnetic perturbations. A disadvantage is the finite lifetime under neutron bombardment that would require periodic replacement of the divertor coils in a reactor, but probably not in ITER because of its limited fluence. Another disadvantage would be poorer blanket coverage unless the divertor coil itself incorporates breeding material. 3 figs

The effect of density on divertor conditions in ASDEX-Upgrade

International Nuclear Information System (INIS)

Pitcher, C.S.; Bosch, H.-S.; Buechl, K.; Field, A.; Fuchs, C.; Haas, G.; Junker, W.; Neu, R.; Neuhauser, J.; Wenzel, U.

1995-01-01

Detailed experimental divertor data are presented on the profiles of density and temperature in the inner and outer divertor fans, the radiated power distribution, the gas pressure and the spectroscopically derived particle fluxes, all as a function of the discharge density. At low and medium density, the inner divertor is cold and dense compared to the outer divertor. At high density, strong X-point MARFE and separatrix radiation partially detaches the inner divertor. Probe measurements which penetrate into the X-point MARFE at the outer divertor are presented. ((orig.))

Engineering design of a radiative divertor for DIII-D

International Nuclear Information System (INIS)

Smith, J.P.; Baxi, C.B.; Bozek, A.S.

1995-10-01

A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The Radiative Divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the Radiative Divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The Radiative Divertor design is very flexible, and will allow physics studies of the effects of slot width and length. Radiative Divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The Radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D Advanced Divertor program form a foundation for the design work on the Radiative Divertor

Assessment of the integration of a He-cooled divertor system in the power conversion system for the dual-coolant blanket concept (TW2-TRP-PPCS12D8)

International Nuclear Information System (INIS)

Norajitra, P.; Kruessmann, R.; Malang, S.; Reimann, G.

2002-12-01

Application of a helium-cooled divertor together with the dual-coolant blanket concept is considered favourable for achieving a high thermal efficiency of the power plant due to its relatively high coolant outlet temperature. A new FZK He-cooled modular divertor concept with integrated pin arrays (HEMP) is introduced. Its main features and function are described in detail. The result of the thermalhydraulic analysis shows that the HEMP divertor concept has the potential of resisting, a heat flow density of at least 10-15 MW/m 2 at a reachable heat transfer coefficient of approx. 60 kW/m 2 K and a reasonable pumping power. Integration of this divertor concept into the power conversion system using a closed Brayton gas turbine system with three-stage compression leads to a net efficiency of the blanket/divertor cycle of about 43%. (orig.)

Scrape-off layer radiation and heat load to the ASDEX Upgrade LYRA divertor

International Nuclear Information System (INIS)

Kallenbach, A.; Kaufmann, M.; Coster, D.P.

1999-01-01

In 1997 the new 'LYRA' divertor went into operation at ASDEX Upgrade and, in parallel, the neutral beam heating power was increased to 20 MW by installation of a second injector leading to a P/R value of 12 MW/m. Experiments have shown that the ASDEX Upgrade LYRA divertor is capable of handling such high heating powers. There is an overall reduction of the maximum heat flux in the LYRA divertor by about a factor of 2 compared with the previous open divertor Div I. This reduction is mainly due to increased radiative losses inside the divertor region, which are caused by an effective reflection of hydrogen neutrals into the hot separatrix region. The main channel of radiative loss is carbon radiation, which cools the divertor plasma down to a few electronvolts, where hydrogen radiation losses become significant. The radiative losses preferentially reduce the power flux at the separatrix, leading to early detachment around the strike point position. With increasing density, the detached region extends upwards on the vertical target. The power fraction radiated in the LYRA divertor is around 45% and nearly independent of the heating power. This value is a factor of 2 higher than the typical radiation fraction in Div I. B2-EIRENE modelling of the performed experiments supports the experimental finding and refines the understanding of loss processes in the divertor region. (author)

Development of database for the divertor recycling in JT-60U and its analysis

Energy Technology Data Exchange (ETDEWEB)

Takizuka, Tomonori; Shimizu, Katsuhiro; Hayashi, Nobuhiko; Asakura, Nobuyuki [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Arakawa, Kazuya [Komatsu, Ltd., Tokyo (Japan)

2003-05-01

We have developed a database for the divertor recycling in JT-60U plasmas. This database makes it possible to investigate behaviors of the neutral-particle flux in plasmas and the ion flux to divertor plates under a condition for core-plasma parameters, such as electron density and heating power. The correlation between the electron density and the heating power is not strong in this database, and parameter scans for the density and the power in wide ranges are realized. On the basis of this database, we have analyzed the ion flux to divertor plates. The divertor-plate ion flux amplified by the recycling grows nonlinearly with the increase of the electron density n{sub e}. Its averaged dependence is a linear growth ({approx}n{sub e}{sup 1.0}) at the low density, and becomes a nonlinear growth ({approx}n{sub e}{sup 1.5}) at the high density. The spread of dependence from the averaged one is very large. This spread is caused mainly by complex physical characteristics of divertor plasmas, though it is little dependent on the heating power. The behavior of ion flux depends strongly on divertor configurations and divertor-plate/first-wall conditions. It is confirmed that the bifurcated transition takes place from the low-recycling divertor plasma at the low density to the high-recycling divertor plasma at the high density. The density at the transition is nearly proportional to the 1/4 power of the heating power. (author)

Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

International Nuclear Information System (INIS)

Kaufmann, M.; Bosch, H.S.; Herrmann, A.

1999-01-01

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. Measurements presented in this paper reveal a reduction of the maximum heat flux 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 significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (author)

Energy and particle control characteristics of the ASDEX Upgrade 'LYRA' divertor

International Nuclear Information System (INIS)

Kaufmann, M.; Bosch, H.-S.; Herrmann, A.

2001-01-01

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. Measurements presented in this paper reveal a reduction of the maximum heat flux 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 significant. They are increased due to an effective reflection of neutrals into the hot separatrix region. B2-Eirene modelling of the performed experiments supports the experimental findings and refines the understanding of loss processes in the divertor region. (and others)

Optimal thermal-hydraulic performance for helium-cooled divertors

International Nuclear Information System (INIS)

Izenson, M.G.; Martin, J.L.

1996-01-01

Normal flow heat exchanger (NFHX) technology offers the potential for cooling divertor panels with reduced pressure drops (<0.5% Δp/p), reduced pumping power (<0.75% pumping/thermal power), and smaller duct sizes than conventional helium heat exchangers. Furthermore, the NFHX can easily be fabricated in the large sizes required for divertors in large tokamaks. Recent experimental and computational results from a program to develop NFHX technology for divertor coolings using porous metal heat transfer media are described. We have tested the thermal and flow characteristics of porous metals and identified the optimal heat transfer material for the divertor heat exchanger. Methods have been developed to create highly conductive thermal bonds between the porous material and a solid substrate. Computational fluid dynamics calculations of flow and heat transfer in the porous metal layer have shown the capability of high thermal effectiveness. An 18-kW NFHX, designed to meet specifications for the international Thermonuclear Experimental Reactor divertor, has been fabricated and tested for thermal and flow performance. Preliminary results confirm design and fabrication methods. 11 refs., 12 figs., 1 tab

Studies of impurity deposition/implantation in JET divertor tiles using SIMS and ion beam techniques

International Nuclear Information System (INIS)

Likonen, J.; Lehto, S.; Coad, J.P.; Renvall, T.; Sajavaara, T.; Ahlgren, T.; Hole, D.E.; Matthews, G.F.; Keinonen, J.

2003-01-01

At the end of C4 campaign at JET, a 1% SiH 4 /99% D 2 mixture and pure 13 CH 4 were injected into the torus from the outer divertor wall and from the top of the vessel, respectively, in order to study material transport and scrape-off layer (SOL) flows. A set of MkIIGB tiles was removed during the 2001 shutdown for surface analysis. The tiles were analysed with secondary ion mass spectrometry (SIMS) and time-of-flight elastic recoil detection analysis (TOF-ERDA). 13 C was detected in the inner divertor wall tiles implying material transport from the top of the vessel. Silicon was detected mainly at the outer divertor wall tiles and very small amounts were found in the inner divertor wall tiles. Si amounts in the inner divertor wall tiles were so low that rigorous conclusions about material transport from divertor outboard to inboard cannot be made

Effect of low density H-mode operation on edge and divertor plasma parameters

International Nuclear Information System (INIS)

Maingi, R.; Mioduszewski, P.K.; Cuthbertson, J.W.

1994-07-01

We present a study of the impact of H-mode operation at low density on divertor plasma parameters on the DIII-D tokamak. The line-average density in H-mode was scanned by variation of the particle exhaust rate, using the recently installed divertor cryo-condensation pump. The maximum decrease (50%) in line-average electron density was accompanied by a factor of 2 increase in the edge electron temperature, and 10% and 20% reductions in the measured core and divertor radiated power, respectively. The measured total power to the inboard divertor target increased by a factor of 3, with the major contribution coming from a factor of 5 increase in the peak heat flux very close to the inner strike point. The measured increase in power at the inboard divertor target was approximately equal to the measured decrease in core and divertor radiation

Divertor heat flux mitigation in the National Spherical Torus Experimenta)

Science.gov (United States)

Soukhanovskii, V. A.; Maingi, R.; Gates, D. A.; Menard, J. E.; Paul, S. F.; Raman, R.; Roquemore, A. L.; Bell, M. G.; Bell, R. E.; Boedo, J. A.; Bush, C. E.; Kaita, R.; Kugel, H. W.; Leblanc, B. P.; Mueller, D.; NSTX Team

2009-02-01

Steady-state handling of divertor heat flux is a critical issue for both ITER and spherical torus-based devices with compact high power density divertors. Significant reduction of heat flux to the divertor plate has been achieved simultaneously with favorable core and pedestal confinement and stability properties in a highly shaped lower single null configuration in the National Spherical Torus Experiment (NSTX) [M. Ono et al., Nucl. Fusion 40, 557 2000] using high magnetic flux expansion at the divertor strike point and the radiative divertor technique. A partial detachment of the outer strike point was achieved with divertor deuterium injection leading to peak flux reduction from 4-6MWm-2to0.5-2MWm-2 in small-ELM 0.8-1.0MA, 4-6MW neutral beam injection-heated H-mode discharges. A self-consistent picture of the outer strike point partial detachment was evident from divertor heat flux profiles and recombination, particle flux and neutral pressure measurements. Analytic scrape-off layer parallel transport models were used for interpretation of NSTX detachment experiments. The modeling showed that the observed peak heat flux reduction and detachment are possible with high radiated power and momentum loss fractions, achievable with divertor gas injection, and nearly impossible to achieve with main electron density, divertor neutral density or recombination increases alone.

Status of National Spherical Torus Experiment Liquid Lithium Divertor

Science.gov (United States)

Kugel, H. W.; Viola, M.; Ellis, R.; Bell, M.; Gerhardt, S.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.; Roquemore, A. L.; Schneider, H.; Timberlake, J.; Zakharov, L.; Nygren, R. E.; Allain, J. P.; Maingi, R.; Soukhanovskii, V.

2009-11-01

Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is the 2009 installation of a Liquid Lithium Divertor (LLD). The 20 cm wide LLD located on the lower outer divertor, consists of four, 80 degree sections; each section is separated by a row of graphite diagnostic tiles. The temperature controlled LLD structure consists of a 0.01cm layer of vacuum flame-sprayed, 50 percent porous molybdenum, on top of 0.02 cm, 316-SS brazed to a 1.9 cm Cu base. The physics design of the LLD encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Critical need for MFE: the Alcator DX advanced divertor test facility

Science.gov (United States)

Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Wolf, S.; Bonoli, P.; Fiore, C.; Granetz, R.; Greenwald, M.; Hutchinson, I.; Hubbard, A.; Hughes, J.; Lin, Y.; Lipschultz, B.; Parker, R.; Porkolab, M.; Reinke, M.; Rice, J.; Shiraiwa, S.; Terry, J.; Theiler, C.; Wallace, G.; White, A.; Whyte, D.; Wukitch, S.

2013-10-01

Three critical challenges must be met before a steady-state, power-producing fusion reactor can be realized: how to (1) safely handle extreme plasma exhaust power, (2) completely suppress material erosion at divertor targets and (3) do this while maintaining a burning plasma core. Advanced divertors such as ``Super X'' and ``X-point target'' may allow a fully detached, low temperature plasma to be produced in the divertor while maintaining a hot boundary layer around a clean plasma core - a potential game-changer for magnetic fusion. No facility currently exists to test these ideas at the required parallel heat flux densities. Alcator DX will be a national facility, employing the high magnetic field technology of Alcator combined with high-power ICRH and LHCD to test advanced divertor concepts at FNSF/DEMO power exhaust densities and plasma pressures. Its extended vacuum vessel contains divertor cassettes with poloidal field coils for conventional, snowflake, super-X and X-point target geometries. Divertor and core plasma performance will be explored in regimes inaccessible in conventional devices. Reactor relevant ICRF and LH drivers will be developed, utilizing high-field side launch platforms for low PMI. Alcator DX will inform the conceptual development and accelerate the readiness-for-deployment of next-step fusion facilities.

Finite element modelling of transport and drift effects in tokamak divertor and SOL

International Nuclear Information System (INIS)

Simard, M.; Marchand, R.; Boucher, C.; Gunn, J.P.

1996-01-01

A finite element code is used to simulate transport of a single-species plasma in the edge and divertor of a tokamak. The physical model is based on Braginskii's fluid equations for the conservation of particles, parallel momentum, ion and electron energy. In modelling recycling, transport of neutral density and energy is treated in the diffusion approximation. The electrostatic potential is obtained from the generalized Ohm's law. It is used to compute the electric field and the associated E x B drift. In a first approximation, transport is assumed to be ambipolar. The system of equations is discretized on an unstructured triangular mesh, thus permitting good spatial resolution near the X-point and an accurate description of divertor plates of arbitrary shape. Special care must be taken to prevent numerical corruption of the highly anisotropic thermal diffusion. Comparisons will be made between simulations and experimental results from TdeV. This will focus, in particular, on density and temperature profiles at the divertor plates, and on the plasma parallel velocity in the SOL. The asymmetry in the power deposited to the inner and outer divertors and the effect of magnetic field reversal will be considered. Comparisons with B2-Eirene simulation results will also be presented

Divertor design for the TITAN reversed-field-pinch reactor

International Nuclear Information System (INIS)

Cooke, P.I.H.; Bathke, C.G.; Blanchard, J.P.; Creedon, R.L.; Grotz, S.P.; Hasan, M.Z.; Orient, G.; Sharafat, S.; Werley, K.A.

1987-01-01

The design of the toroidal-field divertor for the TITAN high-power-density reversed-field-pinch reactor is described. The heat flux on the divertor target is limited to acceptable levels (≤ 10 MW/m 2 ) for liquid-lithium cooling by use of an open divertor geometry, strong radiation from the core and edge plasma, and careful shaping of the target surface. The divertor coils are based on the Integrated-Blanket-Coil approach to minimize the loss in breeding-blanket coverage due to the divertor. A tungsten-rhenium armour plate, chosen for reasons of sputtering resistance, and good thermal and mechanical properties, protects the vanadium-alloy coolant tubes

Small angle slot divertor concept for long pulse advanced tokamaks

Science.gov (United States)

Guo, H. Y.; Sang, C. F.; Stangeby, P. C.; Lao, L. L.; Taylor, T. S.; Thomas, D. M.

2017-04-01

SOLPS-EIRENE edge code analysis shows that a gas-tight slot divertor geometry with a small-angle (glancing-incidence) target, named the small angle slot (SAS) divertor, can achieve cold, dissipative/detached divertor conditions at relatively low values of plasma density at the outside midplane separatrix. SAS exhibits the following key features: (1) strong enhancement of the buildup of neutral density in a localized region near the plasma strike point on the divertor target; (2) spreading of the cooling front across the divertor target with the slot gradually flaring out from the strike point, thus effectively reducing both heat flux and erosion on the entire divertor target surface. Such a divertor may potentially provide a power and particle handling solution for long pulse advanced tokamaks.

Real-time mass measurement of dust particles deposited on vessel wall in a divertor simulator using quartz crystal microbalances

International Nuclear Information System (INIS)

Tateishi, Mizuki; Koga, Kazunori; Katayama, Ryu; Yamashita, Daisuke; Kamataki, Kunihiro; Seo, Hyunwoong; Itagaki, Naho; Shiratani, Masaharu; Ashikawa, Naoko; Masuzaki, Suguru; Nishimura, Kiyohiko; Sagara, Akio

2015-01-01

We are developing a dust monitoring method using quartz crystal microbalances (QCMs) equipped with a dust eliminating filter. Here we report a dust eliminating ratio of the filter and first measurement results of the QCMs in a divertor simulator. The volume of spherical dust in unit area on the filter and QCM under the filter were 2.09 × 10 −9 and 1.22 × 10 −10 m 3 m −2 , respectively. Thus, the dust eliminating ratio of the filter is 94.2%. The QCM without the filter gives deposition rate due to radicals and dust particles, whereas the QCM with the filter gives deposition rate predominantly due to radicals. From the results, we deduce information of mass fraction of dust particles in deposits

Reactor application of an improved bundle divertor

International Nuclear Information System (INIS)

Yang, T.F.; Ruck, G.W.; Lee, A.Y.; Smeltzer, G.; Prevenslik, T.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

Comparison of 2D simulations of detached divertor plasmas with divertor Thomson measurements in the DIII-D tokamak

Directory of Open Access Journals (Sweden)

T.D. Rognlien

2017-08-01

Full Text Available A modeling study is reported using new 2D data from DIII-D tokamak divertor plasmas and improved 2D transport model that includes large cross-field drifts for the numerically difficult low anomalous transport regime associated with the H-mode. The data set, which spans a range of plasma densities for both forward and reverse toroidal magnetic field (Bt, is provided by divertor Thomson scattering (DTS. Measurements utilizing X-point sweeping give corresponding 2D profiles of electron temperature (Te and density (ne across both divertor legs for individual discharges. The simulations focus on the open magnetic field-line regions, though they also include a small region of closed field lines. The calculations show the same features of in/out divertor plasma asymmetries as measured in the experiment, with the normal Bt direction (ion ∇B drift toward the X-point having higher ne and lower Te in the inner divertor leg than outer. Corresponding emission data for total radiated power shows a strong inner-divertor/outer-divertor asymmetry that is reproduced by the simulations. These 2D UEDGE transport simulations are enabled for steep-gradient H-mode conditions by newly implemented algorithms to control isolated grid-scale irregularities.

Development of divertor pumping system with superpermeable membrane

International Nuclear Information System (INIS)

Nakamura, Y.; Ohyabu, N.; Suzuki, H.; Nakahara, Y.; Livshits, A.; Notkin, M.; Alimov, V.; Busnyuk, A.

2000-01-01

A new divertor pumping system with superpermeable membranes of group Va-metals (Nb, V) is now under research and development. Properties of membrane pumping were investigated with the use of a plasma device simulating divertor plasma conditions. The deposition of metal (Fe) and non-metal (C) impurities on the membrane upstream surface results in a degradation of plasma driven superpermeation at the membrane temperature T m m ≥800 deg. C. The same temperature effect on superpermeation is observed at sputtering of membrane surface by energetic plasma ions. In addition, the first application of the membrane pumping to fusion devices has been carried out and a deuterium pumping through the membrane was demonstrated under the conditions of divertor plasma in the JFT-2M tokamak

Towards a physics-integrated view on divertor pumping

International Nuclear Information System (INIS)

Day, Chr.; Gleason-González, C.; Hauer, V.; Igitkhanov, Y.; Kalupin, D.; Varoutis, S.

2014-01-01

Highlights: • Physics-integrated design approaches are to be preferred over approaches based on simple requirement lists. • A physics-integrated assessment is presented for the divertor vacuum pumping system based on detachment onset conditions for the divertor. • This approach considers density dependent pump albedo to reflect the effects of gas recycling at the divertor and the changes in flow regime with density. • A comparison with DEMO indicates that the divertor pumping system for a pulsed DEMO scales less than linearly with fusion power. - Abstract: One key requirement to design the inner fuel cycle of a divertor tokamak is defined by the torus vessel gas throughput and composition, and the sub-divertor neutral pressure at which the exhaust gas has to be pumped. This paper illustrates how divertor physics aspects can be translated to requirements on the divertor vacuum pumping system. An example workflow is presented that links the realization of detachment conditions with the sub-divertor neutral gas flow patterns in order to determine the appropriate number of torus vacuum pumps. For the example case of a fusion DEMO size machine, it was found that 7 actively pumping cryopumps (ITER-type) are necessary to handle the gas throughput that is needed to manage the heat flux and densities related to detachment onset

Simulation of the ASDEX divertor performance after hardening

International Nuclear Information System (INIS)

Schneider, W.; Lackner, K.; Neuhauser, J.; Wunderlich, R.

1985-05-01

Two combined computer models - a fluid description of the plasma scrape-off layer (SOLID) and a Monte-Carlo code for the neutral gas dynamics (DEGAS) - are used to assess changes in the divertor performance expected from the modifications in geometry needed for hardening the ASDEX divertor chamber for long-pulse, high-power heating. Stand-alone DEGAS calculations with assumed fixed scrape-off plasma parameters predict a doubling of the neutral escape probability, which, however, still remains so low, that achievement of the high divertor recycling regime can be expected over roughly the same operational regime as before modifications. This conclusion is also supported by fully self-consistent calculations with the combined model. Due to the reduced divertor, a significant reduction is predicted in the divertor time constant, which is expected to affect transient phenomena. (orig.)

The lithium vapor box divertor

International Nuclear Information System (INIS)

Goldston, R J; Schwartz, J; Myers, R

2016-01-01

It has long been recognized that volumetric dissipation of the plasma heat flux from a fusion power system is preferable to its localized impingement on a material surface. Volumetric dissipation mitigates both the anticipated very high heat flux and intense particle-induced damage due to sputtering. Recent projections to a tokamak demonstration power plant suggest an immense upstream parallel heat flux, of order 20 GW m −2 , implying that fully detached operation may be a requirement for the success of fusion power. Building on pioneering work on the use of lithium by Nagayama et al and by Ono et al as well as earlier work on the gas box divertor by Watkins and Rebut, we present here a concept for a lithium vapor box divertor, in which lithium vapor extracts momentum and energy from a fusion-power-plant divertor plasma, using fully volumetric processes. At the high powers and pressures that are projected this requires a high density of lithium vapor, which must be isolated from the main plasma in order to avoid lithium build-up on the chamber walls or in the plasma. Isolation is achieved through a powerful multi-box differential pumping scheme available only for condensable vapors. The preliminary box-wise calculations are encouraging, but much more work is required to demonstrate the practical viability of this scheme, taking into account at least 2D plasma and vapor flows within and between the vapor boxes and out of the vapor boxes to the main plasma. (paper)

Divertor characterization experiments

International Nuclear Information System (INIS)

Porter, G.D.; Allen, S.; Fenstermacher, M.; Hill, D.; Brown, M.; Jong, R.A.; Rognlien, T.; Rensink, M.; Smith, G.; Stambaugh, R.; Mahdavi, M.A.; Leonard, A.; West, P., Evans, T.

1996-01-01

Recent DIII-D experiments with enhanced Scrape-off Layer (SOL) diagnostics permit detailed characterization of the SOL and divertor plasma under various operating conditions. We observe two distinct plasma modes: attached and detached divertor plasmas. Detached plasmas are characterized by plate temperatures of only 1 to 2 eV. Simulation of detached plasmas using the UEDGE code indicate that volume recombination and charge exchange play an important role in achieving detachment. When the power delivered to the plate is reduced by enhanced radiation to the point that recycled neutrals can no longer be efficiently ionized, the plate temperature drops from around 10 eV to 1-2 eV. The low temperature region extends further off the plate as the power continues to be reduced, and charge exchange processes remove momentum, reducing the plasma flow. Volume recombination becomes important when the plasma flow is reduced sufficiently to permit recombination to compete with flow to the plate

Divertor remote handling for DEMO: Concept design and preliminary FMECA studies

Energy Technology Data Exchange (ETDEWEB)

Carfora, D., E-mail: dario.carfora@gmail.com [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Di Gironimo, G. [ENEA/CREATE/Università degli studi Napoli Federico II, 80125 Napoli (Italy); Järvenpää, J. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland); Huhtala, K. [Tampere University of Technology, Korkeakoulunkatu 6, 33720 Tampere (Finland); Määttä, T.; Siuko, M. [VTT Technical Research Centre of Finland, P.O. Box 1300, FI-33101 Tampere (Finland)

2015-10-15

Highlights: • Concept design of the RH system for the DEMO fusion power plant. • Divertor mover: hydraulic telescopic boom concept design. • An alternative solution to ITER rack and pinion divertor mover (CMM). • Divertor cassettes end effector studies. • FMECA studies started on the DEMO divertor mover. - Abstract: The paper describes a concept design of a remote handling (RH) system for replacing divertor cassettes and cooling pipes in future DEMO fusion power plant. In DEMO reactor design important considerations are the reactor availability and reliable maintenance operations. The proposed divertor mover is a hydraulic telescopic boom driven from the transportation cask through the maintenance tunnel of the reactor. The boom is divided in three sections and it is driving an end-effector in order to perform the scheduled operations of maintenance inside the vacuum vessel. Two alternative designs of the end effector to grip and manipulate the divertor cassette are presented in this work. Both concepts are hydraulically actuated, based on ITER previous studies. The divertor cassette end-effector consists of a lifting arm linked to the divertor mover, a tilting plate, a cantilever arm and a hook-plate. Taking advantage of the ITER RH background and experience, the proposed hydraulic RH system is compared with the rack and pinion system currently designed for ITER and is an object of simulations at Divertor Test Platform (DTP2) in VTT's Labs of Tampere, Finland. Pros and cons will be put in evidence.

Divertor design through shape optimization

International Nuclear Information System (INIS)

Dekeyser, W.; Baelmans, M.; Reiter, D.

2012-01-01

Due to the conflicting requirements, complex physical processes and large number of design variables, divertor design for next step fusion reactors is a challenging problem, often relying on large numbers of computationally expensive numerical simulations. In this paper, we attempt to partially automate the design process by solving an appropriate shape optimization problem. Design requirements are incorporated in a cost functional which measures the performance of a certain design. By means of changes in the divertor shape, which in turn lead to changes in the plasma state, this cost functional can be minimized. Using advanced adjoint methods, optimal solutions are computed very efficiently. The approach is illustrated by designing divertor targets for optimal power load spreading, using a simplified edge plasma model (copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Analyses of microstructure, composition and retention of hydrogen isotopes in divertor tiles of JET with the ITER-like wall

Science.gov (United States)

Masuzaki, S.; Tokitani, M.; Otsuka, T.; Oya, Y.; Hatano, Y.; Miyamoto, M.; Sakamoto, R.; Ashikawa, N.; Sakurada, S.; Uemura, Y.; Azuma, K.; Yumizuru, K.; Oyaizu, M.; Suzuki, T.; Kurotaki, H.; Hamaguchi, D.; Isobe, K.; Asakura, N.; Widdowson, A.; Heinola, K.; Jachmich, S.; Rubel, M.; contributors, JET

2017-12-01

Results of the comprehensive surface analyses of divertor tiles and dusts retrieved from JET after the first ITER-like wall campaign (2011-2012) are presented. The samples cored from the divertor tiles were analyzed. Numerous nano-size bubble-like structures were observed in the deposition layer on the apron of the inner divertor tile, and a beryllium dust with the same structures were found in the matter collected from the inner divertor after the campaign. This suggests that the nano-size bubble-like structures can make the deposition layer to become brittle and may lead to cracking followed by dust generation. X-ray photoelectron spectroscopy analyses of chemical states of species in the deposition layers identified the formation of beryllium-tungsten intermetallic compounds on an inner vertical tile. Different tritium retention profiles along the divertor tiles were observed at the top surfaces and at deeper regions of the tiles by using the imaging plate technique.

Advanced divertor concepts

International Nuclear Information System (INIS)

Ohyabu, N.; Komori, A.; Sagara, A.; Suzuki, H.; Morisaki, T.; Masuzaki, S.; Watanabe, T.; Noda, N.; Motojima, O.

1996-01-01

LHD divertor development program has generated various innovative divertor concepts and technologies which will help to improve the plasma performance in both helical and tokamak devices. They are two divertor operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement). Local island divertor geometry has also been proposed. This new divertor has been successfully tested in the CHS device and is planned to be installed in the LHD device. In addition, technological development of new efficient hydrogen pumping schemes (carbon sheet pump and membrane pump) are being pursued for enhancement of the divertor control capability. 17 refs., 8 figs

Divertor heat flux control and plasma-material interaction

International Nuclear Information System (INIS)

Kikuchi, Yusuke; Nagata, Masayoshi; Sawada, Keiji; Takamura, Shuichi; Ueda, Yoshio

2014-01-01

Development of reliable radiative-cooling divertors is essential in DEMO reactor because it uses low-activation materials with low heat removal and the plasma heat flux exhausted from the confined region is 5 times as large as in ITER. It is important to predict precisely the heat and particle flux toward the divertor plate by simulation. In this present article, theoretical and experimental data of the reflection, secondary emission and surface recombination coefficients of the divertor plate by ion bombardment are given and their effects on the power transmission coefficient are discussed. In addition, some topics such as the erosion process of the divertor plate by ELM and the plasma disruption, the thermal shielding due to the vapor layer on the divertor plate and the formation of fuzz structure on W by helium plasma irradiation, are described. (author)

First tests of diagnostic mirrors in a tokamak divertor: An overview of experiments in DIII-D

International Nuclear Information System (INIS)

Litnovsky, A.; Rudakov, D.L.; De Temmerman, G.; Wienhold, P.; Philipps, V.; Samm, U.; McLean, A.G.; West, W.P.; Wong, C.P.C.; Brooks, N.H.; Watkins, J.G.; Wampler, W.R.; Stangeby, P.C.; Boedo, J.A.; Moyer, R.A.; Allen, S.L.; Fenstermacher, M.E.; Groth, M.; Lasnier, C.J.; Boivin, R.L.

2008-01-01

Mirrors will be used in ITER in all optical diagnostic systems observing the plasma radiation in the ultraviolet, visible and infrared ranges. Diagnostic mirrors in ITER will suffer from electromagnetic radiation, energetic particles and neutron irradiation. Erosion due to impact of fast neutrals from plasma and deposition of plasma impurities may significantly degrade optical and polarization characteristics of mirrors influencing the overall performance of the respective diagnostics. Therefore, maintaining the best possible performance of mirrors is of the crucial importance for the ITER optical diagnostics. Mirrors in ITER divertor are expected to suffer from deposition of impurities. The dedicated experiment in a tokamak divertor was needed to address this issue. Investigations with molybdenum diagnostic mirrors were made in DIII-D divertor. Mirror samples were exposed at different temperatures in the private flux region to a series of ELMy H-mode discharges with partially detached divertor plasmas. An increase of temperature of mirrors during the exposure generally led to the mitigation of carbon deposition, primarily due to temperature-enhanced chemical erosion of carbon layers by D atoms. Finally, for the mirrors exposed at the temperature of ∼160 o C neither carbon deposition nor degradation of optical properties was detected

Study of the radiation in divertor plasmas; Etude du rayonnement dans les plasmas de divertor

Energy Technology Data Exchange (ETDEWEB)

Laugier, F

2000-10-19

We have studied the cooling of the edge plasma by radiation in the divertor volume, in order to optimize the extraction of power in tokamaks and to limit the wall erosion. In attached divertor plasmas experiments, the concentration of intrinsic impurities at the edge is related to the response of the wall to the incident energy flow of plasma, depending on a phenomenological law. We carried out an analysis of the radiation according to this law and to the control parameters of the discharges. The largest radiated fraction and best synergy are obtained when the concentration of intrinsic impurities strongly increases with the energy of incident plasma. On the other hand, the erosion of the wall is stronger. In detached plasmas, we proved that the performances in terms of incident plasma energy loss and pressure loss are optimal when the density of the slowest neutrals is strong at the edge and when their radial penetration is small. On Tore Supra, we highlighted the correlations between the maximum Mach number of incident plasma flow, the radiation front and the penetration of the neutrals. A simple diagnostic based on the localization of the maximum Mach number proves that detached mode is not optimal on Tore Supra, because the radial penetration of the slowest neutrals is not sufficiently small. In the last part, we obtained the three-dimensional topology of the radiation in the ergodic divertor using a spectral analysis code and boundary conditions consistent with the temperature distribution on the wall. The radiation is maximum in front of the divertor modules. As a consequence, radiated power is underestimated by standards measurements of Tore Supra that are located between the modules. We finally showed that the profiles of temperature along the field lines are modulated, this is specific to the ergodic divertor. (author)

Modeling results for a linear simulator of a divertor

International Nuclear Information System (INIS)

Hooper, E.B.; Brown, M.D.; Byers, J.A.; Casper, T.A.; Cohen, B.I.; Cohen, R.H.; Jackson, M.C.; Kaiser, T.B.; Molvik, A.W.; Nevins, W.M.; Nilson, D.G.; Pearlstein, L.D.; Rognlien, T.D.

1993-01-01

A divertor simulator, IDEAL, has been proposed by S. Cohen to study the difficult power-handling requirements of the tokamak program in general and the ITER program in particular. Projections of the power density in the ITER divertor reach ∼ 1 Gw/m 2 along the magnetic fieldlines and > 10 MW/m 2 on a surface inclined at a shallow angle to the fieldlines. These power densities are substantially greater than can be handled reliably on the surface, so new techniques are required to reduce the power density to a reasonable level. Although the divertor physics must be demonstrated in tokamaks, a linear device could contribute to the development because of its flexibility, the easy access to the plasma and to tested components, and long pulse operation (essentially cw). However, a decision to build a simulator requires not just the recognition of its programmatic value, but also confidence that it can meet the required parameters at an affordable cost. Accordingly, as reported here, it was decided to examine the physics of the proposed device, including kinetic effects resulting from the intense heating required to reach the plasma parameters, and to conduct an independent cost estimate. The detailed role of the simulator in a divertor program is not explored in this report

Divertor design and its integration into the ITER-FEAT machine

International Nuclear Information System (INIS)

Janeschitz, G.; Antipenkov, A.; Federici, G.; Ibbott, C.; Kukushkin, A.; Ladd, P.; Martin, E.; Tivey, R.

2001-01-01

The physics of the edge and divertor plasma is strongly coupled with the divertor and the fuel cycle design. Due to the limited space available the design as well as the remote maintenance approach for the ITER divertor are highly optimized to allow maximum space for the divertor plasma. Several auxiliary systems (e.g. in vessel viewing, glow discharge electrodes...) as well as a part of the pumping and fuelling system have to be integrated together with the divertor into the lower level of the ITER machine. Two main options exist for the choice of the plasma-facing material in the divertor, i.e. W and CFC. Based on already existing R and D results one can be optimistic that the material choice will be mainly based on physics considerations and material issues (e.g. C-T co-deposition). The requirements for the ITER fuel cycle arise from plasma physics as well as from the envisaged operation scenarios. Due to the complex dynamic relationship of the fuel cycle subsystems among themselves and with the plasma, codes are employed for their optimization. This paper elaborates these interacting issues and gives the latest design status. (author)

LHD helical divertor

International Nuclear Information System (INIS)

Ohyabu, N.; Watanabe, T.; Ji Hantao

1993-07-01

The Large Helical Device (LHD) now under construction is a heliotron/torsatron device with a closed divertor system. The edge LHD magnetic structure has been studied in detail. A peculiar feature of the configuration is existence of edge surface layers, a complicated three dimensional magnetic structure which does not, however, seem to hamper the expected divertor functions. Two divertor operational modes are being considered for the LHD experiment, high density, cold radiative divertor operation as a safe heat removal scheme and high temperature divertor plasma operation. In the latter operation, a divertor plasma with temperature of a few kev, generated by efficient pumping, expects to lead to significant improvement in core plasma confinement. Conceptual designs of the LHD divertor components are under way. (author)

Visible spectroscopy in the DIII-D divertor

International Nuclear Information System (INIS)

Brooks, N.H.; Fehling, D.; Hillis, D.L.; Klepper, C.C.; Naumenko, N.; Tugarinov, S.; Whyte, D.G.

1996-06-01

Spectroscopy measurements in the DIII-D divertor have been carried out with a survey spectrometer which provides simultaneous registration of the visible spectrum over the region 400--900 nm with a resolution of 0.2 nm. Broad spectral coverage is achieved through use of a fiberoptic transformer assembly to map the curved focal plane of a fast (f/3) Rowland spectrograph into a rastered format on the rectangular sensor area of a two-dimensional CCD camera. Vertical grouping of pixels during CCD readout integrates the signal intensity over the height of each spectral segment in the rastered image, minimizing readout time. For the full visible spectrum, readout time is 50 ms. Faster response time (< 10 ms) may be obtained by selecting for readout just a small number of the twenty spectral segments in the image on the CCD. Simultaneous recording of low charge states of carbon, oxygen and injected impurities has yielded information about gas recycling and impurity behavior at the divertor strike points. Transport of lithium to the divertor region during lithium pellet injection has been studied, as well as cumulative deposition of lithium on the divertor targets from pellet injection over many successive discharges

Modelling of Power Fluxes during Thermal Quenches

International Nuclear Information System (INIS)

Konz, C.; Coster, D. P.; Lackner, K.; Pautasso, G.

2005-01-01

outboard and inboard divertor heat pulses. Our simulations show a strong variation of the width of the power deposition profile with the pre-disruptive divertor conditions. In the experiment, power deposition profiles broaden significantly during plasma disruptions. Within the scope our transport model in B2, this behaviour is found for the cases with low densities, while profile widths change only little for the cases with strong neutral puffing. Further investigation, both experimental and theoretical, is needed for a better understanding of the thermal quench phase of disruptions. (Author)

Long-term erosion and re-deposition of carbon in the divertor region of JT-60U

International Nuclear Information System (INIS)

Gotoh, Y.; Tanabe, T.; Ishimoto, Y.; Masaki, K.; Arai, T.; Kubo, H.; Tsuzuki, K.; Miya, N.

2006-01-01

Erosion and redeposition profiles of carbon tiles used in the W-shaped divertor of JT-60U with all carbon plasma facing wall are studied. The inner divertor is mostly covered by carbon redeposited layers, while the outer divertor mostly eroded. In the dome region, the erosion dominates on the inner dome-wing, while redeposited on the outer dome-wing. The redeposited layers on the outer dome-wing show very clear columnar structures indicating local carbon transport form the outer divertor to the outer dome-wing. The weight gain by the redeposition extrapolated to the whole divertor area is 0.55 kg. Since the extrapolated total erosion is about 0.33 kg, the remaining 0.22 kg must be originated from the main chamber erosion. Significant amount of the redeposition is caused locally by multiple processes of erosion, ionization and prompt redeposition toward inboard direction owing to gyration along magnetic filed line. This inboard transport is one of the reasons for small redeposition on the plasma shadowed area of the W shaped divertor of JT-60U with pumping slots placed at the bottoms side

A snowflake divertor: a possible solution to the power exhaust problem for tokamaks

Energy Technology Data Exchange (ETDEWEB)

Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Umansky, M. V.

2012-11-21

This paper summarizes recent progress in the theory of a snowflake divertor, a possible path to reduce both steady-state and intermittent heat loads on the divertor plates to an acceptable level. The most important feature of a SF divertor is the presence of a large zone of a very weak poloidal magnetic field around the poloidal field (PF) null. Qualitative explanation of a variety of new features characteristic of a SF divertor is provided based on simple scaling relations. The main part of the paper is focused on the concept of spreading of the heat flux by curvature-driven convection near the PF null. References to experimental results from the NSTX and TCV tokamaks are provided.

Structural analysis of the ITER Divertor toroidal rails

Energy Technology Data Exchange (ETDEWEB)

Viganò, F., E-mail: Fabio.Vigano@LTCalcoli.it [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Escourbiac, F.; Gicquel, S.; Komarov, V. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Lucca, F. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy); Merola, M. [ITER Organization, Route de Vinon sur Verdon, 13115 St. Paul lez Durance (France); Ngnitewe, R. [L.T. Calcoli SaS, Piazza Prinetti 26/B, 23807 Merate (Italy)

2013-10-15

The Divertor is one of the most technically challenging components of the ITER machine, which has the main function of extracting the power conducted in the scrape-off layer while maintaining the plasma purity. There are 54 Divertor cassettes installed in the vacuum vessel (VV). Each cassette body (CB) is fastened on the inner and outer concentric Divertor toroidal rails. The comprehensive assessment (in accordance with the Structural Design Criteria for ITER In-vessel Components: ITER SDC-IC) of the Divertor toroidal rails has been performed during design activity based on performing of thermal and stress analyses at operating conditions of neutron stage of ITER operation. This paper outlines the engineering aspects of the ITER Divertor toroidal rails and focuses on some critical regions of the present design highlighted by the performed structural assessment. The structural assessment has been performed with help of using Finite Element (FE) Abaqus code and based on criteria given by ITER SDC-IC.

Actively convected liquid metal divertor

International Nuclear Information System (INIS)

Shimada, Michiya; Hirooka, Yoshi

2014-01-01

The use of actively convected liquid metals with j × B force is proposed to facilitate heat handling by the divertor, a challenging issue associated with magnetic fusion experiments such as ITER. This issue will be aggravated even more for DEMO and power reactors because the divertor heat load will be significantly higher and yet the use of copper would not be allowed as the heat sink material. Instead, reduced activation ferritic/martensitic steel alloys with heat conductivities substantially lower than that of copper, will be used as the structural materials. The present proposal is to fill the lower part of the vacuum vessel with liquid metals with relatively low melting points and low chemical activities including Ga and Sn. The divertor modules, equipped with electrodes and cooling tubes, are immersed in the liquid metal. The electrode, placed in the middle of the liquid metal, can be biased positively or negatively with respect to the module. The j × B force due to the current between the electrode and the module provides a rotating motion for the liquid metal around the electrodes. The rise in liquid temperature at the separatrix hit point can be maintained at acceptable levels from the operation point of view. As the rotation speed increases, the current in the liquid metal is expected to decrease due to the v × B electromotive force. This rotating motion in the poloidal plane will reduce the divertor heat load significantly. Another important benefit of the convected liquid metal divertor is the fast recovery from unmitigated disruptions. Also, the liquid metal divertor concept eliminates the erosion problem. (letter)

The MAST improved divertor

International Nuclear Information System (INIS)

Darke, A.C.; Hayward, R.J.; Counsell, G.F.; Hawkins, K.

2005-01-01

The Mega Amp Spherical Tokamak (MAST) at Culham is one of the leading world machines studying the spherical tokamak (ST) concept. At the time of the initial construction in 1998 little was known about the sort of divertor structures that would be required in an ST. The machine was therefore provided with relatively rudimentary structures that were designed mostly to protect important components from the hot plasma. While these have served the machine well it was accepted that they might not be suitable when operating MAST to its full potential. The years of experience of operating MAST have led to the design, manufacture and now installation of a new divertor, the MAST improved divertor (MID), that should be able to cope with the full performance of the machine. The design is based on imbricated (fan-shaped) disks of tiles at the top and bottom of the machine for the outer strike points, giving an excellent compromise between power handling and diagnostic access, with substantial new centre column strike point armour and a shaped plate in between. High purity graphite is chosen as the plasma facing material in preference to CFC since in this case it has a better balance of performance and cost. The lower imbricated disk is insulated in alternate sectors for studies of divertor biasing and extensive diagnostics and additional inboard gas injection are included

Application of the radiating divertor approach to innovative tokamak divertor concepts

Energy Technology Data Exchange (ETDEWEB)

Petrie, T.W., E-mail: petrie@fusion.gat.com [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Allen, S.L.; Fenstermacher, M.E. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Groebner, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Holcomb, C.T. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Kolemen, E. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); La Haye, R.J. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Leonard, A.W.; Luce, T.C. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); McLean, A.G. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Maingi, R. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Moyer, R.A. [University of California San Diego, 9500 Gilman Dr., La Jolla, CA 92093-0417 (United States); Solomon, W.M. [Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543-0451 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, 700 East Ave, Livermore, CA 94550 (United States); Turco, F. [Columbia University, 2960 Broadway, New York, NY 10027 (United States); Watkins, J.G. [Sandia National Laboratory, PO Box 5800, Albuquerque, NM 87185 (United States)

2015-08-15

We survey the results of recent DIII-D experiments that tested the effectiveness of three innovative tokamak divertor concepts in reducing divertor heat flux while still maintaining acceptable energy confinement under neon/deuterium-based radiating divertor (RD) conditions: (1) magnetically unbalanced high performance double-null divertor (DND) plasmas, (2) high performance double-null “Snowflake” (SF-DN) plasmas, and (3) single-null H-mode plasmas having different isolation from their divertor targets. In general, all three concepts adapt well to RD conditions, achieving significant reduction in divertor heat flux (q{sub ⊥p}) and maintaining high performance metrics, e.g., 50–70% reduction in peak divertor heat flux for DND and SF-DN plasmas that are characterized by β{sub N} ≅ 3.0 and H{sub 98(y,2)} ≈ 1.35. It is also demonstrated that q{sub ⊥p} could be reduced ≈50% by extending the parallel connection length (L{sub ||-XPT}) in the scrape-off layer between the X-point and divertor targets over a variety of the RD and non-RD environments tested.

A numerical study of plasma detachment conditions in JET divertor plasmas

Energy Technology Data Exchange (ETDEWEB)

Simonini, R; Corrigan, G; Radford, G; Spence, J; Taroni, A; Weber, S [Commission of the European Communities, Abingdon (United Kingdom). JET Joint Undertaking

1994-07-01

Simulation results obtained with the EDGE2D/U code confirm that for a given particle inventory in the SOL (including the divertor), the main parameter determining whether or not particle, momentum and energy detachment occurs, is the residual power P - P{sub lost}, where P is the total power entering the SOL and P{sub lost} is the power lost by transport to walls and by volume losses in the SOL outside the region where detachment takes place. For particle contents leading to reasonable values of the separatrix mid-plane density, detachment is found if the residual power is low enough. Typically the residual power must be inferior to 3 MW for good detachment, with the exact value depending on the geometry of the divertor, the transport assumptions and the neutral recirculation scheme. The results show that divertor plasma conditions relevant for the study of power exhaust and impurity control problems are possible in JET. 9 refs., 2 figs., 1 tab.

Engineering design of a Radiative Divertor for DIII-D

International Nuclear Information System (INIS)

Smith, J.P.; Allen, S.L.; Anderson, P.M.; Baxi, C.B.; Chin, E.; Fenstermacher, M.E.; Hill, D.N.; Hollerbach, M.A.; Hyatt, A.W.; Junge, R.; Mahdavi, M.A.; Porter, G.D.; Redler, K.; Reis, E.E.; Schaffer, M.J.; Sevier, D.L.; Stambaugh, R.D.

1995-01-01

A new divertor called the Radiative Divertor is presently being designed for the DIII-D tokamak. Input from tokamak experiments and modeling form the basis for the new design. The Radiative Divertor is intended to reduce the heat flux on the divertor plates by dispersing the power with radiation. Gas puffing experiments in the current open divertor have shown a reduction of the divertor heat flux with either deuterium or impurity puffing. However, either the plasma density (D 2 ) or the core Z eff (impurities) increases in these experiments. The radiative divertor uses a slot structure to isolate the divertor plasma region from the area surrounding the core plasma. Modeling has shown that the Radiative Divertor hardware will provide better baffling and particle control and thereby minimize the effect of the gas puffing in the divertor region on the plasma core. In addition, the Radiative Divertor structure will allow density control in plasma shapes with high triangularity (>0.8) required for advanced tokamak operation. The divertor structure allows for operation in either double or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. Biasing is an integral part of the design and is based on experience at the Tokamak de Varennes (TdeV) and DIII-D. Boron nitride tiles electrically insulate the inner and outer strike points and a low current electrode is used to apply a radial electric field to the scrape-off layer. TdeV has shown that biasing can provide particle and impurity control. The design is extremely flexible, and will allow physics studies of the effect of slot width and height. This is extremely important, as the amount of chamber volume needed for the divertor in future machines such as International Thermonuclear Experiment Reactor (ITER) and Tokamak Physics Experiment (TPX) must be determined. (orig./WL)

Physics design and experimental study of tokamak divertor

International Nuclear Information System (INIS)

Yan Jiancheng; Gao Qingdi; Yan Longwen; Wang Mingxu; Deng Baiquan; Zhang Fu; Zhang Nianman; Ran Hong; Cheng Fayin; Tang Yiwu; Chen Xiaoping

2007-06-01

The divertor configuration of HL-2A tokamak is optimized, and the plasma performance in divertor is simulated with B2-code. The effects of collisionality on plasma-wall transition in the scrape-off layer of divertor are investigated, high performances of the divertor plasma in HL-2A are simulated, and a quasi- stationary RS operation mode is established with the plasma controlled by LHCD and NBI. HL-2A tokamak has been successfully operated in divertor configuration. The major parameters: plasma current I p =320 kA, toroidal field B t =2.2 T, plasma discharger duration T d =1580 ms ware achieved at the end of 2004. The preliminary experimental researches of advanced diverter have been carried out. Design studies of divertor target plate for high power density fusion reactor have been carried out, especially, the physical processes on the surface of flowing liquid lithium target plate. The exploration research of improving divertor ash removal efficiency and reducing tritium inventory resulting from applying the RF ponderomotive force potential is studied. The optimization structure design studies of FEB-E reactor divertor are performed. High flux thermal shock experiments were carried on tungsten and carbon based materials. Hot Isostatic Press (HIP) method was employed to bond tungsten to copper alloys. Electron beam simulated thermal fatigue tests were also carried out to W/Cu bondings. Thermal desorption and surface modification of He + implanted into tungsten have been studied. (authors)

Lower hybrid wave edge power loss quantification on the Alcator C-Mod tokamak

Science.gov (United States)

Faust, I. C.; Brunner, D.; LaBombard, B.; Parker, R. R.; Terry, J. L.; Whyte, D. G.; Baek, S. G.; Edlund, E.; Hubbard, A. E.; Hughes, J. W.; Kuang, A. Q.; Reinke, M. L.; Shiraiwa, S.; Wallace, G. M.; Walk, J. R.

2016-05-01

For the first time, the power deposition of lower hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal, and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt ( t Radiofrequency (LHRF) power. LHRF power was found to absorb more strongly in the edge at higher densities. It is found that a majority of this edge-deposited power is promptly conducted to the divertor. This correlates with the loss of current drive efficiency at high density previously observed on Alcator C-Mod, and displaying characteristics that contrast with the local RF edge absorption seen on other tokamaks. Measurements of ionization in the active divertor show dramatic changes due to LHRF power, implying that divertor region can be a key for the LHRF edge power deposition physics. These observations support the existence of a loss mechanism near the edge for LHRF at high density ( n e > 1.0 × 10 20 (m-3)). Results will be shown addressing the distribution of power within the SOL, including the toroidal symmetry and radial distribution. These characteristics are important for deducing the cause of the reduced LHCD efficiency at high density and motivate the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch.

Divertor Heat Flux Reduction and Detachment in the National Spherical Torus eXperiment.

Science.gov (United States)

Soukhanovskii, Vsevolod

2007-11-01

Steady-state handling of the heat flux is a critical divertor issue for both the International Thermonuclear Experimental Reactor and spherical torus (ST) devices. Because of an inherently compact divertor, it was thought that ST-based devices might not be able to fully utilize radiative and dissipative divertor techniques based on induced power and momentum loss. However, initial experiments conducted in the National Spherical Torus Experiment in an open geometry horizontal carbon plate divertor using 0.8 MA 2-6 MW NBI-heated lower single null H-mode plasmas at the lower end of elongations κ=1.8-2.4 and triangularities δ=0.45-0.75 demonstrated that high divertor peak heat fluxes, up to 6-10 MW/ m^2, could be reduced by 50-75% using a high-recycling radiative divertor regime with D2 injection. Furthermore, similar reduction was obtained with a partially detached divertor (PDD) at high D2 injection rates, however, it was accompanied by an X-point MARFE that quickly led to confinement degradation. Another approach takes advantage of the ST relation between strong shaping and high performance, and utilizes the poloidal magnetic flux expansion in the divertor region. Up to 60 % reduction in divertor peak heat flux was achieved at similar levels of scrape-off layer power by varying plasma shaping and thereby increasing the outer strike point (OSP) poloidal flux expansion from 4-6 to 18-22. In recent experiments conducted in highly-shaped 1.0-1.2 MA 6 MW NBI heated H-mode plasmas with divertor D2 injection at rates up to 10^22 s-1, a PDD regime with OSP peak heat flux 0.5-1.5 MW/m^2 was obtained without noticeable confinement degradation. Calculations based on a two point scrape-off layer model with parameterized power and momentum losses show that the short parallel connection length at the OSP sets the upper limit on the radiative exhaust channel, and both the impurity radiation and large momentum sink achievable only at high divertor neutral pressures are required

Steady state and transient power handling in JET

International Nuclear Information System (INIS)

Matthews, G.F.

2003-01-01

Steady state and transient power deposition profiles have been measured in the JET MIIGB divertor using improved diagnostics techniques involving the use of fast infra-red, thermocouples and Langmuir probe arrays. In unfuelled type I ELMy H-modes a very narrow power profile is observed at the outer target which we associate with the ion channel. Systematic parameter scans have been carried out and our analysis shows that the average power width scaling is consistent with a classical dependence of perpendicular transport in the SOL. Using the fast IR capability the factors such as rise time, broadening, variability and in/out asymmetry have been studied and lead to the conclusion that type I ELMs in ITER may fall just below the material ablation limits. JET disruptions are very different from type I ELMs in that only a small fraction of the thermal energy reaches the divertor and what does arrive is distributed uniformly over the divertor area. This is very different from the current ITER assumption which puts most of the energy from the thermal quench onto the divertor strike points. (author)

Magnetic divertor design for the compact reversed-field pinch reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Miller, R.L.; Krakowski, R.A.

1984-01-01

A recently completed design of a pumped-limiter-based Compact Reversed-Field Pinch Reactor is used to estimate for the first time the impact of magnetic divertors. A range of divertor options for the low-toroidal-field RFP is examined, and a design selection is made constrained by consideration of field ripple (magnetic island), blanket displacement, recirculating power, cost, heat flux, and access. Design choices based on diversion of minority (toroidal) field lead to a preference for (poloidally) symmetric or bundle divertor geometries

Divertor plasma modification by divertor biasing and edge ergodization in JFT-2M

International Nuclear Information System (INIS)

Shoji, T.; Nagashima, K.; Tamai, H.; Ohdachi, S.; Miura, Y.; Ohasa, K.; Maeda, H.; Ohyabu, N.; Leonard, A.W.; Aikawa, H.; Fujita, T.; Hoshino, K.; Kawashima, H.; Matsuda, T.; Maeno, M.; Mori, M.; Ogawa, H.; Shimada, M.; Uehara, K.; Yamauchi, T.

1995-01-01

The effects of divertor biasing and edge ergodization on the divertor plasma have been investigated in the JFT-2M tokamak. Experimental results show; (1) The differential divertor biasing can change the in/out asymmetry of the divertor plasma. It especially changes the density on the ion side divertor plasma. The in/out electron pressure difference has a good correlation with the biasing current. (2) The unipolar divertor biasing can change the density profile of divertor plasma. The radial electric field and shear flow are the cause for this change. (3) The electron temperature of the divertor plasma in the H-mode with frequent ELMs induced by edge ergodization is lower than that of usual H-mode. That is due to the enhancement of the radial particle flux by frequent ELMs, ((orig.))

Design integration of liquid surface divertors

International Nuclear Information System (INIS)

Nygren, R.E.; Cowgill, D.F.; Ulrickson, M.A.; Nelson, B.E.; Fogarty, P.J.; Rognlien, T.D.; Rensink, M.E.; Hassanein, A.; Smolentsev, S.S.; Kotschenreuther, M.

2004-01-01

The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the divertor to enhance thermal mixing. For divertors with flowing liquid metals (or other electrically conductive fluids) MHD (magneto-hydrodynamics) effects are a major concern and can produce forces that redirect flow and suppress turbulence. An evaluation of Flibe (a molten salt) as a working fluid was done to assess a case in which the MHD forces could be largely neglected. Initial studies indicate that, for a tokamak with high power density, an integrated Flibe first wall and divertor does not seem workable. We have continued work with molten salts and replaced Flibe with Flinabe, a mixture of lithium, sodium and beryllium fluorides, that has some potential because of its lower melting temperature. Sn and Sn-Li have also been considered, and the initial evaluations on heat removal with minimal plasma contamination show promise, although the complicated 3D MHD flows cannot yet be fully modeled. Particle pumping in these design concepts is accomplished by conventional means (ports and pumps). However, trapping of hydrogen in these flowing liquids seems plausible and novel concepts for entrapping helium are also being studied

Steady state and transient power handling in JET

International Nuclear Information System (INIS)

Matthews, G.F.

2002-01-01

Recent JET experiments and analysis have demonstrated the importance of edge collisionality for the physics of divertor power loading both during and between ELMs. Since collisionality decreases strongly with machine size, JET routinely operates in an ITER relevant regime which is difficult or impossible to access in smaller devices. This new understanding has enabled us to develop more physically justifiable scalings for static and transient power deposition in ITER and demonstrates a need for kinetic models when simulating edge behaviour in JET and ITER. Steady state power loading in ITER is likely to be within limits provided that the divertor plasma is kept in the high recycling or detached regime. Extrapolations of the typical type I ELMs found in JET to ITER highlight the importance of developing regimes characterised by small ELMs, if surface ablation is to be avoided. Disruptive power loads measured in the JET divertor appear far more benign than would be expected from current ITER assumptions. (author)

Design and performance study of the helium-cooled T-tube divertor concept

International Nuclear Information System (INIS)

Ihli, T.; Raffray, A.R.; Abdel-Khalik, S.I.; Shin, S.

2007-01-01

The ARIES-CS study has been launched with the goal of developing through physics and engineering optimization an attractive power plant concept based on a compact stellarator configuration. The study included an effort to characterize the divertor location and corresponding heat load distribution, and to develop a He-cooled divertor concept that could accommodate a heat flux of at least 10 MW/m 2 , and that would integrate well with the other power core components. This paper describes the design study of this divertor concept, which, although developed for a compact stellarator, is well suited for a tokamak configuration also

Radiation loss and global energy balance of ohmically heated divertor discharge in JT-60 tokamak

International Nuclear Information System (INIS)

Koide, Yoshihiko; Yamada, Kimio; Yoshida, Hidetoshi; Nakamura, Hiroo; Niikura, Setsuo; Tsuji, Shunji

1986-03-01

Divertor experiment in JT-60 with a small divertor chamber has been successfully performed up to 1.6 MA discharge. Several divertor effects were experimentally confirmed as follows. Radiation loss in main plasma saturates with the increase of plasma current and its ratio to the input power is about 20 % at 1.5 MA. The rest of input power is exhausted into the divertor chamber and a half of it is dissipated as the radiation loss. Impurity accumulation is not observed during a few sec without internal MHD activity and gross impurity confinement time is several hundred msec. (author)

Development of liquid lithium divertor for fusion reactor

International Nuclear Information System (INIS)

Evtihkin, V. A.; Lyublinskij, I. E.; Vertkov, A.V.; Chumanov, A.V.; Shpolyanskij, V.N.

2000-01-01

Development of divertor is one of the most acute problems of the tokamak fusion reactor. The use of such materials as tungsten, beryllium, graphite and CFC's enabled to solve the problem to a certain extent fulfilling the need of the ITER project. The problem still rests unsolved for the DEMO-type reactors. Lithium if used as a material for high heat flux components may provide a successful solution of the problem. A concept of Li divertor based on the use of capillary-pore structures (CPS) is proposed and is being validated by a complex of experimental research and engineering developments. An optional concept of Li divertor for power removal at 400 MW in steady-state (DEMO-S project) is presented. The complex of experimental research is under way to prove the serviceability of the Li CPS in different conditions that would be realized in divertor

ADX: a high field, high power density, Advanced Divertor test eXperiment

Science.gov (United States)

Vieira, R.; Labombard, B.; Marmar, E.; Irby, J.; Shiraiwa, S.; Terry, J.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; ADX Team

2014-10-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment (ADX) - a tokamak specifically designed to address critical gaps in the world fusion research program on the pathway to FNSF/DEMO. This high field (6.5 tesla, 1.5 MA), high power density (P/S ~ 1.5 MW/m2) facility would utilize Alcator magnet technology to test innovative divertor concepts for next-step DT fusion devices (FNSF, DEMO) at reactor-level boundary plasma pressures and parallel heat flux densities while producing high performance core plasma conditions. The experimental platform would also test advanced lower hybrid current drive (LHCD) and ion-cyclotron range of frequency (ICRF) actuators and wave physics at the plasma densities and magnetic field strengths of a DEMO, with the unique ability to deploy launcher structures both on the low-magnetic-field side and the high-field side - a location where energetic plasma-material interactions can be controlled and wave physics is most favorable for efficient current drive, heating and flow drive. This innovative experiment would perform plasma science and technology R&D necessary to inform the conceptual development and accelerate the readiness-for-deployment of FNSF/DEMO - in a timely manner, on a cost-effective research platform. Supported by DE-FC02-99ER54512.

Heat removal capability of divertor coaxial tube assembly

International Nuclear Information System (INIS)

Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

1994-05-01

To deal with high power flowing in the divertor region, an advanced divertor concept with gas target has been proposed for use in ITER/EDA. The concept uses a divertor channel to remove the radiated power while allowing neutrals to recirculate. Candidate channel wall designs include a tube array design where many coaxial tubes are arranged in the toroidal direction to make louver. The coaxial tube consists of a Be protection tube encases many supply tubes wound helically around a return tube. V-alloy and hardened Cu-alloy have been proposed for use in the supply and return tubes. Some coolants have also been proposed for the design including pressurized He and liquid metals, because these coolants are consistent with the selection of coolants for the blanket and also meet the requirement of high temperature operation. In the coaxial tube design, the coolant area is restricted and brittle Be material is used under severe thermal cyclings. Thus, to obtain the coaxial tube with sufficient safety margin for the expected fusion power excursion, it is essential to understand its applicability limit. The paper discusses heat removal capability of the coaxial tube and recommends some design modifications. (author)

Modeling of combined effects of divertor closure and advanced magnetic configuration on detachment in DIII-D by SOLPS

Science.gov (United States)

Si, H.; Guo, H. Y.; Covele, B.; Leonard, A. W.; Watkins, J. G.; Thomas, D.; Ding, R.

2018-05-01

One of the major challenges facing the design and operation of next-step high-power steady-state fusion devices is to develop a divertor solution for handling power exhaust, while ensuring acceptable divertor target plate erosion, which necessitates access to divertor detachment at relative low main plasma densities compatible with current drive and high plasma confinement. Detailed modeling with SOLPS is carried out to examine the effect of divertor closure on detachment with the normal single null divertor (SD) configuration, as well as one of the advanced divertor configurations, such as x-divertor (XD) respectively. The SOLPS modeling for a high confinement plasma in DIII-D finds that increasing divertor closure with SD reduces the upstream separatrix density at the onset of detachment from 1.18× {{10}19} {{m}-3} to 0.88× {{10}19} {{m}-3} . Moreover, coupling the divertor closure with XD further promotes the onset of divertor detachment at a still lower upstream separatrix density, down to the value of 0.67× {{10}19} {{m}-3} , thus, showing that divertor closure and advanced magnetic configuration can work synergistically to facilitate divertor detachment.

Divertor modelling for conceptual studies of tokamak fusion reactor FDS-III

International Nuclear Information System (INIS)

Chen Yiping; Liu Songlin

2010-01-01

Divertor modelling for the conceptual studies of tokamak fusion reactor FDS-III was carried out by using the edge plasma code package B2.5-Eirene (SOLPS5.0). The modelling was performed by taking real MHD equilibrium and divertor geometry of the reactor into account. The profiles of plasma temperature, density and heat fluxes in the computational region and at the target plates have been obtained. The modelling results show that, with the fusion power P fu =2.6 GW and the edge density N edge =6.0x10 19 l/m 3 , the peak values of electron and ion heat fluxes at the outer target plate of divertor are respectively 93.92 MW/m 2 and 58.50 MW/m 2 . According to the modelling results it is suggested that some methods for reducing the heat fluxes at the target plates should be used in order to get acceptable level of power flux at the target plates for the divertor design of the reactor.

Divertor extreme ultraviolet (EUV) survey spectroscopy in DIII-D

Science.gov (United States)

McLean, Adam; Allen, Steve; Ellis, Ron; Jarvinen, Aaro; Soukhanovskii, Vlad; Boivin, Rejean; Gonzales, Eduardo; Holmes, Ian; Kulchar, James; Leonard, Anthony; Williams, Bob; Taussig, Doug; Thomas, Dan; Marcy, Grant

2017-10-01

An extreme ultraviolet spectrograph measuring resonant emissions of D and C in the lower divertor has been added to DIII-D to help resolve an 2X discrepancy between bolometrically measured radiated power and that predicted by boundary codes for DIII-D, JET and ASDEX-U. With 290 and 450 gr/mm gratings, the DivSPRED spectrometer, an 0.3 m flat-field McPherson model 251, measures ground state transitions for D (the Lyman series) and C (e.g., C IV, 155 nm) which account for >75% of radiated power in the divertor. Combined with Thomson scattering and imaging in the DIII-D divertor, measurements of position, temperature and fractional power emission from plasma components are made and compared to UEDGE/SOLPS-ITER. Mechanical, optical, electrical, vacuum, and shielding aspects of DivSPRED are presented. Work supported under USDOE Cooperative Agreement DE-FC02-04ER54698 and DE-AC52-07NA27344, and by the LLNL Laboratory Directed R&D Program, project #17-ERD-020.

Plans of LHD divertor experiment

International Nuclear Information System (INIS)

Ohyabu, Nobuyoshi; Komori, Akio; Sagara, Akio; Noda, Nobuaki; Motojima, Osamu

1996-01-01

Scenarios of the LHD divertor experiment are presented. In the LHD divertor experimental program, various innovative divertor concepts and technologies, developed during its design phase will be utilized to improve the plasma performance. Two divertor operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement) are among them. Local island divertor geometry has also been proposed. This new divertor has been successfully tested in the CHS device and is planned to be installed in the LHD device. In addition, technological development of new efficient hydrogen pumping schemes (carbon sheet pump and membrane pump) are being pursued for enhancement of the divertor control capability. (author)

Physics aspects of the dynamic ergodic divertor (DED)

International Nuclear Information System (INIS)

Finken, Karl H.; Kobayashi, Masahiro; Abdullaev, Sadrilla S.; Jakubowski, Marcin

2003-01-01

The Dynamic Ergonic Divertor (DED) is presently being installed in the TEXTOR tokamak. It consists of 16 helical coils wound helically around the torus at the high field side (HFS). The perturbation currents in these coils generate predominantly islands of m=10...14 and n=4 leading both to rather closed ergodic and to open laminar structures. In the 'laminar mode', the DED forms a helical divertor. 3D modelling (2D finite element/1 D finite volume) of the plasma transport in the laminar zone has started. By the 'dynamic' operation of the DED, the heat is deposited to a wide area and forces are transferred from the currents in the DED-coils to the plasma edge. (author)

Active control of divertor heat and particle fluxes in EAST towards advanced steady state operations

Energy Technology Data Exchange (ETDEWEB)

Wang, L., E-mail: lwang@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Dalian University of Technology, Dalian 116024 (China); Guo, H.Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); General Atomics, P. O. Box 85608, San Diego, CA 92186 (United States); Li, J.; Wan, B.N.; Gong, X.Z.; Zhang, X.D.; Hu, J.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Liang, Y. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Association EURATOM-FZJ, D-52425 Jülich (Germany); Xu, G.S. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Zou, X.L. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); Loarte, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul Lez Durance (France); Maingi, R.; Menard, J.E. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Luo, G.N.; Gao, X.; Hu, L.Q.; Gan, K.F.; Liu, S.C.; Wang, H.Q.; Chen, R. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); and others

2015-08-15

Significant progress has been made in EAST towards advanced steady state operations by active control of divertor heat and particle fluxes. Many innovative techniques have been developed to mitigate transient ELM and stationary heat fluxes on the divertor target plates. It has been found that lower hybrid current drive (LHCD) can lead to edge plasma ergodization, striation of the stationary heat flux and lower ELM transient heat and particle fluxes. With multi-pulse supersonic molecular beam injection (SMBI) to quantitatively regulate the divertor particle flux, the divertor power footprint pattern can be actively modified. H-modes have been extended over 30 s in EAST with the divertor peak heat flux and the target temperature being controlled well below 2 MW/m{sup 2} and 250 °C, respectively, by integrating these new methods, coupled with advanced lithium wall conditioning and internal divertor pumping, along with an edge coherent mode to provide continuous particle and power exhaust.

The investigation of structure, chemical composition, hydrogen isotope trapping and release processes in deposition layers on surfaces exposed to DIII-D divertor plasma

International Nuclear Information System (INIS)

Buzhinskij, O.I.; Opimach, I.V.; Barsuk, V.A.; Arkhipov, I.I.; Whyte, D.; Wampler, W.R.

1998-05-01

The exposure of ATG graphite sample to DIII-D divertor plasma was provided by the DiMES (Divertor Material Evaluation System) mechanism. The graphite sample arranged to receive the parallel heat flux on a small region of the surface was exposed to 600ms of outer strike point plasma. The sample was constructed to collect the eroded material directed downward into a trapping zone onto s Si disk collector. The average heat flux onto the graphite sample during the exposure was about 200W/cm 2 , and the parallel heat flux was about 10 KW/cm 2 . After the exposure the graphite sample and Si collector disk were analyzed using SEM, NRA, RBS, Auger spectroscopy. IR and Raman spectroscopy. The thermal desorption was studied also. The deposited coating on graphite sample is amorphous carbon layer. Just upstream of the high heat flux zone the redeposition layer has a globular structure. The deposition layer on Si disk is composed also from carbon but has a diamond-like structure. The areal density of C and D in the deposited layer on Si disk varied in poloidal and toroidal directions. The maximum D/C areal density ratio is about 0.23, maximum carbon density is about 3.8 x 10 18 cm -2 , maximum D area density is about 3 x 10 17 cm 2 . The thermal desorption spectrum had a peak at 1,250K

Transport in the plasma edge specific connection to the wall in the Tore Supra ergodic divertor experiments

International Nuclear Information System (INIS)

Grosman, A.; Ghendrih, P.; DeMichelis, C.; Monier-Garbet, P.; Vallet, J.C.; Capes, H.; Chatelier, M.; Geraud, A.; Goniche, M.; Grisolia, C.; Guilhem, D.; Harris, G.; Hess, W.; Nguyen, F.; Poutchy, L.; Samain, A.

1992-01-01

The ergodic divertor experiments in TORE SUPRA can be analysed along two main lines. The first one refers to the change of the heat and particle transport in the ergodized zone. This is especially true for the electron heat transport which is enhanced in the edge layer. But other distinctive features give evidence of the importance of the parallel connexion length between the plasma edge and the wall. The field lines, which are stochastic in the major part of the perturbed layer (10-15 cm) are such that, in the outermost layer (3 cm), the connexion topology is regular. This has obvious effects on the particle and power deposition, but also on the plasma parameters, and consequently influences the particle recycling and impurity shielding processes. The TORE SUPRA ergodic divertor experiments are reviewed in this framework

Radiation transport effects in divertor plasmas generated during a tokamak reactor disruption

International Nuclear Information System (INIS)

Peterson, R.R.; MacFarlane, J.J.; Wang, P.

1994-01-01

Vaporization of material from tokamak divertors during disruptions is a critical issue for tokamak reactors from ITER to commercial power plants. Radiation transport from the vaporized material onto the remaining divertor surface plays an important role in the total mass loss to the divertor. Radiation transport in such a vapor is very difficult to calculate in full detail, and this paper quantifies the sensitivity of the divertor mass loss to uncertainties in the radiation transport. Specifically, the paper presents the results of computer simulations of the vaporization of a graphite coated divertor during a tokamak disruption with ITER CDA parameters. The results show that a factor of 100 change in the radiation conductivity changes the mass loss by more than a factor of two

Model of divertor biasing and control of scrape-off layer and divertor plasmas

International Nuclear Information System (INIS)

Nagasaki, K.; Itoh, K.; Itoh, S.

1991-02-01

Analytic model of the divertor biasing is described. For the given plasma and energy sources from the core plasma, the heat and particle flux densities on the divertor plate as well as scrape-off-layer (SOL)/divertor plasmas are analyzed in a slab model. Using a two-dimensional model, the effects of the divertor biasing and SOL current are studied. The conditions to balance the plasma temperature or sheath potential on different divertor plates are obtained. Effect of the SOL current on the heat channel width is also discussed. (author)

A carbon-metal brazing for divertor plates in fusion devices

International Nuclear Information System (INIS)

Matsuda, T.; Matsumoto, T.; Miki, S.; Sogabe, T.; Okada, M.; Kubota, Y.; Sagara, A.; Noda, N.; Motojima, O.; Hino, T.; Yamashina, T.

1993-01-01

A divertor unit, which consists of carbon armors brazed to a copper cooling channel, is under development for fusion devices. Isotropic graphite (IG-430U) and CFC (CX-2002U) are used for the armor, and a copper for the cooling tube. A technique named as dissolution and deposit of base metal was employed for brazing. The reliability of the brazed components was evaluated both by 4-point bending test and thermal shock test. According to the results of a 4-point bending test under the temperature ranged from RT to 800 C in a vacuum, it was found that the strength of the brazed surface at RT was maintained up to the higher temperature, 600 C. High heat load test has been also performed on the brazed sample in order to find whether the samples meet the requirement of the divertor plates of LHD (Large Helical Device). Active Cooling Teststand (ACT:NIFS) with electron beam power of 100kW was used. In LHD, it is presumed that the maximum heat flux is 10MW/m 2 . In addition, the surface temperature of divertor has to be kept below 1,200 C to avoid RES, by active cooling. The heat load test showed that the brazing components of CX-2002U (flat plate type CFC-Cu brazed) was stable at 1,300 C under a heat flux of 10MW/m 2 , when the flow velocity of cooling water was 6m/s. No damage nor deterioration was found at the brazed zone after the heat load test

Conceptual design studies for the European DEMO divertor: Rationale and first results

International Nuclear Information System (INIS)

You, J.H.; Mazzone, G.; Visca, E.; Bachmann, Ch.; Autissier, E.; Barrett, T.; Cocilovo, V.; Crescenzi, F.; Domalapally, P.K.; Dongiovanni, D.; Entler, S.; Federici, G.; Frosi, P.; Fursdon, M.; Greuner, H.; Hancock, D.; Marzullo, D.; McIntosh, S.; Müller, A.V.; Porfiri, M.T.

2016-01-01

Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

Conceptual design studies for the European DEMO divertor: Rationale and first results

Energy Technology Data Exchange (ETDEWEB)

You, J.H., E-mail: you@ipp.mpg.de [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Mazzone, G.; Visca, E. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Bachmann, Ch. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Autissier, E. [CEA, IRFM, F-13108 Saint Paul Lez Durance (France); Barrett, T. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Cocilovo, V.; Crescenzi, F. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Domalapally, P.K. [Research Cnter Rez, Hlavní 130, 250 68 Husinec–Řež (Czech Republic); Dongiovanni, D. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Entler, S. [Institute of Plasma Physics CAS, Za Slovankou 3, 182 00 Praha 8 (Czech Republic); Federici, G. [EUROfusion PMU, c/o IPP, Boltzmann Str. 2, 85748 Garching (Germany); Frosi, P. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); Fursdon, M. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Greuner, H. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Hancock, D. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Marzullo, D. [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); McIntosh, S. [CCFE, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Müller, A.V. [Max Planck Institute for Plasma Physics, Boltzmann Str. 2, 85748 Garching (Germany); Porfiri, M.T. [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Italy); and others

2016-11-01

Highlights: • A brief overview is given on the overall R&D activities of the work package Divertor which is a project of the EUROfusion Consortium. • The rationale of the hydraulic, thermal and structural design scheme is described. • The first results obtained for the preliminary DEMO divertor cassette model are presented. - Abstract: In the European fusion roadmap, reliable power handling has been defined as one of the most critical challenges for realizing a commercially viable fusion power. In this context, the divertor is the key in-vessel component, as it is responsible for power exhaust and impurity removal for which divertor target is subjected to very high heat flux loads. To this end, an integrated R&D project was launched in the EUROfusion Consortium in order to deliver a holistic conceptual design solution together with the core technologies for the entire divertor system of a DEMO reactor. The work package ‘Divertor’ consists of two project areas: ‘Cassette design and integration’ and ‘Target development’. The essential mission of the project is to develop and verify advanced design concepts and the required technologies for a divertor system being capable of meeting the physical and system requirements defined for the next-generation European DEMO reactor. In this contribution, a brief overview is presented of the works from the first project year (2014). Focus is put on the loads specification, design boundary conditions, materials requirements, design approaches, and R&D strategy. Initial ideas and first estimates are presented.

Radiative divertor plasmas with convection in DIII-D

International Nuclear Information System (INIS)

Leornard, A.W.; Porter, G.D.; Wood, R.D.

1998-01-01

The radiation of divertor heat flux on DIII-D is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE has reproduced many of the observed experimental features

Boundary plasma control with the ergodic divertor

International Nuclear Information System (INIS)

Ghendrih, Ph.; Becoulet, M.; Beyer, P.

1999-01-01

Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

Boundary plasma control with the ergodic divertor

International Nuclear Information System (INIS)

Ghendrih, Ph.; Becoulet, M.; Beyer, P.

2001-01-01

Ergodic divertor experiments on Tore Supra provide evidence of significant control of plasma-wall interaction. Theoretical investigation of the laminar region (i.e. governed by parallel transport) indicates that control of the plasma state at the target plate can be achieved with plasma states similar to that observed with the axisymmetric divertor. Analysis of the temperature field with a 2-D test particle code allows one to recover the observed spatial modulation and shows that an intrinsic barrier appears to develop at the separatrix. Energy deposition peaking, analysed with a 3-D code, is strongly reduced when moderate transverse transport is considered. Possible control of upstream parameters can thus be achieved in the ergodic region, for instance a lowering of the parallel energy flux by cross field transport. (author)

The DIII-D Radiative Divertor Project: Status and plans

International Nuclear Information System (INIS)

Smith, J.P.; Baxi, C.B.; Bozek, A.S.

1996-10-01

New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the second and final phase, in 1998. The phased approach enables the continuation of the divertor characterization research in the lower divertor while providing pumping for density control in high triangularity, single- or double-null advanced tokamak discharges. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. By puffing neutral gas into the divertor region, a reduction in the heat flux on the target plates will be be demonstrated without a large rise in core density. This reduction in heat flux is accomplished by dispersing the power with radiation in the divertor region. Experiments and modeling have formed the basis for the new design. The capability of the DIII-D cryogenic system is being upgraded as part of this project. The increased capability of the cryogenic system will allow delivery of liquid helium and nitrogen to three new cryopumps. Physics studies on the effects of slot width and length can be accomplished easily with the design of the Radiative Divertor. The slot width can be varied by installing graphite tiles of different geometry. The change in slot length, the distance from the X-point to the target plate, requires relocating the structure vertically and can be completed in about 6-8 weeks. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Required diagnostic modifications will be minimal for Phase 1, but extensive for Phase 2 installation. These Phase 2 diagnostics will be required to fully diagnose the high triangularity discharges in the divertor slots

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

International Nuclear Information System (INIS)

Matveev, D; Kirschner, A; Litnovsky, A; Borodin, D; Samm, U; Schmid, K; Komm, M; Van Oost, G

2014-01-01

An estimation of the contribution of gaps to beryllium deposition and resulting tritium retention in the divertor of ITER is presented. Deposition of beryllium layers in gaps of the full tungsten divertor is simulated with the 3D-GAPS code. For gaps aligned along the poloidal direction, non-shaped and shaped solutions are compared. Plasma and impurity ion fluxes from Schmid (2008 Nucl. Fusion 48 105004) are used as input. Ion penetration into gaps is considered to be geometrical along magnetic field lines. The effect of realistic ion penetration into gaps is discussed. In total, gaps in the divertor are estimated to contribute about 0.3 mgT s −1 to the overall tritium retention dominated by toroidal gaps, which are not shaped. This amount corresponds to about 7800 ITER discharges up to the safety limit of 1 kg in-vessel tritium; excluding, however, tritium release during wall baking and retention at plasma-wetted and remote areas. (paper)

Latest status of manufacturing activity of ITER divertor and engineering issues on tungsten divertor

International Nuclear Information System (INIS)

Suzuki, Satoshi

2011-01-01

Divertors for ITER are now in construction. In the present chapter, the specification and the latest status of manufacturing of ITER divertors are presented. In addition, issues in the development of divertors for the fusion demo reactor are given on the basis of experiences on the ITER divertor development. (J.P.N.)

Energy transport to the divertor plates of ASDEX-Upgrade during ELMy H-mode phases

International Nuclear Information System (INIS)

Herrmann, A.; Laux, M.; Coster, D.; Neuhauser, J.; Reiter, D.; Schneider, R.; Weinlich, M.

1995-01-01

The energy flux to the ASDEX-Upgrade divertor plates is routinely measured by themography and Langmuir probes. The thermographically observed power decay length at the target plate is about 1 cm near the inboard separatrix. During an edge localized mode (ELM) of type I the density profiles are significantly, changed; an additional contribution occurs characterized by a power decay length in the order of 10 cm outside the separatrix and additional power is deposited into the private flux region. It is supposed that this is due to the changing, contribution of energy conduction versus convection. Results of ELM-modelling using the coupled B2-EIRENE code reproduce the main features of the experimental observations. The sheath transmission factor is calculated by combining themography and Langmuir probe data. ((orig.))

Proposal of an alternative upper divertor in ASDEX Upgrade supported by EMC3-EIRENE simulations

Directory of Open Access Journals (Sweden)

T. Lunt

2017-08-01

Full Text Available We discuss the benefits of installing a pair of in-vessel coils with currents |Ifx| ≲ 50 kAt in the upper divertor of ASDEX Upgrade (AUG to study a series of ‘alternative’ divertor configurations, like the Snowflake (SF and the X-divertor (XD, that are currently considered as alternative solutions for the power exhaust problem. The possibility of operating the standard lower single-null (SN and double-null (DN would be preserved. Potential effects to reduce the peak parallel- and/or perpendicular heat flux are predicted from a simple geometrical-diffusive model as well as by numerical EMC3-EIRENE simulations for pure deuterium attached conditions with spatially constant diffusion coefficients. Beyond that a series of other potential transport- and radiation related heat flux mitigation effects are identified and could be studied experimentally with the modified upper divertor in the high-power divertor Tokamak AUG.

Two-dimensional numerical study of ELMs-induced erosion of tungsten divertor target tiles with different edge shapes

International Nuclear Information System (INIS)

Huang, Yan; Sun, Jizhong; Hu, Wanpeng; Sang, Chaofeng; Wang, Dezhen

2016-01-01

Highlights: • Thermal performance of three edge-shaped divertor tiles was assessed numerically. • All the divertor tiles exposed to type-I ELMs like ITER's will melt. • The rounded edge tile thermally performs the best in all tiles of interest. • The incident energy flux density was evaluated with structural effects considered. - Abstract: Thermal performance of the divertor tile with different edge shapes was assessed numerically along the poloidal direction by a two-dimensional heat conduction model with considering the geometrical effects of castellated divertor tiles on the properties of its adjacent plasma. The energy flux density distribution arriving at the castellated divertor tile surface was evaluated by a two-dimension-in-space and three-dimension-in-velocity particle-in-cell plus Monte Carlo Collisions code and then the obtained energy flux distribution was used as input for the heat conduction model. The simulation results showed that the divertor tiles with any edge shape of interest (rectangular edge, slanted edge, and rounded edge) would melt, especially, in the edge surface region of facing plasma poloidally under typical heat flux density of a transient event of type-I ELMs for ITER, deposition energy of 1 MJ/m"2 in a duration of 600 μs. In comparison with uniform energy deposition, the vaporizing erosion was reduced greatly but the melting erosion was aggravated noticeably in the edge area of plasma facing diveror tile. Of three studied edge shapes, the simulation results indicated that the divertor plate with rounded edge was the most resistant to the thermal erosion.

Results of the H-mode experiments with JT-60 outer and lower divertors

International Nuclear Information System (INIS)

Nakamura, Hiroo; Tsuji, Shunji; Nagami, Masayuki

1989-08-01

In JT-60, hydrogen H-mode experiments with outer and lower divertors were performed. In the outer divertor, H-mode were obtained, similar to the ones observed in the other lower/upper divertors. Its threshold absorbed power and electron density were 16 MW and 1.8 x 10 19 m -3 . In the two combined heatings with NB+ICRF and NB+LHRF, H-mode discharges are also obtained. Moreover, in new configuration of lower divertor, H-mode phases without and with ELM are obtained. Typical results of the lower divertor are shown to compare the H-mode characteristics between the two configurations. Improvement of the energy confinement time in the two divertors was limited to 10 %. Analyses on ballooning/interchange instabilities were carried out with precise equlibria of JT-60. These results showed that the both modes were enough stable. (author)

Comparison of Ne and Ar seeded radiative divertor plasmas in JT-60U

Energy Technology Data Exchange (ETDEWEB)

Nakano, T., E-mail: nakano.tomohide@jaea.go.jp

2015-08-15

In H-mode plasmas with Ne, Ar and a mixture of Ne and Ar injection, the divertor radiation power fractions amongst these impurities in addition to an intrinsic impurity, C, are investigated. In plasmas with the inner divertor plasma attached, carbon is the biggest radiator, whichever impurity, Ne, Ar or a mixture of Ar and Ne is injected. In contrast, in plasmas with the inner divertor plasma detached, Ne is the biggest radiator due to a significantly high recombination radiation from Ne VIII. Ar is always a minor contributor in plasmas with the inner divertor both attached and detached.

A new scaling for divertor detachment

Science.gov (United States)

Goldston, R. J.; Reinke, M. L.; Schwartz, J. A.

2017-05-01

The ITER design, and future reactor designs, depend on divertor ‘detachment,’ whether partial, pronounced or complete, to limit heat flux to plasma-facing components and to limit surface erosion due to sputtering. It would be valuable to have a measure of the difficulty of achieving detachment as a function of machine parameters, such as input power, magnetic field, major radius, etc. Frequently the parallel heat flux, estimated typically as proportional to P sep/R or P sep B/R, is used as a proxy for this difficulty. Here we argue that impurity cooling is dependent on the upstream density, which itself must be limited by a Greenwald-like scaling. Taking this into account self-consistently, we find the impurity fraction required for detachment scales dominantly as power divided by poloidal magnetic field. The absence of any explicit scaling with machine size is concerning, as P sep surely must increase greatly for an economic fusion system, while increases in the poloidal field strength are limited by coil technology and plasma physics. This result should be challenged by comparison with 2D divertor codes and with measurements on existing experiments. Nonetheless, it suggests that higher magnetic field, stronger shaping, double-null operation, ‘advanced’ divertor configurations, as well as alternate means to handle heat flux such as metallic liquid and/or vapor targets merit greater attention.

Plasma shape control calculations for BPX divertor design

International Nuclear Information System (INIS)

Strickler, D.J.; Neilson, G.H.; Jardin, S.C.; Pomphrey, N.

1991-01-01

The Burning Plasma Experiment (BPX) divertor is to be capable of withstanding heat loads corresponding to ignited operation and 500 MW of fusion power for a current rise time and flattop lasting several seconds. The poloidal field (PF), diagnostic, and feedback equilibrium control systems must provide precise X-point position control in order to sweep the separatrices across the divertor target surface and optimally distribute the heat loads. A control matrix MHD equilibrium code, BEQ, and the Tokamak Simulation Code (TSC) are used to compute preprogrammed double-null (DN) divertor sweep trajectories that maximize sweep distance while simultaneously satisfying a set of strict constraints: minimum lengths of the field lines between the X-point and strike points, minimum spacing between the inboard plasma edge and the limiter, maximum spacing between the outboard plasma edge and the ICRF antennas, minimum safety factor, and linked poloidal flux. A sequence of DN diverted equilibria and a consistent TSC fiducial discharge simulation are used in evaluating the performance of the BPX divertor shape and possible modifications. 5 refs., 10 figs

Optimization for steady-state and hybrid operations of ITER by using scaling models of divertor heat load

International Nuclear Information System (INIS)

Murakami, Yoshiki; Itami, Kiyoshi; Sugihara, Masayoshi; Fujieda, Hirobumi.

1992-09-01

Steady-state and hybrid mode operations of ITER are investigated by 0-D power balance calculations assuming no radiation and charge-exchange cooling in divertor region. Operation points are optimized with respect to divertor heat load which must be reduced to the level of ignition mode (∼5 MW/m 2 ). Dependence of the divertor heat load on the variety of the models, i.e., constant-χ model, Bohm-type-χ model and JT-60U empirical scaling model, is also discussed. The divertor heat load increases linearly with the fusion power (P FUS ) in all models. The possible highest fusion power much differs for each model with an allowable divertor heat load. The heat load evaluated by constant-χ model is, for example, about 1.8 times larger than that by Bohm-type-χ model at P FUS = 750 MW. Effect of reduction of the helium accumulation, improvements of the confinement capability and the current-drive efficiency are also investigated aiming at lowering the divertor heat load. It is found that NBI power should be larger than about 60 MW to obtain a burn time longer than 2000 s. The optimized operation point, where the minimum divertor heat load is achieved, does not depend on the model and is the point with the minimum-P FUS and the maximum-P NBI . When P FUS = 690 MW and P NBI = 110 MW, the divertor heat load can be reduced to the level of ignition mode without impurity seeding if H = 2.2 is achieved. Controllability of the current-profile is also discussed. (J.P.N.)

Discharge power dependence of carbon dust flux in a divertor simulator

International Nuclear Information System (INIS)

Nishiyama, Katsushi; Morita, Yasuhiko; Uchida, Giichiro; Yamashita, Daisuke; Kamataki, Kunihiro; Seo, Hyunwoong; Itagaki, Naho; Koga, Kazunori; Shiratani, Masaharu; Ashikawa, Naoko; Masuzaki, Suguru; Nishimura, Kiyohiko; Sagara, Akio; Bornholdt, Sven; Kersten, Holger

2013-01-01

In fusion devices, dust particles are generated due to plasma–wall interactions and may cause safety or operational problems. Therefore it is necessary to clarify the generation and transport mechanisms of dust particles. Here we have measured energy influx from H 2 plasmas toward a graphite target using a calorimetric probe and compared the results with the dust flux toward a dust collecting substrate set on the reactor wall. The dust flux decreases with increasing the energy influx. For the higher discharge power, the more number of dust particles tend to redeposit onto the graphite target due to the higher ion drag force and hence the dust flux toward the reactor wall becomes smaller. The results show that dust inventory depends strongly on energy influx to graphite divertor plates in fusion devices

Design, R&D and commissioning of EAST tungsten divertor

Science.gov (United States)

Yao, D. M.; Luo, G. N.; Zhou, Z. B.; Cao, L.; Li, Q.; Wang, W. J.; Li, L.; Qin, S. G.; Shi, Y. L.; Liu, G. H.; Li, J. G.

2016-02-01

After commissioning in 2005, the EAST superconducting tokamak had been operated with its water cooled divertors for eight campaigns up to 2012, employing graphite as plasma facing material. With increase in heating power over 20 MW in recent years, the heat flux going to the divertors rises rapidly over 10 MW m-2 for steady state operation. To accommodate the rapid increasing heat load in EAST, the bolting graphite tile divertor must be upgraded. An ITER-like tungsten (W) divertor has been designed and developed; and firstly used for the upper divertor of EAST. The EAST upper W divertor is modular structure with 80 modules in total. Eighty sets of W/Cu plasma-facing components (PFC) with each set consisting of an outer vertical target (OVT), an inner vertical target (IVT) and a DOME, are attached to 80 stainless steel cassette bodies (CB) by pins. The monoblock W/Cu-PFCs have been developed for the strike points of both OVT and IVT, and the flat type W/Cu-PFCs for the DOME and the baffle parts of both OVT and IVT, employing so-called hot isostatic pressing (HIP) technology for tungsten to CuCrZr heat sink bonding, and electron beam welding for CuCrZr to CuCrZr and CuCrZr to other material bonding. Both monoblock and flat type PFC mockups passed high heat flux (HHF) testing by means of electron beam facilities. The 80 divertor modules were installed in EAST in 2014 and results of the first commissioning are presented in this paper.

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

Science.gov (United States)

Frerichs, H.; Schmitz, O.; Covele, B.; Feng, Y.; Guo, H. Y.; Hill, D.

2018-05-01

Numerical simulations of toroidal asymmetries in a tightly baffled small angle slot (SAS) divertor on the DIII-D tokamak show that toroidal asymmetries in divertor closure result in (non-axisymmetric) local onset of detachment within a density window of 10-15% on top of the nominal threshold separatrix density. The SAS divertor is explored at DIII-D for improving access to cold, dissipative/detached divertor conditions. The narrow width of the slot divertor coupled with a small magnetic field line-to-target angle facilitates the buildup of neutral density, thereby increasing radiative and neutrals-related (atoms and molecules) losses in the divertor. Small changes in the strike point location can be expected to have a large impact on divertor conditions. The combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field configuration causes the strike point to move along the divertor target plate, possibly leaving the divertor slot at some locations. The latter extreme case essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade the performance of the slot divertor. Such a strike point dislocation is approximated by a finite gap in the divertor baffle for which 3D edge plasma and neutral gas simulations are performed with the EMC3-EIRENE code.

Models for poloidal divertors

International Nuclear Information System (INIS)

Post, D.E.; Heifetz, D.; Petravic, M.

1982-07-01

Recent progress in models for poloidal divertors has both helped to explain current divertor experiments and contributed significantly to design efforts for future large tokamak (INTOR, etc.) divertor systems. These models range in sophistication from zero-dimensional treatments and dimensional analysis to two-dimensional models for plasma and neutral particle transport which include a wide variety of atomic and molecular processes as well as detailed treatments of the plasma-wall interaction. This paper presents a brief review of some of these models, describing the physics and approximations involved in each model. We discuss the wide variety of physics necessary for a comprehensive description of poloidal divertors. To illustrate the progress in models for poloidal divertors, we discuss some of our recent work as typical examples of the kinds of calculations being done

Models for poloidal divertors

Energy Technology Data Exchange (ETDEWEB)

Post, D.E.; Heifetz, D.; Petravic, M.

1982-07-01

Recent progress in models for poloidal divertors has both helped to explain current divertor experiments and contributed significantly to design efforts for future large tokamak (INTOR, etc.) divertor systems. These models range in sophistication from zero-dimensional treatments and dimensional analysis to two-dimensional models for plasma and neutral particle transport which include a wide variety of atomic and molecular processes as well as detailed treatments of the plasma-wall interaction. This paper presents a brief review of some of these models, describing the physics and approximations involved in each model. We discuss the wide variety of physics necessary for a comprehensive description of poloidal divertors. To illustrate the progress in models for poloidal divertors, we discuss some of our recent work as typical examples of the kinds of calculations being done.

Design, fabrication, and testing of a helium-cooled module for the ITER divertor

International Nuclear Information System (INIS)

Baxi, C.B.; Smith, J.P.; Youchison, D.

1994-08-01

The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m 2 . The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m 2 applied over a surface area of 20 cm 2 . The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of this effort we conclude that helium cooling of the ITER divertor is feasible without requiring a very large helium pressure or a large pumping power

Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

Energy Technology Data Exchange (ETDEWEB)

Kugel, W.; Bell, M.; Berzak,L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren,R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov L. E.

2008-09-26

Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW~1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with longpulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

International Nuclear Information System (INIS)

Kugel, W.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.; Harjes, H.; Kaita, R.; Kallman, J.; Maingi, R.; Majeski, R.; Mansfield, D.; Menard, J.; Nygren, R. E.; Soukhanovskii, V.; Stotler, D.; Wakeland, P.; Zakharov, L. E.

2008-01-01

Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW∼1), to enable ne scan capability (x2) in the H-mode, to test the ability to operate at significantly lower density for future ST-CTF reactor designs (e.g., ne/nGW = 0.25), and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization

Detached divertor plasmas in Alcator C-Mod: A study of the role of atomic physics

International Nuclear Information System (INIS)

Lipschultz, B.; Boswell, C.; Goetz, J.A.

1999-01-01

Detailed profiles of the volumetric recombination occurring in Alcator C-Mod plasmas are presented. During detachment the recombination sink is compared to the divertor plate sink as well as the divertor ion source. Depending on plasma conditions, volume recombination removes between 10 and 75% of the ions before they reach the plates. A second, equally important process that leads to a drop in plate ion current is inferred to be a reduction in divertor ion source, which is correlated with a drop in power flowing into the ionization region and the pressure loss of detachment. For high n e the divertor recombination can cross the separatrix near the x-point, cool the core and lead to a disruption. Experimental measurements show a difference in ion and neutral velocities for H-mode detached plasmas. The resulting ion-neutral collisions are found to be more efficacious than recombination in removing momentum from the ions. The neutral component of volumetric power emission from the divertor has been measured by means of a novel filtering technique to be substantial (∼ 20% of the total divertor volumetric emission). (author)

Analysis of heat transfer and erosion effects on ITER divertor plasma facing components induced by slow high-power transients

International Nuclear Information System (INIS)

Federici, G.; Raffray, A.R.; Chiocchio, S.; Esser, B.; Dietz, J.; Igitkhanov, Y.; Janeschitz, G.

1995-01-01

This paper presents the results of an analysis carried out to investigate the thermal response of ITER divertor plasma facing components (PFC's) clad with Be, W, and CFC, to high-recycling, high-power thermal transients (i.e. 10--30 MW/m 2 ) which are anticipated to last up to a few seconds. The armour erosion and surface melting are estimated for the different plasma facing materials (PFM's) together with the maximum heat flux to the coolant, and armour/heat-sink interface temperature. The analysis assumes that intense target evaporation will lead to high radiative power losses in the plasma in front of the target which self-protects the target. The cases analyzed clarify the influence of several key parameters such as the plasma heat flux to the target, the loss of the melt layer, the duration of the event, the thickness of the armour, and comparison is made with cases without vapor shielding. Finally, some implications for the performance and lifetime of divertor PFC's clad with different PFM's are discussed

Spectroscopic diagnostics for liquid lithium divertor studies on National Spherical Torus Experiment

International Nuclear Information System (INIS)

Soukhanovskii, V. A.; Roquemore, A. L.; Bell, R. E.; Kaita, R.; Kugel, H. W.

2010-01-01

The use of lithium-coated plasma facing components for plasma density control is studied in the National Spherical Torus Experiment (NSTX). A recently installed liquid lithium divertor (LLD) module has a porous molybdenum surface, separated by a stainless steel liner from a heated copper substrate. Lithium is deposited on the LLD from two evaporators. Two new spectroscopic diagnostics are installed to study the plasma surface interactions on the LLD: (1) A 20-element absolute extreme ultraviolet (AXUV) diode array with a 6 nm bandpass filter centered at 121.6 nm (the Lyman-α transition) for spatially resolved divertor recycling rate measurements in the highly reflective LLD environment, and (2) an ultraviolet-visible-near infrared R=0.67 m imaging Czerny-Turner spectrometer for spatially resolved divertor D I, Li I-II, C I-IV, Mo I, D 2 , LiD, CD emission and ion temperature on and around the LLD module. The use of photometrically calibrated measurements together with atomic physics factors enables studies of recycling and impurity particle fluxes as functions of LLD temperature, ion flux, and divertor geometry.

Innovative divertor concepts for LHD

International Nuclear Information System (INIS)

Ohyabu, N.; Komori, A.; Akaishi, K.

1994-07-01

We are developing various innovative divertor concepts which improve the LHD plasma performance. These are two divertor magnetic geometries (helical and local island divertors), three operational scenarios (radiative cooling in the high density, cold boundary, confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode like confinement improvement) and technological development of new efficient hydrogen pumping schemes. (author)

He-cooled divertor development for DEMO

International Nuclear Information System (INIS)

Norajitra, P.; Giniyatulin, R.; Ihli, T.; Janeschitz, G.; Krauss, W.; Kruessmann, R.; Kuznetsov, V.; Mazul, I.; Widak, V.; Ovchinnikov, I.; Ruprecht, R.; Zeep, B.

2007-01-01

Goal of the He-cooled divertor development for future fusion power plants is to resist a high heat flux of at least 10 MW/m 2 . The development includes the fields of design, analyses, and experiments. A helium-cooled modular jet concept (HEMJ) has been defined as reference solution, which is based on jet impingement cooling. In cooperation with the Efremov Institute, work was aimed at construction and high heat flux tests of prototypical tungsten mockups to demonstrate their manufacturability and their performances. A helium loop was built for this purpose to simulate the realistic thermo-hydraulics conditions close to those of DEMO (10 MPa He, 600 deg. C). The first high heat flux test results confirm the feasibility and the performance of the divertor design

ALPS - advanced limiter-divertor plasma-facing systems

International Nuclear Information System (INIS)

Allain, J. P.; Bastasz, R.; Brooks, J. N.; Evans, T.; Hassanein, A.; Luckhardt, S.; Maingi, R.; Mattas, R. F.; McCarthy, K.; Mioduszewski, P.; Mogahed, E.; Moir, R.; Molokov, S.; Morely, N.; Nygren, R.; Reed, C.; Rognlien, T.; Ruzic, D.; Sviatoslavsky, I.; Sze, D.; Tillack, M.; Ulrickson, M.; Wade, P. M.; Wong, C.; Wooley, R.

1999-01-01

The Advanced Limiter-divertor Plasma-facing Systems (ALPS) program was initiated in order to evaluate the potential for improved performance and lifetime for plasma-facing systems. The main goal of the program is to demonstrate the advantages of advanced limiter/divertor systems over conventional systems in terms of power density capability, component lifetime, and power conversion efficiency, while providing for safe operation and minimizing impurity concerns for the plasma. Most of the work to date has been applied to free surface liquids. A multi-disciplinary team from several institutions has been organized to address the key issues associated with these systems. The main performance goals for advanced limiters and diverters are a peak heat flux of >50 MW/m 2 ,elimination of a lifetime limit for erosion, and the ability to extract useful heat at high power conversion efficiency (approximately40%). The evaluation of various options is being conducted through a combination of laboratory experiments, modeling of key processes, and conceptual design studies. The current emphasis for the work is on the effects of free surface liquids on plasma edge performance

T-12 divertor experiment

Energy Technology Data Exchange (ETDEWEB)

Bortnikov, A V; Brevnov, N N; Gerasimov, S N; Zhukovskii, V G; Kuznetsov, N V; Naftulin, S M; Pergament, V I; Khimchenko, L N [Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow. Inst. Atomnoj Ehnergii

1981-01-01

In designing tokamak devices and reactors, in the last few years, the use of elongated-cross-section plasma discharges has been proposed to improve the economic and physical parameters. Application of a quadrupole poloidal magnetic field necessary for sustaining the elongated discharge cross-section serves, in this case, to create the magnetic configuration of an axisymmetric poloidal divertor. To-day, the creation of such a combination, including an elongated plasma cross-section and a divertor and using the outer poloidal magnetic field coils, seems to be the most reasonable approach, from the point of view of design and technology. Such a divertor was produced and studied at the T-12 tokamak. A stable equilibrium configuration of a finger-ring tokamak with a divertor has been produced by superposing the magnetic fields of the plasma current, the external quadrupole coils and the copper shell currents; the reactor blanket can fulfil the function of the latter. It is shown that both a symmetric magnetic configuration with two divertors and a droplet configuration with a single divertor may be realized by controlling the plasma column position with respect to the equatorial plane. The stability of the plasma column against vertical displacement depends on this position and the distance between the separatrix points. Vertical instability stabilization has been observed. The divertor layer efficiently screens the plasma from the impurity influx from the wall and unloads the wall from particle and energy fluxes. The results obtained from the tokamak T-12 experiment have demonstrated the capability of a system with outer poloidal field coils and a copper shell providing an elongated-cross-section plasma column with poloidal divertors.

Versator divertor experiment: preliminary designs

International Nuclear Information System (INIS)

Wan, A.S.; Yang, T.F.

1984-08-01

The emergence of magnetic divertors as an impurity control and ash removal mechanism for future tokamak reactors bring on the need for further experimental verification of the divertor merits and their ability to operate at reactor relevant conditions, such as with auxiliary heating. This paper presents preliminary designs of a bundle and a poloidal divertor for Versator II, which can operate in conjunction with the existing 150 kW of LHRF heating or LH current drive. The bundle divertor option also features a new divertor configuration which should improve the engineering and physics results of the DITE experiment. Further design optimization in both physics and engineering designs are currently under way

Snowflake divertor configuration studies for NSTX-Upgrade

International Nuclear Information System (INIS)

Soukhanovskii, V.A.

2011-01-01

Snowflake divertor experiments in NSTX provide basis for PMI development toward NSTX-Upgrade. Snowflake configuration formation was followed by radiative detachment. Significant reduction of steady-state divertor heat flux observed in snowflake divertor. Impulsive heat loads due to Type I ELMs are partially mitigated in snowflake divertor. Magnetic control of snowflake divertor configuration is being developed. Plasma material interface development is critical for NSTX-U success. Four divertor coils should enable flexibility in boundary shaping and control in NSTX-U. Snowflake divertor experiments in NSTX provide good basis for PMI development in NSTX-Upgrade. FY 2009-2010 snowflake divertor experiments in NSTX: (1) Helped understand control of magnetic properties; (2) Core H-mode confinement unchanged; (3) Core and edge carbon concentration reduced; and (4) Divertor heat flux significantly reduced - (a) Steady-state reduction due to geometry and radiative detachment, (b) Encouraging results for transient heat flux handling, (c) Combined with impurity-seeded radiative divertor. Outlook for snowflake divertor in NSTX-Upgrade: (1) 2D fluid modeling of snowflake divertor properties scaling - (a) Edge and divertor transport, radiation, detachment threshold, (b) Compatibility with cryo-pump and lithium conditioning; (2) Magnetic control development; and (3) PFC development - PFC alignment and PFC material choice.

Divertor cooling device

International Nuclear Information System (INIS)

Nakayama, Tadakazu; Hayashi, Katsumi; Handa, Hiroyuki

1993-01-01

Cooling water for a divertor cooling system cools the divertor, thereafter, passes through pipelines connecting the exit pipelines of the divertor cooling system and the inlet pipelines of a blanket cooling system and is introduced to the blanket cooling system in a vacuum vessel. It undergoes emission of neutrons, and cooling water in the divertor cooling system containing a great amount of N-16 which is generated by radioactivation of O-16 is introduced to the blanket cooling system in the vacuum vessel by way of pipelines, and after cooling, passes through exit pipelines of the blanket cooling system and is introduced to the outside of the vacuum vessel. Radiation of N-16 in the cooling water is decayed sufficiently with passage of time during cooling of the blanket, thereby enabling to decrease the amount of shielding materials such as facilities and pipelines, and ensure spaces. (N.H.)

Chemically deposited tungsten fibre-reinforced tungsten – The way to a mock-up for divertor applications

Directory of Open Access Journals (Sweden)

J. Riesch

2016-12-01

Full Text Available The development of advanced materials is essential for sophisticated energy systems like a future fusion reactor. Tungsten fibre-reinforced tungsten composites (Wf/W utilize extrinsic toughening mechanisms and therefore overcome the intrinsic brittleness of tungsten at low temperature and its sensitivity to operational embrittlement. This material has been successfully produced and tested during the last years and the focus is now put on the technological realisation for the use in plasma facing components of fusion devices. In this contribution, we present a way to utilize Wf/W composites for divertor applications by a fabrication route based on the chemical vapour deposition (CVD of tungsten. Mock-ups based on the ITER typical design can be realized by the implementation of Wf/W tiles. A concept based on a layered deposition approach allows the production of such tiles in the required geometry. One fibre layer after the other is positioned and ingrown into the W-matrix until the final sample size is reached. Charpy impact tests on these samples showed an increased fracture energy mainly due to the ductile deformation of the tungsten fibres. The use of Wf/W could broaden the operation temperature window of tungsten significantly and mitigate problems of deep cracking occurring typically in cyclic high heat flux loading. Textile techniques are utilized to optimise the tungsten wire positioning and process speed of preform production. A new device dedicated to the chemical deposition of W enhances significantly, the available machine time for processing and optimisation. Modelling shows that good deposition results are achievable by the use of a convectional flow and a directed temperature profile in an infiltration process.

Model-based radiation scalings for the ITER-like divertors of JET and ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Aho-Mantila, L., E-mail: leena.aho-mantila@vtt.fi [VTT Technical Research Centre of Finland, FI-02044 VTT (Finland); Bonnin, X. [LSPM – CNRS, Université Paris 13, Sorbonne Paris Cité, F-93430 Villetaneuse (France); Coster, D.P. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Lowry, C. [EFDA JET CSU, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Wischmeier, M. [Max-Planck Institut für Plasmaphysik, D-85748 Garching (Germany); Brezinsek, S. [Forschungszentrum Jülich, Institut für Energie- und Klimaforschung Plasmaphysik, 52425 Jülich (Germany); Federici, G. [EFDA PPP& T Department, D-85748 Garching (Germany)

2015-08-15

Effects of N-seeding in L-mode experiments in ASDEX Upgrade and JET are analysed numerically with the SOLPS5.0 code package. The modelling yields 3 qualitatively different radiative regimes with increasing N concentration, when initially attached outer divertor conditions are studied. The radiation pattern is observed to evolve asymmetrically, with radiation increasing first in the inner divertor, then in the outer divertor, and finally on closed field lines above the X-point. The properties of these radiative regimes are observed to be sensitive to cross-field drifts and they differ between the two devices. The modelled scaling of the divertor radiated power with the divertor neutral pressure is similar to an experimental scaling law for H-mode radiation. The same parametric dependencies are not observed in simulations without drifts.

Energy and particle transport in the radiative divertor plasmas of DIII-D

International Nuclear Information System (INIS)

Leonard, A.W.; Allen, S.L.; Brooks, N.H.

1997-06-01

It has been argued that divertor energy transport dominated by parallel electron thermal conduction, or q parallel = -kT 5/2 2 dT e /ds parallel, leads to severe localization of the intense radiating region and ultimately limits the fraction of energy flux that can be radiated before striking the divertor target. This is due to the strong T 5/2 e dependence of electron heat conduction which results in very short spatial scales of the T e gradient at high power densities and low temperatures where deuterium and impurities radiate most effectively. However, we have greatly exceeded this constraint on DIII-D with deuterium gas puffing which reduces the peak heat flux to the divertor plate a factor of 5 while distributing the divertor radiation over a long length

Innovations in the LHD divertor program

International Nuclear Information System (INIS)

Ohyabu, N.; Komori, A.; Noda, N.; Morisaki, T.; Sagara, A.; Suzuki, H.; Watanabe, T.; Motojima, O.; Takase, H.

1995-01-01

Various innovative divertor concepts have been developed to improve the LHD plasma performance. They are two divertor magnetic geometries (helical divertor configurations with and without n/m=1/1 island) and two operational scenarios (confinement improvement by generating high temperature divertor plasma and simultaneous achievement of radiative cooling and H-mode-like confinement improvement). In addition, technological development of new efficient hydrogen pumping schemes are being pursued for enhancing the divertor control capability. 16 refs., 4 figs

Divertor modeling for the design of the National Centralized Tokamak with high beta steady-state plasmas

International Nuclear Information System (INIS)

Kawashima, H.; Sakurai, S.; Shimizu, K.; Takizuka, T.; Tamai, H.; Matsukawa, M.; Fujita, T.; Miura, Y.

2006-01-01

The modification of the JT-60U to a fully superconducting coil tokamak, National Centralized Tokamak (NCT) facility, has been programmed to accomplish the high beta steady-state plasma research. A 2D divertor simulation code, SOLDOR/NEUT2D, is applied to the construction of a database for optimum design of the divertor. A semi-closed divertor configuration with vertical target is adopted as the first conceptual divertor design on NCT. With an anticipated SOL power flux of 12 MW at the high beta steady-state operation, the peak heat load on the divertor target is evaluated to be ∼16 MW/m 2 . Effects of divertor geometry and intense gas puffing are demonstrated with a view to reduce the heat load. For the simulation of divertor pumping, we find that the pumping efficiency increases by a factor of 2∼3 by narrowing the divertor gap from 20 to 5 cm. An attractive feature in reducing the heat load and improving the particle controllability has been obtained for a new divertor design due to a recent progress in NCT design

Control of divertor configuration in JT-60

International Nuclear Information System (INIS)

Yoshino, R.; Kukuchi, M.; Ninomiya, H.; Yoshida, H.; Tsuji, S.; Hosogane, N.; Seki, S.

1985-01-01

The control algorithm of JT-60 divertor configuration is presented. JT-60 has five types of poloidal magnetic field coil with each power supply in order to regulate the control objectives mentioned above. However, if one controls each objective by each coil current independently, there must inevitably occur large interaction between control objectives. Because the relation between control objectives and coil currents is complicated. This situation may be the same with a fusion reactor device. For making it possible to control each objective independently without causing large interaction, the authors adopt the noninteracting control algorithm. Hence, this report demonstrates the availability of this method to the control of JT-60 divertor configuration

Migration and deposition of 13C in the full-tungsten ASDEX Upgrade tokamak

International Nuclear Information System (INIS)

Hakola, A; Aho-Mantila, L; Groth, M; Kurki-Suonio, T; Makkonen, T; Likonen, J; Koivuranta, S; Krieger, K; Mayer, M; Mueller, H W; Neu, R; Rohde, V

2010-01-01

The migration of carbon in low-density, low-confinement plasmas of ASDEX Upgrade was studied by injecting 13 C into the main chamber of the torus at the end of the 2007 experimental campaign. A selection of standard tungsten-coated lower-divertor and main-chamber tiles as well as a complete set of lower-divertor tiles with an uncoated poloidal marker stripe were removed from one poloidal cross section and analysed using secondary ion mass spectrometry. The poloidal deposition profiles of 13 C on both the tungsten-coated tiles and on the uncoated graphite areas of the marker tiles were measured and compared. For the W-coated lower-divertor tiles, 13 C was deposited mainly on the high-field side tiles, while barely detectable amounts of 13 C were observed on low-field side samples. In contrast, on the uncoated marker stripes the deposition was equally pronounced in the high-field and low-field side divertor. The marker-tile results are in agreement with those obtained from graphite tiles after the 2003 and 2005 13 C experiments in ASDEX Upgrade. In the case of W-coated tiles, the 13 C measurements were complemented by determining the total amount of deposited carbon ( 12 C) on the tiles, which also shows strong deposition at the inner parts of the lower divertor. The estimated deposition of 13 C on W at the divertor areas was less than 1.5% of the injected amount of 13 C atoms. The 13 C analyses of the main-chamber tiles and small silicon samples mounted in remote areas revealed significant deposition in the upper divertor, in upper parts of the heat shield, in the limiter region close to the injection valve, and below the roof baffle. Approximately 8% of the injected 13 C is estimated to have accumulated in these regions. Possible reasons for the different deposition patterns on W and on graphite in different regions of the torus are discussed.

Surface studies of tungsten erosion and deposition in JT-60U

International Nuclear Information System (INIS)

Ueda, Y.; Fukumoto, M.; Nishikawa, M.; Tanabe, T.; Miya, N.; Arai, T.; Masaki, K.; Ishimoto, Y.; Tsuzuki, K.; Asakura, N.

2007-01-01

In order to study tungsten erosion and migration in JT-60U, 13 W tiles have been installed in the outer divertor region and tungsten deposition on graphite tiles was measured. Dense local tungsten deposition was observed on a CFC tile toroidally adjacent to the W tiles, which resulted from prompt ionization and short range migration of tungsten along field lines. Tungsten deposition with relatively high surface density was found on an inner divertor tile around standard inner strike positions and on an outer wing tile of a dome. On the outer wing tile, tungsten deposition was relatively high compared with carbon deposition. In addition, roughly uniform tungsten depth distribution near the upper edge of the inner divertor tile was observed. This could be due to lift-up of strike point positions in selected 25 shots and tungsten flow in the SOL plasma

Physics design requirements for the National Spherical Torus Experiment liquid lithium divertor

International Nuclear Information System (INIS)

Kugel, H.; Bell, M.; Berzak, L.; Brooks, A.; Ellis, R.; Gerhardt, S.P.; Kaita, R.; Kallman, J.; Majeski, R.; Mansfield, D.K.; Menard, J.; Stotler, D.; Zakharov, L.E.; Maingi, Rajesh; Nygren, R.E.; Soukhanovskii, V.; Wakeland, P.

2009-01-01

Recent National Spherical Tokamak Experiment (NSTX) high-power divertor experiments have shown significant and recurring benefits of solid lithium coatings on plasma facing components (PFCs) to the performance of divertor plasmas in both L- and H-mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15 25% ne decrease from present highest non-inductionless fraction discharges which often evolve toward the density limit, ne/nGW 1), to enable ne scan capability (2) in the H-mode, to test the ability to operate at significantly lower density (e.g., ne/nGW = 0.25), for future reactor designs based on the Spherical Tokamak, and eventually to investigate high heat-flux power handling (10 MW/m2) with long pulse discharges (>1.5 s). The first step (LLD-1) physics design encompasses the desired plasma requirements, the experimental capabilities and conditions, power handling, radial location, pumping capability, operating temperature, lithium filling, MHD forces, and diagnostics for control and characterization.

Modeling of thermal effects on TIBER II divertor during plasma disruptions

International Nuclear Information System (INIS)

Bruhn, M.L.; Perkins, L.J.

1987-01-01

Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs

Development of a compact W-shaped pumped divertor in JT-60U

Energy Technology Data Exchange (ETDEWEB)

Sakurai, S.; Hosogane, N.; Masaki, K.; Kodama, K.; Sasajima, T.; Kishiya, K.; Takahashi, S.; Shimizu, K.; Akino, N.; Miyo, Y.; Hiratsuka, H.; Saidoh, M. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Inoue, M.; Umakoshi, T.; Onozuka, M.; Morimoto, M. [Mitsubishi Heavy Industries, Wadasaki-cho, Hyogo-ku, Kobe-shi, 642 (Japan)

1998-09-01

In JT-60U, the modification to a W-shaped pumped divertor will be completed in May 1997, aiming to realize sufficient reduction in heat flux to the targets and good H-mode confinement simultaneously. W-shaped geometry is optimized not only for forming radiative divertor plasmas and reducing the back flow of neutral particles but also for allowing various experimental configurations. Toroidally and poloidally segmented divertor plates, dome and baffles are arranged in a W-shaped poloidal configuration. The pumping speed can be changed during a shot by variable shutter valves in the three pumping ports under the outer baffle. The net throughput is enough for particle control in the steady radiative operations with high power NBI heating. Carbon fiber composite (CFC) tiles are used for the divertor targets and the divertor throat where large heat flux is expected. Gaps between two adjacent segments are carefully sealed to suppress the leak of neutral gas from the exhaust duct below the divertor and baffles. The strength of the whole structure is confirmed by an electromagnetic force analysis and structural analysis carried out for disruptions of 3 MA discharges with a halo current. (orig.) 11 refs.

Mechanical Design of the NSTX Liquid Lithium Divertor

Energy Technology Data Exchange (ETDEWEB)

R. Ellis, R. Kaita, H. Kugel, G. Paluzzi, M. Viola and R. Nygren

2009-02-19

The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuumcompatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

Enhancing the DEMO divertor target by interlayer engineering

Energy Technology Data Exchange (ETDEWEB)

Barrett, T.R., E-mail: tom.barrett@ccfe.ac.uk [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); McIntosh, S.C.; Fursdon, M.; Hancock, D.; Timmis, W.; Coleman, M. [CCFE, Culham Science Centre, Oxfordshire OX14 3DB (United Kingdom); Rieth, M.; Reiser, J. [Karlsruhe Institute for Technology, IMF-I, D-7602 Karlsruhe (Germany)

2015-10-15

Highlights: • The European ‘near-term’ DEMO forsees a water-cooled divertor. • Divertor targets typically use an interlayer between the armour and structure. • Engineering the properties of the interlayer can yield large gains in performance. • A response surface based design search and optimisation method is used. • A new design passes linear-elastic code rules up to applied heat flux of 18 MW/m{sup 2}. - Abstract: A robust water-cooled divertor target plate solution for DEMO has to date remained elusive. Common to all contemporary concepts is an interlayer at the boundary between the tungsten armour and the cooling structure. In this paper we show by design optimisation that an effectively designed interlayer can produce dramatic gains in power handling. By engineering the interlayer as part of the design study, it is found that divertor performance is enhanced by either a low conductivity ‘Thermal Break’ interlayer or an ‘Ultra-Compliant’ interlayer. For a 10 MW/m{sup 2} surface heat flux we find that a thermal conductivity of 15 W/mK and elastic modulus of 1 GPa are effective. A design is proposed which passes linear-elastic code rules up to an applied heat flux of 18 MW/m{sup 2}.

Enhancing the DEMO divertor target by interlayer engineering

International Nuclear Information System (INIS)

Barrett, T.R.; McIntosh, S.C.; Fursdon, M.; Hancock, D.; Timmis, W.; Coleman, M.; Rieth, M.; Reiser, J.

2015-01-01

Highlights: • The European ‘near-term’ DEMO forsees a water-cooled divertor. • Divertor targets typically use an interlayer between the armour and structure. • Engineering the properties of the interlayer can yield large gains in performance. • A response surface based design search and optimisation method is used. • A new design passes linear-elastic code rules up to applied heat flux of 18 MW/m"2. - Abstract: A robust water-cooled divertor target plate solution for DEMO has to date remained elusive. Common to all contemporary concepts is an interlayer at the boundary between the tungsten armour and the cooling structure. In this paper we show by design optimisation that an effectively designed interlayer can produce dramatic gains in power handling. By engineering the interlayer as part of the design study, it is found that divertor performance is enhanced by either a low conductivity ‘Thermal Break’ interlayer or an ‘Ultra-Compliant’ interlayer. For a 10 MW/m"2 surface heat flux we find that a thermal conductivity of 15 W/mK and elastic modulus of 1 GPa are effective. A design is proposed which passes linear-elastic code rules up to an applied heat flux of 18 MW/m"2.

Mechanical Design of the NSTX Liquid Lithium Divertor

International Nuclear Information System (INIS)

Ellis, R.; Kaita, R.; Kugel, H.; Paluzzi, G.; Viola, M.; Nygren, R.

2009-01-01

The Liquid Lithium Divertor (LLD) on NSTX will be the first test of a fully-toroidal liquid lithium divertor in a high-power magnetic confinement device. It will replace part of the lower outboard divertor between a specified inside and outside radius, and ultimately provide a lithium surface exposed to the plasma with enough depth to absorb a significant particle flux. There are numerous technical challenges involved in the design. The lithium layer must be as thin as possible, and maintained at a temperature between 200 and 400 degrees Celsius to minimize lithium evaporation. This requirement leads to the use of a thick copper substrate, with a thin stainless steel layer bonded to the plasma-facing surface. A porous molybdenum layer is then plasma-sprayed onto the stainless steel, to provide a coating that facilitates full wetting of the surface by the liquid lithium. Other challenges include the design of a robust, vacuum compatible heating and cooling system for the LLD. Replacement graphite tiles that provided the proper interface between the existing outer divertor and the LLD also had to be designed, as well as accommodation for special LLD diagnostics. This paper describes the mechanical design of the LLD, and presents analyses showing the performance limits of the LLD.

Utilization of vanadium alloys in the DIII-D Radiative Divertor Program

International Nuclear Information System (INIS)

Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

1995-10-01

Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics (GA), in conjunction with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan for the utilization of vanadium alloys as part of the Radiative Divertor (RD) upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy (DOE). This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components, and (2) demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming Radiative Divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development (R and D) efforts to support fabrication development and to resolve key issues related to environmental effects

Utilization of vanadium alloys in the DIII-D radiative divertor program

International Nuclear Information System (INIS)

Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

1996-01-01

Vanadium alloys are attractive candidate structural materials for fusion power plants because of their potential for minimum environmental impact due to low neutron activation and rapid activation decay. They also possess favorable material properties for operation in a fusion environment. General Atomics in conjunction with Argonne National Laboratory and Oak Ridge National Laboratory has developed a plan for the utilization of vanadium alloys as part of the radiative divertor upgrade for the DIII-D tokamak. The plan will be carried out in conjunction with General Atomics and the Materials Program of the US Department of Energy. This application of a vanadium alloy will provide a meaningful step in the development of advanced materials for fusion power devices by: (1) developing necessary materials processing technology for the fabrication of large vanadium alloy components and (2) demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak environment. The program consists of three phases: first, small vanadium alloy coupon samples will be exposed in DIII-D at positions in the vessel floor and within the pumping plenum region of the existing divertor structure; second, a small vanadium alloy component will be installed in the existing divertor, and third, during the forthcoming radiative divertor modification, scheduled for completion in mid-1997, the upper section of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. This program also includes research and development efforts to support fabrication development and to resolve key issues related to environmental effects. (orig.)

European DEMO divertor target: Operational requirements and material-design interface

Directory of Open Access Journals (Sweden)

J.H. You

2016-12-01

Full Text Available Recently, an integrated program of conceptual design activities for the European DEMO reactor was launched in the framework of the EUROfusion Consortium, where reliable power handling capability was identified as one of the most critical scientific as well as technological challenges for a DEMO reactor. The divertor is the key in-vessel plasma-facing component being in charge of power exhaust and removal of impurity particles. The DEMO divertor target will have to withstand extreme thermal loads where the local peak heat flux is expected to reach up to 20 MW/m2 during slow transient events in DEMO. To assure sufficient heat removal capability of the divertor target against normal and transient operational scenarios under expected cumulative neutron dose of up to 13 dpa is one of the fundamental engineering challenges imposed on target design. To develop the design of the DEMO divertor and related technologies, an R&D work package ‘Divertor’ has been set up in this consortium. The subproject ‘Target Development’ is devoted to the development of the conceptual design and the core technologies of the plasma-facing target. Devising and implementing novel structural heat sink materials (e.g. W/Cu composites to advanced target design concepts is one of the major objectives of this subproject. In this paper, the underlying design requirements imposed by the envisaged power exhaust goal and the prominent material-design interface issues are discussed. In addition, the candidate design concepts being currently considered are presented together with the related material issues. Finally, the first results achieved so far are presented.

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

International Nuclear Information System (INIS)

El-Morshedy, Salah El-Din; Hassanein, Ahmed

2009-01-01

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m 2 plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

Transient thermal hydraulic modeling and analysis of ITER divertor plate system

Energy Technology Data Exchange (ETDEWEB)

El-Morshedy, Salah El-Din [Argonne National Laboratory, Argonne, IL (United States); Atomic Energy Authority, Cairo (Egypt)], E-mail: selmorshedy@etrr2-aea.org.eg; Hassanein, Ahmed [Purdue University, West Lafayette, IN (United States)], E-mail: hassanein@purdue.edu

2009-12-15

A mathematical model has been developed/updated to simulate the steady state and transient thermal-hydraulics of the International Thermonuclear Experimental Reactor (ITER) divertor module. The model predicts the thermal response of the armour coating, divertor plate structural materials and coolant channels. The selected heat transfer correlations cover all operating conditions of ITER under both normal and off-normal situations. The model also accounts for the melting, vaporization, and solidification of the armour material. The developed model is to provide a quick benchmark of the HEIGHTS multidimensional comprehensive simulation package. The present model divides the coolant channels into a specified axial regions and the divertor plate into a specified radial zones, then a two-dimensional heat conduction calculation is created to predict the temperature distribution for both steady and transient states. The model is benchmarked against experimental data performed at Sandia National Laboratory for both bare and swirl tape coolant channel mockups. The results show very good agreements with the data for steady and transient states. The model is then used to predict the thermal behavior of the ITER plasma facing and structural materials due to plasma instability event where 60 MJ/m{sup 2} plasma energy is deposited over 500 ms. The results for ITER divertor response is analyzed and compared with HEIGHTS results.

Scrape-off layer and divertor theory meeting: Proceedings

International Nuclear Information System (INIS)

1994-03-01

This report contains viewgraphs on the following topics: fluid modelling of neutrals in the SOL and divertor; instabilities of gas-fueled divertors: theory and adaptive simulations; stability of ionization fronts of gaseous divertor plasmas; monte carlo calculation of heat transport; reduced charge model for edge impurity flows; thermally collapsed solutions for gaseous/radiative divertors; adaptive grid methods in transport simulation; advanced numerical solution algorithms applied to the multispecies edge plasma equations; two-dimensional edge plasma simulation using the multigrid method; neutral behavior and the effects of neutral-neutral and neutral-ion elastic scattering in the ITER gaseous divertor; particle throughput in the TPX divertor; marfes in tokamaks; a comparative study of the limiter and divertor edge plasmas in TEXT-U; issues of toroidal tokamak-type divertor simulators; ASDEX upgrade; the ITER divertor; the DIII-D divertor program and TPX divertor; DEGAS 2: a transmission/escape probabilities model for neutral particle transport: comparison with DEGAS 2; a collisional radiative model of hydrogen for high recycling divertors; comparison of fluid and non- fluid neutral models in B2.5; DIII-D radiative divertor simulations; 3-D fluid simulations of turbulence from conducting wall mode; turbulence and drifts in SOL plasmas; recent results for 1 1/2-D ITER gas target divertor modelling; evaluation of pumping and fueling in coupled core, SOL, and divertor chamber calculations; and ITER gas target divertors: comparison of volume recombination and large radial transport scenarios using DEGAS

A survey of problems in divertor and edge plasma theory

International Nuclear Information System (INIS)

Boozer, A.; Braams, B.; Weitzner, H.; Hazeltine, R.; Houlberg, W.; Oktay, E.; Sadowski, W.; Wootton, A.

1992-01-01

Theoretical physics problems related to divertor design are presented, organized by the region in which they occur. Some of the open questions in edge physics are presented from a theoretician's point of view. After a cursory sketch of the fluid models of the edge plasma and their numerical realization, the following topics are taken up: time-dependent problems, non-axisymmetric effects, anomalous transport in the scrape-off layer, edge kinetic theory, sheath effects and boundary conditions in divertors, electric field effects, atomic and molecular data issues, impurity transport in the divertor region, poloidally localized power dissipation (MARFEs and dense gas targets), helium ash removal, and neutral transport. The report ends with a summary of selected problems of particular significance and a brief bibliography of survey articles and related conference proceedings

The isotope effect on divertor conditions and neutral pumping in horizontal divertor configurations in JET-ILW Ohmic plasmas

Directory of Open Access Journals (Sweden)

J. Uljanovs

2017-08-01

Full Text Available Understanding the impact of isotope mass and divertor configuration on the divertor conditions and neutral pressures is critical for predicting the performance of the ITER divertor in DT operation. To address this need, ohmically heated hydrogen and deuterium plasma experiments were conducted in JET with the ITER-like wall in varying divertor configurations. In this study, these plasmas are simulated with EDGE2D-EIRENE outfitted with a sub-divertor model, to predict the neutral pressures in the plenum with similar fashion to the experiments. EDGE2D-EIRENE predictions show that the increased isotope mass results in up to a 25% increase in peak electron densities and 15% increase in peak ion saturation current at the outer target in deuterium when compared to hydrogen for all horizontal divertor configurations. Indicating that a change from hydrogen to deuterium as main fuel decreases the neutral mean free path, leading to higher neutral density in the divertor. Consequently, this mechanism also leads to higher neutral pressures in the sub-divertor. The experimental data provided by the hydrogen and deuterium ohmic discharges shows that closer proximity of the outer strike point to the pumping plenum results in a higher neutral pressure in the sub-divertor. The diaphragm capacitance gauge pressure measurements show that a two to three-fold increase in sub-divertor pressure was achieved in the corner and nearby horizontal configurations compared to the far-horizontal configurations, likely due to ballistic transport (with respect to the plasma facing components of the neutrals into the sub-divertor. The corner divertor configuration also indicates that a neutral expansion occurs during detachment, resulting in a sub-divertor neutral density plateau as a function of upstream density at the outer-mid plane.

Modelling of the material transport and layer formation in the divertor of JET: Comparison of ITER-like wall with full carbon wall conditions

International Nuclear Information System (INIS)

Kirschner, A.; Matveev, D.; Borodin, D.; Airila, M.; Brezinsek, S.; Groth, M.; Wiesen, S.; Widdowson, A.; Beal, J.; Esser, H.G.; Likonen, J.; Bekris, N.; Ding, R.

2015-01-01

Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented

Modelling of the material transport and layer formation in the divertor of JET: Comparison of ITER-like wall with full carbon wall conditions

Energy Technology Data Exchange (ETDEWEB)

Kirschner, A., E-mail: a.kirschner@fz-juelich.de [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Matveev, D.; Borodin, D. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Airila, M. [VTT Technical Research Centre of Finland, 02044 VTT (Finland); Brezinsek, S. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Groth, M. [Aalto University, Otakaari 4, 02015 Espoo (Finland); Wiesen, S. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Widdowson, A. [Culham Centre for Fusion Energy, Abingdon OX14 3DB (United Kingdom); Beal, J. [York Plasma Institute, Department of Physics, University of York, Heslington, York YO10 5DD (United Kingdom); Esser, H.G. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Trilateral Euregio Cluster, 52425 Jülich (Germany); Likonen, J. [VTT Technical Research Centre of Finland, 02044 VTT (Finland); Bekris, N. [Karlsruhe Institute of Technology, Institute for Technical Physics, Hermann-von-Helmholtz-Platz 1, Bau 451, 76344 Eggenstein-Leopoldshafen (Germany); Ding, R. [Institute of Plasma Physics, Chinese Academy of Sciences, P.O. Box 1126, Hefei, Anhui 230031 (China)

2015-08-15

Impurity transport within the inner JET divertor has been modelled with ERO to estimate the transport to and the resulting deposition at remote areas. Various parametric studies involving divertor plasma conditions and strike point position have been performed. In JET-ILW (beryllium main chamber and tungsten divertor) beryllium, flowing from the main chamber into the divertor and then effectively reflected at the tungsten divertor tiles, is transported to remote areas. The tungsten flux to remote areas in L-Mode is in comparison to the beryllium flux negligible due to small sputtering. However, tungsten is sputtered during ELMs in H-Mode conditions. Nevertheless, depending on the plasma conditions, strike point position and the location of the remote area, the maximum resulting tungsten flux to remote areas is at least ∼3 times lower than the corresponding beryllium flux. Modelled beryllium and tungsten deposition on a rotating collector probe located below tile 5 is in good agreement with measurements if the beryllium influx into the inner divertor is assumed to be in the range of 0.1% relative to the deuterium ion flux and erosion due to fast charge exchange neutrals is considered. Comparison between JET-ILW and JET-C is presented.

Impurity diagnosis of a KSTAR graphite divertor tile using laser induced breakdown spectroscopy technique

Energy Technology Data Exchange (ETDEWEB)

Kim, Minju; Cho, Min Sang; Cho, Byoung Ick, E-mail: bicho@gist.ac.kr

2017-04-15

Laser induced breakdown spectroscopy (LIBS) has been tested to diagnose impurity elements on a Korea Superconducting Tokamak Advanced Research (KSTAR) divertor tile. Spectral lines of various impurity elements such as iron, chromium, and nickel were detected from the divertor surface. The variation of spectra with consecutive laser pulses demonstrates the potential for depth profiling analysis for the deposited impurity layer. The LIBS plasma parameters have been qualitatively determined from analysis of the relative line intensities and linewidths for each element. The validity of this analysis has been checked with atomic spectral simulations.

Estimation of peak heat flux onto the targets for CFETR with extended divertor leg

International Nuclear Information System (INIS)

Zhang, Chuanjia; Chen, Bin; Xing, Zhe; Wu, Haosheng; Mao, Shifeng; Luo, Zhengping; Peng, Xuebing; Ye, Minyou

2016-01-01

Highlights: • A hypothetical geometry is assumed to extend the outer divertor leg in CFETR. • Density scan SOLPS simulation is done to study the peak heat flux onto target. • Attached–detached regime transition in out divertor occurs at lower puffing rate. • Unexpected delay of attached–detached regime transition occurs in inner divertor. - Abstract: China Fusion Engineering Test Reactor (CFETR) is now in conceptual design phase. CFETR is proposed as a good complement to ITER for demonstrating of fusion energy. Divertor is a crucial component which faces the plasmas and handles huge heat power for CFETR and future fusion reactor. To explore an effective way for heat exhaust, various methods to reduce the heat flux to divertor target should be considered for CFETR. In this work, the effect of extended out divertor leg on the peak heat flux is studied. The magnetic configuration of the long leg divertor is obtained by EFIT and Tokamak Simulation Code (TSC), while a hypothetical geometry is assumed to extend the out divertor leg as long as possible inside vacuum vessel. A SOLPS simulation is performed to study peak heat flux of the long leg divertor for CFETR. D 2 gas puffing is used and increasing of the puffing rate means increase of plasma density. Both peak heat flux onto inner and outer targets are below 10 MW/m 2 is achieved. A comparison between the peak heat flux between long leg and conventional divertor shows that an attached–detached regime transition of out divertor occurs at lower gas puffing gas puffing rate for long leg divertor. While for the inner divertor, even the configuration is almost the same, the situation is opposite.

Estimation of peak heat flux onto the targets for CFETR with extended divertor leg

Energy Technology Data Exchange (ETDEWEB)

Zhang, Chuanjia; Chen, Bin [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Xing, Zhe [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Wu, Haosheng [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Mao, Shifeng, E-mail: sfmao@ustc.edu.cn [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Luo, Zhengping; Peng, Xuebing [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China); Ye, Minyou [School of Nuclear Science and Technology, University of Science and Technology of China, 96 Jinzhai Road, Hefei, Anhui 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei, Anhui 230031 (China)

2016-11-01

Highlights: • A hypothetical geometry is assumed to extend the outer divertor leg in CFETR. • Density scan SOLPS simulation is done to study the peak heat flux onto target. • Attached–detached regime transition in out divertor occurs at lower puffing rate. • Unexpected delay of attached–detached regime transition occurs in inner divertor. - Abstract: China Fusion Engineering Test Reactor (CFETR) is now in conceptual design phase. CFETR is proposed as a good complement to ITER for demonstrating of fusion energy. Divertor is a crucial component which faces the plasmas and handles huge heat power for CFETR and future fusion reactor. To explore an effective way for heat exhaust, various methods to reduce the heat flux to divertor target should be considered for CFETR. In this work, the effect of extended out divertor leg on the peak heat flux is studied. The magnetic configuration of the long leg divertor is obtained by EFIT and Tokamak Simulation Code (TSC), while a hypothetical geometry is assumed to extend the out divertor leg as long as possible inside vacuum vessel. A SOLPS simulation is performed to study peak heat flux of the long leg divertor for CFETR. D{sub 2} gas puffing is used and increasing of the puffing rate means increase of plasma density. Both peak heat flux onto inner and outer targets are below 10 MW/m{sup 2} is achieved. A comparison between the peak heat flux between long leg and conventional divertor shows that an attached–detached regime transition of out divertor occurs at lower gas puffing gas puffing rate for long leg divertor. While for the inner divertor, even the configuration is almost the same, the situation is opposite.

Facilities for technology testing of ITER divertor concepts, models, and prototypes in a plasma environment

International Nuclear Information System (INIS)

Cohen, S.A.

1991-12-01

The exhaust of power and fusion-reaction products from ITER plasma are critical physics and technology issues from performance, safety, and reliability perspectives. Because of inadequate pulse length, fluence, flux, scrape-off layer plasma temperature and density, and other parameters, the present generation of tokamaks, linear plasma devices, or energetic beam facilities are unable to perform adequate technology testing of divertor components, though they are essential contributors to many physics issues such as edge-plasma transport and disruption effects and control. This Technical Requirements Documents presents a description of the capabilities and parameters divertor test facilities should have to perform accelerated life testing on predominantly technological divertor issues such as basic divertor concepts, heat load limits, thermal fatigue, tritium inventory and erosion/redeposition. The cost effectiveness of such divertor technology testing is also discussed

Role of the pump limiter throat-ergodic divertor effect on edge plasma

International Nuclear Information System (INIS)

Grosman, A.; Samain, A.; Ghendrih, P.; Capes, H.; Morera, J.P.

1988-01-01

A large part of the Tore Supra programme is devoted to plasma edge studies. Two types of such density control apparatus have been implemented, a set of pumps limiters and the ergodic divertor. The goal of the present paper is to investigate the effect of the pump limiter throat on pumping efficiency. We present also the possibilities of the ergodic divertor device to facilitate plasma pumping and power exhaust

Thermal effects of divertor sweeping in ITER

International Nuclear Information System (INIS)

Wesley, J.C.

1992-01-01

In this paper, thermal effects of magnetically sweeping the separatrix strike point on the outer divertor target of the International Thermonuclear Fusion Reactor (ITER) are calculated. For the 0. 2 Hz x ± 12 cm sweep scenario proposed for ITER operations, the thermal capability of a generic target design is found to be slightly inadequate (by ∼ 5%) to accommodate the full degree of plasma scrape-off peaking postulated as a design basis. The principal problem identified is that the 5 s sweep period is long relative to the 1. 4 s thermal time constant of the divertor target. An increase of the sweep frequency to ∼ 1 Hz is suggested: this increase would provide a power handling margin of ∼ 25% relative to present operational criteria

Numerical exploration of non-axisymmetric divertor closure in the small angle slot (SAS) divertor at DIII-D

Science.gov (United States)

Frerichs, Heinke; Schmitz, Oliver; Covele, Brent; Guo, Houyang; Hill, David; Feng, Yuhe

2017-10-01

In the Small Angle Slot (SAS) divertor in DIII-D, the combination of misaligned slot structure and non-axisymmetric perturbations to the magnetic field causes the strike point to vary radially along the divertor slot and even leave it at some toroidal locations. This effect essentially introduces an opening in the divertor slot from where recycling neutrals can easily escape, and thereby degrade performance of the slot divertor. This effect has been approximated by a finite gap in the divertor baffle. Simulations with EMC3-EIRENE show that a toroidally localized loss of divertor closure can result in non-axisymmetric divertor densities and temperatures. This introduces a density window of 10-15% on top of the nominal threshold separatrix density during which a non-axisymmetric onset of local detachment occurs, initially leaving the gap and up to 60 deg beyond that still attached. Conversely, the impact of such toroidally localized divertor perturbations on the toroidal symmetry of midplane separatrix conditions is small. This work has been funded by the U.S. Department of Energy under Early Career Award Grant DE-SC0013911, and Grant DE-FC02-04ER54698.

Assessment of radiation maps during activated divertor moving in the ITER building

International Nuclear Information System (INIS)

Ying Dongchuan; Zeng Qin; Qiu Yuefeng; Dang Tongqiang; Wu Yican; Loughlin, Michael

2011-01-01

As the main interface components between plasma and vacuum vessel, the divertor is foreseen to be removed to the hot cell for refurbishment during the 20 years of ITER operation. During this process, the activated divertor will cause a large increase of radiation in the ITER building. 3D analysis has been performed to assess the radiation maps throughout the ITER building for assisting the shielding design for personnel and sensitive equipment. The activation of the divertor has been determined by coupled neutron transport and inventory calculations, radiation maps have been obtained from gamma transport calculations. The neutron and gamma transport calculations have been performed by MCNP5 code with FENDL2.1library. The inventory calculations have been performed by FISPACT2007 code with EAF-2007 library. The results of these 3D decay gamma radiation maps are presented by pictures in this paper, including the biological dose maps and maps of heat deposition in electronic equipment.

ASDEX upgrade - definition of a tokamak experiment with a reactor compatible polaoidal divertor

International Nuclear Information System (INIS)

1982-03-01

ASDEX Upgrade is intended as the next experimental step after ASDEX. It is designed to investigate the physics of a divertor tokamak as closely as possible to fusion reactor requirements, without thermonuclear heating. It is characterized by a poloidal divertor configuration with divertor coils located outside the toroidal field coils, by machine parameters which allow a line density within the plasma boundary sufficient to screen fast CX particles from the plasma core, by a scrape-off layer essentially opaque to neutrals produced at the target plates, and, finally, by an auxiliary heating power high enough for producing a reactor-like power flux density through the plasma boundary. Design considerations on the basis of physical and technical constraints yielded the tokamak system optimized with respect to effort and costs as described in the following. It uses normal-conducting coil systems, is the size of ASDEX, and has a field of 3.9 T, a plasma current of up to 1.5 MA, and a pulse duration of 10 s. To provide the required power flux density, an ICRH power of 10 MW is needed. For comparison, a superconducting version is under investigation. (orig.)

Parametric analysis of the thermal effects on the divertor in tokamaks during plasma disruptions

International Nuclear Information System (INIS)

Bruhn, M.L.

1988-04-01

Plasma disruptions are an ever present danger to the plasma-facing components in today's tokamak fusion reactors. This threat results from our lack of understanding and limited ability to control this complex phenomenon. In particular, severe energy deposition occurs on the divertor component of the double-null configured tokamak reactor during such disruptions. A hybrid computational model developed to estimate and graphically illustrate global thermal effects of disruptions on the divertor plates is described in detail. The quasi-two-dimensional computer code, TADDPAK (Thermal Analysis Divertor during Disruptions PAcKage), is used to conduct parametric analysis for the TIBER II Tokamak Engineering Test Reactor Design. The dependence of these thermal effects on divertor material choice, disruption pulse length, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is investigated for this reactor design. Results and conclusions from this analysis are presented. Improvements to this model and issues that require further investigation are discussed. Cursory analysis for ITER (International Thermonuclear Experimental Reactor) is also presented in the appendix. 75 refs., 49 figs., 10 tabs

Preliminary analysis of the efficiency of non-standard divertor configurations in DEMO

Directory of Open Access Journals (Sweden)

F. Subba

2017-08-01

Full Text Available The standard Single Null (SN divertor is currently expected to be installed in DEMO. However, a number of alternative configurations are being evaluated in parallel as backup solutions, in case the standard divertor does not extrapolate successfully from ITER to a fusion power plant. We used the SOLPS code to produce a preliminary analysis of two such configurations, the X-Divertor (XD and the Super X-Divertor (SX, and compare them to the SN solution. Considering the nominal power flowing into the SOL (PSOL = 150 MW, we estimated the amplitude of the acceptable DEMO operational space. The acceptability criterion was chosen as plasma temperature at the target lower than 5eV, providing low sputtering and at least partial detachment, while the operational space was defined in terms of the electron density at the outboard mid-plane separatrix and of the seeded impurity (Ar only in the present study concentration. It was found that both the XD and the SXD extend the DEMO operational space, although the advantages detected so far are not dramatic. The most promising configuration seems to be the XD, which can produce acceptable target temperatures at moderate outboard mid-plane electron density (nomp=4.5×1019 m−3 and Zeff= 1.3.

Engineering and design aspects related to the development of the ITER divertor

International Nuclear Information System (INIS)

Dietz, J.; Chiocchio, S.; Antipenkov, A.

1994-01-01

Most of the divertor concepts proposed for the Next Step devices relied on the exhaust of the SOL power to target plates which intersect the magnetic field fines. The resulting highly peaked thermal load, together with the concentrated fluxes of energetic particles, posed severe design constraints and ultimately led to unacceptably short target lifetime. The ITER high density gas target divertor concept is based on transferring the nominal power perpendicular to the magnetic field lines from the plasma edge onto large surfaces and on dissipating the particles' energy through atomic and molecular mechanisms. While the basic ideas for this approach have been motivated by recent results in present tokamaks, a full assessment of this concept still requires extensive experimental and modelling work. The paper describes the engineering and design aspects involving the development of the ITER divertor and shows how the physics assumptions translate into engineering requirements, and how the additional existing constraints (such as the limited space, neutron load, electromagnetic effects, compatibility with other components, remote maintainability) have been taken into account for the design definition. The concept developed takes advantage of the spatial separation of the several physics phenomena anticipated to take place in the divertor, thus relaxing the needs to accommodate in the same region opposing requirements

Characterized the pattern of the material deposition in the HL-2A tokamak

Energy Technology Data Exchange (ETDEWEB)

Cai, Laizhong, E-mail: cailz@swip.ac.cn [Southwestern Institute of Physics, P.O.Box 432, Chengdu, Sichuan 610041 (China); Wang, Jianbao; Wu, Ting; Zeng, Xiaoxiao [Southwestern Institute of Physics, P.O.Box 432, Chengdu, Sichuan 610041 (China); Hai, Ran; Ding, Hongbin [Dalian University of Technology, Dalian, Liaoning 116024 (China)

2017-03-15

Since the divertor geometry of a tokamak has a strong impact on the material erosion and deposition on the wall and HL-2A has a unique divertor configuration, it is necessary to investigate the material deposition pattern in HL-2A although a few results on other tokamaks have already been published. In this paper, tiles retrieved from the vessel are analyzed ex-situ by SIMS, SEM and laser-induced breakdown spectroscopy (LIBS). And deposition behind the lower divertor is in-situ measured by a quartz crystal microbalance (QMB). The deposition in HL-2A displays a complex pattern and clear localization characteristic. The thickness of the deposition layer varies in the range of 0-4μm. And in-situ diagnostic of QMB indicates that the average thickness of the deposition layer per pulse is over ten nanometers. In addition, the results imply that Si, Fe and D have different behaviors during the material deposition in HL-2A.

Numerical studies on divertor experiments

International Nuclear Information System (INIS)

Ueda, N.; Itoh, K.; Itoh, S.-I.; Tanaka, M.; Hasegawa, M.; Shoji, T.; Sugihara, M.

1988-04-01

Numerical analysis on the divertor experiments such as JFT-2M tokamak is made by use of the two-dimensional time-dependent simulation code. The plasma in the scrape-off layer (SOL) and divertor region is solved for the given particle and heat sources from the main plasma, Γ p and Q T . Effect of the direction of the toroidal magnetic field is studied. It is found that the heat flux which is proportional to b vector x ∇T i has influences on the divertor plasmas, but has a small effect on the parameters on the midplane in the framework of the fluid model. Parameter survey on Γ p and Q T is made. The transient response of the SOL/divertor plasma to the sudden change of Γ p and Q T is studied. Time delay in the SOL and divertor region is calculated. (author)

Comparative divertor-transport study for helical devices

International Nuclear Information System (INIS)

Feng, Y.; Sardei, F.; Kobayashi, M.

2008-10-01

Using the island divertors (ID) of W7-AS and W7-X and the helical divertor (HD) of LHD as examples, the paper presents a comparative divertor transport study for three typical helical devices of different machine-size following two distinct divertor concepts, aiming at identifying common physics issues/effects for mutual validation and combined studies. Based on EMC3/EIRENE simulations supported by experimental results, the paper first reviews and compares the essential transport features of the W7-AS ID and the LHD HD in order to build a base and framework for a predictive study of W7-X. Revealed is the fundamental role of the low-order magnetic islands in both divertor concepts. Preliminary EMC3/EIRENE simulation results for W7-X are presented and discussed with respect to W7-AS and LHD in order to show how the individual field and divertor topologies affect the divertor transport and performance. For instance, a high recycling regime which is absent from W7-AS and LHD is expected for W7-X. Topics addressed are restricted to the basic function elements of a divertor such as particle flux enhancement and impurity retention. In particular, the divertor function on reducing the influx of intrinsic impurities is examined for all the three devices under different divertor plasma conditions. Special attention is paid to examining the island screening potential of intrinsic impurities which has been predicted for all the three devices under high divertor collisionality conditions. The results are discussed in conjunction with the experimental observations for high density divertor plasmas in W7-AS and LHD. (author)

Development of divertor remote maintenance system

Energy Technology Data Exchange (ETDEWEB)

Takeda, Nobukazu; Oka, Kiyoshi; Akou, Kentaro; Takiguchi, Yuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-04-01

The ITER divertor is categorized as a scheduled maintenance component because of extreme heat and particle loads it is exposed to by plasma. It is also highly activated by 14 MeV neutrons. Reliable remote handling equipment and tools are required for divertor maintenance under intense gamma radiation. To facilitate remote maintenance, the divertor is segmented into 60 cassettes, and each cassette weighing about 25 tons and maintained and replaced through four maintenance ports each 90 degrees. Divertor cassettes must be transported toroidally and radially for replacement through maintenance ports. Remote handling involving cassette movers and carriers for toroidal and radial transport has been developed. Under the ITER R and D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.S. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team. (author)

Development of divertor remote maintenance system

International Nuclear Information System (INIS)

Takeda, Nobukazu; Oka, Kiyoshi; Akou, Kentaro; Takiguchi, Yuji

1998-01-01

The ITER divertor is categorized as a scheduled maintenance component because of extreme heat and particle loads it is exposed to by plasma. It is also highly activated by 14 MeV neutrons. Reliable remote handling equipment and tools are required for divertor maintenance under intense gamma radiation. To facilitate remote maintenance, the divertor is segmented into 60 cassettes, and each cassette weighing about 25 tons and maintained and replaced through four maintenance ports each 90 degrees. Divertor cassettes must be transported toroidally and radially for replacement through maintenance ports. Remote handling involving cassette movers and carriers for toroidal and radial transport has been developed. Under the ITER R and D program, technology critical to divertor cassette maintenance is being developed jointly by Japan, E.U., and U.S. home teams. This paper summarizes divertor remote maintenance design and the status of technology development by the Japan Home Team. (author)

Radiative divertor plasmas with convection in DIII-D

International Nuclear Information System (INIS)

Leonard, A.W.; Porter, G.D.; Wood, R.D.; Allen, S.L.; Boedo, J.; Brooks, N.H.; Evans, T.E.; Fenstermacher, M.E.; Hill, D.N.; Isler, R.C.; Lasnier, C.J.; Lehmer, R.D.; Mahdavi, M.A.; Maingi, R.; Moyer, R.A.; Petrie, T.W.; Schaffer, M.J.; Wade, M.R.; Watkins, J.G.; West, W.P.; Whyte, D.G.

1998-01-01

The radiation of divertor heat flux on DIII-D [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] is shown to greatly exceed the limits imposed by assumptions of energy transport dominated by electron thermal conduction parallel to the magnetic field. Approximately 90% of the power flowing into the divertor is dissipated through low-Z radiation and plasma recombination. The dissipation is made possible by an extended region of low electron temperature in the divertor. A one-dimensional analysis of the parallel heat flux finds that the electron temperature profile is incompatible with conduction-dominated parallel transport. Plasma flow at up to the ion acoustic speed, produced by upstream ionization, can account for the parallel heat flux. Modeling with the two-dimensional fluid code UEDGE [T. Rognlien, J. L. Milovich, M. E. Rensink, and G. D. Porter, J. Nucl. Mater. 196 endash 198, 347 (1992)] has reproduced many of the observed experimental features. copyright 1998 American Institute of Physics

Development of a high-heat flux cooling element with potential application in a near-term fusion power plant divertor

Energy Technology Data Exchange (ETDEWEB)

Nicholas, Jack Robert, E-mail: jack.nicholas@eng.ox.ac.uk [Osney Thermo-Fluids Laboratory, University of Oxford, Oxford (United Kingdom); Ireland, Peter [Osney Thermo-Fluids Laboratory, University of Oxford, Oxford (United Kingdom); Hancock, David [CCFE, Culham, Oxfordshire (United Kingdom); Robertson, Dan [Rolls-Royce Plc., Derby, Derbyshire (United Kingdom)

2015-10-15

Highlights: • Laminate jet impingement system introduced for high pressure operation (17 MPa+). • Numerical thermo-fluid analysis on baseline geometry. • Cascade impingement shown to reduce divertor mass flow rate requirements and increase fluid temperature change. • Numerical thermo-fluid analysis validated using scaled experiments with air. - Abstract: A low temperature jet impingement based heat sink module has been developed for potential application in a near-term fusion power plant divertor. The design is composed of a number of hexagonal CuCrZr sheets bonded together in a stack to form a laminate structure. This method allows the production of complex flow paths using relatively simple manufacturing techniques. The thermo-fluid performance of a baseline design employing cascade jet impingement has been assessed and compared to a non-cascade case. Experimental validation of the numerical work was carried out on a scaled model using air as the working fluid. Local heat transfer coefficients were obtained on the surface using surface temperature data from thermochromic liquid crystals.

Manufacturing and joining technologies for helium cooled divertors

International Nuclear Information System (INIS)

Aktaa, J.; Basuki, W.W.; Weber, T.; Norajitra, P.; Krauss, W.; Konys, J.

2014-01-01

Highlights: • The manufacturing and joining technologies developed at KIT for helium cooled divertors are reviewed and critically discussed. • Various technologies have been pursued and further developed aiming divertor components with very high quality and sufficient reliability. • Very promising routes have been found for which however still R and D works are necessary. • Technologies developed are also useful for other divertor and even blanket concepts, particularly those with tungsten armor. - Abstract: In the helium cooled (HC) divertor, developed at KIT for a fusion power plant, tungsten has been selected as armor as well as structural material due to its crucial properties: high melting point, very low sputtering yield, good thermal conductivity, high temperature strength, low thermal expansion and low activation. Thereby the armor tungsten is attached to the structural tungsten by thermally conductive joint. Due to the brittleness of tungsten at low temperatures its use as structural material is limited to the high temperature part of the component and a structural joint to the reduced activation ferritic martensitic steel EUROFER97 is foreseen. Hence, to realize the selected hybrid material concept reliable tungsten–steel and tungsten–tungsten joints have been developed and will be reported in this paper. In addition, the modular design of the HC divertor requires tungsten armor tiles and tungsten structural thimbles to be manufactured in high numbers with very high quality. Due to the high strength and low temperature brittleness of tungsten special manufacturing techniques need to be developed for the production of parts with no cavities inside and/or surface flaws. The main achievement in developing the respective manufacturing technologies will be presented and discussed. To achieve the objectives mentioned above various manufacturing and joining technologies are pursued. Their later applicability depends on the level of development

Progress in ergodic divertor operation on Tore Supra

International Nuclear Information System (INIS)

Ghendrih, Ph.; Becoulet, M.; Colas, L.; Grosman, A.; Guirlet, R.; Gunn, J.; Loarer, T.; Azeroual, A.; Basiuk, V.; Beaumont, B.; Becoulet, A.; Bremond, S.; Bucalossi, J.; Capes, H.; Corre, Y.; Costanzo, L.; Michelis, C. de; Devynck, P.; Feron, S.; Friant, C.; Garbet, X.; Giannella, R.; Grisolia, C.; Hess, W.; Hogan, J.; Ladurelle, L.; Laugier, F.; Martin, G.; Mattioli, M.; Meslin, B.; Monier-Garbet, P.; Moulin, D.; Nguyen, F.; Pascal, J.Y.; Pecquet, A.L.; Pegourie, B.; Reichle, R.; Saint-Laurent, F.; Vallet, J.C.; Zabiego, M.

1999-09-01

Upgrade of the Tore ergodic divertor has led to significant progress in ergodic divertor physics. The disruptive limit governed by the stochastization of the outer magnetic surfaces is found to occur for a value of the Chirikov parameter reaching 2 on the magnetic surface q = 2 + 3 / 12. This experimentally observed robustness allows one to operate at very low safety factor on the separatrix (q ∼ 2). Numerical analysis of ballooning turbulence in a stochastic layer indicates that the decay of the density fluctuations is in associated with an increase of the fluctuating electric drift velocity. The bottom line is then an enhanced cross-field transport in the vicinity of the target plates. This lowering of confinement appears to be compensated by an intrinsic transport barrier on the electron temperature. The 3-D response of the temperature field is computed with a fluid code. The intrinsic transport barrier at the separatrix, reported experimentally, can be recovered together with small amplitude temperature modulations in the divertor volume. Experimental evidence of the 3 density regimes (linear, high recycling and detachment) is reported. The low critical density values for these transitions indicate that similar parallel physics govern the axisymmetric and ergodic divertor, despite the open configuration of the latter. Measurement and understanding of these density regimes provide a means for feedback control of plasma density and an improvement in ICRH coupling scenarios. Experimental data also indicated that particle control with the vented target plates is effective. Increase of impurity control and radiation efficiency are recalled. Global power balance has been analysed. These results confirm the enhanced radiation capacity of the ergodic divertor. (author)

NSTX Tangential Divertor Camera

International Nuclear Information System (INIS)

Roquemore, A.L.; Ted Biewer; Johnson, D.; Zweben, S.J.; Nobuhiro Nishino; Soukhanovskii, V.A.

2004-01-01

Strong magnetic field shear around the divertor x-point is numerically predicted to lead to strong spatial asymmetries in turbulence driven particle fluxes. To visualize the turbulence and associated impurity line emission near the lower x-point region, a new tangential observation port has been recently installed on NSTX. A reentrant sapphire window with a moveable in-vessel mirror images the divertor region from the center stack out to R 80 cm and views the x-point for most plasma configurations. A coherent fiber optic bundle transmits the image through a remotely selected filter to a fast camera, for example a 40500 frames/sec Photron CCD camera. A gas puffer located in the lower inboard divertor will localize the turbulence in the region near the x-point. Edge fluid and turbulent codes UEDGE and BOUT will be used to interpret impurity and deuterium emission fluctuation measurements in the divertor

Theory of Advanced Magnetic Divertors

Science.gov (United States)

Kotschenreuther, Michael; Valanju, Prashant; Mahajan, Swadesh; Covele, Brent

2013-10-01

The magnetic field structure in the SOL is the most important determinant of divertor physics. A comprehensive analytical and numerical methodology is developed to investigate SOL magnetic fields in the backdrop of two advanced divertor geometries- the X-divertor (XD) proposed and discussed in 2004, and the snowflake divertor (SFD) of 2007-2010. The analysis shows that XD and SFD represent very distinct and readily distinguishable magnetic geometries, epitomized through a differentiating metric, the Divertor Index (DI). In terms of this simple metric, the XD (DI > 1) and the SFD (DI XD flux surfaces are less convergent, in fact, divergent (flaring). These different SOL magnetics imply different physics, particularly with respect to detachment dynamics. It is also shown that some experiments on NSTX and DIII-D match both the prescription and the predictions of the 2004 XD paper. Work supported under US-DOE projects DE-FG02-04ER54742 and DE-FG02-04ER54754.

Diagnostics for Evaluating Performance of NSTX Liquid Lihium Divertor

Science.gov (United States)

Kaita, R.; Kugel, H.; Kallman, J.; Leblanc, B.; Paul, S.; Roquemore, A. L.; Skinner, C.; Soukhanovskii, V.; Maingi, R.; Ahn, J.-W.; Wilgen, J.; Allain, J.-P.; Taylor, C.

2009-11-01

A Liquid Lithium Divertor (LLD) is being installed on NSTX to investigate particle control and power handling with liquid lithium as plasma-facing component (PFC). The LLD is expected to provide a low-recycling plasma-facing component (PFC). To study the effects of such a PFC on plasma performance, a variety of edge measurements are required. Since its surface is highly reflective at visible wavelengths, a Lyman-alpha detector array will be used to monitor the recycling. To understand changes in edge transport, electron temperature and density measurements will be made with Langmuir probes mounted in PFC's near the LLD, and the edge sightlines of a multipoint Thomson scattering system. A frequency-scanning reflectometer will also provide scrapeoff layer electron density profiles. The LLD response to heat loads will be examined with infrared cameras and thermocouples. Diagnostics are also needed to measure the erosion and codeposition of lithium. They include quartz deposition monitors and a retractable probe for exposing samples to the plasma.

Fabrication of divertor cassette for ITER

International Nuclear Information System (INIS)

Sanguinetti, G.P.

2008-01-01

The Divertor is the component located on the bottom of the ITER vacuum vessel, whose main function is to adsorb the high thermal flux generated by the plasma whilst keeping the plasma impurity at a reasonable low level. The divertor consist of 54 units, each comprising outer components, facing the plasma and a component supporting the plasma facing components (PFC) and providing coolant distribution to them (divertor cassette). The divertor cassette is a box structure, butt welded and machined, made from plates and forgins of austenitic stainless steels. The cassette fabrication, which is in detail described, includes manufacturing of the attachments of the PFC to the cassette, the coolant distribution channels, and the cassette to vacuum vessel locking system. The divertor cassette is a pressure component (the cooling water runs at 40 bar) and therefore divertor cassette design, fabrication and service shall comply with the European PED and the applicable French law for the ITER. (orig.)

A large divertor manipulator for ASDEX Upgrade

Energy Technology Data Exchange (ETDEWEB)

Herrmann, Albrecht, E-mail: albrecht.herrmann@ipp.mpg.de; Jaksic, Nikola; Leitenstern, Peter; Greuner, Henri; Krieger, Karl; Marné, Pascal de; Oberkofler, Martin; Rohde, Volker; Schall, Gerd

2015-10-15

Highlights: • A large divertor manipulator for ASDEX Upgrade is developed and tested. • It allows replacing a relevant part of the divertor by dedicated targets and probes. • Modified solid standard targets. • Electrical and mechanical probes for dedicated investigations. • Test of actively cooled component. - Abstract: In 2013 a new bulk tungsten divertor, Div-III, was installed in ASDEX Upgrade (AUG). During the concept and design phase of Div-III the option of adaptable divertor instrumentation and divertor modification as contribution for divertor investigations in preparation of ITER was given a high priority. To gain flexibility for the test of divertor modifications without affecting the operational space of AUG, the large divertor manipulator, DIM-II, was designed and installed. DIM-II allows to retract 2 out of 128 outer divertor target tiles including the water cooled support structure into a target exchange box and to replace these targets without breaking the vacuum of the AUG vessel. DIM-II is based on a carriage-rail system with a driving rod pushing a front-end with the target module into the divertor position for plasma operation. Three types of front-ends are foreseen for physics investigations: (i) modified standard targets clamped to the standard cooling structure, (ii) dedicated front-ends making use of the whole available volume of about 230 × 160 × 80 mm{sup 3} and (iii) actively cooled/heated targets for cooling water temperatures up to 230 °C. This paper presents the DIM-II design including the FEM calculations for the modified divertor support structure and the front-end options, as well as the test procedure and operation mode.

The remote exchange of the JET divertor

International Nuclear Information System (INIS)

Pick, M.

1999-01-01

In 1997 a series of experiments were performed in the JET machine using deuterium-tritium (D-T) mixtures and resulting in discharges with record breaking fusion power and fusion energy. The experiments demonstrated a key technology required for fusion, namely the on-line operation of a tritium fuel re-processing plant. These experiments left the inside of the JET vessel inaccessible to manned access for approximately one year. During this time, the complete Mark IIA divertor, a major system within the torus, was successfully removed and replaced with a new divertor design, the Mark II Gas Box divertor, using only remote handling techniques. This was the first application of the JET remote handling system and a demonstration of a further key ITER technology. The paper explains the methodology and operational approach taken to achieve the results using the remote handling system developed at JET. It describes the remote handling equipment including the force-reflecting servo-manipulator, the specialised tools designed, the facilities needed, and the trials, planning and training carried out to ensure the safe, reliable and rapid completion of the remote handling tasks. The planned tasks are outlined including the execution of the novel procedure for a remote, sub-millimetre precision, dimensional survey of the divertor support structure using digital photogrammetry. Furthermore the paper shows how the adaptability of the system was used to successfully undertake a large number of unplanned tasks including the removal of damaged tiles, a damaged diagnostic system and the vacuum cleaning of diagnostic windows. (author)

Impact of the impurity seeding for divertor protection on the performance of fusion reactors

Science.gov (United States)

Siccinio, Mattia; Fable, Emiliano; Angioni, Clemente; Saarelma, Samuli; Scarabosio, Andrea; Zohm, Hartmut

2017-10-01

A 0D divertor and scrape-off layer (SOL) model has been coupled to the 1.5D core transport code ASTRA. The resulting numerical tool has been employed for various parameter scans in order to identify the most convenient choices for the operation of electricity producing fusion devices with seeded impurities for the divertor protection. In particular, the repercussions of such radiative species on the main plasma through the fuel dilution have been taken into account. The main result we found is that, when the limits on the maximum tolerable divertor heat flux are enforced, the curves at constant electrical power output are closed on themselves in the R-BT plane, i.e. no improvement would descend from a further increase of R or BT once the maximum has been reached. This occurrence appears as an intrinsic physical limit for all devices where a radiative SOL is needed to deal with the power exhaust. Furthermore, the relative importance of the different power loss channels (e.g. hydrogen radiation, charge exchange, perpendicular transport and impurity radiation), through which the power entering the SOL is dissipated before reaching the target plate, is investigated with our model.

Study of high-Z target plate materials in the divertor of ASDEX-Upgrade

Energy Technology Data Exchange (ETDEWEB)

Hirsch, S; Asmussen, K; Engelhardt, W; Field, A R; Fussmann, G; Lieder, G; Naujoks, D; Neu, R; Radtke, R; Wenzel, U [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1994-12-31

The reduction of divertor tile erosion is a challenging problem in present and future tokamaks. Until now, graphite has almost exclusively been used for divertor plates, and it is estimated that unacceptable amounts of material would be eroded under reactor relevant conditions where power fluxes to the target plates as high as 20 MW/m{sup 2} are expected. In a high-recycling divertor with relatively low temperature (5 eVdivertor plates, in-situ studies of the erosion of various divertor target materials have been performed by means of passive spectroscopy. From our spectroscopic observations we infer that under high density divertor conditions the advantages of high-Z materials become fully efficient. (author) 6 refs., 2 figs.

Comment on “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)

International Nuclear Information System (INIS)

Ryutov, D. D.; Cohen, R. H.; Rognlien, T. D.; Soukhanovskii, V. A.; Umansky, M. V.

2014-01-01

In the recently published paper “Magnetic geometry and physics of advanced divertors: The X-divertor and the snowflake” [Phys. Plasmas 20, 102507 (2013)], the authors raise interesting and important issues concerning divertor physics and design. However, the paper contains significant errors: (a) The conceptual framework used in it for the evaluation of divertor “quality” is reduced to the assessment of the magnetic field structure in the outer Scrape-Off Layer. This framework is incorrect because processes affecting the pedestal, the private flux region and all of the divertor legs (four, in the case of a snowflake) are an inseparable part of divertor operation. (b) The concept of the divertor index focuses on only one feature of the magnetic field structure and can be quite misleading when applied to divertor design. (c) The suggestion to rename the divertor configurations experimentally realized on NSTX (National Spherical Torus Experiment) and DIII-D (Doublet III-D) from snowflakes to X-divertors is not justified: it is not based on comparison of these configurations with the prototypical X-divertor, and it ignores the fact that the NSTX and DIII-D poloidal magnetic field geometries fit very well into the snowflake “two-null” prescription

THERMAL HYDRAULIC ANALYSIS OF FIRE DIVERTOR

International Nuclear Information System (INIS)

C.B. bAXI; M.A. ULRICKSON; D.E. DRIMEYER; P. HEITZENROEDER

2000-01-01

The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and with alpha dominated heating. The outer divertor and baffle of FIRE are water cooled. The worst thermal condition for the outer divertor and baffle is the baseline D-T operating mode (10 T, 6.6 MA, 20 s) with a plasma exhaust power of 67 MW and a peak heat flux of 20 MW/m 2 . A swirl tape (ST) heat transfer enhancement method is used in the outer divertor cooling channels to increase the heat transfer coefficient and the critical heat flux (CHF). The plasma-facing surface consists of tungsten brush. The finite element (FE) analysis shows that for an inlet water temperature of 30 C, inlet pressure of 1.5 MPa and a flow velocity of 10 m/s, the incident critical heat flux is greater than 30 MW/m 2 . The peak copper temperature is 490 C, peak tungsten temperature is 1560 C, and the pressure drop is less than 0.5 MPa. All these results fulfill the design requirements

Evaluation of helium cooling for fusion divertors

International Nuclear Information System (INIS)

Baxi, C.B.

1993-09-01

The divertors of future fusion reactors will have a power throughput of several hundred MW. The peak heat flux on the diverter surface is estimated to be 5 to 15 MW/m 2 at an average heat flux of 2 MW/m 2 . The divertors have a requirement of both minimum temperature (100 degrees C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermo-mechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are: (1) Manifold sizes; (2) Pumping power; and (3) Leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques are considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled diverter module was designed and fabricated by GA for steady-state heat flux of 10 MW/m 2 . This module was recently tested at Sandia National Laboratories. At an inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW/m 2 . The pumping power required was less than 1% of the power removed. These results verified the design prediction

Plasma flow in the DIII-D divertor

International Nuclear Information System (INIS)

Boedo, J.A.; Porter, G.D.; Schaffer, M.J.

1998-07-01

Indications that flows in the divertor can exhibit complex behavior have been obtained from 2-D modeling but so far remain mostly unconfirmed by experiment. An important feature of flow physics is that of flow reversal. Flow reversal has been predicted analytically and it is expected when the ionization source arising from neutral or impurity ionization in the divertor region is large, creating a high pressure zone. Plasma flows arise to equilibrate the pressure. A radiative divertor regime has been proposed in order to reduce the heat and particle fluxes to the divertor target plates. In this regime, the energy and momentum of the plasma are dissipated into neutral gas introduced in the divertor region, cooling the plasma by collisional, radiative and other atomic processes so that the plasma becomes detached from the target plates. These regimes have been the subject of extensive studies in DIII-D to evaluate their energy and particle transport properties, but only recently it has been proposed that the energy transport over large regions of the divertor must be dominated by convection instead of conduction. It is therefore important to understand the role of the plasma conditions and geometry on determining the region of convection-dominated plasma in order to properly control the heat and particle fluxes to the target plates and hence, divertor performance. The authors have observed complex structures in the deuterium ion flows in the DIII-D divertor. Features observed include reverse flow, convective flow over a large volume of the divertor and stagnant flow. They have measured large gradients in the plasma potential across the separatrix in the divertor and determined that these gradients induce poloidal flows that can potentially affect the particle balance in the divertor

Design of DIII-D advanced divertor

International Nuclear Information System (INIS)

Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thruston, G.

1989-01-01

The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffing. The Advanced Divertor has two principal components: ( 1) a toroidally symmetric baffle; and (2) a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. 2 refs., 4 figs

Design of DIII-D Advanced Divertor

International Nuclear Information System (INIS)

Smith, J.P.; Baxi, C.B.; Reis, E.; Schaffer, M.; Thurston, G.

1989-11-01

The Advanced Divertor is a modification being designed for the plasma chamber of the DIII-D tokamak in order to optimize the divertor configuration and allow a broader range of experiments to be carried out. The Advanced Divertor will enable two classes of physics experiments to be run in DIII-D: Divertor biasing and Divertor baffling. The Advanced Divertor has two principal components: a toroidally symmetric baffle; and a continuous ring electrode. The tokamak can be run in baffle, bias, or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D plasma control system. The baffle will contain approximately 50,000 l/s pumping for particle removal in the outer bottom corner of the vacuum vessel. The strike point will be positioned at the entrance aperture for the baffle mode. The aperture geometry is designed to facilitate a large particle influx plus a high probability that backstreaming particles will be reionized and redirected to the aperture. Where the baffling plates meet, gas sealing is required to prevent recycling of neutrals back into the plasma. The electrode is a continuous water-cooled ring, armored with graphite. The ring is electrically isolated from the vessel wall and is biasable to 1 kV and 20 kA. The outer leg of the divertor will be positioned on the graphite covered ring during biasing experiments. The supports for the ring are radially flexible to handle the differential thermal growth between the ring and the vessel wall but stiff in the vertical direction to restrain the ring against large disruption forces. The coolant and electrical feeds are designed in a similar manner. All the feeds are supported from and maintain a 5 kV isolation to the vessel wall. 2 refs., 4 figs

Comparison between a pumped-limiter and a divertor for the next step machines

International Nuclear Information System (INIS)

Harrison, F.F.A.

1985-01-01

The paper presents a simple description of the physics issues which influence the conceptual design of a pumped-limiter and single-null poloidal divertor in a next step, long burn tokamak of NET/INTOR scale. Predicted performance of the limiter and divertor are compared in regard to localised recycling, sputtering of the plasma collection surfaces, penetration of sputtered impurities into the fusion plasma, surface power loading and exhaust of helium ash. It is concluded that the performance of the divertor is superior and that it can be predicted with a reasonable degree of confidence. The viability of the limiter remains in doubt but the concept cannot be rejected at the present time

Heat loads to divertor nearby components from secondary radiation evolved during plasma instabilities

Energy Technology Data Exchange (ETDEWEB)

Sizyuk, V., E-mail: vsizyuk@purdue.edu; Hassanein, A., E-mail: hassanein@purdue.edu [Center for Materials under Extreme Environment, School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)

2015-01-15

A fundamental issue in tokamak operation related to power exhaust during plasma instabilities is the understanding of heat and particle transport from the core plasma into the scrape-off layer and to plasma-facing materials. During abnormal and disruptive operation in tokamaks, radiation transport processes play a critical role in divertor/edge-generated plasma dynamics and are very important in determining overall lifetimes of the divertor and nearby components. This is equivalent to or greater than the effect of the direct impact of escaped core plasma on the divertor plate. We have developed and implemented comprehensive enhanced physical and numerical models in the upgraded HEIGHTS package for simulating detailed photon and particle transport in the evolved edge plasma during various instabilities. The paper describes details of a newly developed 3D Monte Carlo radiation transport model, including optimization methods of generated plasma opacities in the full range of expected photon spectra. Response of the ITER divertor's nearby surfaces due to radiation from the divertor-developed plasma was simulated by using actual full 3D reactor design and magnetic configurations. We analyzed in detail the radiation emission spectra and compared the emission of both carbon and tungsten as divertor plate materials. The integrated 3D simulation predicted unexpectedly high damage risk to the open stainless steel legs of the dome structure in the current ITER design from the intense radiation during a disruption on the tungsten divertor plate.

Possible divertor solutions for a fusion reactor. Pt. I. Physical aspects based on present day divertor operation

International Nuclear Information System (INIS)

Kallenbach, A.; Bosch, H.-S.; De Pena Hempel, S.; Dux, R.; Kaufmann, M.; Mertens, V.; Neuhauser, J.; Suttrop, W.; Zohm, H.

1997-01-01

For pt.II see ibid., p.109-117 (1997). With an anticipated power flux across the separatrix of up to 300 MW of an ITER-like fusion reactor, conventional measures of power spread lead to a peak power load at the target plates in the order of 30 MW m -2 , far beyond the technically feasible limit for stationary operation. Radiative cooling by seed impurities appears to be the most promising plasma-physical option to reduce the target power load, but extrapolations of present experiments predict an only marginally tolerable increase of the plasma effective charge Z eff . Key points will be the achievement of very high electron densities, leading to more effective radiative cooling by δP rad /δZ eff ∝n e 2 while keeping the edge temperature within its optimum range. This range is bounded from below by the H→L mode temperature threshold due to confinement requirements, whereas the upper boundary is given by the ideal ballooning stability limit which is connected to type-I ELM activity which may cause non-tolerable divertor heat loads. The completely detached H-mode (CDH) in ASDEX Upgrade demonstrates radiative H-mode operation within this operational range exhibiting high-frequent type-III ELMs and target power load in the order of 10% of the heating power. At present, open questions on high density reactor operation are related to radiative instabilities as well as edge transport enhancement and H-mode impairment observed in several tokamaks under high density conditions. Measures to overcome these detrimental effects will be investigated with improved divertor concepts in the near future. The possible problems connected to high density reactor operation can be relaxed, if the design of plasma facing components with higher heat flux endurance is successful. (orig.)

The ITER divertor cassette project meeting

International Nuclear Information System (INIS)

Merola, M.; Riccardi, B.; Tivey, R.

1999-01-01

The Divertor Cassette Project topical meeting was held on May 26-28, 1999 at the ENEA Brasimone Research Centre in Camugnano (Bologna), Italy. Specialists from all the four Parties and the JCT participated in the meeting. It was concluded that the Divertor Cassette Project has significantly contributed to solving a large part of the critical issues of the ITER divertor design

Design analysis of the ITER divertor

International Nuclear Information System (INIS)

Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F.; Merola, M.; Riccardi, B.; Petrizzi, L.; Villari, R.

2007-01-01

The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

Design analysis of the ITER divertor

Energy Technology Data Exchange (ETDEWEB)

Samuelli, G.; Marin, A.; Roccella, M.; Lucca, F. [L.T. Calcoli SaS, Merate (Lecco) (Italy); Merola, M. [ITER Team, Cadarache (France); Riccardi, B. [EFDA CSU Garching (Germany); Petrizzi, L.; Villari, R. [CRE ENEA sulla Fusione Frascati, Roma (Italy)

2007-07-01

The divertor is one of the most challenging components of the ITER machine. Its function is to reduce the impurity in the plasma and consists essentially of two parts: the plasma facing components (PFCs) and a massive support structure called the cassette body (CB). Considerable R and D effort (developed by EFDA CSU GARCHING and the ITER International Team together with the EU Associations and the EU Industries) has been spent in designing divertor components capable of withstanding the expected electromagnetic (EM) loads and to take into account the latest ITER design conditions. In support of such efforts extensive and very detailed Neutronic, Thermal, EM and Structural analyses have been performed. A summary of the analyses performed will be presented. One of the main result is a typical exercise of integration between the different kind of analyses and the importance of keeping the consistency between the different assumptions and simplifications. The models used for the numerical analyses include a detailed geometrical description of the CB, the inlet, outlet hydraulic manifolds, the CB to vacuum vessel locking system and three configurations of the PFU. The effect of electrical bridging, both in poloidal and toroidal direction, of the PFU castellation, due to a possible melting at the W mono-block or tiles, occurring during the plasma disruptions, has been analyzed. For all these configurations 2 VDE scenarios including the effect of the Toroidal Field Variation and the HaloCurrent with the related out of plane induced EM forces have been extensively analyzed and a detailed poloidal and radial distribution of the nuclear heating has been used for the neutronic flux on the divertor components. The aim of this activity is to produce a comprehensive design and assessment of the ITER divertor via: -The estimation of the neutronic heat deposition and shielding capability; -The calculation of the related thermal and mechanical effects and the comparison of the

Engineering of the divertor injection tokamak experiment (DITE)

International Nuclear Information System (INIS)

Plummer, K.M.; Bayes, D.V.; Bell, D.; Burt, J.; Galloway, F.; Sanders, B.C.; Skelton, D.E.; Varley, G.L.

1976-01-01

The DITE assembly has been constructed to study the effect of powerful neutral injection and the use of magnetic divertors in Tokamak systems. In addition, the plasma is stabilized by a position controlled feed-back vertical field system developed from results on the CLEO experiment, and added to DITE later in the design stage. The machine is designed for an ultimate plasma current of 340 kA, having a minor radius of 23 cm at q = 2, on a major radius of 113 cm. The 28 kG Bphi field, from 16 liquid nitrogen cooled coils has a 2% ripple at the edge of the plasma. The divertor is a ''bundle'' type, the present design of which is limited to operating in a Bphi field of 18 kG. Neutral Injection, initially by two, and ultimately by four injectors, is intended to supply about 1,500 kW of beam power. The engineering is now complete and the machine commissioned; this paper describes the up-to-date design of the machine and includes some of our experiences during design, construction and commissioning

Engineering design of a toroidal divertor for the EBT-S fusion device. Final report, Phase II. EBT-S divertor project

International Nuclear Information System (INIS)

Mai, L.P.; Malick, F.S.

1981-01-01

The mechanical, structural, thermal, electrical, and vacuum design of a magnetic toroidal divertor system for the Elmo Bumpy Torus (EBT-S) is presented. The EBT-S is a toroidal magnetic fusion device located at the ORNL that operates under steady state conditions. The engineering of the divertor was performed during the second of three phases of a program aimed at the selection, design, fabrication, and installation of a magnetic divertor for EBT-S. The magnetic analysis of the toroidal divertor was performed during Phase I of the program and has been reported in a separate document. In addition to the details of the divertor design, the modest modifications that are required to the EBT-S device and facility to accommodate the divertor system are presented

Module of lithium divertor for KTM tokamak

Energy Technology Data Exchange (ETDEWEB)

Lyublinski, I., E-mail: yublinski@yandex.ru [FSUE ' Red Star' , Moscow (Russian Federation); Vertkov, A.; Evtikhin, V.; Balakirev, V.; Ionov, D.; Zharkov, M. [FSUE ' Red Star' , Moscow (Russian Federation); Tazhibayeva, I. [IAE NNC RK, Kurchatov (Kazakhstan); Mirnov, S. [TRINITI, Troitsk, Moscow Region (Russian Federation); Khomiakov, S.; Mitin, D. [OJSC Dollezhal Institute, Moscow (Russian Federation); Mazzitelli, G. [ENEA RC Frascati (Italy); Agostini, P. [ENEA RC Brasimone (Italy)

2012-10-15

Highlights: Black-Right-Pointing-Pointer Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Black-Right-Pointing-Pointer Capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. Black-Right-Pointing-Pointer Lithium divertor module for KTM tokamak is under development. Black-Right-Pointing-Pointer Lithium filled tungsten felt is offered as the base plasma facing material of divertor. Black-Right-Pointing-Pointer Results of this project addresses to the progress in the field of fusion neutrons source and fusion energy source creation. - Abstract: Activity on projects of ITER and DEMO reactors has shown that solution of problems of divertor target plates and other plasma facing elements (PFEs) based on the solid plasma facing materials cause serious difficulties. Problems of PFE degradation, tritium accumulation and plasma pollution can be overcome by the use of liquid lithium-metal with low Z. Application of lithium will allow to create a self-renewal and MHD stable liquid metal surface of the in-vessel devices possessing practically unlimited service life; to reduce power flux due to intensive re-irradiation on lithium atoms in plasma periphery that will essentially facilitate a problem of heat removal from PFE; to reduce Z{sub eff} of plasma to minimally possible level close to 1; to exclude tritium accumulation, that is provided with absence of dust products and an opportunity of the active control of the tritium contents in liquid lithium. Realization of these advantages is based on use of so-called lithium capillary-porous system (CPS) - new material in which liquid lithium fill a solid matrix from porous material. The progress in development of lithium technology and also activity in lithium experiments in the tokamaks TFTR, T-11M, T-10, FTU, NSTX, HT-7 and stellarator TJ II permits of solving the problems in development of

Understanding the SOL flow in L-mode plasma on divertor tokamaks, and its influence on the plasma transport

International Nuclear Information System (INIS)

Asakura, Nobuyuki

2007-01-01

Significant progress has been made in understanding the driving mechanisms in SOL mass transport along the magnetic field lines (SOL flow). SOL flow measurements by Mach probes and impurity plume have been performed in L-mode plasma at various poloidal locations in divertor tokamaks. All results showed common SOL flow patterns: subsonic flow with parallel Mach number (M parallel ) of 0.2-1 was generated from the Low-Field-Side (LFS) SOL to the High-Field-Side (HFS) divertor for the ion ∇B drift towards the divertor. The SOL flow pattern was formed mainly by LFS-enhanced asymmetry in diffusion and by classical drifts. In addition, divertor detachment and/or intense puffing-and-pump enhanced the HFS SOL flow. Most codes have incorporated drift effects, and asymmetric diffusion was modelled to simulate the fast SOL flow. Influences of the fast SOL flow on the impurity flow in the SOL, shielding from core plasma, and deposition profile, were directly observed in experiments

Recent advances towards a lithium vapor box divertor

Directory of Open Access Journals (Sweden)

R.J. Goldston

2017-08-01

Full Text Available Fusion power plants are likely to require near complete detachment of the divertor plasma from the divertor target plates, in order to have both acceptable heat flux at the target to avoid prompt damage and also acceptable plasma temperature at the target surface, to minimize long-term erosion. However hydrogenic and impurity puffing experiments show that detached operation leads easily to x-point MARFEs, impure plasmas, degradation in confinement, and lower helium pressure at the exhaust. The concept of the Lithium Vapor Box Divertor is to use local evaporation and strong differential pumping through condensation to localize low-Z gas-phase material that absorbs the plasma heat flux and so achieve detachment while avoiding these difficulties. The vapor localization has been confirmed using preliminary Navier–Stokes calculations. We use ADAS calculations of εcool, the plasma energy lost per injected lithium atom, to estimate the lithium vapor pressure, and so temperature, required for detachment, taking into account power balance. We also develop a simple model of detachment to evaluate the required upstream density, based on further taking into account dynamic pressure balance. A remarkable general result is found, not just for lithium-vapor-induced detachment, that the upstream density divided by the Greenwald-limit density scales as nup/nGW ∝ (P5/8/B3/8 Tdet1/2/(εcool+γTdet, with no explicit size scaling. Tdet is the temperature just before strong pressure loss, assumed to be ∼ ½ of the ionization potential of the dominant recycling species, and γ is the sheath heat transmission factor.

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

Czech Academy of Sciences Publication Activity Database

Matveev, D.; Kirschner, A.; Schmid, K.; Litnovsky, A.; Borodin, D.; Komm, Michael; Van Oost, G.; Samm, U.

-, T159 (2014), 014063-014063 ISSN 0031-8949 Institutional support: RVO:61389021 Keywords : plasma * tokamak * tritium retention * ITER * castellated surfaces * gaps * divertor * impurity deposition Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.126, year: 2014 http://iopscience.iop.org/1402-4896/2014/T159/014063/

Westinghouse compact poloidal divertor reference design

International Nuclear Information System (INIS)

Yang, T.F.; Lee, A.Y.; Ruck, G.W.

1977-08-01

A feasible compact poloidal divertor system has been designed as an impurity control and vacuum vessel first-wall protection option for the TNS tokamak. The divertor coils are inside the TF coil array and vacuum vessel. The poloidal divertor is formed by a pair of coil sets with zero net current. Each set consists of a number of coils forming a dish-shaped washer-like ring. The magnetic flux in the space between the coil sets is compressed vertically to limit the height and to expand the horizontal width of the particle and energy burial chamber which is located in the gap between the coil sets. The intensity of the poloidal field is increased to make the pitch angle of the flux lines very large so that the diverted particles can be intercepted by a large number of panels oriented at a small angle with respect to the flux lines. They are carefully shaped and designed such that the entire surfaces are exposed to the incident particles and are not shadowed by each other. Large collecting surface areas can be obtained. Flowing liquid lithium film and solid metal panels have been considered as the particle collectors. The power density for the former is designed at 1 MW/m 2 and for the latter 0.5 MW/m 2 . The major mechanical, thermal, and vacuum problems have been evaluated in sufficient detail so that the advantages and difficulties are identified. A complete functional picture is presented

Interaction of ICRF power and edge plasma in Tore Supra ergodic divertor configuration

International Nuclear Information System (INIS)

Nguyen, F.; Grosman, A.; Basiuk, V.; Fraboulet, D.; Beaumont, B.; Becoulet, A.; Ghendrih, Ph.; Ladurelle, L.; Meslin, B.

2000-01-01

The coupling of ICRF power to plasma is a crucial problem in Tore Supra for high power and long pulse operations and depends greatly on the edge parameters, in particular on the edge density. Conversely, the behaviour of the bulk plasma is related to the edge conditions and the injection of RF power also induces major modifications on the edge plasma. Moreover, the Ergodic Divertor (ED) of Tore Supra imposes a complex configuration at the edge due to the presence of the magnetic perturbation. Several diagnostics are available to study the interaction of ICRF power with the edge plasma: Langmuir probes on the ED modules, infra red (IR) cameras, charge exchange neutral analysers. In minority heating scheme, the edge density is very sensitive to any perturbation in the high recycling regime which is always found in the ED configuration for relevant plasma parameters. Partially detached regimes, with or without inhomogeneities of density and temperature induced by the flux tubes of the laminar layer, are obtained for high resistance coupling values. The coupling is then not very robust and feedback control or antenna automatic matching techniques are developed. In fast wave electron heating scheme with ED, various fast wave absorption mechanisms (minority heating, Mode Conversion, Alfven resonance) are present at the plasma edge due to the large size of the plasma. The ICRF coupling is difficult due to the low fast wave direct electron damping, even with high hydrogen minority scheme. An increase of the injected ICRF power could improve this situation

Conceptual design of CFETR divertor remote handling compatible structure

International Nuclear Information System (INIS)

Dai, Huaichu; Yao, Damao; Cao, Lei; Zhou, Zibo; Li, Lei

2016-01-01

Highlights: • Conceptual design for the CFETR divertor have been proposed, especially the divertor remote handling compatible structure. • The degrees of freedom of the divertor are analyzed in order to validate the design the divertor supports structure. • Besides the ITER-like scheme, a new scheme for the divertor remote handling compatible supports is proposed, that is the rack and pinion mechanism. • The installation/removel process is verified through simulation in Delmia in order to check design quality for remote handling requirements. - Abstract: Divertor is one of key components of tokamak fusion reactor. The CFETR is China Fusion Engineering Test Reactor. Its divertor will expose to tritium environment and neutron radiation. Materials of the divertor will be radioactived, and cannot be handled by personnel directly. To develop structure which compatible with robots handle for installation, maintenance and removing is required. This paper introduces a conceptual design of CFETR divertor module which compatible with remote handling end-effectors. The divertor module is confined by inner and outer support. The inner support is only confined divertor module radial, toroidal and vertical moving freedom degrees, but not confined rotating freedom degrees. The outer support is the structure that can confine rotating freedom degrees and should also be compatible with remote handling end-effectors.

Conceptual design of CFETR divertor remote handling compatible structure

Energy Technology Data Exchange (ETDEWEB)

Dai, Huaichu, E-mail: yaodm@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei (China); Yao, Damao; Cao, Lei; Zhou, Zibo; Li, Lei [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

2016-11-15

Highlights: • Conceptual design for the CFETR divertor have been proposed, especially the divertor remote handling compatible structure. • The degrees of freedom of the divertor are analyzed in order to validate the design the divertor supports structure. • Besides the ITER-like scheme, a new scheme for the divertor remote handling compatible supports is proposed, that is the rack and pinion mechanism. • The installation/removel process is verified through simulation in Delmia in order to check design quality for remote handling requirements. - Abstract: Divertor is one of key components of tokamak fusion reactor. The CFETR is China Fusion Engineering Test Reactor. Its divertor will expose to tritium environment and neutron radiation. Materials of the divertor will be radioactived, and cannot be handled by personnel directly. To develop structure which compatible with robots handle for installation, maintenance and removing is required. This paper introduces a conceptual design of CFETR divertor module which compatible with remote handling end-effectors. The divertor module is confined by inner and outer support. The inner support is only confined divertor module radial, toroidal and vertical moving freedom degrees, but not confined rotating freedom degrees. The outer support is the structure that can confine rotating freedom degrees and should also be compatible with remote handling end-effectors.

Rapidly Moving Divertor Plates In A Tokamak

International Nuclear Information System (INIS)

Zweben, S.

2011-01-01

It may be possible to replace conventional actively cooled tokamak divertor plates with a set of rapidly moving, passively cooled divertor plates on rails. These plates would absorb the plasma heat flux with their thermal inertia for ∼10-30 sec, and would then be removed from the vessel for processing. When outside the tokamak, these plates could be cooled, cleaned, recoated, inspected, and then returned to the vessel in an automated loop. This scheme could provide nearoptimal divertor surfaces at all times, and avoid the need to stop machine operation for repair of damaged or eroded plates. We describe various possible divertor plate designs and access geometries, and discuss an initial design for a movable and removable divertor module for NSTX-U.

Atomic and molecular processes in JT-60U divertor plasmas

Energy Technology Data Exchange (ETDEWEB)

Takenaga, H.; Shimizu, K.; Itami, K. [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1997-01-01

Atomic and molecular data are indispensable for the understanding of the divertor characteristics, because behavior of particles in the divertor plasma is closely related to the atomic and molecular processes. In the divertor configuration, heat and particles escaping from the main plasma flow onto the divertor plate along the magnetic field lines. In the divertor region, helium ash must be effectively exhausted, and radiation must be enhanced for the reduction of the heat load onto the divertor plate. In order to exhaust helium ash effectively, the difference between behavior of neutral hydrogen (including deuterium and tritium) and helium in the divertor plasma should be understood. Radiation from the divertor plasma generally caused by the impurities which produced by the erosion of the divertor plate and/or injected by gas-puffing. Therefore, it is important to understand impurity behavior in the divertor plasma. The ions hitting the divertor plate recycle through the processes of neutralization, reflection, absorption and desorption at the divertor plates and molecular dissociation, charge-exchange reaction and ionization in the divertor plasma. Behavior of hydrogen, helium and impurities in the divertor plasmas can not be understood without the atomic and molecular data. In this report, recent results of the divertor study related to the atomic and molecular processes in JT-60U were summarized. Behavior of neural deuterium and helium was discussed in section 2. In section 3, the comparisons between the modelling of the carbon impurity transport and the measurements of C II and C IV were discussed. In section 4, characteristics of the radiative divertor using Ne puffing were reported. The new diagnostic method for the electron density and temperature in the divertor plasmas using the intensity ratios of He I lines was described in section 5. (author)

Snowflake divertor experiments on TCV

International Nuclear Information System (INIS)

Piras, F; Coda, S; Duval, B P; Labit, B; Marki, J; Moret, J-M; Pitzschke, A; Sauter, O; Medvedev, S Yu

2010-01-01

An ELMy H-mode 'snowflake' (SF) divertor is established and studied for the first time in the TCV tokamak. The H-mode access and the edge localized mode (ELM) dynamics are compared with a conventional single-null diverted configuration. The SF configuration exhibits 15% higher confinement and 2-3 times lower ELM frequency. Ideal MHD stability analysis suggests enhanced stability of the SF H-mode pedestal to mid- to high-toroidal-mode-number modes. The capability of the SF to redistribute the edge power on the additional strike points has been confirmed experimentally.

Characteristics of toroidal energy deposition asymmetries in ASDEX

International Nuclear Information System (INIS)

Evans, T.E.; Neuhauser, J.; Leuterer, F.; Mueller, E.R.

1990-01-01

Large toroidal and poloidal asymmetries with characteristics which are sensitively dependent on q a , the vertical position of the plasma, and the type of additional heating are observed in the energy flow to the ASDEX divertor target plates. The largest asymmetries and total energy depositions are observed during lower hybrid wave injection experiments with approximately 50% of the input energy going to the combined divertor targets and shields. A maximum localized energy density loading of 10 MJ/m 2 is typical under these conditions. Measurements of the asymmetries are consistent with a model in which magnetic islands and ergodicity due to intrinsic magnetic perturbations dominate the energy transpot across the primary magnetic separatrix. The results emphasize the essential role of resonant magnetic perturbations in determining the performance of tokamaks and demonstrate that non-axisymmetric effects caused by small perturbations become increasingly important in determining the transport properties as the injected power is increased. (orig.)

ATHENA calculation model for the ITER-FEAT divertor cooling system. Final report with updates

International Nuclear Information System (INIS)

Eriksson, John; Sjoeberg, A.; Sponton, L.L.

2001-05-01

An ATHENA model of the ITER-FEAT divertor cooling system has been developed for the purpose of calculating and evaluating consequences of different thermal-hydraulic accidents as specified in the Accident Analysis Specifications for the ITER-FEAT Generic Site Safety Report. The model is able to assess situations for a variety of conceivable operational transients from small flow disturbances to more critical conditions such as total blackout caused by a loss of offsite and emergency power. The main objective for analyzing this type of scenarios is to determine margins against jeopardizing the integrity of the divertor cooling system components and pipings. The model of the divertor primary heat transport system encompasses the divertor cassettes, the port limiter systems, the pressurizer, the heat exchanger and all feed and return pipes of these components. The development was pursued according to practices and procedures outlined in the ATHENA code manuals using available modelling components such as volumes, junctions, heat structures and process controls

Heat flux management via advanced magnetic divertor configurations and divertor detachment

Energy Technology Data Exchange (ETDEWEB)

Kolemen, E., E-mail: ekolemen@princeton.edu [Princeton University, Princeton, NJ 08544 (United States); Allen, S.L. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bray, B.D. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Fenstermacher, M.E. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Humphreys, D.A.; Hyatt, A.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Lasnier, C.J. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Leonard, A.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Makowski, M.A.; McLean, A.G. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Maingi, R.; Nazikian, R. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Petrie, T.W. [General Atomics, PO Box 85608, San Diego, CA 92186-5608 (United States); Soukhanovskii, V.A. [Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Unterberg, E.A. [Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831 (United States)

2015-08-15

The snowflake divertor (SFD) control and detachment control to manage the heat flux at the divertor are successfully demonstrated at DIII-D. Results of the development and implementation of these two heat flux reduction control methods are presented. The SFD control algorithm calculates the position of the two null-points in real-time and controls shaping coil currents to achieve and stabilize various snowflake configurations. Detachment control stabilizes the detachment front fixed at specified distance between the strike point and the X-point throughout the shot.

Comparison between stellarator and tokamak divertor transport

International Nuclear Information System (INIS)

Feng, Y.; Lunt, T.; Kobayashi, M.; Reiter, D.

2010-11-01

The paper compares the essential divertor transport features of the poloidal divertor, which is well-developed for tokamaks, and the non-axisymmetric divertors currently investigated on helical devices. It aims at surveying the fundamental similarities and differences in divertor concept and geometry, and their consequences for how the divertor functions. In particular, the importance of various transport terms governing axisymmetric and helical scrape-off-layers (SOLs) is examined, with special attention being paid to energy, momentum and impurity transport. Tokamak and stellarator SOLs are compared by identifying key geometric parameters through which the governing physics can be illustrated by simple models and estimates. More quantitative assessments rely nevertheless on the modeling using EMC3-EIRENE code. Most of the theoretical results are discussed in conjunction with experimental observations. (author)

The divertor remote maintenance project

International Nuclear Information System (INIS)

Maisonnier, D.; Martin, E.; Akou, K.

2001-01-01

Remote replacement of the ITER divertor will be required several times during the life of ITER. To facilitate its regular exchange, the divertor is assembled in the ITER vacuum vessel from 60 cassettes. Radial movers transport each cassette along radial rails through the handling ports and into the vessel where a toroidal mover lifts and transports the cassette around a pair of toroidal rails. Once at its final position the cassette is locked to the toroidal rails and is accurately aligned in both poloidal and toroidal directions. A further requirement on the divertor is to minimise the amount of activated waste to be sent to a repository. To this end the cassettes have been designed to allow the remote replacement, in a hot cell, of their plasma facing components. The paper describes the two facilities built at ENEA Brasimone, Italy, whose aim is to demonstrate the reliable remote maintenance of the divertor cassettes. (author)

The divertor remote maintenance project

International Nuclear Information System (INIS)

Maisonnier, D.; Martin, E.; Akou, K.

1999-01-01

Remote replacement of the ITER divertor will be required several times during the life of ITER. To facilitate its regular exchange, the divertor is assembled in the ITER vacuum vessel from 60 cassettes. Radial movers transport each cassette along radial rails through the handling ports and into the vessel where a toroidal mover lifts and transports the cassette around a pair of toroidal rails. Once at its final position the cassette is locked to the toroidal rails and is accurately aligned in both poloidal and toroidal directions. A further requirement on the divertor is to minimise the amount of activated waste to be sent to a repository. To this end the cassettes have been designed to allow the remote replacement, in a hot cell, of their plasma facing components. The paper describes the two facilities built at ENEA Brasimone, Italy, whose aim is to demonstrate the reliable remote maintenance of the divertor cassettes. (author)

SOLPS simulations of X-divertor in NSTX-U

Science.gov (United States)

Chen, Zhongping; Kotschenreuther, Mike; Mahajan, Swadesh

2017-10-01

The X-divertor (XD) geometry in NSTX-U has demonstrated, in SOLPS simulations, a better performance than the standard divertor (SD) regarding detachment: achieving detachment with a lower upstream density and stabilizing the detachment front near the target. The benefits of such a localized front is that the power exhaust requirement can be satisfied without the radiation front encroaching on the core plasma. It is also found by our simulations that at similar states of detachment the XD outperforms the SD by reducing the heat fluxes to the target and maintaining higher upstream temperatures. These advantages are attributed to the unique geometric characteristics of XD - poloidal flaring near the target. The detailed physical mechanisms behind the better XD performance that is found in the simulations will be examined. Work supported by US DOE under DE-FG02-04ER54742 and SC 0012956.

Tungsten nano-tendril growth in the Alcator C-Mod divertor

International Nuclear Information System (INIS)

Wright, G.M.; Brunner, D.; Labombard, B.; Lipschultz, B.; Terry, J.L.; Whyte, D.G.; Baldwin, M.J.; Doerner, R.P.

2012-01-01

Growth of tungsten nano-tendrils (‘fuzz’) has been observed for the first time in the divertor region of a high-power density tokamak experiment. After 14 consecutive helium L-mode discharges in Alcator C-Mod, the tip of a tungsten Langmuir probe at the outer strike point was fully covered with a layer of nano-tendrils. The thickness of the individual nano-tendrils (50–100 nm) and the depth of the layer (600 ± 150 nm) are consistent with observations from experiments on linear plasma devices. The observation of tungsten fuzz in a tokamak may have important implications for material erosion, dust formation, divertor lifetime and tokamak operations in next-step devices. (letter)

Deuterium depth profiling in JT-60U W-shaped divertor tiles by nuclear reaction analysis

International Nuclear Information System (INIS)

Hayashi, T.; Ochiai, K.; Masaki, K.; Gotoh, Y.; Kutsukake, C.; Arai, T.; Nishitani, T.; Miya, N.

2006-01-01

Deuterium concentrations and depth profiles in plasma-facing graphite tiles used in the divertor of JAERI Tokamak-60 Upgrade (JT-60U) were investigated by nuclear reaction analysis (NRA). The highest deuterium concentration of D/ 12 C of 0.053 was found in the outer dome wing tile, where the deuterium accumulated probably through the deuterium-carbon co-deposition. In the outer and inner divertor target tiles, the D/ 12 C data were lower than 0.006. Additionally, the maximum (H + D)/ 12 C in the dome top tile was estimated to be 0.023 from the results of NRA and secondary ion mass spectroscopy (SIMS). Orbit following Monte-Carlo (OFMC) simulation showed energetic deuterons caused by neutral beam injections (NBI) were implanted into the dome region with high heat flux. Furthermore, the surface temperature and conditions such as deposition and erosion significantly influenced the accumulation process of deuterium. The deuterium depth profile, scanning electron microscope (SEM) observation and OFMC simulation indicated the deuterium was considered to accumulate through three processes: the deuterium-carbon co-deposition, the implantation of energetic deuterons and the deuterium diffusion into the bulk

Power balance for impurity seeded ergodic divertor discharges in Tore Supra

Energy Technology Data Exchange (ETDEWEB)

Reichle, R.; Vallet, J.C.; Basiuk, V.; Chantant, M.; Giannella, R.; Guirlet, R.; Mitteau, R

1999-10-15

The main emphasis of the present contribution is the search for independent indications concerning local radiation enhancement in the vicinity of target plates by the application of the power-balance analysis to impurity seeded discharges and the use of spatially resolved calorimetry. All methods confirm the existence of an important local radiation enhancement in front of target objects. Their torlly localised nature and the lack of short scale toroidal resolution of the bolometers cause an underestimate of the radiation. Observed corrective factors for radiative divertor discharges range between 1.1 and 1.6 depending on the density. the impurity seeded discharges show during the impurity injection a reduction of this factor which can be explained by the radiation distribution becoming toroidally more symmetric due to larger ionisation lengths. The consistency of the results seems to validate the assumption of a fall off length of the radiation of 40 cm as used for the radiation extrapolation from the outboard calorimetry. The localised extra radiation falling onto one individual neutralizer is about 1 kW. This should be compared with spectroscopic radiation measurements and modelling to see whether there is room for charge exchange neutrals to be important. (authors)

Power balance for impurity seeded ergodic divertor discharges in Tore Supra

International Nuclear Information System (INIS)

Reichle, R.; Vallet, J.C.; Basiuk, V.; Chantant, M.; Giannella, R.; Guirlet, R.; Mitteau, R.

1999-01-01

The main emphasis of the present contribution is the search for independent indications concerning local radiation enhancement in the vicinity of target plates by the application of the power-balance analysis to impurity seeded discharges and the use of spatially resolved calorimetry. All methods confirm the existence of an important local radiation enhancement in front of target objects. Their toroidally localised nature and the lack of short scale toroidal resolution of the bolometers cause an underestimate of the radiation. Observed corrective factors for radiative divertor discharges range between 1.1 and 1.6 depending on the density. the impurity seeded discharges show during the impurity injection a reduction of this factor which can be explained by the radiation distribution becoming toroidally more symmetric due to larger ionisation lengths. The consistency of the results seems to validate the assumption of a fall off length of the radiation of 40 cm as used for the radiation extrapolation from the outboard calorimetry. The localised extra radiation falling onto one individual neutralizer is about 1 kW. This should be compared with spectroscopic radiation measurements and modelling to see whether there is room for charge exchange neutrals to be important. (authors)

Hybrid formulation of radiation transport in optically thick divertor plasmas

Energy Technology Data Exchange (ETDEWEB)

Rosato, J.; Marandet, Y.; Bufferand, H.; Stamm, R. [PIIM, UMR 7345 Aix-Marseille Universite / CNRS, Centre de St-Jerome, Marseille (France); Reiter, D. [IEK-4 Plasmaphysik, Forschungszentrum Juelich GmbH, Juelich (Germany)

2016-08-15

Kinetic Monte Carlo simulations of coupled atom-radiation transport in optically thick divertor plasmas can be computationally very demanding, in particular in ITER relevant conditions or even larger devices, e.g. for power plant divertor studies. At high (∝ 10{sup 15} cm{sup -3}) atomic densities, it can be shown that sufficiently large divertors behave in certain areas like a black body near the first resonance line of hydrogen (Lyman α). This suggests that, at least in part, the use of continuum model (radiation hydrodynamics) can be sufficiently accurate, while being less time consuming. In this work, we report on the development of a hybrid model devoted to switch automatically between a kinetic and a continuum description according to the plasma conditions. Calculations of the photo-excitation rate in a homogeneous slab are performed as an illustration. The outlined hybrid concept might be also applicable to neutral atom transport, due to mathematical analogy of transport equations for neutrals and radiation. (copyright 2016 The Authors. Contributions to Plasma Physics published by Wiley-VCH Verlag GmbH and Co. KGaA Weinheim. This)

'EU divertor celebration day'

International Nuclear Information System (INIS)

Merola, M.

2002-01-01

The meeting 'EU divertor celebration day' organized on 16 January 2002 at Plansee AG, Reutte, Austria was held on the occasion of the completion of manufacturing activities of a complete set of near full-scale prototypes of divertor components including the vertical target, the dome liner and the cassette body. About 30 participants attended the meeting including Dr. Robert Aymar, ITER Director, representatives from EFDA, CEA, ENEA, IPP and others

ITER tungsten divertor design development and qualification program

Energy Technology Data Exchange (ETDEWEB)

Hirai, T., E-mail: takeshi.hirai@iter.org [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Escourbiac, F.; Carpentier-Chouchana, S.; Fedosov, A.; Ferrand, L.; Jokinen, T.; Komarov, V.; Kukushkin, A.; Merola, M.; Mitteau, R.; Pitts, R.A.; Shu, W.; Sugihara, M. [ITER Organization, Route de Vinon sur Verdon, F-13115 Saint Paul lez Durance (France); Riccardi, B. [F4E, c/ Josep Pla, n.2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Suzuki, S. [JAEA, Fusion Research and Development Directorate JAEA, 801-1 Mukouyama, Naka, Ibaragi 311-0193 (Japan); Villari, R. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, I-00044 Frascati, Rome (Italy)

2013-10-15

Highlights: • Detailed design development plan for the ITER tungsten divertor. • Latest status of the ITER tungsten divertor design. • Brief overview of qualification program for the ITER tungsten divertor and status of R and D activity. -- Abstract: In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R and D activity is summarized in this paper.

High temperature divertor plasma operation

International Nuclear Information System (INIS)

Ohyabu, Nobuyoshi.

1991-02-01

High temperature divertor plasma operation has been proposed, which is expected to enhance the core energy confinement and eliminates the heat removal problem. In this approach, the heat flux is guided through divertor channel to a remote area with a large target surface, resulting in low heat load on the target plate. This allows pumping of the particles escaping from the core and hence maintaining of the high divertor temperature, which is comparable to the core temperature. The energy confinement is then determined by the diffusion coefficient of the core plasma, which has been observed to be much lower than the thermal diffusivity. (author)

Implications of steady-state operation on divertor design

International Nuclear Information System (INIS)

Sevier, D.L.; Reis, E.E.; Baxi, C.B.; Silke, G.W.; Wong, C.P.C.; Hill, D.N.

1996-01-01

As fusion experiments progress towards long pulse or steady state operation, plasma facing components are undergoing a significant change in their design. This change represents the transition from inertially cooled pulsed systems to steady state designs of significant power handling capacity. A limited number of Plasma Facing Component (PFC) systems are in operation or planning to address this steady state challenge at low heat flux. However in most divertor designs components are required to operate at heat fluxes at 5 MW/m 2 or above. The need for data in this area has resulted in a significant amount of thermal/hydraulic and thermal fatigue testing being done on prototypical elements. Short pulse design solutions are not adequate for longer pulse experiments and the areas of thermal design, structural design, material selection, maintainability, and lifetime prediction are undergoing significant changes. A prudent engineering approach will guide us through the transitional phase of divertor design to steady-state power plant components. This paper reviews the design implications in this transition to steady state machines and the status of the community efforts to meet evolving design requirements. 54 refs., 5 figs., 2 tabs

NSTX plasma operation with a Liquid Lithium Divertor

Energy Technology Data Exchange (ETDEWEB)

Kugel, H.W., E-mail: hkugel@pppl.gov [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Allain, J.P. [Purdue University, West Lafayette, IN 47907 (United States); Bell, M.G.; Bell, R.E.; Diallo, A.; Ellis, R.; Gerhardt, S.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Heim, B. [Purdue University, West Lafayette, IN 47907 (United States); Jaworski, M.A.; Kaita, R.; Kallman, J.; Kaye, S.; LeBlanc, B.P. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Maingi, R.; McLean, A. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Menard, J.; Mueller, D. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Nygren, R. [Sandia National Laboratories, Albuquerque, NM 87185 (United States); Ono, M.; Paul, S.F. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); and others

2012-10-15

Highlights: Black-Right-Pointing-Pointer NSTX 2010 experiments tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium molybdenum divertor surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. Black-Right-Pointing-Pointer Noteworthy improvements in plasma performance with the plasma strike point on the liquid lithium molybdenum divertor were obtained similar to those obtained previously with lithiated graphite. The role of lithium impurities in this result is discussed. Black-Right-Pointing-Pointer Inspection of the liquid lithium molybdenum divertor after the Campaign indicated mechanical damage to supports, and other hardware resulting from forces following plasma current disruptions. - Abstract: NSTX 2010 experiments were conducted using a molybdenum Liquid Lithium Divertor (LLD) surface installed on the outer part of the lower divertor. This tested the effectiveness of maintaining the deuterium retention properties of a static liquid lithium surface when refreshed by lithium evaporation as an approximation to a flowing liquid lithium surface. The LLD molybdenum front face has a 45% porosity to provide sufficient wetting to spread 37 g of lithium, and to retain it in the presence of magnetic forces. Lithium Evaporators were used to deposit lithium on the LLD surface. At the beginning of discharges, the LLD lithium surface ranged from solid to liquefied depending on the amount of applied and plasma heating. Noteworthy improvements in plasma performance were obtained similar to those obtained previously with lithiated graphite, e.g., ELM-free, quiescent edge, H-modes. During these experiments with the plasma outer strike point on the LLD, the rate of deuterium retention in the LLD, as indicated by the fueling needed to achieve and maintain stable plasma conditions, was the about the same as that for solid lithium coatings on the graphite prior to the installation of the

Divertor plasma studies on DIII-D: Experiment and modeling

International Nuclear Information System (INIS)

West, W.P.; Brooks, N.H.; Allen, S.L.

1996-09-01

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma provides separation between the first wall and the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it, the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma wall interaction. In this paper the authors describe a set of experiments on DIII-D designed to provide detailed two dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. They also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modeling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process

The design of a new JET divertor for high triangularity and high current scenarios

Energy Technology Data Exchange (ETDEWEB)

Chappuis, P. E-mail: philippe.chappuis@cea.fr; Damiani, C.; Guerin, C.; Hurd, F.; Loarte, A.; Lomas, P.; Lorenz, A.; Pamela, J.; Peacock, A.; Portafaix, C.; Rapp, J.; Riccardo, V.; Rimini, F.; Saibene, G.; Salavy, J.F.; Sauce, Y.; Sartori, R.; Solano, E.; Thomas, E.; Thomas, P.; Tsitrone, E.; Valeta, M.P

2003-09-01

A new divertor (MKII-HP) has been designed to be implemented in JET as part of a possible enhancement programme of the JET facility (JET EP). The aim is to handle up to 40 MW of injected power for 10 s with plasma triangularities up to 0.5 while keeping enough flexibility for other scenarios. The divertor is shaped to optimise the wetting fraction without exposing sharp edges or metallic parts and the general design allows for high halo currents.

Heat and particle transport of sol/divertor plasma in the W-shaped divertor on JT-60U

International Nuclear Information System (INIS)

Asakura, N.; Sakurai, S.; Hosogane, N.

1999-01-01

The plasma profile and parallel flow in the scrape-off layer (SOL) were systematically measured using Mach probes installed at the midplane and the divertor x-point. Quantitative evaluation of a parallel flow: naturally produced in a torus to keep the pressure constant along the field line, was consistent with the measurement. Geometry effects of the W-shaped divertor on the divertor plasma and particle recycling at the newly installed baffle plates were evaluated quantitatively using the edge plasma data. (author)

Power deposition for ion cyclotron heating in large tokamaks

International Nuclear Information System (INIS)

Hellsten, T.; Villard, L.

1988-01-01

The power deposition profiles during minority ion cyclotron heating are analysed in large tokamaks by using the global, toroidal wave code LION. For tokamaks with large aspect ratio and with circular cross-section, the wave is focused on the magnetic axis and can be absorbed there by cyclotron absorption when the cyclotron resonance passes through the magnetic axis. The power deposition profile is then essentially determined by the Doppler broadening of the ion cyclotron resonance. For equilibria either non-circular or with a small aspect ratio the power deposition profile depends also on the strength of the damping. In this case the power deposition profile can be expressed as a sum of two power deposition profiles. One is related to the power absorbed in a single pass, and its shape is similar to that obtained for large aspect ratio and circular cross-section. The other profile is obtained by calculating the power deposition in the limit of weak damping, in which case the wave electric field is almost constant along the cyclotron resonance layer. A heuristic formula for the power deposition is given. The formula includes a number of calibration curves and functions which has been calculated with the LION code for JET relevant equilibria. The formula enables calculation of the power deposition profile in a simple way when the launched wave spectrum and damping coefficients are known. (author). 7 refs, 11 figs

Divertors for helical devices: Concepts, plans, results and problems

International Nuclear Information System (INIS)

Koenig, R.; Grigull, P.; McCormick, K.

2003-01-01

With LHD and W7-X stellarator development is now taking a large leap forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control, and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large stellarators were carefully prepared in smaller scale devices like Heliotron E, CHS and W7-AS. While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller scale experiments like Heliotron-J, CHS and NCSX will be used for the further development of divertor concepts. The two divertor configurations that are presently being investigated, are the helical and the island divertor, as well as the local island divertor (LID), which was successfully demonstrated on CHS and just went into operation on LHD. Presently, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor which will allow quasi continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi steady-state operating scenario in a newly found high density H-mode operating regime, which benefits from high energy and extremely low impurity confinement times, with edge radiation levels of up to 90 % and sufficient neutral compression in the subdivertor region (> 10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios and toroidal asymmetries due to symmetry breaking error fields, etc. will be discussed. (orig.)

Divertors for Helical Devices: Concepts, Plans, Results, and Problems

International Nuclear Information System (INIS)

Koenig, R.; Grigull, P.; McCormick, K.

2004-01-01

With Large Helical Device (LHD) and Wendelstein 7-X (W7-X), the development of helical devices is now taking a large step forward on the path to a steady-state fusion reactor. Important issues that need to be settled in these machines are particle flux and heat control and the impact of divertors on plasma performance in future continuously burning fusion plasmas. The divertor concepts that will initially be explored in these large machines were prepared in smaller-scale devices like Heliotron E, Compact Helical System (CHS), and Wendelstein 7-AS (W7-AS). While advanced divertor scenarios relevant for W7-X were already studied in W7-AS, other smaller-scale experiments like Heliotron-J, CHS, and National Compact Stellarator Experiment will be used for the further development of divertor concepts. The two divertor configurations that are being investigated are the helical and the island divertor, as well as the local island divertor, which was successfully demonstrated on CHS and just went into operation on LHD. At present, on its route to a fully closed helical divertor, LHD operates in an open helical divertor configuration. W7-X will be equipped right from the start with an actively cooled discrete island divertor that will allow quasi-continuous operation. The divertor design is very similar to the one explored on W7-AS. For sufficiently large island sizes and not too long field line connection lengths, this divertor gives access to a partially detached quasi-steady-state operating scenario in a newly found high-density H-mode operating regime, which benefits from high energy and low impurity confinement times, with edge radiation levels of up to 90% and sufficient neutral compression in the subdivertor region (>10) for active pumping. The basic physics of the different divertor concepts and associated implementation problems, like asymmetries due to drifts, accessibility of essential operating scenarios, toroidal asymmetries due to symmetry breaking error fields

Physical study of experimental fusion breeder FEB divertor

International Nuclear Information System (INIS)

Zhu Yukun; Zhou Xiaobing; Huang Jinhua; Feng Kaiming; Deng Peizhi; Huo Tiejun

1999-10-01

The physical study of FEB divertor is presented. In order to improve the impurity control and increase ion-neutral interactions in the divertor, the configuration of the divertor is optimized to be the close type in the engineering design activity compared with the open type in the early conceptual activity. The operation mode of the divertor is designed to be partial detached plasma mode under conditions of combination gas-puffing with impurity injection. The position of gas-puffing is optimized to be at the torus mid-plane with NEWT1D code from the viewpoint of impurity retention and radiation in the scrape-off layer/divertor region. Boron is chosen as the injected impurity. The effect of boron impurity injection is evaluated from the reduced heat load on the divertor target. The plasma pressure drop along the scrape-off layer/divertor region is estimated with the two-point transport model and impurity radiation model in the dynamic gas target concept. The simulation results show that the plasma pressure drop factor f p is not only related to the radiation fraction f rad but also related greatly to the stagnation point density n s

Multi-Fluid Modeling of Low-Recycling Divertor Regimes

International Nuclear Information System (INIS)

Smirnov, R.D.; Pigarov, A.Y.; Krasheninnikov, S.I.; Rognlien, T.D.; Soukhanovskii, V.A.; Rensink, M.E.; Maingi, R.; Skinner, C.H.; Stotler, D.P.; Bell, R.E.; Kugel, H.W.

2010-01-01

The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate.

Development of a radiative divertor for DIII-D

International Nuclear Information System (INIS)

Allen, S.L.; Brooks, N.H.; Campbell, R.B.; Fenstermacher, M.E.; Hill, D.N.; Hyatt, A.W.; Knoll, D.; Lasnier, C.J.; Lazarus, E.A.; Leonard, A.W.; Lippmann, S.I.; Mahdavi, M.A.; Maingi, R.; Meyer, W.; Moyer, R.A.; Petrie, T.W.; Porter, G.D.; Rensink, M.E.; Rognlien, T.D.; Schaffer, M.J.; Smith, J.P.; Staebler, G.M.; Stambaugh, R.D.; West, W.P.; Wood, R.D.

1995-01-01

We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while τ E remains similar 2 times ITER-89P scaling. However, n e increases with D 2 puffing, and Z eff increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity (δ∼0.8) is important for high τ E VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.))

Development of a radiative divertor for DIII-D

Energy Technology Data Exchange (ETDEWEB)

Allen, S.L. [Lawrence Livermore National Lab., CA (United States); Brooks, N.H. [General Atomics, San Diego, CA (United States); Campbell, R.B. [Sandia National Labs., Albuquerque, NM (United States); Fenstermacher, M.E. [Lawrence Livermore National Lab., CA (United States); Hill, D.N. [Lawrence Livermore National Lab., CA (United States); Hyatt, A.W. [General Atomics, San Diego, CA (United States); Knoll, D.; Lasnier, C.J. [Lawrence Livermore National Lab., CA (United States); Lazarus, E.A. [Oak Ridge National Lab., TN (United States); Leonard, A.W. [General Atomics, San Diego, CA (United States); Lippmann, S.I. [General Atomics, San Diego, CA (United States); Mahdavi, M.A. [General Atomics, San Diego, CA (United States); Maingi, R. [Oak Ridge National Lab., TN (United States); Meyer, W. [Lawrence Livermore National Lab., CA (United States); Moyer, R.A. [California Univ., Los Angeles, CA (United States); Petrie, T.W. [General Atomics, San Diego, CA (United States); Porter, G.D. [Lawrence Livermore National Lab., CA (United States); Rensink, M.E. [Lawrence Livermore National Lab., CA (United States); Rognlien, T.D. [Lawrence Livermore National Lab., CA (United States); Schaffer, M.J. [General Atomics, San Diego, CA (United States); Smith, J.P. [General Atomics, San Diego, CA (United States); Staebler, G.M. [General Atomics, San Diego, CA (United States); Stambaugh, R.D. [General Atomics, San Diego, CA (United States); West, W.P. [General Atomics, San Diego, CA (United States); Wood, R.D. [Lawrence Livermore National Lab., CA (United States)

1995-04-01

We have used experiments and modeling to develop a new radiative divertor configuration for DIII-D. Gas puffing experiments with the existing open divertor have shown the creation of a localized ( similar 10 cm diameter) radiation zone which results in substantial reduction (3-10) in the divertor heat flux while {tau}{sub E} remains similar 2 times ITER-89P scaling. However, n{sub e} increases with D{sub 2} puffing, and Z{sub eff} increases with neon puffing. Divertor structures are required to minimize the effects on the core plasma. The UEDGE fluid code, benchmarked with DIII-D data, and the DEGAS neutrals transport code are used to estimate the effectiveness of divertor configurations; slots reduce the core ionization more than baffles. The overall divertor shape is set by confinement studies which indicate that high triangularity ({delta}{approx}0.8) is important for high {tau}{sub E} VH-modes. Results from engineering feasibility studies, including diagnostic access, will be presented. ((orig.)).

Is Carbon a Realistic Choice for ITER's Divertor?

International Nuclear Information System (INIS)

Skinner, C.H.; Federici, G.

2005-01-01

Tritium retention by co-deposition with carbon on the divertor target plate is predicted to limit ITER's DT burning plasma operations (e.g. to about 100 pulses for the worst conditions) before the in-vessel tritium inventory limit, currently set at 350 g, is reached. At this point, ITER will only be able to continue its burning plasma program if technology is available that is capable of rapidly removing large quantities of tritium from the vessel with over 90% efficiency. The removal rate required is four orders of magnitude faster than that demonstrated in current tokamaks. Eighteen years after the observation of co-deposition on JET and TFTR, such technology is nowhere in sight. The inexorable conclusion is that either a major initiative in tritium removal should be funded or that research priorities for ITER should focus on metal alternatives

Divertor erosion in DIII-D

International Nuclear Information System (INIS)

Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.

1998-05-01

Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point of two divertor plasma conditions: (1) attached (Te > 40 eV) ELMing plasmas and (2) detached (Te 10 cm/year, even with incident heat flux 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood and that effective sputtering yields are > 10%. In ELM-free discharges, this erosion rate can account for the rate of carbon accumulation in the core plasma. Divertor plasma detachment eliminates physical sputtering, while spectroscopically measured chemical erosion yields are also found to be low (Y(C/D + ) ≤ 2.0 x 10 -3 ). This leads to suppression of net erosion at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates (∼ 10 microm/s) at the OSP of an attached plasma. Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

Modeling of thermal effects on TIBER II [Tokamak Ignition/Burn Experimental Reactor] divertor during plasma disruption

International Nuclear Information System (INIS)

Bruhn, M.L.; Perkins, L.J.

1987-01-01

Mapping the disruption power flow from the mid-plane of the TIBER Engineering Test Reactor to its divertor and calculating the resulting thermal effects are accomplished through the modification and coupling of three presently existing computer codes. The resulting computer code TADDPAK (Thermal Analysis Divertor during Disruption PAcKage) provides three-dimensional graphic presentations of time and positional dependent thermal effects on a poloidal cross section of the double-null-divertor configured reactor. These thermal effects include incident heat flux, surface temperature, vaporization rate, total vaporization, and melting depth. The dependence of these thermal effects on material choice, disruption pulse shape, and the characteristic thickness of the plasma scrape-off layer is determined through parametric analysis with TADDPAK. This computer code is designed to be a convenient, rapid, and user-friendly modeling tool which can be easily adapted to most tokamak double-null-divertor reactor designs. 14 refs

Divertor plasma physics experiments on the DIII-D tokamak

International Nuclear Information System (INIS)

Mahdavi, M.A.; Allen, S.L.; Evans, T.E.

1996-10-01

In this paper we present an overview of the results and conclusions of our most recent divertor physics and development work. Using an array of new divertor diagnostics we have measured the plasma parameters over the entire divertor volume and gained new insights into several divertor physics issues. We present direct experimental evidence for momentum loss along the field lines, large heat convection, and copious volume recombination during detachment. These observations are supported by improved UEDGE modeling incorporating impurity radiation. We have demonstrated divertor exhaust enrichment of neon and argon by action of a forced scrape off layer (SOL) flow and demonstrated divertor pumping as a substitute for conventional wall conditioning. We have observed a divertor radiation zone with a parallel extent that is an order of magnitude larger than that estimated from a 1-D conduction limited model of plasma at coronal equilibrium. Using density profile control by divertor pumping and pellet injection we have attained H-mode confinement at densities above the Greenwald limit. Erosion rates of several candidate ITER plasma facing materials are measured and compared with predictions of a numerical model

DC switch power supply for vacuum-arc coatings deposition

International Nuclear Information System (INIS)

Zalesskij, D.Yu.; Volkov, Yu.Ya.; Vasil'ev, V.V.; Kozhushko, V.V.; Luchaninov, A.A.; Strel'nitskij, V.E.

2008-01-01

Special DC Switch Power Supply for vacuum-arc deposition was developed and tested in the mode of depositing Al and AlN films. Maximum output power was 6 kW, maximum output current - 120 A, open-circuit voltage - 150 V. The Power Supply allows to adjust and stabilize output current in a wide range. Testing of the Power Supply revealed an advantages over the standard 'Bulat-6' power supply, especially for deposition of non-conductive AlN films.

Two-dimensional divertor modeling and scaling laws

International Nuclear Information System (INIS)

Catto, P.J.; Connor, J.W.; Knoll, D.A.

1996-01-01

Two-dimensional numerical models of divertors contain large numbers of dimensionless parameters that must be varied to investigate all operating regimes of interest. To simplify the task and gain insight into divertor operation, we employ similarity techniques to investigate whether model systems of equations plus boundary conditions in the steady state admit scaling transformations that lead to useful divertor similarity scaling laws. A short mean free path neutral-plasma model of the divertor region below the x-point is adopted in which all perpendicular transport is due to the neutrals. We illustrate how the results can be used to benchmark large computer simulations by employing a modified version of UEDGE which contains a neutral fluid model. (orig.)

Divertor cassette movers prototypes for ITER

International Nuclear Information System (INIS)

Bogusch, E.; Batz, R.; Bieber, O.; Gottfried, R.; Cerdan, G.

1998-01-01

Following competitive tendering, in October 1996 Siemens was contracted by the European Commission to design and supply an assembly of four Divertor Cassette Movers Prototypes including the control and command systems for the movers proper. The assembly consisting of one Cassette Toroidal Mover (CTM), one Radial Mover Tractor (TRC), one Second Cassette Carrier (SCC), and one Radial Cassette Carrier (RCC) represents key components of the Divertor Test Platform at Brasimone, one of the seven large R+D projects for ITER. By detailed design, high-precision manufacturing and testing of these devices, Siemens contributed to the verification of an important task within the European R and D program towards ITER construction. Replacement of the divertor cassettes is a scheduled maintenance operation throughout the life of ITER. The successful fabrication and testing of the Divertor Cassette Movers Prototypes is all important milestone to verify this delicate operation. (authors)

ELM induced divertor heat loads on TCV

Energy Technology Data Exchange (ETDEWEB)

Marki, J., E-mail: janos.marki@epfl.c [Centre de Recherches en Physique des Plasmas (CRPP), Ecole Polytechnique Federale de Lausanne (EPFL), Association Euratom - Confederation Suisse, CH-1015 Lausanne (Switzerland); Pitts, R.A. [Centre de Recherches en Physique des Plasmas (CRPP), Ecole Polytechnique Federale de Lausanne (EPFL), Association Euratom - Confederation Suisse, CH-1015 Lausanne (Switzerland); Horacek, J. [Institute of Plasma Physics, Association EUROATOM-IPP.CR, Za Slovankou 3, 182 00 Prague 8 (Czech Republic); Tskhakaya, D. [Association EURATOM-OAW, Institut fuer Theoretische Physik, A-6020 Innsbruck (Austria)

2009-06-15

Results are presented for heat loads at the TCV outer divertor target during ELMing H-mode using a fast IR camera. Benefitting from a recent surface cleaning of the entire first wall graphite armour, a comparison of the transient thermal response of freshly cleaned and untreated tile surfaces (coated with thick co-deposited layers) has been performed. The latter routinely exhibit temperature transients exceeding those of the clean ones by a factor approx3, even if co-deposition throughout the first days of operation following the cleaning process leads to the steady regrowth of thin layers. Filaments are occasionally observed during the ELM heat flux rise phase, showing a spatial structure consistent with energy release at discrete toroidal locations in the outer midplane vicinity and with individual filaments carrying approx1% of the total ELM energy. The temporal waveform of the ELM heat load is found to be in good agreement with the collisionless free streaming particle model.

ELM induced divertor heat loads on TCV

Science.gov (United States)

Marki, J.; Pitts, R. A.; Horacek, J.; Tskhakaya, D.; TCV Team

2009-06-01

Results are presented for heat loads at the TCV outer divertor target during ELMing H-mode using a fast IR camera. Benefitting from a recent surface cleaning of the entire first wall graphite armour, a comparison of the transient thermal response of freshly cleaned and untreated tile surfaces (coated with thick co-deposited layers) has been performed. The latter routinely exhibit temperature transients exceeding those of the clean ones by a factor ˜3, even if co-deposition throughout the first days of operation following the cleaning process leads to the steady regrowth of thin layers. Filaments are occasionally observed during the ELM heat flux rise phase, showing a spatial structure consistent with energy release at discrete toroidal locations in the outer midplane vicinity and with individual filaments carrying ˜1% of the total ELM energy. The temporal waveform of the ELM heat load is found to be in good agreement with the collisionless free streaming particle model.

A mechanism for large divertor plasma energy loss via lithium radiation in tokamaks

Science.gov (United States)

Rognlien, T. D.; Meier, E. T.; Soukhanovskii, V. A.

2012-10-01

Lithium has been used as a wall-conditioning element in a number of tokamaks over the years, including TFTR, FTU, and NSTX, where core plasma energy confinement and particle control are often found to improve following such conditioning. Here the possible role of Li in providing substantial energy loss for divertor plasmas via line radiation is reported. A multi-charge-state 2D UEDGE fluid model is used where the hydrogenic and Li ions and neutrals are each evolved as separate species and separate equations are solved for the electron and ion temperatures. It is shown that a sufficient level of Li neutrals evolving from the divertor surface via sputtering or evaporation can induce energy detachment of the divertor plasma, yielding a strongly radiating zone near the divertor where ionization and recombination from/to neutral Li can radiate most of the power flowing into the scrape-off layer while maintaining low core contamination. A local peaking of Li emissivity for electron temperatures near 1 eV appears to play an important role in the detachment of the mixed deuterium/Li plasma. Evidence of such behavior from NSTX discharges will be discussed.

Effects of low and high mode number tearing modes in divertor tokamaks

International Nuclear Information System (INIS)

Punjabi, Alkesh; Ali, Halima; Boozer, Allen; Evans, Todd

2007-01-01

The topological effects of magnetic perturbations on a divertor tokamak, such as DIII-D, are studied using field-line maps that were developed by Punjabi et al. [A. Punjabi, A. Verma, and A. Boozer, Phys. Rev. Lett. 69, 3322 (1992)]. The studies consider both long-wavelength perturbations, such as those of m=1, n=1 tearing modes, and localized perturbations, which are represented as a magnetic dipole. The parameters of the dipole map are set using DIII-D data from shot 115467 in which the C-coils were activated [J. L. Luxon and L. E. Davis, Fusion Technol. 8, 441 (1985)]. The long-wavelength perturbations alter the structure of the interception of magnetic field lines with the divertor plates, but the interception is in sharp lines. The dipole perturbations cause a spreading of the interception of the field lines with the divertor plates, which alleviates problems associated with heat deposition. Magnetic field lines are the trajectories of a one-and-a-half degree of freedom Hamiltonian, which strongly constrains the topological features of the lines. Although the field line maps that we use do not accurately represent the trajectories through ordinary space of individual field lines, they do represent their topological structure

Dynamic behavior of detached recombining plasmas during ELM-like plasma heat pulses in the divertor plasma simulator NAGDIS-II

International Nuclear Information System (INIS)

Uesugi, Y.; Hattori, N.; Nishijima, D.; Ohno, N.; Takamura, S.

2001-01-01

It has been recognized that the ELMs associated with a good confinement at the edge, such as H-mode, must bring an enormous energy to the divertor target plate through SOL and detached plasmas. The understanding of the ELM energy transport through SOL to the divertor target is rather poor at the moment, which leads to an ambiguous estimation of the deposited heat load on the divertor target in ITER. In the present work the ELM-like plasma heat pulse is generated by rf heating in a linear divertor plasma simulator. Energetic electrons with an energy range 10-40 eV are effectively generated by rf heating in low temperature plasmas with (T e )< ∼1 eV. It is observed experimentally that the energetic electrons ionize the highly excited Rydberg atoms quickly, bringing a rapid increase of the ion particle flux to the target, and make the detached plasmas attached to the target. Detailed physical processes about the interaction between the heat pulse with conduction and convection, and detached recombining plasmas are discussed

Divertor load footprint of ELMs in pellet triggering and pacing experiments at JET

Energy Technology Data Exchange (ETDEWEB)

Frigione, D., E-mail: domenico.frigione@frascati.enea.it [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Garzotti, L. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Lennholm, M. [EFDA CSU, Culham Science Centre, OX14 3DB (United Kingdom); Alper, B. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Artaserse, G. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Bennett, P. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Giovannozzi, E. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Eich, T. [Max Planck Institute for Plasma Physics, Garching (Germany); Kocsis, G. [WIGNER RCP RMI, POB 49, 1525 Budapest (Hungary); Lang, P.T. [Max Planck Institute for Plasma Physics, Garching (Germany); Maddaluno, G. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Mooney, R. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Rack, M. [Institut für Energieforschung – Plasmaphysik, Forschungszentrum Jülich, 52425 Jülich (Germany); Sips, G. [EFDA CSU, Culham Science Centre, OX14 3DB (United Kingdom); Tvalashvili, G. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom); Viola, B. [Unità Tecnica Fusione, ENEA C.R. Frascati, via E. Fermi 45, 00044 Frascati (Roma) (Italy); Wilkes, D. [CCFE, Culham Science Centre, OX14 3DB (United Kingdom)

2015-08-15

An investigation of pellet pacing and triggering of Edge Localized Modes (ELMs) was carried out in the frame of ELM mitigation studies aimed at reducing their damaging effects on the plasma-facing components (PFCs). The divertor power load footprint of triggered ELMs was compared with gas puffing controlled ELMs. Small pellets, corresponding to a few per cent of the target plasma particle inventory, were used to minimize the fueling effect and the total particle throughput. There is no evidence that pellets can reduce the divertor power load with respect to gas fueling when operating at the same ELM frequency. The line average density and the energy confinement time remained constant when the gas was progressively substituted by pellets. The launch from the Vertical High Field Side (VHFS) confirmed to be more efficient in ELM triggering than from the Low Field Side (LFS) while the power load footprint remained the same both in time evolution and in spatial distribution when changing the injection geometry.

Optimization design study of an innovative divertor concept for future experimental tokamak-type fusion reactors

International Nuclear Information System (INIS)

Willem Janssens, Ir.; Crutzen, Y.; Farfaletti-Casali, F.; Matera, R.

1991-01-01

The design optimization study of an innovative divertor concept for future experimental tokamak-type fusion devices is both an answer to the actual problems encountered in the multilayer divertor proposals and an illustration of a rational modelling philosophy and optimization strategy for the development of a new divertor structure. Instead of using mechanical attachment or metallurgical bonding of the protective material to the heat sink as in most actual divertor concepts, the so-called brush divertor in this study uses an array of unidirectional fibers penetrating in both the protective armor and the underling composite heat sink. Although the approach is fully concentrated on the divertor performance, including both a description of its function from the theoretical point of view and an overview of the problems related to the materials choice and evaluation, both the approach followed in the numerical modelling and the judgment of the results are thought to be valid also for other applications. Therefore the spin-off of the study must be situated in both the technological progress towards a feasible divertor solution, which introduces no additional physical uncertainties, and in the general area of the thermo-mechanical finite-element modelling on both macro-and microscale. The brush divertor itself embodies the use, and thus the modelling, of advanced materials such as tailor-made metal matrix composites and dispersion strengthened metals, and is shown to offer large potential advantages, demanding however and experimental validation under working conditions. It is clearly indicated where the need originates for an integrated experimental program which must allow to verify the basic modelling assumptions in order to arrive at the use of numerical computation as a powerful and realistic tool of structural testing and life-time prediction

Overview of erosion–deposition diagnostic tools for the ITER-Like Wall in the JET tokamak

International Nuclear Information System (INIS)

Rubel, M.; Coad, J.P.; Widdowson, A.; Matthews, G.F.; Esser, H.G.; Hirai, T.; Likonen, J.; Linke, J.; Lungu, C.P.; Mayer, M.; Pedrick, L.; Ruset, C.

2013-01-01

This paper presents scientific and technical issues related to the development of erosion–deposition diagnostic tools for JET operated with the ITER-Like Wall: beryllium and tungsten marker tiles and several types of wall probes installed in the main chamber and in the divertor. Markers tiles are the standard limiter and divertor components additionally coated first with a thin sandwich of Ni–Be and Mo–W for, beryllium and tungsten markers, respectively. Both types of markers are embedded in regular arrays of limiter and divertor tiles. Coated W–Be probes are also inserted in the Be-covered Inconel cladding tiles on the central column. Other types of erosion–deposition diagnostic tools are: rotating collectors, deposition traps, louver clips, quartz microbalance and mirrors for the First Mirror Test at JET for ITER. The specific role of these tools is discussed in detail

Multi-fluid modeling of low-recycling divertor regimes

International Nuclear Information System (INIS)

Smirnov, R.D.; Pigarov, A.Yu.; Krasheninnikov, S.I.; Rognlien, T.D.; Soukhanovskii, V.A.; Rensink, M.E.; Maingi, R.; Skinner, C.H.; Stotler, D.P.; Bell, R.E.; Kugel, H.W.

2010-01-01

The low-recycling regimes of divertor operation in a single-null NSTX magnetic configuration are studied using computer simulations with the edge plasma transport code UEDGE. The edge plasma transport properties pertinent to the low-recycling regimes are demonstrated. These include the flux-limited character of the parallel heat transport and the high plasma temperatures with the flattened profiles in the scrape-off-layer. It is shown that to maintain the balance of particle fluxes at the core interface the deuterium gas puffing rate should increase as the divertor recycling coefficient decreases. The radial profiles of the heat load to the outer divertor plate, the upstream radial plasma profiles, and the effects of the cross-field plasma transport in the low-recycling regimes are discussed. It is also shown that recycling of lithium impurities evaporating from the divertor plate at high surface temperatures can reverse the low-recycling divertor operational regime to the high-recycling one and may cause thermal instability of the divertor plate (copyright 2010 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

A solid tungsten divertor for ASDEX Upgrade

International Nuclear Information System (INIS)

Herrmann, A; Greuner, H; Jaksic, N; Böswirth, B; Maier, H; Neu, R; Vorbrugg, S

2011-01-01

The conceptual design of a solid tungsten divertor for ASDEX Upgrade (AUG) is presented. The Div-III design is compatible with the existing divertor structure. It re-establishes the energy and heat receiving capability of a graphite divertor and overcomes the limitations of tungsten coatings. In addition, a solid tungsten divertor allows us to investigate erosion and bulk deuterium retention as well as test castellation and target tilting. The design criteria as well as calculations of forces due to halo and eddy currents are presented. The thermal properties of the proposed sandwich structure are calculated with finite element method models. After extensive testing of a target tile in the high heat flux test facility GLADIS, two solid tungsten tiles were installed in AUG for in-situ testing.

Operation method for thermonuclear device and divertor for it

International Nuclear Information System (INIS)

Kotake, Michiko; Yoshioka, Ken; Fukumoto, Hideshi; Okazaki, Takashi; Kinoshita, Shigemi; Takeuchi, Kazuhiro.

1992-01-01

Divertor plates are disposed subsequently along with circumferential direction of a vacuum vessel in a region where magnetic fluxed generated from the divertor coils are injected toward a container wall. Each of the divertor plates is moved in a state that the injection position of the magnetic fluxes enter to the vacuum vessel is kept constant. Alternatively, each of the divertor plates is inclined at an angle facing the injection direction of plasma particle fluxes, or it is inclined so that the angle between the injection surface and the magnetic fluxes makes an acute angle. Since each of the divertor coils is moved in the state of keeping the injection position of the magnetic fluxes during firing of plasmas, in other words, with on change of the current of the divertor coils, the position of the magnetic fluxed is kept at a predetermined condition. Accordingly, charged particles are prevented from concentrating locally without causing eddy current in the coils and the vacuum vessel, which can contribute to the reduction of the wear of the divertor plates. (N.H.)

An analytical erosion model for divertor plates and limiter experiments in CHS

International Nuclear Information System (INIS)

Sagara, A.; Noda, N.; Akiyama, R.; Arimoto, H.; Idei, H.; Iguchi, H.; Kaneko, O.; Kohmoto, T.; Kubo, S.; Matsuoka, K.; Morita, S.; Motojima, O.; Nishimura, K.; Okamura, S.; Takahasi, C.; Takita, Y.; Yamada, I.; Matsunami, N.; Rice, J.; Yamada, H.; Shoji, T.; Ueda, M.

1992-01-01

A self-consistent analytical solution for net erosion of a divertor plate which is set perpendicular to magnetic field lines is presented. The primary flux profile of hydrogen and impurities except redepositing particles is externally given as well as the return ratio of sputtered atoms to the plate. In the direction along the divertor trace, all conditions are uniform. The ionization mean free path is assumed constant to simplify equations. The analytical solution is compared with net erosion experiments carried out in compact helical system (CHS) by exposing a graphite target to a neutral beam heated plasma column introduced perpendicularly to the target along the magnetic field lines through a 2 cm slit opend on a graphite limiter. After exposure to 98 discharges, the target surface is analyzed with Rutherford backscattering method. Deposition profiles of Ti and O impurities are very well explained with the analytical predictions. (orig.)

Innovative design for FAST divertor compatible with remote handling, electromagnetic and mechanical analyses

Energy Technology Data Exchange (ETDEWEB)

Di Gironimo, Giuseppe, E-mail: giuseppe.digironimo@unina.it [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Cacace, Maurizio [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Crescenzi, Fabio [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Labate, Carmelenzo [CREATE, University of Naples Parthenope, Via Acton 38, 80133 Napoli (Italy); Lanzotti, Antonio [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Lucca, Flavio [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Marzullo, Domenico; Mozzillo, Rocco [CREATE, University of Naples Federico II, P.le Tecchio 80, 80125 Napoli (Italy); Pagani, Irene [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy); Ramogida, Giuseppe; Roccella, Selanna [ENEA, Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Viganò, Fabio [LT Calcoli srl, Piazza Prinetti 26/B, 23807 Merate, LC (Italy)

2015-10-15

Highlights: • The conceptual design of FAST divertor has been carried out through a continuous process of requirements refinement and design optimization (V-model approach), in order to achieve a design suited to the needs, RH compatible and ITER-like. • Thermal, structural and electromagnetic analyses have been performed, resulting in requirements refinement. • FAST divertor is now characterized by more realistic, reliable and functional features, satisfying thermo-mechanical capabilities and the remote handling (RH) compatibility. - Abstract: Divertor is a crucial component in Tokamaks, aiming to exhaust the heat power and particles fluxes coming from the plasma during discharges. This paper focuses on the optimization process of FAST divertor, aimed at achieving required thermo-mechanical capabilities and the remote handling (RH) compatibility. Divertor RH system final layout has been chosen between different concept solutions proposed and analyzed within the principles of Theory of Inventive Problem Solving (TRIZ). The design was aided by kinematic simulations performed using Digital Mock-Up capabilities of Catia software. Considerable electromagnetic (EM) analysis efforts and top-down CAD approach enabled the design of a final and consistent concept, starting from a very first dimensioning for EM loads. In the final version here presented, the divertor cassette supports a set of tungsten (W) actively cooled tiles which compose the inner and outer vertical targets, facing the plasma and exhausting the main part of heat flux. W-tiles are assembled together considering a minimum gap tolerance (0.1–0.5 mm) to be mandatorily respected. Cooling channels have been re-dimensioned to optimize the geometry and the layout of coolant volume inside the cassette has been modified as well to enhance the general efficiency.

EU R and D on divertor components

International Nuclear Information System (INIS)

Merola, M.; Daenner, W.; Pick, M.

2005-01-01

Since the last SOFT conference held in Helsinki in 2002, substantial progress has been made in the EU R and D on the divertor components. A number of activities have been completed and new ones have been launched. The present paper gives an update of the works carried out by the EU Participating Team in support of the development of the divertor, which is one of the most challenging components of the next-step ITER machine. The following topics are covered: (1) the further development and consolidation of suitable technologies for the production of high heat-flux components, which culminated with the successful manufacturing and testing of a full-scale vertical target prototype; (2) the completion of the post-irradiation testing of divertor mock-ups and samples; (3) the preparation for the hydraulic and assembly tests of a complete set of full-scale divertor components; (4) the on-going R and D on the definition of workable acceptance criteria for the procurement of ITER high heat-flux components; (5) the activities in support of the divertor design

Particle control in the DIII-D advanced divertor

International Nuclear Information System (INIS)

Schaffer, M.J.; Lippmann, S.I.; Mahdavi, M.A.; Petrie, T.W.; Stambaugh, R.D.; Hogan, J.; Klepper, C.C.; Mioduszewski, P.; Owen, L.; Hill, D.N.; Rensink, M.; Buchenauer, D.

1991-11-01

A new, electrically biasable, semi-closed divertor was installed and operated in the D3-D lower outside divertor location. The semi-closed divertor has yielded static gas pressure buildups in the pumping plenum in excess of 10 mtorr. (The planned cryogenic pumping is not yet installed). Electrical bias controls the distribution of particle recycle between the inner and outer divertors by rvec E x rvec B drifts. Depending on sign, bias increases or decreases the plenum gas pressure. Bias greatly reduce the sensitivity of plenum pressure to separatrix position. In particular, rvec E x rvec B drifts in the D3-D geometry can direct plasma across a divertor target and then optimally into the pumping aperture. Bias, even without active pumping, has also demonstrated a limited control of ELMing H-mode plasma density. 5 refs., 8 figs

Initial results from the Tokapole-II poloidal divertor device

International Nuclear Information System (INIS)

Biddle, A.P.; Dexter, R.N.; Groebner, R.J.; Holly, D.J.; Lipschultz, B.; Phillips, M.W.; Prager, S.C.; Sprott, J.C.

1979-01-01

The latest in a series of internal-ring devices, called Tokapole II, has recently begun operation at the University of Wisconsin. Its purpose is to permit the study of the production and confinement of hot, dense plasmas in either a toroidal octupole (with or without toroidal field) or a tokamak with a four-node poloidal divertor. The characteristics of the device and the results of its initial operation are described here. Quantitative measurements of impurity concentration and radiated power have been made. Poloidal divertor equilibria of square and dee shapes have been produced, and an axisymmetric instability has been observed with the inverse dee. Electron cyclotron resonance heating is used to initiate the breakdown near the axis and to control the initial influx of impurities. A 2-MW RF source at the second harmonic of the ion cyclotron frequency is available and has been used to double the ion temperature when operated at low power with an unoptimized antenna. Initial results of operation as a pure octupole with poloidal Ohmic heating suggest a tokamak-like scaling of density (n proportional to Bsub(p)) and confinement time (tau proportional to n). (author)

Effects of low-Z and high-Z impurities on divertor detachment and plasma confinement

Directory of Open Access Journals (Sweden)

H.Q. Wang

2017-08-01

Full Text Available The impurity-seeded detached divertor is essential for heat exhaust in ITER and other reactor-relevant devices. Dedicated experiments with injection of N2, Ne and Ar have been performed in DIII-D to assess the impact of the different impurities on divertor detachment and confinement. Seeding with N2, Ne and Ar all promote divertor detachment, greatly reducing heat flux near the strike point. The upstream plasma density at the onset of detachment decreases with increasing impurity-puffing flow rates. For all injected impurity species, the confinement and pedestal pressure are correlated with the impurity content and the ratio of separatrix loss power to the l-H transition threshold power. As the divertor plasma approaches detachment, the high-Z impurity seeding tends to degrade the core confinement owing to the increased core radiation. In particular, Ar injection with up to 50% of the injected power radiating in the core cools the pedestal and core plasmas, thus significantly degrading the confinement. As for Ne seeding, medium confinement with H98∼0.8 can be maintained during the detachment phase with the pedestal temperature being reduced by about 50%. In contrast, in the N2 seeded plasmas, radiation is predominately confined in the boundary plasma, which leads to less effect on the confinement and pedestal. In the case of strong N2 gas puffing, the confinement recovers during the detachment, from ∼20% reduction at the onset of the detachment to greater than unity comparable to that before the seeding. The core and pedestal temperatures feature a reduction of 30% from the initial attached phase and remain nearly constant during the detachment phase. The improvement in confinement appears to arise from the increase in pedestal and core density despite the temperature reduction.

Plasma characteristics of the end-cell of the GAMMA 10 tandem mirror for the divertor simulation experiment

International Nuclear Information System (INIS)

Nakashima, Y.; Sakamoto, M.; Yoshikawa, M.; Takeda, H.; Ichimura, K.; Hosoi, K.; Hirata, M.; Ichimura, M.; Ikezoe, R.; Imai, T.; Kariya, T.; Katanuma, I.; Kohagura, J.; Minami, R.; Numakura, T.; Oki, K.; Ueda, H.; Asakura, Nobuyuki; Furuta, T.; Hatayama, A.; Toma, M.; Hirooka, Y.; Masuzaki, S.; Sagara, A.; Shoji, M.; Kado, S.; Matsuura, H.; Nagata, S.; Nishino, N.; Ohno, N.; Tonegawa, A.; Ueda, Y.

2012-11-01

In this paper, detailed characteristics and controllability of plasmas emitted from the end-cell of the GAMMA 10 tandem mirror are described from the viewpoint of divertor simulation studies. The energy analysis of ion flux by using end-loss ion energy analyzer (ELIEA) proved that the obtained high ion temperature (100 - 400 eV) was comparable to SOL plasma parameters in toroidal devices and was controlled by changing the ICRF power. Parallel ion temperature T i∥ determined from the probe and calorimeter shows a linear relationship with the ICRF power in the central-cell and agrees with the results of ELIEA. Additional ICRF heating revealed a significant enhancement of particle flux, which indicated an effectiveness of additional plasma heating in adjacent cells toward the improvement of the performance. Superimposing the ECH pulse of 380 kW, 5 ms attained the maximum heat-flux more than 10 MW/m 2 on axis. This value comes up to the heat-load of the divertor plate of ITER, which gives a clear prospect of generating the required heat density for divertor studies by building up heating systems to the end-mirror cell. Initial results of plasma irradiation experiment and construction of new divertor module are also described. (author)

Design of the Wendelstein 7-X inertially cooled Test Divertor Unit Scraper Element

Energy Technology Data Exchange (ETDEWEB)

Lumsdaine, Arnold, E-mail: lumsdainea@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Boscary, Jean [Max Planck Institute for Plasma Physics, Garching (Germany); Fellinger, Joris [Max Planck Institute for Plasma Physics, Greifswald (Germany); Harris, Jeff [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Hölbe, Hauke; König, Ralf [Max Planck Institute for Plasma Physics, Greifswald (Germany); Lore, Jeremy; McGinnis, Dean [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Neilson, Hutch; Titus, Peter [Princeton Plasma Physics Lab, Princeton, NJ (United States); Tretter, Jörg [Max Planck Institute for Plasma Physics, Garching (Germany)

2015-10-15

Highlights: • The justification for the installation of the Test Divertor Unit Scraper Element is given. • Specially designed operational scenarios for the component are presented. • Plans for the design of the component are detailed. - Abstract: The Wendelstein 7-X stellarator is scheduled to begin operation in 2015, and to achieve full power steady-state operation in 2019. Computational simulations have indicated that for certain plasma configurations in the steady-state operation, the ends of the divertor targets may receive heat fluxes beyond their qualified technological limit. To address this issue, a high heat-flux “scraper element” (HHF-SE) has been designed that can protect the sensitive divertor target region. The surface profile of the HHF-SE has been carefully designed to meet challenging engineering requirements and severe spatial limitations through an iterative process involving physics simulations, engineering analysis, and computer aided design rendering. The desire to examine how the scraper element interacts with the plasma, both in terms of how it protects the divertor, and how it affects the neutral pumping efficiency, has led to the consideration of installing an inertially cooled version during the short pulse operation phase. This Test Divertor Unit Scraper Element (TDU-SE) would replicate the surface profile of the HHF-SE. The design and instrumentation of this component must be completed carefully in order to satisfy the requirements of the machine operation, as well as to support the possible installation of the HHF-SE for steady-state operation.

Divertor impurity monitor for the International Thermonuclear Experimental Reactor

Science.gov (United States)

Sugie, T.; Ogawa, H.; Nishitani, T.; Kasai, S.; Katsunuma, J.; Maruo, M.; Ebisawa, K.; Ando, T.; Kita, Y.

1999-01-01

The divertor impurity monitoring system of the International Thermonuclear Experimental Reactor has been designed. The main functions of this system are to identify impurity species and to measure the two-dimensional distributions of the particle influxes in the divertor plasmas. The wavelength range is 200-1000 nm. The viewing fans are realized by molybdenum mirrors located in the divertor cassette. With additional viewing fans seeing through the gap between the divertor cassettes, the region approximately from the divertor leg to the x point will be observed. The light from the divertor region passes through the quartz windows on the divertor port plug and the cryostat, and goes through the dog-leg optics in the biological shield. Three different type of spectrometers: (i) survey spectrometers for impurity species monitoring, (ii) filter spectrometers for the particle influx measurement with the spatial resolution of 10 mm and the time resolution of 1 ms, and (iii) high dispersion spectrometers for high resolution wavelength measurements are designed. These spectrometers are installed just behind the biological shield (for λthe transmission loss in fiber and in the diagnostic room (for λ⩾450 nm) from the point of view of accessibility and flexibility. The optics have been optimized by a ray trace analysis. As a result, 10-15 mm spatial resolution will be achieved in all regions of the divertor.

Geometrical properties of a 'snowflake' divertor

International Nuclear Information System (INIS)

Ryutov, D. D.

2007-01-01

Using a simple set of poloidal field coils, one can reach the situation in which the null of the poloidal magnetic field in the divertor region is of second order, not of first order as in the usual X-point divertor. Then, the separatrix in the vicinity of the null point splits the poloidal plane not into four sectors, but into six sectors, making the whole structure look like a snowflake (hence the name). This arrangement allows one to spread the heat load over a much broader area than in the case of a standard divertor. A disadvantage of this configuration is that it is topologically unstable, and, with the current in the plasma varying with time, it would switch either to the standard X-point mode, or to the mode with two X-points close to each other. To avoid this problem, it is suggested to have a current in the divertor coils that is roughly 5% higher than in an ''optimum'' regime (the one in which a snowflake separatrix is formed). In this mode, the configuration becomes stable and can be controlled by varying the current in the divertor coils in concert with the plasma current; on the other hand, a strong flaring of the scrape-off layer still remains in force. Geometrical properties of this configuration are analyzed. Potential advantages and disadvantages of this scheme are discussed

Density dependence of SOL power width in ASDEX upgrade L-Mode

Directory of Open Access Journals (Sweden)

B. Sieglin

2017-08-01

A recent study [4] with an open divertor configuration found an asymmetry of the power fall-off length between inner and outer target with a smaller power fall-off length λq,i on the inner divertor target. Measurements with a closed divertor configuration find a similar asymmetry for low recycling divertor conditions. It is found, in the experiment, that the in/out asymmetry λq,i/λq,o is strongly increasing with increasing density. Most notably the heat flux density at the inner divertor target is reducing with increasing λq,i whilst the total power onto each divertor target stays constant. It is found that λq,o exhibits no significant density dependence for hydrogen and deuterium but increases with about the square root of the electron density for helium. The difference between H,D and He could be due to the different recycling behaviour in the divertor. These findings may help current modelling attempts to parametrize the density dependence of the widening of the power channel and thus allow for detailed comparison to both divertor effects like recycling or increased upstream SOL cross field transport.

Divertor erosion in DIII-D

International Nuclear Information System (INIS)

Whyte, D.G.; Bastasz, R.; Wampler, W.R.; Brooks, J.N.; West, W.P.; Wong, C.P.C.; Buzhinskij, O.I.; Opimach, I.V.

1998-08-01

Net erosion rates of carbon target plates have been measured in situ for the DIII-D lower divertor. The principal method of obtaining this data is the DiMES sample probe. Recent experiments have focused on erosion at the outer strike-point (OSP) of two divertor plasma conditions: attached (T e > 40 eV) ELMing plasmas, and detached (T e 2 . In this case, measurements and modeling agree for both gross and net carbon erosion, showing the near-surface transport and redeposition of the carbon is well understood. In the attached cases, physical sputtering (with enhancement from self-sputtering and oblique incidence) is dominant, and the effective sputtering yield, Y, is greater than 10%. In ELM-free discharges, the total OSP net erosion rate is equal to the rate of carbon accumulation in the core plasma. For the detached divertor cases, the cold incident plasma eliminates physical sputtering. Attempts to measure chemically eroded hydrocarbon molecules spectroscopically indicate an upper limit of Y ≤ 0.1% for the chemical sputtering yield. Net erosion is suppressed at the outer strike-point, which becomes a region of net redeposition (∼ 4 cm/exposure-year). The private flux wall is measured to be a region of net redeposition with dense, high neutral pressure, attached divertor plasmas. Leading edges intercepting parallel heat flux (∼ 50 MW/m 2 ) have very high net erosion rates at the OSP of an attached plasma (∼ 10 microm/s > 1,000x erosion rate of aligned surfaces). Leading edge erosion, and subsequent carbon redeposition, caused by tile gaps can account for half of the deuterium codeposition in the DIII-D divertor

Divertor tungsten tiles erosion in the region of the castellated gaps

Energy Technology Data Exchange (ETDEWEB)

Hu, Wanpeng, E-mail: wangdez@dlut.edu.cn; Sang, Chaofeng; Sun, Zhenyue; Wang, Dezhen

2016-11-01

Highlights: • Simulation of the tungsten tiles erosion by different impurities in the divertor gap region is done by using a 2d3v Particle-In-Cell code. • High-Z impurity causes the largest erosion rate on W tile. • The peak physical sputtering erosion rate locates at the plasma-facing corners. - Abstract: Erosion of tungsten (W) is a very important issue for the future fusion device. The castellated divertor makes it more complicated due to complex geometry of the gap between the tiles. In this work, the plasma behaviors and resulting W tile erosion in the divertor tile gap region are studied by using a two dimension-in-space and three dimension-in-velocity (2d3 v) Particle-In-Cell (PIC) code. Deuterium ions (D{sup +}) and electrons are traced self-consistently in the simulation to provide the plasma background. Since there are lots of impurities, which may make a great impact on the tile erosion, in the divertor region to radiate the power, the erosion of W tile by different species are thus considered. The contributions of deuterium and impurities: Li, C, Ne, and Ar, to the W erosion, are studied under EAST conditions to show a straightforward insight. It is observed that the physical sputtering of W tile by impurities is much higher than that by the D ions, and the peak erosion region locates at the plasma-facing corners.

Moving Divertor Plates in a Tokamak

International Nuclear Information System (INIS)

Zweben, S.J.; Zhang, H.

2009-01-01

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions

Moving Divertor Plates in a Tokamak

Energy Technology Data Exchange (ETDEWEB)

S.J. Zweben, H. Zhang

2009-02-12

Moving divertor plates could help solve some of the problems of the tokamak divertor through mechanical ingenuity rather than plasma physics. These plates would be passively heated on each pass through the tokamak and cooled and reprocessed outside the tokamak. There are many design options using varying plate shapes, orientations, motions, coatings, and compositions.

Divertor IR thermography on Alcator C-Moda)

Science.gov (United States)

Terry, J. L.; LaBombard, B.; Brunner, D.; Payne, J.; Wurden, G. A.

2010-10-01

Alcator C-Mod is a particularly challenging environment for thermography. It presents issues that will similarly face ITER, including low-emissivity metal targets, low-Z surface films, and closed divertor geometry. In order to make measurements of the incident divertor heat flux using IR thermography, the C-Mod divertor has been modified and instrumented. A 6° toroidal sector has been given a 2° toroidal ramp in order to eliminate magnetic field-line shadowing by imperfectly aligned divertor tiles. This sector is viewed from above by a toroidally displaced IR camera and is instrumented with thermocouples and calorimeters. The camera provides time histories of surface temperatures that are used to compute incident heat-flux profiles. The camera sensitivity is calibrated in situ using the embedded thermocouples, thus correcting for changes and nonuniformities in surface emissivity due to surface coatings.

Accounting of the Power Balance for Neutral-beam heated H-Mode Plasmas in NSTX

International Nuclear Information System (INIS)

Paul, S.F.; Maingi, R.; Soukhanovskii, V.; Kaye, S.M.; Kugel, H.

2004-01-01

A survey of the dependence of power balance on input power, shape, and plasma current was conducted for neutral-beam-heated plasmas in the National Spherical Torus Experiment (NSTX). Measurements of heat to the divertor strike plates and divertor and core radiation were taken over a wide range of plasma conditions. The different conditions were obtained by inducing a L-mode to H-mode transition, changing the divertor configuration [lower single null (LSN) vs. double-null (DND)] and conducting a NBI power scan in H-mode. 60-70% of the net input power is accounted for in the LSN discharges with 20% of power lost as fast ions, 30-45% incident on the divertor plates, up to 10% radiated in the core, and about 12% radiated in the divertor. In contrast, the power accountability in DND is 85-90%. A comparison of DND and LSN data show that the remaining power in the LSN is likely to be directed to the upper divertor

R(and)D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project

International Nuclear Information System (INIS)

Hirai, T.; Maier, H.; Rubel, M.

2006-01-01

The ITER-like Wall Project was initiated at JET, with the goal of testing the reference material combination chosen for ITER: beryllium (Be) in the main chamber (wall and limiters) and tungsten (W) in the divertor. The major aims are to study the tritium retention, material mixing, melt layer behavior and to optimize plasma operation scenarios with a full metal wall. The project requires major design and engineering efforts in R(and)D: (i) bulk W tile, (ii) W coatings on carbon fibre composites (CFC) (iii) Be coatings on Inconel, (iv) Be marker tiles. For the W divertor, two R(and)D tasks were initiated: (1) development of a conceptual design for a bulk W tile as the main outer divertor target plate, and (2) W coating selection from 14 different samples produced by various techniques for the other divertor plates and neutral beam shine. The bulk W tile must withstand power loads of 7 MW/m 2 for 10 s. JET divertor plates are not actively cooled, therefore, heat capacity of the tiles is an important design parameter. In addition to power handling, mechanical structural stability under electromagnetic forces and compatibility with remote handling are the key requirements in the design. The design has been completed. The test-tile survived 100 pulses at 7 MW/m 2 for 10 s in the electron beam facility, JUDITH. The W coatings with different thickness, thin ( 2 and 200 pulses at 10 MW/m 2 for 5 s. In all tested samples cracks developed perpendicularly to the fiber bundles in CFC because of contraction of the coating in the cooling phase. Coatings were also exposed to 1000 ELM-like loading pulses. The thin coatings showed fatigue leading to delamination, whereas for thick coatings better resistance in ELM-like loading. As a result of R(and)D a full W divertor was decided: bulk metal at the outer divertor and W coating at other areas. Be-related R(and)D activities are in two areas. Production of 8-9 μm layers on inner wall cladding Inconel tiles ensures the full coating of

Thermomechanical simulation of WEST actively cooled upper divertor

International Nuclear Information System (INIS)

Batal, T.; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

2016-01-01

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m 2 . This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m 2 , and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m 2 . The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

Thermomechanical simulation of WEST actively cooled upper divertor

Energy Technology Data Exchange (ETDEWEB)

Batal, T., E-mail: tristan.batal@cea.fr; Richou, M.; Guilhem, D.; Firdaouss, M.; Larroque, S.; Ferlay, F.; Missirlian, M.; Bucalossi, J.

2016-11-15

The Tore Supra tokamak is being transformed in an x-point divertor fusion device in the frame of the WEST (W-for tungsten-Environment in Steady-state Tokamak) project, launched in support to the ITER tungsten divertor strategy. The WEST project aims to test ITER-like W monoblock Plasma Facing Units (PFU). This ITER-like divertor will be tested under long plasma discharge up to 1000 s, with high heat flux density up to 20 MW/m{sup 2}. This paper presents the results of ANSYS thermal-structural simulations of the WEST upper divertor. The upper divertor is made of twelve 30° sectors, each one composed of 38 PFU. The PFUs are actively cooled CuCrZr heat sinks and the incidence surface is coated with a thin tungsten layer. The fixing system is made of pins engaged in slotted holes. Besides, the fixing system of the sector assembly is the same as WEST lower divertor, so one upper divertor sector can be used indifferently in upper or Lower position during transitional operation phases in WEST. The total surface of the upper divertor is 8 m{sup 2}, and it has to be able to extract up to 4 MW in steady-state, with peak heat flux values up to 8 MW/m{sup 2}. The fixing system was designed to handle structural loads such as forces and torques resulting from halo and eddy current, respectively, especially during disruptions and Vertical Displacement Event (VDE). The torque resulting from eddy current is first calculated thanks to an internal CEA ANSYS APDL routine. Then the ANSYS structural and thermal-structural simulations of the PFU are presented, and its design is validated thanks to A-level RCC-MRx criteria. Finally, the most conservative load case is determined in order to validate the design of the pins and the support structure.

Deposition Measurements in NSTX

Science.gov (United States)

Skinner, C. H.; Kugel, H. W.; Hogan, J. T.; Wampler, W. R.

2004-11-01

Two quartz microbalances have been used to record deposition on the National Spherical Torus Experiment. The experimental configuration mimics a typical diagnostic window or mirror. An RS232 link was used to acquire the quartz crystal frequency and the deposited thickness was recorded continuously with 0.01 nm resolution. Nuclear Reaction Analysis of the deposit was consistent with the measurement of the total deposited mass from the change in crystal frequency. We will present measurements of the variation of deposition with plasma conditions. The transport of carbon impurities in NSTX has been modelled with the BBQ code. Preliminary calculations indicated a negligible fraction of carbon generated at the divertor plates in quiescent discharges directly reaches the outer wall, and that transient events are responsible for the deposition.

ATHENA simulations of divertor pump trip and loss of heat sink transients for the GSSR

Energy Technology Data Exchange (ETDEWEB)

Sjoeberg, A

2001-04-01

The ITER-FEAT Generic Site Safety Report includes evaluations of the consequences of various types of conceivable transients that may occur during operation. The transients that have to be considered in this respect are specified in the Accident Analysis Specifications document of the safety report. For the divertor primary heat transport system the ranges of transients include amongst others a trip of the main circulation pump in the divertor cooling loop as well as a loss of heat sink, both initiated at full fusion power operation. The thermal-hydraulic consequences related to the coolability of the divertor primary heat transport system components for these two transients have been evaluated and summarized in the safety report and in the current report an overview of those efforts and associated outcome is provided. The analyses have been made with the ATHENA thermal-hydraulic code using a separately developed ATHENA model of the ITER-FEAT divertor cooling system. The results from the analyses indicate that for the pump trip transient the margin against overheating of critical highly loaded parts of the divertor cassette is small but seems sufficient. In case of the loss of heat sink transient the conservative analysis reveals that the pressurizer safety valve will be opened for an extended period of time and the long term transient development indicates a risk of completely filling up the pressurizer vessel. Thus the margins against jeopardizing the integrity of the divertor cooling system with the current design are for this case small but can for a long term operation at associate conditions pose a problem.

Towards the procurement of the ITER divertor

International Nuclear Information System (INIS)

Merola, M.; Tivey, R.; Martin, A.; Pick, M.

2006-01-01

The procurement of the ITER divertor is planned to start in 2009. On the basis of the present common understanding of the sharing of the ITER components, the Japanese Participating Team (JAPT) will supply the outer vertical target, the Russian Federation (RF) PT the dome liner and will perform the high heat flux testing, the EU PT will supply the inner vertical targets and the cassette bodies, including final assembly of the divertor plasma-facing components (PFCs). The manufacturing of the PFCs of the ITER divertor represents a challenging endeavor due to the high technologies which are involved, and due to the unprecedented series production. To mitigate the associated risks, special arrangements need to be put in place prior to and during procurement to ensure quality and to keep to the time schedule. Before procurement can start, an ITER review of the qualification and production capability of each candidate PT is planned. Well in advance of the assumed start of the procurement, each PT which would like to contribute to the divertor PFC procurement, should first demonstrate its technical qualification to carry out the procurement with the required quality, and in an efficient and timely manner. Appropriate precautions, like subdivision of the procurement into stages, are also to be adopted during the procurement phase to mitigate the consequences of possible unexpected manufacturing problems. In preparation for writing the procurement specification for the vertical targets, the topic of setting acceptance criteria is also being addressed. This activity has the objective of defining workable acceptance criteria for the PFC armour joints. A complete set of analyses is also in progress to assess the latest design modifications against the design requirements. This task includes neutronic, shielding, thermo-mechanical and electromagnetic analyses. More than half of the ITER plasma parameters that must be measured and the related diagnostics are located in the

Alternative divertor target concepts for next step fusion devices

Science.gov (United States)

Mazul, I. V.

2016-12-01

The operational conditions of a divertor target in the next steps of fusion devices are more severe in comparison with ITER. The current divertor designs and technologies have a limited application concerning these conditions, and so new design concepts/technologies are required. The main reasons which practically prevent the use of the traditional motionless solid divertor target are analyzed. We describe several alternative divertor target concepts in this paper. The comparative analysis of these concepts (including the advantages and the drawbacks) is made and the prospects for their practical implementation are prioritized. The concept of the swept divertor target with a liquid metal interlayer between the moving armour and motionless heat-sink is presented in more detail. The critical issues of this design are listed and outlined, and the possible experiments are presented.

Neutral particle retention in the JET MK I divertor

International Nuclear Information System (INIS)

Ehrenberg, J.K.; Campbell, D.J.; Harbour, P.J.; Horton, L.D.; Loarte, A.; McCormick, G.K.; Monk, R.D.; Saibene, G.R.; Simonini, R.; Taroni, A.; Stamp, M.F.

1997-01-01

Retention of neutral deuterium and nitrogen in the JET MK I divertor has been investigated. Results show that ohmic plasma detachment reduces deuterium retention, that the magnetic divertor configuration has some influence on the achievable deuterium retention, and that nitrogen in nitrogen-seeded steady state detached H-mode discharges accumulates in the divertor. (orig.)

Variation of particle exhaust with changes in divertor magnetic balance

International Nuclear Information System (INIS)

Petrie, T.W.; Allen, S.L.; Brooks, N.H.

2006-01-01

Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e. the degree to which the divertor topology is single-null or double-null (DN) and (2) the direction of the of B x ∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the B x ∇B ion particle drift direction. Our data suggests that the presence of B x ∇B and E x B ion particle drifts in the scrape-off layer and divertor(s) play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density. These results have implications for particle control in ITER and other future tokamaks

Vanadium alloys for the radiative divertor program of DIII-D

International Nuclear Information System (INIS)

Smith, J.P.; Johnson, W.R.; Stambaugh, R.D.; Trester, P.W.; Smith, D.; Bloom, E.

1995-10-01

Vanadium alloys provide an attractive solution for fusion power plants as they exhibit a potential for low environmental impact due to low level of activation from neutron fluence and a relatively short half-life. They also have attractive material properties for use in a reactor. General Atomics along with Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL), has developed a plan to utilize vanadium alloys as part of the Radiative Divertor Project (RDP) modification for the DIII-D tokamak. The goal for using vanadium alloys is to provide a meaningful step towards developing advanced materials for fusion power applications by demonstrating the in-service behavior of a vanadium alloy (V-4Cr-4Ti) in a tokamak in conjunction with developing essential fabrication technology for the manufacture of full-scale vanadium alloy components. A phased approach towards utilizing vanadium in DIII-D is being used starting with small coupons and samples, advancing to a small component, and finally a portion of the new double-null, slotted divertor will be fabricated from vanadium alloy product forms. A major portion of the program is research and development to support fabrication and resolve key issues related to environmental effects

Thermal response to heat fluxes of the W7-AS divertor surface submitted to surface modification under high temperature treatment

Energy Technology Data Exchange (ETDEWEB)

Hildebrandt, D., E-mail: dieter.hildebrandt@ipp.mpg.d [Euratom Association, Max-Planck-Institut fuer Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstrasse 1, 17491 Greifswald (Germany); Duebner, A. [Euratom Association, Max-Planck-Institut fuer Plasmaphysik, Teilinstitut Greifswald, Wendelsteinstrasse 1, 17491 Greifswald (Germany); Greuner, H.; Wiltner, A. [Teilinstitut Garching, Boltzmannstr. 2, 85748 Garching (Germany)

2009-06-15

Some target tiles of the W7-AS divertor has been investigated with respect to their thermal behaviour at the surface during power loading with well-defined heat fluxes in the Gladis facility. The primary aim was to examine uncertainties in the determination of heat fluxes derived from IR-thermography during operation of W7-AS. It is found that the derived heat flux profiles are strongly influenced by the local distribution of plasma-deposited contamination analyzed by AES and SIMS. With the observed actual surface conditions characterized by redeposited contamination equivalent up to about 1 mum thickness, the heat fluxes were partially overestimated up to a factor of 4 during operation of W7-AS. This uncertainty is observed to be significantly reduced after heat treatment at temperatures beyond 700 deg. C attained at power flux densities of 10.5 MW/m{sup 2} and durations longer than 5 s.

Thermal response to heat fluxes of the W7-AS divertor surface submitted to surface modification under high temperature treatment

International Nuclear Information System (INIS)

Hildebrandt, D.; Duebner, A.; Greuner, H.; Wiltner, A.

2009-01-01

Some target tiles of the W7-AS divertor has been investigated with respect to their thermal behaviour at the surface during power loading with well-defined heat fluxes in the Gladis facility. The primary aim was to examine uncertainties in the determination of heat fluxes derived from IR-thermography during operation of W7-AS. It is found that the derived heat flux profiles are strongly influenced by the local distribution of plasma-deposited contamination analyzed by AES and SIMS. With the observed actual surface conditions characterized by redeposited contamination equivalent up to about 1 μm thickness, the heat fluxes were partially overestimated up to a factor of 4 during operation of W7-AS. This uncertainty is observed to be significantly reduced after heat treatment at temperatures beyond 700 deg. C attained at power flux densities of 10.5 MW/m 2 and durations longer than 5 s.

Active Control of Power Exhaust in Strongly Heated ASDEX Upgrade Plasmas

Science.gov (United States)

Dux, Ralph; Kallenbach, Arne; Bernert, Matthias; Eich, Thomas; Fuchs, Christoph; Giannone, Louis; Herrmann, Albrecht; Schweinzer, Josef; Treutterer, Wolfgang

2012-10-01

Due to the absence of carbon as an intrinsic low-Z radiator, and tight limits for the acceptable power load on the divertor target, ITER will rely on impurity seeding for radiative power dissipation and for generation of partial detachment. The injection of more than one radiating species is required to optimise the power removal in the main plasma and in the divertor region, i.e. a low-Z species for radiation in the divertor and a medium-Z species for radiation in the outer core plasma. In ASDEX Upgrade, a set of robust sensors, which is suitable to feedback control the radiated power in the main chamber and the divertor as well as the electron temperature at the target, has been developed. Different feedback schemes were applied in H-mode discharges with a maximum heating power of up to 23,W, i.e. at ITER values of P/R (power per major radius) to control all combinations of power flux into the divertor region, power flux onto the target or electron temperature at the target through injection of nitrogen as the divertor radiator and argon as the main chamber radiator. Even at the highest heating powers the peak heat flux density at the target is kept at benign values. The control schemes and the plasma behaviour in these discharges will be discussed.

Divertor development for ITER

International Nuclear Information System (INIS)

Janeschitz, G.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Matera, R.; Martin, E.; Parker, R.; Tivey, R.; Pacher, H.D.

1998-01-01

The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near at the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented. (orig.)

Divertor, thermonuclear device and method of neutralizing high temperature plasma

International Nuclear Information System (INIS)

Ikegami, Hideo.

1995-01-01

The thermonuclear device comprises a thermonuclear reactor for taking place fusion reactions to emit fusion plasmas, and a divertor made of a hydrogen occluding material, and the divertor is disposed at a position being in contact with the fusion plasmas after nuclear fusion reaction. The divertor is heated by fusion plasmas after nuclear fusion reaction, and hydrogen is released from the hydrogen occluding material as a constituent material. A gas blanket is formed by the released hydrogen to cool and neutralize the supplied high temperature nuclear fusion plasmas. This prevents the high temperature plasmas from hitting against the divertor, elimination of the divertor by melting and evaporation, and solve a problem of processing a divertor activated by neutrons. In addition, it is possible to utilize hydrogen isotopes of fuels effectively and remove unnecessary helium. Inflow of impurities from out of the system can also be prevented. (N.H.)

The ITER divertor cassette project

International Nuclear Information System (INIS)

Ulrickson, M.; Tivey, R.; Akiba, M.

2001-01-01

The divertor ''Large Project'' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

The ITER divertor cassette project

International Nuclear Information System (INIS)

Ulrickson, M.; Tivey, R.; Akiba, M.

1999-01-01

The divertor 'Large Project' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

Two-dimensional analysis of limiter/divertor transition in scrape-off layer plasmas

International Nuclear Information System (INIS)

Ueda, N.; Itoh, K.; Itoh, S.I.

1989-01-01

The structures of scrape-off layer and divertor plasmas have been studied numerically with a neutral code and a two-dimensional fluid code. Doublet-III is taken as an example for an open divertor configuration. A decisive parameter is the distance between the plasma surface (determined by the magnetic separatrix) and the limiter, which is varied in order to assess the interaction of the plasma with the limiter as well as the effect of neutrals on the main plasma. The minimum value of the limiter clearance needed to prevent plasma-limiter interaction is determined. The scaling of the edge temperature and the dependence of the e-folding length of the scrape-off layer plasma on the heating power are obtained. (author). 16 refs, 17 figs

Analysis of particle transport in a gas target divertor

Energy Technology Data Exchange (ETDEWEB)

Ohtsu, Shigeki; Tanaka, Satoru [Tokyo Univ. (Japan). Faculty of Engineering

1996-10-01

2-dimensional modelling of divertor plasma was performed with three types of the divertor geometry configuration. Pumping is effective to reduce neutral recycling to core region in the configuration without baffle. In baffle configuration, a good shielding of neutrals in the divertor region can be achieved. The dome configuration reduces plasma density near the null region and flow shear near the separatrix. (author)

Scaling of divertor heat flux profile widths in DIII-D

International Nuclear Information System (INIS)

Lasnier, C.J.; Makowski, M.A.; Boedo, J.A.; Allen, S.L.; Brooks, N.H.; Hill, D.N.; Leonard, A.W.; Watkins, J.G.; West, W.P.

2011-01-01

New scalings of the dependence of divertor heat flux peak and profile width, important parameters for the design of future large tokamaks, have been obtained from recent DIII-D experiments. We find the peak heat flux depends linearly on input power, decreases linearly with increasing density, and increases linearly with plasma current. The profile width has a weak dependence on input power, is independent of density up to the onset of detachment, and is inversely proportional to the plasma current. We compare these results with previously published scalings, and present mathematical expressions incorporating these results.

Steady-state tokamak reactor with non-divertor impurity control: STARFIRE

International Nuclear Information System (INIS)

Baker, C.C.

1980-01-01

STARFIRE is a conceptual design study of a commercial tokamak fusion electric power plant. Particular emphasis has been placed on simplifying the reactor concept by developing design concepts to produce a steady-state tokamak with non-divertor impurity control and helium ash removal. The concepts of plasma current drive using lower hybrid rf waves and a limiter/vacuum system for reactor applications are described

Analysis of divertor asymmetry using a simple five-point model

International Nuclear Information System (INIS)

Hayashi, Nobuhiko; Takizuka, Tomonori; Hatayama, Akiyoshi; Ogasawara, Masatada.

1997-03-01

A simple five-point model of the scrape-off layer (SOL) plasma outside the separatrix of a diverted tokamak has been developed to study the inside/outside divertor asymmetry. The SOL current, gas pumping/puffing in the divertor region, and divertor plate biasing are included in this model. Gas pumping/puffing and biasing are shown to control divertor asymmetry. In addition, the SOL current is found to form asymmetric solutions without external controls of gas pumping/puffing and biasing. (author)

Interpretation of low ionized impurity distributions in the ASDEX Upgrade divertor

International Nuclear Information System (INIS)

Lieder, G.; Napiontek, B.; Radtke, R.; Field, A.; Fussmann, G.; Kallenbach, A.; Kiemer, K.; Mayer, H.M.

1993-01-01

Design studies for reactor-like devices, like ITER, have particularly emphasized the importance of erosion and transport of material from the divertor target plates. In this context experimental measurements which can lead to a better understanding of the underlying physics are highly desirable. We discuss the spatial profiles of line emission from impurities measured in the divertor of ASDEX Upgrade with a recently developed multi-chord divertor spectrometer system. These profiles are obtained from observations in the ultra-violet/visible spectral range. The divertor spectrometer system was developed particularly to measure the erosion of the divertor plates and to study transport of the impurities and the ionization and recombination processes in the divertor region. (author) 6 refs., 3 figs., 2 tabs

Interpretation of low ionized impurity distributions in the ASDEX Upgrade divertor

Energy Technology Data Exchange (ETDEWEB)

Lieder, G; Napiontek, B; Radtke, R; Field, A; Fussmann, G; Kallenbach, A; Kiemer, K; Mayer, H M [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

1994-12-31

Design studies for reactor-like devices, like ITER, have particularly emphasized the importance of erosion and transport of material from the divertor target plates. In this context experimental measurements which can lead to a better understanding of the underlying physics are highly desirable. We discuss the spatial profiles of line emission from impurities measured in the divertor of ASDEX Upgrade with a recently developed multi-chord divertor spectrometer system. These profiles are obtained from observations in the ultra-violet/visible spectral range. The divertor spectrometer system was developed particularly to measure the erosion of the divertor plates and to study transport of the impurities and the ionization and recombination processes in the divertor region. (author) 6 refs., 3 figs., 2 tabs.

Influence of stray light for divertor spectroscopy in ITER

International Nuclear Information System (INIS)

Kajita, Shin; Veshchev, Evgeny; Lisgo, Steve; Barnsley, Robin; Morgan, Philip; Walsh, Michael; Ogawa, Hiroaki; Sugie, Tatsuo; Itami, Kiyoshi

2015-01-01

The influence of stray light in the divertor spectroscopy system in ITER is quantitatively investigated using a ray tracing simulation. Simulation results show that the stray light is negligible at positions in the divertor where the plasma emission is strong. However, it is also shown that the stray light can be significantly greater than the real signal if the plasma intensity is low. Deuterium and beryllium emissions are used for the assessment; for beryllium cases in particular, since the emission profile may be non-uniform in the divertor region, the influence of stray light can be non-negligible at some positions, e.g., above the divertor dome

Liquid metal cooled divertor for ARIES

International Nuclear Information System (INIS)

Muraviev, E.

1995-01-01

A liquid metal, Ga-cooled divertor design was completed for the double null ARIES-II divertor design. The design analysis indicated a surface heat flux removal capability of up to 15 MW/m 2 , and its relative easy maintenance. Design issues of configuration, thermal hydraulics, thermal stresses, liquid metal loop and safety effects were evaluated. For coolant flow control, it was found that it is necessary to use some part of the blanket cooling ducts for the draining of liquid metal from the top divertor. In order to minimize the inventory of Ga, it was recommended that the liquid metal loop equipment should be located as close to the torus as possible. More detailed analysis of transient conditions especially under accident conditions was identified as an issue that will need to be addressed

Hydrogen isotopes retention in divertor tiles of DIII-D tokamak

International Nuclear Information System (INIS)

Skorodumov, B.G.; Buzhinskij, O.I.; West, W.P.; Ulanov, V.G.

1996-01-01

The absolute concentration of hydrogen isotopes in graphite divertor tiles coated with boron carbide after the exposure in DIII-D during 16 operational weeks of the 1993 campaign was obtained using the 14 MeV neutron-induced recoil detection (NERD) method. It is shown that the absolute concentration of H in tile's surface layers correlates with thickness of the deposited layers. The graphite tile without boron carbide coating had a H concentration similar to that of the tile with the thickest deposited layer. Deuterium and tritium were not detected in any of the investigated tiles. The proposed method can be used for the determination of the thickness of coatings without sample destruction. Thus, the thickness of boron carbide coatings on the tiles obtained with this method varied from 80 to 115 μm, which corresponded well to electron microscope data. (orig.)

A new fully automatic PIM tool to replicate two component tungsten DEMO divertor parts

International Nuclear Information System (INIS)

Antusch, Steffen; Commin, Lorelei; Heneka, Jochen; Piotter, Volker; Plewa, Klaus; Walter, Heinz

2013-01-01

Highlights: • Development of a fully automatic 2C-PIM tool. • Replicate fusion relevant components in one step without additional brazing. • No cracks or gaps in the seam of the joining zone visible. • For both material combinations a solid bond of the material interface was achieved. • PIM is a powerful process for mass production as well as for joining even complex shaped parts. -- Abstract: At Karlsruhe Institute of Technology (KIT), divertor design concepts for future nuclear fusion power plants beyond ITER are intensively investigated. One promising KIT divertor design concept for the future DEMO power reactor is based on modular He-cooled finger units. The manufacturing of such parts by mechanical machining such as milling and turning, however, is extremely cost and time intensive because tungsten is very hard and brittle. Powder Injection Molding (PIM) has been adapted to tungsten processing at KIT since a couple of years. This production method is deemed promising in view of large-scale production of tungsten parts with high near-net-shape precision, hence, offering an advantage of cost-saving process compared to conventional machining. The properties of the effectively and successfully manufactured divertor part tile consisting only of pure tungsten are a microstructure without cracks and a high density (>98% T.D.). Based on the achieved results a new fully automatic multicomponent PIM tool was developed and allows the replication and joining without brazing of fusion relevant components of different materials in one step and the creation of composite materials. This contribution describes the process route to design and engineer a new fully automatic 2C-PIM tool, including the filling simulation and the implementing of the tool. The complete technological fabrication process of tungsten 2C-PIM, including material and feedstock (powder and binder) development, injection molding, and heat-treatment of real DEMO divertor parts is outlined

Progress towards RF heated steady-state plasma operations on LHD by employing ICRF heating methods and improved divertor plates

International Nuclear Information System (INIS)

Kumazawa, R.; Mutoh, T.; Saito, K.

2008-10-01

A long pulse plasma discharge experiment was carried out using RF heating power in the Large Helical Device (LHD), a currentless magnetic confining system. Progress in long pulse operation is summarized since the 10th experimental campaign (2006). A scaling relation of the plasma duration time to the applied RF power has been derived from the experimental data so far collected. It indicates that there exists a critical divertor temperature and consequently a critical RF heating power P RFcrit =0.65 MW. The area on the graph of the duration time versus the RF heating power was extended over the scaling relation by replacing divertor plates with new ones with better heat conductivity. The cause of the plasma collapse at the end of the long pulse operation was found to be the penetration of metal impurities. Many thin flakes consisting of heavy metals and graphite in stratified layers were found on the divertor plates and it was thought that they were the cause of impurity metals penetrating into the plasma. In a simulation involving injecting a graphite-coated Fe pellet to the plasma it was found that 230 Eμm in the diameter of the Fe pellet sphere was the critical size which led the plasma to collapse. A mode-conversion heating method was examined in place of the minority ICRF heating which has been employed in almost all the long-pulse plasma discharges. It was found that this method was much better from the viewpoint of achieving uniformity of the plasma heat load to the divertors. It is expected that P RFcrit will be increased by using the mode-conversion heating method. (author)

Thermo-mechanical and damage analyses of EAST carbon divertor under type-I ELMy H-mode operation

Energy Technology Data Exchange (ETDEWEB)

Li, W.X. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Ye, M.Y. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Wu, S.T. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China); Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Qian, X.Y.; Zhu, C.C. [School of Nuclear Science and Technology, University of Science and Technology of China, Hefei 230026 (China)

2016-04-15

Highlights: • Type-I ELMy H-mode is one of the most severe operating environment in tokamak. • An actual time-history heat load has been used in thermo-mechanical analysis. • The analysis results are time-dependent during the whole discharge process. • The analysis could be very useful in evaluating the operational capability of the divertor. - Abstract: The lower carbon divertor has been used since 2008 in EAST, and many significant physical results, like the 410 s long pulse discharge and the 32 s H-mode operation, have been achieved. As the carbon divertor will still be used in the next few years while the injected auxiliary heating power would be increased gradually, it’s necessary to evaluate the operational capability of the carbon divertor under the heat loads during future operation. In this paper, an actual time-history heat load during type-I ELMy H-mode from EAST experiment, as one of the most severe operating environment in tokamak, has been used in the calculation and analysis. The finite element (FE) thermal and mechanical calculations have been carried out to analysis the stress and deformation of the carbon divertor during the heat loads. According to the results, the main impact on the overall temperature comes from the relative stable phase before and after the type-I ELMs and local peak load, and the transient thermal load such as type-I ELMy only has a significant effect on the surface temperature of the graphite tiles. The carbon divertor would work with high stress near the screw bolts in the current operational conditions, because of high preload and conservative frictional coefficient between the bolts and heatsink. For the future operation, new plasma facing materials (PFM) and divertor technology should be developed.

An X-point ergodic divertor

International Nuclear Information System (INIS)

Chu, M.S.; Jensen, T.H.; La Haye, R.J.; Taylor, T.S.; Evans, T.E.

1991-10-01

A new ergodic divertor is proposed. It utilizes a system of external (n = 3) coils arranged to generate overlapping magnetic islands in the edge region of a diverted tokamak and connect the randomized field lines to the external (cold) divertor plate. The novel feature in the configuration is the placement of the external coils close to the X-point. A realistic design of the external coil set is studied by using the field line tracing method for a low aspect ratio (A ≅ 3) tokamak. Two types of effects are observed. First, by placing the coils close to the X-point, where the poloidal magnetic field is weak and the rational surfaces are closely packed only a moderate amount of current in the external coils is needed to ergodize the edge region. This ergodized edge enhances the edge transport in the X-point region and leads to the potential of edge profile control and the avoidance of edge localized modes (ELMs). Furthermore, the trajectories of the field lines close to the X-point are modified by the external coil set, causing the hit points on the external divertor plates to be randomized and spread out in the major radius direction. A time-dependent modulation of the currents in the external (n = 3) coils can potentially spread the heat flux more uniformly on the divertor plate avoiding high concentration of the heat flux. 10 refs., 9 figs

Variation of Particle Control with Changes in Divertor Geometry

International Nuclear Information System (INIS)

Petrie, T W; Allen, S L; Brooks, N H; Fenstermacher, M E; Ferron, J R; Greenfield, C M; Groth, M; Hyatt, A W; Leonard, A W; Luce, T C; Mahdavi, M A; Murakami, M; Porter, G D; Rensink, M E; Schaffer, M J; Wade, M R; Watkins, J G; West, W P; Wolf, N S

2004-01-01

Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of Bx(divergent)B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the Bx(divergent)B ion particle drift direction. Our data suggests that the presence of Bx(divergent)B and ExB ion particle drifts in the scrapeoff layer (SOL) and divertors play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density n e,PED . In the lower range of densities considered in this study, i.e., n e,PED / n GREENWALD <0.4, particle exhaust rates are also found to be approximately proportional to the deuterium recycling intensity in front of the respective plenum entrance. Our results are shown to have implications for particle control in ITER and other future tokamaks

Variation of particle control with changes in divertor geometry

International Nuclear Information System (INIS)

Petrie, T.W.; Allen, S.L.; Brooks, N.H.; Fenstermacher, M.E.; Groth, M.; Porter, G.D.; Rensink, M.E.; Wolf, N.S.; Ferron, J.R.; Greenfield, C.M.; Hyatt, A.W.; Leonard, A.W.; Luce, T.C.; Mahdavi, M.A.; Schaffer, M.J.; West, W.P.; Murakami, M.; Wade, M.R.; Watkins, J.G.

2005-01-01

Recent experiments on DIII-D point to the importance of two factors in determining how effectively the deuterium particle inventory in a tokamak plasma can be controlled through pumping at the divertor target(s): (1) the divertor magnetic balance, i.e., the degree to which the divertor topology is single-null (SN) or double-null (DN), and (2) the direction of the of Bx∇B ion drift with respect to the X-point(s). Changes in divertor magnetic balance near the DN shape have a much stronger effect on the particle exhaust rate at the inner divertor target(s) than on the particle exhaust rate at the outer divertor target(s). The particle exhaust rate for the DN shape is strongest at the outer strike point opposite the Bx∇B ion particle drift direction. Our data suggests that the presence of Bx∇B and ExB ion particle drifts in the scrapeoff layer (SOL) and divertors play an important role in the particle exhaust rates of DN and near-DN plasmas. Particle exhaust rates are shown to depend strongly on the edge (pedestal) density n e,PED . In the lower range of densities considered in this study, i.e., n e,PED /n GREENWALD <0.4, particle exhaust rates are also found to be approximately proportional to the deuterium recycling intensity in front of the respective plenum entrance. Our results are shown to have implications for particle control in ITER and other future tokamaks. (author)

Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Krakowski, R.A.; Miller, R.L.

1985-01-01

Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line tracings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented

Design manufacturing and thermo-mechanical testing of a relevant size mono block divertor prototype

International Nuclear Information System (INIS)

Cardella, A.; Vieider, G.; Di Pietro, E.; Orsini, A.; Febvre, M.; Guerreschi, U.; Reheis, N.; Bruno, L.

1994-01-01

Following a technological development of joining techniques between carbon fibre composite tiles and metallic tubes, and the manufacturing and testing of small size actively cooled mock-ups, a relevant size divertor prototype has been designed, manufactured and tested. The prototype consisted of a series of metallic tubes surrounded by CFC tiles, cooling collectors and a supporting system representative of a divertor dump plate for high power reactors. The tubes have been preliminary tested at the CEA 200 kW electron beam facility with uniform fluxes up to 5 MW/m 2 to select the best five tubes, which together with a sixth non tested tube have been then assembled to form the prototype. This has been tested at the JET high power neutral beam injector test facility. After screening tests the prototype has been subjected to thermal cycling at more than 15 MW/m 2 . (author) 12 refs.; 4 figs

Divertor heat and particle control experiments on the DIII-D tokamak

International Nuclear Information System (INIS)

Mahdavi, M.A.; Baker, D.R.; Allen, S.L.

1994-05-01

In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D 2 gas puffing and impurity injection were studied separately and in, both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models

Characteristics of divertor plasma and scrape-off layer in JT-60U

International Nuclear Information System (INIS)

Itami, K.; Shimada, M.; Hosogane, N.

1992-01-01

Heat flux to the divertor is measured by thermography and the heat transport in the scrape-off layer is studied in beam heated discharges of JT-60U. The heat flux onto the divertor is ∝50% of total beam power at maximum. The in-out asymmetry of the heat flux P HEAT in /P HEAT out is as large as 20-40% when the ion grad-B drift is toward the divertor. Differences in P HEAT in /P HEA T out due to the direction of ion grad-B drift are as large as large as ∝40%. A scaling of the peaking factor Y of heat flux, defined by Y=2πRfq max /P HEAT , is obtained for beam heated discharges in JT-60U with a wide range of plasma parameters. The Y corresponds to the inverse of the thickness of the scrape-off layer. From a statistical analysis, it is found that the peaking factor Y of heat flux scales as P HEAT 0.49±0.18 anti n e -0.45±0.22 q eff -0.67±0.18 . (orig.)

Structural design of the DIII-D radiative divertor

International Nuclear Information System (INIS)

Reis, E.E.; Smith, J.P.; Baxi, C.B.; Bozek, A.S.; Chin, E.; Hollerbach, M.A.; Laughon, G.J.; Sevier, D.L.

1996-10-01

The divertor of the DIII-D tokamak is being modified to operate as a slot type, dissipative divertor. This modification, called the Radiative Divertor Program (RDP) is being carried out in two phases. The design and analysis is complete and hardware is being fabricated for the first phase. This first phase consists of an upper divertor baffle and cryopump to provide some density control for high triangularity, single or double null discharges. Installation of the first phase is scheduled to start in October, 1996. The second phase provides pumping at all four divertor strike points of double null high triangularity discharges and baffling of the neutral particles from transport back to the core plasma. Studies of the effects of varying the slot length and width of the divertor can be easily accomplished with the design of RDP hardware. Static and dynamic analyses of the baffle structures, new cryopumps, and feedlines were performed during the preliminary and final design phases. Disruption loads and differential thermal displacements must be accommodated in the design of these components. With the full RDP hardware installed, the plasma current in DIII-D will be a maximum of 3.0 MA. Plasma disruptions induce toroidal currents in the cryopump, producing complex dynamic loads. Simultaneously, the vacuum vessel vibrations impose a sinusoidal base excitation to the supports for the cryopump. Static and dynamic analyses of the cryopump demonstrate that the stresses due to disruption and thermal loadings satisfy the stress and deflection criteria

Snowflake divertor plasmas on TCV

International Nuclear Information System (INIS)

Piras, F; Coda, S; Furno, I; Moret, J-M; Sauter, O; Turri, G; Bencze, A; Duval, B P; Felici, F; Pochelon, A; Zucca, C; Pitts, R A; Tal, B

2009-01-01

Starting from a standard single null X-point configuration, a second order null divertor (snowflake (SF)) has been successfully created on the Tokamak a Configuration Variable (TCV) tokamak. The magnetic properties of this innovative configuration have been analysed and compared with a standard X-point configuration. For the SF divertor, the connection length and the flux expansion close to the separatrix exceed those of the standard X-point by more than a factor of 2. The magnetic shear in the plasma edge is also larger for the SF configuration.

Analysis of sweeping heat loads on divertor plate materials

International Nuclear Information System (INIS)

Hassanein, A.

1991-01-01

The heat flux on the divertor plate of a fusion reactor is probably one of the most limiting constraints on its lifetime. The current heat flux profile on the outer divertor plate of a device like ITER is highly peaked with narrow profile. The peak heat flux can be as high as 30--40 MW/m 2 with full width at half maximum (FWHM) is in the order of a few centimeters. Sweeping the separatrix along the divertor plate is one of the options proposed to reduce the thermomechanical effects of this highly peaked narrow profile distribution. The effectiveness of the sweeping process is investigated parametrically for various design values. The optimum sweeping parameters of a particular heat load will depend on the design of the divertor plate as well as on the profile of such a heat load. In general, moving a highly peaked heat load results in substantial reduction of the thermomechanical effects on the divertor plate. 3 refs., 8 figs

FLP: a field line plotting code for bundle divertor design

International Nuclear Information System (INIS)

Ruchti, C.

1981-01-01

A computer code was developed to aid in the design of bundle divertors. The code can handle discrete toroidal field coils and various divertor coil configurations. All coils must be composed of straight line segments. The code runs on the PDP-10 and displays plots of the configuration, field lines, and field ripple. It automatically chooses the coil currents to connect the separatrix produced by the divertor to the outer edge of the plasma and calculates the required coil cross sections. Several divertor designs are illustrated to show how the code works

Divertor pumping system with NBI cryopump for JT-60

International Nuclear Information System (INIS)

Akino, Noboru; Kuriyama, Masaaki; Ohga, Tokumichi; Seki, Hiroshi; Tanai, Yutaka

1998-08-01

The pumping system for JT-60 W-shape divertor with the NBI cryopump have been developed. The pumping speed achieved in the divertor region was 13-15 m 3 /s for deuterium gas with three units of the NBI cryopumps. In a simulation experiment of helium ash exhaust through the divertor, pumping of a mixed gas of helium and deuterium has been demonstrated using the NBI cryosorption pumps covered with an argon condensed layer. Control of neutral particle pressure in the divertor region became possible by having remodeled an aperture of the existing fast shutter, which is installed between the JT-60 vacuum vessel and NBI beam-line, to be regulated. (author)

Observation of an improved energy-confinement regime in neutral-beam--heated divertor discharges in the DIII-D tokamak

International Nuclear Information System (INIS)

Burrell, K.H.; Ejima, S.; Schissel, D.P.

1987-01-01

Tokamak discharges using the expanded boundary divertor in the DIII-D device exhibit H-mode confinement. With neutral-beam power up to 6 MW, energy confinement remains comparable to the Ohmic value at a plasma current of 1 MA. Confinement is also independent of plasma density and toroidal field. Confinement increases with plasma current, but the exact functional dependence is, as yet, uncertain. These results show that the H mode can be achieved in a reactor-compatible open divertor configuration

Control of thermal therapies with moving power deposition field

International Nuclear Information System (INIS)

Arora, Dhiraj; Minor, Mark A; Skliar, Mikhail; Roemer, Robert B

2006-01-01

A thermal therapy feedback control approach to control thermal dose using a moving power deposition field is developed and evaluated using simulations. A normal tissue safety objective is incorporated in the controller design by imposing constraints on temperature elevations at selected normal tissue locations. The proposed control technique consists of two stages. The first stage uses a model-based sliding mode controller that dynamically generates an 'ideal' power deposition profile which is generally unrealizable with available heating modalities. Subsequently, in order to approximately realize this spatially distributed idealized power deposition, a constrained quadratic optimizer is implemented to compute intensities and dwell times for a set of pre-selected power deposition fields created by a scanned focused transducer. The dwell times for various power deposition profiles are dynamically generated online as opposed to the commonly employed a priori-decided heating strategies. Dynamic intensity and trajectory generation safeguards the treatment outcome against modelling uncertainties and unknown disturbances. The controller is designed to enforce simultaneous activation of multiple normal tissue temperature constraints by rapidly switching between various power deposition profiles. The hypothesis behind the controller design is that the simultaneous activation of multiple constraints substantially reduces treatment time without compromising normal tissue safety. The controller performance and robustness with respect to parameter uncertainties is evaluated using simulations. The results demonstrate that the proposed controller can successfully deliver the desired thermal dose to the target while maintaining the temperatures at the user-specified normal tissue locations at or below the maximum allowable values. Although demonstrated for the case of a scanned focused ultrasound transducer, the developed approach can be extended to other heating modalities with

Modelling of profile control with LH wave injection in the HL-2A single-null divertor plasma

International Nuclear Information System (INIS)

Gao Qingdi; Yuan Baoshan; Li Fangzhu; Wang, Aike; Budny, R.V.

2005-01-01

In the HL-2A tokamak a single-null divertor configuration has been established. The separatrix of the single-null diverted plasma was identified with a filament model, and the determined striking area on the target plate is in agreement with the measurements of electric probe array. Higher power LH wave (1.5MW) is injected to the diverted plasma with a nearly symmetric spectrum. Dominant electron heating and current profile control are investigated with numerical simulation. Plasma heating by electron Landau interaction results in operation scenarios of preferentially dominant electron heating. Due to the off-axis driven current, an optimized q-profile is formed, and an enhanced confinement regime with steep electron temperature gradient is produced. The clear decrease of the electron thermal conductivity in the LH power deposition region shows that an electron-ITB is developed. When higher LH power injects into the target plasma that is heated by NBI (0.5MW), the ion temperature has a large increment in addition to the high increase of electron temperature. The temperature profiles indicate that an enhanced core confinement is established with both ion-ITB and electron-ITB developed. (author)

Turbulent Simulations of Divertor Detachment Based On BOUT + + Framework

Science.gov (United States)

Chen, Bin; Xu, Xueqiao; Xia, Tianyang; Ye, Minyou

2015-11-01

China Fusion Engineering Testing Reactor is under conceptual design, acting as a bridge between ITER and DEMO. The detached divertor operation offers great promise for a reduction of heat flux onto divertor target plates for acceptable erosion. Therefore, a density scan is performed via an increase of D2 gas puffing rates in the range of 0 . 0 ~ 5 . 0 ×1023s-1 by using the B2-Eirene/SOLPS 5.0 code package to study the heat flux control and impurity screening property. As the density increases, it shows a gradually change of the divertor operation status, from low-recycling regime to high-recycling regime and finally to detachment. Significant radiation loss inside the confined plasma in the divertor region during detachment leads to strong parallel density and temperature gradients. Based on the SOLPS simulations, BOUT + + simulations will be presented to investigate the stability and turbulent transport under divertor plasma detachment, particularly the strong parallel gradient driven instabilities and enhanced plasma turbulence to spread heat flux over larger surface areas. The correlation between outer mid-plane and divertor turbulence and the related transport will be analyzed. Prepared by LLNL under Contract DE-AC52-07NA27344. LLNL-ABS-675075.

CIT divertor conceptual design

International Nuclear Information System (INIS)

Wesley, J.C.; Sevier, D.L.

1988-06-01

A conceptual design of the divertor target assembly for the 1.75-m CIT baseline device has been developed. The divertor target assembly consists of four toroidal arrays of pyrolytic graphite plates that cover the inside surface of the ends of the vacuum vessel in the locations where the magnetic separatrices of the plasma intersect the vessel wall. During the course of the plasma discharge, the currents on the poloidal field coils that establish the plasma equilibrium are varied to sweep the separatrix strike locations across the divertor targets. This spreads the plasma heat loading over sufficient area to keep the peak target surface temperature within allowable limits. The required magnetic sweep (/+-/5 cm for the inside strike location and /+-/12 cm for the outside strike location) can be affected by programming either the external poloidal strike location) can be effected by programming either the external poloidal field (PF) coils or the internal PF control coils plus the external PF solenoid coils (PF1 and PF2). The ensuing variations in the elongation and triangularity of the plasma are modest, and fall within the ranges of plasma elongation and triangularity specified in the CIT General Requirements Document. 17 figs., 13 tabs

Plasma diagnostics for the DIII-D divertor upgrade (abstract)

International Nuclear Information System (INIS)

Hill, D.N.; Futch, A.; Buchenauer, D.; Doerner, R.; Lehmer, R.; Schmitz, L.; Klepper, C.C.; Menon, M.; Leikind, B.; Lippmann, S.; Mahdavi, M.A.; Schaffer, M.; Smith, J.; Salmonson, J.; Watkins, J.

1990-01-01

The DIII-D tokamak is being upgraded to allow for divertor biasing, baffling, and pumping experiments. This paper gives an overview of the new diagnostics added to DIII-D as part of this advanced divertor program. They include tile current monitors, fast reciprocating Langmuir probes, a fixed probe array in the divertor, fast neutral pressure gauges, and H α measurements with TV cameras and fiber optics coupled to a high-resolution spectrometer

Optimization of a bundle divertor for FED

International Nuclear Information System (INIS)

Hively, L.M.; Rothe, K.E.; Minkoff, M.

1982-01-01

Optimal double-T bundle divertor configurations have been obtained for the Fusion Engineering Device (FED). On-axis ripple is minimized, while satisfying a series of engineering constraints. The ensuing non-linear optimization problem is solved via a sequence of quadratic programming subproblems, using the VMCON algorithm. The resulting divertor designs are substantially improved over previous configurations

Engineering, installation, testing, and initial operation of the DIII-D Advanced Divertor

International Nuclear Information System (INIS)

Andersen, P.M.; Baxi, C.B.; Reis, E.E.; Schaffer, M.J.; Smith, J.P.

1990-09-01

The Advanced Divertor (AD) for General Atomics tokamak, DIII-D, was installed in the summer of 1990. The AD has enabled two classes of physics experiments to be run: divertor biasing and divertor baffling. Both are new experiments for DIII-D. The AD has two principal components: (1) a continuous ring electrode; and (2) a toroidally symmetric baffle. The tokamak can be run in bias baffle or standard DIII-D divertor modes by accurate positioning of the outer divertor strike point through the use of the DIII-D control system. The paper covers design, analysis, fabrication, installation, instrumentation, testing, initial operation, and future plans for the Advanced Divertor from an engineering viewpoint. 2 refs., 5 figs

Effects of divertor geometry and pumping on plasma performance on DIII-D

International Nuclear Information System (INIS)

Allen, S.L.; Hill, D.N.; Porter, G.D.

1997-06-01

This paper reports the status of an ongoing investigation to discern the influence of the divertor and plasma geometry on the confinement of both ELM-free and ELMing discharges in DIII-D. The ultimate goal is to achieve a high-performance core plasma which coexists with an advanced divertor plasma. The divertor plasma must reduce the heat flux to acceptable levels; the current technique disperses the heat flux over a wide area by radiation (a radiative divertor). To date, we have obtained our best performance in double-null (DN) high-triangularity (δ ∼ 0.8) ELM-free discharges. As discussed in detail elsewhere, there are several advantages for both the core and divertor plasma with highly-shaped DN operation. Previous radiative-divertor experiments with D 2 injection in DN high-δ ELMing H-mode have shown that this configuration is more sensitive to gas puffing (τ decreases). Moving the X-point away from the target plate (to ∼15 cm above the plate) decreases this sensitivity. Preliminary measurements also indicate that gas puffing reduces the divertor heat flux but does not reduce the plasma pressure along the field line. The up/down heat flux balance can be varied magnetically (by changing the distance between the separatrices), with a slight magnetic imbalance required to balance the heat flux. The overall mission of the Radiative Divertor Project (RDP) is to install a fully pumped and baffled high-δ DN divertor. To date, however, both the DIII-D divertor diagnostics and pump were optimized for lower single-null (LSN) low-δ (δ∼ 0.4) plasmas, so much of the divertor physics has been performed in LSN; these results are discussed in Section 2. As part of the first phase of the RDP, we have installed a new high-δ USN divertor baffle and pump; these results are discussed in Section 3. Both divertor and core parameters are discussed in each case

Limiter and divertor systems - conceptual and mechanical design for Aditya Tokamak upgrade

International Nuclear Information System (INIS)

Patel, Kaushal; Rathod, Kulav; Jadeja, Kumarpalsinh A.

2015-01-01

Existing Aditya tokamak with limiter configuration is being upgraded into a machine to have both the limiter and divertor configurations. Necessary modifications have been carried out to accommodate divertor coils by replacing the old vacuum vessel with a new circular section vacuum vessel. The upgraded Aditya tokamak will have different set of limiters and divertors, such as Safety limiter, Toroidal Inner limiter, outer limiter of smaller toroidal extent, Upper and lower divertor plates. The limiter and divertor locations inside the Aditya tokamak upgrade are decided based on the numerical simulation of the plasma equilibrium profiles. Initially graphite will be used as plasma facing material (PFM) in all the limiter and divertor plates. The dimensions of the limiter and divertor tiles are decided based on their installation inside the vacuum vessel as well as on the total plasma heat loads (∼ 1 MW) falling on them. Depending upon the heat loads; the thickness of graphite tiles for limiter and divertor plates is estimated. Shaped graphite tiles will be fixed on specially designed support structures made out of SS-304L inside the torus shaped vacuum vessel. In this paper mechanical structural design of limiter and divertor of Aditya Upgrade Tokamak is presented. (author)

Divertor target profiles and recycling studies in TCV single null lower standard discharges

International Nuclear Information System (INIS)

Pitts, R.A.; Nieswand, C.; Weisen, H.

1996-05-01

A 'standard', single null lower diverted discharge has been developed to enable continuous monitoring of the first wall conditions and to characterise the effectiveness and influence of wall conditioning in the TCV tokamak. Measurements over a period encompassing nearly 2000 ohmic discharges of varying configuration and input power show the global confinement time and main plasma impurity concentrations to be good general indicators of the first wall condition, whilst divertor target profiles demonstrate strikingly the short term beneficial effects of He glow. Good agreement, consistent with a reduction in recycling at the plates is found between the predictions of the fluid code UEDGE and the observed outer divertor profiles of T e and n e before and after He glow. (author) 5 figs., 7 refs

Controlling marginally detached divertor plasmas

Science.gov (United States)

Eldon, D.; Kolemen, E.; Barton, J. L.; Briesemeister, A. R.; Humphreys, D. A.; Leonard, A. W.; Maingi, R.; Makowski, M. A.; McLean, A. G.; Moser, A. L.; Stangeby, P. C.

2017-06-01

A new control system at DIII-D has stabilized the inter-ELM detached divertor plasma state for H-mode in close proximity to the threshold for reattachment, thus demonstrating the ability to maintain detachment with minimal gas puffing. When the same control system was instead ordered to hold the plasma at the threshold (here defined as T e  =  5 eV near the divertor target plate), the resulting T e profiles separated into two groups with one group consistent with marginal detachment, and the other with marginal attachment. The plasma dithers between the attached and detached states when the control system attempts to hold at the threshold. The control system is upgraded from the one described in Kolemen et al (2015 J. Nucl. Mater. 463 1186) and it handles ELMing plasmas by using real time D α measurements to remove during-ELM slices from real time T e measurements derived from divertor Thomson scattering. The difference between measured and requested inter-ELM T e is passed to a PID (proportional-integral-derivative) controller to determine gas puff commands. While some degree of detachment is essential for the health of ITER’s divertor, more deeply detached plasmas have greater radiative losses and, at the extreme, confinement degradation, making it desirable to limit detachment to the minimum level needed to protect the target plate (Kolemen et al 2015 J. Nucl. Mater. 463 1186). However, the observed bifurcation in plasma conditions at the outer strike point with the ion B   ×  \

Radiative and SOL experiments in open and baffled divertors on DIII-D

International Nuclear Information System (INIS)

Allen, S.L.; Brooks, N.H.; Bastasz, R.

1998-11-01

The authors present recent progress towards an understanding of the physical processes in the divertor and scrape-off-layer (SOL) plasmas in DIII-D. This has been made possible by a combination of new diagnostics, improved computational models, and changes in divertor geometry. They have focused primarily on ELMing H-mode discharges. The physics of Partially Detached Divertor (PDD) plasmas, with divertor heat flux reduction by divertor radiation enhancement using D 2 puffing, has been studied in 2-D, and a model of the heat and particle transport has been developed that includes conduction, convection, ionization, recombination, and flows. Plasma and impurity particle flows have been measured with Mach probes and spectroscopy and these flows have been compared with the UEDGE model. The model now includes self-consistent calculations of carbon impurities. Impurity radiation has been increased in the divertor and SOL with puff and pump techniques using SOL D 2 puffing, divertor cryopumping, and argon puffing. The important physical processes in plasma-wall interactions have been examined with a DiMES probe, plasma characterization near the divertor plate, and the REDEP code. Experiments comparing single-null (SN) plasma operation in baffled and open divertors have demonstrated a change in the edge plasma profiles. These results are consistent with a reduction in the core ionization source calculated with UEDGE. Divertor particle control in ELMing H-mode with pumping and baffling has resulted in reduction in H-mode core densities to n e /n gw ∼ 0.25. Divertor particle exhaust and heat flux has been studied as the plasma shape was varied from a lower SN, to a balanced double null (DN), and finally to an upper SN

Manufacturing W fibre-reinforced Cu composite pipes for application as heat sink in divertor targets of future nuclear fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Mueller, Alexander v.; You, Jeong-Ha [Max-Planck-Institut fuer Plasmaphysik, 85748 Garching (Germany); Ewert, Dagmar [Institut fuer Textil- und Verfahrenstechnik Denkendorf, 73770 Denkendorf (Germany); Siefken, Udo [Louis Renner GmbH, 85221 Dachau (Germany)

2016-07-01

An important plasma-facing component (PFC) in future nuclear fusion reactors is the so-called divertor which allows power exhaust and removal of impurities from the main plasma. The most highly loaded parts of a divertor are the target plates which have to withstand intense particle bombardment. This intense particle bombardment leads to high heat fluxes onto the target plates which in turn lead to severe thermomechanical loads. With regard to future nuclear fusion reactors, an improvement of the performance of divertor targets is desirable in order to ensure reliable long term operation of such PFCs. The performance of a divertor target is most closely linked to the properties of the materials that are used for its design. W fibre-reinforced Cu (Wf/Cu) composites are regarded as promising heat sink materials in this respect. These materials do not only feature adequate thermophysical and mechanical properties, they do also offer metallurgical flexibility as their microstructure and hence their macroscopic properties can be tailored. The contribution will point out how Wf/Cu composites can be used to realise an advanced design of a divertor target and how these materials can be fabricated by means of liquid Cu infiltration.

High power RF window deposition apparatus, method, and device

Science.gov (United States)

Ives, Lawrence R.; Lucovsky, Gerald; Zeller, Daniel

2017-07-04

A process for forming a coating for an RF window which has improved secondary electron emission and reduced multipactor for high power RF waveguides is formed from a substrate with low loss tangent and desirable mechanical characteristics. The substrate has an RPAO deposition layer applied which oxygenates the surface of the substrate to remove carbon impurities, thereafter has an RPAN deposition layer applied to nitrogen activate the surface of the substrate, after which a TiN deposition layer is applied using Titanium tert-butoxide. The TiN deposition layer is capped with a final RPAN deposition layer of nitridation to reduce the bound oxygen in the TiN deposition layer. The resulting RF window has greatly improved titanium layer adhesion, reduced multipactor, and is able to withstand greater RF power levels than provided by the prior art.

Features and Initial Results of the DIII-D Advanced Tokamak Radiative Divertor

International Nuclear Information System (INIS)

R.C. O'Neill; A.S. Bozek; M.E. Friend; C.B. Baxi; E.E. Reis; M.A. Mahdavi; D.G. Nilson; S.L. Allen; W.P. West

1999-01-01

The Radiative Divertor Program of DIII-D is in its final phase with the installation of the cryopump and baffle structure (Phase 1B Divertor) in the upper inner radius of the DIII-D vacuum vessel at the end of this calendar year. This divertor, in conjunction with the Advanced Divertor and the Phase 1A Divertor, located in the lower and upper outer radius of the DIII-D vacuum vessel respectively, provides pumping for density control of the plasma while minimizing the effects on the core confinement. Each divertor consists of a cryobelium cooling ring and a shielded protective structure. The cryo/helium-cooled pumps of all three diverters exhaust helium from the plasma. The protective shielded structure or baffle structure, in the case of the diverters located at the top of the vacuum vessel, provides baffling of neutral charged particles and minimize the flow of impurities back into the core of the plasma. The baffles, which consist of water-cooled panels that allow for the attachment of tiles of various sizes and shapes, house gas puff systems. The intent of the puffing systems is to inject gas in and around the divertor to minimize the heat flux on specific areas on the divertor and its components. The reduction of the heat flux on the divertor minimizes the impurities that are generated from excess heat on divertor components, specifically tiles. Experiments involving the gas puff systems and the divertor structures have shown the heat flux can be spread over a large area of the divertor, reducing the peak heat flux in specific areas. The three diverters also incorporate a variety of diagnostic tools such as halo current monitors, magnetic probes and thermocouples to monitor certain plasma characteristics as well as determine the effectiveness of the cryopumps and baffle configurations. The diverters were designed to optimize pumping performance and to withstand the electromagnetic loads from both halo currents and toroidal induced currents. Incorporated also

First results from the dynamic ergodic divertor at TEXTOR

International Nuclear Information System (INIS)

Lehnen, M.; Abdullaev, S.S.; Biel, W.; Brezinsek, S.; Finken, K.H.; Harting, D.; Hellermann, M. von; Jakubowski, M.; Jaspers, R.; Kobayashi, M.; Koslowski, H.R.; Kraemer-Flecken, A.; Matsunaga, G.; Pospieszczyk, A.; Reiter, D.; Van Rompuy, T.; Samm, U.; Schmitz, O.; Sergienko, G.; Unterberg, B.; Wolf, R.; Zimmermann, O.

2005-01-01

Experimental results from the dynamic ergodic divertor (DED) at TEXTOR are given, describing the complex structure of the edge plasma and the properties of the divertor as well as its influence on the plasma rotation

New achievements of the Divertor Test Platform programme for the ITER divertor remote maintenance R and D

International Nuclear Information System (INIS)

Damiani, C.; Baldi, L.; Galbiati, L.; Irving, M.; Lorenzelli, L.; Micciche, G.; Muro, L.; Nucci, S.; Varocchi, G.; Poggianti, A.; Fermani, G.; Maisonnier, D.; Palmer, J.; Martin, E.; Friconneau, J.P.; Gravez, P.; Takeda, N.

2001-01-01

The divertor assembly for the ITER fusion reactor consists of a number of rail mounted cassettes (54 now in ITER FEAT) located in the bottom region of the vacuum vessel. These cassettes shall be removed/installed remotely during the life of the reactor by means of specific devices. To demonstrate and optimise the feasibility of the in-vessel maintenance process the Divertor Test Platform (DTP) has been established at the ENEA Research Centre in Brasimone, Italy, as a major part of the large ITER R and D project L7. A first set of tests has been already carried out and reported during 1998, when the basic feasibility of the divertor replacement was demonstrated. In the present period (January 1999-July 2000), new activities, including both site tests and other 'external' R and D works, have been carried out in order to refine and improve the ITER divertor maintenance scenario. These include the study of abnormal maintenance operations and of possible handling equipment failure and its consequences; the procurement and testing of new sub-systems (e.g. a force reflection manipulator arm), and the development of remote handling techniques including a virtual reality system. Following a short description of the DTP, this paper reports on the new results and achievements, draws the relevant conclusions, and finally discusses future activities

Modeling of impurity spectroscopy in the divertor and SOL of DIII-D using the 1D multifluid model NEWT1D

International Nuclear Information System (INIS)

West, W.P.; Evans, T.E.; Brooks, N.H.

1996-10-01

NEWT1D, a one dimensional multifluid model of the scrape-off layer and divertor plasma, has been used to model the plasma including the distribution of carbon ionization states in the SOL and divertor of ELMing H-mode at two injected power levels in DIII-D. Comparison of the code predictions to the measured divertor and scrape-off layer (SOL) plasma density and temperature shows good agreement. Comparison of the predicted line emissions to the spectroscopic data suggests that physically sputtered carbon from the strike point is not transported up the flux tube; a distributed source of carbon a few centimeters up the flux tube is required to achieve reasonable agreement

Divertor plate for thermonuclear reactor

International Nuclear Information System (INIS)

Yamazaki, Seiichiro; Sato, Keisuke; Nishio, Satoshi.

1993-01-01

In a divertor plate for a thermonuclear reactor, adjacent cooling pipes are electrically insulated from each other and pipes made of a gradient functional material prepared by compositing ceramics having an insulation property and metals are metallurgically joined to at least one portion of each of the cooling pipes. Electric current caused upon occurrence of plasma disruption is interrupted by the insulation portion, so that a large circuit is not formed and electromagnetic force is decreased to such a extent that the divertor plate is not ruptured. Since a header of the cooling pipes can be installed at any optional position, the installation space can be reduced. Further, since inlet and exit collection headers can be disposed on both ends of the cooling pipes, it is possible to shorten the length of the cooling pipe of the divertor plate corresponded to high heat fluxes and reduce the pressure loss on the side of coolants to about 1/2. Further, turn back portions of small radius of curvature of the cooling pipes are eliminated to reduce the cost and extend the lifetime and, in addition, protection tiles can be attached easily. (N.H.)

Constrained ripple optimization of Tokamak bundle divertors

International Nuclear Information System (INIS)

Hively, L.M.; Rome, J.A.; Lynch, V.E.; Lyon, J.F.; Fowler, R.H.; Peng, Y-K.M.; Dory, R.A.

1983-02-01

Magnetic field ripple from a tokamak bundle divertor is localized to a small toroidal sector and must be treated differently from the usual (distributed) toroidal field (TF) coil ripple. Generally, in a tokamak with an unoptimized divertor design, all of the banana-trapped fast ions are quickly lost due to banana drift diffusion or to trapping between the 1/R variation in absolute value vector B ω B and local field maxima due to the divertor. A computer code has been written to optimize automatically on-axis ripple subject to these constraints, while varying up to nine design parameters. Optimum configurations have low on-axis ripple ( 0 ) are lost. However, because finite-sized TF coils have not been used in this study, the flux bundle is not expanded

The WEST project: Current status of the ITER-like tungsten divertor

International Nuclear Information System (INIS)

Missirlian, M.; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

2014-01-01

Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues

The WEST project: Current status of the ITER-like tungsten divertor

Energy Technology Data Exchange (ETDEWEB)

Missirlian, M., E-mail: marc.missirlian@cea.fr; Bucalossi, J.; Corre, Y.; Ferlay, F.; Firdaouss, M.; Garin, P.; Grosman, A.; Guilhem, D.; Gunn, J.; Languille, P.; Lipa, M.; Richou, M.; Tsitrone, E.

2014-10-15

Highlights: • We presented the ITER-like W components occurred for the WEST divertor. • The main features including key elements of the design were detailed. • The main results of studies investigating the integration constraints or issues were reported. • The WEST ITER-like divertor design reached a mature stage to enable the launching of the procurement phase. - Abstract: The WEST (W – for tungsten – Environment in Steady-state Tokamak) project is an upgrade of Tore Supra from a limiter based tokamak with carbon PFCs into an X-point divertor tokamak with full-tungsten armour while keeping its long discharge capability. The WEST project will primarily offer the key capability of testing for the first time the ITER technology in real plasma environment. In particular, the main divertor (i.e. the lower divertor) of the WEST project will be based on actively cooled tungsten monoblock components and will follow as closely as possible the design and the assembling technology, foreseen for the ITER divertor units. The current design of WEST ITER-like tungsten divertor has now reached a mature stage following the 2013 WEST Final Design Review. This paper presents the key elements of the design, reports the technological requirements and reviews the main design and integration issues.

Upgraded divertor Thomson scattering system on DIII-D

Energy Technology Data Exchange (ETDEWEB)

Glass, F., E-mail: glassf@fusion.gat.com; Carlstrom, T. N.; Du, D.; Taussig, D. A.; Boivin, R. L. [General Atomics, P.O. Box 85608, San Diego, California 92186-5608 (United States); McLean, A. G. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94550 (United States)

2016-11-15

A design to extend the unique divertor Thomson scattering system on DIII-D to allow measurements of electron temperature and density in high triangularity plasmas is presented. Access to this region is selectable on a shot-by-shot basis by redirecting the laser beam of the existing divertor Thomson system inboard — beneath the lower floor using a moveable, high-damage threshold, in-vacuum mirror — and then redirecting again vertically. The currently measured divertor region remains available with this mirror retracted. Scattered light is collected from viewchords near the divertor floor using in-vacuum, high temperature optical elements and relayed through the port window, before being coupled into optical fiber bundles. At higher elevations from the floor, measurements are made by dynamically re-focusing the existing divertor system collection optics. Nd:YAG laser timing, analysis of the scattered light spectrum via polychromators, data acquisition, and calibration are all handled by existing systems or methods of the current multi-pulse Thomson scattering system. Existing filtered polychromators with 7 spectral channels are employed to provide maximum measurement breadth (T{sub e} in the range of 0.5 eV–2 keV, n{sub e} in the range of 5 × 10{sup 18}–1 × 10{sup 21} m{sup 3}) for both low T{sub e} in detachment and high T{sub e} measurement up beyond the separatrix.

Physics conclusions in support of ITER W divertor monoblock shaping

Directory of Open Access Journals (Sweden)

R.A. Pitts

2017-08-01

Full Text Available The key remaining physics design issue for the ITER tungsten (W divertor is the question of monoblock (MB front surface shaping in the high heat flux target areas of the actively cooled targets. Engineering tolerance specifications impose a challenging maximum radial step between toroidally adjacent MBs of 0.3mm. Assuming optical projection of the parallel heat loads, magnetic shadowing of these edges is required if quasi-steady state melting is to be avoided under certain conditions during burning plasma operation and transiently during edge localized mode (ELM or disruption induced power loading. An experiment on JET in 2013 designed to investigate the consequences of transient W edge melting on ITER, found significant deficits in the edge power loads expected on the basis of simple geometric arguments, throwing doubt on the understanding of edge loading at glancing field line angles. As a result, a coordinated multi-experiment and simulation effort was initiated via the International Tokamak Physics Activity (ITPA and through ITER contracts, aimed at improving the physics basis supporting a MB shaping decision from the point of view both of edge power loading and melt dynamics. This paper reports on the outcome of this activity, concluding first that the geometrical approximation for leading edge power loading on radially misaligned poloidal leading edges is indeed valid. On this basis, the behaviour of shaped and unshaped monoblock surfaces under stationary and transient loads, with and without melting, is compared in order to examine the consequences of melting, or power overload in context of the benefit, or not, of shaping. The paper concludes that MB top surface shaping is recommended to shadow poloidal gap edges in the high heat flux areas of the ITER divertor targets.

A computational study of operating regimes for poloidal divertors

International Nuclear Information System (INIS)

Petravic, M.; Heifetz, D.; Post, D.

1982-01-01

We have identified three theoretical operating regimes for poloidal divertors. These regimes are determined by the geometry of the divertor and the input energy and particle fluxes, and are characterized by the divertor plasma density and temperature. A fully self-consistent two-dimensional model for the plasma and neutral atom and molecule transport was used to study poloidal divertor operation. Extensions of our previous calculations important to this study were the inclusion of parallel electron and ion thermal conduction. We find that the key physics in divertor operation is the neutral recycling near the neutralizer plate. This can be parametrized by R = GAMMAsub(P)/GAMMAsub(O), the ratio of particle flux striking the neutralizer plate to the particle flux entering the divertor. Values of R approx. equal to 1 can be produced by large pumping rates near the neutralizer plates resulting in low neutral recycling and a high temperature, low density divertor plasma. By decreasing the pumping near the neutralizer plate, R can be raised to an intermediate value of 5-10, the plasma temperature lowered by the same factor, and the density raised by a factor of 10-30. In this regime, escape of the neutrals back to the main plasma is virtually blocked. By further restricting the pumping, R can be raised to twenty or more, thereby lowering the temperature by a factor of twenty or more and raising the density by a factor of ninety or more. Such high density regimes have been observed on D-III and appear to offer the most promise for impurity control and particle control on large reactor experiments such as INTOR or FED. In this paper, we explore the range 3 < R < 16. (orig.)

Development of impurity seeding and radiation enhancement in the helical divertor of LHD

International Nuclear Information System (INIS)

Mukai, K.; Masuzaki, S.; Peterson, B.J.

2014-10-01

Impurity seeding for the reduction of the divertor heat load was conducted in the Large Helical Device (LHD) using neon (Ne) and krypton (Kr) puffing. Enhanced radiation loss and reduction of the divertor heat load were observed without significant changes in stored energy and line averaged density. In the LHD, the radiated power fraction of the heating power, P rad /P heating , where P rad and P heating are the total radiation power and the heating power, respectively, is limited up to around 30% in hydrogen plasmas even for high density plasma just below the radiative collapse (n e,bar > 1 × 10 20 m -3 ), where n e,bar is the line averaged density. With Ne seeding, the ratio could be raised to 52% in spite of much lower density (n e,bar ∼ 1.3 × 10 19 m -3 ), albeit with a slight reduction in confinement. P rad /P heating ∼ 30% could be sustained for 3.4 s using multi-pulse Ne seeding in relatively high density plasmas in this study (n e,bar ∼ 4 × 10 19 m -3 ). Radiation profile measurement with an InfraRed Imaging Video Bolometer (IRVB) was conducted during Ne seeding. The localized supplemental radiation was observed along the helical divertor X-points (HDXs). The enhanced radiation structure was similar to the Ne radiation profile estimated by EMC3-EIRENE code. Kr seeding was also conducted to the plasma of n e,bar ∼ 3.1 × 10 19 m -3 . P rad /P heating ∼ 25% was obtained without significant change in stored energy. The enhancement of the plasma radiation has a slower time constant compared with the Ne seeding. The supplemental radiation area of the Kr seeded plasma moved from the HDXs to the core plasma. High-valent Kr ions are considered to be the dominant radiators from the plasma core region. (author)

Investigation of electron parallel pressure balance in the scrape-off layer of deuterium-based radiative divertor discharges IN DIII-D

International Nuclear Information System (INIS)

Petrie, T.W.; Carlstrom, T.N.; Allen, S.L.

1996-10-01

Electron density, temperature, and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: I p = 1.4 MA, q 95 = 4.2, and P input ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of ∼ 2 between the midplane separatrix and the X-point, and then an additional ∼3--5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power maintains good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate, and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7--30 eV

Improved performance of the W7-AS stellarator with the new island divertor

International Nuclear Information System (INIS)

Brakel, R.; Grigull, P.; McCormick, K.

2003-01-01

The island divertor concept has successfully been realized at W7-AS. The divertor gives access to a new NBI-heated high density regime with densities up to 4x1020 m-3 and energy confinement well above customary scalings (high density H-mode, HDH). In parallel, impurity confinement dramatically decreases. This prevents impurity accumulation, and allows for quasi-stationary high density discharges with excellent confinement and with radiation profiles peaked at the edge. At the highest densities partial detachment occurs with a radiated power fraction up to 90% at a tolerable mitigation of plasma energy. The HDH-mode appears promising with respect to the requirements of both confinement and exhaust. Major experimental results, such as the lack of a high recycling phase preceding detachment, are also predicted by the EMC3/EIRENE code. (author)

A tangentially viewing VUV TV system for the DIII-D divertor

International Nuclear Information System (INIS)

Nilson, D.G.; Ellis, R.; Fenstermacher, M.E.; Brewis, G.; Jalufka, N.